MCFR_UCLA_Ottewitte_thesis.txt

UNIVERSITY OF CALIFORNIA

Los Angeles

Configuration of a Molten Chloride Fast Reactor
on a Thorium Fuel Cycle

to Current Nuclear Fuel Cycle Concerns

A digsertation submitted in partial satisfaction of
the requirements for the degree

Doctor of Philosophy in Engineering

by

Eric Heinz Ottewitte

1982
 

The dissertation of Eric Heinz Ottewitte is approved,

 

 

Moses Greenfield

ClE e AHO&;M

Steven A, Moszkowski

 

e

Gef/id Ce Pomraning
2 z{flfi%

William Kastenberg, Commiitee Chair

University of Califormia, Los Angeles

ii
 

Table of Contents
1.0 Introduction
1ol Overview of Current Nuclear Politics and Technology
1e1e1 Nuclear Fuel Cycle Problems and Their Politics
1a1e2 Technical Solution: Thorium Fuel Cycle
1e¢143 Technical Solution: Fast Halide Reactor

1e1e3e1 Advantages of fluid~fuel reactors
1elelde2 Introduction to the MSFR

1,2 Historical Review of lMolten Salt Activities
Te2e1 American Activity

1e1 Development of the USBR
1.2 Development of the MCFR

o2
o2
1e242 European Aciivity

Poland/Switzerland, led by Miecyzslaw Taube
France

261
242
0243 Engla.nd
2¢4 Soviet Union
1.2.3 Summary MSR State-of-the-Art
2.0 Contemporary Conceins Which Affect Choice of an Advanced Concept

21 Nuclear Fuel Cycle Waste Management

2e.1+1 Characterization of Spent Fuel Wastes

2.1.1.1 Gases

Zeiele2 Solid fission products

2elelel Actinides

241+144 Unique biological hazard of plutonium

2e1+2 Management of Spent Fuel Wastes

2.2 Non-Proliferation
22«1 Development of US Policy

2e2¢2 Relationship Between Nuclear Electric Fower and
Nuclear Weapons Deveiopment

iii
 

2e3

2,4

2e2e3

2.2.4

2e2e5H

2e2e241
2ele242
2e2e243

Extent of the spent fuel problem
Question of reprocessing
Attitudes on return of Pu

Which Reactor Fuel Contains the Least Proliferation

Danger?
202.3.1
2424342

Weapons-—-grade material

Unique 233, daughter radiation
Preignition '
238 . . .

Pu high heat generation
Denaturing and other technical fixes
Summary

The Significance of Stockpiles

2elelal
2e2e4e2

Peacefnl vs, military explosives
What a2bout unsafeguarded production reactors ?

Fuel Cycle Vulnerability to Diversion

2424541
2e2eDec
2e2e503
2e2e5e4
2e2e5eD
2424546

Once-~through fuel cycle
Solid-fuel reprocessing cycle
Molten~salt fuel cycle

To reprocess or not to reprocess
To breed or not to breed

Fusion reactor vulnerability

Fuel Utilization

2e301
20342

20343

24344

Earth's Resources

Use of Taorium

Intercomparison of Reactor Concepts with Regard to Fuel
Utilization

Use of Both Th and U Reserves

Strategic Security-

2edel

Present Susceptibility

2adelal
20441.2
2edalel
2a4e1e4

Subnational blackmail
Sabotage

War

Summary

iv
2ede2 Potential Remedy with an-MSR
2.5 Summary Design Principals for an Advanced Reactor System
2.5«1 HWaste Management Restraints
2e5¢2 Non-Proliferation Restraints
2659¢2 Fuel Utilization Restraints

2¢5«4 Strategic Securiiy Restraints

2.6 Promise and Unigueness of ithe Molten Chloride Fast Reactor
on a Thorium Fuel Cycle

24641 Advantages of the llolten State

24642 Advantazes Stemming from the Exclusive Use of Therium
24643 Advantages of Continuous Reprocessirg

2ebed Advantages of a Very Fast Nevtron épectrum

2.6.5 Spinoff Ability to Digest Existing Spent Fuels

24646 Inherent Disadvantages and Limitations ‘o MCFR

30 Concept Design
3.1 Choosing the Reactor Configuration

*3.1.1 Maximizing Breeding by Minimizing Neutron Moderation
in the Core

3e1elel BG potential
3e1e1,2 Minimizing neutron moderation

3e1e2 Maxrimizing Breeding by Minimizing Leakage

31421 Effect of an inner blanket

3ela3e2 Size of the outer blanket

3a1e2e3 Hoderating the blanket neutron spectrum
{0 enhance capture

Je1e264 Neutron reflector and damage-shield
 

Je2

3.1.3 COI‘e a.nd

le1ed

Blanket Design

Choice of geometry

Choice of number, size,and spacing of iubes
Location of tubes

Axial blanket and neutron leakage

Method of Reactivity Shimming

3e1e5 Summary Guidelines on Reactor Configuration

Reactor Thermohydraulics

3e2al

3.2.2

3e243

3,244

Means of Cooling the lloliten Szlt Fuel

Out-of-core heat exchanger

In-core heat exchanger across tubes
In-core direct contact heat exchange
Blanket cooling

Power Density in an ISFR

- 3424244
13424245

3424246
3e2e247
3424248
Primary

3624341

3024302
3624343

Inherent high power density in FKSFRs
Realistic range of power densities

High neuiron flux levels and radiation damage
in fast reactors

Impetus for high temperature operatiou
Adjusting power density through criticel mass
Reactor design power

Reactor power iistribution

Summary

Coolant Velocity

Maximizing velocity apd minimizing pumping
power

Corrosion dependence on velocity
Velocities in similar systems

Operating Temperatures

3e2e441
3e2e442
3e2e443
3e2eded
3e2e4ded

3424446
3.2.441

Upper limits on fuel salt temperature

Hinimum fuel salt temperature

Minimum secondary coolant temperature
Resirictions on AT across heat exchanger walls
Restrictions on temrzrature differences wiinin
the primary circuit

{4t,t} in other HCFRs

Sumnary

vi
 

3.3

Je245 Design of the Primary Circuit Heat Exchanger

1,2,6

e /¢

3e2e561
3024542
3e24543

Leat transfer coefficient

Minimizing fuel inventory in the heat exchanger
Frimary fluid volume in heat exchanger and
associated plena and piping

Primary Circuit Arrangement

General guidelines

. Location of pumps and heat exchanger

Parallel =mbck-o-nels
Tube channels in series .
Summary

3.2.7 Reactor Thermohydraulics Summary

Choosing the Salt Composition

3e3.1

3e342

3343

Introduction

3.341«1 Neutron spectrum
3.341.2 Chemistry

3e3e1e3 Eutectic melting point
3e3e1e4 Heat transfer parameters
3.341e5 Densities

3e3e1e6 Viscosities

3e341e7 Specific heats

3.3.1.8 Thermal conductivities
3e3e149 Summary

Choice of Halogen

U W Lo W W
wwwwf»wwww
NNNN:ONI\)NN
\DOD—-JO\:H-PUJN"‘

Neutron moderating effect
Neutron absorption

Transmutation products

Melting point

Beiling point and vapor pressure
Chemical behavior

Cost and availability

Density

Summary

Choice of Actinide Chemical States (Stoichiometry)

Actinides inherent to an MCFR(Th)
Melting and boiling points
Chemical behavior

Choice between UCLl, and UCl

3 in the core
Chemical states in

the blanfet

vii
 

3.3¢4 Choice of Carrier Salt Cation

3e3eD

3.4 Structural

3ede

3ede?2

3ede3

3e3ede1 Neutron moderating effect
3e3ede2 Neutron absorption
2.3e4¢e3 Transmutation products
3e3eded Melting point
Je3ed4eH Boiling point and vapor pressure
3e3e446 Chemical behavior
3e3edeT Cost and availability
3e3e448 Density
3¢3s449 Heat transfer coefficient
3e3e441C Summary
Choice of Relative Proportions of Actinides and Carrier
Salt
343541 Neutron physics of the ThCl 4/11013 ratio in the
core mixture
Je3e 50 2 Density
Je3e5e3 Viscosity
3e3¢5¢4 Specific heat
3e3e95¢5 Thermal conductivity
3e3e5.6 Heat transfer coefficient
3¢3¢5«7 Choice of core mix
3¢3e5.8 Choice of blanket mix
3¢3e5.9 Butectic mix for flush salt
Materials
General Consideration and Criteria
3.4¢1e1 Plant-life econoxzics
304e142 Corrosion, general
3e¢4e1e3 Corrosion by electrochemical attack
Jedele4 Corrosion protection
3edeled Thermal and radiation-induced expansion

Chemical Reactions in an MCFR

3040 201
3eda2.2

Reactions with fission~produced mutants

The effect of UC14 presence in the core salt

Candidate Materizls for an MCFR

Overview on metals

do alloys

Graphite

Heat Exchanger considerations

viii
 

3.5

304.4
3eded
30406

3edeT

Materials for Core/Blanket Interface (Tubes)
iaterials for Reactor Vessel
Maierial for a Lead Secondary Circuit

3.44641 Steam generator tubes
3e4e642 Remainder of the circuit

Material for a Helium Secondary Circuit

Physics of MCFR(Th) and Its Fuel Cycle

3501

3a5e2

3e5e3

3454

3e¢5¢D

3e546

Nuclear Models of an MCFR

3454141 Definition of the reactor geometry
3.54142 Use of spherical geometry for neutronics
calculations

Neutronic Calculational Methods

3.5.241 Spatial mesh
356242 Angular quadrature
349¢2e3 Convergence criteria

Neutron Cross Sections

345+341 Bondarenko 26 group Set

3e5¢342 Twenty-six groups, Pl set from ENDF/B-IV
3¢De3el} Effecis of resonarce self-shielding

Transmutation Chains and Fquations

3e¢5¢441 Chains for actinide tragfimutations
345¢4e2 Chain for build-up of U-parented radio-
activity

Principal Physics Metrics

3e5¢5¢1 BG: actual breeding gain

3e5¢5¢2 BGCX: BG extended to zero neutron leakapge

3e5¢5¢3 BGP: BG potentizl zssociated with the core
spectrum

305.5.4 ¢ (COI‘E)
3450505 <EV 0

3¢5¢5.6 Power density

265¢5.7 ¢ {vessel), total and > 100KeV
3e5s5¢8 Fuel inventory and doubling time
Fuel Cycle Modelling

3¢5¢6s1 Fuel cycle model

ix
346

305.7

3e5e8

30549

345610

3e5611

345612

2 Equilibrium fuel cycle
«3 On the choice of U removal rates ¢ and b

Reactor Design: Configuration Trade Studies

3¢5«7e1 Inner blanket study

3eHeTe2 Optimum outer blanket thickness
3¢5aT7«3 Reflector

3e5¢Te4 Core tube material

Reactor Design: Controlling the Core >>°U/234y Ratio

3.5¢8.1 Basis for studying the ratio

3e548.2 Effect of the ratio on spectrum, BG,and Pu presence

3eDe8s3 Effect on reactor fissile inventory and doubling
time

3¢5¢8.4 Deducing the ratio on equilibrium cycle

Reactor Design: Choosing the Core Salt ThCl4/U01 Ratio

3
3.5;9.1 Effect on BG
3e5¢9.2 Effect on reactor fissile inventory and doubling
time
2e5¢9e3 Effect on power density and flux leve§§3 234
34549.4 Effect on actinide fission rates and U/“°™y ratio

Reactor Design: Choosing the Core Carrier Salt (¥aCl)
Content .

3,5¢1041 Effect on BG

3451042 Effect on reactor fissile inventsry and doubling
time :

34561043 Effect on pover,iensiiy and flux levels

3¢5410.4 Effect on EBXU 2°3U ratio

Mutation Effects

3e5s11e1 --Significance of §3§inides which emit alphas
3.5¢11e2 Significance of U production

3eDe11e3 Impact of fission product concentration upon

neutron physics performance

Summary

Safety & Kinetics

34641

Fhysics of Reactor Safety and Xinetics
W6e1a1 Effective delayed neutron fraciion
:6.1.,2 Prompt neutron lifetime, 1,

3
2.6,
3e6e1e3 Temperatus coefficient of relativity
3ebe2 Analysis of Normal Operations

3643

3e¢7 Tuel Processing

3eTe1

3eTo2

3073
374

Reactor startup
Reactor control
Reactor stability and inventory ratio
Reactor shutdown

Change in physical properties due %o transmutations
of Accident Situations

Small leakages

Loss of flow

Structural failure

Emergency cooling

Comparison of molten salt reactors to others
Precipitation-out of cutectic mixtures
Boiling~off of sali mixtures

Containment

Resistance to extermal threat

0 Molten salt combustion support

Principal Salt Reprocessing Methods

Chemistry of the heavy elements

Solvent extraction from aqueous solution
Volatility processes

Pyrometallurgiczl processing

Molten salt elezirolysis

Core Salt Processing

3eTe

*je

347
Je7
37
3e7
3eTe
3e7
3.7
3eTe

NNI\):\JI‘OI\)N

o
\ooa - OZW\N YR -

I\)N
.

Blanke%

Starter

3eTedel

Recovery options

Continuous removal of mutant gases
Removal of non-gaseous fission products
Control of the oxygen levels

Removal of sulfur impurities
Maintaining the chlorine stoichiometry
Storing troublesome fission products by
using them as carrier salts

Comparison to MSBR reprocessing
Materials requirements

Salt Processing
Fuels

Starter fuel for the core

xi
 

3.8

3¢9

147e4e2 Preparation of ThCl, for the blanket

4

3¢TeH Fuel Cycle Summary

Safeguards

3,81 National Fuel Inventory

3,8,2 In~plant Diversion Potential

34843 Facility Modification

3.8+4 Desired Radioactivity in Transported Breeding Gain
3¢8.5 Radiological Sabotage Threat

Access and Haintenance

3+9.1 Compactness of an IMCFR

3¢9.2 Intrinsic Reliability

34943 lNSRE Experience

3+9+4 Comparison of Primary Circuit Configurations
3495 _Reactor Maintance/Replacement Procedure
3496 EReactor Shielding

3,10 Auxiliary Plant

3.11

3+1041 Power Cycle Options
34104141 Intermediate ligquid coolant for steam cycle
34104742 Coolant for gas cycle

3.10,2 Secondary Hardware: Pumps

341043 rilling, Draiqing, and Dump Sysiems

3010s4 Overall Plant Size

Ecofiomics

Je11a1 British Study Results

3el1lelel - Capital-costs
3e11e142 Fuel cycle costs
Je11e143 Summa.ry

13
 

34.11.2 Transportation Costs
3e11e3 OQutage Penalties
341144 Other Cests

4,0 Summary

4e1 Pu, Proliferation, Safeguards, Security,and Waste
Management

4,2 Reactor Design

4.3 Technical and Economic Feasibility

5.0 References

xiii
1938 August 1

1957--1961

1962

1962

1962-1963

1963

1963-1973

1968-1969

1973-1975

197 5~Present

VITA
Born, Cincinnati, Ohioc

Chemical Engineering Cooperative Student
with the Aircraft Nuclear Propulsian Proj-
ect, General Electric Co., Cincinnati, Chio

Ch.E. University of Cincinnati

Summer Job as Technical Zditor, Mound Laboe
ratory, Miamisburg, Ohio

Laboratory Assistant, Phoenix Memorial Lab-
oratory, Ann Arbor, Michigan

M5. Nuclear Zngineering, University of
Michigan, Ann Arbor

Member Technical Staff, Atomics Internaticnai,
Canoga Park, Califormia

Atanics International Assignment to the National
Neutron Cross Section Center, Brookhaven National
Laboratory -

Guest Scientist at Swiss Federal Institute for

Reactor Research (EIR), Wuerenlingen, Switzer=-
land

Scientist at Idaho National Engineering Lab.;
Instructor with University of Idaho at Idaho
Falls

xiv
ABSTRACT OF THE DISSERTATION

Configuration of a Moltien Chloride Fast Reactor
on a Thorium Fuel Cycle

to Current Nuclear Fuel Cycle Concerns

by
Eric Heinz Ottewittie
Doctor of Philosophy in Engineering
University of California, Los Angeles, 1982

Professor William B. Kastenberg, Chair

Current concerns about the nuclear fuel cycle seem to center on

waste management, non-proliferation, and optimum fuel utilization

 

(including use of thorium). This {thesis attempts to design a fast
molten-salt reactor on the thorium fuel cycle to address these concerns
and then analyzes its potential performance, The result features

1¢ A simplified easy-to-replace skewed-tube geometry for the core

2. A ve}&-hard neutron spectrum which allows the useful con-
sumption of all the actinides (no actinide waste)

Je Reduced proliferation risks on the equilbrium cycle compared
to conventional fuel cycles because of the absence of car-
cinogenic, chemically-separable plutonium and the presence of
232U which gives a tell-iale signal and is hazardous to work
with

4e A breeding gain in the neighborhood of 0.3

XV
1.0 INTRODUCTION
1.1 Overview of Current Nuclear Politics and Technology
1elel Nuclear Fuel Cycle Problems and Their Politics

In recent years a number of concerns cover the nuclear fuel cycle .
nave arisen. A principal irritant is plutonium (Pu) production in the
existing light water reactor (LWR) uranium fuel cycle and in the fast
breeder (FBR) extension thereto. Concerned individuals fear that the
presence of Pu stockpiles mey promote the proliferation of nuclear
weapons, covertly or overtly, among nationzl and subnzational (eege
terrorist) groups.

In respornse to this concern, President Ford announced that the US
would no longer consider reprocessing of LWR fuel to be a foregone
conclusion. The alternative of course is to stockpile or to bury spent
fuels, thereby creating a problem of actinide waste management, That
would appear to be less than an optimum ecclogical alternative as much

238

of the urenium mined (mostly U) weculd never be used: mining trans-
forms mountain containing 10,000 partis of embedded uranium from natural
to machine - processed state for the recovery of T2 parts fuel, Of
that, reactors consume 7 - 10 parts; a few paris change into plutonium
and the rest currently becomes other -~ actinide waste,

If commercizl FBRs ever appear, the associaied fuel cycle must
21so contend with higher actinides such as americium and curium, Heat
source rate and alvha decay toxicity may economically prohibit their
fabrication into fuel rods,

These and other concerns also influenced President Carier to slow

dovm the plans for commercial FBRs.
1e1e2 Technical Solution: Thorium Fuel Cycle
The thorium fuel cycle solves some of these problems, By
starting with a material which is lower in atomic weight (A=232
vse 238) this fuel cycle eventually produces far less alpha-active
waste (especially carcinogenic and weapons-grade Pu) than the 238U/

233

Pu fuel cycle., Though U from the Th fuel cycle makes justi as

232

good a nuclear explosive, ii 1s always accompanied by U whose

dauchter decay productis emit 2 highly-penetratinz semma radiation.
This makes 233U hazardous to work withe It also signals its location
and transport.
1e143 Technical Solution: Molten Salt Fast Reactor

1ele3e1 Advantaczes of fluid=-uel reactors. Fluid-fuel reactors
continuously add fuel and renmove fission products and require no fuel
refabrication, Molten-salt reactors (MSR) have been more extansively
developed than other fluid-fuel power sysiems. They appear to offer
advantzges in limiting the proliferation of nuclear explosives: the
fuel cycle inventory of weapons~grade maierial outside the reactor is
very small and concentrated at the power plants; little stockpiling
or shipping of weapons-grade fissile materials occurs.
1eT14362 lLatroduction to the MSFR, Beginning shortly after World War
IT but extending into the TO's several laboratories looked at molten
sclt fast reactor (MSFR) concepts, Interest generally followed the
fortunes of the thermal MSR and so has waned in recent y=ars, Through-
out this time, the inherent absence of cladding in an MSR has hinted
of a neutron economy sufficient for breeding,

A thermal reactor doesn't breed easily; it requires careful

design and continuous reprocessing to minimize nonproduciive neutron
caztures in fission products, core sfiructural materials, and control
poisonse An MSFR is not so sensitive; it has no moderator and little
internal siructure. Cl and limited alkali (4ypically Na) are the
lightest materials present. Therefore, the neutron spectum remains
fast and fission products absorb far fewer neutrons,

Numerous workers have already studied the MSFR(U/Pu) as a
breeder, fission product burner, and ulira-nigh, fast-flux fest
facility with very hard neuiron spectrume This thesis systematically
examines the technical bases of !ISFRs and extends MSFR studies into
the thorium fuel cycle, MSFR(Th), seeking 2 system which can maintain
criticality in steady=-state oneration without accumulating actinides
deirimentally.

238U/Pu cycles differ in their neumtron eco-

The Th/233U and the
nomicse As a result, interest in the thermal molten sal{ breeder
reactor (WSBR) has centered on its special usefulness for breeding on
the Th/233U cyclee. Interest in the MSFR has focused on exceptional

2
breeding gain (BG) with the 38

U/Pu cycle [USFR(J/Pu)]: up %o
Oeb= 0e8e For an MSFR (Th) even with an anticipated reduction of 0.2«
O0e3y, BG= 0Oe3~0.5 may be possible,

Though not usually emphasized, the very hard specirum and good

neutron economy of an MSFR offer the prospect of usefully consuming

 

all actinide products: sooner or later all the transactinides fission,
If some capture initizlly, that only (generally) further deforms the
nucleus, thereby increasing the probability of fission, Thus the MSFR
offers strong potential for resolving current concerms, particularly

on a2 thorium fuel cycle,
 

1e2 Historical Review of Molten Salt Activities

The history of the MSR is as old and complex as the history
of nuclear power itself, The ups and downs have followed those
of the parent technology but the swings have been if anything more
violent,

1e241 American Activities
1e2+141 Develoovment of the SR« In 1947 ORNL began a stuly on the
dbhysics, chenistry, and engineering o7 traniume= and thorium=bearinz
molten fluorides, The potential for very high temperatures and power
density interested the aircraft propulsion project.

MSR technology first appeared in the open literature in 1957
(Briant and Weinberg[1]). Next, Bettis, et al. [2,3] and Ergen, et
ale [4] reported on the Aircraft Reactor Experiment: 2 beryllium-—
moderated, thermal reactor fueled with a UF4/NaF/ZrF2 mix and con-
tained in Inconel, This reactor successfully operated in 1954 for
more than 90,000 kihr without incident, at thermal powers up to 2.5
MW and temperatures as high as 1650°F.

This marked the start of an ORNL program to develop 2 thermal
Molten Salt Breeder Reactor (IISBR) for economic civilian power [5].
It included

1. Evaluating the most promising designs

2e Pinpointiry specific development problems

34 Developing materials for fuels, containers, and moderators

4+ Developing components, especially pumps, valves, and flanges

suitable for extended use with molten salts a%t 130000
5. Developing sfipplémentary chemical processes for recovering
valuable compenents (other than urenium) from spent fuel:

6. Developing and demcfistrafing the maintainability éf afi
MSR’system.

In 1963, Alexander [6] summarized the Cak Ridge development:

l. 1-Z: simplicity of the reacher core and the semicantinucus
fuel handling apparatus lead to low capital costs and in-
creased piant avoiiability.

2. The simplicity and_ccntinuous nature of fissilie and fertile
stream processing methods lead to negligible fuel cycle
ccsts in cn-site p-ants.

3., The negative temper-ature coefficient of regctivity inherent
in the thermal expansicn of the fuel proviaés safety advant-
ages over other reactor cancepts.

L. The internally-~cooled reactor offers competitive nuclear
performance; the externally-cooled reactor, superior per-
formance.

Thié effortALéd tb the design,iconstructioh, androperatlon-gf
the 8~th Molten Sait Reactcr ixperiment (MSRi). Critical operaticn
of the MSRE spanned the period from June 1965 to December 1969, dur-
ing which the reactor accumulated over 13,000 equivalent full-power
hours of operaticn and demenstrated remarkably high levels of opera-
bility, availability, and maintainability. -~

IA recent years, Battelie Nofthwest Lafiorat&fy (BWWL) has made
critical experiments on MSBR configurations [7], Several United

States miversities have studied the chemistry of molten salts [8].
 

ANL. identified the need [9] for in-pile corrosion testing at high burne
ups to clarify thc effect of noble-metal deposition on container metal,

Presently all work has stopped, A current problem was that
Hastelloy-ll limited the temperatures of the OHRNL MSBR: the additicn
of carbides into the grain boundaries keeps He from forming and
swelling there,but at high temperatures the carbides disappear into
the grains. E. Zeoroski of EPRI feels that the problem with the
MSBR was whether the system parts would hold together very long [10].
142.1.2 Development of the HCFRe Goodman et al [11] proposed a
molten~chloride fast reactor (MCFR) in 1952; Scatchard et al [12]
reviewed the chemical problers invelved. In 1955, Bulmer et 21,

[13] designed and evaluated = 500 MWtk MCFR, exterrnally cooled,

At that time the work was secret and only later decléSsified.
Chlorides of sodium, magnesium, uranium, and plutonium made

up the fuel salt., The blanket was depleted uranium oxide, cocled by
sodium, Thelr report contains numerous irade studies, some still
usefue.

Bulmer chose chlorides over fluorides to limit neutron moder=
ation, However, he felt tkat the strong 35Cl(n,p) reaction would
require enrichment in 37Cl. M, Taube [14=16] of the Institute
of Nuclear Research, Warsaw shared this view, More recently both
Taube [17] and Faugeras [18] have mentioned that fluoride salts may

233

work well with U in a fast reactor in spite of the extra moder—
ation, The MSFR work then shifted to Argonne Nationsl Laboratory
(ANL).

E. Se. Bettis (ORNL) feels that [19] ORNL's look at the MCFR was
very superficial since it was outside their thermal reactor charter.
They meinly criticized the high fuel inventory tied up in the heat
exchanger, Bettis thought that M. Taube had a good idea in direct
cooling (in-core, no fuel circulation) with boiling Hg [20], but

that insufficient quantities of Hg exist. ORNL looked umsuccessfully
at molten lead direct cooling: corrosion proved to be the Achilles
Heel,

In 1967, ANL swmarized [21] the fuel properties and nuclear
rerformance of fast reactors fueled with uranium and plutonium tri-
chlorides dissolved in chlorides of alkalis and alkaline-earths,
Included were the physical and chemical properties of the fuel, and
the heat removal and neutronzcs for one homogenecus reactor and two
internally=cooled reactors, The optimum core volumezfor 1000 MWe
power proved to be 10,000 liters for each types FEach exhibited
favorable characteristics of high breeding ratio, large negative
temperature coefficients of reactivity, and low fuel=cycle cost.
However, the unatiractive characteristics of large plutonium
inventory, large volume, complex design, and container material
problems indicated the need for a sizable program 4to develop the
MCFR(U/Pu).

Recently L. Ee licNeese [22] offered the following comments on
the !CFR:

1« Resistance of Mo to salt corrosion seems to depend directly

on the oxidation states present, as expressed by the "redox"
potentials, This corresponds to the directions of the Swiss

chemiczl research,
2 Re graphite vs. Mo as a structural material: graphite is
2 step up in technology requirements. It is also subject
to radiation damzse. ORNL stopped the engineering on it
for MSBR.

3. Re in=core vs, out-of-core cooling: .inwcore offers lower
inventory (SI), but hizh structural radiation damege. The
extra shielding for out-of-core is inconsequentiale Though
the out—of-core (Pu) version oriers a breeding gain of Q.6
~0s7 to counteract the extra inventory, the economic mea~
sare of interest (a2t least for U/Pu cycles) is approz=
imately 312 BG-1. “n summary out—of-core will be easiest,
but may be less eccaomic in terms of fuel cosis,

FPission product burner concepts are of littie interest:
only 1291 separation has been contemplated.

4. The Swiss molten chloride fast reactor designs presume that
2, stream of aboutl one ppm can be continuously directed for
reprocessing and returned., For this:

a, the chemistry exists (at least for fluorine
systems);

be the engineering ability exists;

ce the smallescale ORNL/MSBR plant was not finished;
and

de scaling up has not been done and would be a sig-
nificant undertakingem— much work to be done,

1242 Buropean Activity

1624241 Pola.nd/Swi'tzerland, Led by Mieczyslaw Taube, M. Taube
 

first, in Poland and later in Switgzerland, looked at a variety of
IISR concepts, primarily on the plutonium cycle. Principal studies
were on

1« A systém of about four fast CFRs and one MSR "burner®

of fission product waste [23]. The overall system con=
sumes its own fission product wastes and still has a
vositive breeding gaine The burner features a thermal flux
trap surrounded by fast reactor molten fuel, The higher the
specific power (IW/1 ) in the system, the lower the steady-
state fission product concentrztion becomes,

2. Optimization of the breeding retio for an HCTR(U/Pu),

theoretically achieving 146 = 148 [241.

The 10 IRl 1-1 power density assumed by Dr. Tauve far exceeds
the 70 - 80ki 1™ technology of ORNLe Thus, little direct ex-
perience applies, Still, L. E. HMcNeese (ORNL) was unable to
immediately foresee any intrinsic difficulty wita 10 M7 171 [22].

Presently, the Swiss work has sfopped, as the electorate
pushes the laboratory from reactor research towards general energy
research and development,

1424242 France, In France, Comme Energe Atomique (CEA) and
Electricite de France (EdF) sponsored development of the MSBR as
backup to the Superphoenix breeder, EdF emphasized lower powers
for near-term feasibility, Their long-term interests included
direct-contact in-core molten-lead cooling. CEA pursued higher
temperatures for long-term application, Their near-term interests

centered on experimental corrosion studies,
 

Fonteray-aux-Roses (CEA) and Pechirney Ugine Kuhlman (PUK)
studied several high temperature UMSR comcepts [18,25,26], a
Th/ 23%U cycle in fast and thermal fluoride-salt reactors and an
MCFR (U/Pu). The MCFR design features in—core cooling (Fig.
1.2-1) and a 7.62 m3 cylindrical core surrounded by a 30 m3
blanket, Table 1.2-I intercompares the main fast and thermal
1Sr. ceneepts of the French.
A CZA/FUK subccmpany, SERS, and Carbenne-Lorraine [27]
carried out experimental studies in 1974-1976 including
1, Tests of the mechanical performance of ihe materials
under tempcrature and radiation
2. Corrcsicn tests in salt at high temperzture
3. Cperaticn of loops at high temperature: one fcr iso-
thermal hydrodynamic studies, a non=isothermal cne with
small flow for heat exchanse studies, and a third non-
isothermal ane for corrosian.

In 1977, if the above results were positive, they were to decide on

a development program whose cbject would be the canstructicn of
a prototype of 25 = 50 Mith,

The American and French atomic organizations were also seek—
ing to agree on the fields of preparation for a 200 Mie MSBR demo-
plant. They were to decide in 1979 on the: construction of this
reactor, depending on the progress with high temperature reactors
and success of the French Superphoenix 1200 M{(e) Liquid Metal
Breeder Reactor (LMFBER).,

10
FUEL REPROCESSING

OVERFLOW for FUEL

HIONVHOX3 LV 3H

 

 

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= n_
<{ ’ J .
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= L) z W= u
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=0 a4 S
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COOLANT UClynacl

 

T

 

 

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w?flMM1 ARG

 

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HEAT EXCHANGER
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i

Figure l.2-1 French MCFR Design with In-Care Cooling

 

 

11
Table 1.2-I COGMPARISON OF FRENCH MOLTEN SALT REACTORS

 

Fast Thermal
CHARACTERISTIC Reactor Reactor
Power, Mw
Thermal 2050 2250
Electrical 1000 1000
Core specific power w/c 255 22
Efficiency % 49 Ly
Fuel, mol % 15 Pu013 71.7 LiF
85 NaCl 16 BeF,
12 Thr
0.3 UF &
4
3 : ja 6 238U 3
Fertile material, mol% 5 013 71.7 LiF
35 NaCl 16  BeF
12 T
0.3 UFh
Secondary coolant AlCl3 Na.BF4
Working fluid Dissociating Steam
Né)h Gas
Core fuel velocity, m/s 2 2.5
Neutron spectrum 0.01 Kev to 10 Mev Thermal
Moderator None Graphite
Structural material
reactor and tubes Molybdenum Alloy Hastelloy=N

(Nickel or Iron
Added)
1¢2,2,3 England, British workers studied MSRs in 1954 and 1965,
lhe MSFR interested them the most and they felt that such a study
would complement the U3 MSBR program; A preliminary study of a
fast system using the 233U/Th cycle and fluoride salts did not
Look encouraginé, sé they refécused on a 238U/Pu cycle and chloride
salts. Work on salt chemistry began in 1965; in 1970-1972 the
procram extende< “c include other materials aspects.

Jerkers at Harwell and Winfrith examined three veriants of
£ 25C0 Mie MCFR(238U/Pu) [28-31]: in-core "direct" cooling by
molten lead drops, in—core ccoling by blanket salt passing through
o tubes,and cut-cf-ccre coo-ing.

The direct scheme enccutered too many proclems. The in-ccre
cooling cecncept appeared to warrant further study but the fuel
inventory did not appear to be as low as first thought; also the
high velocities and high pumping pressures presented serious
design problems. ‘The corrosion limits and strength of molybdenum
or its alloys at reactor temperatures representied a large urknown.

With out~of-core cooling, at first the fuel inventory in the
reactor circuit was tco high. Hcwever, compzct layouts and higher
(but still achievable) heat exchanger volumetric ratings reduced
it to within reason. Figure 1.2-2 illustrates the design. The

16 2 -1 13

neutron flux was 3 x 10 ncm ~ s = in the reactor and 3 x 10

An the heat exchanger.

1.2.,2., Soviet Unicn. Very little is known -about Soviet Union .
engineering studies. However, extensive salt thermodynamic and

physical property studies, evidenced in Section 3.3, suggest a

13
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

. | Y _“‘r“—]
L0 |
— ;
S Vo — '
i ./‘ ~ :!\_ - 3 }‘
Tt | ~= —
» + -
4 .
8 l' x I
; !
| A VaT TR
L= Be- } NN~ T
SO ‘ H,
: ]
-~ ‘ '
|
p " 1A
/ \
~ s ’ -
| ==t N\
- e oaim ! S
I
O b 34 M
SCALE
| CORE
2 BLANKET ¢ LEAD OUTLET
4 FUEL SALT PUMP—8off It HIGH TEMPERATURE
S BLANKET COOLER— 40off RESISTANT LINING
6 BLANKET COOLER PUMP-4o0ff 12 DRAIN LINE
7 ISOLATION VALVE i3 FLOW DISTRIBUTOR
8 LEAD INLET 4 SUPPORT FLANGE
Pige

1e2=2 British 6000 Mith MCFR Design; Lead-conled, Oubmof-Core,

14
 

much larger supporting research program than ir the Western
world, A USSR review book on Liquid-salt Nuclear Reactors
regards MSRs as reactors of future with compact and promising
fuel cycles, They conclude that the developmental focus must
be on the extermal fuel cycle, ergo the USSR attention to salt
properties,
1e2e3 Summary MSR Statewof-the-Art

Both the 2 Mith ARE and the 8 MWth MSRE demonstrated ex-~
tended successful MSR performance but under conditions less
strenuous than those considered for a 1000 MiWe or MCFR. The
question remains: will a circulating=fuel system hold together
long enough to be practical? ore specifically, how will the
candidate container materials -« Mo alloys, graphite, or compos=
ites with these = hold up at high temperature to corrosion by
the variety of mutants and oxidation states which high burnup
produces? The answer to such questions will come from in-pile
corrosion testing. The French have an on-going experimental
program in such engineering studies, but it may be proprietary.

Much work needs +to be done in developing and scaling up
the reprocessing system.

With out~of=core cooling most of the fuel circulates
outside the core. This threatens the reactor stability. In-
ventory reduction requires compact layouts and high heat

exchanger efficiency,

15
2,0 ANALYSIS OF CONTEMPORY CONCERNS ABOUT THE NUCLEAR FUEL CYCLE

AND OTHER CONSIDERATIONS AFFECTING CEOICE OF AN ADVANCED CONCEPT

2e¢1 MNuclear Fuel Cycle Waste Management

One primary facet of the nuclear waste problem is that reactor
operation induces short-and intermediate-lived radiocactivity in
materials which had “een stable or only long-lived radiocactive, The
2clution is to altermatively store the activated materials until they
decay, or to transmute them back into harmless stable or quasi-stable
nuclides, When operation produces nuclides which poison the enviro-
ment 2xnd 1ast long, then ecology prefers the added speed of transe
mutation,

Fuel cycle wastes occur in preparation of the fuel (mill tail-
ings), reactor operation (spent fuel and activation of air and water),
and reactor decommissioning (activation products of strustural
materials), The latter are geperally nonvolatile, bound up in the
structural material, and extremely difficult to release to the
enviroment, even in accident situations, The mill teilings are in-
herently low-ievel, Holdup tanks and stacks with large dilution
factors manage the air and water activations, Managzament of the
spent fuel poses thz major problem: all others pale in comparison.

In the absence of US reprocessing,spent fuel has accumulated
recently such as to saturate the utility storage pools.s To prevent
loss of nuclear generation, DOE plans to find away-from-reactor
storage capacity for 810 MT of spent fuel by 1984 and at least
25,000 NT by 1996, Foreign spent fuel will add to these require=-

ments [32]e

16
24141 Characterization of.Spent Fuel Wastes

Fissian products and transmuted actinides emit most of the
radiatian in spent-fuel, Some of these radioisotcpes exist for
long. times, comparable with a human life span. Those with half-
life of a few to 50 years are particularly hazardous because they
radiate faster than the longer-~lived ones. How lang each element
remains in a human or animal depends on 1ts biological-elimination
half-life, which may be very long.

Some half-lives exceed the life span of our society. A concern
then arises that nuclear power might burden the genera?ions to come

with accumulated waste, part-cularly the transactinides (Figure

  
 

2. 1_1 ) .
Relafilve
Toxicity
. 1 &
N
w-l"
: ]
Figure 241w 1
1974
.Bazard from Fuel 1
1
Wastes [33] m*:
4 actinides
134
1™

 

 

 

17
 

2.1.1.1 Cases. Spent fuel gases generally include He and T2 from
ternary fissicn, Hz, Dz, T2, and He frcm charged-particle-out thresn-
cid reacticns, and fissicn procduct gases {Xe, Kr, and Iz). In either
a molten salt fuel or in disscluticn of a solid fuel, gases serarate

cut easily.

The IAEA Code cf Practice for Safe Handling cf Radicnuclides

{341 classifies radicnuclices into fcur groups according to thsir

racicotoxicily per unit activiiy, primarily upen irhalaticn. £n 40Pk

(Th) cculd produce the following nuclides in gaseous form (half-lives

in parentheses):

1. Greup I (very high -adictoxiciiy): ncns
5 5 SS
2. Greup II (high radiotoxicity): 301 (3x1C7y, , Lo (8.0 &),
ard 231 (20.8 1)
: . 1€, 38...
3. Greup IIT (mcderate radictexdicity): ~ F {109.8m), “ Ci
_ 8 . 7 . )
(37.2m), E%® (16.72y), kr (76m), B (2.29 1), BT
(52.6m), Y% (6.6n), P%e (9.1n) .
9 i \
L. Greup IV {low radictericity): X (12.33 v, 12% ".oxlC7y))

L
1 \ 132
BlL.m (11.9 dj, 7“Xe (5.25 d) o

Cf these only 3H, 1291, 85Krm, and 3601 (in order of increasing
hazard) have half-lives long enough tc warrant processing: the rest
just require short-term holdup before releasé to the atmosphere,

Low hazard tritium relatively emits only weak radiation; its
maximum allowable concentration is amang the highest far any
radicactive material. Furthermore the body excretes it rapidiy end

it cannct tioclegicaily ccncentrate in the envircnment, in feood chains

18
or in man. It decays with a 12.3 years half-life to harmless helium
and thus poses no Lcng term hazard. It also has applicatien for the

fusion prcgram. Nevertheless, it must be manitored and is hard to

contain.

8 -
SKrm has a similar half-life but is far more toxiz. A4s an
inert gas, it is easy to isclate. Both 12912 and 36012 gases
might be returned to the reactor as halide salt. There they will

most likely transmuie to harmless 3701 and lBOKr.

2.1.1.2 Solid fissicn products. Storage for about a year (TFig.
2.1-2) reduces most fissicn product radioactivity to more manage-—
atle amounts. Cf the remaining radioisotcpes the most hazardous are
thcse which metzbolize and beccme a part of living crganisms, and

decay with half-iives comparable to a human life-span.

2.1.1.3 Actinides. Management of actinide wastes has caused
much controversy in recent years [35-49], Their long half-life
makes actinides a quasi-permanent burden unless one can transmute or
otherwise remove them from the enviroment, Alpha-emitters threaten the
most , particularly when ingested. Plutconium is the worst of these:
inside the body it seeks out and locates on banes, which makes it
carcinogenic. Secticn 2.1.1.4 discusses this further,

Most aétinides though usually in zn unwieldy form can also

be used for weapons Sectien 2.2 analyzes this.

19
Activity (curie/watt)

 

 

 

 

 

 

 

 

 

 

10 )
|
-14 1 day 10 days 100 days 1 year 10 years 100 years 1000 y.
10 s l | ] ! | ] |
T T T T T T
10 10° 10° 1a’ 108 10° 10t

time, seconds

Fize 2,1=2 Fission Product Decay

20
2.1.1.4 Unique biological hazard of plutonium [50]. Most chemical
elements react chemically with biclogical systems, some of them
detrimentally. Cthers may be safe in elemental form, but not as
compounds. Individual isotopes can also harm living matter by einlt~
ting dangerous radiation. All the isétopes of actinide elements
radiate. Plutonium predominates amang these elements because it
comprises five quasi-stable isotcpes and constitutes the first
significant element above starter uranium on the A scale.

The main isotopes of plutonium, in order of decreasing
importance are 239Pu, 2L‘OPu, 2hlPu, 238Pu, and 242Pu. This reflects
their usual cencentraticns and activity rates. In high-burnup
LMFBR fuel, Pu-241 presents the principal short-term hazard (Table
2.1-1), 239by emits 5.15 leV alpha particles which range only
3.6 cin in air and less than L5/f&m in water. In each ionizing
collision with air or water meolecules, the alpha leses 35 eV.

After about 120,000 collisians, it has lost all of its kinetic
energy. It then stops, captures two electrcns, and changes into
a peutral helium atom. Noble gas helium does not affect tke

human body.

Cutside the human body the alpha radiation poses no threat:
5.1 MeV alpha-particles can penetrate the skin from outside only
to a depth of hé/Ukm. This is still within the epidermis layer
which regenerates very quickly.

The short track length of an alpha~particle in living tissue
concentrates the energy abscrption in a relatively small volume

thereby increasing its affect. Thus to compare toxicity of alpha
¥ N

21
TABLE 2.1=I

RADIOACTIVITY IN FUEL 120 DAYS AFTER DISCHARGE FROM
A 100-MWD/kg, 1000-MWE FBR

 

 

 

 

 

Equilibrium ngéag"“ Level Alter
Decay . Discha;g_e 2Y2 o Sooline
Isotope Mode Half Life |Composition Curies per kg | Kilocuries
' (mM'I?Trhx) of Discharge per M7 9[
' Isotope Fuel Mix
U-232 a 1736y 0.014 2,09 x 10% 0.0
U-237 a |6.754d 3.0 356 0.0
U-238 e |s.5x107y | 909kg 33sx10 | .0
y-239 8 |23.5m 1.2 2 %1077 0.0
Np-236 | @ |22k (57%) | 0.000034 | 5.5 0.0
Np-237 | 8 |2.14x178y| 17 0.706 0.0
Np-2318 | 8 |2.104 0.112 224. 0.0
Np-239 | B |2.35d 179 310, 0.0
Pu-236 | a [2.85y 0.0026 5.04 x 10° 0.0
Pu-238 | @ |89y 246, 1.63 x 107 4.1
Pu-239 | « |24,400 y 40.4 kg 61.38 2.5
Pu-240 | & |6760y 17.2 kg 221. 3.8
Pu-241 | & {13y 2.3 kg 1.13 x 10° 260.0
Pu-242 | @ |379,000y 1.2 kg 3.89 0.0
Pu-243 | B |4.98h 0.05 0.1 0.0
Am-241] @ 453y 176 3,25 x 10° 0.6
Am-242( 8,7 |16k (81%) | 0.064 9.6 0.0
Am-243| « |8000y 96.5 1.85 x 10° 17.9
Am-2e4| 8,y |26m 0.014 2.7 x107° 0.0
Cm-242] a |l162.5d 16.2 2.1 x 10° 34.0
Cm-243] a |35y 0.257 4.2 x 104 0.0
Cm-244| & |18.4y 6.50 7.95 x 10% 0.5

 

 

 

 

22
with the well-known results for X-rays one must multiply the energy
absorptian for alpha-particles by the "Quality Factor", varying from
10 to 50.

The most dangerous situation arises ‘when plutaniwn eaters tae
human body. Alpha-particles can then penetrate into the tissues to
a depth of AS’}Anu Tissues generally are shielded by mémbranes,
but these are thirner than l‘}i.m.

As water ccmprises mcre than 70% by weight of human tissue,
the icnizing penetraticn prucduces hyd-ogen perecxfide tarcugh radio;
ysis. Peroxide acts as a very strang poisen inside living tissue:
Ithrough a series of chemical reactians it changes the structiurs in
the enzymes which catalyze biochemisiry insicde the livingcells
Prolongéd irradiaticn by alprha-particles also deforms:nucleic acids,
the carriers of genetic informaticn.

Cur widest experience concerning biolcgical effects of alpha
emitters is with fléRa. Seventy years of handling suggests taat
the body burden limit for occupatianal exposure tc it be 0.1 ‘}LCi
(O.lupLg). Experiments an dogs and other animal; indicate that

Pu produces about five times greater biological effect.

Plutonium, like calcium, strantium, radium, and cerium, forms
non-saluble phosphates which deposit in the bonas. For these elements

the skeletan is the "critical" organ.
,gglufiggigfi binds strengly to the sialprotein on the bane surface;
the most abundant component cf bone, callagen, does not bind plutanium

to any significant extent. This localizaticn is what makes 239py

23
 

Q
so carcinogenic. The lowest average dose rate of 23’Pu in bones
causing death of rats from osteosarcoma was 57 rad (about 15 times
less than the lethal dose rate from 9OSr).
In contrast thorium and uranium exhibit low radiocactivity. Alpha-
. 232U 233 . . .
emitters and U also endanger biological systems but they don't

"seek bones" like plutonium. This recommends a Th/233U fuel cycle
over a 238U/Fu cne.

2.1.2 Management of Spent Fuel Wasties

One can either store long—lived spent fuel wastes or transmute
them into non-=hazardous products: +those which are stable or near-
stable, or which shortly deccy into same (Figure 2.1-3). Studies
to date would transmute wast2 in Controlled Thermonuclear Reactors
[ 52-55], the Savannah River Plant high-flux production reactors
[56,57], power reactors [56=60], accelerators [61], and others
[62-64]. Taube [23] and others [hd] have shown that, of the trouble~

90Sr and 13705 most resist transmitation:

some radiation emitters,
they have low cross sections (¢ 1 barn) for thermal neutron capture
transmutation into 91Zr and 138Ba. However, a thermal flux trap in
a fast molten salt reactor could provide sufficiently high flux to
reduce the effective half lives from 29 and 30.1 years down to 2 and
9 years, respectively {23]. A solid fuel reactor could not do this
because of technical limits on specific power.

Sometimes, the same properties which rendsr an isotope hazard-

- . . . s 0
ous, also make it attractive as a commercial radiatlon source. 7 Sr

and.137Cs are examples. However, commercial demand does not threat-

en to exceed supply.

24
 

Half-l1ife, years

>»10%9 -1

    
  

 
 
    

 

10
10 stableLhort

nuclides fliving

\{.P. F.P. //
P
\spontareouse /

beta cEcay //
quasi stable \\

F.P. \\ /, S 9

    
   

9
10 10

     
 

on
o
ry

1
O
Q

 

 
 

1g

transmut

   

     

long living
F.P.

  

10 10

10

Pigure 2,1=3, Transmutations of Fission Products

25
 

Actinide "wastes" accumulate when nuclides resiét fission-
ing; Figure 2.1-4 indicates the prominent actinides of the Th
and U fuel cycles and their fission {,endencies. More explicitly
Figure 2.1-5 shows that fission génerally does not predominate over
capture until rather high neutron energies are reached. However,

&; does exceed o, for 23 L‘U at mich lower energies tha.ri for other
actinides. Thus, a verv fast snectrur reactor, esoecially cn the
therium fuel evele can uniocuely zvaid accumulatine actinides.

According to their median flux energies (Tabie‘ 2.1-I1), fast re-
actors must have many more neutrons above the F‘iguré 2¢1=5 crossover
points than thermal reactors., MSFRs which minimize neutron
mederaticn (elastic scé.tter;ng) will have the hardest neutrcn
spectrum of all power reactors. Reactors with highly-enriched
fuels (Table 2,1-II) will also exhibit a faster spectrum: en-
richment eliminates from the core 23 8U or 23 2Th, which primarily
undergo inc].a:.‘bic_ scatter.

In considering the reactivity contribution of each nuclide
note that the reactivity crossover point in Figure 2.1-5 where

N = ’ng:. / O‘c exceeds 1, will occur at even lower energies

(roughly where S¢/6. =0.3-0.4). Thus, nuclides usually considered
as actinide wastes do not severely detract from criticality or

breeding gain potential in a very fast spectrum reactor.

26
232'1}1__,_2331,1_;—-23 Z"u---p--'?'B 5U—-v=-23 6U—=-23 7Np---23 8}:"4:4—--'---23 9Pu

fissicn

fissicn

fissien

238 23%y 2 0py  2lpy L BL2py 203y Pl

i

fissicn

fissim

Fig. 2.1-4 Burnup Chains far the <°Th/°>%y ma 2% /23%u Fuel Cycles

Table 241-11.

Fast Reactor Comparison

 

 

Fuel Mediun Flux
Cycle Reactor -Energy
7 238
U/Pu  ANL 1000 MW(e) LIFBR 130 keV
AT 1000 MW(e) LIIF3R 180 keV
Fast Na ZPPRs 190 keV
GCTR Lattice 187 keV
1000 MW (e) GCFR Design 176 keV
MCFR Ein-core cooling) 198 keV
MCFR {intermal blanket,
out-of=core cooling 370 keV
Ultra-High Fast Flux 471 keV
Molten Chloride Test
Reactor
Th/233U MCFR (out-ogsgore cooling,
high U enriched) 700 keV

27
18.0

1.04

230y

Gf/Oc

 

1 MeV

 

 

0.1
100 keV
E neutron

Figure 2.1-5 Enhanced fissionability of fertile nuclides at higher

peutron energies,

28
2.2 Non~proliferation

2e2¢1 Development of US Policy

US nuclear energy policy was from the first based on a Keen
awareness of the dangerous aspect of nuclear electric power, The
Acheson=Lilienthal Revort of 1946 saw a close association between
the civilian and military aspecis of nuclear energye. It recommendec
international ownership of nuclear explosive materials,

The related US proposal to the United Nations failed in part
because the Soviet Union would net participate, After that the U5
withdrew into a periocd of secrecy. Eventually we became more relaxed
about the development of nuclzar energy for peaceful purposes and
abandoned the secrecy. Because we mistakenly thousghi civilian
reactor safeguards could be stretched to cover the more danzerous
elements in the fuel cycle ~ such as plutonium renrocessing - we
allowed plans for the use of plutonium fo go forward unhampered.
There is where the damage was done [65].

In 1957 the International Atomic Energy Agency IAEA was estab;
lished primarily to monitor the flow of commercial nuclear materials
and equipment among member countries, The charter advocates conirol
over “excess" quantities of plutonium, but does not restrict their
use,

The nuclear policymakers of the fifties and the sixties
apparently did not realize the security implications of easy access
to nuclear explosive material in national stockpiles., The prospect
of many nations possessing substantial nuclear explosive materials

seemed very far away, A4lso, nuclear weapons were thoughi to be

29
enormously difficult to design and fabricate, The US near-monopoly
on the technology, fuels, and equipment for civilian nuclear power
activities worldwide seemed to ensure US control of the situation,
Fledgling nuclaar power programs seemed to be unrelated to the devel-
opment of nuclear weapons., The earlier prescience of the Acheson-
Lilienthal group that they had everything to do with it was ignored.

There was also some genuine confusion on the technical side.

It was once widely thought, for exemple, that you could not make
nuclear weapons with plutonium derived from spent power reactor fuel,
The Acheson-Lilienthal report seems to have misled many, including
the IAEA into believiang you could denature ovlutonium. As 2 result,
many of those responsible for protection against military diversion
of plutonium were relying upon technological barriers which did not
existe

On October 28, 1976, President Ford concludéd that "avoidance
of proliferation must come before economic interests"e Therefore,
we should defer reprocessing until "there is sound feason to conclude
that the world community can effectively overcome the associated
risks of proliferation.”

Two factors contributed to Ford's decision and to the sub-
sequent development of President Carter's non=proliferation policy.
The first is how the safeguards deficiency impacts international
security: we cannot control the plutonium or highly-enriched uranium
of other nations should they suddenly decide to abrogate agreements
and aprropriate explosive material for weapons. At the same time

the extensive zrorth of the nuclear industry, worldiride, is increasing

30
the availability of plutonium,

The second factor affecting current policies is the dubious
economics of reprocessing and recycle of plutonium in light water
reactors, This makes the early introduction of plutonium into
international irade as unnecessary as it is dangerous,

Presidents Ford and Carter both advocated that we restrict
access to dangerous materials, pause in the commitment to plutonium
separztion and use, znd search for alternatives to national stocke-
piling, Their efforts to persuade our allies and trading partners
mct at best, with mixed success, It is hardly surprising that this
US policy shift on plutonium produced widespread irritation, alarm,
and even the cynical suggesiion fthat the Americans were less inter=-
ested in proliferation than in perpetuating a commercial advantage,

Both Britain and France have emphasized the security advantages
of their reprocessing plans, The British suggest that if reprocessing
does not relieve the accumilation of spent fuel, non-nuclear-weapon
states will have to develop indigenevus reprocessing facilities:
this will certainly pose a serious danger of proliferation, Britain
argues that the best way to discourage this development is to confine
reprocessing services to the states which already have the bomb [66],

Carter subsequently submitted a bill, the Nuclear Non-Proliferation
Act of 1977« It would impose stricter rules on US nuclear exports
which would block their use for explosives., Opposition to this act
argued that whatever relationship exists cannot be reversed and that
it is so remote that further controls aren't necessary,

Sam }icDowell of DOZ Safeguards and Security has defined non-pro-

3
liferation as prevention of weapons capability in a nonweapons
state [67]s The goal of an alternmate fuel cycle would be to deter
or reveal diversion such that suitable time would be provided which
allowed other countried to negotiate or induce that country to stop,.

Zebroski of EPRI has suggested [10] that the primary objective
should be to seek improved resistance to diversion from civilian
power fuel cycles,

The heart of the controversy lies in the access to nuclear
explosive materizls which the nuclear eleciric power indusiry pro-

vides, Section 2,2,2 discusses +this further,

242.2 Relationshiv Between Nuclear Electric Power and

Nuclear Weapons Development

The fi.S. generally prodfices plutofiium for weapons purposes in
large special purpose reactors and separates it from fhe irradiated
fuel in govermment reprocessing plants. However, nuclear power
generation electricity also produces plutonium; commercial spent
fuel reprocessing planis will separate it, physically creating large
stockpiles, Also the same technolozy (and in some cases the same
plants) which enrich uranium for fuel can also enrich it further for
Wweapons use,

The essential point is this: getting the requisite explosive
material iz still the most difficult and time-consuming item in the
initial vroduction of nuclear weapons,* The operation of civilian
nuclear power reactors and plants for fissile fuel geparation and/or

enrichment tend to remove this bottleneck, Then if a country has

32
prevared secretly in advance, it can quickly manufacture nuclear
warneads once it decides to do sn,.

This is now a real threat, not present when the basic internate
ional rules for nuclear itrade were formulated twenty years ago. The
reason is that the civilian nuclear power industry has grown enorme
ouslye. By any reasonable measure (the plutonium production rate or

the size of urenium enrichment facilities being wufilized), it exceeds

the scale of the worldfs military nuclear orozramse In fact, in most
countries the quantities of plutonium is spent reactor fuel, if sep-
arzted out and stored, will dwarf any plausible military needs,
2.2.2.1 Txtent of the svent fuel problems Over the next decade Zurope
and Japan plan to send 3 G3 (2etric kilotons) of spent fuel tc Winde
scale, La Hague reportedly will process about 6 Gg sfient fuel for
Japan, the Federal Republic of Germany , Sweden, Switerland, Bel=-
gium, Holland, and Austria, This means separating about 75 Mg (metric
tons) plutonium, or enough for about 10,000 nuclear weapons. Clearly
heavy financial investments ride on the cutcome: when transport
charges are included, the Buropean contracts represeni almost three
billion dollars in business. The intense internafional competition

in nuclear commerce, accompanied by heavy investment and national

 

*¥See the 1975 Encyclopedia Americana article on Huclear Weapons by
John Foster, then the Defense Department's RED chief and a former
director of the Livermore Laboratory: "the only difficult part of
making a fission bomb of some sort is the preparation of a supply of

fissionable material of adequate purity; the design of the bomb itselfl

is relatively zasy",

33
 

pride, tends to obscure the proliferation threat,
2.2.2,2 Question of reprocessing. [32,65,66], Current U,S, policy
supports the relatively safe fuel cycle activities and discourages
the more dangerous ones. A primary U,S,. imperative is to promote come
mon rules for internmational nuclear trade, But before formulating
rules, much less implementing them, one mmst ascertain just what is
dengerous and just how much the spread of nuclear weazons threatens
individual countries and world security. Clearly 2 common understand-
ing does not yet exist, as witness US opposition to Buropean export
sales of pluitoniun reprocessing and uranium enrichment_facilitieso

Recent U,S. Administratisns have shunned reprocessing. Many
analysts [68] believe this directly increases the risks of both
diversion and proliferation, With no reprocessing, the increasing
amount of long-cooled spent fuel, and the increasing number of lo-
cations with such fuel [69] increase diversion risks, The likeliw
hood of "proliferation" increases because concern for a reliable
supply stimulates development of independent enrichment capabilities,

The capability to develop fissile production now exists in fifty
countries, with 10~20 additionzl possible in the next decade [70;71].
A secure fuel supply and reprocessing service through international
commerce would economically discourage these countries from devel-
oping their own., The economic inhibition is substantial since initial
small-scale operations are highly unecomomic relative to ma?ure MAN U=
facturing and production capabilities in supplier countries,

Section 2.2.5 examines the vulnerability of different fuel

cycles to diversion,
Pe2e2¢3 Attitudes on return of Pu, Of prime importance to pro-
liferation is the question: who gets the bred Pw —the reprocessor
or the spent fuel owner? If plutonium eventually displaces uranium
as the primary fuel for power reactors, how can indigenous repro-
cessing be successfully discouraged outside the nuclear weapon
states or a big suppliers club? The goal, long-sought by many,

of fuel "independence'" does not equate to simply shifting fuel
dependence from the United States to Europe, These

commercizl and political pressures, which earlier led to reprocessing
ccuid yet force Fu return even witihcut adequate internaticnal
rrctecticn against preliferazicn.

Presently ncn-Zuratom countries need US permissicn beth to
transfer US-supplied spent fuel to cther reprocessoré like “ind—
scale and La Hague and to get Pu back. Pu can return cnly "under
conditions that...warn the U5 of any diversion well in advange of
the Yime" diverted fiaterial could be transformed into weapons.
Ingland has pledged the material will be rsturned to ils owners
"enly in a form that will reduce the risks" of proliferation. In
his report cn the ‘/indscale inquiry, British Judge Parker took
refuge in time: 'this matter can be alleviated to some extent

oy technical fixes", he wrote.

35
2.2.3 Which Reactor Fuel Contains the Least Proliferaticn Danger?

All reactors either produce Pu or 2'33U, or start with fully-
enriched 235U. Thus weapans—grade fuel is intrinsically recover-
able from all reactor fuel cycles. However, which fuel attracts
nationalist and subnational groups the least as a weapon? The
answer may depend on unraveling complex interrelated factors,
whicn are examined here,
2.2.3.1 Veapons-grade material. For 233U or 235U, weapons grade
means isotopic content enrichment in U exceeding 12 or 20%,
respeptively. In the broad sense weapcns~grade means any material
which can undergo chemical :eparatia to yield Pu or the uranium
enrichments menticned above. Thus dilution with any other element

including thorium does not lower the weapons grade as chemistry
can separate out the U again.

At one time it was thought that reactor-grade plutonium (pro-
duced in thermal reactors) was not of weapons grade [66]s The U,.S.
covernment has now stated unambiguously that they have produced and
successfully tested militarily important nuclear weapons, using
reactor srade plutonium. Even simpnle weapon designs reliably produce
highly powerful (kiloton) explosions.

Similarly; the TARA Safeguards Technical Manual now provides the
following cuidance: plutonium of any grade, in either metal, oxids
or nitrate form can be out in a form suitable for the namufacture of

nuclear explosive devices in a2 matter of days to weeks [65].

36
 

2.2.3.2 Unique 23 2U Daughter Radiatian. All weapons-~grade fuel
can be left in a highly-radiocactive state (by means of fission
products). The associated biological hazérd would then discourage
its use by subnational weapon makers. Special national labor-
atories,could readily clean these up bui ever-present 23 2U inpur-
ity in 233U will continue to decay into hazardecus ‘(-emitters (Sec.
3.5.4). In less than a week after high level decontamination, the
233

gamma activity in recovered Th and U becomes so great that
fabrication by direct methods can be permitted anly on a scheduled
radiation dosage basis [72].

> + 3 3 3 3 £ = 2320 >

The characteristic har ‘-gamma signal cf the daughter

radiation also reveals the location and transport of this material.

2a2e303" Preigniti_@.‘ The-expiosive yvield of .a f'issicn bomb
depends on how long the imploded configuration stays together.
Multiplied source neutrons fram spantaneous 21”0Pu fissians cause
gearly energy r;alease which produces a premature disassembly.
This greatly reduces the' total energy yield of the devics.
Cansequently, a military might shun hizh burnup plutonium

(high in %%Pu), but a subnational group might still desire it.
In pure actinide metal only spaitanecus fissian can con—

tribute copious source neutrons. However actinide compounds pro—
duce intrinsic preignition-triggers through (o ,n) reactions.
Table 2.2-I compares the typical activities from the'se neutron
sources. No Th/cycle isctope can match the 2L‘LOPu spantaneocus
fissicn rate. 232[] comes closest in half-life (Figure 2.2-1),

but its concentration lies magnitudes lower. Thus the neutron

37
Table 242«1 Weavon Preignition Triggers in Pertinent Fuels

 

 

 

 

232U 231‘U le,opu
K half-life (3) 71.7 2.45%10° 6540,
& (&/sec/nuclide) 3.1x100°  g.ox1™H 3.xc R
ol energy (MeV) 543 L8 5.2
SF haif-life (y) 8x10% 2x10%° 1.3x10Mt
Ve (n/dis) <2 22 ~2
M5F (n/sec/nuclide) bx107% 2x10™24 . 3x10°17
Typical isotopie 1 1 1
enrichment 1000 5 5
A {n/sec/atom) 6310727 1x10722 1x10"2°
fim(n/sec/atom)*
\ in F 1x107 18 2%10~17 8x10”
in CL x10=20 2x1071 x10~20
in O 2x10~20 1x10~1 x10~20

 

* Neutron yields per 108

‘A 5 MeV-alphas are 1200, 11, and 7 for

targets of F,Cl, and @, respectively according to Roberts [73]

38

 
 

-LIFE LYEARS)

 

SPONTANEQUS FISSION HALF

 

 

 

4 35 3% 37 ¥ 39 40 4 R

Fige 2.2-1 Spantanecus fission half-life of even—even nuclides versus

Zz/A.

39
 

production rate from 232U and.23hU in with 233U falls many
magnitudes below that from 2l“OPu in Pu.

With high C1,F, or O concentration (as in salt or oxide
compounds) the (o,n ) neutron rate (Table 2.2-3) eguals or
exceeds the spontanecus fissior. neutron rate of metallic Pu
(7x 10_20n/s Pu atcm, 20% 2l‘OPu). F is much stronger than
Ci cr O, in this regard. This should appreciably reduce the ex—
233

rlcsive thr=at frem U weapons in fluoride, chlcride, or oxide

chemical form. The («,n) neutrons will also present a streng

actlvation hazard except when heavily shielded.

2.0.0 .0 238?u hign heat cere-ation. 238Pu nas Iound cansiderabie
applicaticn as a radioisotope heat source. Because its high heat
generation rate will melt practical sizes of the metal, the usual
form is PuC,. For a large enocugh uncocled mass, even the oxide can
melt. Cxide sources also preduce neutrons through 180 (e\n )

reactians. Together, the neutrons and the heat make plutonium

high in 22¢

Pu content cumbersome to work with.

. 238
Two principal rcutes produce ~7 Pu,

239y (n,2n) 238y, (2.2-1)

235 (n, ¥) 3% (n,y) P (n,¥) 2B (2.2-2)
Route (2.2;1) enhances productions for very fast plutonium reactors;
Route (2.2-2), for high enriched 233y or %%y reacters. In high

238

burnup U-fuelled LiRs, Pu isotopic content reaches 1%. In a

Th/233U or a high 23%)—enriched reactor, with little or no 2%

238

present Pu should dominate any plutonium separated out.

40
 

2424345 Denaturing and other technical fixes [32], One technical
fiz, the Civex process, has received a fair amount of international
attention, It returns pluftonium to the customer in the form of
fabricated fuel containing enough radioactive products 4o preclude
easy chemical exfraction of bomb material, However, within a few
years after leaving the reactor, radioactivity decay forfeits this
protection, Since most spent fuel reprocessed in this century will
have cooled for even longer than that, the Civex process cannot
contribute to the solution of the problems we must worry about now,.
Civex also must face other problems, Industry will resist
irradiated fuels becafise ther increase occupational and public
health exposures, Likewise the owners of the material, especially
foreign countries, are not likely to accept the dictates of the
reprocessor on the form in which plutonium is to be returned to
them, This would also argue against indusitrial recycle of solide

233

form “°°U daughters).

The simplest answer to denaturing the BG export from an MCFR
will be to take it directly from the core; fission products, Y -
emitters.and all.

2024346 Summarye It remains yet to say if there exist any signif-
icant proliferation differences among any of the fuels, Ve can
categorize them as follows:

1e U (low-enrichment in 230y or 2330)

2a Pu
3, 2%y

41
 

2334

4.

Fu, 235U, and 233U are all weapons-grade, [Fuelling reactors
With sub-weapons—grade materials does little to limit proliferation:
the reactor will soon produce weapons—grade fuel from the diluert,

238

i.e, Pu from U. Thus, no reactor fuel cycle can avoid weapons—

grade material, Also, any irradiated 238Uécontaining fuel consti=-

tutes a terrorist threat from TNT-exploSion aerosol-dispersal of

carcinogenic Pue
All but very fresh 233U exhibits a deadly penetrating daughter

radiavicne This hazard discourages the assembly and holding of

, 233

U vomb. Hven if onme chamically separates out the daughters,
they soon build in againe

Thus, in summary,no "proliferation~proof™ fuel exists,but
the Th/233U cycle does offer some advantages over the U/Pu one:
the presence of 232U daughter radiation and the absence of carcin=-
ogenic Pu,.

26244 The Sicnificance of Stockpiles

2e2ele1 Peaceful vs, military explosives, The Nonprolifleration
Treaty forbids non-nuclear-weapon states to acquire nuclear exploe-
sive devices, whether labelled military or peaceful (they blow up
the same way). Peaceful explosive services could still be provided
under sirict international physical control,

The same rules should apply to something very similar, a dis-
assembled bomb, That being so, what about the nuclear explosive
material itself? The only answer consistent with the Treaty's

prohibition on nuclear explosive devices is = don't label it "peace=-

42
ful" until you prove you can safeguard it [65].

2.2.4.2 What about unsafeguarded production reactors? [65]s It is
contended for example, that no country choosing to build nuclear
weapons would turn to its civilian power reactors for the requisite
explosive materials; to divert material in this way would risk detec-
tion by the IAEA inspectors, and in addition would provide too poor
a grade of plutonium to interest weaponeers, Under this self-serv-
ing theory, if weapons material is wanted, a special~purpose unsafe=
guarded reactor would be built, I% is possible at the moment to do
this legally in countries noi pariy to the Nonproliferation Treaty
and therefore not subject to inspection of all its indigenous nuclear
facilities. This underlines fthe need to exterd the req;iggments of
the freaty to nmonsignatory nations by conditioning nficlear trade on
acceptance of internafional agreemenfig and inspection on all nuclear
activities within importing countries.

There is increasing pressure to do this, and a recent bill be-
fore the Congress would make this a condition for U.S. nuclear ex~
portse

Even if legal, however, the construcfion of a special purpose
plutonium production reactor signals a country's intention to build
bombs and, in the present climate, risks premeture interception of
its attempt to obtain explosive materizl for nuclear weaponse.

To avoid interference, a country mighi stockpile separated

 

plutonium from spent power plant fuel openly and legally. A dew-
fense establishment can design and fabricate a bomb in privacy;

the illegel activity is then confined to a swift, almost one-

43
 

sten process: appropriation from its storage place of the nec-
essary plutonium, fabrication, and insertion into the waiting bomb,
It is surely the quickest, cheapest, and least risky route to nuc;
lear weapons. So long as individual nations are permitted to kggg
nuclear explosive stockpiles they are in effect, in possession of
an optiorn to make nuclear weapons almost literally overnight,

In other words, from the moment spent reactor fuel is trans-
lated into sevarated plutonium and stored, the element of "timely"
warning, on which our present safeguards system has been relying,
evaporates, The same is true, of course, for stockpiles of highly
enriched uranium,

It is important to understand that so far as safeguards are
concerned a stock of nuclear erxplosive material is ailot more like
a2 bomb than it is like a reactors No one would dream of suggest-
ing that nuclear explosive devices, regardiess of how labelled,
should be exported under international safeguards, The Nonprolif=
eration Treaty settled once and for all the notion that nuclear
explosives came in two categories - pilitary and peaceful, Under
the treaty no such distinction is permitted. Yet strip away the
electronics and the conventional high explosives and label the
plutonium as intended for peaceful purposes and many nmuclear spokes-
man, at home and abroad, will tell you that if subject to occasional
inspections it is a perfectly safe proposition: just like safe=
guarding power reactors [65]e

The worldwide opportunities for diversion are clearly dominated

by the existence of large volumes of spent fuel stored in many hun-
dreds of locations, and especially long = cooled spent fuel which is
not returned to supplier countries [10],

The crux of the proliferation issue is not to completeliy avoid
Pus. Due to advanced technology, any country can produce it without
resoriing to nuclear power cycles, Rather it is to aveid having
Pu lie around in larze quantities where diversion is less detectable,
In an MCFR(Pu), Pu is always in solution, except for that fuel which
is required to start a new reactor. Reprocessing is done within
the plant boundaries continously and the Pu then fed back into the

system,

45
 

24245 Fuel Cycle Vulnerability to Diversion

The opportunity for national or subnational diversion of
weapons—grade material will depend on the size of the fuel cycle
inventorye Transport outside secure boundaries and decentraliz=
ation of the inventory enhances the potential for subnational
diversion: while under transport the material will be in a more
easy-to-handle form and probably under less security; decentrali-
zation also implies less security.

2.2451 Once=throush fuel cycle. Figure 2,2-1 diagrams the trail
of weapons-grade material in a once-through fuel cycle, assuming
that the initial fuel elements are low=enrichment U. After gener-
ation in the reactor, it passes through about three storage/queue
depotss Possibly one more storage depot labelled 'permanent" may
appear later, once it is fixed what that will be, Only one out-of
plant transfer may be necessarye The inventory in the long term
storage depots will grow with time, eventually becoming very large.
Although the Pu there is not separated out, a national laboratory
could easily do so,

One should also note the presence of the enrichment plant:
though it doesn't produce weapons grade-material for this fuel cycle,

it could,

46
Operations Storage

JE A

 

 

U enrichment
plant

|

Reactor

 

 

 

 

 

Figure 2,2=1
Once~through Fuel Cycle

Legend

e e e e o e =2

- = = = intraplant transfer

interplant transfer

 

 

 

224542 Solid fuel reprocessing cycle. Im solid-fuel recycle (Fig,
2.2-2) we note that weapons-grade fuel will accumulate in three plants
and maybe six storage/queue depots. Four interplant transfer operations
occur, counting BG export, It is not hard to envision the equivalent

of ten cores being tied up in the various queues,. plantis,. storags,.

and transports,

2e2+5.3 Molten-salt fuel cycle, Here almosi all weapons-grade
material stays (Fig. 242-3) within the plant boundaries, With out-
of-core cooling, the equivaleni of 4-~5 cores inventory is possible,

but they remain within the walls and mostly in the primary circuit,

Only BG export requires out-ofeplant transport.

47
 

Operations

Reactor ,

Storage

 

 

  
   

  

Piping and
Heat Exchancger

 

 

Reprocessing

 

 

Stockpile of

233U BG for

Other Plant
Startup

 

 

 

 

 

gggort of

UC].3

- = = - intra plant transfer

======= interplant transfer

Fige 242=3 Molten=Salt Fuel Cycle

49
 

232.5.4 To reprocess or not 1o reprocess, The once-~through fuel
cycle, starting with low~enrichment U, seems to display fairly low
vulnerability to subnational diversion, but creates a cumulating
stockpile available for national use, It also creates a long~term
waste management problem and poorly uses our resources,

The molien. salt fuel cycle clearly seems to be less vulnerable
than the solid-fuel reprocessing cycle. On the present rough scale,
the molten salt fuel cycle must be rzted as comparable to the
"once=through" in vulnerability,

The question now occurs = would you export an MSR with its
in-plant fuel cycle capability to a non-weapons state? Or, a
different question: which would you export, the "once-through"
or the "molten" fuel cycle? Assuming that you could alter the MCFR
to make Pu from 238U but that also the non-weapons state keeps
its spent fuel from a2 once-through cycle (see section 2.2,2,3),
either cycle is bad for national proliferation, This seems 1o
say forget about exporting unless you can maintain control: IAEA,
US, or guuer,

On the domestic scene, the reduced market would definitely
favor the "molten" cycle as it solves more problems and makes more

fuel economically available,

50
 

2e2e5¢5 To breed or not to breed, Assuming that the potential
1o breed excess plutonium or highly=-enriched 233U exists, one
must next ask - should we? The alternative is fto use the excess
neutrons to achieve higher fission product levels, a smaller
critical mass, or some other benefit,

Breeding obviously provides initial fuel for starting up
other reactors, The alternative is to start [CFRs up with Pu
from spent LWR fuel, or highly-enriched 235U. But either of
the latter two fuels would make it easier to manufacture weapons,
Also they would cause the thorium fuel cycle in this reactor to
depend on -=he U/Pu fuel cycls, With a system of thermal MSRs

the export 233U BG could be dilubed below weapons grade, but

238U diluent fiould evene

Th diluent is chemically separable and
tuzlly mean Pu production,

Based on these considerations, this study seeks a breeding
MCFR. Only the first genmeration of MCFR(Th) reactors should then
nge@ ?35U or Pu for startun.
2e2e5¢5 Fusion reactor vulnerability. Lest anyone think that
fusion reactors are the panacea, note that the current literature is
full of fissile breeding ideas; producing weaponsw-grade 233U or Pu,
In the US, current thinking regards such a hybrid as neeessary to a

near-term economical reactor, In the USSR it is their prime goal,

51
2.3 el Utilization
2.3.1 Barth's Resources

209 (Z = 83) nature offers on earth only the chemical

Beyond
elements thorium and uranium (Z = 90 and 92). Only these two
materials, in present, or feasibly-transmuted form, can sustain
a neutron=induced fission chain reaction.

Only 0.72% of U is 235U. The enrichment process loses some of
that. Present reactors, without recycle, consume just 10% of the
235U before assizning the rest to semipermanent siorage, Thus we
presently waste 9Ge 94% of the uranium mined, an affront io the en-
vironment. If, instead, the reactor could usefully consume ail of
the actinide, considerably less eyesore and mining expense would
result. Bconomics might then justify the retrieval of extensive
low=grade ores. At a time of energy shortage the present policy
seems foolhardy except that 1t reflects the restrictions posed by
economics, proliferation concern, and spent fuel waste management.
These must first be overcome.

Thorium abounds three times as much as uranium in the earth's
crust, and offers that much more nuclear fuel energy. The avail-
abie uranium ores increase greatly if one includes not only the
classical ores { »1000. ppm uranium) but the very abundant granites
with 80 ppm uranium and thorium, GCranite is the main constituent
of the earth's crust (up to 20 km deep). Extraction from these
ores could become economical if fission reactors fully used all the
Th and U present. Together with fusion reactors, such a fuel cycle

would then open up all the principal world resources of nuclear fuel.

52
2¢342 Use of Thorium

Nature has significantly inhibited the use of Th by not pro-
viding a thorium isotope which can sustain a neuiron chain reaction,
Despite some unique advantages of a Th/233U cycle over a U/Pu one;
Th use must still prove its economicse A reactor which yields high
BG in the Th cycle would help a lot.

High melting points hinder reprocessing of 2 solid Th fuel:
Tho, melts at 3050°C; Th metal, at 1750°C. Breeding is very mar-
ginal, except with a fast reactor. Reducing non-productive neutron
captures (ee.ge to fission products, core structural materials, and
control poisons) vecomes crucial, That suggests fluid fuel reactors:
they can continuously remove fission products while adding fuel on-
ling, and some variants require no core structure.

2e¢3¢3 Intercomparison of Reactor
Concepts with Regard to Fuel Utilization

Much attention currently centers on optimum uranium usage by
candidate reactor concepts, The parameter of Megawatt days of
electric energy produced per mined metriq ton of uranium (MWde/
mined HTU) provides a suitable yardsticke Note that the numerator
includes energy production from thorium while the denominator ignores
thorium consumption, Consider as a benchmark that all the uranium
were somehow (ideally) consumed, Then without any energy production
from Th, the rating would be 380,000 MWide/mined MTU, The advanced
concepts in Table 2,3-I, currently under study in the US, do not
come anywhere near this idealistic figure; not even when including

energy production from The

23
Table 2¢3=]

SOME 30-YEAR FUEL CYCLE PERFORMANCE COMPARISONS [74]

 

Energy Production
Reactor Fuel Cycle MWde/Mined MTU

Uranium Cxcles:

once~through uranium 1860

uranium recycle only, 2270
Pu storage .

uranium and single Pu 2650
pags recycle

uranium and Pu recycle 2800

Thorium Cycles:
93 w/o 235U, recycle of 3420
235y ana 233y

PURs 3 w/o 237U, recycle of 2900
Pu and 233U
3 w/o 235U, crossed 2870

progeny Trecycle

Spectrum Shift (Thorium):

 

93 w/o 235U, recycle of 5550
235U ard 233U
235
3 w/o U23§ecycle of 3450
Pu and U
LWBR Thorium Cycle 6800
CANDU Uranium Cycle 5800
Thorium Cycle 10440
HTGR Thorium Cycle:
once-through 2620
233U recycle 4240

 

¥ 3w/o 235U as U0, core; feed of output Pu into ThO2 with recycle

of Pu; feed of 33y %ggo uo, (natural) with recycle of Pu into
ThO, and recycle of U into U0,

>4
Zebroski [10] concludes that the evolutionary extension of
LWR burnup offers the best near-ferm hope for improving usage,
However, higher burnup will also produce greater statistical vari-
ations in fuel failure, which will limit these extensions, On the
long term, an MCFR should come much closer or exceed the ideal
limit because all actinides are consumed and it requires only thor=
ium feed (except for startup fuel on early reactors),

2e3e4 Use of Both Th and U Reserves

One might desire that any new reactor concept be able to
operate on both the Th and U cycle, so as to evenitually allow use
of all our reserves, MCFR siudies have already indicated large
potential breeding gaint on the U/Pu cycle, providing Pu is accept-
able, In that event, one could also supply spent U fuel as blanket

feed, thereby solving that waste management problem,

95
 

2.4 Strategic Security

Although it doesn't normally get public attention, still

strategic security has its place on the 1list of contemporary concerns

due to current events:

Te

2e

3e

de

It

a fight

The lonz-and short-term goals of the USSR
Other political conflicts which could lead to world war
(viz: Israel-Arab, China vs,..Taiwan or USSR, Arab-Arsd, etc.)
The growing=-pains and frustration or have-nots and emerging
nations
The increasing pelitical turmoil and unrest within America
stemming from problems of cities, alienation of extremist
groups, and, again, have-nots,

might be successfully argued that fat men dofi't look for

- "yon Cassius hath a lean and hungry look", Then insecuritiy

traces to the failure~to~appear of cheap nuclear electricity, coupled

with the disappearance of the cheap oil~fired energy source, Thus the

world still hungers for a cheap abundani energy source %o solve its

WQES,

obvious

Wetd like to think the MCFR is it, Until that becomes more

though, it would be well {0 secure our US energy supply

irom all threat,
2.4.1 Present Susceptibility

At present, US energy supplies, key to our might and resource-
fullness, are all above board (ground level) for everyone to see
and aim at. This includes dams, oil tankers and depots, fossil-
fuelled plants, and nuclear reactors. Actually only the containe
ment building of the latter protrudes above ground but that still

leaves it vulnerable.

2eliel.l Subnational Blackmaile Should a terrorist group seek a
pmeans of national blackmail, the threat of destruction to a dam
or a nuclear plant might seem credible.

Given sufficient explosire a dam would seem to be a better
target, both in size and certainty cf wreaking financial havoc.
L¥R destruction would likely cause more perscnal injury. A fossil =
fuelled plant destruction would not necessarily threaten public
lives., Bombing a city or a poisonous gas {like ohlorine) plant

might pose the next largest threats.

2.:.1.2 Sabotage. In time of war or unrest, sabotage to our sys—
tem could rear its ugly head. Surprise attacks might often occur.
In addition to the above targets, one might then add many more ncn-
nuclear possibilities such as poisoning water systems or destroying
any energy supply. For sheer havoc, though, destroying all water,

supply.to LWR cores as well as their containment would do well.

o1
 

2.4.1.3 War. In event of war, sabotage could be one front of the
attack. In addition planes and missiles might be directed at nu-
clear plants in hopes of accomplishing what the aggressor didn't
want to or couldn't send an atomic bomb to do: perhaps an atomic
blast would have destroyed a valuable industrial complex whereas
an LWR core release will simply harm and demoralize a lot of

people.
2.L.2 Potential Remedy with an MSR

The dominant vulnerability of the above energy sources would
seem to be the presence of the potential hazard above or near the
earth's surface. Moving the nazard lower mitigates the danger.
Dams can't go underground. Nuclear power plants can, but it costs
a lot. An MSR uniquely offers

(1) Continucus removal of the principal hazard from the sys-

tem: +the gaseous and volatile fission products. One can
store these far underground in a recessed tank for security.

(2) Ability to transfer the whole core further underground

at a moment's notice to a (possibly natural convection-)
cooled holding tank.
Thus the MSR may allow the security advantage of underground siting

without all the costs and other difficulties.

58
 

2elie Summar

Analysis suggests that an LWR may be a likely target for atiack
or threat of attack. The goals of such an attack would be

(1) Destruction of the energy production capability

(2) Release of gasecus and volatile fission products into

the air.

This threai to energy and peopie probably exceeds that from a
dam which tends to be located farther from people. Also water must
follow the lay of the land, which authorities could observe and some -
what predict. Transport of fission product gases, they could not very
well .

At the same time, nuclear plants can be the toughest nuts to crack
as they are more compact and self-contained. With a little more hard-
ening, such as an MSR allows, they could be America's energy ace-in-
the hole and basis for survival.

On the long term the MCFR and all other novel concepts should be
explored and re-explored for their ability to provide that panacea of
a cheap abundant energy source = a far betier means of dealing with

yon hungry Cassius than any protective shells one can devise.

59
2.5 Summary Design Principals for an Advanced System
2e9e1 Waste Management Restraints

Spent fuel contairs several products whose activity remains
high after a years storage, Of these, the alpha~emitters can extenw
sively damage intermals when ingested, Plutonium presents a special
hazard because it locates on bones: the whole skeleton of man be-
comes the critical organe. Although our society has learned how to
handle other toxic industrial wastes (arsemic, mercury, chlorine),
control is not perfect and accidents have occurred. Thus, we prefer
to avoid plutonium production if that does not severely curiail
our energy production, A reactor whicn usefully consumes all <he
actinides and has nil Pu in its system would seem very atiractive,

2652 Mon-proliferation Restrainis

Despite earlier misconceptions plutonium of any grade can be
put in a form suitable for the manufacture of nuclear explosive
devices in a matter of days to weeks, Thus, from the moment a
non-weapons state separates plutonium from spent reactor fuel and
stores it, the element of "{imely warning", on which our present
sateguards system relies, vanishes,

The same holds true of course for stockpiles of highly enriched
(in 2337 or 25y) uranium, Thus 259U or 257U mist be diluted with
238U to avoid classification as weapons=-grade, However, subsequent
reactor irradiation again produces plutonium which separates easily
by chemistry. Thus no fuel cycle can avoid producing weapons—grade
fuel, Of course one can minimize the proliferation hazard by using

extended LYR cycles and not reprocessing.

60
 

Both of the last two Presidential Administrations have suggested this
fall-back course, However, to get any significant improvement in
fuel utilization, reprocessing is necessary. Then one seeks to
minimize the transpori of weapons-grade and radioactive materials,
Molten-salt frel cycles with in-house reprocessing offer low pro-
liferation hazard comparable to that of the once~throush cycle,

2¢45¢3 Fuel Uftilization Restraints

Kno'm nuclear fuel supplies contain up to 1015Gw-years of
energye. However much of this industry cennot tap because ores are
uneconomic to extract and the current reactor fuel cycle uses less
than onz cie=thousandth of that which 1t does extraci, The absence
of any naturally-occuring "fissile' isotope puts Th at a further
disadvantage.

A reactor conceot which could zvproach full utilizafion of all
actinide fuel mizht stand a chance of providing the economics neces=-
sary to accomplish the extraction of low-grade ores, including both
uranium and thorium for fission reactors. Such a concept should ex-
hibit a high BG to overcome thoriumts fissile disadvantage,

2¢5e¢4 Strategic Securitiy Restraints

Dams and current nuclear power plants provide attractive targets
to internal and external attack, !MSRs offer a hardening fto such attack
by continuous removal of the major radiation hazard from the system and
ability 1o transfer the whole core further underground at a momenf%
notice. However, if an NHCFR provided the break through to universal

cheap electricity, that should be the greatest advance to security.
2.6 Promise and Uniqueness of the Molten Chloride Fast
Reactor on a Thorium Fuel Cycle

The promise and uniqueness of this concept stems from the
molten salt state, the exclusive use of thorium, continuous repro-—
cessing, and tne very fast neutron spectrum,

20641 Advantazes of the Molten State

Having the fuel in a2 fluid state allows out-of-core cooling,
thereby avoiding structural components in fields of significant ra-
diation damage. It also eliminates labor and material costs assoe
ciated with fuel element decladding, dissolution, and fabrication,
Fuel handling by vumps and piping should be less complex than solid
fuel handling. ©“he simplifi:d core should greatly ease the plant
design, and increase its reliability and availabiliiy, thereby de-
creasing cost,.

A fluid state system also facilitates on-siie close~coupled
fuel reprocessing (Section 2,6.2)s Out=of=core cooling of the
molten salt allows the primary circuit to operate at low pressure:
+this reduces the severity of the environment and allows materials
such as graphite for pipinge That should reduce fuel cycle cosis,

Elimination of fuel cladding and structural material signif=-
icantly improves the neutron economy of the reactor: more neutrous
are available for breeding, reduced critical mass, or other tradeoffs,

The molten state allows high power densities (up to 10 M
per liter) and high temperature operation while at low pressure:
fuel melting is part of the design and so is not a problem; fuel

vaporization does not occur until extreme temperatures, contributing

62
to low operating pressures, Only the contciner materials impose
temperature limitse, With o alloys 900°C appears to be an upper
limite Graphite structures should tolerate considerably higher
temperatures, This could significantly advance thermal efficiency
which could attendantly reduce therma]l pollution of the environment,
Hizh temperature operation affords applications with existing

steam technology, or with He secondary coolant for gas turbines

or process heat, High power density also promoies small sizes
which saves plant costs.

The eutectic nature of the hzlide salts facilitates low tem~
perature overation in the near term by minimizing chemically-reducing
corrosion oroblems with Mo-F2 alloys. With graphite, these problems
nay aot existe Later, high temperature pperation leading to higher
efficiency process heat =t compositions away from the eutectic nadir
can be implemented when more is kmown about material corrosion,
Off-eutectic compositions can also mean higher BG and thermal con-
ductivity of salt,

Halide salts also offer superior thermal and radiation stabil-
ity. This inhibits the formation of other compounds, thereby prevent-
ing corrosion,

The molten nature greatly benefits- safety as well: increase in
temperature causes a strong decrease in fuel density. The inherent
stability of this negative temperature coefficient will limit ex=
cursions, A self-regulating system may be possible, avoiding the
need for control elements,.

The fluid state of the fuel also enhances safety under abnor-

63
 

mal condition through a combination of properties:
(1) Pluidity facilitates removal from the reactor to ever-
safe containers
(2) High heat capacity of fuel restricts temperature rise
on loss of normal cocling
(3) Low salt vapor pressure mininizes the effect of any
temperature rise
2e6e2 Adventazes Stemming from the Ixclusive Use of Thorium
The nuclear industry has exclusively used the uranium fuel
cycle to date, mainly because no "fissile" thorium isotope erisis,
However, once a thorium cycle proves viable, numerous adventazes
accrue, Because the burning of thorium begins much lower on the
atomic weight scale than does uranium, little plutonium or higher
actinides accurmilate, The plutonium which does appear will be
mostly 238Pu, a strong heat source which melts in concentrated form

unless cooled, That hinders its use as a weapone

The predominant products in the cycle will be 232Th, 233Pa,

233U, 234U, and 235U. The only significant hazard associated with
these will be gamma radiation from 232U decay daughters, As the
I.CFR concept requires no handling (as in fuel element fabrication)
this will be a2 plus; it prohibitedly hazardizes the constructior of
Weapons,
2+643 Advantzges of On~site Continuous Reprocessing
Reprocessing on~site minimizes the fissile inventory, environ-

mental hazard, and proliferation danger of the whole system: fuel

is not tied up in other plaants or their temporary storage depois,

64
 

It is also not under transport to or from such locations, eliminc~
ting highjaciking, sabotage, and transporiation accidents, Finally,
it never even occurs in a form or container suitable for transport,
except for breeding gain export,

Continuous reprocessing minimizes the fuel inventory cutside
the reactor: +the iaventory in the core, piping, and heat exchangers
reoresents almost the entire fuel cycle inventory., Continuous re-
processing also removes hazardous and neutron absorbing fission prode
ucte while adding fresh shim fuel from the blanket, This greatly
reduces the potential radioleogiczl danger of the reactor while in~
creasing ihe neutron economy. Processing on-line allows on-=line
refueliing, That reduces domtime and obviates mechanical shim
devices,; increasing the neuiron economy,

24644 Advantages of a Very Fast Neutron Spectrum

An [ICFR exhibits a very fast neutron energy spectrum: chlorine
(atomic weight A%36) comnstitutes the lightest major element; the
salt contains only small amounts of Na (4=23) or K (A=39), Thus
no strong elastic scatferers pervade to moderate the neutron's energy.

Choosing 2 high molar content of UCl, in the core salt further

3
inhibits the neutron moderation. The principal moderating mechanism
then defaults to heavy element inelastic scatter, Feeding only
233UCl3 (~ no ThCl4) into the core reduces even that relative to
fission,

Table 241=II compared the median flux energy of the resulting

neutron spectrum to that for other fast reactors, Since the neutron

capture cross section decreases wiih energy and many isotopes ex=

65
 

hibit threshold behavior for fission (Figure 2,1-5), the hard spectrum
enhances fission over capture, The result is that
1. Reactivity per unii mass increases, decreasing the critical
size
2e More actinides now fission usefully rather than just trans-

mate into a higher-A actinide

o 233 232

3. Th U (n,2n) cross section for producing “~ U, 2 pro-
liferaiion detferrent, increases

4, Parasitic neuiron capture by fission products and structural
materials decreases, thereby improving the neutron economy,
and decreasing the sensitivity of those meterials

233

5« Boih Pz and 234U tend more to fissione This eliminates
the need to remove and hold 233Pa for decay into 233U.

In a plutonium=fueled MCFR, the high actinide density,.the
absence of core internals, and the very fast neutron spectrum can
combine to raise BG up to Oe7e A 233U/Th system could probably
only achieve BG of 042 10 045+ Should one prefer a BG near zero,
the good neutron economy can be diverted to other advantages such
as smaller blankets, slower fission product cleanup, added Th in
the core mix to reduce pover density, cooling in-core, or operation
a2t (lower) eutectic nadir operating temperatures,

The fast neutron spectrum also implies low fission cross sec-
tions relative to 2 thermal neutron spectrums To accomplish the
same pover density as in a thermal system, the flux levels must

16 -2 -1

exceed 10 necn sec o The lack of cvore intermals limits ra=

diation damage problems in the out=of-core cooled MCFR. On the plus

66
 

side, such flux levels might benefit MCFR variants such as high flux

[75] reactors for molecular studies and radio-medicine prouuction [76]

and flux-irap burner reactors for troublesome fission product trans-

mutation [23].

24645 Spinoff Ability fo Dizest Existing Spent Fuels

The MCFR should be able to usefully consume 2ll existing spent

fuels,

In principle one could bring in spent LWR oxide fuel, declad

it (radioactivity makes this the most cumbersome step), coavert it

to chloride just like 'h and U (natural or depleied) oxides, and

breed it in the blanket region. That would, of course, mean pro-

ducing Pu., One might not wznt to export that capability and mey

need to ensure that any expcrtable MCFR designed for the Th cycle

cannot be easily so modified,

24546 Inherent Disadvantages and Limitations to MCFRs

Expected problem areas and concerns for an MCFR will be that

Te

2e

3e

be

Se

Qut-of-core cooling and local reprocessing produce high

fuel inveniory in-plant; on the other hand little inven-
tory exists out-of=plant or in tramspori

Hlolten salt fuel transfers heat poorly compared with so=-
dium in an L#FBR

The high melting voint (.~ 56000) of suitable fuel salts
necessitates pre-~heating in many places

The high melting point of the fuel salt limits the A1
across a heat exchanger, less the salt freeze, Conseguently
one mast increase the mass flow rate

The presence of fission products in the fuel salt necessitaies

67
Te

9

2 high standard of plant reliability and leak fightness
Corrosion and high temperature limit the choice for stiruc-
tural meterials

High neutron flux damages sitructural materials

The reprocessing plant requires development

Mutants in the fuel salt, including sulfur from chlorine

mutation, corrodee-

68
3.0 CONCEPT DESIGN

MCFR(Th) neutronics will depend on the choice of reactor con-
figuration, method of cooling, power density, operating temperatures,
salt compositions, structural materials, and method of shim and
safety conirol.s Sectiorn 2 provides directives for choosing: use-
fully consume the actinides, use thorium as feed fuel, and maximize
breedinse These all coincide with a supra-guide~line: keep the
core neutron spectrum as hard as possible, An intrinsic guideline
is technical and economical feasibility,

3.1 Choosing the Reactor Configuration

Subsections 1.1 and 3e1.2 analyze how to mazimize breeding
by minimizing neutron moderation and leakage, Following subsections
apply that knowledge to evaluate spherical vs, cylindrical geometry,
one-zone reactor vs. core and blanket, other blanket aspects, re-
flector, damage shield, and method of shimming,
3.1e1 Maximizing Breeding by Minimizing Neutron lloderation in the Core
3e1s1e1 BC potentizl. Conversion of fertile to fissile fuel depends
upon capturing the excess neutrons in the chain reactione. Each time
a 233U atom absorbs a neutron, the number of neutrons potentially
available for breeding is [77]
V3-1-ols + Fp (9~ 1)-A-L (3-1)

1+,

Potential BR=

where

B3R

breeding ratio

233

'%3 = noe of fiszion neutrons from U fission

23

o 2 s
-fia=:no. of fission neutrons from Th fission

59
o, = noe of neutron captures in 233U per 233

U fission
A = no. of neutron captures in structural materizls per
233U fission
L = no. of neutrons which leak from the reactor per 233 U
fission

2 ..
F2 = noe Of 32’1‘11 fissions per 2?‘)BU fission

Rearranginz terms one gets

mefo o -oem i) 0

L
where 3 =7-1

Yow in the neutron ener ;y range of fast rezctors the ctositive
contributing factors 31 Fz, and ~ 2 increase with _r;.eutron energy
while 0(3 and A decrease, L varies little with neutron energy,
3y ,1) 5 and &X 3 also vary slowly. F2 and A vary pronouncedly but are
much smaller. Thus BG potential (i.e; available neutrans) increases
with spectral hardness but not dramatically so. For this and other
reascns of secticn 2., ane desires as hard a spectrum as possible,
but not at great expense of scme other desirable parameter such as
actual fertile captures.

For a rough estimate of the BG potential we assume A=0 (no

structural materials) and L=0 (zero neutron leakage). As 23 2Th

 

has a high threshold for fissicn, it is also reasonable to let F2=0.
YWe then have
BR= Va- (I+ch) = '[3_ 1 (3-3)
) ety
where Yl = L
LoV

70
2
7l varies strongly with neutron energy, more sSo for 39Pu than

223 232

U {Figure 3.1-1}. F, 1s greater for 23%) than for *3°1h.

239 233, e s n
Consequently, Pu and U differ little in BG at 1CC keV; but

by near C.4, at 1 Mev.

In studies of an MCFR/Pu), Taube calcuiated maximum BGs of

0.7-0.8. 1his agrees with Figure 3.1-1 and equaticn (3-1), assuming

. . - 238 .. . . .
scme ceniributicn (F) frem ~77U fissicn. {n the basis cf Figure

- - ~ T | o - 2 T
3.1-1 and To% C, we guess a BC potential fecr Th/ 33u near C.L.

Secticn 3.5.7 confirms this.

 

3
Figure 3.1-1 2
Value of Eta (vL) for
Fissile Nuclides

1

0

 

 

P — =

10‘2 100 10-1- 104 105 108

Neutron Energy (eV)

T
 

3.1.1.2 Minimizing ileutron Ijoderaticn. This translates to mini-_'
mizing neutron scatter, i.e. avoiding structural materizl in the
reactor and admixture to the fissile fuel, With separate core

and blanket zones, the core fuel shouwid ccntain as high a frac -

tion of 233U atoms as possible for spectrum hardness. It also im-
plies keeping the core and blanket separate: mixing fertile and
fissile fuel in a single reactor zone such as in the MSBR mocder-
ates the scectrum.
3.1.2 laximlzirg Breeding by Minimizing Leakage

Sectian 3.1.1.1 sougni the maximum breeding possible when nc
reuircns 2scapad the fertilzs blanicet (L=0) and when A=0. Fractizat
aids ftowaris low T might be aa inner blanket, 2 thick outer blanket,

¢

and a gcod neutrcn reflector. Reactor ccnfiguratioh affects A thrcugh
the addition of structural materials.
3.1.2.1 Efrcct of an inner blanket. With an inner blanket (Figure
3.1-2), neutrons enter as if from multipie cores. Those which do
not react are rot yet lost from the reactor: +they reenter the ccre

and cuter blanket where they can further react or turn arocund.

Figure 3J.1-2
Three-zone Spherical

Reactor

 

12
3.1.2.2 Size of the cuter blanket. Irregardless cf an inner blanket,

useful capture of neutrcns (breeding) requires an outer blanket.
ANL studied [21] 1.5 to 4 feet (0.46-1.22 m) thicknesses;

English and Swiss analysts typically assumed 1.0 m. The present
work (Ssction 3.5.1) found 2 meters to be effectively infinite to BG
and very effective in reducing damage to the vessel wall.

2,1.2.5 Mecerating the plarket neuircn spa2cirum LC Snrance capiure.
Ir a fast reactcr,neutrcn cepture decr2ases witn inereasing neuiran
energy (Fig 3.1-3). Thus tc increase fertile capturewe need tc sof-
ter, the neutrcn spectrum in the bianket withcout seriously degrading

i+ in the ccre.

 

E- 1
E |

100 g [Tn-232] g
z i

o, (b) 10 T %
1 F
o §

01 o l,- : e S | il 3

O 19° 10> 2% 197

)
Neutron Enerzy (eV)

Fig. 3s 1=3 -Thorium Capture Cross Sectlon

ORNL studied the effect of a graphite moderator.in the
blanket [53]. ‘Although the fertile capture rate did increse,
reducing the necessary blanket thickness, the overall BG

decreased (Fig. 3.1-~4) due to simultaneous increase in parasitic

~ neutron capture.

Another way of moderating the blanket is tc use a (lighter)

73
Fig. 3.1<4
Influence of a
- BG
Graphite Moderator

in the Zlaenxet upcen

£a [13]

 

Thickness of Grapnitie !oderator Section, cnm

3.1.2.L rsuiren raflocter amd domase-snicid. A neulron rellacter
cutside the reactor vessel will cbviously reduce L. Few of the re-
flected neutrcens will make it through the outer blanket to return
to the core. A thermalizing reflectcr would therefore have little
effect upcn the ccre spectrum, particularly if the cuter blanket
regicn is fairly thick (in mean free paths).

Such a reflector weuid enhance neutren capiurs in the blanket
(put in carrier sait as well as in fartile). It would alsc decrease
Py scmewnat. The net changs in ZG wculd depend cr ncw the ccempeting
capture cross sections change as neutron energy dscreases. This 1is

expressed by the crcss secticn ratio

Q. a O (£) + b Opy (B)
6 (€)

 

14
 

where Ha = nhalide: c¢hlorine or fluorine

Alk alkali or alkaline earth

asb

coefficients dependent upon the molar ratios,

For a typical 65/35 ThClA/NaCl mix, Table 3,1-I shows the results,
At very fast and thermal neuiron emergies Cl parasitic capiure vre-
dominates., In the broad intermediate range of 20 ev to 1 lieV Th fer-

tile capture is greater, Thus we ovrefer that neutrons in a ThCl

4

blanket slow down, but not all the way to eV energies, This recommends

an inelastic or a heavy scatterer more than a light elastic one for

the reflector material, This would also recommend a high actinide-to-

carrier salt ratio in the blznket,

Choice of beryllium as reflector material would: enhance the
outer blanket breeding through scattering and (n,2n) reactions.
However, beryllium is toxic, relatively scarce and quite light.
Graphite, in contrast, costs little.

Cu and Ni are well known for their ability to reflect fast
neutrons, Fe and Pb reflect almost as well and cost less. All
four qualify as inelastic or heavy=elastic acatters. However,
reflector studies in Section 3.5.6.3 indicate that with a
blanket thickness of 200 cm, the choice of reflector material
matters little. Then the low cost and low activation of graphite
should predominate..

Borrowing a page from fusion test reactor design, one could
reduce damsge to the vessel with an inper liner. Neutrons

striking it change direction and lose energy; they are then less

15
 

Table 3.1 Ratio of Cl to Th Neutron Abscorption in ThClA

 

 

 

O abs (b)

Group Energy Range Cl Th gngTh
1 6.5-10.5 MeV 0,215 0.012 7.6

2 L.O=6.5 0. 146 0.02 33.1

3 24 5mly 0.057 0.04 6,

L lel=2.5 0.020 0,08 1.13
5 OuB=1.4 0.0063 0.14 0e20
6 Oulp=Ge8 0.0021 0.17 T.056
7 0e2=0.4 0.0009 0.19 0.022
8 0. 1~0.2 0.0011 0.27 0,018
9 L6.5-100 keV 0,0031 0,42 0.034
10 21.,5=46.5 0.0103 Ov56 0.084
11 10.0=-21.5 0.0109 Q.75 0,066
12 L e65=10 0.0189 1.35 0.064
13 2.15=1.65 0.0194 2,10 0.042
14 1.0=2.15 0.0221 3.30 0.030
15 L65=1000 eV  0.1670 5,0 0.152
16 215-465 0.3730 11. 0.154
17 100=-215 0.1557 19. 0.037
18 16.45-100 0.3660 28. 0.059
19 21.515.5 0.7081 L47. 0.068
20 10.0=21.5 1.1625 12. 04,0
21 Le65=10 | 1.795 Q.46 1747
22 2.15=4.65 2.771 0.67 18.8
23 1,0-2,15 L4231 0.99 19.4
21 0.465=1,0 6.27L  Luhb 19.6
25 04215=0.465 9.33 2.11 20.1
26 0.0252 eV 28.69 7.56 17 .2

likely to damage the reactor vessel.

This damage shield should

be cheap, suggesting graphite, and might be in partitions for

easy replacement. One could also combine damage shield, vessel,

and reflector into one thick-walled graphite vessel.

cost the least to make or replace.

would also be the easiest to dispose of.

16

This might

Due to low activation it
 

3.1.3 Core and Blanket Design

3.1.3.,1 Choice of geometry. Spherical geometry offers
1. Minimum neutron leakage in a critical reactor (thereby
maximum BG)
2. Minimum chielding and plant size
3., Minimum critical mass

L. Simule theoretical computation

The first two factors directly affect reactor design. Minimum
critical mass actually matters littie: the fuel utilization
issue pertains more to total cycle inventory while proliferation
deais with the availability of attractive weapons~grade or car-
cinogenic material., The fourth factor affects just the methods
used for survey calculations.

Three main designs come to mind in spherical gecmetry. The
first copies the thermal MSBR: . core and blanket materials, i.e.
fissile and fertile, flowing together homogenecusly. The second
copies the usual breeder reactor: a minimumesize core surrounded
by blanket. The third adds an inner blanket (Figure 3.1-2).

Inherent disadvantages with spherical containers may be

1. Difficult fabrication of structural materials

2. Difficult replacement of structurai materials

3. Stresses due to material curvature necessitating a

stronger, thicker core/blanket interface

L. Non=-uniform flow velocities in the core, possibly

creating additional stress.

17
 

Cylindrical geometry offers choices similar to Fig. 3.1=2:
BG in (c) of Figure 3.1-6 may be slightly higher than in (b)
when the core fuel is all fissile and no fertile; BG in (a) will
probably be lower due to spectrum degradation. Likewise, re-
placement of the annular core (c¢) by near—equivalent-area tubes
(d) may further increase BG. Either each tube approaches
criticality or they approach each other in distance close

enough to avoid spectrum degradation.

‘|||I|" ‘lllll" "I%!!l'
o0

Figure 3.1-6 Alternate Cyllndrlcal Geometries (Top View)

Design (d) also suggests quick and cheap replacement, a
simple solution to the problems of radiation damage and corrosion
in the reactor. Structural tubes obviously produce no fission
products; if made of graphite, they also radiate no gammas. That
should facilitate their handiing and disposal. Graphite also
costs little., A design with straight tubes would limit re-

placement downtime. Design (d) is also easier to make and replace

than (e).

3.1.3.2 Choice of number, size, and spacing of tubes. Pro-
ceeding with the Figure 3.1=6 (d) concept, Figure 3.1-7 defines
some annular arrays of N tubes. The dotted-line c¢ircle passing

through the tube locates their centers. To minimize core spectrum

18
 

degradation the circle might be several mean free paths in diameter
so that most neutrons leaving the side don't reach ftubes on the
other side, Alternatively tubes might abut one a2nother so that most
neutrons leaving a tube either go permanently into the blanket or

directly into another core tube, To minimize structural material, N

should be small,

o o oC oo
o 0
o) (22) (8.8) (95) (8.2
o o c o Cn© Op 0o’
N=3 Nc4 H=8& TEYA N=7T Nad

Figure 3.1=-7, Annular array of H-tubes

A questior certain to arise over core tubes is the interdeven~
dence of their neutronics ani kineticse. One prefers:ejither close or
near-zero coupling between them: intermediate loose-coupling could
lead to oscillatiions, not quickly damped. 3Behavior as independent
reactors (zero-coupling) would probably require multiple control
panels and operators, i.c. separate full-power cores sharing the sare
blanket, This option might warrant attention as the tubes would share
the same reprocessing plant and other facilities as well without
affecting each other's operation, Since the reactor physics of each
reactor would simply copy that of a conventional_core and blanket
reactor, this study will not ireat it separately. Close neutronic
coupling implies physical closeness as well (separated by less than
one mean free path) and individually far subcritical, Then, shui-

down of one tube channel would shutdown the whole plant,.

9
3.1.3.3 Location of tubes. To gain more insight into tube
location, consider again the arnular core geometry of Figure
3.1-6(c). Fixing the reactor radius, Figure 3.1-8 depicts
designs with inner blanket (IB) diameter ranging from no IB to

no cuter blanket.

 

Fig. 3.1-8 Designs with Varying IB Diameters

Pertinent metrics include breeding gain, medium flux energy in
each region, and critical mass. Section 3.56e1 shows that the
largest breeding gain occurs with no inner blanket. This would
corfespond.to placing the tubes adjacent to one another in the

reactor center.

3.1.3.4 Axial blanket and neutron leakage, Neutrons which leak
from the reactor cannot breed. They also require shielding.
Lengthening the core helps but an axial blanket stops axial
ieakage best. Figure 3.1-9 suggests some designs which accom—
plish that by'dafipening the chain reaction in the tubes near top
and bottom of the reactor vessel.

The overriding concern in choosing between these designs

80
 

 

\ ettty S
Vertically Horizontally - -
Skewed 4/ Skewed

 

 

 

 

 

 

 

 

Ballooning Side~Entering
and —--

- Returning

— e b
|
}—

 

 

 

 

 

 

End
Diverging

 

 

 

 

Figure 3.1-9. Alternate Designs Which Create an Axial
Blanket in the Reactor Vessel

81
 

should be eame of making and replacing the tubes: otherwise one
must design for corrosion and high neutron radiation exposure.
The two skewed tube designs should excel in this regard because
the tubes are straight. The chain reaction in these would con-
fine itself to prolate or oblute spherical gecmetries. The

critical geometry for the other three designs, wouid more
resemble a cylinder.

The next two designs exhibit large bends. Presumably, one-
would use straight pieces tc the extent possibie and fasten them
to the bends, a cumbersome orocsdure in a radiation environment.
The bailconed tube design presents even more difficulty to make
and replace. -

In conclusicn, the use of a few (N=3 or 4) large skewed tubes
should present the least engineering difficulty in their fab-

rication, operation, and replacement. It will also minimize

neutron spectrum degradation and neutron parasitic capture.

32
3.1.4 Method of Keactivity Shimming

The concentration ratio [233U]/[U + Th + FP] provides a
suitable though only approximate measure of the ratio of neutron
producers to neutron absorbers. As the reactor operatés,
this ratio naturally decreases in the core, thereby losing reac-
tivity. Replacement by a mix of higher ratio restores the
reactivity.

In Figure 3.1-10 the blanket breeds 233uc1h and the re—
processing plant separates it from ThClh and NaCl (Section 3.7).
Addition of 4, Na or Th reduces UClh to UC13:

r

-+

5+ UClL ——» HCLl + UCJ.3
Na + UClA —n  NaCl + UCl3
Th + 40014 —e ThCll+ + 4UCIL3

The separated 23311013 should immediately mix in a shim tank with
highlyeradicactive reprocessed core fuel (for non-proliferation
purposes). As shim mix flows into the core, irradiated fuel

passes out to the FP remover. If necessary, a ThClh/NaCl mixture

could be used to provide negative shim.

 

Figure 3.1-10 |

 

 

 

 

 

 

 

 

 

 

 

 

COFC B\thd
MSFR(Th) Reprocessing and
Reactivity Shimming Scheme
¥ [
Pesitwe
S"l.lfl\ T““K
P 1 i
FP v/Th
Removgr ‘;S;::: ::_r

 

 

 

 

83
 

3ele5 Summary Guidelines on Reactor Configuration

The first priority is to maximize BG. Hardening the ccre
neutron spectrum enhances the BG potential but not necessarily
the achievable BG. Hard spectrum also increases the fission
probability for all the actinides,

Spherical geometry maximizes BG by minimizing neutron leak-
age but presents engineering difficuities. It also limits the
power density: since critical masses are smailer in spherical
geometry, power densities rise. Varying the core length in
cylindrical geometry provides more choice in the critical mass
and power density. The spectrum remains hard.

One can also use multiocle, small cylindrical tubes: some
spectrum softening occurs there, depending upon their proximity.
The optimum configuration may be a small number (e.g. three) of
skewed tubes penetrating a cylindrical tank. The effective
geometry of the neutron distribution will then closely approach

a sphere.

84
 

3«2 Reactor Thermohydraulics
An MCFR will require only low pressures for circulation and repro-
cessing removal., Much of the MSBR technology in thermohydraulics should

apply as well to the MCFR.

3.2.1 Means cf Ccooiing the Molten Salt Fuel

e consider three means of cooling moiten salt fuel (Pig-

ure 3.2-1): cui-of-ccre neat exchange, ih--cCre heat exchange acress

tubes, and in=-ccrae direct ccatact with a secandary coclant.

3.2.1.1 Out—cf-core heat exchenge., Here the molten fuel flcws
cut cf the core through piring to the external heat exchanger,
great.y ircreasing the sysi:m inventory. As 3 resuii, mcst de-
layed neutrons emanate outs.de the core. This affects reacter
control (section 3.6.2.2), and the neutromns act.ivate- the primary
circuit and its envirens. The large fuel inventory might pena-
lize the economics of a =2 —or Pu-based piant; for an MCFR(Th)
the penalty matters less because thorium abounds more, the feed
requires no enrichment cr fabrication, and the plant achieves
near-fuil actinide utiliza.ticn. Large fuel inventory also works
against non-proliferation goals, but not as seriously as if it
occurred out=of-plant in transport, reprocessing, or fuel fab-
rication operations.

The absence of structural material inside the reactor maximizes
the neutron economy and minimizes softening the neutron spectrum. The
high emphasis an these characteristics in Section 2 prompts us to adopt

this method of cooling for this study.

85
™r-

Means

O'L_lt—Of—COI'e
feat exchanger

In-core
direct
contact

 

In=core heat
exchange across
tubes

 

rige 3e42=1 Means of Cooling Molien Salt Fuel

Scheme

 

 

[T 3

 

 

 

 

86
 

Nelsan et al [21] examined two MCFR(Pu) designs with out~
of-core cooling: his main comment was an economical concern for
the large Pu inventory. Lane [ 78] mentioned out-of=core cool~
ing for a high~flux (1016 n sec T cm-z) materials~testing fast
reactor. Winfrith examined at least four out~of-core cooling
variants, They recommended further study with both He and lead
secondary coclants.

carlier Taube studies [19] concentratad on in-core conling.
Taube cautioned that loss of delayed neutrans out-of-core might
hurt reactor control. Later work emphasized out—of-core cool-
ing to achieve high flux in owrner and test reactors and high
3G in power reactorse.

3.2.1.2 In=core heat exchanze across tubes. Taube and Ligou [79]

238401 , blanket salt

flowing through 23,000 Mo=alloy tubes in the core. A pump

analyzed a 2030 MWth MCFR(U/Pu) ccoled by

stirred the fuel in the vessel to a speed of 2m/sec to increase
the heat transfcr. That much Mo meant large neutron absorption
(3.3% of the core total). The core inlet and outlet temperatures
were 750 and 793°C. A main question was the weldability of
the Mo-alloy tubes.

A Winfrith group examined [29] Taube's work.. They found
his breeding gains slightly high due to expected deficiencies
in the Bondarenko cross sections for 23 8U and Pu. Also they
found the physical properties of the salt not to be as good
as what he used. These differences would increase tii€ maxdimum

fuel salt temperature to 1160°C. There one approaches limits of

87
of corrosion and of sirength of ilo and its alloys, Taube has also
suggested tubes of steel-reinforced graphite.

Winfrith earlier studied the use of molten lead cooling but
limited themselves to an 80000 salt tenperatures Within that re-
striction they found poor performance, especially low BG due to
neutron absorption by heat exchange materials,

dith helium cooling, HWinfrith allowed the salt to reach 97000;
the outlet e temperature wes SSOOCg

ANL studied two sodium~cocled MCFR(Pu)s [21].
3e2e1ed In-core direct-contact heat exchange, Here an immiscible
£1uid like molien lead [29], boiling mercury [20], or boiling AlCl,
[20] mixes with the fuel directly in the core. It then separates
out and passes out of the core to a heat exchanger. iThe direct con-
tact of molten fuel with molfen coolant affords very good heat trans-
fer, elimination of coolant tubes (and cladding), and possibile ex-
traction of fission products.

The principal problems occur with mixing and separating the
fuel and coolant, and corrosion. The need for extensive research
to overcome these led to the abandonment of this method of cooling,
3.2.1.4 3Blanket cooling. A Th blanket will generate little heat,
Constant bleedoff of the blanket salt (for reprocessing) and re=-
placement by cold HaCl/ThCl4 may satisfy the heat removal require-

ments,.

88
3,2.2 Power Density in an MSFR

3.2.2.1 Inherently high power density in MSFRs. The absence of struc—
tural materials allows high fissile fuel atom density. Absence of nau-
tron moderating materials, and limited presence of fertile and carrier
salts keeps the spectrum hard. Together, these results produce high v
(=96;/ G ) and small critical dimensions. To then produce the

same power as a thermal reactor requires high power densities.

3.2.2.2 Reallstic range of power densities. No actual experience
with 1SFRs exists. ORNL considered power densities of 5~10 Md/
liter in a study of an MSFR fast flux test facility. Taube's MCFR
(Pu) designs range from 0.2 to 11.1 MW/). British designs had 0.36
Mi/l.
CRNL operated a thermal molten salt reactor up to 80 kfl/ 1.
Peak power densities in existing water-cooled reactors include 1.3, 1.5,
2.5, and 4.4 Mi/1 in m, HFBR, ATR, and HFIR [78], respectively.
A coolant velocity of 120 ft/s might perfit 10 MW per liter of core in
HFIR [’(8]. For the sodium—cooled fast reactors Melekes Cll-2, Phoenix
250, and FFTF, core power densities average 2.5, 0.46, and 1.0 Md/1.
The power per unit volume ccolant for the sodium- and water-cooled
reactors above is a factor of two higher., Thus a peak power density of
10 Mi7/1 for a fast molten salt reactor does not appear infeasible, but
this st.udy‘should not exceed it without good cause and a lower
one would present less engineering difficulty.
Flow velocity and heat exchanger size (i.e. cut—of-core fuel

inventory) may constitute the real operational limits.

89
 

3.2.2.3 High neutron flux levels and radiation damage in fast reactors.

The follewing integral expresses the power density:

Pr) = Xni‘issile Of Cb (B) & (3.2-1)
where

P(r)

the power density at the point indicated by vector r,

N fissile = the atom density of the fissile fuel at r,
c ;= the fission cross section for the fissile atoms at r,
$ = +the energy-dependent neutron flux at r,
dE = +the differential of the neutron energy over which the

integration is to occur.
Because O £ is several magni*udes smaller in a fast neutron spectrum
than in a thermal cne, § must increase that much to get the same power
- - - i - »
density. N fissile I° also higher here, but that leads to small critical
volume and higher power density (Section 3.2.2.1) which makes @ even

higher. Because of high @ (> 1016 n em 2

sec™?), ancther limit to power
density could be radiation damage to structural materials (at thehlan.
ket core interface).

The number of displacements per atom and the helium production rate
gauge radiation damage far intermediate and heavy mass materials. In
the past 0.1 MeV and > 1.C MeV fluence served as somewhat coarser (but
still valid) éa.figes. This emphasiies ~_tha:t' most 211 6f the neutmné in a
fast reactor cause damage wnereas midny don't in a thermal reactor.
Coupled with the higher flux levels here, _tha.t bodes a llml'b to the life
of s.;:ructura.l materials within the core environs. The solution there is
t; perlodically.replacé the skewed tub_es.

Cut near the reactor perimeter the 2-meter molten salt blanket has

greatly reduced the flux in the pressure vessel down to ~ 101:3 n cm-2

g0
: this gives a 30 year fluence of ~v1022 n cm-2 for which much

5 -1
experience already exists, Keeping the vessel exposure low is ime
portant enough to discourage the use of any regular arrays of solid
materizl in the blanket through which neutrons could stream,

3e2e2¢4 Impetus for high temperature operation. Heat transfers from

the core according to

q=m C, AT, (3.2=2)
Here
g = the heat removal rate = the heat generation rate under equi=-
librium conditions
m = the mass flow rate, proportional to the fluid velocity (v),
density, and the flow area
C = the heat capacity of the fluid (energy per unit mass per unit
temperature)
J&Tf = the temperature increase of the fluid from entrance to exit

of the core,

Cp and the density are relatively fixed; only v, the flow area,
and ATf can varye. Raising v would increase pumping power requirements
and corrosione. Increasing the flow area may dictate more ThCl4 or
NaCl to avoid supercriticality: that, in turn would soften the spectrum
and reduce the BG potential,

With regard to increasing Amf, Section 3,3 already requires inlet

temperatures of 55000 and higher, Thus an MSFR inherenily generates

91
high temperature heat. This offers advantages (touted elsewhere) as
well as problems. The problems may include leakage through seal expan-
sion and melting of metal. Also, in numerdus fused-salt systems, high
temperature gradients at equilibrium enhance chemical corrosion [80].
Thus, in short summary, increased heat removal requirements will

mest  likely increase corrosicn and/er soften the spectrum.

3,2.2.5 Adjusting power density through critical mass. Sectian 3.5
sets the minimum critical voiume for an MCFR(Th) at about 60 liters of
0.06 ™. Fer a 10 Gith reactor this would imply a power density of
167 Miyth/l. Even for a 2250 !Wth reactor (1000 Mde) it spells 38 Mdth/l.
The power density limits of szction 3.2.2.2 (10 i&/1) dictate a larger
critical volume.

Cne way to get it is to increase leakage from the core (but not from
the reactor) by reducing the core dimensicn in one or more directions. A
thin cylinder is one logical solution: above a given length-to=diameter
ratio the length can infinitely increase without appreciably changing
keff‘ However, space requirements and return piping do impose practical
limits on the length.

Multiple thin cylinders resolve that dilemma. In regular array
(Figure 3.1-7) they approach an annular core geametry with internal blanket
(IB). As the IB diameter increases (annulus moves radially outward in

Fig. 3.1=-8) the critical volume further increases.

92
 

In the skewed tube configuration the effective geometry (where
the tubes come together) can vary from one of minimal core leakage
when angles between tubes are equal and near perpendicular, to one of
high leakage when ftubes penetrate the vessel top and bottom or from
the vessel sides (Figs 3e1=9),

A second way to increase the critical veolume would be to increase
neutron absorption by allowing a higher equilibrium concentration of
fission products., However, breeding zzin would suffer,

A third way is to increase the carrier salt concentiration (de-
crease the molar percentage of actinide salt), However, this will
sofien the neuiron spectrum and decrease the BG potential, (This
would not decrease the BG proliferation hazard of the material: any
material which can sustain a chain reaction in a fast reactor, can do
so even easier in a bomb configuration,)

A fourih method is to add in fertile materizl, This will soften
the spectrum somewhat but without parasitic capture of neutrons, In-
deed, this option looks very favorable; section 3,3 discusses it

further.

93
3.2.2.6 Reactor design power. For a given power density, adding more
fuel (e.g. lengthening the core) can increase the total reactor power.
This should increase the economics since the capital equipment, includ-
ing that for reprocessing, changes mare slowly. However, two or more
snall reactors at a plant nay excel over ane larger ane: though each
cne requires separate instrumentation and operating crew, while one is
down the other still uses the reprocessing system and produces eleciric
power. Without significantly coupling the cores, the blankets of
sevaral reactors might also abut one another so as to eliminate re-
flector and enhance breeding.

A popuiar size for intercomparisan of designs is 1000 ifve.

3.2.2.7 HReactor vower distribution. Zonewise, fission of bred 233U
should generate more power in the blanket than high-threshold 232Th
fission would. The usual concerns regarding power peaking vanish: the

fuel is already melted, and burnup is homogenous due to liquid mixing.

- 3.2.2.8 Summary. The 2-meter thick blanket protects the vessel from
the high fluxes accompanying high power density. The frequency of
skewed core tubes replacement will still depend on the flux levels, In-
creasing the core volume therefore increases the tube lifetime, The most

constructive way to do that will be

(1) The use of multiple thin skewed tubes

(2) Dilution of the fuel with fertile salt

94
 

3,2.,3 Primary Coolant Velocity

Section 3.2.2.4 pointed out that v, flow area, and AT; are the

only variables for removing heat from the core.

3.2.3.1 Maximizing velocity and minimizing pumping power,  ToO maX-
imize velocity ane desires as few bends as possible. Flow through a
sphere like Fig. 3.2-2, unless baffled samehow, would proceed more

slowly near the surface than through the middle. This could lead to
surface hot spots as well us neck corrosian. The ideal ge-metry for

maximum v would seem to be large-diameter pipes as described in section

Jelele

Fig. 3. 2"'2

Typical Spherical Reactor

 

For canstant mass flow rate the coolant velocity, v , will vary in=-
versely as the tube cross sectional area. Pumping power increases as v3;
for v {15 m/s, pumping uses 5 = 8% of the power. Friction losses in

small pipes will add to this. This recommends a few large pipes for

 

maximum flow velocitye.

95
 

3.203.2 Corrosion dependence on velocity. How velocity affects cor-
rosion depends upon the corrosion mechanism which, in turn, depends on
the metal and its environment [80], For corrosion by activation polare
ization (Curve B, Fig. l.2-2), velocity has no effect,

Corrosion by cathodic diffusion does increase with velocity as in
Curve A/Section te Small amounts of an oxidizer, e.g. dissolved oxygen
in acids or water, czause this, Easily passivated materizls such as
stainless steel and titanium frequently resist this corrosion better af
high velocity: there the increased agitation causes an active-tow
passive transition (curve A&, sections 1 and 2).

Some metals resist corrcsion in certain mediums by forming massive
protective films on their surfaces, These differ.from the usual pas~
sivating films in that they are readily visible and Such'less tenacious,
Extremely-high velociiies may mechanically damage or remove these films,
resulting in accelerated attack (curve C)s This is called erosion cor-

rosion,

 

 

 

 

 

f c Corrosion Mechanism
- 8
S A: Cathedie diffusian
¢ B: Activation polarization
S| 4 C: Erosion
0 1 v 2

VeloCity e

Figure 3.2-3 Behavior of Different Corrcsion Mechanisms

96
 

3e24,3e3 Velocities in similar systems, In an LMFBR core sodium
typically flows at 8 m/se In the HFIR core, the coolant reaches
120 £t/s (37 m/s).

In theoretical studies, Taube [17,79] considered velocities of
10416 m/s in the heat exchanger, choosing 14 m/s as optimume. The
British [29] varied v from 4~11 m/s in their NCFR heat exchanger. ORNL
designed an MCFR heat exchanger with a fuel salt velocity of 6.1 m/s,.

3e244 Overating Temperatures

Numerous criteria will limit the primary circuit temperature,
Figure 3,2=4 illustrates the discussion.

3.24441 Upper limits on fusl salt tcmperature, A high fuel sali
temperature could increase the feasible AT and thereby provide extra
hezt transfer driving force, This would then reduce the needed inter-
mediate heat exchanger area, and therzto the external fuel inventory.
Temperatures above 900°C allow operation with He as the secondary
coolant. Even higher temperature would allow MHD topping cycles and
increased efficiency irn conventional heat exchangers,

However, the temperature musn't exceed the salt boiling point or
cause damage to the piping materizl in or out of the core, Temperature
increases the rate of almost all chemical reactions, Corrosion fre-
"quently accelerates exponentially with temperature once past some
threshold value, Refractory metals such as Ho and Mo alloys have high
melting point and {thermal conductivity and resist corrosion well,

With graphite piping, the fuel salt temperature limit might extend to

150000. (Melting points of graphite or Mo are much higher ,)

I
86

 

temperature

1. corrosion
j 2. oxidation

Temperature SUpper 1imits on fuel salt
B

Figure 3424 3. Other chemical reactions
fuel L« structural melting
Temperature ° ( 5. fuel salt boiling
6

Constraints salt y . excessive vapor pressure
in Heat .
Exchanger ,70000 _— _ safety margin

 

Design primary circuit AT; across heat
safety exchanger =~ AT across core
margin

- on temperature difference
fuel salt‘)'b < across heat exchanger tubes
freezing 1. static stresses

point 2. dynamlc stresases
: 3. migration of elloy con—
stituents leading to

 

 

 

 

///J accelerated corrosion
minimum tem-— secondary ]
perature for coolant
process heat —t~~
applications
safety
margin |
sécondary C L
ccolant freeging 7
point
End A heat End B
exchanger

position
The British [29] considered 970°C with He cooling and 810 with Pb
cooling. Taube studied 860-1100"C and chose 950°C, ANL [21] stayed
with 740°C. The French considered 1000-1300°C for high temperature
applications but warned of problems to be overcome,
3,2.4.2 Minimum fuel salt {temperature, The minimum inlet temperature
must exceed the salt melting point with an adegquate safety margin.
Although melting points of actinide chlorides decrease with transmute-
ion up the A-scale (Section 3.6.2.3), impurities also build in which
might raise the melting point through the formation of complex com=
pounds,

Table 3,3-VI suggests 2 minimum T of 65000 to allow some

fuel
range in composition; addins a BOOC margin glives TOOOC. For comparable
MCFR(Pu) concepts, the British chose 650-670°C; Taube used 750°C;

ANL, 625°C; OZNL, 566°C.

3e2.4.3 Hinimum secondary coolant temperatures. Secondary coolant
salts with 2 variety of melting points are available, at least in
principle (Section 3¢10.1e1). In practice other problems and

properties may limit the choice. For process heat applications, the

secondary cocolant temperature must be fairly high,

99
 

3e2sded Restrictions on AT across heat exchanger wallsz, For fuel
salt flowing through a hollow tube,the maximum stress occurs at the
outside cylindrical surface of the tube walle It surpasses the wall

material strength O (psi), such as the yield strength (YS), when the

AT across the wall exceeds

 

o(1-9) _ § [z, 2. (=)
AT, = X E _SE[O‘*b In bla

1 at
2 '|n b’a_ - bi_o_z (302"'3)

 

where ¢ = volumetric heat generation rate, Btu/hr in3
k = thermal conductivity, Btu/hr F in
= coefficisnt of lincar thermal expansion, F'1
E = modulus of elasticity, psi
D = Poisson's ratio
a = inside radius, in
b = outside radius, in
All three terms in this expression are positive; the second numerator
term is generally quite small compared to the first,
A safety margin should most likely be in ATw itself; this will
allow for abnormal transient conditions which produce
(1) High fuel salt temperature
(2) Low secondary coolant temperature
Two materials are of principal concern, Mo and graphite, Table
3,2-1 lists a2 range of properties for graphite, To find the largest
AT _, property Set 1 maximizes G and R, while minimizing o{and E,
Complementary Set 2 should give the smallest 1}Tw. Table 3,2-I1 shows

the results for graphite and 1o under the principal assumptions of

100
 

Thermophysical Properties of Mo and Graphite
Graphite Graphite

 

 

Material Mo* Graphite** Set 1 Set 2
E(106psi) L8 0.5—1.é 0.5 1.2
& (10571 2.8 1.0-22 1,0 2.2
v (Btu/hr F in) 6.5 T oi=9.7 9.7 T ol
¥ 0.32 0.25 0.25 0.25
0.29-YS (1o3psi) 99.8 —_— — —_
UTS (10°psi) 101. 0524 2.5 0.5

* arc cast, hot-rolied
** electrode quality
UTS: ultimate tensile strength

YS: yield strength

101
Table 3.2-I1
Caiculated Stress Limits A'I‘w Across Heat Exchanger Tube Walls

 

G- core
volume
Material  IR/CR (10°psi) () At (€)
o 0.20/0.25 in* 99.8 0.25 586
99.8 0.5 596
99.3 1.0 601
101. 1.0 601
Le5/5.0 mm%* 99.3 1.2 581
0.20/0,25 inx  99.8 2.0 604,
439 0.25 288
49.9 0.50 289
{ LS.9 1.0 250
Le5/5.C am**  LS.9 1.0 298
0.20/0.25 in* L45.9 2.0 230.4
Graphite 0.20/0.25 in* 2.4 0.5 1313
Set 1 24 1.0 4317
1.2 1.0 2157
Graphite 0425 Q.25 68
Set 2 0.25 0.5 76 .7
0.25 1.0 81
0.25 Q.5 162
0.50 1.0 166
* Ax = 50 mils
#* Ax = 19.7 mils

102
| (1) @ = YS or UTS
| (2) A O,20-inch radius tube with 50~-mil wall thickness ( a
popular size)

(3) g = reactor power/(core fuel salt vcolume of 1 m3 and perti-

nent variations thereof)

With graphite Set 1 parameters, the siy's *he limit almost. Sef 2
parameters would allow a maximum szw of 162-16600, depending on the
exact power densiiy, A safety factor of {two would lower the maximum
AT, to 70-80°C.

For llo the fixed range of parameters appears to offer more certain-
ty. The z2llowable zyrw‘= 60C C seems very adequate, easily accomodating
a safety fzctor of two, One sbserves thawu

1e¢ The results do not depent on g which varies inversely with

core volume; i.e. the second numerator term in Eqe 342-3 is

generally small

 

2o - W2ll thickness does not set the szw limit, at least dowm to

20 mils
3. The most sensitive quantity is o/«E
In conclusion, llo seems to have an adequate safeiy margin; graphite
utility may depend on what kind of quality control can be achieved at

reasonable cost, The two fluids in the heat exchanger should counter-

flow to minimize the A‘I‘w.

103
 

3e2e4e5 Restrictions on ftemverature differences within the primary
circuit, The maximum Zle within the primary circuit will be the
difference between core inlet and outlet temperatures, A similar [&Tf
occurs betfween heat exchanger inlet and outlet, NS3R experience in-
dicates that when

(1) The reaction between the fuel salt and a structural alloy

constituent stronzgly depends on temperature, and

(2) Temperature differs greatly within the circuit,
then the constituent dissolves at the high temperature regicns and
deposits in colder regions,

The scope of this study Joes not permit quantification of this re-
striction; however, the poteztial mass transfer does discourage going
far beyond the circuit AT, in similar reactor designé’(below).
3e2e4e6 [ AT, T] in other MCFRs, Figure 3.2-5 summarizes the resulis
of previous MCFR(Pu) designs, Taube's [17] and the British He-cooled
designs [29] are perhaps the boldest in using higher T's and AT's,
Many desisms use a lower melting salt (with lower actinide molar con-
tent) to operate the fuel salt below 70000.
3elelte] Summary, Based on the above discussion, this study considers

(1) A minimum fuel salt temperature of 700°C

(2) A referemce AT, = 200°C; advanced, 300°C

(3) A reference Amiéx = 200°¢; advanced, 3OOOC

(4) A reference LMTD = 150°C; advanced, 200°C
The advanced AT's exceed those of earlier studies (except Taube's)

and are to be avoided if possible.

104
o]

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1000 : C ;
e . C{l A :
Dl b fwg‘ ' i
s [ S o i
Temperature 800 Ol &Qa "<i|]' L
(°c) b b 8
600 i ?/‘ -.;f" |
400 T i ,.*l" ’ T 'l v
. e i o
, - Pl e t-r- '
T i"!” !‘T?J;f'i i
200 L. P b ]
Study ornt 13 a8 Paubot ! Dritish=12? British-22Y British-3>’
Year 1956 1906 197080 1975 1975 1975
Intermediate Na or ta
Coolant Na F 3alt Na Pb Pb He
AT fuel (°¢) 167 115 200 150 160 300
%
o (°c) 345~
across tubes 120 —— 429 113 189 185
Maxe AT 350~
aoross tubes 167 — 520 140 210 270

*
Figure 3,2~5, Review of MCFR(Pu) Heat exchanger Design {Afl%

* See IMigure 3,2~6 for clarification of diagrams if necesasary
*H defined in Section 3,251
3e245 Desicn of the Primary Circuit Heat Exchanger
3e265e1 EHeat transfer coefficient, The heat exchanger plays z crit-
| ical role: virtually all the heat generated in the core must pass
through its tube walls into the secondary coolant, Glasstone and

Sesonske [81] show that:

 

 

= 2TR 3 '
q U(2TTR, L &) &T (3.2-1)
1= R ., R @/R) 1 (342-2)
U h 2. k
i W hi.c.
where
q = heat transfer rate (cal/s)

U = overall heat transfer coefficient (cal/s en® )
R = cutside tube radius (cm)

R.= inside radius (cm) of tube through which fuel salt flows

 

k = wall thermal conductivity (cal/s om X)
ieCe= Film coefficient of intermediate coolant (cal/s cmZK): this
term is generally non-controlling in (2) above
L = fupe length
N = no, of tubes
z&Tw = temperature difference between fuel coolant and infermediate
coolant (across the tube walls), Generally one uses the log-
AT, — ATa
In (BT, /ATa)

zero-dimensional approximation, as defined by Fig. 3.2=6,

mean temperature difference, LMTD = y for a

The heat transfer coefficient for the fuel is

Bt M (342-3)

106
 

 

 

AT ///////// AT _ AT
f A%
:53’
2>
AT
a
Fnd a: End b:
fuel out fuel in
14Ce in ieCe oOut

Figure 3.,2-6 Heat Exchanger AT Definitions

where k = thermal conductivity of the fuel(cal/som X)

D

1}

diameter of tubes throush which it passes {(cm)

For turbulent flow (Re >1000) of fluids with 0.7< Pr ¢ 700
through circular tubes of L/D:>60, the empirical Sieder-Tate equation
gives

¥u = 0,023 ReO'BPr1/3 (/u&[/xw)°‘14 (3.2-4)
where "b" refers to {the arithmetic bulk temperaturs, and "w" to the

average wall temperature, and where

107
O
<
D

 

Re =
&
Pr=—aL
R
v = linear fuel velocity (m/s)

il

fuel density (g/cmB)

fuel viscosity (centpoises = g/m s)

fuel specific heat capacity (cal/gm X)

C
P
To use the above convenient units we must convert centimeters to meters

1 03.2-0. 65

within Re and Pr, Doing this ( x 0,023 = 7.97) and grouping

terms gives

he (E‘é’%%fi) - 7,97 (vp) 0.8 | 0467 (cp//L)O'B/uw'O'MD‘O'Z
(342-5)
Equation (3,2=-5) indicases what makes a good film coefficient:
high values far/o 'V Xy cp’1é“4 and 1/D in that apfifoximate order
of prioriiy.

From Section 3.3.5.3 h. is typically 1.2 cal/s em'K. Assuming R .=

0,20 in, B = 0,25 in, and k= 0.32 cal/s em’C, get

 

1 . 0:25x 25412 1,25 1
T ~0.8x 1.2 0.32 B, (3.2-6)

= Bl§3 + 5—%3——.+ smaller term

U =0u67 cal/s cn® K
The importance of the second term relative to the first in

(3,2-2) or (3.2-6) is

 

 

 

hy ) h R, Roln(Rd/Ri) ] he R, 1o (Rd/Ri)
h/ - R e -_ k ®
0 W W

0.42 here

i
o
e |&
N
ol%i

i

108
 

Alternately for IR = 4.5 mm, OR = 5.0 mm as in Table 3.2~IT

h
L o= 1.2 x 0.451In (50/45).= 0.18
h’ 0.32

This shows that the fuel salt film coefficient controls the heat removal:
going to a thinner tube will not help much. The significance of Mo
would be that it would allow a wider and thicker (therefore more reli.
able tube without heat. transfer penalty.

For comparison, the ORNL study reported [13]

 

 

. -1

Us|—t—r 4+ —2 ¢ -t - 0.15
0.30 0.5 1.8

They used the same dimensions as here, but their fuel salt hf was three

times smaller: they assumed 17% UCl3 fuel salt and 83% carrier salts

{MgCl., & NaCl) while this work presumes high actinide content. The wall

2
material k in the ORNL study was 6.5 times lower (80 Ni/20 Mo there vs.

pure Mo here).

3.2.5.2 Minimizing fuel inventory in the heat exchanger. The ofix-of-
core fuel inventory affects reactof stability (Section 3.6.2,3}, U
inverory costs, and out;of-coré‘criticality.- The -fuel volume in N
exchanger tubes 1is

2
vfubes = NLTrRi ' (3-2-7)

A reasonably-ccnservative guess of the additional fuel in heat ex—
changer plenum and piping appears to be (section 3.6.2.3) ane additional

equivalent volumes ‘Then

109
 

2
OeCe vou‘t-of—core ~ 2 NLT Ri

(342-8)

Rearranging equation (3.5=1) gives

There

nat e

This

or

Given

sity
suffi
vance

would

2 ¥ LW=q/UR AT

fore

2
Voo = By g/ U R, AT,

The condition for stability (section 3.64243) is that v

»

xceed fthree times the core volume, Vc, leCa:

vo;,c. < 3 VC

means

2
R q<3V U2 AT

2
V,AT, > R “q/ 3R

U = 0.67 cal/sec em’ K

R;= 0620 in = 0,508 cm

R= 025 in 00535 cm

= 2250 x 106 watis x 042388y cal = 537.5 x 106 cal/sec

watt sec

VCATW > 109 m3 K

Figure 3.2-7 plots this relation, adding boundaries of power den-
{10 MWth/liter and.AEw<:300. A broad reference region lies
ciently far from the limiting boundaries so as to not require ad-
d design development, To further broaden the reference region

require smaller Ri or larger U,

110
 

Ou‘l'cr flTm“ lelf

. in Other Desfim
Direction of
Neutron Spectrum

Dea)f‘ad.a-‘hf:n \
/ \
\ .

20+ /\ L (1125 MWy [2)
\

4 \ \
. / \ \
{5 - \
. AR $
Critical \\ ) o nc:rcble
Volume | ' ! requ'or\
Vo, (,‘,,_3)

0.5 4

A \ , \
\ /
10— \ e +(2.25 MW, (L)
N
\ .
wz— Ve AT=10q n? K

Sfa_b'.\'.k‘ Limi¥
/\ \
~
7

oMWy (). .
_f. Poq::r&'f)! *‘\, Lumd'

Al

 

e i
—

 

 

 

T Y
o 100 2060 36 400

AT, (x)

O

Figure 3,2-7, Locating the Optimum ATW and VC

111
' Lzrger U implies higher kw and hf and lower effective wall thick-
ness Ax = R In((R O/Ri). Both pure Mo and graphite exhibit high k
(Table 342-1)s Low Ax implies keeping the walls as thin as possible,

| This dictates a metal with good resistance to corrosion, especially

‘ with 2 high temperature-gradient present. It must also withstand the
radiation damage from the delayed neutrons; either that or be cheap

and easily replaced.

3.2.5.3 Volume in heat exchanger and associated plena and piping.

From equations 3.2-1, 3.2-6, and 3.2-8

0.239 cal

 

 

 

2250 MWth X =t = U(2T R, L N) AT,
100 MH s
'
= UR oc AT
O 2 W
R
8 -
v 2:25x239x10° B2 w0t e
% 0.67 x AT (K) R, cm?
2
- R‘
VOC(mB) = ..E_g—— X e (cm)
ATw(look) Ry,

For cases of (R;,Ro) = either (0.20 in, 0.25 in) or (0.45 em, 0.50 cm),

get

3.3

oc
ATw(loaK)

Considering ATW —= 150 K from Section 3.2.4.7, 8et

— 3
Voc ~ Zm

112
 

3.2,6 Primary Circuit Arrangement
3.24641 General Guidelines, Sections 2.5.4, 3.6, and 3.7 and 3,9
show the need for transferring the core and blanket fluids to remote
storage tanks, and flushing the primary circuit. Sections 3,6,2 and
3.8 emphasize minimizing the fuel salt inventory. Section 2,5.,5
mentions the importance of reliability and plant factor.
3424542 Location of pumps and heat exchancers., Both pumps and heat
exchangers should closely surround the reactor as in Figure 1.,2-2 so
as to minimize the fuel inventory in pipinge Shielding from core
neutrons, beyond that provided by the blanket, is rot warranted as the
primary coolant itself emits delayed reutrons, Placing the pump between
the heat exchanger and the ccre inlet imposes a lower overating tempera-
ture on it; this should lengthen its service and easeiits designe
3e2e643 Parallel subchannels, As each tube channel (of N total) leaves
the core it could split into M>2 subchannels, each with a small pump
and/or heat-exchanger., Such redundancy should increase load factor
and safety of operation, Small size would also make it easier and
cheaper to build and replace the units, and easier to avoid criticality
outside the core. However, as the number of pumps increases, the chance
for a single pump failure does also, Thus the optimum plant load factor

could occur for a smaller number of pumps and/or heat exchangers.

113
 

Several alternate arrangements are possible (Figure 3,2-8), Plan
A schedules a single pump and heat exchanger for each channel, Plan 3
splits the flow outside the core into M>2 parallel paths, each with a
punp and heat exchanger. Plans C and D only duplicate pumps., Thus in
3; Cy or D, 2 single pump failure does not terminate all flow,

Approaches A, B, and C differ significantily from the multiple cool=-
ing loop concept of current reactor designs: the principal N loops have
no common container as in Plan D: each circulates independently. Plan
D probably is infeasible because of the

(1) Desire to compactize space so as to keep down the fuel in-

ventory (minimum piping per heat exchanger)

(2) Difficulty in avoiding a critical configuration

if a single pump fails with plan A, not only must that channel
flow cease, but the others' must also: +the whole reactor must shut
down less the chain reaction continue to generate heat in the broken
channel as well as in the two good ones, 1In plan 3B or C, failure of
one pump again affects the whole reactsr: +the flow and power reduce
by 1/M in the bad channel and the other N-1 core channels must also
due to the close neutronic coupling,

If M=2 subchannels in plan C represent regular and backup paths,
then each can handle the full flow and one is always ready in backup -
just need to switch a valve, PFigure 3.,2-9 shows a top view of such 2
plan, Use of multiple (M22) subchannels especially for pumps, should
increase reliability and ability for reduced power operation, but may
also increase fuel inventory;auxiliary heating, and friction (due to

added bends).

114
 

 

 

 

7]

 

 

 

 

 

 

 

 

 

Plan A Plan B

 

 

 

 

 

 

 

———a et —— —— o —
—— e p— —— — o - —

 

 

Plan C Plan D

Figure 3.,2-8, Alternate Pump and Heat Exchanger
Arrangements for Each Core Channel

115
 

/r\_;_/ Shielding

Pumps

=

Heactor vessel’

  

. - m . ~ o i
3 - - . DD e L L T *at - P
ige 3.2-9 Tcop View cf Plzn 7 itk Altermatz Subcharnals 3o Snie

Calls

Fig. 3.2-10 Schematic Diagram
of Series Cannection of Three

Skewed Tube Channels

 

 

 

 

116
 

3.,2.644 Tube channels in series, ror a skewed core (recommended in
section 3.143¢4) the inlet and outlet of a tube occur on opposite
sides of the reactor. To avoid excessive piping and fuel inventory
the outlet of one tube shouid connect to the inlet of another, which
lies vertically belows This means connecting all N channels in series,
Fige 3¢2=10 attempts to describe this. It's as if someone ran a
needle and thread through a cylindrical cushion; coming out the top,
the thread returns outside the cushion to another entrance point on
the bottoms, This continues until the thread reiurns to the starting
position.

The series conneciion causes failure of one pump or aeat exchanger
to shut down the whole system. But this does not differ from plan 4,
and one can still employ duplicate subchannels as in.% or C above,

Figure 3.,2-11 shows a B-type horizontal M=2 design. Here the
tubes peneirate sides rather than the top and bottomes This should
simplify the reactor support and reduce the out-ofwcore inventory as
well, A physical model would be three pencils lying across each other,
penetrating an approximately-square cylinder of (blanket) jello.
3.2.6.§ Summary. In the reference concevi a small number of skewed
tubes connect in series, Alternate subcharnels with pumps in each
could provide backup, but would add frictvion losses, On-~line backup
(to avoid reactor shutdown) would also increase inveniory and auxil-
iary heating. Keeping N to 3 or 4 should help the system reliability
as well as reduce the time required for repair and maintenance. It

also minimizes neutron spectrum degradation and parasitic absorpticn.

117
 

 

 

— — = — R .

 

 

 

Bocikup HX & Pump
(BHXP)

     

 

 

 

 

 

 

 

 

 

 

 

 

PLAN Cne of Three zZmergency
VIieid Lines to Core Salt
Safety Tank
1 Blanket ‘ 1
HX Core tube HX
L - - - F - — 4
( T =) !
‘ Blanket l T
Blanket Salt Drain \‘J
to Safety Tank
ELEVATION
VIizd

Figure 3.2-11 B-Type Horizontal M-2 Layout of Pumps and Heat
Lxchangers.

Arrows denote core sait flow.

118
Limits on power density may come from
1e Radiation damage to the high flux
2o Hizh temperature of the fuel coolant causing leakage through
a seal expansion, melting of metal, or chemical corrosion
3, High fuel inventory in the heat exchanger, affecting eco-~
nomics,dcubling time, and reactor conirol,
Use of replaceable graphite for the material in contect with the fuel
coolant throughout the reactor and much of the primary circuit might

eliminate much of concerns (1) and (2).

119
3.3 Choosing the Salt Composition

The abundance and low neutrcn moderation of chlorine maké
sodium and actinide chlorides the preferred molten salts for fast
reactors on the U/Pu cycle (Section 1.2). Here we ccncentrate on

the thorium fuel cyéle which could have different priorities.

3.3.1 Intrcducticn

 

The top cancerns in chéosing the sélt ccmpositian are neutran-
ics (especially neutron spectrum), chemistry, and meiting point.
Cther desirabie tréifis are

1. Good heat transfer properties: high thermal canductivity

and specific heat cazacity

2. High boiling point to minimize vapar pressure

3. Adequate technological cr laboratory experier;ce

L. Low fuel salt viscosity to minimize pumping ccsts

5. Low price and good availability

6., Non—toxic

J::.1.1 Neutron spectrum. We desire as hard a neutron spectrum in
the core as possible, though not at the expense of BG. Sifice non-—
fissile constituents mainly downscatter and capture neutrans, this
means maximizing the ratic of 233U to all other isotopes. |
In the blanket, a softer neutron spectrum enhances neutron cap-
ture by the fertile material. Thus neutrcn moderators might help
‘there, if they don't absorb neutrcns as well. However this would

not be good for an inner blanket as it would then soften the ccre

neutrcn spectrum.

120
If the core tubes were many mean free paths apart, then each
tube would individually approach delayed critical and have a hard

spectrum, but their ccupling would be loose or near-zero.

3.3.1.2 Chemistry. Using the molten salt fuel as the primary cool=
ant of a nuclear reacter presents many novel problems in both reactor
design and system chemistry. The success of the CRNL Molten 3alt
Reactor Program shows that it can be done at least up to their pow-
er levels. Experience there also revealed that the chemical states
often behave as if in equilibrium. This allows much progress tcowards
understanding interactions within the salt (chemical stability) and
between the salt and its envircnment (corrosion) since many of the
equilibria are grossly predictable.

The free energy of formation of the compounds, AgG, measures
the chemical stability of the salts, specifically their ability to
resist forming other compounds which precipitate out. Taube has
deduced the temperature-dependent A G for salts of interest (Fig-
ure 3.3-1;[82].

Candidate reactants in the system which threaten salt stability

l. Atmosphere constituents oxygen, water, and carban diaxide
2. Structural materials (Sectiom 3.4)
3. Fissian products and other mutants
Radiolysis also affects salt stability.
Chemical corrosicn involves several sciences (Fig. 3.3-2).
Physical chemistry and metallurgy describe the physical, chemical,

and mechanical behavior. Thermodynamics reveals the spontaneous

121
 

 

—400

 

300 500 1000 1500 2000
TEMPERATURE [°K]

Fig. 3.3-1 Free Energy of Formation for Chlorides [82]

122
Electrochemisiry

 

Physical

Corrosion <«——————— ietallurgy
chemistry

Thermodynamics

Figure 3,3-2. Factors Affecting Corrosion of a Metal

direction of a reactiocn ard whether or not corrosion can occur.
Electrochemistry describes electrode kinetics: the dissolution and

plating cut of different materials in electrical contact.

3.3.1.3 éutecticrmeltingApoint. Figure 3.3-=3 depicts a typical
eutectic behavior: the mixing of two salts lowers the melting
point far below that for either salt by itself. The amount of low=
ering depends on the molar ratio of the two salts; excepticnally low
temperature is possible at the eutectic point, or nadir. This may
be of special value for the blanket region where litile heat 1is
generated.

A fuel mix with melting point well below 700°C would minimize
auxiliary heating when shut down and allow a large temperature rise
in the core without reaching high outlet temperatures (> 1000°¢).

Low melting points of the component salts also will facilitate their

initial dissolutione.

123
Carrier salt Actinide salt

like NaCl like ThCl4
e — = — = e m - e -
700°~1100°¢C
Melting
Temperature
300°-700°%¢

 

 

 

 

0 50 100

Actinide MNolar Proportion

Fizure 3,3-3, Tuteziic [lelting Point Behavior for = ixture

Adding 2 third component like a second actinide szlt in
Pigure 34,3-4 can further lower the melting point, especially when all
three are in near-—-equal proportions., Unfortunately many multicome

ponent (ternary or higher) phase diagrams are unkmown or in unavail-

able Russzian literature,

 

 

~T70°C

Figure 3.3-4. Ternary phase diagram for Nac1/9u013/frh014

124
3.3.1.4 Heat transfer parameters. Sectian 3.2.5.1 showed how MCFR

heat removal varies according to

g 047 ©.33 :
h; ~ p O® kfi (.gf ) J//Jw-0.4
/ /

For survey purposes here let

o (bulk temperature ) i//b{(wall temperature ).
-~

Then

O-B |- 0067 .O'L]-7 0033
- {
he /J < /6 Cp

The exponent magnitudes measure the.relstive significance of the con-

stituent parameters,

As cother factors disccurage the use of fluorides in the core, this
section studies oniy chliorides. Infermation an density ( fD ) is read-
ily available for pure salts and a satisfactory algorithm exists far

mixtures. Viscosity (/}1 ) and specific heat (Cp) data also suffice.

Thermal conductivities are very sparse.

3.3.1.5 Densities. Desyatnik et al [83] studied molten mirtures of
UClh and alkaii chlorides. They found the density of individual salts

to decrease linearly with temperature, getting

n

(LiC1) = 1.8807 = 0317 DKL £ 0.0011  g/ca’
1000

(NaCl) = 2.1332 = 0.5405 T() +- 0,0010
1000

(xc1) 2.1751 - 0.6060399— + 0.009
1000

(RbCl) = 3.,1069 ~ 0.8799 TE) + 0.0013
1000

125
4 (CsCL) = 3.7726 — 1.0716 ZE)L L 0.0011
f

1000

/4

£ (BeCl,) = 2.276 - 1.10 ]g—é%-L + 0.0011
(MgCl,) = 1.976 - 0.302 T(K) 4+ 0,001l

1000
9 (CaCl,) = 2.5261- 0.4225 TgK% + 0.0011
10

J (SrCl,) = 3.3896- 0.5781 %%l + 0.0011
‘ T(K) +
O (BaCl,)} = 4.0152- 0.6813 0.0011

l —

P (PoCLy)= 6.6213 - 2.0536 2L 4 0,012
/ 1000

/,-J (ThCL, )= 6.2570 - 2.7030 IK) + 0.015

; 1000
f
O (UCL3) = 7.520 - 2.472 )
/' 1000
;
o (vey,) =-5.6251 - 2,292 ZEL 40,0021

1000

Kinosz and Haupin [ 84] ‘give the density of NaCl and KCl as
/ (LiCl) = (1.463 + 0.002464 T)/(1 + 0.002 T)
’ (NaCl) = {1.611 + 0.001897 T7)/(1 + 0.001538 7)
(KC1) = (1.568 + 0,001619 7)/(1 + 0.001429 T)
/ (MgCl,)= (1,682 + 0.0044628 T)/(1 + 0.002857 T)
/3 (CaCl, )= (2.115 + 0.003768 T)/(1 + 0.002 T)

126
where ¥ = T(K} = 973 = T(OC)- 700. A spot check at 1000 K showed
fair agreement for NaCl and KC1, getting /o = 1.596 and 1.552 by
Kinosz vs 1.593+0.001 and 1.569+0.001 by Desyatnik.

Equivalencing the two formulas one gets

Kinosz R (LiC1) = 1.463 (1 - O.-000316 )
f)(NaCl) = 1.611 (1 - 6.000360".')
/o(z«:m)' = 1.568 (1 — 0.000396T)
F(b1g012)= 1.682 (1 - 0.0002047)
/0 (CaCl,)= 2.115 (1 = 0.000218 T)

O(LiCl) = 1.461 (1 - 0.0C02967T)
Desyetnik /p (NaCl = 1.607 (1 - 0.0C0336 T)"
/J (KC1) = 1.585 (1 = 0.000382 T)
/o (RbC1l)= 2.251 (1 - 0.000391T)
/o (CsCl)= 2.735 (1 - 0.000392 T)
/fi(PbClz)=h.623 (1 ~ 0,000444 T)
P (ThCl, )=3.627 (1 - 0.000745 T)

)

/o (U013)= 5.115 (1 - 0,000483 T
/fi (Ucl, )= 3.395 (1 - 0.0006757")

The Kinosz equivalencing is by

b
—_ - b
1+c¢T 14+¢7T
The resulting slopes (v coefficients) are expectedly higher than

for Degyatnik

127
For a mixture of salts Desyatnik showed [83]} that linear addition

of the molar (specific) volumes of each salt

P = LI wifp ]

gives densities high compared io experiment. Direct addition of
densities would give even higher errors, in the same direction. In
binary mixtures with carrier salts like NaCl, the maximum deviation
occurs a2t about 60 mole % actinide salt. A spot check on a 50=50
UclL 4/1:5.01 mixture at 1000K predicted /c (50/50, 1000K) = 2,91 whereas
the measured was 2.871+4¢ 0.,0024, Formation of complex salts, e.g,

3 KC1 + UClB-—-b-K3UC15, causes this deviation from linearity.
3.3,1.6 Viscosities, Section 3.2.5 showed that low viscosity in-
creases heat transfer. Fizurs 3.3=5 displays the temperature-dependent
viscosity for individual salis [85,86] beginning at their melting
points,

The viscosity of salt mixtures is gemerally unavailable; Figure
3,3-6 does show viscosity behavior for two salt mixtures: first Mg012
with NaCl [87]; then with ThCl, [88]s (Note that the two sets of pure
Mg012 measurements do not agree. The (latter) Russian work is partic-
ularly suspect because of other errors found there, Still the data are
informative here,) The bumps correspond to formation of complex
moclecules,

Comparison of these diagrams with phase diagrams Figs. 3.3-7 and
3.3-28 suggests a correlation with eutectic effects which, in turm,
depend on complex molecule formation. In general it appears feasible
to choose a2 molar composition where the viscosity is close to the low-
est individual salt viscosity. Therefore, for ThCl4/Nacl, UClB/NaCl,

UCl4/NaCl or combinations thereof we can roughly presume

128
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

      

 

D b b ——— ——— ———
1 - —— > 4
—— - - .

o=

L et e e

 

 

 

TS IT I T Tt tn S oo
e e
== f_om T e Tk — g )

 

4 5 6 7 8 9 10 11 12 13 14

" Temperature (100°)

Figure 3.3-5 Viscosity of Individual Molten Salts [85; 86].

129
= T TRttt T T - Poo - It

e e e e e L (L

 
 
  

 

 

 

 

 

 

 

VI S Cos-d'y
(ep)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Hsuz 'ThCh;
ThCl, mole ofe

Figure 3.3-6 Viscosity Behavior for Two Kolten Salt Binary

Mixtures [87, 88]

130
900

 

 

 

 

 

 

 

 

 

800 |
700 F
Tamp=
ature
(°c)
600
500 }
J.LOO . i . g
0 20 40 60 80 {00
MLl, MOLE % NaCl NaCl

Figure 3.3-7. The MgCL,~NaCl Binary Phase Diagram [13]

131
M) = Mgy (B

From reference [85]

Myacy = 186 x 107 exp [3308/1.987 T(K)] cn.

3.31.7 Specific Heat. Perry"s Handbook [89] recommends for fused

salts at and slightly z2bove the melting point:

- ( cal ) - 8.1 ( cal{oc ) ) (no. atoms)

D c 2tom male molecule
- gn} C
Molecular weight (gms/mole)

Data on chlorides (Table 3,3-I) from the Molten Salts Jandbook [85]
supports this formula for 2- and 3-atom molecules, For 4-atom mole-
cules, the coefficient increises to 8.5 cal/°C 2tom mole,

The next question is the temperature dependence ‘of the specific

heat, Table 3,3=II shows how experimental {03_1

-

ac, / dTE [90] for

some F Salts compare to their E;P°-1 d/o/dT}and coefficients of linear
sion gd_ CO
expansi P i * p and /Do represent values extrapolated to T=0

0=
p
fluctuations of the two parameters do not correlate very well however,

Kelvin, On the average & parzllels (C de/dT) fair; individual
Still this represents the best empirical correlation which could be
developed within the scope of this study., On this basis we approximate

c, (cal/em C) = cp° [1-al(X) ]

wnere
Cp = Cp (T} /L1 =T ()]

Table 3.3~III calculates some salts of interest. Figure 3,3-8 corre-

lates the resulting Z.CP [900°] } so as to deduce the missing value

for MgCl (for which o« was unavailable),

132
Table 3.,3=-1

Dependency of Specific Heat on No. Atoms per Molecule [85]

t

cp N Cp"
( cal(oc ) ( No. atoms) cal/ °c )
Salt mole molecuie molecule mole Qtom
LiCL 15.0 2 7«5
NaCl 16.0 2 8.0
KCL 16.0 2 8.0
CsCl 18.0 2 9,0
AgClL 16 .0 2 8.z
CuCl 15.9 2 g0
TiCl 1744 2 8.7
Avg. 16.3 2 8.16
NiCl2 2L.0 3 8.0
MgCl, 22.0 3 Te3
C&Clz 23 .6 3 7«9
SrCl, 27.2,26.7 3 9.0
BaCl 26.3,25.0 3 8,6
Ca012 24.C 3 8.0
Hg012 25.0 3 8.3
ZrC.L2 2Le1 3 8.03
Pb012 23 .6 3 7.93
I-InCI2 22.6 3 7.53
FeC'.L2 2hely 3 8.13
Avg. 2442 3 8.06
LaCl, 3747 L 9.4
PrCl3 32.0 L 8
Nd013 35.0 L 8.8
AlCl3 31.2 L 7.8
GdCl3 33 .7 L 8.4
HoCl3 353 4 8.l
ET013 33 07 1{. 8.14.
BiCl3 343 L 8.6
FeC.L3 32.0 4 8.0
f‘.vg. 3349 4 8047

133
Table 343-I1
Approximate Equivalence Between Specific-Heat
Temperature-Dependence and That for Other Physical

Parameters, Based con Heasured I Salt Hixfures

 

 

105 X CO=-
ol 100 doy 10 4P ot eveem

Hix Fracticns CE 0 K)d= (0 1)aT sion
NaF/ZrF 4 50/50 28.37 23.0 29,06
LiF/NaF 60/10 4o W47 2i.40 27.1
LiF/KF 50/50 32.12 25.70 34.0
LiF /RbF L3/57 36,52 26495 36.5
LiF/NaF/ZrF 4 55/22/23 24,51 23 .54 30.6
NaF/Zth/UF L 56/39/5 | 20.16 22.16 28.3
RbF/Zth/UF L L8/18/L, 3441 21.86 27 .8
LiF/KF/UF,  48/L8/L 51.42 20475 32.4
NaF/LiF/ZrF, 20/55/21/k 31.L46 22,54 29.3

/E,

Average 34210 2412 313

134
Determination of C (1) = ¢ [1- o 2(K)] and € (900 °¢) for
Chloride Salts of Interest

Salt

 

LiCl
NaC.
KCL

RbCL
CsCl

CaCl

SrCl

Ba.Cl2
PbCl
’I.’hGl/4
ucl

Ucl

‘s
(cal/%c
mole!

15.
16,
16,
16.2
18,0
23.6
2740
25,6
23.6
L2.5
3k.
b2.5

Mol

Ut

42,39

58,44

74456
120.92
168.36
110.99
153453
208.25
278,10
373.85
344,439
379.84

Table 343 = III

¢ [T ]

p-mp
(cal/"¢
£m

0.3539
0.2738
0.2146
0.1340
0.1069
0.2126

01703

0,1229
0.0849
0.1137
0.0987
0.1119

135

T

614
801

776

(
oy oy C

 

C'D

ey

po C_zm_
2.96  0.4799 0.3132
3.65 0.450%4 0.257°%
3.98 0.368,  0.196L
Le13 0.2263  0,1197
L.O4 0.1701  0.0895
2.07 0.2713  0.2054
2,17 042267 061690
2.20 01688  0.1253
L.10 0.1213  0.0645
5.57 02713  0.0940
Le2h, 041873 0.0941
5.26  0.2049  0.0785
 

 
  
 

\\ L -‘.__-_ = . ] .
0.2 ~--—————————\ —— ¥~\—alkaline-earth -—
L , L~ — chlorides )

e S S
cp[%n°c] T, N

, o |
—_— - Ih \ - . - . 1

(cal/OC %n.)l -—-—*———-—j—'gflo‘:id‘.es s et ,
: - - i !

 

 

| = : {
| 1 1 1 1

Molecular Weight

Fig. 3.3=8 Correlation of Specific Heat with Salt Molecular Weight

to Deduce C_ r900°c] for MgCl, e

136
3.3s1.8 Thermal conductivity. Figure 343=9 displays the measured
thermal conductivities of individual salts, all from Russian sources [91-
94]e Por other salts and sal$ mixtures, one must{ rely on phenomenolog-—
ical modelss Knowing rooa—~temperature k for a salt does not help in
predicting k(T) of the liquid: Figure 3.3~10 shows how k changes with
temperature as a sali pesses from solid to liquid state,
Gembill ({95] proposed that two distinct mechanisms determine k
for fused saltis:
1e Atomic or lattice conduction along the shori-ranzge atomic or
molecular order present in liquids
2e Ionic trzusfer by e drift of free charged ions beiween the
atoms, Gambill successfully correlated existirg sa2lt mixture

(mostly flucrides) data at their melting points by

caL/s
- O.S O.B 1.8
ki <cm = = 0.04308 T o /Omp / u fa.tomic_
(343~1)
where Top = melting point temperature (X)

/0 = density (g/cc)

M = avg, molecular weight of mixture
¥ = thermal conductivity (cal/s cm K)

L tomic = 04438 ~ 0,09 N (atomic part)

Ce
N ) .
avg = average number of ilons per dissociated molecule

(€ege 5 for ThCl, and 2 for NaCl)

4

137
 

   

Thermal o
Conduciivity
(¥/m K) '

 

=] PbC\z 3uess

 

 

 

 

 

 

500 600 700 8L 900 1000

Temperature (°C)

Figure 3.3-9 Thermal Conductivity of Individual Salts

138
 

 

The rmal
Cmdud(fihs

(W/m-C)

 

 

 

 

 

az

 

0 200 4b0 - 600 goo
Tempersture (°¢)

7iZe 343=10 Thermel Conductivity of T.-IgCl2 from Solid $¢ Liguid Siates

Wnen Navg > 54 ion drifting predominates and this approach fails
(£

mixtures of interest as core salt, Nan generally falls below 3, The

atomic 1S Bily or actuzlly negative), However, with fused salt

effect of complexing, e.g. 3KC1 + UClj—*-K§1 (IICJ_4)"3 will generally
increase Havg’ however,

For k(Tmp) of individual chloride salts Gambill's formula differs
markedly from experimental Russian data (Table 3,3~IV and Figure 3,3-11),
The consistent dependence of the theory correction factors with salt
molecular weight (Figure 343=11) suggests

(1) A different exponent than 1.8 for M (more like 0.8)

(2) A slightly different correlation wita Navg

For temperature dependence of k, Gambill [95] derived the following

expression

139
"Table 3.3-1V

Comparison of Gambill's Formula to Experiment for Individual Chloride

 

 

Salts
/ Koo Rme em
T al \ g
«Pe m.pe /mC ncal

Salt ‘Iavg. Lat. (K) M (g/ce) \s K cm)_ (;“F(cm)_ goate
LiCl 2 0.258 887  42.39 1.50 8.09 3.61 0ol
KCl 2 1049  7L.56 1.54 3.25 2.27 0.698
AbCl 2 988 120.92 2.24 1.78 1.79 1.006
CsCl 2 919 168.36 2.79 1.13 1.29 1.142
MgCl, 3 0.168 981 9s5.22 1.68 3.34 2.45 0.734
CaCl, 3 1045 110.99 2.09 3411 2,67 0.859
5r€l, 3 1146 158.53 2.73 2.12 2452 1.189
BaCl, 3 1236 208.25 317 1.52 247 1.625

140
correction curve

 

TEET T LTI ll.llrll:rlrlrilml!_‘_. e
= TTIIITT ‘.”Ha.w_wltx- T T s e e s
S - ; TR T o

 

  
 

 

~|1.]1/]'eyeball average

 

240

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Lo .- - - EEEE - [

S ~ : . - B g - . [
: - o - — - - v . S ae- . -

T et b en e = el o - ‘l/

ST T o L T T e e N S
Vo s e e T o e oL oTouTTL ;n_LHu.'..,H.‘_A

 

 

 

 

 

 

 

 

 

oI o — - / mou
- At . — - ——— —d e -l T T
I e = A LT LT/
I TTIIIIIT LTI e\ ———
o - ot e = A\ Y=
— b e — T e T L Tl I, N
— oo i e b g e T e e
i L L rm i el T LT N\ T

 

1&6 ’

Exf;er;ment - to -~
Corr:‘?:’h;n Factor
(hEXP /hctk)

at T;np.

Theo

141

Salt Molecular Weight

Fig. 3.3~11 Correlation of Experiment—~to—(Gambill's) Theory Differences with Molecular Weight
 

kt-otal (T )= ®ygea (T2) g‘pfimuc /;ZZD:) [5;2‘3}
+(1- &m;c)[m)][ ]/1

The usefulness of this formula depends upon the confidence develooped
in it; Cambill had no data with which to test it. Attemots to fit
the availapble (Russian) data for individual szlis fail drematically,
Inspection of the terms reveals the cause: k. . (first term) ore-
dominantly depends on/oi/2 which has a negative T-coefficieat while
experimentally dk/dT is posiiives For K: onic (second term) the YT
factor overrides the negaetive T-devendence in/o, but only up to a
certzin iemperaiure. Overall this approach falls.

Perry's Chemical Ingineering Handbook [96] describes numerous
other formulae but these generally pertain Yo organics and other non-
related fluids; also they either do not freat temperature dependence
or they grossly disagree.

In Table 3.3~V the electrostatic~bonded alkali and alkaline

earth chlorides of intermediate and heavy molecular weight appear to

follow
%: = 0,024 x I ¢ V(m‘.-I/m°K2)
where i1 = molecular weight

V = cation valence
For highly-covalent UCl4, dk/dT = 0,0023 x MV(mH/sz), a magnitude
amaller. For other covalent molecules, e.g; SiCl4, k actually de-
creases with T. As PbCl, belongs to this chemical class, its dk/dT

<howld be about =0.25 millitatts/mek,

142
Table 3.3- V
Correlation of Temperature Dependence of Salt Thermal Conductivity with

Catian Valence Number and Molecular Weight [91-54],

 

Cation [dk/aT]/H-v
dk/dT 5 Valence ¢
Ciass Salt (mi /m =< ) v M (107 W/m-Kz)
21k211 LiCl 0 1 42439 0
calorice .
FaCl 2.1 1 58e 44 35.9
KC1 243 1 74456 3049
RbCl 2.0 1 120.92 2448
CsCl 4.0 1 168.36 23,8
alkaline  MgCl, L7 2 95,22 2447
earth )
chloride a0l .1 2 110.99 2340
| SrCl, 6.8 2 158.53 21.5
Ba012 10.5 2 208.25 2542
(covalent)
actinide = 9oL 3.5 L 379.84 . 243
chlorides
other S1CL, ~0.25 4 169,90 ~0437
covalent ]
chlorides GeCl, ~0.15 4 214 4,41 ~0.18
SnCl, =0.,20 4 260,50 ~0,19
PC1, -0.30 3 137433 C «0.73
AsCl, -0.25 2 181.28 =Ce69

143
The few measurements of chloride mixtures available — RbCL/UCL A
and CsC1/UC1 g mixtures (Fige 3¢3=12 and 343-13) [94] — show that k
fluctuates dramatically with molar composition., The oscillations may
correspond to the formation of complex molecules just as occurred with
viscosity data, This behavior suggests that one might yet achieve high
k =k(pure carrier salt), even for high actinide molar composition and
vhen k¥ (pure actinide salt) is much lower.

Figure Je3=14 compares Gambill's predictions f{o the above date,
with and without the single~sali correction factors from Fige 3¢3-11,
In nost cases omission of the correction factor gives better agreement.

llote here how k very definifely increases with temperature: the
slope remains positive up through at least 60 mole % actinide salt, The
average velue of (dk/dT) /MeV from these six cases iSs 245 x 1070 W/m'K2
(or 640 x 10"9 cal/sec cm K2) in close agreement with the value of 2,3

-6

z 10 W m-K2 for UCl4

In summary, Gambill's formule for k(TmU) proved successful with

fluoride mixtures. With chloride salis it appears to succeed better
with salt mixtures than with the individual salts, Certain mixture
compesiiions should give a k comparable to that of the individual
carrier sali. For the mixtures of concern here, the covalent actinide
chloride tends to control the temperature depemdence of k: it wvaries
much less than the carrier chloride salt, To obtain high thermal con-
ductivity, both experiment (Figs 343-9) and theory (Gambill's formula
for high thermal conductivity) tend to favor salts with low M. In

Gambill's formula the factors Tmp and fgmp tend to cancel one another,

144
 

Thermal J
Cond nd\vr{'-’ 07 o D
(W w' x™) 7;[\\-7\0(/\
0.5 2 !
0.3 1 1 1 : 1
0 20 40 60 go 100

Mo‘t a/l-" UC‘Q

Figure 3.3~12 Thermal conductivity isotherms for fused UCl4-Rb01

 

 

mixtures at different temperature (K): 1) 883;
2) 993; 3) 962 [94]
Thermal
Conducfiu't‘l\’
(Wi k)
Mole %o .UC14
Figure 3e3=13 Thermzl conductivity isotherms for fused UCl4-CsCl

mixtures at different temperatures (K) 1) 883; 2)

923; 3) 943 [94]

145
0.9

 

( mI;IK )0,5

0.3

0.9

0.5

 

 

 

 

 

 

L UCL b 60%
20% UG L LO% uclL, % ucy,

. 8% RbClL - 60% RbCL.. . . . 0% RbCL .

!
" ' — —

.____.__._o — ——— —— - ,.,__ R

I )

. I .
X

el - - - X - —_— e
sy Lo ‘~—--—-.——--—l--—“'} pm—— 1 —
4LCO 7C0 400 700 400 700

80% CsCl 60% CsCL _ LO% CsCl ]
- - - - ‘ e - b o
T T e
— — - . —X _—

B X e
A ! ! / t ] ' [ p b e
400 700 400 700 300 700

Temperature (Y) of Salt Mixture

Figure 3.3-14 Comparisan of Thecretical

Legend

{k} with Russian Ex—
periments on RbCL/UCL , and CsC1/UCL , Mixtures

. experiment [ 94 ]

X Gambill prediction at Tmp

o Gambill prediction at Tmp corrected according
to individual salt data = (Fig. 3.3-11)

146
3.3.1.9 Summary, ‘e seek maximum 233U content and hard spectrum in
the fuel salt ard a softer spectrum in the outer blanket.

The CRNL MSR program shows that fuel salt circulation is feasible:
It ié facilitated in part by lowering of melting points in |
eutectic muilti-component mixtures. Chemical stability and
corrosion of the molten salts are fairly predictable.

Heat transfer depends on fuel sait density (/:) thermal
conductivity (k), viscosity Q/()- and specific heat (Cp).
Ip(T),/pL(T), and Cp (Tm.p.) are well known for single salts;
much less so for salt mixtures. Knowledge bf k (T for mizture is
better, but still skimpy. HMcjor uncertainties in predicting h. for 2

nixzture should stem from all bui/g(T).

147
Ja3+2 Choice of Halogen

This study centers on Cl and F for which much literature exists,
Br and I are also possibilities which we peruse only superficially;
economics, availability, and salt stability probably preclude their
use,
3434241 Neutron moderating effect, The halogen will be the principal
*light" material of the systeme Thus chlorine (4=36) will produce a
much harder neutron spectrum than fluorine (A=19), satisfying a
principal core guideline, A soft neutron spectrum in the blanket may
enhance BG,

Je2e¢2+2 Neuiron absorpiion, [ absorbs sigmificantly less than Cl

 

(Table 3,3-VI), which further recommends F for the blanket, Numerous
workers have suggested enriching the chlorine in 3701 as 3501 contrib-

utes most to absorption at thermal energies, However, despite the

37

fact that ~ 'ClL has the low level density of a magic nucleus (=20

neutrons), its Se~wave strength function (fl:/D) and gamma width fl,will
be near the same at fast neutron energies. Thus no immediate evidence

exists that enrichment in will reduce the fast neutron absorption.

Table 3,3-VI Halogen Neutron Absorption Cross Section

Spectrum-averaged Absorption Cross Section (mb)
Typical 200 cm

Halogen Tincal Core Blanket
F De3 4e3
Cl 17 9e5

148
3.3.2.3 Transmutaticn products. One might still enrich 3 7Cl S0 &as
to avoid producing sulfur and phosphorus (Figure 3.3-15). Enrichment
would be relatively easy because '

1. 3761 has a (high) natural abundance of 24.5% to start with

2. The weight ratio for separation, 37=35 =0.057 is relatively
35

large
3. Chlorine is gaseous at room temperature
Meither 1‘91“, 35 Cl, nor 3 7Cl transmute into long-lived muclides.
The chlorine isotopes - create a stroanger short-term radiation source
than 1917', but the usual Ber, and LiF camponents in a fluorine salt

6 L . .
mixture procuce 05, (1.6 x 10 years) and tritium (i2.3 years).

BY
1o (100%) 19 (n,x) 16N~ -,‘E—;a-»mo

‘ : o 18
2. (100%) 19g (n,2n) 185 %*o

3. (100%) P (n,y) °OF LY 20y,

11 s
35 35 -
4. (15.771%) °°c1 (a,p) “75 _A 35
17 U4 T T A
N 35 32, 4
5¢  [75.TT5) cl (nye) °°P / o
17 s T s

6. (24.23%) M (n,Y)3801 E'Y 38yr
17

3Tm

Te  (75.77%) 3501 (n, 2n) 345 é’b’ 34
32 m

Fize 3e3=15 Principal Neutron-Induced Reactions of the Candidate

Halogens Which Produce Impurities and Radiocactivity

149
3.3.2.4 Melting point [977. Salts which melt above room temperature
require preheating to start up a power plant: +the higher the meliing
voint, the greater the

1, Heating requirements,

2, Difficul*y of plant operation,

}. Poszibnility of freeze;up during plant shutdowm,

Chlorides stay liquid at much lower temperatures than fluorides.
For actinides (Fig.‘3.3-16) and alkaline earths (Pige 3¢3=17), the
differences reach 400-700°C; for alkalies (Fig, 3.3-17), only 100°C.
Futectic mixes behave simila-ly (Figures 3.3-18 and 3,3~19). This

discourazes the use oif fluorzdes in the bvlanket,

    

1600

1200

800

M.P. Temperature (°c)

 

400 =T

Atomic Number Z of the Salt

Cation- (¥, Cl, Br, I)

Fig. 3.316 Halide Melting Points for Pertinent Actinides

150
1LC0

 

  
 
 
 
  
    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1200

 

 

SRR @1

chemical Earths

Eof 42 chemi-

 

 

 

 

 

e B P s ats
i clscbrn b b et : ;f r .\_\_ TV
il rmeom e CmE T e ]
sl Alkalies S Alkaline ¢

 

 

 

 

 

‘ -t cal state) 5_:_::3'

Sl A e

Sy N

i 2 L TR T
00 |— T e T

B =i/ EES :

Tt T

 

 

 

 
 
 
 
 

 

 

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

= S b

TR [

 

 

 

 

 

 

 

Methng Point Tempevafure (OC)
¥

 

 

 

 

400 o 0 40 o 40 8O

Atomic Number Z of the Salt Cation

Figure 3.3-17. el-inz Points for Ealides of Alkalies
(Li. ¥e, X, Rb, Cs) and Alkaline Zarths
(8e, Mg, Ca, Sr, Ba),.

“00‘1 “004

UF*

1

900 A 900 4

100 " TDO d

500 - 500 |

  

Melting Point Temperature (*c)

 

 

Mc.Hn.r\z Ponit Temperatuve (°c)

 

 

 

. 300 - r v 200
o. oS [0 0.

Mole Fraction

 

o5 O
Mole Fraction

Fige 3 03""1.8 Fig . 3,3-19

Typical Eutectic Melting Points

Typical Eutectic Melting Points
for Chloride Mixes

for Fluoride Mixes

151
5r and I also show (Figure 3.3-16) systematically lower melting

points, but availability and physiochemical properties may limit them.

Section 3.3~1 menticned how eutectic termary fiixtures can lowver
the melting point even further. However secticn 3.3.4 shows that

significant lowering only comes with low actinide molar content.

3.3.2.5 Boiling point and vapor pressure. Figure 3.3-70 shows how
the vapor pressure for metal chlorides depends on temperature; boil-
ing occurs where the vapor pressure reaches ane atmosphere. Core
salts with high boiling points will emit low vapor pressure at the
reactor operating temperatures. Tbis means low system pressure and
enhanced safety. Depending cn corrosicn rates, it may also permit
the use of thin-walled (more efficient) heat—exchanger tubes. The
higher the boiling point, the greater the uliimate temperatures
which may be employed, Boiling points above 1500-1600°C would
allow maximum high temperature operation with graphite, Thus one
desires a carrier salt with low melting voint and hizh boilineg ovoint,
Since the eutectic behavior applies only to the melting noints
these zoals are not as contradictory as they first seem.

Comparing the boiling points of chlorides and fluorides,
(Fige 3¢3-21) one concludes that generally the fluorides are less
volatile, especially for alkaline earths, Unfortunately the higher
melting points and moderating properties of the fluorides discourage
their use in the core., Should vapor pressure at desired itemperatures

become 2 dominating problem, they could provide relief,

152
 

 

   

 

10
MoCl5
268
] AICL,  ZrCla .
o 33:'1i= CrCls  NiCla MnCle ZrCls CuyC! 1750
TiCla NBCLS / FeCly [e1s (g7  [nso 1207 1490 L_ 1780
' 136 1 \265/ /33 N 1026 N, AL ues N 1E2T~
- < . - i. .
7 // ~
/// Q"M '/ -
7 ' . ~
n
=2
10+
— /
= 7/
-4 o
05+
i
4 =
>
w
wn
6 W .
104 & Maxmum
[+
' fuel temperature
jn this reactor
-4
10 -
)
-1
10

 

 

200 400 600 800 1000 1200 1400 1600
TemperAaTURE  [°C]

Figure 3.3-20 Vappr Pressure for Metal Chlorides [17]

153
 

 

 

 

2400

 

 

 

 

 

 

 

 

 

 

 

 

2000

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

On
(&
o

 

 

 

 

 

 

1200

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

800

 

 

 

 

B, I’ Temperature (oc)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Toimeme Alkaldes =mmmmimine sy T} Earths
o e ("1 chemi e == '_:r'fi.‘.‘ St (+2 chemi- 7
P TIEE T T 0y state) o i e o e T e . )
S S e e oAl state) =
S e et e e e e e e e e e
400 {=isor= i 1= "—r —o ==t B

 

 

 

 

 

 

 

 

    

Atomic Number Z of the Salt Cation

Fige 343=21 Boiling Points for Halides of Alkalies (Li, Na, ¥, Rb,

Cs) and Alkaline Barths (3e, Mg, Ca, Sr, Bz) [97]

154
3.3.2.6 Chemical behavior. This section surveys the chemical
stability of the halides in the system (tendency to break up) and
their corrosiveness (tendency to chemically attack the structural

materials in the primary circuit at the temperature of concern),

Figures 3.3-22 compares the stability of some chlofide and fluo-—
ride compcunds: the lower the A G, the more stable. NaCl end iir
ere particuiariy good. In general chlorides are less stable than
fluofides. This pattern of decreasing stability with nalide atomic
number continues through bromine and iodine.

The smell AC values for MoCl, and MoCly inhibit their fermatien,
wiiicn is why Ho is a preferrei siructural materiat in a chloride
system. Graphite also resists chemical attack (fcrma%ion of CF, and
ocL,) well. The appreciable AC for halides of Fe, Cr, and Ni makes
them unsuitable as structural material in contact with salt.

Chlorides generally dissalve in H,05 fluorides do not. Inside
the molten salt circuit the presence of oxygen or water will produce
oxy—chlorides (CCl) which corrode. Cutside the reactor, water sclu-
bility may facilitate cleanup of leaks or spills.

AG for other halides is less available; they are generally less

’

stable,

155
961

Hymerical Yaluea givan for 40

1100 4

Fuel
Componenta

-100
nalBOOK
-IOOfl
(KJInolF)
-370 BaF
L up
§
-UF!
- b= -
%00 ?UFB, ;nP“

-5004

 

-€00

~ LIP

 

L0Gux in

Struotural
Matarials

HBP“j

b iLe
Crf, I~ Fef,

C'P5fi

 

ZLLP-ae?z 1oiucion

Plsslon
?roaucta

aa?3 i fiuP:
JDPu 1F Tep
3
zre,
ciPp 4 .
w Lmtt
[~ erp
Laly, -4 UOF’
Ser -1

 

 

Flgure 3.3-22, Free Energy of Formation

+ 100 +

-100 {

6ulOOOK

-200 +

(xJ/moiC )

=300

-hoo 4

-5C0

 

=b0u

Puel
Components

| ucy,

ThClH e,

—
w

PuClylpact

'Amcl)

—

NaCL <

 

Secuctural
Materials

noc1z1t“°011
[ cciy

FP:Clz
CrC12~

 

Pission
Producta

TcCIJ*

Snc12,F

CeCl, |
0T
LlLl’j

SeC1, 4
Hact, "}

 

PdCl2

SeCl
TeClz

irCk

PrCi

SmCl2
RuCl
CacCl

 

at 1000 ¥ for Fluorides and Chllorides [17]
3e3e2s] Cost and availability. The abundance of seawater provides
obvious cost and availability advantages for chlorine over Zluworine,
The scarcity and cost of Be for a F/Li/Be mix (necessary to achieve

low F melting temperatures) tilts the scale even more in favor of Cl.
3e3e2.8 Density, The power needed for pumping the core salt depends
directly on its density; heat transfer <z the other hand increases with

/00.8

(O.BBg/cmB), much less, Typical densities revorted for salts in the

o Thus, lead (10.5g/cm3) requires Aigh pumping power; sodium

fluoride HSBR and chloride IMCFR (Pu) concepts are 3.2 and 3.1g/cm3
respectively, indiceting little difference due to halide choice,
le3eZe9 Summarys. In the core Cl vermits z hard neutron spectrum; ¥
does not, Heavier halogens zre less well studied, scarcer, and un-
necessary. I excels only in lower vapor pressure, Ifi the blanket,
chloride costs less, melis lower, and ic more chemically stzble, but
fluorine, which moderates better and absorbs fewer neutrons, could
considerably enhance breeding in the blanket, The main problem with F

there is high melting point,

157
3.3.3 Choice of Actinide Chemical States{Stoichiometry)
This section assumes the use of Cl salts in the core, and Cl or

F salts in the blatket.

3,3.3.1 Actinides inherent to an MCFR(Th). Since only thorium crosses
the plant boundary as feed replacement material, the principal actin-
ides in the system will be Th and U isctopes with Ae232-234. Pa, Np,
and Pu should total less than 1% (section 3.5.8.1). Pu starter fuel
in early reactors will produce some im and Cm but {hese fission use-

fully. Later reactors'wiil,Stdrt up on bred 233U.

3.3.3.2 Melting and boiling points. In Figure 3.3-16 the meltirnz
points for actinides in the -4 (IV) state clearly lie belcw those in
the (ITI) state. Eutectic mixtures tend to follow this frend. This
recommends the (IV) state for the blanket which geneéates less heat
and tolerates moderater much better (Sec. 3.3.1.1).

Few salt boiling points have been measured. As ITII-state melting
points exceed those of the IV state, boiling points should also. This

recommends the (III) state for the core salt.

3.3.3.3 Chemical benavior. UClA and UCl3 can coexist in equilibrium.
In the core some UC_L3 wirl picl up free chlarine to form UClh. By
itselfl UClh corrodes, but not when UCl3 predominates (i.e. when the
stoichiometry is near 3 Cl atoms to 1 U atom). ThClh may be present
in the core as a critical mass diluent or negative shimming agent, or
simpiy be carried over in blanket reprocessing. It can also coexist
with UCl3 without any deleterious effect.

Corrosion also increases with temperature. Temperature in the

blanket should be fairly low unless other conditions warrant a high

158
melting temperature.

3.3.3.4 Choice between UC.‘L3 and UCl‘E+ in the core. Hard spectrum,
BG potential, and corrosion should mainiy dictate the core com-
position. The higher uranium density and beiling point and

lower corrosiveness of UCl3 far outweigh the lower melting point

of UClh.

3.3.3.5 Chemical states in the blanket. As a chloride, Th exists
only as ThClh and can coexist with UCl, or UCl,. UCL, helps lower

3 L 4
the melting point in the blanket where little heat is generated.

159
3.3.L Choice of Carrier Salt Cation

Elements which promise chemical stability and high boiling point
(Fig. 3.3-21) as a halide carrier salt include the alkalis Li, Na and
K, and the alkaline earths Be, Mg, and Ca. Inadequate data limit the
studies of the heavier alkalies and alkaline earths Rb, Cs, Sr, and

Ba. Pb is included because of interest generated by British studies.

3.3.4.1 Neutron moderating effect. The average lethargy increment
per collisionfE measures the effectiveness cf various materials as
moderators. At most neutron energies %dapproximates the group—-
\ise elastic downscatter cross sectiem ( qdis the elastic scatter
crcss secticn and YU is the group lethargy width). The evidence
from § (Figure 3.3-23) and 557 /6w (Table 3.3-VII) pretsy uck
agree: L1 and Be moderzte very strongly; Na and Mg, intermediate; K

and Ca downscatter the least; Rb, Cs, Sr, Ba, and Pb should also.

19

Fig. 3.5-23 "1 1
Variation of é y

the average lethargy

 

increment per colli-~

0.07 J

 

gjion wvith atomic num~

 

o v f lI IJ fl_l; 1l “:, _T' N e o2

ber Z. Atomic Number, 2

160
19t

Group

1

o 3 o v - jeL N

10
11

12

13
14

Table

Neutron Energy

6.5 = 10.5 MeV

4.0 ~ 6.5
2.5 ~ 4.0
1.4 = 2.5
0.8 = 1.4
0.4 - 0,8
0.2 - 0.4
0.1 - 0.2
16.5-100
21.5-46.5
10.0-21.5
4.65-10.0
2.15-4.65
1.0 - 2.15

Reactor spactrum average
Fission spectrum averuge

MeV
MeV
MeV
MeV
MeV
MeV
MeV
keV
keV
keV
keV
keV

keV

3.3-VII FElastic downscatter cross sections [98]

Li-7
0.746
1.00
0.910
0.613
0.642
0.532
1.474
0.379
0.337
0.357
0.357
0.367
0.371
0374

0. 727
0.744

Cross Sections (b) by Material

Be
0.373
0.410
0.832
0.565
0.971
1.12
1.23
1.51
1.51
1.57
1.60

1.63

1.63

1.63

14017
0.832

Ng.
0.244
0.230
0.2,8
024
0.350

0.510

0.475
0.1460
0.582
0.472
0.549
0.878
10.97

0.680

0.398
0e325

Mg
0.120
0.159
0,160

0,226

0,304,

0.400
0.634
0.547
J.895
0.421
0.421
0.369
0.369
0.369

0.416
0,297

K
0,146
0.196
0.241
0.191
0.135
0.116
0,147
0.174
0.168
0.179
0.130
0.240
0.275
0.131

0,159
0,175

Ca
0.148
0.212
0.269
0.184
0.231
0.175
0.192
O.144
0.096
0.096
0.121
0.160
0.155
0,192

0.192
04209

MCFR

0.501
3.18
8.46
17.15
19.92
22.03
14442
7.86
L.00
1.48
0.778
0.133
0.0372

0.011

Fission

1.7
9.2
18.6
27.
20.
1h.
6.
2.3
0.9
0«3
0.1
The choice of C1 (Ax=36) as the core halide (Section 3.3.2) en-
courages cheoosing a carrier salt caticn (CSC) with A236: e.z. X or
Ca. This is not crucial, however, as Cl well ocutnumbers CSC in the salt
mix?

Inelastic scattering also degrades the neutreon spectrume. The
fissicn and reactor spectrum averages in Table 3.3-VIII reccmmend X

end Ca the mest (again); Pb and ¥a, the least.

3.,3.L.2 Neutron absorpticn. The cation cress sections all measure
just millibarns (Table 3.3=I¥); the anions Cl and F (Table 3.3~VI)
abscrb neutrons much more. cStall i’ff for X, L1, and Be exceed thcse

for Pb, Mg, Na, and Ca. by magnitudes.

3.3.4.3 Transmtaticn products. Neutron absorption can transmute the
C5C into markedly-different products. The law of energy conservation
fixes which channels the compound nucleus can decay into., All (n, ¥ )
reactions are possible. Table 3.3-X searches for others which have
positive % =values. Fig. 3.3-24 lists the cnes which also yield sig=-
nificant transmitation productse

Li transmutes into tritium and helium, leaving a surplus of halogen.
Be activates to (1.6 x 106 year) 1OBe.’ Na transforms into Mg, changing
the chemical valency of the salt system, The very fast nevtron spec-

trum enhances (n, 2n) production of 2.6-year-rédioa.ctive 22Na.. Neutron

. 6 .
absorpticn by 2 Mg (11% abundance) produces Al, resulting in the chem—

162
€91

Grou

n

N o W

Table 3,3~VIII,

Meutron Linergy
6.5 — 10.5 MeV

4,0 - &.5 MeV
2.5 = 4.0 MeV
1.4 = 2.5 MeV
0.8 — 1.4 MeV
0.4 = 0.8 MeV

0.2 = 0.4 MeV

FS(%
1.7
9.2

i8.6

27.

20

14,

0. 50.
3.2
8.5
17..

20,

22.

Fission Spectrum (FS) Average

Reactor I'lux Spectrum (RFS3)

Avcrage

Comparison of Inelastic Cross Sactions [98]

Croas Sections (barns)

 

IJi"E

0.40
0.30
0.25
0.25
0.16
0.03

0.

0.185

0.114

Be

0.054

ka
0.65
0.65
0.65
0.58

0.48

0.13
0.

0.472
0.303

[;';-&_

0.84
0.85
0.75
0..,0

0.01

0.

O.

0.80
0.50
0.16
0.01

0.

0.

0.

Ca Fb

1.15 2,50
0.76 2.10
0.05 1023
0. 0655
O. 0429
0. 0,01
0. O

0.342 0.092 0.0990.67

0.165 0.035 0.034 0.34

Pb; lla » Mg> Li > (Be, K, Ca)
ol

Table 3.3-IX Neutron Capture Comparisan of Candidate Carrier Salt Elements [98]

 

 

 

 

 

IFisgion Core
Infinitely-Dilute Cross Section (barns) for . _ Spectrum Spectrum
Croup Neutron Energy Be Na - Mg K Ca Li=6 b () (%)
1 6.5 -~ 10.5 MeV 0.030 0.050 0.060 0.330 0.02 2.06 0,000 1.7 0.501
2 LeO = 6,5 0.070 0.005 0.003 0.240 0.01 0.10 0,000 9.2 3.18
3 2.5 = L0 0.095 0.0002 0.0002 0.150 0.00 0.16 0,001 18.6 8.46
L 1.4 - 2.5 0.04,0 0.0002 0.0003 0.065 0,00 0.25 0,001 27« 17.15
5 0.8 ~ 1.4 0.003 0.,0002 0.0004 0,025 0.30  0.003 20. 19.92
6 0.4 = 0.8 0,000 0,0003 0.0004 0.003 0«50 04004 14. 22,03
7 0.2 = 0.4 0.0006 0.0003 0.00} 2.00 0,006 6, 14.42
8 0.1 =02 Y 0.0012 0.0040 0,006 0.95 0,006 2.3 7.86
9 465 = 100 keV 0.0016 0.0005 0.009 0.70 0,005 0.9 4.0
10 21.5 = 46.5 0.0026 0.0005 0.014 0.85 0,004 0.3 448
11 10 = 26.5 0.001 0.0005 0.005 1.20 0,002 0.1 - 0.778
12 L,65~10 0.001 0.0001 0.033 1.0 0,001 0. 0.133
13 2.15=4.65 0,100 0.0002 0.200 2.60  0.001 0. 0.0372
14 1.0 =2.15 ¢ 0,000 0.0003 0.009 3.90 0,001 0. 0.0110
15 465 =1000 eV : 0.005 0.0004 0.013 + 5,70 0.001 0. 1.72-3
FSA(b 0.036 0.0016 0.0017 0.079 0.0013  0.029 0.0022
csaibg 0.018 0,0009 0.0010 0.041 0.0004 0,046 0,0035

Canclusion: (Li,K) > Be > ( Pb, lig, Na, Ca)

* in natural lithlum
Table 3.3-% @ =-Values for Scme Neutron Reacticns with Candidate

Alkali and Alkaline Earth Isotopes [99]

 

Kat . Q-value (MeV) for

Target Abund. (n,2n) (n,p) (n,o )
Isotoe ()  Reaction  Reactin  Redotim
b 7.5 5.7 2.7 4.8
i 92.5 7.3 -10.4 unknown
%Re 100. -1.7 . -12.8 -0.6
“Na 0 ~11.1 €3.6 +1.95
23N a 100 —12.4 -3.6 3.9
2hg 79 -16.5 L7 -2.6
Mg 10 =7.3 -3.1 +0.48
25y 11 ~11.1 -7.7 =54
3% 93 -13.1 +0.22 +1.36
% 0.012 7.8 +2.3 +3.9
bl 6.7 -10. -1.7 ~0.10
k0ca 97 ~15.6 0.5 £1.75
tha 0 ~8.3 +1.2 +5.2L
Lca 0.65 -11.5 -2.7 +0.35
b3ca 0.14 ~7.9 ~1.0 +2.3
bhoq 2.1 -11.1 —4.5 2.7
4ca 0.003 ~10.4 6.9 unknown
4ca 0.15 9.9 unknown unknown

165
Table 3.3=% (Continued)

 

Nat, O=value (MeV) for
Target  Abunde (n,2n) (n,p) (ny A )
Isotove (%) . Reactian Reaction Reaction
85 72,17 ~10.5 0.11 0.99
8Tab 27.83 -9.9 3.1 —1.2
Bl 0.56 ~11.8 -0.10 0.68
&gy, 9.9 ~11.5 ~0.99 1.13
&sr 7.0 -84 0.51 3.21
8881' 82.6 -11.1 —lyeli6 ~0.76
Blhes 2.0y 6.9 2.1 6.2
B3cs 100 ~9.0 0.35 L.38
DG, 0.10 ~10.3 0.34 6,76
B2, 0,095 ~9.6 ~0.1,0 5,96
BLB& 201-(» "'9.3 -1028 2-21
B, 6.5 7.2 0.57 6,94
1365, 7.8 ~9.2 -2.0 o2l
B%, 7.9 8.5 “4.0 3.77
0Ly, 1.4 -8.2 0.02 8.20
206p,, 2401 -8.1 ~0.74 7.12
207py 22.1 6.7 ~0.65 7.89
208y, 5244 7.4 .2 6.05
2
3
Lo
5.
e

Te

Fe

(75%)  CLi(n, o)t fi%fl—dae
(92.5%) Lil(n, Y)2 “Ee
6
(100%) “Be(n, ¥ )03 lebxl10 ¥ . 105
3
(1003%) 23Na(n, Y)24Na 1-2—;—1‘_?—»241@
" (n, 2n)22Na gif}igl-i'-gzNe

(11.01 73)26Mg(n, \(,)27,145' 2:/35_{’.“&2_*‘27 AL

(9343%) 39K(n,p)39A §§25§;~;-39K

/

5
" (n, )30 ZTEIEL 36, 25
AT
(6475%) 41K(n,\/)42K l2hr 4%,

/’d—

Figure 3,3=24 Priccipal Weutron-induced Reactions Which Produce

Impurities and Radioactivity from the Candidate

Cations

Netural abundances are in parentheses

167
10,

11.

124

13

14.

15«

16,

17

184

19.

204

21,

224

0 d
(96494%) 4OCa(n,o< )37Ar _3_4_8_.....3701

8
(72.17%) 2930 (n,20) %m0 4L 2% o

€\p Y, 8
i (1, )36 18265 486
? (fib/

e e BB %y

(27.83%) 87Rb(n,2_n) Rb 18,854 5% Kr 89.1,

Ve
F ,a’r 504 5d

8g' -
C e e )G, e B
/ . ‘ /

 

" (n, p)87KI' 76 m 87

Rb
Sy

 

" (n,N)BA'Br 31:8 T - 843'.1‘
&

(82.6%) Bsp(a n, ¥)%%zr(a, v )% 2L -

¥
504 5d /?
1 13 d 6
(1och)  '33cs(n, Y ) 34 s(2, ) 3%¢s(a, ¥) 20cs -/—;—?-- 35,
3 ;
AE206y ' 23x106y
" (n,2n)!3%Cs ——-—-—-—6 474, 132z, ¢ 1325,
/2
0" (n, p)133){e _2%\/_(1_._13305

1€8
23

2.

25.

26.

27 .

28,

29,

30.

31.

(roag) Pies (m, &) P e P
4 r

(11.2%) B37pa (n,p) B7ag 2017, U7,
pY

(71.9%) 38a (n,Y) P%a B2 ming U9,
P Y

v o
. (nyal) 135Xe _2_'3___, 135Cs
Py

(1.4%) 20kpy, (n,od) 201Hg
206
(24.1%) 7 Pb (n, k) 2Py i‘i—??—» 2033,

(22.1%) 2Teb (n,o) Vg

(52.4%) 2%Cpb (n,«) POhg 220 o OO
/e Y

169
ically-aggressive AlCl3 as well as a small change in valeacy. Trans—
mutaticn of K produces long-lived radioactivity and divalent calcium.
The magic proton number of Ca causes the existence of numerous stable
and quasistable Ca isotopes. These mostly transmute into themselves,
yielding nil radioactivity.

Rb produces long~lived SSKr and 9OSr plus considerable short-lived
radicactivity and different elements (especially Sr). Sr irradiation

90

principally leads to * Sr.

Cs tramsmutaticn products appear innocucus. Ba will directly
produce 13705. Protm-magic b will generally produce stable nuclei
but highly-exothermic (n, ) reactions will yield chemically different

Tl and Hg.

3.3.5.4 Melting point. This section relies on the limifed avazilable
fusibility diagrams for salt mixtures [87, 100-322]. Most of these
come from terse articles in remote Russian journals; some disagree with
one another, This limits us to only qualitative conclusions.

Low core salt melfing-point helps avoid high ocutlet temperature
(e.ge > 1000 C). Figure 3.3-25 displays the approximate melting points
for UCl3 mixes. Alkali chlorides tend to form lower temperature eutec-
tics with UCl3 than do alkaline earth chlorides., The bumps signal com-
plex molecules like K3U016 (3KC1'U013). Desyatnik et al [100] verify
that complex uranium groupings increase in frequency as the alkali-

cation radius increases,

170
 

 

 

BeClz ¢ UCY,
Se(]
oY
800 - &%
T00 - $
N\ /
- 1/,
T
+2
5
2, I\ ‘
=
=
2
500 - ‘
0 20 40 60 8b 100
lolar Pez_:'cent Ucl, in IrIJ:rture

3

Fig. 3.3~25 Pusion Diagrams for Binary UC

1T

l3 Salt Mixtures
Many of the eutectic melting points occur at lowUCl3 molar content.

In Contrast; hard spectrum and high BG potential favor high actinide

cantent.

Table 3.3-XI shows the possible UCl3 molar contents under various
temperature restrictions. RbCL excels for a SSOOC limit; KCl or RbCL
for 600°C. At 650°C all alkali chlorides show potential, though KCl
and CsCl excel. Alkaline earth chlorides don't work until the low temp-—

erature limit exceeds 7OOOC.

If one wents a high~actinide 30 NaCl/70 UCl; mix then the lowest
core salt temperature allowab.e (with a little safety margin) is 7500C.
With a 50/50 NaCl/UCl3 mix th: temperature can drop to nominally 650°C.

Section 3.3.1.3 mentioned that ternary and higher compoment mix~
tures can further lower the melting point. Figure 3.3-26 presents four
ternary diagrams for UClB-UClthkCl, vhere Ak=alkali or alkaline earth.
Cne desires to optimize the U()lB/'UCll+ ratio so as to hald down temperature
(which affects corrosion) as well as the content of corrosive UClh.
Keeping the U/Ak ratio high maximizes spectrum hardness and BG potential.
Unfortunately, the isotherms and the eutectic mep. tend awey from hizh
gg&B cantent., Ternary mixtures with two AkCl components behave similarly
(Fig. 3.3-27). In summary one must tradeoff low melting point (low
U-~III content) and corrosion (high Ak content) against large spectrum

hardness and BG potential (high U~III content).

172
Table 3.3-XI. Candidate Core Salt Compositions Under Varicus

Temperature Restrictions

 

 

 

 

Lowest Temp. UcCl Lowest Temp. UClL
to be En=- Molgr to be fn- Molgr
countered Carrier Content countered Carrier Cantent
_ (%) Salt (%) (°¢) Salt (%)
500 RbCl ~15 700 LiCl 0=560
530 LiCl ~25 NaCl 18~60
NaCl ~ 35 KC1 1068
CsCl 25 RbC1 5-70
RbCl 1545 CsCl LO=70
| MgCl,  ~ 38
600 LiCl 10-40 CaCl,  25-55
NaCl 30=45 SrCl,  60-70
KCl 50~60
RbCL 15-55 750 1iCl o-70
CsCL ~55 NaCl 1275
KC1 6=75
650 LiCl 050 | RbCl 0=75
NaCl 20-55 CsCl 0=75
KC1 15-65 MgCl, 0-50
RbC1 8-40 CaCl,  15-65
CsCl L5565 Srcl,  50-80
CaCl, BaCl,  70-75

 

 

 

173
 

 

 

] 2.'-'1(:( - LL:I &

 
   
 

358
e,

\

 
  

 

335"

52 O Ul

  

 

 

Fig . 3 -3‘-26

 

 

Ternary UClB/UClL Fusibility Diagrams

174
 

 

The blanket salt will generate little heat; therefore, a low m.p.

1 temperature will facilitate its handling. High ThCl, content maximizes
BG there. Carrier-salts whose eutectic mixtures with ThClI+ (Fig. 3.3~
28) seem to qualify are PbCl,, LiCl, NaCl, and KCl. They also form
suitable eutectics with the small amownts of UClh present (Fig. 3.3-29).
PbCl2 has a sitight disadvantage in that Pb is divalent, further in-
creasing parasitic capture by chlorine.

Under an arbitrary operating design limit of ane U atom per 10
Th atoms and low Ak content (to avoid blanket power cooling and BG loss)
ternary UClA/ThClh/AkCl (Fig. 3.3-30) mixtures do not significantly
lower the meiting point: all eutectic points lie at high UClh and

carrier salt contents,

175
Melting Poin* Temperature (°c)

900

800

700

600

500

4C0

300

 

 

 

 

1

20 40

Molar Percent ThCl

4

60

in Mixture

80

100

Fig. 3.3-28 Fusion Diagrams for Binary ThCl4 Salt Systems

176
Melting Point (°0)

700~

600~

 

S

_,,,
\oiififir'

   
    

 

 

 

50074 2
% 0\
N
1
400~
300 : : ; :
0 20 40 60 80 100

Molar Percent Actinide Chloride

ThCly

UCl,

Figure 3,3-29, Pusion Diagrams for Binary Mixtures of Ac(IV)

and Carrier Chloride Salts

177
 

 

 

 

 

 

 

Fige 3.3-30 Ternary UCl h/ThGlL Fusibility Diagrams

178
 

Addition of a second carrier salt does lower Tmp (down to 325-

T

335° ¢) at ~ 20% ThCl, content (Fig. 343=31); however, though 'Li and

4
Pb may not be too absorbing, the high Cl/Th ratio will damage BG and

neutron spectrum,.

Lict

 

 

Fig. 3¢3- 31. ThCl 4/1,101/?11012 Fusibility Diagram

Fluoride blanket salt presents another altermative, TLip (riz.
3.3-32) could best lower the molting point and thermalize the blanket
neutron spectrum without parasitically capturing neutrons. NaP looks
next best, However fluorides solidify near TOOOC and eutectic effects

do not appear until ThF, molar content falls disastrously (to BG) low

4
to 200 (Fige 3e3-32); chlorides melt much lower.

119
Melting Point Temperature (OC)

1100

1000

900

800

700

600

 

500 A

400 -+

300

 

 

 

T v ¥

0 20 40 60 80 100

Mol % ThF4 in Mixture

Figure 3,3-32, Fusion Diagrams for Binary ThF, Salt Systems

4

180
   
   
 

Fig. 3.3-33
ThF LL/L iF/BeF, and
ThF L/N aF/ZrF L
Fusibility

Diagrams

LF ?YI:FI

LF lfl'hfi

THF‘ s

    
  
 
 

TEMRERATURE iry *C
CUMFOSITION 1N mate 7,

 

 

TEMPERATURES ARE IN *°C
CCMPUSITION IN mow %

Minina 870

 

£E-618
-443 _—
of &S
" Siy
QA /™ N
AN
N w° L ey
g v A
&) R o >
he)
. N i ‘
Naf -~ = ' o ~ ~ 2F,
290 : T %2 a - 92

181
 

The eutectic points which minimize KCl content are 5.5 u013/u5 UClh/
49.5 XC1 for a m.p. of 314°C, or 17 UC1,/26 UCL, /57 XC1 for a mep. of

518°C. With NaCl ane might opt for a m.p. of 339°C at 17 ueL,/

4.5 UCL, /42.5 NaCl or 432°C at 21.5 UCLy/29.5 UCL, /49 NaCl. The CaCl,

ternary offers no attractive eutectic points.

Nane of the above points have UClB/UClh> 1; that could spell
corrosion troubies (Sect. 3.4.2.2). Still, temperature affecis cor-
rosicn as well, so that high UClh courd yet be worthwhiie. Of the
five cases mentioned above the mix with mep. 33900 locks the best.

Ternary mixtures of ThFh/UFh/AkF behave like their chloride
counterpart: no eutectic poimts occur with high ThFh content.
Mixing in two carrier salis with ThF, doesn't help either (Fig.
3,3-33): +the well known MSBR salt Flibe Ber/LiF) combines with
ThF, or UF, %o reach melting points of 400~500°C, but only at 12
ThF4 (for 50000). Flibe also suffers because Be is scarce and Li
produces T unless enriched in 7Li,

Figze Je3=6 showed aigher halides %o have lower melting points,
but they nave other problems,
3u3ele5 3Boilinr noint and vnvor nressurz, Chlorides of Cz, X, la
and Mg all boil in the same range, 1400=1600°C, Their differences
are insignificant because UCl4 (792°C) and ThCl4 (92800) pose much
lower limits. The boiling points of UCl, and PuCl, are unknown,

3 3
but their melting points compare with those of the carrier salts.

182
 

3.3.4.6 Chemical behavior. Halogens react vigorously with the al—
kali metals to form a very stable binary ccmpound. Salts of alkaline
earths also exhibit large free energy for formatien (Fig. 3.3-12).

Fb halides are not so stable: O0BW. [13] found PoCl, 4o react with the

structural materials available then, They zlso found Zrél2 to cause

2 Snowy Ddrecitiiate.

In addition to having high melting points, alkali halides conduct
electricity in the molten state, and otherwise behave like electro-
valent compounds. Uith the excepticn of lithium fluoride, they readily

dissolve in water.

3.3.h7 Cest and availability. NaCl is very abundant and cheap. This
reduces both direct capital and fuel cycle costs; it also allows other

options such as substitution of fresh for radicactive haCl.
3.3.4.8 Density. The densities of the principal (lighter) candidate

carrier salts (Table 3.3-XITI) differ little compared to those of ThCL,

UClB, and UClh.
Due to differences in molecular weight, a 50-50 molar mix of
actinide and carrier salt means about an €515 weight mix, respectively.

Thus the actinide salt density controls the final density.

3,3.,4.9 Heat Transfer Coefficient. Section 3.3.l1.4 pointed cut that
_ 102 -0y A 03B | OFT
h:g /J/ll CF }l
Section 3.3.4.8 showed that the actinide salt controls the density.
Sectiom 3.3.1.6 suggests thafiyu(fuel salt,hixture):/AA(carrier

salt). Fig. 3.3-5 suggests avoiding most of the alkaline earth salts:

183
 

Table 3.3=XII Density (g/cmj) of Chloride Salts at 800°C

 

Salt | Desyztnik et al [83] Kinosz and Heunip (84]
LiCl l.42 1.46
NaCl 1.55 1.61
KC1l 1.52 1.57
RbCli 2.16

CsCl 2.62

PbCl, Lel2

ThCJ.A 3.36

U(.‘.L3 L.87

UClI+ 3.17

BeCl, 1.10

MgCl, 1.65 1.68
CaCl2 2.07 2.11
SrCl2 2.77

BaCl2 3.28

the alkalil chlorides and PbCl 5 offer lox-:ez;/uvalues.

In Table 3.3-IIT1, the lighter chlorides offer the higher specific
heats. In Figure 3.3=9 they alsc exhibit higher thermal canductivities
in the 700-9C0°C range, CsCl is an exceptim.

Ignoringja because the actinide density contrels, a heat transfer

figure—~of-merit (FCM) for carrier salts is CpO'33 k0'67 L.'O°b7 ,

184
 

The results for Section 3.3.1 data (Figure 3.3-34) and the above

discussion recommend the lighter-weight—salts, and all the alkali
chlorides.

A FCY including the density factor‘fp‘a (Fise 3¢3-35) aczin
discourages only the heavier alkaline earth chlorides, CsCl and FhCl,

look very good then.

In summary, LiCl, MgCl2, and NaCl look to be the most attractive,

2

sive,; however, as i1 generallrs does not include the effects cof actinide

in +hz% order. SrCl,. and BaClz}the least. This study is not conclu-
caloride a2dmizture,

3.3.4.10. Choice in other studies. ORNL in 1956 [13] chose a

3 NaCl/ 2 MgC12 mix based on the:eutectic point between these 4wo,
However, as pointed out above, the advantage of a second carrier
salt disappears when the actinide salt predominates.

Similarly, ORNL in 1963 [§] and ANL in 1967 [21] reported
studies on carrier salt mixtu;es of NaCl, KCl andegClz but with
only 30 and 18-50% mole of actinide chloride present, respectively.

Taube [17] generally assumes NaCl alone as carrier salt even
with low actinide contents. He mentions alsc further study of
NaF/ZrF,, but Section 3.3.2.9 precludes F from the core. The

British [29] also choose NaCl by itself.,

185
 

 

- - - l‘ — e ——— e e
_ —— —_— — - - —— - —_ Coae —

e ——— e ——— e — e — —
- ———— e

 

  
  

L C(Jf:fh

chlorides

 

   

|
alkaline

  

 

 

 

 

0.6

0.8
[900°C](ca/g °c)%*3°

p .
x k[850°¢](H/mix) 07

x)u[85000](cp)_°'47

C

186

Molecular Weight
Heat Transfer Coefficient Figure-of-Merit for Chloride Salts of Interest

Figure 3-3"340
T T ke

1.2 ST ©OMglly .

   
 

 

o _ _ S chiorides __ _

 

 

100 _______X NOCI .

   

0.8 T T alkaline earth T T T T T T
oottt T \chiorides T I T

 

 

 

 

 

 

 

 

40 80 120 160 200 240 280

Salt Molecular Weight
Figure 3,3=35. Heat Transfer Coefficient Figfiré-of-Merit for
Chloride Salts of Interest, Including Density

Foll = » [850°c] (/o) *° x C, [900%¢](ce1/5%) "+

x k [850°Cc]W/m )0 €7 x/u,[850°C](cp)"o'47

187
 

3e3+4+t1  Summarve Table 3.3-XIII summarizes the” intercozparison of
carrier salt cations. The goal of keeping the core spectrum hard dis-
courages use of Be and Li there because of elastic downscatter; Ca and
X dovmscatter the least. lNa undergoes the most inelastic scatter; Ca
and ¥, the least. Parasitic neutron capture though small, exceeds by
magnitudes for Li and K over that for Ca, Mg, and Na. Ca transmutes
the least into radioactivity or troublesome chemicals.

Caticn choice on the basis of eutectic melting point depends on
the range of acceptable actinide molar contents. Boiling points far
carrier chlorides olay no role as they far exceed those for actinide
chlerides. . The chenical stability of the 2lxali chlorides (NaCl and
KC.L) surpasses that of the aixaiine earth chiorides (i-lgCl2 and CaClz).
laCl abounds by far the most in nature and costs the least., Salt density
affects pumping power needs but the higher densities of the actinide
chlorides dwarf any differences due to carrier salt choice. The light
alkalis transfer heat the best.

In summary, ¥aCl costs little and exhibits good physical and
chemical properties. K and Ca feature better nuclear properties for
the core sait but the much higher Cl concentration obscures them. Na
and 7Li look best for the blanket sali. Final choicc must weigh these
relative advantages plus the location of eutectic melting points near

the desired actinide molar fraction.

188
 

Table 3.3-XII1.

nuclear properties

elastic scatter
inelastic scatter
neutron absorption
radioactivity

impurity mutants
overall nuclear

physical properties
mop-

bePe Vapor press.
heat transfer coef.

chemical behavior

salt stability

economics

cost
disposal

overall comparison

core
blanket

Intercomparison of Salt Cations

rating according to core salt goals

 

good intermediate poor

K,Ca Na,Mg,Pb Li,Be

K,Ca,B Li,Mg Na,Pb

Mg,Ca,'Li Pb,Be Li,K

Na 9

Ca,Pb g, Li,Cs Li,Be,Na,K
Rb,5r,Ba

Ca,Sr,0s K,Na,Be,Ba Mg,L1,Pb

Ca K,Mg,MNa,Pb Li,Be

K,~a,Li,Pb Mg,Ca Be

al. o.k.

Li, Mg Na,f,Ca Ba,dr,

Pb ,kb,Cs
Na,K,Rb,Cs Mg,Ca,Pb
Sr,Ba
Na Ca,X,Mg,Li,Pb Be

(same as radioactivity)

Na 7 . K,Ca,Mg,Pb,Sr,Ba Li,Be
Na,'Li Mg,Ca,K,Pb Li,Be

139
 

3.3.5 Choice of Helative Proporticns of Actinide and Carrier Salt .

In the core we want to avoid nuclei which neither fission nor breed,
but only dovmscatter and capture neutrans. This means maximizing the
fissile enrdichment and the ratio of actinide-~to—~carrier salt. It is
also important to maximize the fuel salt heat transfer ccefficient at
an average fuel temperature of 800 -85000; hf comprises the parameters

D ., C and k {Section 3.2.5).
'f’ T/A y p! £ ( )
3.3.5.1 Feutren physics of the ThC.LL/‘UClz ratic in the core mixcurs.

Considerable ThCld inherently resides in between the core tubes as
blanket salt 2nd interacts. neutronically with the czore. ~However, as

a first approximation, Sectioa 3,951 ignmores that modellinz heter-

r

1/

ogeneity as well as that of the multiple core tubes. Thus the ThCl
UCL, ratio here refers solely to that within the primary loop.
The addition of ThCla to the core salt mixture should
1. Markediy increase the critical core radius, thereby decreasing
the power density. This causes
a. the core flux level t¢ decrease, which
(1) extends the life ¢f the core tubes, and
(2) decreases the damage flux to the vessel, which ex-
tends its life
be. the primary coolant power density in the heat exchanger to
decrease, which
(1) decreases out-of-core fuel inventory
(2) eases heat exchanger stresses
2. Lead to an optimum Th/U ratio where the primary circuit fis-

sile-inventory is a mizimum. The 1956 ORNL study [13]
 

obtained a similar result
Je Increase internal (core) breeding, which
ae reduces the necessary size of the outer blanket
be reduces the import of a reactor
4. Soften the neutron spectrum, which
2e decreases the BG potantial of the reactor
be 1nhibiis the useful consumption of higher zctinides
ce nullifies z smzll part of the flux decrease from the
dilution because the averege fission cross sections
become smaller,

In sunmary there are both pros and cons re Th presence which are
difficult to quantify on a ccmmon basis., However, Tualitatively, 3G
potential and the useful conzumpiion of the actinide§ argue for low
Th concentration ([Th]) while [Th] ——== O could mean too freguent z
replacement of the core tubes, Practically, reactor operation could
start with high [Th] and wide tubes and later change to low [Th] and
smaller tubes as confidence in the tube wall life develops, This ver-
satility would probably require only the inclusion of a few extra tube
ports in the reactor vessel,
3e3e5+2 Density. This study principally involves U013’ Nall, and
ThClq. NaCl molar content may vary only between 30 and 70% to achieve
eutectic melting point lowering, Pigure 3,3~36 shows the density for
various ThCl4/U013 ratios using the densities and the prescription for
their addition from section 3¢3e1e5e

For a Pu013/2 UCl3/3.16 NaCl mix, Taube and Davudi [123] listed

p =292 g/ce at 850°C; for 15 PuCl3/85 NaCl, 2 = 2.12 gfcc. These
values both seem low compared to Fige 343-36 (though different
actinide salts are involved). In the same repori these authors quote
/cr= 3,526 g/cc for a 1 Pu013/2 UCl3/3.65 ¥aCl mix at 850°C, which
seems nighe

The 1963 ORNL paper [6] listed a density of 3.08 g/cc at 643°C
for a 45 NaCl/25 KCl/3O (PuCl3 + UCl3) mix, Since density decreases
vith increasing temperature, that value supports Figure 3.3-36. The
earlier 1956 ORNL study [13] used /0 = 25 g/cc for 2 3 NaCl, 2 HgClz,

1 (UCl3 + PuCl3) salt, which also seems consistent,

3434563 Viscesitye 5Section 3e31.6 recommended

//% (mi:ture):i/bt(carrier salt)
Assuming JaCl carrier salt, -hen Ifrom Figs 3.3=5 at SBOOC

//LL= 0.012 gn/s ca

For a 1 PuCl3/2 8013/3.65 HaCl mixture, Taube and Dawudi [123]

assumed 2 much higher//%-= 0,05 (1-0,001(1=850°C)). Similarly, at
649°C OMIL used 0,0662 + 0.0080 for a 45 NaCl/25 KC1/30 (UC1, + PuCl,)
salt in 1963 [6] and 0,10 for 2 3 NaCl/2 MgClz/ 1 (UCl3 + PuCl3) salt
in 1956 [13]. However, neither Taube nor ORIL indicate any experi-
mental basis whereas the//%.recommended nerc is basad on recens
(Russian) measurements [124-125]s This reduction in:/utby more than
a factor of three increases hf vy roughly 50% over previous studies,
3e3e5.4 Specific heat, From the study of sfiecific heats for 3-

component salts (Section 3.341.7)

o 8.1 cal/°C (2m2 + 4m4 + 5m5)
c (cal/g C) = X =
P yrieDe atom mole m2M2 + m, M, + m.I

44 55
and Cp (T) = Cp, (-7 (X)]

192
[ThCl4]/[U013]

Molar Ratio

 

 

Density

(g/cc)

 

 

 

 

 

Holar Content (%)

(TaCl, + UCL,)

Figure 3.3-36 Density at 850°C for NaCl/ThCl 1+/UC}.3 Mixtures Based
an Janz 5t .Al1 [112] Data and Linear Addition of

Specific Volumes of Component Salts.
where
m2 = mole fraction of the 2-atom alkali carrier salt
m. = mole fraction of the 4-atom UCl3

m. = mole fraction of the 5-atom ThCl

4

Moyl s Mo = corresponding molecular weighte

0 - -—
cC =¢C 1T K
o= % o /01-=T  (K)]

o = average coefficienti of linear expansion

= Z mil)(i
1

-4 q -
O(; = 34655 4.23, and 5.57 x 10 °¢™" for WaCl, UC1,, 2nd ThCl

4
Figure 3.3-37 shows the results at the Tmp for each material..

For a Tixed temperature, Cp changes much less with molar composition.

Taube and Dawudi {123] assumed c, ; 5.094 and 0,068 x (1-0.0005
(7 -850°C) ) cal/em®°C for 15 PuC13/85 NaCl and 1 PuCly/2 UCL,/3.16
¥aCl mixtures, fairly consistent with Figure 3.3-37. However for a
1 PuC13/2 U013/3.65 ¥aCl mix 2t 850°C they used 2 rather high value of
0s24 cal/gm C. This agrees better with the 1963 ORVL [6] value of
0u3+ 0403 cal/gn’C at 649°C for a 45 ¥aCl/25 KCL/30 (UoL, + Pucl,)
mix and the 1956 ORML [13] valiue of 0.2 cal/an’C at 649°C for a 3 Wacl/
2 Mg012/1 (ucl, + PuClB) salt,

The values used in this work are intermediate to the above extremes,
Although they admittedly contain large uncertainty, at least Section

3e3e1e7 describes their basis; Taube and ORNL do not describe theirs,

194
 

‘Th{l /Uc13

N
0
Lo
o

—

 

NaCl Molar rracvion

Figure 3.3-37 Specific Heat at T, for NaCl/ThClb/UCl3 Mixtures

3e3e5.5 Thermal conductivity.

Section 34341.8 estimated
k(T ) = 0,04308 07 0.8/ 18 ¢ cal/s cm K
o/~ 7* mp /Omp atomic '

k (7) =k (Tmp) + 6.0 x 107 (T—Tmp) MV

and ’

Table XIV shows.the oalculated[k} at T=p_

D

; T=T 4+ 100°C, and T-350°C.
The higher k~values seem to occur for high NaCl and ThCl

contents,
A

195

 
961

 

Table 3.,3-XIV Calculation of Thermal Conductivity for NaCl/‘l‘hClL‘L/L_J_Ql3 Mixtures

 

 

NaCl k ( gg 262%‘
ngizit ThCr%! /Uang Tmp /:n.p : _ a}t at'. Tmp at

( 9) Ratio  (C)  (g/cc) M N v f atomic Top 4 100c  800°C
30 0 740 h37  295.1 3.40 0.132 1.57 1.94 1.97
50 0 610 4.02  198.9 3.00 0.168 1.69 1.93 2 .26
70 0 540 3.35 142.7 2.60 0.204 2.10 2424 2.52
30 0.5 (713) 3.96 263.2  3.633 0.111 1.61 2.03 2.18
50 0.5 (540) 3.85 204.7 3.167 0.153 1.63 1.90 2.46
70 0.5 (502) 3.25 146.2 2,700 0.195 2.00 2415 252
30 1.0  (700)  3.81  267.2  3.750 0.100 1,68 2,12 2434
50 1.0 (505) 3.79 207.5 3.250 0.146 1.61 1.89 2.58
70 1.0 (4,82) 3.21 147 .9 2.750 0.190 1.96 2,12 2453
30 3.0 660 3.66 273.2  3.925 0.0848 1.81 2429 2472
50 3.0 400 3.83  211.8  3.375 0.1343 1.59 1.89 2.95
70 3.0 L25 3.21 150.5 2,825 0.1838 1.89 2.05 2459
Table XV compares some of these {k} (with T-600°C) to those
assumed in other studies, ordered by increasing actinide salt content,
ORML {13] and Taube values seem rather high, Close agreement with the

other |k{ (z2 x 1073

Cal/s cm K) lends credence to the values assumed
in this worxe.
Jele5ab Heat transfer coefficient. Yith the above bases, and assuming

(1) The alzorithm developed for h,. in Section 3¢2.51

f

(2) A salt velocity in the heat erchanger of 14 meters per second

(3) An inside tube dizmeter of 0440 inches

(4) u=
/

A (for these survey purposes),

7 well

my s 3~ < i X - 3

Thne film coefficient for various mixtures and temperatures becomes

h_= T.37 i]A)O'B (10-3)0'67 2 (g/cc)o'8 k (‘IO-3 cal/s cm ¥)

T
Qe dT Oe2
(1.46) x (1.016) < Cp (ca.l/g K)0,3,3.

0.5367
The results (Table 3.3-XVI ) differ very little with mixture or
temperature., Considering the approx:mations made for the input para-~
neters, no significant trend is detected. Roughly
hf = 1,2 cal/sec cm2 X
The only other velue available for comparison is the ORNL [13]

hf = 0438 cal/s cm2 Ke Table 3,3=XVITI shows fthat the factor of 3

increase here is predominantly due to a doubling of fuel wvelocity
and an 8-fold decrease in viscosity. The latter is mostly duve to
more recent information and higher temperature, and partly contingeni

on picking a composition with eutectic-like effects on viscosity

(SEC. 3.3.1.6).

197
 

Table 3.3=XV  Comparison of Calculated Thermal Conductivities to

Those Assumed in Previous MCFR Studies

Entries arranged in order of decreasing carrier salt concentration

 

k
Carrier Actinide (IIO-BCal
Study salt salt (%) sec cm k )
CRKNL/56 [13] 50 KNaCl 17 (OC1, + 649 el
33 MgCl, ' PuCl3)'
ANL/C [21] 82 NaCl 5.4 PuCl,
oot 12.6 UCl3 55205 2.14-
CRlL/é3 [6] 45 WaCl 30 (Puly+ 649 1.7
25 KC1 UClB)
&TL/A [21] 70 NaCl 30 Pucl 682.5 2.1
This work 70 NaCl 30 U013 640 2.2
Taube [17] 55 kaCl 15 PuCL, 850 3.6
30 UCL,
650 2.9
ANL/B [21] 50 NaCl 22.5 PuCl3 615 1.6
27.5 UCLy
This work 50 NaCl 50 UC1, 610 1.7
This work 50 NaCl 25 ThCl 605 1.9
25 UCJ.B“

193
Table 3.3=-XVI  Calculated Heat Transfer Ccefficients

Molar Composition by (cal/sec en? X)

at at

 

 

o 1 a
LhClb/UCIB NaCl Typ + 100 850 C
0 30 1.22 1.22

50 1.20 1.23

70 1.22 1.18

0.5 30 1.17 1.19

50 1.15 1.19

70 1.16 1.11

1.0 30 1.17 1.19

50 1.13 1.18

70 lo1k 1.09

3.0 30 1.20 1.24

50 1.15 1.20

70 1.12 1.06

Table 3,3-XVII Comparison of hf Be-ween This Work and the CRNL/567137 Stud-

199

Effcet of
This work/
Parameter _CRNL 56 [13] This “ork CRNL-56
salt Temp (°¢) 732 (582) —_—
Yolar % NaCl 83 70 —_—
/c(g/cué) 2.5 2.7 ('2—'1)0'8= 1.06
(en/s en) 0.10 0.01L46 (0'012)‘0 47 2.9
c, (cal/gnc) 0.20 0.15 (Efiig)o "33 0.91
-3
10 “cal 12,0. 67
X. (m) Lel 2.12 ( ) 0.61,
N‘(Tfi/S) 6.1 14. (ZA.]_ 0. 8: 1094
20 l-lh
hf(cal/s m C) 0.38 1.14 (ngé) = 3.00
 

3e3.5.7 Choice of core mix. A pure fissile fuel would produce
the hardest neutron spectrum but an MSR requires carrier salt
admixture to reduce the melting point. Adding fertile salt
to the core mix helps reach BG potential by reducing leakage
through increased fertile capture, but it also lowers the BG
potential by softening the core neutron spectrum. Another
consequence will be to increase the critical mass of fissile
kkuel and the core size, This will be good for power density
(Fig. 3.5~15) and in-cors to out-of-core inveniory ratio (for
control purposes), but there are limits to economical plant
size.

An alternative to fertile fuel in the core is to increase
the size of the outer blanket. However, this also increases
the size of the overall reactor system.

The ORNL/56 design limited the actinide concentration to
17 mole precent actinide chloride. Of this, 2/3 of the actinide
was fertile material. The prime motivation was to keep the
liquidus temperature below SOOOC, to avoid what were then exotic
ctructural meterials,

The British [29] chose an actinide molar conceniration of
30-4055 the Fremch [23], 30%,following the earlier lead of LNL [217.
The. readoas for these choices are not clearly stated, However, we
suspect they include ~. . -.

l. Keeping the power density low

2. Keeping down the plant fissile inventory

3. Improving the thermophysical properties.

This study assumed 70% for most of the physics analyses so as to
 

fully explore the advantages of a hard spectrum: power density is
reduced by using a geomeiry with high surface/volume ratio and limited
fertile fuel presence (limited reduction in enrichment), By and large
<hie enrichment problem is ignored because all fuels have weapons capa~
bility but 232U;bearing 233U, the least so, The better thermophysical
properiies from higher carrier salt content are sacrificed for the sake
of the highly~desirable hard spectrum and attendant high BG potential.
3ele5¢8 Choice of blanket mix, Since the blanket generates little
heat, it can avoid high temperatures, This should

1. Preclude corrcsion of the reactor vessel

2. Allow more choices Zor the reactor vessel material

3. Reduce auxiliary hezting needs

For low temperature operation the salt composition should lie
near the eutectic point(s), but on the side high in actinide content
for bigh BG, The French and Swiss MCFR(U/Pu) desigas assume 65% us,

and 353 NaCl; the British, 40% UCl, and 60% NaCl, For a NaCl/UCl

3 3
miz the eutectic point occurs at about 65% NaCl; for NaCl/ThClA, at

about 70% WaCl. Ve assume here a 65% ThCl, and 35% NaCl, XCl, or

4
7LiCl mite Table 31.,3=XIIT tends to favor = fzrther look at 7LiF

for. the blankei, Pb012 is an outside possibility yet,
3.3.,5.9 Butectic mix for flush salt. Section 349.5 proposes flush
operations for rcmoving residual radioactive salt in core and blanket,

The best mix for this purpose would probably be the eutectic composi-

tion of the blanket feed: with the NaCl/ThCl system, any mix between

a
25 and 50% NaCl would be good (Fige 343-22).

201
3¢4 Structural Materials

To find materizls of suitable sirength and endurance one must
anticipate the environment and the response of materials to it,
Environment includes radiation, temperature, pressure, fluid compo-
sition, chemical additives, condensation, and vaporization. Haterial
response depends on mechanical stremgths, metallurgical phases, and
chemical interactions (corrosion).

Requirenents on the material will include suifable thermophysiczl
oroperties, ability to fabricate and weld, product availability, and

existence of pertinent experience for the material,

3.4.,1 feneral Ccpsiderations and Criteria

3.4e1e1 Plant-life economics, Initial study usually suggests several
economical meterizls. The material cost is important, but the {rue
installed cost depends also on size, pipe schedule, system complexity,
joint make-up, fabrication techniques, and labor rates, Dovmiime and
life expectancy, functions of the material response, will affect the
ongoing maintenance costs, Fortunately, metals usually have xmown
corrosion rates,
3.4.1.2 Corrosion, general [80, 126]. The most important engineer-
ing task is likely to be finding materials which resist corrosion by
fuel and blanket salt mixtures containing fission products and other
impuritiess This may require a irade-off with system conditions such
as lowering temperature, decreasing velocity, or removing oxidizers,

Alternately one might sacrifice some corrosion resistance
for higher strength; also, the best material might not be readily

available, The most desirable material will probably combine

202
 

moderate cost with reasonzble life. Sometimes the environment condi-
tions change with {time: The structural materizl must handle that or
be changed also, ZIxperimental loops under similar reactor oper-ting
conditions are thus inveluzable: +they allow one %0 examine aciuzl cor-
roded pipe, valves, and fitiings.
Hany factors can affect corrosion, some of which may be obscure,
The questions below help in evaluzting them:
1e ilnat 1s the composition of the corrosive fluid?
2. What is the concentration (specific gravity, pH, efc.)?
3. What is the operatinz temperature and pressure?
Le Is water present ai any time?
Se Is air present, or is the opportunity for air-leaks high?
6o Is the system ever flushed, rinsed, or drained out?
Ta Is a slight amount of corrosion objectionable from a contam-
ination standpoini?
8s Has any specific trouble or problem been experienced with
certain materials?
9¢ ‘That materizls are being, or have been, used for valves
and Tittizgs?
10e What were the comparative lives o the materials used?
11e Are there special fabrication, handling, or installcotion
problems?
12, hat is the estimated instzlled cost or capitalized cost
for the projected life of the piping?
Metallurgical structure markedly affects corrosion resistance:

fortunately it can often be altered, Physical chemistry and its

203
various disciplines are most useful for studying the mechanisms
of corrosicn reactiocns, the surface conditions,of metals, and other
basic properties.
3e4.1.3 Corrosion by electrochemical attack [80]. Under chemical
attack, metal ions leave anodic areas of the surface and enter into
the solution, thus dissolving the surface, This creates an etched
affect. In the solution they chemically react with other elements
such as oxygen, chlorine, etc., to form nommetallic compounds; plateout
of these darkens the surface witk = protective, passive film., ¥xamples
include rust on irom, oxide layer on aluminun, and "passivation"™ of
stainless steel by immersion in niiric acid.

While they remain intacw, such films generally protect the metal

or 2t least vetard further degradation. However, high stream veloci-

 

ties, vibration, and thermal shock can all break the film conftinuity.
The pickling of pipe to remove mill scale illusirates controlled cof;
rosion by direct attack,

Galvenic action, or bimetal corrosion, exemplifies another common
mechanicme, Two dissimilar metzls establish an elecirical potentizl
vhen they contact, or connect by an electrical pathway, in the presence
of a conducting solution (electrolyte).

In bolted joints, such as pipe flanges, the relative sizes of the
anodic and cathodic areas become important., An anodic area that is
small in relation to the cathodic area accelerates the corrosion.
Where flanges and bolts are dissimilar metals, the boliing metal

should be cathedic to the flange metal,

204
 

3.441s4 Corrosion protection [80, 126]. The means of reducing
corrosicn include

(1) Use of corrosion-resistant metals and alloys

(2) Protective coatings such as paint, electroplating, etc,

(3) Cathodic protection by the use of 2 sacrificial metal

higher in the electropotential series

Impervious graphite - for handling hydrofluoric, sulfuric, and
nitric acids = has some temperafure znd concentration limits., It is
unsuitable for use with free bromine, fluorine, and iodine, as well
as chromic acid and sulfur trioxide, Normally, its poraosity mzkes
graphite unsuitable as a2 piping material; however, various resins
used as impregrmants will proiuce ar impervious material. The pipe
manufaciurer should know what corrosives are to be handled so he
can provide the proper impregnant, A concern here would be the
effects of high radiation on the organic resin as well as the
graphite.

The use of lined piping sysfems for corrosive surfaces is rap-
idly increasing, Linings include glass, plasiics, elasiomers, and
various metals. Lined pipe systems usually rely on carbon steel as
a main struciural component: ease of fabrication and low cost permit
its use with most linings and manufacturing techniques, Glass;lined
steel pipe has one of the broadest ranges of corrosion resistance of
2ll,s It's smooth surface improves product flow, but it has poor
impact and thermal shock resistance, requiring care and handling
during installation and maintenence, Thermal shock resistance could
be a valid concern where one plans to completely and quickly remove

the fuel from the primary circult, several {imes per year.

205
3.4.1.5 Thermal and radiation~induced expansicn. Temperature and
radiation will cause most materials to expand. This will affect the
choice of materials for core tubes, reactor vessel, and wt—of-reactor
equipment (pipes, pumps, heat exchangers) for primery, secondary,

and tertiary I:ircuits. Table 3.4=1 gauges thermal expansion for éome

materials of interest: both Mo and graphite look attractive.

Table 3.4=I Coefficients of Thermal Zxpansion for Some Candidate

Structural Materials [ 86,27]

Cecefficient

Material Temp. Range('C) ( 10’6Z °C~!
Fxtruded graphite
- along the grain 20-100 2

1000 4
~ against the grain 20-100 3

1000 6
Mo 25 5
Hastelloy N 20~100 11
Fe 25 12
Ni 25 12-13
20455 25 17
Croloy 20-100 18

206
 

Core tubes need to be far enocugh apart that expansion does not
cause stresses by forcing them together. Design for langitudinal ex-
pansicn will prevent bowing and/or breakage (of graphite).

Proper layout should accomodate expansion in ocut=of=-reactor piping.
Expaneion effects in a pump or heat exchanger are more critical: for—

tunately radiation levels there, from delayed neutrons,are near three

magnitudes below that in the core.

3.L.2 Chemical Reactions in an MCFR

Compatible ccntainment materials are those which do not form
stabie chlcrides. ZExamples =re graphite, iron, nickel, and the re-
fractery metals. The chloridesof the pure fuel salt are hignly
stable, reacting little with materials they contact. However,
over the reactor life the core and blanket salts will mutate into
a variety of,compositions, begetiing additional chemical species:
fissicn products and free chlorine from figsion; other mutants
through neutron capture; oxygen creeping in through seals. For-
tunately, fluid systems can reprocess continuously (Section 3.7),
which helps to keep these impurities low.

Chemical reactions between these salt mixtures and inccm=
satible-container materials could

1. Precipitate fissile material leading to criticality safety

problems

2. In the presence of high temperatures, high velocities, and

high radiaticn fields, cause corrosion leading to leakage

and maintenance problems.

207
3.4.2.1 Reactions with fissimn-produced mutants. When uranium fissimns,
the UCl3 mclecular structure breaks up, crgating three chlorine ions and
two fission products (FP). Table 3.4~IIL [9] gives the valencies of.tnese
these-FP ia their most stable state (largest free.edergy of formation
[17,82,86,127] at reattor temperatures).

The FP will oanly form chlorides if the free enthalpy of formatim
exceeds about 20 kilojoules per mole chlorine: FP up to and including
molybdenum in Fig. 3.4-1. The gases xenmn, Efyptcn, and probably iodiné
and bromine are inert to chloridization; so are the more inert metals
such as palladium, technetium, ruthenium, rhedium, and probably tellur- '
ium. Based on MSRE experience , they will either entrain with the inert

gases or plate out an metal surfaces.

Valencies of the Fissiorn=Product
Table 344-11 Elements in ilolten Chlorides

 

Valency Blenents
-2 Se, Te
-1 I
O Kr, Xe
+1 Rb, Ag, In, Cs
+2 sr, 2r, Yo, P4, Cd, Sr, Ba, Sm, Bu
+3 Y, T¢, Ru, Rh, Sb, La, Ce, Pr, N4, Pm, Gd
+5 ND

208
 

 

Principal fuel Fissian Remarks
and structural . products
material
04 -Mo metal Pd{12.86), Tc(4.0%), Rh(1.73]),
(struct.) 2™ Ru(3.144), 1(6.17), Te(7.65)
Graphite » Xe(21.2), Kr(0.84), Br
- ccl e
= 4 &
U ole no chlorine for
- . the synthesis of
3 Al possible metal
g - chlorides
. a
™ -
= 1004 -MoCl wl -MoCl (15.15)}
P X “
x (corrosion -floCl, (0.28) ©
3
o product) c
[Ty Q
™~ vt
e -AgCl  (1.338) "
-GnCY1. (.57 -
' -cac1§ (0.58) 3
- -SnCl, (0.32) S
S -InCl° (0.08) &
m W 2
E 200~ -UCl4 (2} @
G m o
* -ucl, -ZrCl, (2.15) 2°
U . Qa o
a 5 o E
- -ZrCl2 _—
g g
c -PuC13 ’ -YC1 (3.021 Lo
" -Prc?3 (4.28) ee
S -EuCl> (1.086) £
3 o
o -CeCl3 (13.88) -
Q -LaCl; (5.78) o
. 3
* €
300 -NaCl -ReC1  (1.05) | S
-CsC1 (13.79%5) T
-SmCl2 (3.73)
-SrC1° (5.48) S
-BaCl2 (9.50) ) ™

 

 

Fig. 3.4=1 The behavior of fission products in the molten chlerides fuel

(Yislds given represent products fer 100 Pu fissioned)[17].

209
 

For the fissioning of 100 Pu atoms the following balance has been

suggested [17]

0,008 segas + 0,003 Brgés +.O.942 Krgas

+ 1,05 BbCL + 5449 's:~012 + 3,03 YC1, + 21,5 ZrCl

3 3

0.29 NBCl. 4 18416 MoCl, + 0428 MoCl. + 4.01 Tc
5 2 3 , met

31.45 Rumet + 1473 Rhmet + 12,66 Pdmet + 1,88 AgCl

+

+

+ 0.66 CdCl2 + 0,06 InCl + 0.324 SuCl, + 0,67 SbCl

2 3
7f65 TeC12 + 6,18 Igas + 21,2 Kegas + 13434 CsCl

-+

+ 9450 3aCl, + 5.70 LaCl, + 13.98 CeCl, + 4.28 PrCl

3
3 + 1044- PmCl

3
3+ 3.74 SmCl3 + 0,60 Eull

3

+ 11,87 KdCl 5

+ 0,03 Cd013 + traces of complex compoufids like CsUCL.
Q

This suggests the production of about 80 kg insoluble metal per
Giwth~year, That amount of metal deposition warrants serious
attention to prevent resirictions and plugging.

Since each Pu atom fissions into iwo product atoms, the average

3 chloride ions ver Pu
2 FP per Pu

neutrality. Chasonov showed that for PuCl, the average valency is

3

close to 2,0, indicatinzg a deficiency of Cl ions. However, the re-

= 145 Yo maintain electro-

 

valency should be

moval of the relatively-inert metals like Ru, Pd, and Te, either
through plateout or deliberately by the reprocessing system, can
quickly lower the average valency, 3Below 1,5, Cl ions are in excess,
producing corrosion.

In summary, tube plugging may occur at some locations and fube
dissolution by corrosion at others, In~pile testing to high buraups

is needed to study these competing effects.

210
 

Experience on molien fluoride systems indicates that irradiation
and fission products do not accelerate corrosion (lMcLain pe800)

Molybdenum as a fission product with a yield of lai (out of 200%)
of all fission products may remain, in part, in metallic form. When
molybdenum is the structural material, the corrosion problems of metal-
lic molybdenum or its alloys are strongly linked with the fission product
behavior in tnis medium. Fission fragment MoCl, will react with UCl3
to form UCLA plus Mo. Likewise the excess chlorine released will react
with the strongest reducing agent present, UClS, to form UClh.
3.4.2.2 The efféct'ofUCl; —resence in the core salfl-tsij. re UC1L,
highly corrodes container me-als, but both theory (stability of cor- |
rosion product chlorides) and labcratory experience show that dilute—UClh
fuel salt attacks little: alloys of ircn, nickel, or refractory metals
such as molybdenum resist chlofide fuel and blanket salts, providing
the UClB/UClh mix contains no mere than a few percent UCL, . Alloys can
also include minor amounts of more reactive metals such as chromium if
the design allows for same surface leaching.

Extrapolating from CEXL experience with fluoride fuel salts at
comparable heat ratings, irradiation should neither release chlorine

from the melt nor enhance the attack of the cantainer materials.

211
3.4.3 Candidate Materials for an MCFR

The high flux levels in an MCFR suggest high radiaticn damage,
However, due to the unique tube configuration the only exposed struc-
ture is the core/blanket interface: the blanket shields the vessel,
reducing its radiation dose by three magnitudes.

Beycnd the high radiation environment of the reactor lie piping,
pumps, and heat exchangers for the primary fuel-ccolant. These face
exposure to high temperatures, delayed neutron irradiation, and corrc—
sicn by salt transmutant impurities.

Materials in the secandary and possible=tertiary circuats wilil
face a much milder enviraument. However, the amounts of materials
there greatiy surpass those ir the primary circuit, thereby sensi-
tizing their cost: except for the high temperature helium case, one
prefers lower operating temperatures so as to allow cheaper materials.
The cost factar, however, must also be balanced with factors of corrosion,
thermal stress, and avoidance of secondary-coolant freezing (especially
for active liquid metal) on the intermediate heat exchanger tubes.

Table 3.4=IIT 1lists the most promising materials for use with the
different fluids. Strength cansideraticns dictate the temperature iimits.
Corrosion tests under MCFR conditions have generally not been carried
out yet. Materials which best resist corrosion by chemical attack
should be those with small free energy of chloride farmation (Fig. 3.4-1).
Low vasor pressure (high boiling and melting points) will also allow
chlorides which form to remain as protective coverings. From avail-

able information (Figse 3.4-1 and 3.4-2) Mo is a strongcardidate..-

212
Table 3.,4-I1T, Materials for Various MCFR Fluid Fnvironment

Approximate Material

 

Fluid Candidate Materials Temperature Limit €
Molten Chloride Stainless steel 600
Salt Hastelloy N 7CO

Molybdenum or TZM - 1000~1200 (?)
Graphite 15C0

Helium Nickel alloys 9C0~1000
Molybdenum or TZM 1000
Graphite 1500

Lead Fecralloy or (Croloy 500
Molybdenum 1000 (?)

Superheated Steam Stainless st=zel or high 550

nickel alloys

Lower Temperature Ircn alloys L50
Steam and water

3.,4.3.1 Overview on iletals, High nickel alloys ard refractory metals
appear to be the only metals compatible with fuel and blanket salt,
Superior sirenzgth and resistance to corrosion and radiation embrittle
ment at high temperature recommend molybdenum alloys (particularly TZH)
for contact with the circulating fluid.

There is some incentive for using molybdenum alloys for all the
components (including pumps) which contact the salt., However massive
molybdenum outside the core would require precautions against its
external oxidation: e.ge enclosing the reactor vessel and other sensi-
tive components in a hot box purged with an inert or slightly reducing
atmosphere, An alternctive to this would be to use duplex material:

a protective Mo layer elecirodeposited or plasma sprayed onto certain

213
 

 

 

 

UcCls oPuCla
9
~ SALT-FUEL
COMPONENTS
1500- MgCle NaCl
J Z2rCla
.C
O
2. 10004
L
- Q/Q/
z MaClx
= 1 1
2 ] WCl,
3 PtCl,
0
— 500- AuCls
]
0 100 200 300

. =000 -t
Free Enthalpy of Formation AG'm [}c}'.molCfl

Fige 3e4=2 Desistance-of Metals to Chloridization Corrosion [17]

214
 

steels,

For the less severe conditions of the reactor vessel and the salt
ducts returning from the heat exchanger, Hastelloy-N c¢r advanced devel-
opments of that alloy might suffice, However, having decided to use
melybdenum for the core/blanket membrane, a change to a dissimilar
metal elsevhere in the salt circuit could cause galvanic corrosion.
Also, nickel alloys resist molten lead (a2 candidate secondary coolant)
very poorly compared with molybdenum; they would probably be subject
to attack when the lead leaked fthrougn the heat exchanger into the salt,
3044342 o 2lloys. Mo offers high temperature sirength, zood irra-
diation resistance, and high thermal conductivity: promising traits
for 2 material in contact with fuel and blankev salts at hizh temper-
atures, However, fabrication of light molybdenum sections requires
welding and heat ftreatment: this technology needs development, but
it is already proceeding for other applications,

Also, the high specific cost of molybdenum is acceptzable only on
a2 limited basis: +the vessel piping, eic, should use materials which
are cheaper and more easily fabricated. Candidates include Mo/Fe,
Ho/Mli and Ti/Zr/Mo (TZ1) alloys. With Mo/Fe indications are that the
swelling for Fe has already peaked at the operating temperatures,
while that for Mo occurs 2t much higher temperatures,

The use of molybdenum or its alloys may permit high fuel salt
temperatures (above 100000). This can reduce fuel salt inventory
through higher power densities and allow gas turbine cycles and/or
process heat applications using helium as an intermediate and final

coolant,

215
Mo has an appreciable absorption cross section for neutrons. Such
reactions also produce significant amounts of technetium. This changes
the nature of the structural material and makes it radiocactive.

Figure 3.4=1 and the discussion in Section 3.4.2 indicated Mo to be

strengly corrosiaon resistant. The most significant reaction from the
standpoint of ccrrosicn shouid be (Fig. 3.3=13)

UClb, + Mo ———i < UCJ.3 + Mt:JCLL2
UCi 4 will be a major censtituent in the blanket from the neutron transmu-
fation of ThCla. Free chloride from fission of U in the core will combine

with UC].3 to form UClL. It also reacts directly with Mo:

Mo + 012 — Mo()l2

Traces of oxygen or water will produce molybdenum oxide, €.ge.

MoCJ.2 + H20 — MoQ + 2 HCL

With all of these reactions, mechanical properties will change.

3.4.3.3 Graphite. Both carbon and graphite resist corrosion
and conduct heat well enough to warrant exdsting uses in heat
exchangers and pumps. A resin-bonded graphite has found wide-
application in the chemical process industries. Both America
and France have used graphite with Li.F/BeF‘2 molten salt: Mce=-
Donnell~Dougl as markets a high-strength, woven graphite for this;
the French company PUK sells a product with graphite connected

to heat exchanger material [27].

The high thermal conductivity results in excellent thermal

216
 

shock resistance. Still one must exert care as carbon and graphite
are weak and brittle compared with metals.

Tensile strength varies between about 500 and 3000 lb/in.z,
and impact resistance is nil. Abrasion resistance is poor. Highe=
temperature stability is good. and they can be used at temperatures
up to 4000 or 5000°F if protected from oxidation (burning).
Silicon-base coatings (silicides or silicen carbide) and iridium
coatings are claimed to give protection up to around 2900°F.

Based on the existing experience as a structural material
for molten fluoride thermal breeders and on periodic replacement
of tubes, we assumed graphiie for the physics studies here: 2 cm

thick for the tube wall anc 4L cm thick for the vessel wall,

1.4.3.4 Heat exchanger considerations. As mentioned in sectian 3.3.2.5
low vapoer pressure of the primary and secondary fused salts permits
thin-walled heat exchangers. These are highly efficient in heat transfer,
thereby decreasing necessary area and fissile volume out-ofecore.

However, assuming close pitch, they will require development

and testing to resist the thermal and vibrat onal strains over

a long working life.

Refractory metals are attractive as they can accommodate high tem—-
pefatufeslfihiie»still affording the good heat transfer of a metallic
material. A strong candidate is Mo, It does have an appreciable cross
section for neutron absorption: +this affects delayed-neutron activation
but not neutron ecanemy or radiaticn damage there. Of greater cancern

is the welding and joining properties of Mo, its cost, and its oxidation.

217
3,4.4 Materials for the Core/Blanket Interface

_Finding an interface material that can withstand a 1016 n cm-'2
éeo;1 flux over an acceptable period of time is indeed a challenge,
It is greatly eased by assuming periodic replacement of the few tubes:
an action far less complicated thar the conventional replacement of
fuel elements; just drain the core and blanket and uncouple the tubes
outside the reactor, This is especially feasible if one uses a cheap
material like graphite. Section 3.9, Access and Haintenance, discusses
this further. Still the tubes may experience a cumulative dose of 1023

n cm-2 in just a few months, Also irradiation of graphite produces

"stored energy'.

ledeH HMaterials for Reactor Vessel

 

The reactor vessel will experience

(1) Low pressure

(2) Temperature higher than the blanket salt meliing point

(3) Radiztion damage from a 10'3 1 em™? sec”! flux.
A suitable material must meet these conditions; it should also mini-
mize

(1) Induced radioactivity so as to maximize access

(2) Cost

Candidates include graphite and prestressed concrete (of low=

activating compositions),

218
 

3e4.6 lMaterial for a Lead Secondary Circuit
34446,1 Steam generator tubes, Mo alloys normally resist lead cor-
rosion well, bul mizht not if a steam generator lezked in small amounts
of oxysene Dissolving Mg or Ca as a getter in the lead would help
lower the oxygen level, but might not suffice, One could also clad
the molybdenum with an alloy which resists "oxygenated" lead; Fecraloy,
2 ferritic sfeel containing 2luminum, forms a protective alumina
film under these conditions,

A third, untried alternative, might be that Fecraloy alone can
resist both steam and oxygepated lead, Forming a duplex with staine
less steel could overcome ifs poor high temperature strength,

The Dossibllity of m2ii-g the neat exchanger tubes of Hastelloy-
¥ haes only slignt attractions: it would still need a Fecraloy or molyb-
denum external cladding to resist the lead,

Y,0e642 Remainder of the circuit., The ducts carrying the lead to
and from the steam generators and zlso the steam generator shells
might be molybdenum, protected from external oxidation either by a
stainless steel cladding or by an inert atmosphere. An altermative

is Fecraloy, externally clad with stainless steel for strengthe.

219
 

3ede 7 Material for a Helium Secondary Circuit

The main change relative 1o the lead-cooled design is bigher.tcm=-
peratures, parallel to a core temperature increase from 810 to 970°C.
This suggests the need for molybdenum alloys, especially on mechanical
grounds, However, even though there is no reason to doubt their cor-
rosion resistance at this temperature, any supporting evidence is come
2letely lacking, British studies showed {that design variations might
lower the sa2lt pump operatinz temperature from 800°C %o 67000. This
would ease the engineering and the selection of materials,

Materials selection for helium cooling should ve much easier than
for lead cooling but so far has been little studied: unidentified orob-
lems could become just as ch:;lenging as with lead-cooled sysiems.

Presumably a helium cooled system could draw on the extefi;ive
work already done with respsct to HTRs and gas turbines: The MCFR
version could actuwally ride on the future HTR program,.

HTR programs have recognised that, in practice, helium is not
an inert coolant since iraces of air, moisture and other oxidizing
gases can cause corrosion and traces of carbonaceous ges can car-
burice structural materials, Thoush much work remains it does seenm
likely that it will be possible to specify

1e The impurity levels acceptable in the helium

2, MNethods of controlling them

3o What intentional additions if any, should be made, for

example, to render the helium reducing,

Selected nickel alloys could be used for {the turbines and other
circuit components, This would restrict the need for molybdenum in

the coolant circuit to heat exchanger tubes and similar applicztions,

220
 

3.5 Physics of MCFR(Th) and Its Fuel Cycle

3.5.1 Nuclear Models of an MCFR

J.5.1.1 Definition of the reactor geometry. Section 3.l1.3 conclud—
ed that the best MCFR arrangement might be a small number of skewed
cylindrical tubes. The chain reaction will center where the tubes
converge. However, as nc fuel salt boundaries exist along the tube
ares, criticalitiy boundaries blur,

High levels of flux and power generation in the primary salt
may extend far along the tubes. This will depend on individual tube
subcriticality, distance of separation between tubes, and angle of
their skew. For example, if they are near critical by themselves and
are distant from one another, then their fluxes will follow a cosine
distribution over the total distance between bends. ‘However, ine
creasing the intertube'coupling will cause the fluxes within the tubes
to fall off more steeply beyond the tube convergence region.

Th concentration and other fuel salt parameters will also affect
neutron distributions because they moderate the neutrons which short-
ens their mean free path. Fluxes will also extend into the blanket
in all directions but with much shorter relaxation lengths.

The choice of tube arrangement will depend upon a detailed trade
study of effective core size: extension of the chain reaction out
along the tubes will increase exposure of the pressure vessel to
neutrons emanating from the tubes; making the core size small re-

duces stability of the system to perturbations.

221
3.5.1.2 Use of spherical geometry for neutronic calculations.

Neutronics analyses of complicated geometiries require expensive Monte

Carlo calculations. For.the present survey purposes, it should .

suffice to use simple one-dimensional spherical geometry featuring
five zones (Fig. 3.5=1): core, core tube wall, blanket, vessel wall,

and reflector.

 

 

 

. —7 __——core L7 — reactor
tube vessel
II wall wall
E core blanket | reflector

 

 

Fig. 3.5~1 Cne-dimensional Spherical Model

All the tube wall material gets lumped into one spherical shell.
The validity of this approach, vs. uniform material distribution in
the spherical core, will depend upon the number of actual tubes. For
only three or four tubes, the current thinking, it should do quite well.

Taking the tube wall thickness to be two cm facilitates compar-—
isons with MCFR(Pu) studies. The vessel wall is four cm thick; the
reflector zone, 4O cm. A check calculation with 20 em reflector-
thickness showed no significant changes when the blanket thickness
was 200 cm (quasi-infinite).

éection 3.5.7.2 studies the effect of blanket thickness, settling
on a value of 200 em for most cases. The critical core radius usually

ranges 25.-= 50 cm.

222
Section 3.3.5.6 gives the densities of individual salts and an
algorithm for combining them. Densities for graphite and nickel were
0.1 and 0.09 atoms cm * barn™l.

A primary difference between an exact calculation and those in
spherical geometry is likely to be the spectrum softening and other
effects of the ThClA/NaCl blanket salt intermediate between the fuel
salt tubes. Depending on the clearance and skew angle between tubes,
a smeared model of such a spherical region would probably include 5=
25% blanket salt by volume.

In this regard, the calcilations may be better cdone in infinite-
cylinder gecmetry. That remzins to be shown however,iand cylindrical
calculations cost more. In the meantime, spherical c;itical U masces
(U) or ¥slumes Gan be corrected to -cylindrical geometry by use of -
shape factors:

U @/D) = U, /SFE/D)

Fige. 3.5=2 displays some experimental and calculated data. A sub=-
set of these (Table 3.5I) was fit with a least squares program %o
produce the foilowing algorithm for use in Section 3.5.5 studies:

log SF(L/D) = 64399 x 107> log(L/D) - 0.6976 (log(L/D))*

- 90820 X 10-3

223
 

 

 

 

 

1.0 -
O —Q
s\
™ 4 \‘\Jn\
0.87 Q
A
o A
£ 0.7 o
3
O
)
=
0.6 4 4
7 A
0.5 + O
OO]+ ! ] 171 I { , ] l f
Ool 0-2 Ool:l. 006 008 1 2
L/D

Fige 3.5~2 Shape Factors for Cylindrical Fast-Neutron Systems [123]

Legend D  high-density metal system
O  low=density metal system _
0O ZPR~III 1l:1 critical assemblies
Table 3,5~ Experimental Shape

Factor (SF) Data [1281 Assumed for This Study

 

 

L/ SeFs _L/D_ SeF.
0.217 04297 0.9 0,96
0,257 04565 1.0 096
03 0.7 1425 0,94
0ud 0,82 Ted 04 91
0.5 0.89 167 0.87
0.6 0,94 2 04825
0e7 0,96 2 04825
0.8 0,96 8,79 - 0429

225
 

Je5.2 Neutronic Calculational Methods

Neutronic analysis of small fasi reactors like MCFRs generally
requires transport theorye. Mcdeling the reactor as a one-=dimensional
symmetric sphere permits use of the well established and efficient
ANISN code [129].
3.5.2.7 Spatial mesh. A spatizl mesh of 1/10 of a2 mean free path
£cr interval would give quasi-exact results in transport theory
calculations, Computer storage and cost generally prohibit this,

A mesh of one mean path (mfp) per interval suffices for most cal-
culations,

In MCFR(Th) calculations, the neutron mean free vath typically
ranges from 1-10 em (Teble 3,5-II), With a core radius of 30-80 cm,
a 200 cm blanket and a 40 cm reflector, a uvniform one mfp interval
mesh would mean a lot of intervals, PFurthermore, most would lie
outside the core, leading to possible numerical instability in the
iterative transport theory calculations, Consquently a coarser mesh
was used (Table 3.5-II), especially outside the core, This worked
fine for most of the survey calculations; 2 few select effects needed
finer detail,
3¢5¢2+.2 Angular quadrature, Most studies were done with an 34 angu-
lar quadrature, 88 check calculations on the smallest of spheres
showed increased neutron leakage as expected, but only increased the
volume by 1.5%, while decreasing BG by ~ 0.01,
3,5.2.3 Convergzence criteria, Initial survey calculations satisfied
a keff convergence of 0,01 and a critical radius search convergence

of 1¢%. Select effects required 0.001 k pp = cOnvVergence,

226
Lee

Table 3.5 Spatial Mesh in MCFR{Th) Neutronic Calculations

Core | Vesgsel
Core Wall Blanket Wall Reflector

Typical mfp (cm) 12 3 10 2 1
Typical dimension (cm) 30-80 2 200 L LO
Renge of dimensions (cm) 21~136 2 100-200 4 20,40
Survey No. Intervals 10 1 10 1 b
(10/16)* Int. Size (cm 3-8 2 20 4 10

" v (mfp 0.25-0.67 0.67 2 2 10
Detail No. Intervals 20 1 go 1 14
(20/1,6 )% Int. Size (em 1.5 2 o7 4 3

" n (mfp 0.1 0.7 0.7 2 3
Multizone No. Intervals 10+10 2 10+10432 1 545
(20/65)* Int. Size (cm; (Re5,vb1l) 1 1. 3. 5. 4 3.5

" " (ulfp (0.2'Vbl) 003 O.l 0.3 0.5 2 3.5
Intermediate No., Intervals 10 1 25 1 L
(10/31)* Int. Size (cm; 2.7 2 8 L 10

" " (mfp 0.23 0-7 008 2 10

* (no. ¢ore intervals/no. non-core intervais)
 

3e5.3 Neutron Cross Sectians

This study required neutron cross sectians for Th, 233U, 23l‘U,
235U, 2330 fission products, C, Na, Ni, and Cl. Data for 236U, Mo,

F, and other carrier salt catlons were desirzble,

3.5.3,1 Bandarenko 26 group set [98]. - The Bondarenkc neutron crbss
sectlon library long ago =stablished itself for fast réazctsr caleu-
lations, Testing has revealed deficiencies in the Pu isotopes, 235U,

238

and U, but these nuclei by and large do not appear in an MCFR(Th).

Less is known in general about Th and 233U crdss sections; they are. cur-
rently undergoing extensive reevaluation for ENDF/B-S; For survey
ournoses nhere, the 3ondarenk: values in Pl-cor-ecied PO form (trans-
port cross sections) should suffice, :

 Nuclei fiiésing from ihe library are Cl and F. Early in the
study, K substituted for Cl: this was conservative as K absorbs
more neutrons. Final calculations used Cl and FPl sets derived from

ENDF/B - IV.

 3.5.3.2 Twenty-six group, P1 set from ENDF/B-IV. Available
late in this study was a 92-group Pl set of cross sections gener-
ated by AMPX [130] from ENDF/3-IV, containing most of the materials
of interest here. This was too large and expensive to run so the
set was collapsed to 26 groups similar (Table 3.5-III) to those in
the Bondarenko library to allow cross—comparison and -usage.
The list of materials included not cnly Cl but also F. Missing
however were Na and C. Comparison of their nuclear properties indicat~
ed Mg would represent Na well, especially since Na content is small

campared to Cl. For C, the Bondarenko set was used ; C ic also a minor

228
 

5
:

SRS EGREBwe~onrwbro

Table 3.5III

Lower Boundaries of Energy Group

Bondarenko Set

10. 5 MeV
6.5 MeV
L.0 MeV
2.5 MeV
1.4 MeV
C.8 MeV
O.4 MeV
Q.2 MeV
C.l MeV

LE.5 keV

21.5 keV

10.0 k2V
L.65 keV
2.15 keV
1.0 keV

465 ev
215. eV
100. eV

L6.5 eV

21.5 eV
10. eV
L+65 eV
2.15 eV
1.0 eV
0.465 eV
0.215 eV

thermal

229

149
6.07
3.68
2423
135
0.82

AMPX-Generated EXDF/B-LV Set

MeV
MeV
MeV
MeV
MeV
MeV
MeV
MeV
MeV
keV
keV
keV
keV
keV
keV
eV
eV
eV
eV
eV
eV
eV
eV
eV
eV
eV
 

constituent.

3.5.3.3 Effects of resonance self-shielding. Résdnance—regicn
cross sectims for the two sets described above were averaged
assuming the materials to be infinitely dilute, i.e. ignoring
resonance self-shielding. In 233U most of the reactiviiy effects
from fission and capture self-shielding, if present, would cancel

each other. Self-shielding for 232Th capture, if present, could
affect 8G.

The Bondarenko-calculated self-shielding factors for capture

(Table 3.5I7 ) show virtually no self-shielding in the top 12

grcups, which cover well ths core and blanket spectra. %“hat little

. 232
does occur in grcup 12 fcr

233U is similar. Thus the infinitely-dilute calculations

Tn, Doppler Broadening partly amelio—

rates.

appear correct to first and maybe second-order approximaticn.

230
 

Table 3.51V

Bondarenko Self-Shielding Factors for

B2 98]

 

"C at Uo

cqual to }l I, at O, equat to

 

 

‘g t st G5 equal to
1!

l 10+

I

 

 

15

2l

 

 

 

 

 

 

 

 

 

 

 

I
I
' ! r ; l
300 ‘ 1.00‘ too! 08 'oe7 1o T1uo {098 & uo 1,00 0,99
500 ! 17001 vool 100 noy § o fioo oo | 10 | 100 1,00
20 | LW % LOUG Lo Lo tuo LU0 e T Lo L Log ) tu
L R
© 300 LU0 LO0T 0.7 0.02 T b0 099 004 100+ 0,99 1097
a0 00 L0000 00 L Lo 09 L log ! 1,00 0.0
2000 1,00 | !.OUi 1,00 in.slv HERCIRRE | Luo 100 100 5 e
| o
S v i
WO Lo 1,00, 0,95 'o S0 41 1097 |oss 11,00 099|093
Codw L : OO 098 0.~ o outd "y Y LLAo » w99 | 0w
2000 11,001,001 0,99 %().-..l L 100 1o iu,gs ; 100 | 1,00 | 0.99
' : ! , E ; :] !
OG0 Lou 0T 002 02 T tae 093 T05) b0 ) 0,93 ¢ 6,96
POU Tu v Yy 002 sty 097 0S8 L0 099 uEs
PII0C LG bt 099 o2 (L 1 ,IO,SZS <100 100 1 0,94
Ly | |
300 1 1.00.097: 087 (0,45 i 1,00 {086 :0,59 , 1.00 | 0.94 | 0.7
900 11,00 0981 0,92 {053 ) oo |09t (0,62 oo {097 | o's0
2100 [ 1,00 Tow | 0,96 30'64 | Loo juve 0,65 0 | 095 | 0,86
| L
i o
300 {1,00:0921 058 1020 034|050 [0.29 1092 ) 0.65 ! 0.4
| 900 100 095, 0,67 1025 087 1058 10,30 | 0,95 [ 071 | 044
| 2100 ‘ 1,00 10,46 i 0,74 lu31 {091 1066 10,32 40,9 | 0,77 48
i
| ‘ !
300 | 1.0010.01 1 0.59 10.21 | 0.82 1044 10,23 {090 | 0,61 { 0,36
o000 | 1,06 10940 0,68 1026 T 057 |0.51 |0.2¢ §093 | 0,66 | 0.39
| 2060 LU0 006 076 035 09l 10,62 1027 0,95 0,73 | 0,43
N L |
| 300098086045 1003 0R2 1037 10,22 2091 | 0,64 | 045
L 900 1099 6l 0054 10T Cosr tuks 10023 Y04 | 070 ¢ o030
| 2100 099094 062 vl i (.94 5U,Su 0,24 40,9 10,71 0,30
, R
300 {0,9610.72| 0.27 100837 0.61 029 10,22 i 0,85 | 0.64 | 0.5
900 : 0,97 .0.76: 0,30 ! 0.0857 0,70 10,30 1022 [ 0,88 | 0,65 | 0.54
2100 {098 0,82 | U3 10,099 | 07D 1034 10,23 |09 | 069 | 055
i ! )
i | I
© 300 11.0011.00 | 1,00 | 100 4 1.00 : 100 11,00 § 1.00 | 1,00 | 1.00
900 + 1,00 - Lo 1,00 100 1 16O 100 1I.UU 21,00 11,001 100
2100 | 100 100 100 Luo "t e g 1,00 1,00 ) 1,00 | 1,00
o
1 300 [ 09110551 0,16 0058 0,37 L0080 0,043 . U538 | 0.22
SOU OO 06T 0 0 0TeE 07 0 T00 T 067 1 026 10,14
i 2100 | 0.96 {0761 0,31 0,097 i! G0 1004 0 077 | 0,36 | 0,
i 1 ! i

231

 

 

e —————

 
3a5.4 Transmutation Chains and Equations
345461 Chains for actinide transmutations Figure 3.5-3 details
the Th cycle burnup chein as it pertains to reactor criticality. 233Pa
branching ~ by neutron capture to 234y or by‘fia- decay to 233y ; is
very important in a thermal reactor where 234U only absorbs neutrons.
However, in 2 fast HCFR(Th) spectrum 234y mostly fissions (Figure
Zel=3) ;-- much more so than any other fthreshold-{ission isotope. It
does not significantly moderate the spectrum or otherwise detract.
33 232Th: 233

2
Production of ~~~Pa follows directly from

233

Th decays

quickly enough to ignore i1ts dresence, Pa decays at the rate of

1n 2/27.0 days = 3.0 x 10-7 disintegrations per second per 233Pa 210Mma

With g = 10" 5 en? s in core (25% of +ime), ~ O oui-~of-core, and

a core-spectrum-averaged cross section of 0.3 barns (Table 3,5-V),

233Pa(n,Y) reactions oceur 0.7 x 107 3-1, more than two magnitudes

233

slower than Pa decay, In the blanket the greatly~reduced flux
lowers the reaction rate even further, A fhermal reactsr with<3é200 =
39b and f = 104 5 en™® s 0,25 yields a similar rate of 1 x 1077
s [1313.

In summary, 234U production from 233Pa parallels that in 2
thermal reactor; but a low-threshold fission cross section raises it
to fissile fuel status and also destroys it faster. Therefore ihe
reduced presence of 234U and the 233U;like reactivity behavior of
234U in an IICFR obviate the need for separating out 233?& to let it
decay as in an USBR, Therefore it suffices in these survey calcula~

tions to igmore the 233p, (n,¥ ) branch, i.e. to construe all 232

233

Th

captures to pass directly to Us

232
£ee

232 e 23y Py

 

23 1;1>a ’__J 233% _____]

2310 et

 

Figure 3.5-3 Th Cycle Burnup Chain for Neutronic Calcul

;236Pu ‘ 238Pu_ e 239Pu
1___237Np ____J
23 5U 236U
(n, ¥, -
LNy 2n) p
ations
 

Table 3.5V

MCFR{Th)-Spectrum Averaged (ross Sections for Some Obscure Reactions

 

Neutron

 

6.5—;0.5MeV
LeO=b.5

2 5=440
1.&"2.5
QuB=1.4
Q.4=0.8
0-2—0011-
0.1-0.2

L6 o 5=100keV
21.5—46-5
10.0=21.5
465=10.0
2 415=4..65
1.0=2.15
1,65-1000 eV
215 465

17 100=- 215

18 46.5-100

19 21.546.5
20 10 =21.5
21=26

E;CRE:C3FSE:ESq)03«JChvwx-th)h'

Core average

 

 

Cross Sectigé {tarns )* > g;zztrum g;zgtizm
Group Energy 230Th(n,)j 33Pa(g,73 33Pa(an) (%) %)

0.006 0.01 1.2 0.501 0.0LC3
0.005 0.02 0.8 3.18 0.272
0.0066 0.05 0.8 8.46 0.835
0.01 0«13 Ca7 17.15 2.78
0.2 0.3 Oal 19.92 6.82
0.034 03 Oe 22.03 16.5
0.04L 0.3 1L.42 18.5
0.05 Ouds 7 .86 17.1
0.12 Q.6 LeO 1Lel
Qo2 lel 1.48 1l.1
0.32 2.2 0.778 Te71
Qo 5e5 0.133 2.70
0.65 5¢3 0.0372 0.521
l.1 ) 0,011 0.84L2
1.7 1.5 nil 0.1522
0.5 10 1.49x10_

25 1.87x10

IQ—O 1-L083"l+

95 3.954

nil

0.035 0.29 C.53
0«12 0.79 0.83

Blkt.average

* cross sections eyeballed from @NDF/B-V.

In calculating burnup (fuel depletion and replenishment), this study

also omitted the

232
3 Th(n,2n) and 233U(n,2n) paths as well as the cal-

c e i 236, ...
culation of isotopes higher than 3 U (Fig. 3.53). These approximatiaons

should not affect the conclusions of this survey with respect to BG or

criticality. The resulting chain was

234
232Th 433U 23LU 235U 236U

Each arrow denotes an (n, ¥ ) reaction.

The corresponding equations, which describe nuclide activity up

through 235U, are

d,/dt = Gy @ Th = (0, &) Uy (345.bm1)
dUh/dt = (S &y Uy = {o,4) U, (3+544=2)
dUS/dt = (64 9 U, - <G, 0% Usg (345.4-3)
Here
Uy = 233y content (acoms)
Ul+ = ¥4y content (ztoms )
Ug = 23% content (atoms)
Th = 2?2Th content (atoms)
g = neutron flux (n cm ® s )
£, = cross section (b) for 232Th(n, Y) reaction
Clu= ¥ . w a0 *3y(n,abs) "
Sy= " v non Fya,my
Sea= " ‘ w ow Blygoaps) v
e e By gy
bue 00w BSynamg) o

(6 &)= total reaction rate, integrated over the neutran spectrum
(reactions 5™t atofi-l)
Table 3+5=VI lists the ANISN—calculated reaction rates for

numerous cases of interest.

235
9te

 

Table 3..5-V1

Core and Blanket Average Reaction Rates (Reaotions/Atom/s) for Some Configuraticns .of:Interest

 

 

k.3
Configuration
- Nuclide Reaction Zene 1,7,Ni,30 0,7,6,30 0,5,0,30 0,3,C,30 3,7,€,30 0,7,C,50 0,7,C,70
232Th capture core 6.43-9 O. 0. 0. 2.40-~-9 O. O.

233y capture core  5.41~9 9.8L~9  9.67=9  9.37-9  2.09~9 9.07=9  7.60=9
blkt. 2.72-10 3.86~10 3.81-10 3.75~10 1.43-10 3.,68-10 3.35-10

absorp. core 7..5.-8 1.65-7 1.60-7 1.53=7 2.57-8 1e45~7 1.11~7

=3y capture corc  9.26m9 1.78-8 1.74-8 1.68-8  3.44~-9 1.61~8 1.31-8
blkt. 3.95-10 5.82«10 5.76-10 5,6510 2.08~10 5e53=10  5.00-10

absorp. core 3.21-8 8.21-8 7 «95-8 746-8  9.42-9 7 .02-8 54088

blkt. 6.16=10 9.31=10 9.20-10 9.00-10 3.04~10 8.78-10 7.85=10

235U absorp. core 5.4,8=-8 1.17-7 Lo l4=T7 1.09-7 1.90-8 1.04—7 8.08-7

blkte 1e92-9 2,759 2,729 2.67-9  9,84-10 2.61~-9 2..36=9

% The nomenclature for each configuration ias

ThClh/UCl3 core ratio, 233U/ZBL’U core ratio, reflector materiel, NaCl molar content (%) in

core salt
 

3.5.4.2 Chain for buildup of “<U-parented radicactivity. < <U
buildup in the fuel merits special attention because it heads a

lengthy decay chain (Figure 3.5=4) which culminates in 208’1‘1, the

emitter of a penetrating 2.61-MeV gamma ray.

 

i - 232 '
Fig. 3.5-4. “U Decay Chain

2317
The high-energy gamma necessitates remote handling of the re-
cycled fuel and complicates the fuel reprocessing and refabrication.
This discourages fuel diversion and theft as well as subsequent weapon
manufacture, For legitimate HMCFR operation, which requires no fuel
fabrication ard continuously reprocesses fuel anyhow, it poses no major

additional problem,

Some 228Th will actually exist prior 1o 232U production due to
232Th decay:
& - &
232y, s 228, {? 228, o 228,
1.39 2 107 ¥ 8Ty 2

This might assist as a deterrent early in reactor overation due
to the 2.6~ileV gammz siznal, but being a different chemical element

Th will easily separate from the U fuel, Therefore it plays no sig-

nificant role here. No other isctope besides 232U should be initially

present,

2
Several paths lead o 232U in a Th/ 23‘U fuel cycle. Each starts

232

from a source material (enclosed in boxes in Figure 3.5=5). Th

233

and
230

U are obvious source materials in core and blanket, About 1 ppm

233

Th normally occurs in thorium; it comes from decay of U which
occurs naturally in most Th ores,
The route

233U(n,2n) 232U

should dominate, The routes

238
 

 

 

232U Buildup Chains., Boxes denote ié’otope which
exist in reactors at startup.

Figure 3,5=5.

Dn(n,zm) P 5) Plrala, v) e ( 2) 232
20m(n, Y) #m( ) 21pa(n, v ) P2pa( p) 24
22n(n, ¥) *Pm( 27) *pa(n,2n) *%pa( 5) 3%
will also contribute.
The threshold nature of these reactions implies that 23 2U will
buildup rmuch more in fast reactors than thermal cnes.
Magnitude comparison of the variocus paths can lead to some
simplificaticns. Ignoring the delay from 25,6k half-life decay
of 211y and 22.1m . half-life decay of “-Th, the differential

equaticns are

239
 

dig [dt = G PN,y + Grpd Ny + Sy @ N3 = (2 + 5 d) Nap

(345.4-4)
d“u/&t = GDZN¢ thZ + Ggof¢ Nco - 6\| d)M” (3.5.4_5)
Nt = Gaey BNz = (hs +Ga® )i © (345.4-6)

where

1,

.. .th ..
N concertration of 1 nuciide

H

(78
i

= 00,02,11,1:,22,23 for Th isotcpes 230 and 232,
Pa isotopes 231,3, and U isotopes 232-3, respectively

16

=
i

total fiux (n 3{{2 s™1) = typically 2 x 10
in cere, 5 x 10°" in Dblanket

2330,

?\13'—" probability (N) for g-decay of
= 300 X 10-7 S_l '
Mg= A\ for X -decay of 2% 2 3.1 x 10'10.'_5"1

Cpoy= SPectrum—averaged cross section ( & ) for the

1
259 Th (u, ¥ ) reacticn.

9

Subscript ¥ denotes (n,}./) reactions to distinguish from N which de-
notes (n, 2n) reactions and from a blank which denotes total neutron

absorption.

240
The procedure for determining most IC%E was to average their group
values ;Gi(g?}over a typical calculated MCFR spectrum. ANISN calcul-
ations did this for Cpe® (Table 3.5~VI). Table 3.5-VII similarly
evaluates O ., d, ,and G,, . svalaations for GFK}*‘ and Gfi_ data est—
imated from the literature gave results too low to warrant more de-
tailled evaluation.

Where pointwise data is unavailable, MCFR spectrum—averaged reac-

tion rates { $63 ) can be estimated from fission-spectrum averages by
(Fs)
po=] 2] o
X,
Here,

=]

It

.ratio of flux to fission-spectrum fraction in the first neu-

tron energy group

= 5x 1015, typically for an MCFR(Th) core

13

= 1 x 10, typically for an HCFR(Th) blanket

Calamand [132] and Gryntakis [133] evaluated the fission spectrum
average for Snon and 6-231‘1 to be 14.2 and 3.3 mb, respectively. Fitting

Calamand's empirical correlation for
G(Fs) - 343, & 2/3 _-0.98156 Fo(MeT)

Then for 233Pa(n,2n) with Ep= A41 (=Q) = 234
i

_ . 233 *
5 1(;31) = 21,0 mb

Table 3.5—ViII summarizes these results along with all the others above.

241
Table 345-VII

Some MCFR(Th) Spectrum-—Averaged Four-Group Cross Sections for Actinide

Transmutations [134]

 

242

C (v)
Tnergy 235, 3o, 232
1 0.8-10. eV 0,01 1,2 1.97
2 46,5=800 keV 0.43 0.18 2,08
3 1=4645 kel 343 O 3.0
4 0.465-1 ke¥ 61, Os 14,
5  thermal 200, o 1554
Spectrum Core 0436 0467 241
Averages Blkt. 1.2 0e25 2.3

Spectrum (%

Core

4942
4843
2e4

0.1

0.

Rlanket

10.8

6642

22,9
0e17

2e1=T
Table 1,5=YTIT fvaluated Production and Destructicn Coefficient for

 

 

a Typical LCFR
Reacticn Reaction Activity(s-¥)7
Ccefficient* Symbol Descriptian gore Blanket
: = 232 =11 -
X ooy C oo ? >21h(n,2n) 7x10 1x107 22
= 2 - ~-10
A oy T (YY)  4xad”? 26
~/ 231 . - -10
11 llxfi 3 Pa(n! () 7x10 7 6x10
~ 13 ‘gumfi 2339a(n,2n) 107 2x107P
23 So?  2Hq(n,2n) 20107 a0
) = 232 -3 =10
73 3
2 11 1 Pa{n,abs) 2x10 7x10
A\ 233, [/ 4~ =7 -7
/,3 3 » 13 pa’ b ) 3.0x10° " 3.0x10
3 232, -8 w17
/;522 o Y (n ,abs) L4x10 1.x10
» oo 22 (o) 3.1x10720 3,1:007%°
T Pm,Y) 7,100 6510712
o 57 23pa(n,abs)  1.1x107° lx10 0
S o7 . . . Sal .
# 7 ® ¢ are production coefficients and j(i; are cestruction coef=-

ficients. Their use in the differential equations helps in visual

checking and bookkeeping of terms

243
 

The probability ratio R of path <3 OTh(n, Y )23 lTh to path 23 2Th

(n,2n)231rh (Fig. 3.5=5) is

R=_ Soo ‘oo
o2N 'NOZ
For N Mmuo = 1{)—6 (1 ppm) mentioned above, get
R = 107 x Zi-lf);g - 107 in core
7 x 10
c o0k SELTT |60 i bramket
1x 1079

Thus the = OTh(n, ¥) 231rn path can be ignored.

In the destruction of N5, )\13 = 3.0 x 10~ s~L far exceeds ¢0'13:
in the core by a factor of nez- 30; in the blanket, by three magnitudes.
In the destruction of N22, 622;1) far exceeds }\22 and t@:lus becomes the
governing time constant (of about 10 mos.) in the approach to 232, eq—

wilibrium concentration.

244
Taking La Place transforms of Equations 3.5.4=4 through <6

23/”23 * 0(11 1t n/}iw /& 22

77
{/‘Q * g3 Moy + Ky 111*“13%132

SQ'{22 -42 22 = d

 

22 T s4 -'_)22

s My - 21 - & 22@{- 02 =2 U 11

.
e T 1S ]

s /0( fl,( /p - /513/}( 13

 

42 = s+/} * O2Y/]{ 02

-
.-.\
J

.
Assuming k. constant = N, for 02 and 23, then 4101 (s) = é /VLR

<

 

 

. i s
.eru (s)=3 4123 ) and substituting, get
| olya, 1 ° i o
e ——Jqn 2l K g
2 S+ Py s S s+ Oy "

+ o3 [ ° o2y ’flc,, ] }
. /]'1 , o—_— C
S 4-/—1"\3 ' S

245
Now initially one would have
o o
2’\3 (n A/lll,
Then

l c '){H ) <
Q = — ‘x23q£l = Doy 4Z
/(27_ s (s+042) { ’ s~/ ez

L Fia

c
= 'XOLY ’z('
Sf'_f),_; cl

Normalizing 232U to 233U cancentration cne has

 

 

N - At
B U‘@ )
N2y T
New ot nt
Xy "(cz:\j ( N75 ','G)u -/317_ 4’/‘57_2 2 —/Jn €
+ e ———

 

./b\\ /'931 /-17-1 —fill

 

- ez : Aat Pt
L o1z Yozy (M‘/N'u)?\ Ny -foia + /Ju_& P - /5.3 < ®

/5‘3 /57—1 /b.r, '/bn,

. \‘&—fl?u.l" -

fu

 

l-e”i%ll.r' | — d_flh t
—_— - .
2 2.
/3 - P
i__e"‘fil)_t] (l - e——fllit }
' Nez [ /5’-!- /3
+ Ky O(O'z.\{ (’_._;.C__ > —_—
N /9 3 - I‘GLL

by oo (2 )

N3

246
 

3 —( 2
and H(_rg’i,az): &(.5,_)~C(1)

 

get :
hJ-Zj. - 0(2_5 G"(_bz& + \i” ;XOZI'{ (ifi) ‘+ (,”6“1 ./321.)

23

+ oy Negy ('“:]'a._;\ H (,5a3) ‘,5;;) (3;5°4-7)

e can now eXplore r122/1\:23 buildup as a function cf 1\102/'1\123, ie€e

0T/ Jas £ (2 00)

To calculate the radicactivity we construct and seolve the decay
equations for the 2320 daughters in Figure 3.5~4. 221’&3, 220&:1,
216 2121, . .

Po, and b all decay relatively instantaneously, so they can
be skipped. For a fuel ocutside the reactor with a given 25 2U con=
centration one then has |

60.6 m
2321] _71.1'_.,.228'[-1-1 .l_.?y‘ha_;zl%i 0.50 208T1 S 06 MeV

6006:&
0. 64

For vurvose of simplification, we take the 232U concentration here as

constant, The differential equations are then

dN5g
= >\22 N2 = Mog Nog
dT

247
\
>0 Fos = *32 Va2

> ol
m N
i

== 0.36 \/\32 Ny, - \)\18 Xg
where

N,, = cecncentraticn of 322'[}

Tog = 'f '_'. 338%
1732 = " " 21 %i

33

Ng = " " 2087y
7\2_2 = 1n 2/72y = 3.1 x 19710 s;'1 |
)\08 = 1n 2/1_.9y - 1221070 &
)\32 = 1n 2/60.6m = 1.9 x 1074 s"{l

)\18 = 1ln 2/3m=3.9x10-3 5_1

Taldng Laplace transforms

Q

.s’”,g "4’(,3 = 0.36 )JL ’v!sz _)lB 44!8

 

Mg = —L {%;+oa@haflfl}

S+ hig

sfl(fl "4/[; = )«oe q‘{og - >\3’L 4{32_

248
T = o5 o+ s
S Mag - /L(Dg = A My = Ao Nos
w5l ey |

S /Y\YL - ’flz?_ = = e /Qu_
Ny rr\

Mg

1

 

S+ )«ZL

 

. -t o ) 2 ° Ao °
# e - S+ e g%" SOk [%52 v 22 (M

 

S-l— )\31 S+)Ot
©
+ >z1. 447_2 )] K
= Mo oMy
S+ Mg (s+hg)(s+ Afi->
+ D2k ),31 Xoe oz | 036 V32 do N2z /Ul?:

 

 

(54-),5) (s+)3,_>(s +)og) T 3(S+)‘\B>(5+)‘52)(5+>:‘:>

There are an infinite set of possible situations and correspen-
ding solutions associated with different reprocessing schemes. Cf

most direct interest is the situation where

a o
'vl(go =My = Mg =

249
Then uo
2\ .,
S= Mg NIS = 03 \/*bz )\ofb )n. )l% ,T;) ’ “‘!,13

.__.L_..._ — { fi(\m,‘)\m_,)efl) + Fz()‘nu)'fi: A"B) +§()\fi1>—5%>’%)§
>‘8 >31)53

vhere
F(a,b,c) = e-at/(b-a)(c—a)_ a

S = emission rate for 2, 1=iieV 208'1‘]. S2mmas

208 - . . c
and the units of )\lSN 18 & < Tl-decays/s. Substituting in I\a23=
23 233

19 gm 233“U metal cm-3 x 6.02 x 10 U atoms per 2333 233’6 R

[)‘;.g , one gets a source rate

23
Se19.x 22922307 036 x 1.9 % 1074 x 1.2 x 107 x3.1 x 10710
{ 233
— —--—"""’__— F,;
N23 AIB )3.?. )DB 1=t

—4 V- o0.049 [ 3
S(scmj) N 22 _____.}___..—-ZF

N—2-3- >1e )32 >‘D‘B

Note that for at least one Fi the expanential numerator should
be near cne and its dencminator very small so thatZFi is large,

Section 3.5.11.2 evaluates S for situations of interest.
31.5.5 Pertinent Physics Metrics

3.5.5.1 BG: actual breeding gain. The canventicnal definition

of BG is

rate of fissile production

BG = rate of fissile destructian

(345.5-1)

The question then arises: which materials are fissile and which
fertile? Conventian says 23 L'U is fettile, but in the very hard
HCFR neutron spectrum it behaves like a fissile: G exceeds Oy over

most of the neutrecn spectrum (Fig. 2.1-5), suggesting a bare

237 2L05 (159 kg),

or “*n (114 kg) {1351, 232y and 8U cannot sustain a chain
233:‘—' 23-4-1}', a-n.d 235" sy

critical mass smaller than that for Np (69 k&),

rzaction, Thus we lacel U as fissiles in an 0932,

232'3?3:1 and 238‘6 2s non-fissilcs,

BG can mean the instantaneous BG at any givén point in time (of
the burnup cycle) or a time-averaged BG such as over a given LMFBR re-
fuelling cycle or reactor life. As the survey nature of this study
precludes a detailed fuel management analysis, this study deals

anly with instantaneous BG, presumably on equilibrium cycle:

BG = Th capture rate

rate of 23 3“U ' 23 l‘U, 23 5U fission

This certainly differs in <°U treatment from that for a thermal Th/>>-U-
MSBR: however each reflects the physical realties of the correspanding

neutron spectra, and this definition does agree with equatian (3.5 5.-1)

above.

251
 

3.5.5.2 BGX: BG extended to zero neutron leakage. BGX
gauges the BG potential of a reactor design in the event that
neutrans which presently escape from the reactor (outer) blanket

(CB) were instead captured there in the same proportims as

present.

/ L;total abs rat?J

~

Th capture -]
BGX = (1+BG)§1 J-_--—._-— £ Z -1

OB

This metric thus indicates what additional BG can be obtained

through a larger blanket.

3.5.5.3 BGP: BG potential asscociated with the core spectrum.

The theoretical limit to BG forms another valuable metric of an MCFR(Th).

Taube calculated previously-unheard—of BG = 0.7-0.8 (BR = 1.7-1.8)
for an MCFR(Pu)e
Section 3.1.1.1 defined BG potentizl as the nfimsgr;of neutrons

available for breeding per 233U fission. That igrpored all actinides
5 232

excep Th .and 233U. The net result depended primarily on 30r

for 233U and indicated 2 potential for BGs up to 0.4 on the 232Th/233U

fuel cycle,

The ideal situatica would be for 232Th to capture neutrons which
leave the core: 1i.e. for none to leak out the blanket nor be lost to
parasitic capture, Then BG = Vy -2, However, for the MCFR here-we
must return from utopia and allow at least for parasitic losses due

to necessary core salt components. This then gives

252
 

E (parasitic captures)

BCP = (V) =2) core .

B

reactor C s

Zreac tor (f:r.ssn.ons )

3.5.5.4 d) « The core—averaged total flux will primarily
core

indicate the rate of damage to, and thereby replacement of, the

graphite tubes.

3e5e5.5 (E c>ore<é « The median core flux energy is found as-

 

suming a constant lethargy flux within a group's boundaries,
This metric proves to be a valuable barometer of B&® (Fig. 3.5-0)

since V and O increase wi:h fast neutron energy while & decreases.

3.5.5.6 Power densitv. Together wita ( E>core ¢ , this metric
determines @ core® 1t also affects heat exchanger design: the

higher the power density in the core fluid salt, the faster must heat

per unit fluid volume be removed in the exchanger.

3,5.5,7 @ (vessel), total and > 100 keV. Unlike the core tubes,
the MCFR design and operation plan does not readily tolerate the re-~
pla.cemént of the reactor vessel, Therefore the neutron damage,

especially measured by the flux above 100 keV, should not exceed ].Ol3

-2
ncm  sec - or so.

253
Flux-
Spectrume
Averaged

(P =27

0456

0655

 

Flux~-Spectrum-Lveraged Energy (keV)

Figure 3.5=6. Variaticn of Utopian BG (Iznoring Parasitic
Losses and Leakage) with Core Flux=Spectrum

Gardness.

254
 

3.5.5.8 Fuel inventory and doubling time. Doubling time (DT) provides
a pertinent measure of the economics and physics of a breeder reactor:
the time required to breed encugh extra fuel to start up another re-
actor. That should really encompass the whole fuel cycle: core +
blanket + primary loop + ineplant stockpiles + transport + fuel fab-
rication + fuel reprocessing + out=of=plant storage.

However, perhaps because solid fuel reactors all require similar
ex-reactor fuel cycies, DI conventionally measures just the time for
duplicating core and tlanket. To facilitate comparison, this study
adopts that definition also except that the MCFR must also include
its full primary circuit cantents:

Ui + U0 + Ub

DT = (3.5-5‘2)
BG production rate

Here
U; = U content inside the reactor vessel; more precisely be-
tween the ex—vessel bends of the tubes (Fig. 3.2-9,-10,
and-11)
U =

o = U inventory in primary circuit beyond (outside) the
reactor vessel: piping, pumps, and heat ekchangers.
Section 3.4 indicates the volume Vo = 23 m3.
Ub = Blanket U content
This should result in a conservative comparison of the MCFR
fuel cycle to the solid fuel cycle of other breeder reactors: solid
fuel will require much more inventory in storage, transportation,
decladding, and fabrication operations; the MCFR, little. Both

cycles should involve similar amounts of chemical processing, but

255
 

the MCFR goihg from liquid to liquid state, may require fewer steps

and therefore less U.

The odd geometry of the skewed tube configuration increases the
in-vessel U content, Uj} considerably beyond the calculated critical
spherical mass, U_. If there was only ane cylindrical tube (N=1)

then the volume Vi assoclated with Ui would be

Vi(N=l) = 'n‘n‘?L/z+ = VS/SF(L/D) (3e5¢5=3)
wiiere

v

s critical volume in spherical geometry

SF = L/D=dependent shape factor (Sece 3.5.1.2)

L = critical cylinder length. With a single cylinder, L = H

H = distance along tube between bends. Considering the 2
meter blanket surrounding the core, we estimate H = 5 m

D = critical cylinder diameter

Rearranging Eq. 3.5.53, get
L
L/D = (TrL3 SF(L/D) / & V) ' (3e545=4)
Since SF depends on L/D, Eqe 3.5.5=4 is transcendental: it must
be solved iteratively or by graphical intersection technigue.

If N> 1 and the tubes lay sufficiently apart from one another
(lcosely coupled), then ' (N > 1)=N vy (N=1). However, this is
not of interest because we prefer strong intertube neutronic coupling
(Section 3.1.3.2).

With strong coupling, if N tubes were not skewed but lay ad-

256
jacent to one another, then their combined cross sectional area
would appreach that of a single large tube, then

Vi(N >l) = Vl (Nzl) Of Eq. 305-5—3.
In contrast, when skewed the maximum, the tubes closely approach

ane ancther only near their midpoints. Then no tube sustains the
neutran chain reaction very well ogtside the close-approach zone,
The critical geometry of that zone should resemble a square cylinder
(L/D=1), somewhat squeezed at its middle into the shape of an hour
glass or a wheat stack (U).

The large mean free paths of fast neutrans tend to negate the
"tubular" protusions of core salt from this geometry: next colliéions

will probably occur in the neutron-absorbing blanket. Then with

L/D=1
2
V/SF(L/D=1) = NWD, D, /4
wherse
Dt = diameter of a single tube
D, = (bV /N T sF(L/p=1) )Y/
Vi(N >1) = NTH th/‘* = H((vs/o.gv)2 N '[T/A)l/B

0.94 H (N vs‘g)l/3 (34545=5)

Figure 3.5=7 displays the results of Eq. 3.5.53 and 3.5.55.
The di.i‘;'erences in V]-_- correspond to the differences between zero and
maximum skew angle respectively, Thus the true results would lie
intermediate, For the present study we take Vi at some value of low
significant figure slightly below the Eq. 345~7 line (for N=3), ' This
should result in a comservative estimate of DT, To get Uy or U we

miltiply V. or V_ by U_/7 .

257
 

 

 

 

 

 

 

 

 

 

 

 

 

A
(m®)

 

 

 

 

 

 

 

 

 

 

 

 

Figure 3.5=T. 7Tariation of In-Vessel Primary-Circuit Volume
with Critical-=Sphere Volume for Maximum~ and

Zero-Angle Tube Skew

258
3,5.6 Fuel Cycle Modelling

3.5.6.1 Fuel cycle model. Figure 3.5-8 describes the actinide flow
scheme. (Section 3.7 describes the chemical aspects of the fuel cycle.)
Th supply to both primary circuit and blanket maintains the Th con=-
centrations. U exits with fractiaonal removal rates p and b respect-

jvely, passing into the shim tanke b 1is actually a net removal rate:

part of the removed blanket U returns.

| fi@v

' Primary BlanRet

Circutr

| (xnel. core)
8

f d}— t bus)
&

Fig. 3.5-8. Actinide Flow Scheme

 

 

 

 

 

 

 

 

 

 

 

Fractian £ of the shim tank inflow feeds the primary circuit,

simultaneous with pUp and fission product removal. The remaining

fraction g = 1=-f is exported as BG.

Extending Sec. 345.4e1, U varies in the blanket according to

259
aUy /a6 = < Ome ) Th B Uy B Uy (3.5.6-1)

3b
de_b/dt = (Cwbdy Upy = (Fsad » Uy, - BUg, (3.5.6-3)

Simjlarly in the primary circuit

Wy /dt = (Gor D) Thy = (5,6 > g Uzg = Flgy
+ 2(pUy, + bUsy) (3e5.6-4)

dilbp/dt =S ) Ugg = (5D o Us = g
+ i'(ppr + bth) (345.6=5)

QU /dt = B &) Uy = Eab D g Usg = PUg,
. ‘%

+ i‘(pUsp + bU5b) (3.5 )

Here the subscript s denotes the spherical core region. The in-

tegrated reaction rates may differ slightly in few-tube geometry,

but they can be adjusted later to reaction rates from assemblies of

higher Th concentration.

We define here

U, = U35 + Ul;s + U55 (3.5.6=7)
U-b = UBb + Ul{.b + Usb (3.5.6_8)
Usp= s vp/vs, i = 3,4, or 5 for 233y 23“1.!, or %,

€ =V Vs
Vp: Volume in the primary circuit

VS= Volume of a critical sphere

260
 

Then, substituting in Eqe. 3.5.6=4 through =6 for the {Uip} -
b T3
T = { Ty ¢>s Ths" <6.34¢>5 UBS- pSUBS + f(pSUBS +D UBb)
¢ Dus
at = <S3\'¢>5 UBS- <6u¢>s UAS- pSULS + f(pSULS + b Uh.b)
5 5 £ S
= (‘Siwq’>s U™ <55Q¢>$ Ugem PSUg, + (g Ugg + b U5b)

3.5.6.2 BEquilibrium fuel cycle. We now study a specific situation of
interest - the equilibrium cycle where compositions have stabilized and
{cfigj /dtz each = O. The starter fuel does not affect the equilibrium

composition; only the feed material and neutron spectrum control it.
This section seeks to determine the relations which express

1. The equilibrium proportions of actinides in primary circuit

and blanket

2. The equilibrium breeding gain and/or reactor doubling time.

ANISN calculations provide the reaction probability rates (per
atom) {6 ¢>) « In addition we specify the core uranium concentration,
and blanket thiciness. Together with a critical core radius search,
this fixes Us ,Ths, and Thb. Nine unknowns remain: UBs' Ubs' U5s'
UBb' th, Ugps f, by, and c. There are seven equations, 3.5.6=1 =
3.5.6~7. That suggests that we also choose the reprocessing rates b
and c¢; fihese will then determine f, and from it,the BG export and
doubling time.

Now setting dUij/dt = 0 for each i and j, then from Eqe. 3.5.6=1
through 3.5.6=3, in t.hé blanket:

261
Coqr &)

Uy = =e————eee—ee T (3.546=9)
3b b + <O—3¢L¢>h h.b

Gy O
U, = _ S Ugp (3,5.6~10)
4b b + <dkd>>b

< O_“Y¢> b
U = e o U (3e5.6=11)
sb b+ <651¢>h Lb

Normalizing uranium contents to 'I‘hb, get

 

 

 

: &
Ty = COme 7 (233Ub a.toms/?'3 2Thb atom)
b + <63“'¢>b .
{63387,
h 54 =
Lb b + <°-Qc¢>:9 ujb
_ {Cuy by
50 L + <G&b>b “Lb
where
Uiy = Usy/Thy

In the core, by equations 3¢5¢6-4 through 3.5.,6-6, get

, 4y Th. + bf U
Ugg = (G ¢ 25 ™s b (3.5.6=12)

<G:"a"b>s + P‘S(l—f)

 

g U, +bfU
Ul:,s - < ch>5 3s Lb (3.5.6_13)
(Ot +p5 (1-f)

add Yy tbf U
Ugg = (Sud)s ks 20 (3e5.6~14)
<GP +p5 (1-f)

 

262
 

Normalizing to U  and setting c = p&:

_ <5~mr 4:'75 ts Y Y

 

 

 

“35 - <§31¢>s + c(1~-f)
COsy d’ls%s * Y
U=
(Shad)s + ¢(1-1)
0 = {Surd)sUss M Usy
o8 (56 Vg + c(1~f)
Here
Uyg = UiV
by = Ths/Us
o e

Thy/Us

o

Now to calculate these requires knowledge of f» Summing Eq,

345.6=5 through 3.5.6=7 gives a recursion formula suitably weighted over

all the actinide cantents:

f +

°U3+ bUb

Normalizing numerator and denominator to U, get

f =
c+b- u, - t’b
Owr iterative procedure will be to
(1) Choose b
(2) Calculate {uibg and ug

(3) Make initial {“is% guess

263

&
CUS + ; <5'.f- ¢>5 ’ Uis : + GSQ.¢>5 USs - <§1\Y¢>s Th.S

4‘-
c +1§’<°"’ QY Uis  + (O &) Ugg = {Spr P ts

(3e546=15)
 

(L) Choose ¢
(5) Calculate f by Zge 3¢506=15
(6} Calculate corresponding {uisi and ug = Z‘uis
(7) If ]us-l |< &, renormalize Zuisz and u_
(8) Reset starter guess and go to (4) until desired range
of ¢ i1s covered.
(9) Return to (1) until desired range of b is covered.
The BG export rate is
'fie = (1-£)(eU, + bUy) (34546-16)
The doubling time with respsct to reactor inventory is then

Ug

T (1) (U, + b

 

(3.5.6-17)
b)

where

Ut = Ul+ UO + Ub (Sec. 3.5.5-8)'

ANISN calculates U_ and reaction rates (Table 3.5~V1),

U, depends on removal rate b (Eq. 3.5.6-9;10 & 11)

Fige 3.5=7 provides the means of getting Ui from Us

Uo = 2=3 mj, as discussed in Sece 3+5.5.8.

3.5.6.3 On the choice of U removal rates ¢ and b. This section
studies how ¢ and b affect the core = 3U/23 4 ratio and blanket

U content (Ub) using reaction rates calculated for an MCFR model
with 30/35/35 NaCl/ThC1, AUCL, molar composition and 7/1 2223/%2hy
ratio in the core salt. Section 3.5.8 shows the importance of high

2330/231“[} ratio in the core salt. Uy, should be minimized because

264
 

it degrades BG export (68) and increases doubling time (DT); it
also generates unwanted power and radicactivity in the blanket.

Figure 3.5-9 shows how increasing b quickly reduces Uy to nil.
The total U inventory (U,) then comprises just the primary circuit
contents, U; + Uj. Meanwhile, fie rises from oblivion to its max=
imum level. Since Ut and UE vary with b in opposite directions,
their quectient DT (EqQ. 3.5.6=17) varies even more strangly with b
(Fig. 3.510). Core 233U/EBL‘U ratic also varies markedly with b in
Fig. 3.510.

All these comparisons argue for b2210-7, but not much beyond
thate Fige. 3.511 shows thzt high ¢ keeps 233U/231*U high and only
slightly affects DT adverse_y. It seems to warrant .c as high as
10_6/5, but not crucially so. Note that ¢ = 1077 implies a primary
Up removal rate of p=¢/% = ¢ Vs/tp. Thus the requirement on p will
typically fall a factor of 10 or more below (be far less stringent
than) that on c.

A rate of 10—6/5 implies the removal of 1 ppm of the U per second,
or a complete processing of that zcne every 11.6 days. Though that
may be doable, a rate of 10-7/5 or full processing in 116 days seems
more reasonable while imposing no severe penalty.

Detailled analysis in other studies shows the same b and ¢ be-

havior patterns there. Thus we adopt b=c=ld-?/s as a reference for

this woric, We also conclude that physical reprocessing limitations
should not prevent us from achieving near-optimum 233U/ZBLU ratio

and DT and small Ub.

265
 

 

T 10°° 16" 1o 0 w

U, Extraction Rate, b (Fraction/s)

Figure 3,5-9, Sensitivity of Blanket Uranium Content (Ub) to

Its Extraction Rate (b)

Bagis: ANISN calculations of critical size and reaction
rates of a core mixture with 30 mole % NaCl, 35

mole % TaCL,, 35 mole % UCL 233y/234y ratio of 7.

3!

266
 

(yrs)

Figure 34,5=10
Sensitivity of DT and
233y/234y patio to

U5 Extraction Rate

233,
234,

~

- Based on ANISN
calculations. of
critical size and
reaction rates of
a core mixture with
30 mole % NaCl
35 mole % ThCl

35 mole % 0013
233U/234"U ratio of 7T

4

¢ = core U extraction
rate (fraction s )

 

 

 

 

 

 

 

 

-4

 

10
10"11

 

 

 

104

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

US Extraction Rate, b

(fraction 8-1)

267
 

DT : | z

- b = 1078
Fige 3e5=11 _— o
Sensitivity of
Doubling Time and 60 -
233y/23% Ratio o - |
Core U Extraction L
Rate
40 < e
| — b= 10'%
————-—‘-——/-‘ ; =
233y

 

Based on ANISN
calculation of - —
critical size and - - -

 

T
reaction-rates o? ‘ 10-11 10—9 10—7 10—5
a core nixture with
30 mole 7 NaCl
d
35 mole % ThCl4 Core U Extraction Rate, c

35 mole % ucl,

233U/234U ratio of 7
b = blkte U extraction rate

(fraction /s)

(fraction 5-1)

268
3.5.7 Reactor Design: Configuration Trade Studie

3,5,7.1 Inner blanket study. Taube studied [17] inner blankets up

to 1.1 m in radius for spherical MCFR(Pu) and (fluoride) MFFR(Pu). 1In
each case the outer blanket remained cne—meter thick. With pure fissile
(no fertile) salt in the core, the BG slightly increased ‘for an MCFR(PU)

and decreased for an MFFR(Pu). Fig... 3e5~12 shows the changes by zone in
the MFFR(Pu study: inner blanket breeding increases with its volume,

vat also softens the core neutran spectrum. The result is increased

core BG but lower total 3G, more so for F than Cl s2lt. Thus for a

  
   
   
 

 

 

 

 

 

 

 

0.207
' 3¢ inaer bl.
0.10
Helative BG care
BG
0.00
010 BG outer bl.
R =160 cm R =280 cm
. I 1 o 1
0 1 2 2.37 3

Relative Inner Blanket Volume. .

Fig. 3.512 Impact of Inner Blanket Volume on BG far a MFFR(Pu)

269
 

fixed reactor dismeter, which controls the overall plant size, an inner
blanket does not appear worthwhile relative to just increasing the out-
er blanket thickness. Furthermore even if marginal BG advantages were
found, the added engineering complexity wofild probably discourage it.

To verify the effecfi of an inner blanket a study was made in spher—
ical gecmetry. Figures 3.5-13 and 3.5-14 show the results for‘a
censtant total blanket thickness of 200 cm. As the IB diameter increas-
es, the annular core moves radiaily cutward. As a result

1. The core volume increases, decreasing the power density

2. The core spectrum softens

3+ Neutron leakage from the reactor increases

Le Inner breeding increases and cuter breeding decreases resulting

in a strong net decrease
The accelerating BG decrease is primarily caused by the increasing neu-

troan leakage and secondly by the softened core spectrum effect on BG

potential,

3.5.7.2 Cptimum outer blanket thickness. A study was made to determine

what constituted an infinite (cuter) blanket thickness to BG. It assumed

no inner blanket. The study cansidered ThClL/UClB molar ratios in

the core of 0,1, and 3, The NaCl molar content in the fuel salt

was 30%. The results (Fig. 3.5-15) indicete that 2 m is near in-
finite.

Figure 3.5-16 shows that the ocuter blanket also shields the
vessel from the high core flux: a two-meter thickness reduces

the damage (> 0.1 MeV)} flux by over 3 magnitudes do:n to the level

of the delayed neutrons.

270
 

 

 

Median
Core Flux
Energy
(MeV)

 

 

 

 

 

 

 

 

 

 

BG

 

 

 

0 40 80 120 160

Inner Blanket Thickness (cm)

Figure 3,5-13, Decline of BG and Median Core Clux Energy
with Tncreasing Inner Blanket Size

Bases: = 30/35/35 molar proportions of Na/ThCl4/UCl3
in the core

« 200 cm total (inner + outer) blanket thickness
- Ni reflector

3.5.7.3 Reflector. A pair of ANISN calculations determined the
effect of reducing the reflector from 4O cm graphite to 20 cm

USinE:a,keff"Precision of 0.001 or 0,1%, P1 cross sections, and

1

S8 quadrature. With a 2 m blanket, the BG dropped from 0.3448

271
 

 

 

Neutron Flux SRR DU r"___ T
-2 2 : P N7 total
(n em ~ sec ) - | BNYLAN]
in Reactor B b \t o s
AN NN
fofi/ | 7 0,1 Mev

Vessel Wall

   
 
 
 
 
 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Outer Blanket
Right Leakage
(neutrons per
fission source
neutron)

0.01

 

0 50 100 150

Inner Blanket Thickness (cm)

FPigure 3,5-14, Increasing Neutron Leakzge into Reactor
Vesgsel Wall with Increasing Inner Blanket
Size,

Bases: - 30/35/35 molar pronortions of NaCl/ThCl¢/UCI3

in the core

-« 200 cm total (inner + outer) blanket thickness
- Ni reflector

272
 

 

 

 

 

 

 

 

BG

 

 

 

 

 

 

 

Blanket Thickness {(m)

Pigure 3¢5~15. Determining the Outer Blanket Thickness Which
Is Quasi=-Infinite to the BG, for Two Core
Salt Mixtures

Bases: blanket salt = 30/70 molar NaCl/ThCl 4
~ 40 cm graphite reflector
<~ no inner blanket

- 30% molar NaCl fraction in core salt

273
 

 

 

 

104

Neutron Flux
in Reactor
Vessel

 

 

 

 

(n cm-gsec-T)

 

 

 

 

 

10'3
1ol2
Total Flux
(ThCl¢/UCl3 = 3)
10! 0e? MeV flux
‘ - (Thel,/ucl, = 3)
0 1 2 3 4 > 4 3

Blanket Thickness (m)

Figure 3.5~16, Typical Decrease in Vessel Flux and
Damage with Increasing Blanket
Thickness, Graphite Rellector

ThCl4/U01 Molar

 

 

Legend Symbol Ratio in gore Salt Flux
0 - 7 01 MeV

3 total
+ 3 > 0e1 MeV
A 5 > 0.1 MeV

274
to 0.3442 or 0.2%. The critical radius increased from 26,11

to 26,1, cm or 0.1% (a critical volume increase of 0.3%). Thus
cutting the reflector thickness in half barely shows up above the
accuracy of the calculaticns. 1his result cerresponds well to that
of the previcus secticn: namely that two meters censtitutes a

near-infinite olanket thickness.

Comparisan of C and i reflecters far a 3/1 Th/U core ratio and 5m
thick blanket showed no detectable differences. For a 0/1 Th/U core ratic
and 1-m thiek blanket, C performed much better (BG=0.106) than Ni (BG=
0.0L0); the critical core rziius differed little. The BG difference
probably stems from the moderating power of C: neutrons reflected by C
stand a much better chance of capture by Th than those reflected by Ni.
This recommends C for cases where.a non=infinite blanket is used. For

> 1.5=2.0 m thick blankets and 65/35 ThClh/NaCl mixture, a neutron re—

flector appears unneeded.

3_.5.7.4_ Core tube material, In an MCFR (Pu) study (751, & 20

MO/SO Fe alloy for the core/blanket wall reduced 3G by 0.04 relative

" %0 a C wall., As the 20% Ho dominated the neutron absorpfions there
(617), 100% Mo tubes might Teduce BG by C.2. The MCFR(Pu) study
assumed in~core cooling with the tubes occuvying 5.9% of the core
volume, In the present MCFH(Th) concept the 2-cm tubes occupy several
times thai proportion of volume but they lie more on the periphery.

Thus a 3G effzct similar fo 0.2 might be expected.

2775
 

3.5.8 Reactor Design: Controlling the Core 233U/231‘U Ratio

Je5.8.1 Basis for studying the ratio. 23 I"'U cantent affects the
Section 2-adduced goals: hard neutran spectrum, high BG, and mine
irum presence of Pu and other heavy actinides. Although 23 L‘U
cantent is not a design (independent) variable but rather a physics
consequence (dependent variable), still design variables like NaCl/
UClB/‘IhCJ.l+ molar composition can influence it. Therefore we study
its effects as if it were an independent variable, loocking at 233 v/
23y o 3,45y and 7.

The MCFR model used as a basis in the study included a core salt

molar composition of 30/70/C NaCl/UClj/ThClh and a 200 cm blanket,

3.5,8.2 Effect of the ratic on spectrum, BG, and Pu presence.
Fig. 3517 shows that 23l‘U softens the neutron spectrum (( E} core
decreases. This is probably due to a higher {G’inel) /<G 56>
ratio for 23l‘U canpared to that for 233 Ue 23 LU also reduces BG and
power density. Table 3.5IX shows that the major BG reduction stems
directly from the lower (’U(S;-— Oc_> value for 23,4‘[1, not from spectral
softening. This same phenomenon decreases the reactivity of high

23 LU systems: that accounts for the larger critical volumes and lower
power densities.

Intuitively, a high ratio will minimize the presence of Pu and
other heavy actinides (the third goal): 234 heads the chain that
leads to them.

In summary, a high 233U/?'B}‘U ratio promotes all three goals of

hard neutron spectrum, high BG, and minimal Pu presence.

276
Average
Core
Flux
Energy
(keV)

Power
Density

Breeding
Gain

Figure 3.5-1T7. Correlation of Decreasing 234U Content With Key Core

880
860
840

820

27
26

25

23

0.34
Q.33
Ce32

0431

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

233U/234U Ratio

Physics Parameiers

 

271

 

 

 

 

 

 
Table 3.5-7% Calculation of Neutron Procduction by the Two Principal

 

 

 

U Isotopes
233U/231»U value for (V) 6c = 6a)
Ratio 2y 2y _EEEE_
3 3.27 1.16 2.74
5 3.26 1.18 2.91
7 3.33 1.02 3.04

3.5.8.3 Zffect on reactor fissile inventory and doubling time.
Because 23LU is a fissile MCFR fuel, the reactor U inventory varies
little in this study. However, the U export rate (Zq. 3.5.6~16) does
increase with decreasing 234y content, reaching 0.25 MT/yr for the
ratio 233U/?'BAU = 7. Figure 3.5-18 shows the consequent effect on
doubling time. This recommends a high 2'33U/231‘U ratio; however,

significant effort to reach higher ratios may not be warranted.

273
 

 

DT (years)

 

 

 

oL

30 4 =i e
3 5 T

 

Core Sait Z*u [P Ratio

Figure 3.5-18. Variation of Doubling Time with Core Salt 233U’/234U
Ratio
Basis: 30/70/0 NaCLfUCl3/ThCl4 molar szlt composition in

the core, 200 cm blanket, C reflector; b = ¢ = 16-7/3

3454804 Deducing the ratio on equilibrium cycle. Given a set of
reaction rates and choosing b and c, the equations of Sece 345.8.2
specify the 233U'/234U ratio, However, this ratio for the fuel cycle
will depend on the reaction rates used. These, in turn, depend on
the ratio assumed apriori in the neutronic analysis of the equil-
ibrium fuel cycle. Figure 3.5-19 suggests that the two converge
near 233U/234U = 95 for b=c=10"| and near 123=13 for b=c=10‘6. This
is for a 30/70/0 IIaCl/UCl3/ThCl , molar compositions Similar behavior
is presumed for other MCFR models.,

The results of later calculations with the refined set of P1 Cl

cross sections (Sece 3.5¢342) suggest a slightly lower value, nearer to

233U/234U = 8% and 115 for b = ¢ = ‘IO-7 and 10-6, respectively,

279
 

 

 

 

-6
Core 233U/234U Ratio =c=10 /s

oa Equilibrium Fuel
Cycle Deduced

 

 

 

 

 

 

 

 

 

from ANISN-Czalculated 10 _“:-“_:5{“:;:::_5;_:7:_'_'
Reaction Rates e b = C = 10-7/5
8 e ——
3 p) T
Core 233U/234U Ratio

Assumed in Calculating Reactiion
Rates with ANISY

Figure 3.5-19. Calculaiional Approach to the Equilibrium Core
233U’/23‘1U Ratio
Bases = 30/70/0 molar p-oportions of Nacl/ThCl4/UCl3 in the

core, 200 cm blanket

3.5.9 Reactor Design: Choosing the Core Salt ThClL/UClQ Ratio
S -

To approach its BG potential (BGX or BGP) the reactor must
capture its excess neutrons in fertile material. One means of re-
ducing neutron leakage frem the reactor is to thicken the ocuter
blanket (Sec. 3.5.6.2) so that the fertile fuel captures or reflects
most of the neutrons which escape the core. Another way is to add
fertile material to the core fuel thereby reducing the core neutron
leakage. However, this also softens the neutron spectrum, which
lowers the BG potential (BGP).

With an MCFR(U/Pu) Taube reported [17] BG up to 0.7-0.8 with such
techniques. The following subsections study the optimum fertile

content in an MCFR(Th) core. They assume a salt mixture of 30%

280
 

NaCl and 70% (UCl3 + ThClh) and a 233U/231‘U ratio of 7/1.

3.5.9.1 Effect on BG. Fige 3.5~20 shows the results of ANISN
calculations. Neutrons leak significantly from a one-meter blanket
so the choice of reflector affects BG. Then cne prefers ThClk/UCl3 =
C.

When the blarnket thickmess exceeds 2.0 m however, few neutrons
leak and the reflector has little effect. [Th] = O then produces
the greatest BG because it engenders the hardestneutron spectrum
(Fige 3.5-21) and consequently, the most excess neutrons (Sec. 3.5¢5¢5).

Fig. 3520 also shows the BG from the 1956 ORNL study [13] of an
MCFR(Pu) which included a Pb core-reflector and a graphite-moderated
thin blanket. The larger ¥ value for Pu causes the high BG. Few
neutrons leak despite low fertile content, but this is because of
high (83%) carrier salt content. BG could actually be higher were
it not for the spectrum softening and neutron absorption by the
carrier salt.
3.5.9.2 Lffect on reactor fissile inventory and doubling times.
Substituting ThClh for UCl3 decreases density, However, the critical
core radius increases near proportionally, and volumé increases even
faster. Thus, overail, Th dilution increases the spherical critical
mass U (Fig. 3.5=22) and the corresponding mass in the tubes Us
(from Fig. 3.57). In contrast, the primary circuit volume outside
the vessel, Vo’ remains fixed, so U, decreases.

While V_ exceeds V., U,=U;+U _+U  decreases. (Ub plays little
role while b210-7.) Eventually the Th proportion reaches a level

where the internal volume Vi exceeds the fixed V,. Then Ut begins

281
 

 

 

 

 

 

L T e . s p—

~__Pb and C,Z0 cm U0, HCFR(Pu) .-

 

 

Breeding

 

 

Galn

 

 

 

 

 

 

 

 

 

 

Fertile/Fissile Ratio in the Core Salt

Figure 3,5~20, Effect of Fertile Presence in the Core Salt upon BG:
Lowering of the BG Potential; Trapping of Leakage

Neutrons

Legend: C, 200 implies a 200 cm ThCl, blanket and a C reflector(s).
The MCFR(Pu) reactor features interlayered blankets and
reflector/moderators

© carbon reflectors, MCFR(Th)
x Ni reflectors, MCFR(Th)
A Be and C reflectors/moderators, HCFR(Pu)

282
 

 

900

 

 

 

 

 

 

 

 

 

 

 

 

 

 

800
- Median 700

' Core-—-
600

Neutrecn

Erdergy 5C0

" (keV)
L0OO
300

 

o 1 2 3 L 5

Th/U Ratio

Fig. 3.5-21  Effect of Th Presence in Core Sal;c, Upon Core Neutron
Spectrum
to increase with increasing Th/U ratio. This U -minimum occurs
near ThClh/UClB = 1-2,

In the 1956 ORNL study [13] of an NCFR(Pu) the minimum plant inven-
tory occurred a.1:: a 238U/Pu ratio of 2 (Fige 345-23), similar to this work.
The MCFR(Pu) total fissile inventory however lies roughly a factor
of 3 below that for the present MCFR(Th) study: the reason is great
dilution of the Pu fuel mixture by carrier salt(83% melar content).

In Fig. 3.524 the BG export rate fJe for b=c=10-7/ s (defined
“ by Eqe 3.5.6=16) stays fairly constant as the core Th/U ratio changes.

By Eqe 3.5.6=17 then, DT behavior must closely follow that of U L
Fig. 3.5-2l shows that the lowest DT clearly occurs at a ThCl L/UC]'B

ratio £ O: somewhere around 1-2.

283
 

 

 

 

     
  

U
Inventory

(u) T

 

| . . X
! == primary circult
. outside the-
;rfikfij vessel

|

T,
|
|

Te0

 

 

 

spherical
; core
1 calculation

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Th/U Ratio in Core Salt

Figure 3.5-22, Dependence of the MHCFR(Th) Reactor U Inventory

upon the Core Th/U Ratio,

284
 

 

 

 

Total Plant

2395,

 

 

 

Inventory

 

(+T)

 

 

 

 

 

 

 

 

 

23811/‘23 Pu Ratio in Core Salt

Figure 3,5~23, MCFR(Pu) Plant Pu Inventory with a 3 NaCl,

2 MgCl,, ! Ucl3 (Puc13) Core Salt,

Ixternal holdup volume = 3,5 m3

285
40 3

 

 

Doubling
Time, DT
(years)

 

 

 

 

 

 

 

 

BG Ixport
Rate, Ue
(tr7/yr)

 

 

 

 

 

 

 

 

 

Th/U Ratio in Core Salt

Figure 3,5-24. Variation of BG Export Rate and Consequent

Doubling Time with Core Th/U Ratioa

286
3.5.7.3 Effect on power density and flux levels

The power density limit of 10 Mi/liter (Section 3.2.2.2) reflects more
a premonition cf problems than a recognition of any specific. ones,
Still, flux levels in the core tubes and the pressure vessel closely
relate to power density (section 3.2.2.3). Figure 3.5-25 shows

how ThClA dilution of the care salt markedly reduces power density
and fluxes alike.

The skewed-tube geometry will further increase the critical mass
(lower the power density). A cylinder with length-to-diameter ratio
(/D) near 1.0 requires about 10% more critical mass in low density
metal sysiems than does a sphere; with L/D~2, about 40% more, The
heterogeneity of multiple skewed tubes further increases critical
mass as does neutran capture and scatter by the blanket ThClL in
between the tubes. Altogether these effects should reduce power
density and fluxes by more than a factor of two compared to spheres.
Aplied to Fig. 3.5-25, the result is that ThClL/UCl3 = 0.1 now

satisfies the power density criteria of < 10 iW/1.

At higher ThClh/UClB ratios the > 100 keV flux in the vessel

“Lor ¢ 102 1 o2 in 30 years. That

falls below 101‘3 n cmm2 S
level will allow operation over the reactor life for many metals.

Meanwhile the core flux level will be «~--102'3 n cm-z mo—l.

Presumably, most materials won't be able to take many months of that.
This emphasizes the importance of our design: straight tubes, easily

replaced, and not touching one another.

287
Power
Density

(/1)

Flg\ll‘e 30 5-25.

 

 

 

<Y Poore
1 | i b (total)

 

dbvessel
(total)

 

 

 

cbvesscszl

 

 

( > 100keV)

 

 

 

Power
Density

 

 

Th/U ratio in Core Salt

Reduction in Flux and Power Density as Th
Presence in the MCFR(Th) Core Salt Softens the

Newtron Energy Spectrum

Bases: 2 m C reflector, 233U/234U in core

7:1, Na/Actinide in core 30/70.

288
 

 

 

233

 

 

 

 

 

 

10 U
Isotopic e e e

 

 

<ce/oy —
in the N — el
Core Szalt

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o 232y

 

 

 

Th/U Ratio in Core Sals

Figure 3,5=2G, Spectrum Softening of Th Presence in Core Salft:
Bffect Upon the Fission-to-Capture Rate Com-

petition for Key Isotopes

289
 

JeDe%ed Effect on actinide fission rates and 233UZEEAU ratio. The
spectrum softening effect of ThCl4 content (Fig. 3.5;21) reduces the
average fission cross section relative to the capture cross section,
especizally for even-A isotopes which do not readily fission (Fig.
3.5-26)s This favors zccumulation of higher-A nuclides which section
24142 showed to be undesirable., Figure 3,5-27 confirms this effect
on the 233U/234U ratio, Section 3.5.8 showed that 234U presence

further softens the spectruz,.

14 '

——— e :
et e el e ek

 

 

 

 

 

 

 

 

 

 

 

 

 

Bquiliorium
233,23 ]
U/ 4U = C = 10_’6(
= C = 10
Ratio in
Core Salt

 

 

 

 

 

TbCl4/UCl3 Yole Ratio

Figure 3,5=-27, Effect of ThCl4 Content in Core Salt Upon

233U/234U Ratio
Bases: 3055 NaCl mole content in salt

1/1 ratio for 233U/234U

290
 

3.5.10 Reactor Design: Core Carrier Salt (NaCl) Content

Like ThClL, NaCl substitution for UCl3 in the primary salt
significantly reduces the fissile inventory outside the core tubes,
but also degrades the neutronic performance in the core. This
section analyzes some of these physics conseguences assuming a core
salt composed of zero ThCl, and a 2333/234U ratio of 7. Results

are compared with those from ’I‘hClI+ substitution (Sece 3.5.9).

3.5.10.1 Effect on BG. NaCl decreases BG through parasitic abe
sorption and moderation of neutrons. Fig. 3.5-28 combines the results
from the present MCFR(Th) study with the behavior derived from an

MCFR(Fu) study by Taube. The results suggest

1, Generally minimizing the NaCl ccntent

2. Absolutely avoiding NaCl molar content 80%

Fig. 3.5=29 compares NaCi substitution relative to ThClh
substitution on an expanded scale. One sees that ThClh diminishes
BG slightly less than NaCl does. The probsble reason is that most
neutron capture in ThGlh contributes to BG (even considering the
four Cl atoms), whereas none of that in NaCl does. Table 3.5IX
supports thise.

~NaCl difference would prob-

With a thinner blanket, the ThClL

ably be much greater.

2.5,10.2 Effect on reactor lissile inventory and doubling time,

Figure 3.5~30 shows the resulis of this study. Similar to Sec.

291
 

Legend

x deduced from
an MCFR(Pu)
study

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Ll e e
MCFR(Th) ,’ZZ.’[H:". S S L R LT
02 S e gete _,-‘}.-h R e S ..,,..UI"._:
- Tevaeeeil- -t e rem e e ey v o byt o
EG 1R :Llli_:: oo I:{-.——* --.—--;:_:-_ @ Calculated fOI‘
—- ie— e pr—— At | 2 i dnf g [} i '] .
i ::*_".—:.{-_:: i ;._Qi::':_ = S MCFR(Th)
Q.1 |TmmTmT Cmnboimd s e o study
- . - bhefrogmtbeeo
Sl e R e
SoTnigozlomiloiiprs Ly tros
—— e} ———— . e ) e b - i C et ———
TS ST L LI T I L T r T

 

carrier salt
salt Fissile Salt Molar Fraction

Fig. 3.5=28 Drastic Felloff in EG at High Carrier Salt Content

 

 

 

 

 

   

 

 

T -3;*'[&5[32 b ) - Legend
onoaiprni L R TR e e )
CillAns o s I o e T ey as above
0.34 =tCarrier -_.§alp:*_. _”I‘;%‘EB?I.E ik fi
. | M ‘ .
vt VSt A o e
i Agtlfl?,‘:}i“{
hee et i i e
£, alli il
0.33 TaEm L
S

 

 

=y
g
i

 

BG 0.32

Ce

Ll

-
ol =
o b s
Hlimmoo

=

 

B

————

 

 

e =
1

pow oy s

 

 

0.31

e

i

 

 

 

T
)
Tt

 

 

0.30

 

 

tporr— ey
i,i .

-4

o
=
-
by e

 

 

 

 

 

 

 

 

 

 

0.29

O
[
o
-
-
O

Fissile Salt Melar Fraction

Fig. 3.529 BG Comparison of Carrier Salt Substitution to ThCl L

(Fertile) Salt Substitution

292 .
 

Doubling Time - DT
(years)

 

 

 

U
Inventories

(81T)

 

 

 

 

 

 

 

 

 

 

. all
all 0 02 Oad 06 0.8 1.0 fissile

carrier (WaCl) 1 salt

sjlt Fissile Salt Molar Fraction

 

Figure 3.5=-30, Effect of Carrier Salt Admixture Upon Fissile

Inventory and Doubling Time

293
 

3.5.9.2, NaCl substitution for UCly increases the fissile inventory
in the core tubes within the vessel. However, the larger inventory
outside the vessel decreases faster. Thus the total inventory
(primary circuit plus blanket) decreases with increasing NaCl
content.

NaCl substitution also decreases the BG export rate, but only
clightly. Uniike ThClh substitution, NaCl substitution continues
to decrease DT down through 30% UCl3 melar content. However, beyond
that point Sec. 3.5.10.1 indicates that BG should fall off drastic-—

ally, with consequent DT increase.

3.5.10.3 Zffect cn power density and flux levels. Figure 3.531
shows how NaCl substituticn for UCl3 decreases the power density and
flux leyels. However, ThClA—substitution causes an even faster dee
crease. In both cases the blanket composition remains constant.

Thus the greater fall-off in flux from adding ThClh to the core salt

must stem from the higher microscopic absorption cross section for Th

than for Na (Table 3.5~X),resulting in less neutron leakage from the
core.
The difference that ThClL makes relative to NaCl might allow

the use of a thinner blanket.,

3.5.10.4 Effect on 233U/231‘U ratio. Substitution of non-fissile
salts for UCl3 softens the neutron spectrum because it substitutes
a source of elastic~ and inelastic- scattered neutrons for one of
fission neutrons. In a softer sfiectrum, the rate of <My (thresh=

old) fission decreases and 233U (-1/v ) capture increases. Both

224

—_—
Total
Core Flux

 

 

 

 

 

 

 

 

 

 

 

> 100 keV
Flux

=2 =1
necm S )

in Rezctor
Vessel

 

 

 

 

 

 

 

 

 

 

Core
Fover

Density
(mA)

 

 

 

 

 

UCl

y Holar Content (%) in Core Salt

Figure 345-31. Effect of Core Salt Dilution on Flux Levels and
Power Density: X - NaCl Substitution, O = ‘I‘hCl4
Substitution

Basis: 7/1 233U/234U ratio in core salt

200 cm Dblanitet
Starting core salt composition:70% U013,30% NaGl,0% ThC1,

295
 

Table 3.5~X

Some Pertinent Microscopic Absorption Cross Sections

Cross Section (mb )
Element Core* Blanket*

Cl 15 6.5
Na 1.0 1.7
Th 230 11

* As calculated for an MCFR(Th) with core salt mixture at 30/35/35
NaCl/‘UCJ.3 /ThCl , molar composition.

changes increase the ZBAU content. 233U capture increase also de-

creasesthe 233U content, giving a double effect. All three effects

work in coucert to decrease the ratic (Fig. 3.532).

Differences between NaCl and ThClh as salt substitutes appear
to be small. NaCl mostly scatters elastically while ThCll+ scatters
more inelastically. The different shapes of the NaCl and ThClh curves
in Fig..3.5-32 may reflect these differences: Th inelastic scatter

should contribute more to spectrum softening at low non-fissile con~

tent where the neutrons still exceed the inelastic threshold energy.

296
 

 

 

 

 

 

 

 

 

 

T e e
Go T e e 2T
Bquilibrium e e ey T
8. o
Ratio in R Ll TR T
i S P Q TTTTLVTL o = o
Core Salt T L - S
Te x - - -’ -
6. | o Tor
0 20 40 60 80

ucl, Molar Content (%) in Core Salt

Figure 3,5=32., Effect of Core Salt Dilution on the Equilibrium

233y/23%y Ratio in the Core Salt: X - NaCl

Substitution, 0 - ThCl4 Substitution

Basis: U reprocessing rates b = ¢ -7/

=10 S
reaction rates based on 233U/234U ratio = 7 in core salt
200 cm blanket
starting core salt composition T0% UCl,, 30% Nacl,
o ThCl,

297
 

3.5.11 Mutation Effects

J.5.11.1 Significance of actinides which emit alphas. Reactor

irradiation transmutes actinides into long-lived alpha-~emitters:

they increase in atomic number (2Z) and weight (A) through (n,Y¥ )

reactions and /fi —decay; they then generally decrease slowly in 2

and A through alpha decay, finally becoming a stable nucleus.

Alpha emitfers are undesirable because

1.
2.
3.
Le

They pose a long-lived radiobiological hazard
They produce neutrons through ( & ,n) reactions
Some produce neutrons also through spontaneous fission

They destroy reprocessing chemicals

The radiobioclogical hazard wi-l depend upon

1.

The probabili‘h;y cf the isotope reaching the p;lblic en-
vironment in an ingestible form

Its radioactive half-life: very short-and very long-lived
isctopes present little danger

The biological~eiiminaticn half-life; . how leong before the
body removes it naturally

The chemical affirity of the isctope for spaéific crgans
or other members of the bcdy. Thosz which seek out bones
are labelled carcincgenic (cancer causing)

The available energy per decay

The cccurrence of secondary radiations by reactions such
as (oL,n) ffir high energy alphas

The quantity of isctope present.

298
 

Combination or these faciors defines the toxicity of an isotope - just
how much of a practical threat it paces, Table 3,5XI includes a tox~
icity classification along with other pertinent physics information,
The moderate helf-lives of Pu plus their early preeminence in the act-
inide burnup chains warrant their highly toxic classification, Carcin-
ogenicity further emphasizes avoiding their.production,

The problem with (o(,n) reactions as well as spontaneous fission
neutrons is that neutrons penetrate gamma shielding: in addition +o
harming workers directly, they can activate the surroundings, presenting
longer-term hazards.

Yumerous light elements (up through Al and Si) underzo these re-
actions with alphas, Fige 3e5=33 shovws some relative yields, F,
though not included there reacts almost as strongly as Be., F, Li,
and Be are candidate materials for (blanket) salt mix; this phenom-
enon argues against their use unless the neutron sigmal and hazerd
are desired for non-proliferation pur-oses, Section 3.7 mentions the
use of Iy O, ligy 2nd a with various revrocessing schemes; <his
phenomenon would discourage their use there somewhat,

The threshold-like behevior of the (o ,n) neutron yields ver re-

action (Fiz, 345-33) makes isotoves emitting higher—energy alphas part

ticularly hezardous. These inoclude 21231, 212Po, 216Po, 220Rn, 2242a,
236Pu, 2420m, and 244Cm. However, when one considers the half-lives

s 2
and quentities actually present, the principal (<(,n) sources are 38Pu,

2420“1 and 239pu while the main S. F. neutron sources iend 1o be 2420m,

244Cm, 238Pu, 240Pu, and 242Pu. These all really emphasize a strong

desire to a2void Pu and higher actinides,

299
 

Table 3.5-XI Hazard Characterization of Potential MCFR(Th) Heavy

Elements [34, .50, 51, 136]

 

 

Alpha S.F. 1 GHe Carcin-~
Half- Alpha Half- FBR ogenic Tox-
life Energy life Conc.. Bone icity
Isotope (years) (MeV) (years) (C/g mix ) Seeker? Class
Pb=212 10,64 1 ' (beta decay)
Fom212 45 s 1.65
=216 0.15 s 6.779
Bi-212 60.6 8.78, 6.05, 6.09
Rn=220 5506 S 6.2883
Ra=22/ 3.66 d 54,6856
=26 1300 LT8L5 1
Th-228 1.913 5.4233, 53405
=230 TeT+4 L.687, L.621 1.5+17
-232 1,40+10 L.01, 3.95 1423
=231 2441 d (beta decay) II
Pa“-231 3.28—{-4, 5-011—--92y5003 l-l+16
=232 72 503203, 542635 813
-233 1.592+5 L&-0821+1 ho783 1.24+17
=234 2¢44+5 Le7lly 4722 2+16
=235 7,0048  LesO0l, 4365 3.5+17
~236 2,342+7  Leb9hy Lk 2+16
-238 4.468+9 Le196, L.15 8.19+15 I1
Np=236 1,346 (beta decay)
=237 2.144_6 h.788, 1.;..770 1+18
Pu-236 2. 85 5.768, 5,721  .3,5+9
~238 87.74 5.4992, 5.4565  4o77+10 Ll yes
~239 2.41144 9016y 5.1k, 5.10 5.5+415 2.5 yes I
-240 6537 5.1683, 5.1238  1.340+11 3.8 yes
=241 T{g,1 L.897 260 yes
2.2 3.76+5 4,901, 4.857 6.75+1C yes
=244 8.3+7 Le589, Le5LS 6.55+10 Jes
Am=241  p32 5.4857, 54430  1.147+14 0.6 I
-2 7380 56275, 5.23L4 2.0+ 14 7.9
Cm=2 162.8 d 6.1129, 6.0696 6.0%+6  34.0

300
 

 

 

 

 

1t

wt L oe
Neu.‘*rnns )
Emitfed por o] B |k
TIacideal Al
gtlihl .f’E lt—L i N
-y «

9% c

T o

lo'a -

lo" -

'L

lb‘” A 1 2 i L

oL L 3» & 5 6 71

By (HeY)

Figure 3.5-33 Energy-Dependence of Major («,n) Reacticns;[137]

31,5112 Significance of 232U Prodiction., Section 3,5.4.2 derived
the equations for producing 232U and its daughter (principally 208Tl)
radioactivity. Figure 3,5=34 shows =he resulting 232U enrichment in
core and blankei salts assuming the consiani 232Th/233U retios indi-
cated., The time to reach equilibrium conceniraiion in the core re-
: £lects the time constant of 1fi5¢>= 10 mos. estimated in Seces 3¢%5¢4e2.
The several-magnitude higher 232U concentration in the core
compared to blanket reflectc the 1000-higher production coefficient
CX23N in that herder neutron spectrum, This means that fresh blanket
U will contein not only far less rfission product hazard but also far

less 232U hazard., Thus it constitutes the biggest threat to prolifer-~

oo
ation., It is therefore very important to mix it with irradiated core

U so as to "spike it" into a2 less proliferation-hazardous form,

301
Core

 

 

 

 

Blanket

 

 

 

 

Time in Reactor (Months)

232

Figure 3.5-34. Buildup of U in Core a2nd Blanket Salts

302
 

e would like to ectimate what level of radiation hazard this
2321} dauzhter activity actuzlly poses to thefi, diversion, and weapoas
manufacture, The minimum critical mass of 233{1 for weapons use will
nominally be 6 kg [139]. With a U metal density of 19,2 g/cm3 (50],

that implies a radius of 4.2 cme The corresvonding flux from a

sphere of such materizl is given [139] by

S
$=,-[2Ra - (@ -RY) n[(g-2)fa-R)]  for a2R
where S = uniform source rate = >\18 le (see Eg, 3e5¢4-8)
2 = radial distance from sphere center
R = sphere radius
For a = R
¢- =
2

At a distance of 1 meter from the sphere

d) = ;j:m [ZR(loo-t-R)— C;Z,"f-ZDcR + lo“'_fiz) In (zr-z +1oo)]

|00

 

For 2.6-MeV photons the dose conversion factor is 3.6 x 10—6 P.em/";fi'

per unit Y flux [81-, Then the dose

 

_ -7 [ Na { _ 2
= 37 <10 [;}-;aj{xs)u)m ;FEE

303
' 3
N } z |
- -0 Nz ——— F. .
= 40xI0 [ T { )‘\3>32A08 for a = 1

‘From Figure 3.5-34 [N22/N23] is typicelly 10-3. Fize 3.5~35 presents
the results for De One sees that after 10 years the dose from 208Tl

aprracaes

D{surface) = 45 Rem/nour

I

D( 1 meter) = 45 miem/hour

fl

To remove most of this zad other radiation one could chemicaliy
separate the uranium in the core sali{ I{rom the 208T1 and all other
actinides and fission productis, However, this would not separate 233U
from 232U. Then, in only fwo weeks the surface dose would again reach
1 Rem/hour. Purthermore, if only 99% removal efficiency was achieved,
2 3 Rem/hour surface dose from 208T1 alone would be present in the
interim,

From this one concludes that a highly efficient scientific group
with laboratory facilities (such 25 a national laboratory} might be
able to make a bomb without getting a deadly radiation dose, but all
others would encounter difficulty,

In addition to these complications, the tell-tale 208Tl signals
would continue to emanate from the separated waste as well as from the
new regenerating source in the separated U, All these froubles must
discourage subnational groups which contemplate taking and using 233U

for weapons purposes.

304
S0¢€

Dose at

Surface

(Rem/hr)

or

Dose at
Distance of
1 Meter

(mRem/hr)

Figure 3.5-35. Variation of Radiation Hazard from 0.1%

 

100

 
    

0. 01 =
1 hr 10 "mo 10 mo

10-3mo 1 day 1 vk,

232

Time Since Chemical Separation out of U Daughters from Fuel Salt

2325 paughter Content in 6 kg of 25°U

I'ollowing Reprocessing
The effect of such activity on a2 conventional solid Fuel cycle,
would probably be to require costly shielding and remote handling fo
protect the workers, This means in fuel fabrication vlants and, pos-
~ioly, 2t the end steges of drocessing dlanis. 4 molten salt fuel
cycle autoustes the reprocessing far more; the shorter time out-of-
core should also reduce the activiiy due to neutron absorption re-
actions with each member of the decay chain.

Meanwhile, in any breedins zain set aside, %the raediation hazard
will stert adding to that from fission products a2ad actinides and in-
crezse Tor several years, Thus the oiggest problem which 232U ore-
zents in a2 molien salt thoriun cycle should ve transport of oreeding
gain: radiction will make th2 fuel unwieldy (reavy shielding and re-
mote handling) and hazardous, thereby somewhat unattractive to diver-
sion and to uveczpons manufacture by subnatioral groups.

Since 232U and 233U are chemically inseparable this radiation
hazard will contiaue to manifest itself for many multiples of the
232U T2—year halflife, i.es for centuries. 3y leaving i ing

- : y le€e « 3y leaving the breeding
gein in chloride form it is ready Vo stariup another plant, without

intermediiate handling,

3.5.11.3 Impact of fission preduct concentration upon neutron physics

performance. ANISN calculations showed that a FP concentration of

b 1.-1

5.8 x 10 aem b =0.10 [Ufissile] in an MCFR(Th) reduces the

3R by 0.007 cut of 1i.34 or 0.5%. Within the accuracy of the cal-

culation, this amount of FP did not aff:ct the critical dimensions.
Figure 3.5-36 shows the FP effect on BG calculated by Taube
et al for an MCFR(Pu): [FP] = 2 x 1074 - 0.10[Pu] appears to cause
a similar BR loss of 0.01/1.45 = 0.7%. [FP] = 1.0 [Pu] seems to

produce a (proporiicnally) larger loss of about 10% in ER.

1e5 S
1.4
le3 -

l.2 4

 

 

 

T ¥ 1 l T
o2 ks L2 L)
Fission Product Concentration (a cm—lb-l)

Fig. 3.5=36 Impact of Fission Products upon the ER of an MCFR(Pu)[17]

l1.-1

[Pu]core = 0.0021 a em ™ b

307
 

3.5.12 Summary

Section 3.5 attempted a first cut at optimizing the reactor des
ign and compesition and also analyzed some of the consequences there-
from. The first six subsections defined methods, data, and models.

Sec. 3.5.7 discouraged the use of an inner blanket and showed
that two meters constitutes an infinite outer blanket, In that event
a reactor reflector contributes little., In addition to achieving
maximum BG, the 2 m=blanket greatly reduces the flux dose to the
reactor vessel., Sec. 3.5.7 also favored graphite as the core tube
wall material because Mo strongly absorbs neutrons.

Section 3.5.8 showed tnat design decisions which favor a high
233U/23£‘U ratio in the core salt produce

1. A hard neutron spectrum there

2. Higher BG and lower DT

3. Minimal presence of Pu and other high-A actinides
Fortunately, within the ranges of interest, all the trade studies
indicate that the 233U/ZBI*U ratio on equilibrium fuel cycle will
exceed 7. That is sufficiently high to achieve the above goals;
additional major efforts are unwarranted.

.Sections 3459 and 3.5.10 revealed that both ThClh and NaCl,
acting as diluents in the core salt, greatly reduce flux levels and
the amount of out~of-core fissile inventory. They also reduce BG.
The combined effects of inventory and BG produce an optimum DT near
a U molar composition of about 30% and a Th/U ratio of 1=2. This

and the lower flux levels explain the preference of other MCFR

308
 

studies for a fissile molar content of 15=30 %. However, low fissile
content will decrease ther233U/23AU ratio, an indication of greater

transplutenium buildup. Thus the true optimum remains to be found

and will depend on final assignment of priorities.

DT of 20-30 years appear feasible, This is high compared to
quoted DT 13-20 years for an MSBR on the Th fuel cycle. However,
an MCFR(Th) should build up less transplutonium.

Efforts to lower DT should focus on the out=of-core inventary.
Note also, that, compared teo conventional reactors, the absence of
fuel fabrication and decladding operations and the minimal trans-
portation and storage needs provide additional fuel inventory ad-
vantages which are not taken into account here.

Section 3.5.11.1 quantified somewhat the basis for avoiding
transplutonium buildup. Sece 3.5.11.2 pinpointed fresh blanket U
as the greatest proliferation hazard. By immediately mixing it with
core salt, one adds not only fission product activity, which is
chemically removable, but also 232U—daughter radiocactivity which
is not. The radioactivity will discourage U handling, but not make
it impossible. Sec. 3.5.11.3 showed that the presence of too much

FP can sericusly redace BG.

309
3.6 Safety & Kinetics
3e6.1 Physics of Reactor Safety and Kinetics
3.6.1.1 Effective delayed neutrcn fraction. Delayed neutrons
frem fission cf 2331;, 235U, and 229Py amount to 0.0027, 0.00635,
and 0.0022 of the respective total neutron yields., At first glance

this suggests that the effective delayed neutron fractio for

7
B/vers
233, . 239 . :
a U=-fueled }CFR exceeds that fer a Pu~fueled [MFBR. However,
most cf the MCFR fuel lies cut-=of-core in the primary circuit.
This makes .3 .. much smzller.
/ eff

Passing through a core less than 2 meters high, with a ve-
locity of 9 meters per second, the fuel is present for only 0.2
seconcs. crcach decay half-i:fe of thne six delayed=-neutron groups
(Table 3.6=I) far exceeds that. Therefore the delayed neutrons
originate near uniformly throughout the primary circuit and /éhff
diminishes by the ratio of fuel in-core to that in the total cir-

cuit: by 0.2-0.3 to 5eff‘= 0o 0005~0,0008,

3,6.1.2 Prompt neutron lifetime,l . 1 increases with the median
fission energy Z_, ranging from 5x1071 for an LMFER (190 keV) to
3:{10_'9 fcr Jezebel and 233U Godiva (near 1.6 MeV). For the inter—

meciate MCFR(Th) with B = 400-700 keV, 1 should = 10‘8 st

3.6.1.3 Temperature ccefficient of reactivity. Reactor stability
depends on several coefficients of reactivity. In a molten salt
reactor, the temperature coefficient due tc expansion dominates.
In an MCFR the very hard neutrcn spectrum nullifies any Doppler

ccefficient of reactivity.
 

Groug

1

o W - W n

2 )
Table 3.6-1 33U Delayed Neutron Characteristics

 

Haif-life (Sec) Relative Abundance (%)
55. 9
21. 30,
Do 25.
2. 28.
0.5 5e
0.3 3.

Section 3.3.5.6 indicatzs a typical temperature coefficient

of salt density of 0.0004 %4/ £ /O C. ANISN calculations (Section

3.5.8.3) reveal that 0.050 4, lowers reactivity by 0.034 Ak/ Xk,

a coefficient of 0.68 AL'/ k / Af’//: . Applying the above gives

-27 x 10

5 Ak o -1 .
— € . Using /Beff = 0.00068 one gets a strong

negative temperature coefficient of =40 ¢ /°C.

For comparison, 2 Pu=fuelied LMFBR typically exhibits

~0.4 to =0.5 ¢ /°C [140], assuming Bopr = 0,0034  €alculation

for a 1000 Mie MSBR gave an isothermal temperature coefficient

of «0.9 x 1077 Ak fx /°c [131]. However this comprised a =4.4

x 107 Lk/k/°C for the negative Doppler coefficient of thorium

and positive coefficients for the thermal spectrum graphite

moderator and fuel salt density. The latter stems from a neg-

ative salt density coefficient of -=0.03 Ak/k/ A/a//o for their

low-actinide fluoride salt mix.

3t
3.6.2 inalysis of Normal Cperations

3.6.2.1 Reactor startup. MCFR coolant circuits and salt supply will

need preheating just like an lSBR does, A salt of subcritical fuel

conicentration or even a low-melting non-fuel salt could {iow test
the circuit. Cne might approach zero power criticality and sub-
sequent rise to power by gradually adding fissile fuel through the

clean-up loop to a suberiticai fuel salt.

3.6.2.2 Reactor control, A iimited study [29] indicated feasible
control of reactor and power level by menipulating-the secondary
circuit conditicns: wmainly varying the flow rate of salt through-

the ccre and heat exchangers. The negative temperature ccefficient
will hold the salt temperatire nearly constant at all power levels.

The absence of tke heat transfer and fuel pin temperature differentials

of a solid-fuel reactor allow this direct respcnse.

3.6.2.3 Reactor stability and inventory ratio. The amount of fuel
in the primary circuit outside the core fixes the effective delayed
neutran fraction and affects the reactor stability. The British

[29] showed that an HCFR(U/Pu) should be stable to small reactivity
perturbations if the fraction of primary circuit salt in the core ex-
ceeded 25he Engineering studies of their MCFR designs met this re-
quiroment with a margin, The primary circuit flow rate seemed to.

have little effect upon the delived neutron -concentration in the core,

312
Table 3.6-II Fuel Inventory in a 11 GHth MCFR(Pu) Plant [23]

Specific heat exchanger power
(canservative data)

Total volume of heat exchanger for
11G4(th)

Puel frazction -of heat exchznger volume
Fuel volume in heat exchanger

Fuel 1n the ductrwork

Total fuel out of cors

Fuel in core

Total fuel in system

Mean specific power of fuel in the
whole system 11GW th -

5.3
Plutonium content of fuel

Power rating of whole system

lKZ-I/cm3
il m3

0.3

3.3
1.0
L3 o
1.0 0
5.3 °
2.1 Gi/r

0.8 gPu/cm3

0.385 kgPu/MW th

Taube anaiyzed a 11 G{th MCFE(Pu) and inventoried (Table 3.6-II)

of the fissile fuel salt is out of the core.

313

the fissile fuel salt in the primary circuit. Assuming a power density

of 11 Mith/liter and conservative heat exchanger pericrmance, 81%

For a given heat ex-

changer rating, as the total plant power decreases the inventory in
the heat exchanger will also by about 0.3 m3 fuel per GWth. Littie
change occurs in the core or piping. Therefore according to ths

British analysis, this plant should operate at lower power.
 

Table 3.0~I11 compares Taube's results to those of 3ritish de-

signs and a 1956 ORWL study. The 3ritish designs contain only 56-645%

of the primery circuit fuel salt out~of-core, revortedly mostly in

the duct work, Unlike Tzube but in agreement with the British designs,

the OREL/1956 study also aftributed 2 large portion of the inventory

to piping between core and heat exchanger.

Table 3.6-II1 Volume Distributions in Some MCFR(Pu) Designs

 

 

 

crRL/56 [13] Taube 23] British [29]
Vol Distribution Volume Distribution
OLUMe  (ut~of-core um out=of=core Quaiitative
(m3) (%) (m%l_ (%) Proportions
Core 3.30 1.0
HX plena 0.35 10.0
HX tubes 1.62 L6.2
HX subtotal 1.97 56.1 3.3 76.7
Piping 1.54 L3.9 1.0 23 .3 dominant
Qut~of=core 3.51 100 L3 100
Subtotal
Total 6081 5.3
% Out=of= . 51.5 81.1 56=64

core

314
Fige 3e6-1 shous a typical British design [29] with large aeat
exchangers to compensate for lower thermal conductivity values for the
fuel salt, They have given considerable thought to reducing duct
iength but find that unconventional methods tend to be less relizble
or less easily maintained, Increasing fuel velocity can reduce duct
cross sectionzl areas, but limits due fo vibration and puaping pover
costs may occur,

The choice of intermediate coolant will also affect oui-~of-core
inventory by its influence on heat exchanger arrangement and temper-
ature conditions,

Je5e2s4 Reactor Shutdowne Basically the ICI'R requires no scram
device because the strong temperature coefficient holds the system
steady, Adding blanket salt or other absorber should quickly reduce
power and temperature, Hovirng part or all of the molten core fuel fo
geometrically-safe, cooled holding tanis will accomplish full shutdovm,.
Criticality safety calculations must be sure to consider criticality
233 23

3
increase as “Ue. There decay heat will

Pa decays (t1 = 27.04d) to
<2

help keep the salt molten, while natural circulation cooling holds

dovm the fuel vapeor pressure,

Je0e245 Chanze in physical pronerties due to Iransmutations, As

232 233

Th and U transmute to actinides higher in Z, the individual salt

melting points (excluding eutectic effects) lower (Fige 346-2). How~-
ever, since uranium isotopes predominate in the burnup chain, very
little higher-Z material appears, Similarly, boiling point and vapor

pressure should also change little,

315
91t

Figur@ 3. 6—1 °

          

I CoRg

I mAMyt
3 ST IRaTRATE HMEAY [NCHANCAR - B ot
4 RAL AT A -t

3 MR XA - den
b RAMRLT COOLAR MMP- 4ot
T el IRATL (OO ANT HEADER
8 T Dxart CO AN 1M dT
1 TIREDATE COOy ANT OUTUEY
O Gak v ( wuvy

1 el .
Yeur ! L_i ‘!

Typical MCFR{Pu) Design with Large Heat lLxchangers to Compensate for Low Thermal

Conductivity of Salt [29]
Melfing
Point

(°c)

Figure 3. 6=24

 

 

 

 

 

 

 

 

      
  
  

 

 

 

 

 

700 ‘ - . - [ I
oo N +3 chemcal stote-—-

 

 

 

Koo

 

 

 

—- - - - _—— L — ——
L e — [ e e

 

 

 

90 91 92 93 9
(fn) (P2) (U) (o) (Pu)

Atomic Number

Helting Point of Some Actinide Chlorides

317
 

- 3.6,3 Analysis of Accident Situations

Cantinual removal of velatile fission products frem the salts
during cperation eliminates several obnoxicus species fram being
present in an accident. Yany of the remaining hazards would stay
in the salt. Thus, no accident can occur which corresvonds in se-
verity of radiocactivity to the meltdowm of solid fuel systems. The
major concern then becomes that no primary circuit failure lead to

critical masses formirg.

3.6.3.1 Small leakages. Leaks bstween core snd blanket pose ne
sericus chemical threat as the salts are compatible. However, other
factors influence whether ti2 pressure difference should force

core salt leaks into the blenket or vice versa,

Higher temperatures in the core salt may produce higher wvapor
pressures there then in the blanket. The core salt also circulates
under pump pressure vwhile the blanket salt need not. If a reserve
tank autcmatically replaced fuel salt leaking into the blanket, the
system could simultanecusly approach supercriticality and high
temperature, a strong temperature coefficient not withstanding. The
high temperature could further abet the detericration. 1In contrast,
blanket salt leaking into the core would anly dampen the criticalitys:
hovever, this could result in degraded system performance.

An anomalous rise in concentration of 233U or fission product
in the blanket, or a rise in temperature there (due to increased
fissions), ar a drop in core salt pressure would sigmal core-to-
blanket leakoge. Leakage of blanket salt into the core would decrease

the core mean temperature and blanket salt pressure.

318
The secondary coolant circuit should be pressurized slightly
higher above the primary one, and the tertiary even nigher. This
will cause less active coolants like helium or lead te leak into
the molten salt. He 1s inert and the salt cleanup system, which
aiready separates the fission product gases, would remove helium
as well. Lead interacts with nickel-bearing allcys, but not with
i{o cr graphite. Traps at certain points in the cirecuit wyculd iocate
the positicn of .ead .eaxage.

The ccntainment building will catch leakage cf volatile fission
products from the primary cirauit or btlarnket tc the air, simiiar
to MSRZ, with zppropriate detection and leaktight barriers, |
3.6.3.2 Loss of flow. Leaks, pipe break, heat exchanger plugging,
and pump failure can all reduce flow in the primary circuit. Figure
3.6=3 analyzes the consequences. In most events the reactar shuts
down and the fuel drains to a safety tank; th=re fission product
decay and-delayed neutron-induced fission continue to generale
heat, but the high capacity of the salt restrains the temperature
rise, -

The effect of a single pump failure depends upen the primary
circuit arrangement. With single punps per channel (Pian A in Fig.
3.2~8) failure of cne pump would shut down one whole core channel.
The fuel sait in that core chanrel must then drain out to stop full
pover production im it. Subsequent replacement by void, blanket; or
carrier salt would cause a large loss of reactivity thereby shutting
down the reactor unless

(1) The separation between chamnels was such that they were

319
 

Flow Reduction

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

02t

 

or
Total Loss
One Multiple Pump Failure or Heat Pipe Minor
Pump Total Power Failure Exchanger Break Leaks
Failure (Electric and Diesel Backup) Plugging
Suitch to Dumj» Fuel to Safety Tank Provided with Natural| |Spillage into
Backup Pump Circulatory Emergency Cooling Epson Salt Bed

 

 

 

 

 

 

Fige 3.6-3 Analysis of Primary Circult Failures

 
individually near crifiical, and
(2) Missing reactivity could be added through enrichment
enhancement
Then one could continue operation on a reduced scale, e.z. 6/7 for
six out of seven channels still operating, until a more opportune
time occurred in which to drain the whole core and replace mal-
functianing equipnment.

With multiple subchannels per core channel, each with pumps
and exchangers as in Fig. 3.2-8-B, C and D, failure of one pump
still allows partial reactor operation until a better time for shut-—
down and repair. If cne operated the Plan B or C subchannels or |
the plan D heat exchanger at less than full efficiency, then their
full use is available in the event of one subchannel failure. This
would however, increase out=cf=core inventory and capital costs
though., Should the core tubes be connected in series as in Secticn
3.2.6.4 then any in~core malfunction (leak or pipe break) would
require system shutdown.

Should both regular and emergency power supply fail, all
pumps would fail unless they are steam turbine driven. However,
electrically driven pumps are easier to include in the containment
and also easier to supply for pre-testing, etce Inertia would
help electric pump run—down rates, or same short~term auxiliary
supply might have to be provided until full dump of fuel has takenr
place. The fluid could also drain through a "freeze wvalve" by
gravity into a natural—convection cooled tank which has a non-
critical geomeiry.

In the event that flow ceases and core salt remains in place,

321
the strong negative temperature coefficient will control the temp-
erature and power: initial temperature increase will decrease
density, which decreases reactivity and thereby power, until some
equilibrium state is reached. HCFR(Pu) studies [29] with a simpli=
fied reactor model showed that the temperature rise of the salt for a
reactivity step of up to $1 should be less than BOOOC; for 7 pumps
failing out of 8, less than 230°C.

In summary, it appears that reactor stability concern over
out-of—core inventory may oreclude the use of cut~of-core subchannels
as well as Aictate series ccnnecticn of the core chamnels. This
means designing the system fer high reliability to minimize shut-
dosm and for rapid repair of heat exchangers, piping, and pumps
wnen shutdown does occur. Rapid repair will entail expedient removal
of all the core salt followed by a salt flushing of all radiocactivity
from the system. The strang temperature coefficient should disspell

any criticality accident concerns.

322
3.6.3.3 Structural failure. Failure of the tubes separating core
and blanket would lead to no chemical or cempatibility problems.
Dilution of the core salt will reduce reactivity; anly if core salt
were replenished as it entered and displaced the blanket, could a
reactivity increase occur. Even then the system could accomodate
temperature rise from moderate salt additions, as ample margin
exists abecve cperating temperature on a short term basis. The
limit to the permissible rate of such salt addition needs to be
established.

Vessel failure would reguire rapid dumping from both the
circuit and the catchpots. Dump tanks should probably be sized
to centain the cantents of cne secondary coolant circuit as well
as all core and blanket salt. Relief valves on the secondary
system will protect the primary circuit from pressurizaticn result-
ing from a major steam generator failure into the secandary system.

In the event of a major circuit failure, operating pressure
would play an important role in the subsequent fate of the fissicn
products and the containment, The molten salts themselves ex-
hibit low vapor pressures. Although high pumping losses can cause
primary circuit pressures up to 460 psi, lead secondary coolant
acts essentially like a hydraulic system with little stcred energy.
Missile formation is therefore unlikely and it should be possible
to demonstrate a containment which will not be breached from this

Cause .

323
 

With a high-pressure helium—cooled system one must ask whether
an accident might aerosolize the core salt. However, a properly=-
designed reactor vessel could withstand the full helium pressure from
a severe rupture between the coolant and primary circuits. Thus
release of activity te the reactor containment cell would not occur
except under a simultonecus double failure. The cell itself is
small and can economically be made in the form of a prestressed
vessel to withstand missile damage and to act as an additional
barrier to fission product release. A final low pressure building
cantainment would prevent release from smell leaks in the earlier

centainment stages.

3.6.3.L Emergency cooling. Unlike solid fuel, molten salt can quickly
transfer into a geometrically-safe tank. Cravity can provide a
failsafe transfer method from reactor to holding tank (Figure 3.2-11):
a pipe wherein the salt is normally kept frozen weould open naturally
upcn loss of electric power.

In the tank, the salt will circulate naturally as heat from
fission product decay transfers through tube walls to air convected
by natural draft towers. Ctherwise pressure would build up from salt
vaporization. The absence of mechanical moving parts will make this
whole system hizhly reliable; multiplicity could add even further in-
tegrity. A forced draft system would probably layout more compactly

324
 

A second cooling option would circulzte naturally a low melting-
point salt or NaK through U-tuves in the tank to boiling water heat
excnangers, Air-cooled condensers situated in a2 normel or forced draft
stack would condense *he steam formed, Alternatively, the heat could
transfer directly to a large boiling pool, thus accomodating decay heat
for a protracted period without makeupe. Condensers or make-up water
would be provided for coniinuous operation,

A catchall salt bed below this apparatus would serve as a backup
for any leaks or breaks, The heat of salt{ formation would greatly
2id in absorbing decay energy. An independent cooling system might
remove decay heat. The bed wowld alsec dilute the fuel salt., Proper
choice of tank diameter would insure subcriticalitye.

Recovery from emergency tanks or beds would require heaters,
drains, and pumps. When carrier or blanket salt is added for dilu~
tion and/or heat absorption by latent heat of formation, a method

tc separate out the dump salt would z2lso be needed,

325
 

3.6.3.5 Comparison of MSRs to others. WMSRs should be biologically
safer because

l. The plant continuously removes velatile fission products

2. The fuel is already molten and in contact with materials

designed for that candition.

The removal of volatiles fram the primary circuit still requires
attention to their presence elsewhere in the plant. However, it
should not be difficult to insure the integrity of a storage medium
below ground. After a suitable decay period scme of the gases may be
releasable to the atmosphere.

In the event a pump fails or other flow loss occurs, ths enly
concern is that decay heat right build up vaporization pressure,
However, the same colls which preheat the fluid could also cool the

fluid.

3.6.3.6. Precipitaticn out of eutectic mixtures. After the temperature
cof a salt mixture falls to the solidification point, the compositiom
of liquid changes, sliding alang the liquidus curve as crystals
separate out (Figure 3.6-4). With low initial UCly molar centent,
cooling precipitates out NaCl crystals, thereby enriching the fluid

in UClB. With high initial UCl3 molar content, cooling produces UCl3
crystals and NaCl-enriched fluid. In either case the liquid migrates
to the nearest eutectic point (nadir) composition and causes a concenw
tration of UClB. Thus one must consider possible criticality situ-
ations and design geometries to prevent them, Probably precipitation

of UCl3 crystals from a UClB-rich fluid is the less-seriofis case,

326
 

Initial Initial
Composition Composition
A 3

I

 

Temperature

 

 

 

 

All

ATl
yaCl

UCl
Molar
Composition

Fig. 3.6=4 Migration of the Liquid Composition Along the Eutectic
Path as the Salt Mixture Cools

3,6.,3.,7 BRBoiling off cf salt mixtures. In the event that the carrier
salt had a boiling point much lower than that of the actinide chlorides,
one could postulate a positive temperature coefficient contribution in
that temperature region. Fortunately evidence from section 53.3.3.5

contradicts this: carrier salts appear to be less volatile than

actinide chlorides.

327
3,6,3,8 Containment. Solid fuel plants generally feature triple
containment: fuel element, reactor vessel, and tha:n the contain-
ment building. The MCFR will feature reduced FP radiocactivity due
to continuous gas cleanup but it will inherently have only double
containment of piping and 2 building. A second radiation hazard
arises from activation of the secondary coolant and nearby equipment
by delayed nevirons in the primary circuit.

To further ceontain these harzards, a low-pressure leak-tight
membrane might be formed arcund the walls, floors and ceiling of
the reactor system. This membrane can also form part of the duct-
ing for the inert gas circuiaticn required to cool the concrete
structure and shielding and inhibit vessel oxidation. Heat losses
with insulation restricting the concrete temperatures to below 7OGC
would be about 200 watts/m2 with 40 ¢m of insulation. This would
equate to about 3 M{ total heat removal by water or air cooling.

The main building would constitute the tertiary containment.
Its volume must be sufficient to contain the stored emergy of any
gases present. In the lead-cooled design these are cover gas vol-
umes al low to moderate pressure. The helium—cooled version may
require a larger outer containment volume and/or intermediate pre-

stressed concrete containment(Sec. 3.6.3.3)

3,6.3.9 Resistance to external threat. Cne can postulate a number
of scenarios wherein external explosion or impact threatens th=
reactor integrity directly or through a loss—of-coolant accident.
These could involve ground or aerial bombs, planes, and space re-
entry projectiles (e.g. meteors, missiles, space laboratory), by

either accident or intention. An individual, a subnaticnal group,

328
 

or another country might intentionally initiate, Nature threatens
with earthquakes, tornadoes, and dam breaks.

In many or most of thase situations one could anticipate the
danger. Most reactors already lie below ground level. Still the
vessels generally sit high enough to be rupture~prone to strong
explosions cr impacts. Such an incident could release volatized
fission products and actinides of_consequence greatly exceeding
a simple bomb cr other initiating event.

To protect against these threats an MCFR could uniquely trans-
fer its fuel to a storage tank (section 3.6.3.3), located remotely

under additicnal earth (Figure 3.6-~5) or in an otherwise hardened

"sunker".

 

 

 

containment
ground
level
core
| storage
tank

 

 

 

Figure 3.,5~5 Hardened Starage Tank

3.6.3.10. Molten salt combustion support. Molten salt itself does
not burn, but will support combustion with solids such as wood, coke,
paper, plasties, cyanides, chlorates, and ammonium salts and with
active metals such as aluminum, sodium, and magnesium. Water from

spray sprinklers or low=velocity fog nozzles provides good fire

protection.

329
 

347 Fuel Processing [17, 81, 141-146 ]

This section analyzes the chemical aspects of {the reactor fuel
cycle: principelly vreparation of the ThCl4 blanket fuel and repro-
cessing of irradiated blanket and core salts,

Some processing should occur continously, some batchwise, Ratch-
wise will generally be easier, more efficient in separation, and more
economical, This study does not deal witk the near-identical concerns
associated with starting early cores up with PuCl3 or 235UCl instead

3
233

of ucl .

3

ThO2 is converted to ThClh, mixed with reprocessed blanket
salt, and fed into the blanket simultaneous with blanket salt
removal. Blanket reprocessing separates out bred'UClh(pre-
dominantly ;233U01h},reduces it to UCL,, and admixes that to
reprocessed core salt, forming the core shim material., This
constitutes a spiked fuel, suitable for BG export as well as
for positive reactivity shim to the MIFR(Th) core (simultaneous
with irradiated core salt removal).

Core salt reprocessing removes fission products, oxides,
corrosion products, and chlorine iransmutation productsj part
continuously and part batchwise. It alsoc readjusts the chlorine
stoichiometry.

3.7.1 Principal Salt Reprocessing Methods

The MSR ccnicept inherently aveids many of the conventicnal
reprocessing stages, especially fuel element dissolution and
refabrication. The absence of these stages removes the need for

high decontamination stages in reprocessing. Proliferation ccncerns

330
 

also discourage high decontamination: residual radioactivity

makes weapcn construction hazardous. Eliminating ail these stages

racucss costs.,

This leaves just the separation stages: core sali cleanup and

bred-U extraction frem the blanket. The most promising methods are by

l. Solvent extracticn from aqueous soluticn
2. Volatility

3. Pyrcmetallurgy

L. Pyrochemistry (molien salt electrolysis)

e first review the chemistry of the heavy elements.

3.7.1.1 Chemistry of the heivv elements[31], The separation
processes, especially those based on solvent extiraction, take ad-
vantage of the scmewhat unusual chemical behavier of the actinides.
Elements in the analcgous lanthanide series all exhibit similar

chemistry: +the presence of three, relatively loosely-bound, outer

electrons causes each atom to exhibit 2 positive valence of 3,

The actinides also all form a tripositive (III) valence
state. However, some also evidence loosely-bound inner elec-
trens. This leads to tetrapositive (IV}, pentapositive (V),
and hexapositive (VI) states.

Tnese higher oxidation states evidence different stabilities
(table 3.7-I1) which facilitates extraction and separation of the
heavy elements. Th and U differ proncuncedly, Th evidencing

mostly just the IV state.

3
Table 3.7-I RELATIVE STABILITIES COF OXIDATICN STATES CF THE

ACTDNIDZ TIZMENTS [81]

 

 

Atomic

NOueases 89 90 91 92 93 9L 95 96 97
Element... Ac Th Pa U MNp P Am Cm Bk
T  tereces e * *? * HH M O SRR MR
TV eennnnese HERE H AR R R X **
“]..'l'll.. **M* * H* H *
V-i..-..... X Ak L *

Legend - most stable state

decreasingly
stable

The IIT and IV states easily preeipitate from aquecus soluticn
as flucrides; the V or VI states do not. The fluorides of the
IIT and IV states do not volatilize; the VI-state fluorides do
at fairly low temperatures. The IV and VI states appreciably
dissolve in certain organic liquids; the III-state nitrates remain
virtually insoluble in these liquids.
3.7.1.2 Solvent extracticn from agueous solution. The extractiaon
o; actinides from aquecus sclution by an organic solvent is the
most advanced process: it has been widely used since about 1951.
The Thorex version extracts 233U from aqueous soluticn of irradiated

1n fuel elements [72}. It relies on the stability differences

332
 

between Th and U in higher valence states (Table 3.7-1).
The facility of solvent extracticon for multi-stage operaticn

without consuming additional heat or chemicals particularly benefits

1. Situations requiring extreme purification, Use of enough
stages can lower the gamma activity from fission products
in the extracted urznium tc belcw that of naturai uranium.
Thus cne might not want to furnish this capapility tc a
foreign naticon.

2. Situaticns where th:s propertiies of two metals parallel
one anicther sc clcsely that a single precipitaticn or
crystallizaticn can not separate to the degree required.
Thus this method may not be needed for Th-U separation.

The dissoluticen step for a chloride salt has to be the simplest
of all: Just add water. Different organic sclvenis separately
extract the U, Th, and wastes. Addition of 012 and CCLL subsequentliy
rechloridizes U and Th.

Disadvantages associated with this method include

l. Extra criticality precautions fer H-mederated fusl soluticns

2. Multipliciiy and complexity of steps

3. Large waste volumes

I, Large shielded space required

5. Additional steps needed to produce solid wastes.
3.7.1.3 Velatility orocesses. Volatile UF (b.p. 56.AOC) easily
separates from fission product and Th fluorides. The‘ability to de-
contaminate to low activiiy levels parallels that for solvent evirzc—.

tion, However, the volatility method requires fewer steps and therefore

333
smaller volumes of highly-radicactive wastes compared with aquecus
pProcessinge.

Volatilization should work especially well for processing a
ThFh/%aF (cr other fluoride carrier) blanket salt mix. Cne would

cxidize +he bred UFL to UFy, separate it by distillation, chloridize

4

it, and reduce it to UClL,. The remaining ThF

1 would return to the

blanket.

Distilling UClh, UCls, or UCl, from a mix of ThClb and chlorides
of the structural materials and fissicn products would be more diffiw
cult: more volatile chloridss compounds exist and their vapor-pres-
sure ranges overlap, Howev:zr, considerzble less develooment effor<

has been expended on method:s for separating volatile chlorides; the

problems may be solvable,

3:7.1,5 Pyrometallurgical processing. In the 1950's ANL developed
pyrometallurgical processes to recover and purify fissile and fertile
material from breeder reactors. Although demcnstrated cn a pilot-
plant scale, much engineering development remained to evolve a
workable and reliable process, especially in fiiew of criticality
restrictions.

A typical process would use a molten chloride flux to contact
the oxide fuel and a molten metal to extract the actinide. Hence the
need for high temperature (pyro). With molten chloride salt fuels
anly the extractant is needed: for chloride fuels, Dillon [28]
proposed a Mg-Zn alloy. A similar metal may work with Th/U cycle
chlorides.

Pyrometallurgical processes decontaminate by factors of only

334
about one hundred. With them one must handle fuel remotely. This
poses no problem for molten salt processing: it requires no fuel
fabrication or other handling; everything is dcne remotely anyhow.
Leaving in racicactivity also inhibits use of the material for
Weapans.

Because of its compactness, pyrometallurgical process can operate
ciose-coupled to the reacter and cn a muich shorter cycle than the
acuecus rcute. If econcmical, a pyrochemical processing plant could
easily be acccmodated within the reactor building due to its small
size., Preliminary work indicates that capital and operating costs
may be high because of the small batch type operaticns needed.

Detracticns include lcw recovery rates and the development

needed to cope with temperatures zbove TOOOOC. High temperature

€quipment of great reliability that can be operated and replaced
remotely 1s hard to design and expensive to test. However, work

already done indicates that the difficulties may be overcome.

3.7.1.5 Molten salt electrolysis. Techniques exist to deposit
actinide oxides and carbides from molten =alt sclution through
electrolytic raduction [143] e.g.

U6,*% (seln) + 267 == 1UC, (solid)
at the cathode. At ths anode:

2C1" (soln) === Cl;(gas) + 2e”
Similarly Taube [147] mentioned reducing actinides directly in their
melten chloride mixture. The Cl, released could be used to chloridize

ThC, for blanket feed (section 3.7.7).

335
 

3.7.2 Core Salt Processing

Cn equilibrium cycle, the core fuel will include carrier alkali
salt (e.g. NaCl), fissile UC13, diluent ThClh, actinide transmutants
(e.g. PaClL, NpClB, and PuC;B), impurity oxides, fission products in
various forms and states (section 3.4.2.1), structural cerrosion
products, and sfifn { from chlorine transmutation)}. Some of the
mutants may fcrm éomplex chlorides like 052U016 angd compounds like
UI and US, which precipitate out at sundry temperatures.

Core shim material rmust replace irradiated core salt near-
ccntinuously so as to maintain criticality. The fuel burnup rate’
(1.6 gm/min for a 2250 }fth plant) fixes the rate of shim replacement.
Actual core salt reprocessing can still occur batchwise or continuously.
The choice will depend in part on the allowable inventory of core salt
in reprocessing and the allowable level of FP concentrations in the
core. The latter depends an
1. The reactivity worth of FP
2. The effect of lower Th/U ratio or increased reactor size
an reactor performance.

3« The effect of ¥P on BG

L« Acceptable radicactivity levels in the primary circuit
(high levels necessitate remote reprocessing and extra

plant shielding).

336
 

3.7.2.1 BRecovery options. The first option is whether to clean up
the salt (remove the bad part) or just recover the 233UCl3 and scrap
the rest since lNaCl is so cheap. Solid fuel cycles conventionally
take the latter approach. Here we choose to clean up the salt because
1. NaCl radiocactivity precludes easy disposal
2. Znrichment in 37Cl(if chosen) would forfeit cheap cost
arguments
3. Discarding non-uranium actinides would mean poor fuel
utilization since every actinide atom can eventually
fission in this spectrum
L. Continuous gas removal gives a gocd start on salt cleanup.
It might be feasible to just remove the parasitic neutron ab—
sorbers and corrcsion agents: let the salt accumulate the rest of
the mostly-radioactive non-volatile mutants,some as substitute
carrier salt. The acdvantages are
1. A suitable place to store non-volatile radioactive wastes
2. The reactor will transmute many of the wastes into a less
hazardous form.
3. The radiocactivity will add to the heat source.
L4, The radicactivity discourages diversion of the core salt
for weapans purposes.
Pctential disadvantages to watch for and control are
1. Change in viscosity and other thermophysical properties
2, DlMutant plate-ocut causing flow blockage, radiation sources,
or other problems.
The enumerated advantages seem real enough to warrant this

basic approach; practicalities may require some modifications,

337
3.,7.2.2 Continuous removal of mutant gases. Fission produces Se,
Xe, Kr, I,, and Bry, which are all gases at reactor temperatures
(section 3.4.2.1). Threshold reacticons élso produce He and HT gas.
Cne toa of fuel from the core of a fast breeder contains 2 x IOA Ci
ot 8%z 100 0i of Be (after cooling 4 gonths), 130 and 0.7 Ci
of LBl (after cooling 4 and 6 months, respectively), 0.13 Ci

1291, and 2200 Ci of T. In an MSR circuit rupture, these weculd

of
all present a radiation hazard. Gas accumulation will also build
up pressure, affecting circulation,

In remedy for the MSBR,CENL proposed a cleanup system which
recirculates gas in lcops across each main salt pump. Injected
helium nucleates bubbles which absorb gases, the small-particle
fog of inert metals (Pd, Tc,Ru,Rh, and Te) and some of the volatile’
chlorides (Fig. 3.7~1). “hile in a hold-up tank to reduce decay
heat, some of the metals and chlorides deposi£ out. Passage through
traps and beds removes Kr, Xe, water, etc., before returning to the
helium injector. About 20% of the bypass flow undergoes a long delay
in which all isotopes except the 1(-year 85Kr decay to an insignificant
level.

Presently 85Kr levels from reactors are small encugh to discharge
directly to the atmospherc. Should nuclear power abound in 30-40
years it may be necessary to separate out 85K; low=temperature
fractionation looks like a promising method.

Eveporation and fractionation can also concentrate tritium for

lengthy storage; this is presently done in solid fuel reprocessing.

336
 

 

  
  
 
 
  
    
  
  
  
  
  
  
  
  

6 Lr 50 X {ast

alow

 

Jas

39 ar 153 1 PrE Estraction

 

vary slow
3 3 52 Te

 

33 As 51 Sb

 

PPS Volatile chlarides

32 de 5¢ Sn

 

A
N s 3 In Low volatile
chlorides

 

FPs

  

 

30 In "0 4

 

 

 

Nan
volatile

   
 

15 Rh PP

 

v Ry Bethls

 

1y T

 

2 Mo Low

 

   
 

£l W Fes volatila

 

cnloriles

 

W Ir

 

64 04

 

61 Fu

 

€2 sa

 

Porlodlic Table

&L Pm Non

 

   
 

60 Na PPA volatile

 

enlerides
5§ o=

 

58 Ce

 

39 Y |57 la

 

38 3r [S6 Ba

 

Low volatile chlorides

 

 

 

37 R |ss ca | FA

 

Fig. 2.7=1 TFiszicn Producis in Molten Chlorides edia
—A——L—-——n

An intermediate half-life makes 137Cs ane of the more
troublesome fission prcducts (Fige 2.1=2). Figure 3.7=2
shows that only a small portion of the A=137 yield comes
directly to 13705. Though the gaseous precursors decay re-
latively quick, continuous gas extraction could remove a lot
of them before non-gaseous 13705 formeds This would
1. Lower the activity of the circuit
2. Avoid the more difficult removal of CsCl, or other

compound later

3. Help isclate Cs for individualized volume reducticn.
 

A
2 g J2 33 5 55 sg

 

 

 

 

il s Tote [ lp w—Cyp —wgy AsL37
from — )
tisaion {@ j
1oa 10
Fi . . — - ependant
lg 3 7 Tiels 10
Feprocessing Characteristics L
0.1 -
of the A = 137 Fission s
1’ ,
Tavmdier anad salr oir Ca-13 a-1y7
rreduct cnain. ¢ T now
Sacaonas 2
1o
10
1 v ——
£xcract fan

~ Rate
a7k L

The main design parameters for gaseous fission product removal

1. The absorption required in the delay bteds
2. The heat removal needed to avoid excessive temperature
rise in the charcoal traps.

Cne form of delay bed im a trough of swimming pool size. Low pressure
steam forms above the pool and passes to condenser qnits. Such a system
would dwarf the reactor in size. A4lternatively, cne might store the
gases safely and reliably at high pressure.
3.7.2.3 ‘Removal of non-rasecus fission products. The inert metals
(section 3.4.2.1) not escaping as fog will likely deposit in var—
ious parts of the primary or FP removal circuits. 3Ixternal processing
will remove the remaining 50-60% of the fission firoducts as nec-—

essary unless on—line treatment processes can be developed.

340
 

Section 3.7.1 mentions three possible rrocessing methods,

Cf these solvent extraction frcm aqueous sclution is the best
established. Illcwever, ecanomics may require the reprocessing plant
to serve nultiple reactor facilities. Reprocessing at a different
location, would mean

1. Increased out-of-plant inventory

2. Increased nazard of radiocactivity, sabotage, and pro=

liferation
Ccmbined with the disadvantages menticned in Secticn 3.7.1.2 this
argues for consideraticn of the less—developed processes of sectiecn
3.7.1. _
3.7.2.4 Contrcl of the oxy-en levels, Cxyzen and oxygen—cantaining
compounds react with UCl; to precipitate uranium o des and oxy-
chlorides. Oxygen can enter the salt through air, water vapor cr
transmutation of F. The vagaricus nature of air and water vapor
entry necessitates keeping the oxygen content well below saturation.
A continuous gas bubbling system (with chemical reducing agent)
should help: so should the appreciable capacity of the salt for
OXygen.

Experiment indicates difficulty with simple methods of salt
cleanup such as the small change in sclubility by temperature
adjustment.. An alternative effective methed routes the bypass
gas flow through a bed, vhere the gas mixes continucusly with
injected liquid NaAlClh. Greater stapility of A1203 causes it
to form over AlCl3 (Fig. 3.7-3). The .solid alumina then separates

out by filtraticn or cyclaone.

341
-!.00-1

|
w
o
o
i

cHiories  [KT-mol™' ]
&
T

1000°%
form

AG

~100-

 

 

 

—100 L —200 ~3d0 , "
1 -
AG_ OXIDES  [KI-Jamel 0]

Figure 3.7-3 Competition Between Formatimn of Cxides and Chlorides at

1000 K.

342
3e7e2sH Removal of sulfur impurities, Several nuclear reactions
convert chlorine to sulfur (Figure 3.7;4). Mass balance and Coulonmb
barriers combine to inhibit most of these reactions; but high flux
levels bolster production, 3501 converts to 325, 335, and 365. 3731
should produce only small amounts of 348 (favoring 2 3701 enrichment
2gain), A natural chlorine-fuelled system will produce more sulfur
than the salt solubility can handle, Phosghorus, though present is
only transitory, decaying quicily to suliur,

The presence of sulfur, or of any other element capable of
compounding with uranium, need not adversely affect the feasibil-
ity of the system: simple zdjustment of temperature and UCl]+ |
content can induce precipitztion in a clean-up circuii. Because
one can reliably predict the production rate of these mutants,
the concentrations in the core could be safely maintained close
to saturation: then even an inefficient removal process would
suffice.

The effect of the sulfur presence will, in part, depend on its
oxidation state. For molten sulfur, the oxidation-state equi-
librium is fairly well fixed; the predominant state tends to be‘
positive (Fig. 3.7-5). The "positivity" also increases with
irrediation (Fig. 3.7-6).
3,7.2.6 Maintaining the chlcrine stoichiometry. Zach actinide atom
initially binds three or four CL atoms. Although fission splits
each actinide atom into two product atcms, the net valency reduces
(in part due to inert gases)} and an excess of chlorine occurs.

This can lead to several corrosicn agents, especially UC14.- (The

343
 

Reacticn Reaction Q-value (MeV)

1. %1 (n,p) BET (%) 35 - 0.615, 9.89
7,28
35Cl
2. 2%1 (n,x) 3% ll*-jf-_d-». 325 | 0.935
3. =6 36
L 3533_ (2,Y) 2°c1 (n,p) 778 8458,1.92
(n, ) 2p —2-,2-,2&-335 2.46
(Y1245 3Ly 6,28
5, 3761 (n,p) s 219'-“_;—_-'1‘-;-3701 -3.98
/
6. Tgi (n,o0) Shp L2:bs | 3bg =1.29
5

Figure 3.7-4 Nuclear Reactions Which Produce Sulfur in an MCFR(Th)

intense fissian fragment irradiation of the salt produces short-
lived ions which quickiy oxidize UCl3 to UClh.)

To hold down the concentration of UClh, and excess Cl in general,
one can react the fuel salt at a modest rate with metal of naturel
uranium or thorium or with other reducinz agents.
3eTe2:s7 Storing troublesome fission products by usinz them as
carrier salis. Section 2.,1,2 points out that, of all the fission

90 137

products, “ Sr and Cs should present the most trouble, However,
Sr and Cs belong to the alkaline earth and alkali classes of ele-

ments examined in section 3.lede Thus a good way to manzge these

344
 

100 -

tulphur species (¥s)

 

 

 

¢ T =T Y
0 00 00

trradiation temperature (°C)

Fig. 3.7=-5 Irradiaticn-tempe-ature dependence of the oxidaticn-

 

state distribution of 275 species [148]

 

 

 

Fig. 3.7-6 Effect of length of irradiaticn time on the --S-

species distributian, 150°C [148]

345
 

long=lived salts might simply be to-use them as carrier salts SrCl2
and CsCl. Unlike natural Sr and Cs, they would not increzse in
radiological hazard, but rather decrease, This would also apply to
some of the other fission products, especially RbCl znd BaClZ.
2.7.2.8 Comparison to MSSRX reprocessing. In FSBR reprocassing

cf the fuel salt (Fig. 3.7-7) a volatility prcecess (3ection 3.7.1.3)
first extracts uranium and immediately returns it to the primary
circuit. Subsequent steps prceceed more slowly and sometimes tor—
233

tuously. They especially include Pa extraction by a ligquid

bismuth contact procsss and -are earth extraction. After 233Pa
decays to 233U, it returns t> the reactcr.

}SBrs must carefully hendle tritium gas produced by 6Li
(n, ® ) reacticns: releases to the environment must be strictly
centrolled.

Reprocessing for an HCFR(Th) should be much simpler: enhanced
233Pa and 23&U fissionability makes it unnecessary to separate and

hold up 233Pa. Also an MCFR produces magnitudes—less tritium,

3.7.2.9 liaterials reouirements. The processing plant may require
special materiais. The transfer lines will probably be molybdenum
tubing; scme of the. largze vessels may be graphite. For an HSBR

a frozen layer of sali protects the wall of the fluorinator from

corrosion,

346
   

   
 

Zower
Razetor

-3233 AWz

      
 

 

uczive

txtracti

Proces3ing

 

 

cine
Separa -
e I tidn 9t
by s 84520 = I 3=z
iT. Fa
frradiated
Fuel ELflozaney

Fmoval

 

 

  

Fluorinatior
itnout Jpgree—=

Therma

= e

 

 

u-233

for sale

 

 

 

 

Figure 3.7-7. SBR Reprocessing Scheme

347

  
 
   
 

Ex::;- ian

 

Fl3t Lan

 

   
   
 

 

 

 

 

4

 
_— .

 

3.7.3 Blanket Salt Processing

Blanket salt processing must extract UClh and convert it
to UClB. Some degree of separation may be initially accomplish-
ed by slowly cooling the salt mix to precipitate out ThC'Ll.l+
until a eutectic composition is reached (Fig. 3.6-1).

A low need for high purity plus a desire for compactness
favor pyroprocessing with chlorides. Should the blanket salt
be a fluoride, oxidation to UFy and subsequent distillation
(section 3.7.3.8) appeals.

UClh reduces to UCl3 reilatively easy using HZ' alkali
(Na,K,etc.), U, or The For metallic reduction, rod, gauze
or turnings are popular. Using alkali (ik = Na or K), the
chemical reaction is

UClh + Ak ——i U013 + AkCL
This results in a 50/50 molar mix; lower in uranium content
than desired. Any AkCl initially present would even further
reduce the UCl3 molar content. This leaves culy Th or H2
as alternatives.

Th metal is toxic and flammable. H, produces HCl, which in
turn might be used to chlorinate the thorium oxide feed (section
3.7+7)s A possible direct route to reducing UClk.may;be by
electrolyéis, releasing 012 which could, in turn, chlorinate ThO

2
feed.

343
 

Blanket salt should well accord batch reprocessing. The frequency
of blanket reprocessing will depend on
1. The tolerable 233U coancentration in the blanket which de~
pends on
da« Allowable power production in blanket which depends on
(1) Maximum desirable heat removal capability from
blanket salt
(2) Neutron damage to reactar vessel
b. The amount of 233U inver:tory to be tolerated in the
blanket, which depends on
(1) Proliferation hazard
(2) Financial worth of <oy
c. The loss in BG it causes
2. The optimum reprocessing economics which depends on
ae 233U separation efficiency
b. Blanket reprocessing costs
3. The demand rate for 233U in the core shim tank which depends
on
a. The tolerable size of the shim tank, which depends on
(1) Criticality hazard
(2) Proliferation Hazard
(3) Financial worth of *2oU
b. Fuel burnup rate in the core.
Regarding BG (lc above), Figure 3.7=-8 shows that Taube found a sig-
nificant BG loss as the Pu/ZBBU ratio went from 10 to 10°<. An

.. 2 .
HMCFR(Th) should behave similarly with 33U/Th ratio,

349
 

 

 

 

1.507
BR
1.40 -
4
1.30 -
107 1072
Delative P concentraticn: U=21% = 1.0

M e A - 230
Fig. 3.7-8, Impact of Fisgzil ‘Pu Contant in ths Blanlkst uren

3,7.L. Starter uels

3.7.L.1 Starter fuel for the core. Until 233U becomes avail-
* Nl i - . e bl baa V- 3 Y 235U

able, :CFi(ih) reactors must start up with nighly-—enricned

or Pu. Pu most clecsely matches the critical dimensicns and

233

characteristics of the =quilibrium U cycla. It also avoids
mining new uranium (at a rate of 10C0 parts U fcr 72 parts 235U)
and uses up existing stocks of Pu.
This conflicts with the desire to avoid Pu, but requires only the

one iransport for the starter core fuel. To start with highly-~enricihed

2 —-— - - * -
BDU vouli later require 2 core seometry caonge (smaller ftube diameter

350
 

or wider skew angle) or thorium diluvion of the 233U. Whenv233U be-
comes available, it can start up new reactors,

Replaceable tubes would also permit variable core dimensions and
compositions One might then even start with a slow neutron spectrum
and core mix of lower enrichment until dufficient 233U is bred 1o oper-
ate the desimm CP2(Th),
3e74de?2 Prevaration of ThCl, for blanket. Th principelly occurs in
rOCKS as Th02, toe same Torm considered for solid fuels. The necessary
technology already exists for converting oxides into chlorides: <he
principal method to recover tliorium end uranium from speant solid fuels
naopens to ve =z caloride wvolc-:ility process whnerein ThCl, occurs zas an

4

intermediate step,

In this process, a gas mixture of 012 and CCl4 attacks the oxides

at 700-85OOC to form volatile ThCl4 and other chlorides, Sintered

metal filters remove solids from the gas; salt sorption traps at con=

trolled temperatures separate the chlorides into different fractions,

One could also form ThCl, while

4
(1) Reducing blanket UCL 4 to UCl3:
Th + 4 UCl4—=-4 UCl3 + ThCl4

(2) “Gettering" excess Cl (released from the UCl, molecule by

4

fission) from the core salt

Th + ACL—» Thr314

However, these methods require metallic Th which is pyrophoric arnd

difficult to obtain from ThOZ.

Another pecssibility would be to use the dCl product from blanket

351
salt reduction along with lots of heat to drive the reaction

e O
ThO2 + 4HC1 ThCl4 + 2H2

Most of the heat would be recoverable and might come from the reactor

itself,

3.7.5 Fuel Cycle Summary

Fig. 3.7-9 summarizes a typical MCFR(Th) fuel cycle. On equi-

librium cycle an MCFR(Th) blanket consumes natural thorium and breeds

233UClh . Blanket reprocessing separates out the UClh, reduces it

to UClB’

and feeds it into the core as shim. Small anounts of Th,

23L‘U, and.232U also carry over in chloride form.

Continuous core reprocessing should result in
p £

1.

2,

Gases vented tc atmosphere following suitable decay
Very small lang-lived FP waste which could remain at the
plant until the end of the plant life

Mo actinide waste

A few.commercial byproducts, e.g. S, nor-radiocactive FP,
and possibly 9OSr and 137Cs radiation sources

Renovated core salt still highly radioactive, mixed in
with the blanket product shim. DMost of this radiocactive,

weapcns-grade mix returns to the core, but some goes cut

as BG in a form non-conducive to intimate handling.

Section 2.5.2 advocates minimum transport of weapons-grade

and radioactive material. The one shipment of starter fuel in

and shipment of EG out constitutes the only major transport, nil

compared to the usual reactor fuel cycle.

352
 

3.8 Safeguards

3.5.1 llational Fuel Inventory

Although cut—of-core cooiing greatly increases the fuel in-—
ventory in the primary circuit, fuel safeguards does not suffer:

l. Fuel diversion from the primary c¢ircuit requires shut-

ting down the operating power plant

<. High radicactivity makes diversicn unwieldy.

The inventory in the reprccessing stream will depend cn the
reprocessing mode (batchwise vs. cantinuous) and on the reprocessing
coals (degree of separations., The only other fuel accumulaticns
will be in ths shim tank ancd in the breeding gain experi, both
highiy radiocactive.

The absence of fuel febrication and fuel elemené reprecessing,
and associated_trahéportation eliminates most out~of-plant inven-
tory. Thus, unlike other reactor fuel cycles, the national in-
ventory of an MCFR basically comprises only the power plant inven-—

tories.
3.8.2 In—-plznt Diversicn Potential

FP and 232U-daughter radicactiviiy in the molten fuel should
discourage diversion prior to reprocessing. Selective reprocessing
could remove just the neutron absorbars and corrosive agents,
while allowing high—-activity non—voiatile emitters to remain.

(This would simultaneously manage some of the cumberscme waste).
After appreciable reactcr cperaticn, long-term radicactivity and heat

generation retes may prohibit salt use for ieapons.

353
The most sensitive (weapons-attractive) material should be the
low=burnup UCl3 or UCl4 extracted from the blanket., Processing re-
motely and then mixing immediately with (reprocessed) radicactive
core salt should safeguard the UClB' Lo stage should ever hold
encugh low=burnup 233UCl3 to werrant seizure for quick bomb

manufzcture.

Je8.3 Facility cdificaticon

Presumably a country cculd modify its MCFm(Th) to produce Pu
by simply substituting UClAL fer ThClh in the blanket. This could
te dene either in the salt scoution or in a separate capsule.

But then a research reactor c¢- other neutrcn sources would serve
2lmecst as well. Thus safeguards cannot prohibit such‘steps: they
can only seek to induce ramifications to, and external sigrals of,
such modification. Signals would include higher heat production
in the blanket and fewer 2.56-iieV photons from 232 daughters.
Designing a blanket with limited heat remcval capability should

preclude significant Pu manufacture.
3.8.L, Desired Radicactivity in Transported Sreeding Gain

The best opportunities for diversion or theft of core salt
would oceur with stockpiles and transports of breeding gain. Still
even though the fuel be packaged for safe handling in transport,
the presence ox 232U daughters and radioactive fission products in
the Tuel will make it unatiractive for handling in weapcns con-
struction. Furthermore, this radicactivity will necessitate un-

wieldy shieldinz during transport. 7he subject radioisotopes should

3%
 

1. Have low neutron cross secticn to avoid neutronics penélty

2. Be nan-volatile so as to not pose a hazard in a transpcrt
accident

3. Be difficult to separate from the core salt.

Troublescme long=lived fission products might be ideal.

3.5.5 Hadiclorical Sabotaze Threat

Fission reactors inherently induce high levels of radioactivity.
A terrorist could deliberately release radiocactive gases and, by
means of explosive devices, radiocactive aercscols. Although an
M3E piant centinucusly purgec its fuel, much of the problem re—
mains: the same amcunt of fission products have still been gen-
erated per l/e and must eventually be disposed of (transported
from in-plant safe storage to some final safe stcrage). Cnly
semipermanent storage at the MSR site can alleviate.the transpor-
tation risk relative to other reactors. But even then, such a
high fission product concentration might alsc attract saboteurs.

In cantrast, a fusiom reactor produces only tritium which is
conceivabiy dispersable. In-plant recyeling further mitigates the
prcblem. However, the only near-term economical fusion reactor is
projected to be the hybrid which will produce weapans-grade fissile

fuel again.

355
 

3.9 Access and Maintenance

The geals fer all nuclear power plants include efficient,
reliable, and safe reactor operation. This entails adeguate
inspecticn and opportunity for maintenance and repair. iith lMSREs
one can drain the fuel, flush the system with clean salt, and then
replace the whole reactor vessel or individual heat exchanger
tnits. However, with out-of-core cooling, the primary circuit
emits delayed neutrens; these activate circuit equipment and the

secondary coolant, inhibiting access and maintenance there.
3.9.1 Ccmpzztress of an [“CrA

Figure 3.%1 indicates the compact laycut cf an MCFR and
compares its cverzll size to that for a typical L:FBR and a
lower-powered MSPR. In British HCFR{Pu) designs, no item of
plant exceeds 6.5 m diameter; this permits prefabrication of the
vessel in the shop, thus reducing construction time and interest

charges.
3.9.2 Intrinsic Feliability

The MSR avoids intricate mechanical devices: it needs

1. Mo complex refueliing mechanisms

2. Tew of the clcse tolerances present in solid=-fuelled re-
actors

3. Mo mechanicai control devices in the core or blanket.

This implies the high system reliability of a chemical or hydraulic

356
plant: pumps, heat exchangers, and containing materials will cperate
reliably so lcng as they do not exczed proven limits of corrcsicn,

thermal expansion, and salt velccity.

3,9.3 MS3EZ Ixoperience

MSRT demonstrated that with proper design crne can maintain an
active fluid-fuel system without undu: difficulty. Such design
should entail easy replacement of pump and heat exchanger or in-
situ repair. As an example, a U-tube heat exchanger allows the
major components to expand and zlsc permits plugging any leaking
tubes from the less—active coolant circuit side. These features

do cost some extra fuel inventory though.

3.9.4 Comparisen cf Primary Circuit Cenfigurations

In nuclear reactors radiation inhibits access to the primary cir-
cuit., The liquid nature of MCFR fuel allows the removal of most
of the radiation sources during reactor shutdown.

Cne would like to design the primary circuit fio allow reduced—
power operaticon while repairing a small ex=—core portion of it in a
shielded cell (Fig. 3.9=2). Then removal of cne ™tube" circuit from
operation would still leave (p-1)/n of the reactor in cperation. Ad-
versely the primary circuit must lie close to the core so as to min-

imize out-of-core fuel tnventory.

357
 

Because of this the radiation fields of {he core and adjacent sub-
channels overlap, Therefore unless robots can do the job, replacement
or repair of any component (heat exchangers, pump, piping, etc.) will
probably require removal of all the radiation sources: 1i.e,, shut
down and clean out of the whole reactor,

In that event one might ask, are multiple loops in each sub-
channel still needed? Probably they still buy you time to wait until
a scheduled maintenance period occurs = like that for reolacing the

in=core tubes,

one of two
subchannels
per core fube

  
 

/- shielding

Teactor vessel

Figure 3,9~2, Top View of a 3~Tube Reactor with Subchannels
in Shiezlded Cells: :

-

358
 

3.9.5 Reactor Maintenance/fieplacement Procedure

Upon shutdown the core fluid may drain or be pumped into safe
gecmetlry tanks. Cperators could then flush the primary circuits,
first with (i_rradiated) blanket salt and then with fresh blanket
mix. The latter might first cleafl‘up tine blanket region
itself. Zach mix would need a separate drain tank. Both tanks
and iines might require awdliary heating. ‘fater or steam couid
accemplisn the final cieznup: chiorides dissolve easiiy in water.

Jith a clean system, operators could dzcouple and replace the
grapiiite tubes just like fuel elements, only much faster:

1. Mo delay time 1s needed for radiozciivity to die down

2. Low radicactivity gr=2atly eases tne handling prcblems

3. Spent tubes will require no extensive storage as for spent

fuel elements; they should be easily disvosed of, |

L. Cnly a feuw (N=3-6) tubes are involved.

Core tubes will preobably require the most frequent replace-
ment of all the solid reactor components due to the combined effects
of radiation, corrosion, and high temperature. Frequency of re-
placement might be similar to that for LR and LIiFBR refuelling
(6-12 mos). Even more frequent intervals can be tolerated without
significant penalty. The costs of replacement tubes will be neg-
ligible alsc. C(ne would like to remove the tubes without dis—
turbing the reactor vessel. Cnce in a great while, the reactor

vessel or a graphite vessel liner could also be replaced,

3.9.6 Reactor Shieiding

Cut-cf-core ccoling may cause scme unique shielding problems

359
 

due to the circulation of delayed neutron emitters. Assuming a
delayed neutrcn fracticn of 0.003, a total-to-core volume ratio

of four,:3-= 3 seccncary neutrans per fissicn, a core power den—
sity of 10 Miith per liter, neutron multiplicaticn facter k in a
5-in. pipe= 2/3 (a reascnable arbitrary guess), ocne gets a neutren

scurce rate cf

 

 

T10 MW leV LD delaryed nemts. cere vol.
8 T e X ——— X —— X
corg fter  l.eozxiWs 20 May 4ot vol.
: . . - -1
1 lder «—L (subertical = B.4 x 10’7 nem?s

I03 ch 1-k mi H\'Pll-tlh.ln

The surface fiux an a cylindrical source is [149]

¢ = E§_i_5‘ & (n,R/A)

where A = neutrcn mean free path (cm)
B = bulldup factor, normally applied in use with photcns to
add scattered flux in with the uncollided flux. Thus
B is usually near 1.0.
n = haif-height of the cylinder measured in units of \; in

evaluating G, h = 6 is the same as h = oo

R = cylinder radius (cm).
tan™ (a/x)

G(ab)= J ixj o~ X Secy J.j) muxG—=G(°°s‘=°)=l-0

Thus on the surface of a 5inch diameter pipe, with ) = 10 cm,

G( a0, 2.5 x 2.54/10) = 0.73 and § = 1.825 B-S=105n co < s L.

Treating a pipe as 2 2 meter-long line source, then at a distance

1 meter away from its midpoint [149]
2
= St . -2 -
¢ =37 (45°,0) = 0323 + 5 2102 n ca2s”!

360
 

vhere a = 1 m (distance away)

‘b- e ' '
F (g, b) = § e® e g
o
In summary, an unshielded 5~in pipe will produce delayed neutron

2 1013 n c:m-2 sec-l nearby.

fluxes of the order 10

This flux level could appreciably activate the primary circuit
environment, causing delay end difficulty in maintenance. Routing
cltiernate subchannels through separate shielded cells outside the
vessel (Fig. 3.7=1) would partly circumvent the problem: then if
one scbchannel broke down, the operitor could divert flow to another.
After a suitable decay time the broken subcircuit could be entered
and fixed.

The delayed neutron radiztion stems from relatively short-lived
emitiers; it will quickly stop once reactor operation does. The high
gamma radiation from fission and activation products lasts much

longer: it will continue during shutdown. Continuous fission

product cleanup helps to keep these gamma radiation levels down.
3.9.7 Radiation Sources in Reprocessing Equipment

Section 3.5.10.1 reported the neutron production from decay
alphas. To avoid these neutrons, the process should avoid use
of light elements, especially Be,F, B, Li, A1, O, Mg, and Na
(in decreasing order of importance).

The presence of 23 2U y especially in core salt reprocessing,
will also produce penetrating radiation. Section 3.5.10.2 dis~

cusses this in more detaile.

361
3.10 Auxiliary Plant

3.10.1 Power Cycle Options

Generally the primary coclant transfers heat to
1. An intermediate liquid coolant which generates steam for
a. turbine, or to
2. A gas coolant which passes directly to high-efficiency gas
turbines and/or process heat applications,
Melting point, corrosion, and activation (by cielayed neutrons) cone

stitute the main concerns in the auxiliary plant.

3.10.1.1 TIntermediate liguid coolant for steam cycle. Liquid
metals and fused salts, beczuse of their high boiling points
and thermal conductivities allow operaticn at high temperatures
and low pressures. This promates excellent heat transfer with-
out expensive and hazardous high pressure equipment. Low pressure
especially permits use of thin heat-exchanger tubes.

Research earlier on Na-graphite reactors and later on LMFBRs
has produced a broad base of Na technology. Similar technology
exists for the low-melting eutectic Na-K mixture. Either one, but
especially NaK, avoids problems of coolant freezing. NASA-Lewis
has also pioneered the use of K with a K~vapor turbine, getting up
to 65% efficiency. That cycle would probably need more study though.

Unfortunately, alkali metals react vigorously with both steam
and molten salt. A small leak on the high-pressure steam side could
spell disaster. If the alkali leaked into the molten salt , removal

by processing would be hard. Sodium would react with UCl3 to form

362
NaCl and pure U metal; the latter could plate out or otherwise
accumuilate.

Lead behaves better chemically as it.reacts little with either
salt or steam and should not enlarge any leaks. Lead costs a lot
but that has not discouraged the British from checosing it. It will
require development of suitable pumrs and components especially in
regards to corrosion. The inertia associated with lead's high density
will also require high strength construction.

Lead will need auxiliary heating to prevent freezing during
low power and start-up conditions. The cooler parts of the lead
circuit could use low alloy steels but regions above SOOOC may
require other materials: the superheater will need duplex tubing

as lead attacks nickel~bearing alloys.

Molten salts compete favorably with liquid metals: they
exhibit thermal conductivities intermediate to water and the
poorer of the liquid -metals, Their specific heat capacities parzllel
water's. Furthermore, an intermediate coolant of molten salt
should more closcly match the primary salt in physical proper—
ties, thereby reducing freezing and thermal stress problems.
They will cost far less then liquid metals. Table 3.10-I shows
that a wide choice in eutectic melting points exists among
candidate salt mixtures.

Considerable experience exists in pumping and general use
of molten salts, especially in salt baths and petroleum refining
apalications which feature ccnstant high temperatures and con-—

tinuous system operation. Salts work well with Hastelloy N.

363
 

Table 3.10-1 Eutectic Melting Points of Some Salt Mixes [112]

Salt Mix M.P. (°c)
ZrF A/NaF 500

NaBF , 408

ThC1l h/PbClz ' 395

92 NaBF 1/8 NaF 384,

2 (LiF)/BeF, 364

Thel, /NaCl | 360-370
ThCl h/PbClz/LiCl 325-335
NaliQ,/Nalo 3 /K0 142
NaAlCl, 108

However steam leakage into the coclant salt could produce
locally high corrosion. Processing difficulties would arise
if the ccolant leaked into the fuel salt. For these latter
reasons the British decided against molten salts for the sec-

ondary coolant.

3.10.1.2 Coolant for gas cycle. 002 or other oxidizing gas leak-

ing into the fuel salt, would react to form TO That could over-

o0
load the oxide removal section of the salt cleanup plant.

Inert helium avoids this problem. He cooling is also under
development for other reactors along with a He gas turbine. It
should work well with molybdenum heat exchanger tubing although
this is not yet fully endorsed.

Both France and the Soviet Unicn are considering gaseous

364
aluminous chloride to power the first stage of gas turbines; dis-
sociating MO, (N0 A) gas, the second stage.

Lower turbine working pressure with gas should permit more
component prefabrication, That should lower installation times
and save on coflstruction and interest charges. Avoiding the steam
generator and tertiary circuit also reduces costs, the extent
depending upon the gas cycle efficiency.

Poor heat transfer properties of gas will penalize the cycle
with a larger heat exchanger. This means higher out-~of—core fuel
salt inventory. One must also analyze the safety of a pressurized

secondary coolant.

3.10.2 Auxi.iary Hardware

3.10.2.1 Pumps. Pumps that circulate molten salts experience un-
usual metallurgical, chemical, physical and mechanical c:énditions.
Salts exhibit temperature-dependent physical properties. At tem—
peratures near 700°C they corrode many structural materials and
deteriorate chemically if exposed to air, water or water vapor.

ORNL successfully tested single-stage centrifugal salt pumps
in the MSRE and in loops at 650°C up to 16,000 hours. Two-stage
semi=-axial pumps may require basic development work. Both pump
designs will require development and testing to successfully in-
crease sizes to the range needed for a 1000 MWe plant.

Both pump designs separate the high~temperature pumping ele~
ments from the primary shaft seal and bearings: this allows the
latter parts to operate near ambient temperatures and to use con-

ventional lubricarntis. An inert gas buffers the space between the

365
high-temperature pumped salt and the primary pump shaft seal.
Elastomeric seals and lubricants must be protected from excessive
gamma heating and radiation damage.

Initially, mechanical pump designs avoided rubbing parts in the
salts; now pumps with journal bearings in the salt have operated
successfully.

No direct experience exisis with high power, continuously-
running lead pumps. The design principles would probably follow
the sodium pump design developed for IMFBRs but with higher power

and strength requirements.

3.10.2.2 Core/blanket seal. Section 3.7 emphasizes guarding against
core fluid leaks more than tianket leaks. Therefore, the core tubes
should be a single entity throughout the reactor witfi no pressure
seal between core and blanket salt.

Cne solution might be to route the core tubes through a secand
tube or sleeve which penetrates the reactor vessel. However, such
a sleeve will itself experience high radiation damage and require
replacement. It would also thermalize neutrons. Thus we prefer to
create a blanket seal on the core tube (Figure 3.10-1). Low blanket
pressure enhances the feasibility for it. One could also design a
catcher for limited blanket leakage as part of the blanket reproces—

sing stream. Then, no seal at all might even work.
reactor
seal on vessel
tlanket salt
alternative
|

blanket
salt
region

 

 

 

 

core salt tube

 

[
|
I
|
l
l
|
l
|
|
|
l
n
!

 

 

blanket
salt
region
\
~
C:;;;z;et TN -
salt \ /
catcher | |
alternative f :

Fig. 3.10=1 Alternative Interfaces Between Core Tube and

Reactor Vessel.

367
3.10.3 Filling, Draining and Dump Systems

Two tanks below the reactor vessel serve for filling and drain-
ing. Each can contain the whole salt inventory of core or blanket,

They provide for quick dumping of the core and/or blanket or for

accumilating salt leaks from the reactor vessel. Two more tanks
function as standbys and as delay tanks for salt removal for pro-
cessing, one each for core and blanket. An emergency ccoling
system would remove the decay heat from any tank through the same
lines used for preheating.

Mechanical pumps or gas pressure can transfer the salt. For
initial fillipg the fuel or *lanket salt will be brought to site
in solid form. Electric immsrsion heaters or steam coils that
pass below the surface of a salt bath melt the salt under inert
gas in swall non-critical batches. Heating a solid bath of salt
from the bottom alone can develop sufficient pressure to rupture
equipment or expel the molten salt through the solid surface.

Once molten.the salt can be pumped.into the réactor.

The primary salt circuits will require preheating to about
600°C ; the secondary, less. In some areas it will be more economic
and convenient to heat whole spaces such as the reactor vault and
processing plant; trace heating is better for longer lengths of pipe-
work as in sgcondary pPipes to the steam generators. Submergsd
centrifugal pumps circulate the salt. They are of a type
which permits no cantact of the salt with the packing gland., The

ORNL HMSRE encountered no particular problems with preheating.

368
The ISR uniquely separates, concentrates, and stores part of its wastes
«7ithin the plant for neutron economy, safety, and corrosion protection
of the reactor. Being already in the form of a concentrated heat
source suggests the possibility of dissipating the decay heat by put«
tinz it to works One use could be for preheating. Felium or a low-
melting point salt like Flibe (36300) could be the heating agent. An
auxiliary heater would orovide a backup heat source, espvecially oricr

to accumulating sufficient fission product waste,

3,10,5 Overall Plant Size

Building laycut in the Zritish MCFR(Pu) design is compact: the
reactor vessel and primary circult are small; the plant requires no
elaborate fuel handling machinery. The skewed=-tube configuration and
tube replacement machinery here forfeit some of this compaciness.
If inline pyrochemical reprocessing proves practical and economical,
it should fit in the building with little size increase. This in-
cludes the attendant control and maintenance areas. If feasible,

high pressure storage of fission gases would also save much space.

369
 

3.11 Zconomics

oceveral earlier sections have mentioned potential economic

advantages with MSRs, and MCFRs in particular. This section

attempts to summarize these and others; many come from an in-

ternally-published British study for an MCFR(Pu) [29]..

7.11.1 British Study Results

The British compared a lead-cooled MCFR(Pu) to an LiFER of year

1990 desiszme

3.11.1.1 Capital costs. Here they found

1.

Le

5.

The MSFR avoids elaborate control systems and complex fuel

handling equipment. This implies cost savings through
compactness and simplicity. However necessary auxiliary
systems increase costse.

Component smallness allows shop fabrication in nearly all
cases,

The simplicity of a chemical type of plant should

a. Base erectlon, testing, and commissioning procedures

be. Reduce construction time with consequent savings in

interest and overhead charges.

If full advantage could be taken of the higher temperature

potential , further savings on plant costs could accrue

with the lead secondary-cooclant/steam plant system.

Should secondary cooling by helium prove feasible (satisfy
safety requirements), the associated temperatures possible

with it might permit use of gas~turbines. Significant

370
savings might then accrue from skipping a tertiary cir-
cuit and using a layout integrated in a pre-stressed con-

crete pressure container vessel,

3.11.1.2 Fuel cycle costs, The British study assumed

1. Off-site solvent extraction of heavy metals

2. Removal of residual fission products

3. Cn-line removal of volatile fission products and salt

clean-up.

L« A reprocessing cycle of three months cooling-off delay

period and three months processing and transport time.

5. A one-month stockzile of fresh fuel at the reactor site.
Comparison of fabrication ard reprocessing costs showed a definite
advantage for an MCFR over the oxide~-fuelled LMFBR, as expected.

If the maximum fuel salt temperatures can increase from the
800°C of the reference lead-cooled case to 920°C, the MCFR(Pu)
inventory could reduce by 25%. It would then approach that for
the oxide-fuelled IMFBR; the doubling time also would then fall
to a value .Ssimilar to that of the LMFBR.

The British also examined the on-line pyrochemical process in
the hope that it would reduce inventory over the aqueous process.
While it appeared technically feasible, the initial-cost estimates
were discouragingly high. These stemmed from

1. Batch processing of small *non—critical" quantities,

2. The remote operation and maintenance involved.

This area needs further study.

371
 

3.11.1.3 Summary. MCFRs appear to offer better economics than
LMFBRs. The fuel cycle costs above assume the well-established
aqueous processing for heavy metal; :.an on-line high-temperature
process only cleans the salt up. Potentially large capital savings
in a high temperature MCFR warrant further study of its plant

design, mainftenapce problems, and safety.

3.11.2 Transvortation Costs

Transportation of convenitional spent reactor fuel incurs large
eosts through

1. The high radiocactivity of the fuel

2. The considerable residuzal heat

3. The presence of fissionable materials.
These necessitate

1. FExtensive shielding materials

2. Small packages or package cooling

3. Cumbersome handling procedures

L. Security (safeguards)

5. Special accident procedures during transport.
The transportation problem largely determines the consuruction sites
for reprocessing plants and the cooling time of the spent fuel before
reprocessing [F44]‘ An MSFR with in-plant reprocessing avoids all of

this except for BG transport.

372
 

d.11.3 Cutage Penalties

An MCFR may experience less downiime than a conventional LWR:
MCFR core tubes should wear out more frequently, but low activity
and the fewness of rmumber should simplify the task. However, problems

related to peripheral equipment malfuncti. .. still apply, and could be

greater with a new concept.
3.11.,, Other Costs

Graphite tubes should be cheap. NaCl will be dirt cheap. T,
abounds in nature more than UOE. An MCFR avoids fuel element fab-

rication and enrichment costs. High flow velocities incur high pump—

ing costse.

313
 

L0 SUMMARY

4«1 Pu, Proliferation, Safeguards, Security, and Waste Management

Much concern with the nuclear fuel cycle centers on Pu due
to its carcinogenicity and its potential use in a weapon. One
desires close control less it pollute the environment or be di-
verted {overtly or covertly) to weapons use. Once chenically sep-
arated (relatively easy to do for Pu) it can be put into a pre~
assembled bomb within days. The main difficulty will be to pro-
perly implode the device. National or large subnational groups
might be able to practice that trick beforehand.

Present and prospective solid fuel c¢ycles involve extensive
fuel logistics including elament fabrication, decladding, and post-
operation storage. Molten salt reactors skip these steps, thereby
reducing much of the pollution and diversion dangers.

A reactor an the Th fuel cycle minimizes production of Pu and
heavy actinides by beginning much lower on the atomic weight scale
and usefully consuming them. Fast reactors succeed even better by
enhancing fission, especially for 23&U. This pertains especially
to fast molten salt reactors with a very hard neutron spectrum.

All reactors use or produce weapons—grade fissile fuel. In
the Th fuel cycle, the predominantly—233u fissile fuel discloses
its presence by emitting a hazardous, characteristic highly-
penetrating gamma ray. Any transformation of the reactor to the
U/Pu fuel cycle will evidence a reduced gamma signal.

An #5R alsc allows one to continuously remove the principal

(gaseous) radiation hazard from the fuel and store it away from

374
 

the reactor. This fact plus the ability to transfer the whole
core further underground at a moment's notice should enhance

security ¢f the plant from attack.

L.2 Reactor Design

Exiensive evaluation of core salt properties favors cheap
NaCl as the carrier sait, ThClL for fertile material, and UCl3
for fissile material. High fissile salt proportion (Th/U = 0.5,
30% NaCl) will yield high BG (up to O.4) and enhance actinide
fission due to hard neutron spectrum. However, it also causes
high deubling time and more frequent replacement of radiation-
damaged core tubes.

Doubling time reduces to near 20 years at lower fissile con-
tents (Th/U = 1 to 2, 30% NaCl). The optimum composition should
lie in between these iwo points, depending on the assignment of
priorities. In earlier years of operation one might operate with
a low fissile cantent to achieve low DT, low_ flux levels, and
softer spectrum. Then in later years, when more is known about

radiation damage and the press for available 233

U is diminished,
the plant could switch to a salt of higher fissile coantent so as
to avoid actinide buildup and produce more surplus neutrons(highe-
er BG).

The blanket region could benefit from a thermalizing carrier
salt like 7LiCl or a Iluoride instead of NaCl, but cost and high
melting points, respectively, discourage those.

The optimum core configuration for BG and DT appears to be

a few frequently-replaced, skewed, graphite tubes in a blanket

375
 

bath 2 meters thick: a cost-effective solution to the high flux
levels inherent to this concepts Inner blanket regions definitely
detract: they decrease BG without decreasing the out-~of-core in-
ventory.

L3 Technical and Economic Feasibility

Fast molten chloride reactors have been cursorily considered
before but mainly for the U/Pu fuel cycle. The ORNL MSR program
showed the feasibility of fuel salt circulation. The combinaticn
of that experience and MCFR research (out-of-pile experiments and
theoretical studies, so far) provide a basis for believing the
concept will work.

Chemical stability and corrosion of molten salts are fairly
predictable. Low vapor pressure of the salts enhances safety and
permits low pressure structural components.

Molten fuel state and cooling ocut~of-core simplify component
design in a radiation environment. They forego complicated
refueling mechanism, close tolerances associated with solid fuel,
and mechanical control devices. Molten state and low vapor
pressure of the salts also offer inherent safety advantages.

Graphite and Mo alloys and coatings appear as promising
candidates for prisary sait containment, both in and out of core.
These high temperature materials may permit high fuel salt
temperatures (above 1000°C). This can reduce fuel salt inventory
in the heat exchanger and allow gas turbine cycles and/or process

heat applications using helium as an intermediate and final coolant.

376
 

Use of the MCFR on the Th fuel cycle should greatly expand the
amount of available fuel: three times as much Th exists as U.
The high BG accompanying very fast neutron spectra in some MCFR
designs will enhance the economic retrievability of extensive

low-grade U and Th ores.

377
1e

2a

Ja

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