Lecture 3: Forensics in Nuclear
Applications
• Readings:

Nuclear Forensics Analysis: Chapter 3 Engineering Issues
• Manipulation of natural radionuclides
• Enrichment

Gas

Laser

Solution
• Fuel cycle

Reactor types

Reactions

Reprocessing

Purification
3-1
Manipulation of Natural Radionuclides
• Exploitation of fissile and fertile isotopes

Fissile: 235U

Fertile: 232Th, 238U
• Enriched 235U

Simplifies engineering aspects of nuclear reactors

Produces uranium enriched in 238U
• 233U and 236U from reactions in reactors
• Pu isotopes from neutron capture on 238U and subsequent capture on
resulting 239Pu
3-2
Mining activities
•
•
Non-nuclear uses of U and Th

Counterweights

Ballast

Munitions

Lantern mantles

Mg-Thoria
Common steps in U and Th mining activities

Preconcentration of ore
 Based on density of ore

Leaching to extract uranium into aqueous
phase
 Calcination prior to leaching
* Removal of carbonaceous or
sulfur compounds
* Destruction of hydrated species
(clay minerals)

Removal or uranium from aqueous phase
 Ion exchange
 Solvent extraction
 Precipitation

Use of cheap materials
3-3
U Ore leaching
• Acid solution methods

Addition of acid provides best results
 Sulfuric (pH 1.5)
* U(VI) soluble in sulfuric
 Anionic sulfate species
* Oxidizing conditions may be needed
 MnO2
* Precipitation of Fe at pH 3.8
• Carbonate leaching

Formation of soluble anionic carbonate species
 UO2(CO3)34
Precipitation of most metal ions in alkali solutions

Bicarbonate prevents precipitation of Na2U2O7
 Formation of Na2U2O7 with further NaOH addition

Gypsum and limestone in the host aquifers necessitates carbonate
leaching
3-4
Recovery of uranium from solutions
•
•
•
Ion exchange

U(VI) anions in sulfate and
carbonate solution
 UO2(CO3)34 UO2(SO4)34
Load onto anion exchange,
elute with acid or NaCl
Solvent extraction

Continuous process

Not well suited for carbonate
solutions

Extraction with alkyl
phosphoric acid, secondary
and tertiary alkylamines
 Chemistry similar to ion
exchange conditions
Chemical precipitation

Addition of base

Peroxide
 Water wash, dissolve in
nitric acid
 Ultimate formation of
(NH4)2U2O7 (ammonium
diuranate), yellowcake
 heating to form U3O8 or
UO3
•
TBP extraction

Based on formation of nitrate
species

UO2(NO3)x2-x + (2-x)NO3- +
2TBP UO2(NO3)2(TBP)2
3-5
Yellowcake to Fluorination
Heat to 350 ºC
Yellowcake (Na2U2O7)
Conversion to UO3
H2 Reduction, 700 ºC
UO2
HF, 450 ºC
UF4
Mg
U metal
F2
UF6
MgF2
3-6
Th ore processing
• Main Th bearing mineral is monazite

Phosphate mineral
 Also contains lanthanides
 strong acid for dissolution results in water soluble salts
* Sulfuric acid at elevated temperature (200 ºC)
• Th goes with lanthanides

Need separation from lanthanides for reactor usage

Separate by precipitation

Lower Th solubility based on difference in oxidation state
 pH 5.5 removes Th
• Oxalate or hydroxide precipitate calcined

Oxide formed, dissolved in 8 M HNO3

Extract in TBP/kerosene
 Separate from Ce by reduction with NaNO2 of H2O2
 Back extract Th into dilute nitric acid
* Any U present stays in organic phase

ThO2 prepared by heating to 700 ºC
3-7
Uranium enrichment
• Different enrichment needs
 3.5 % 235U for light water reactors
 > 90 % 235U for submarine reactors
 235U enrichment below 10 % cannot be
used for a device
Critical mass decreases with increased
enrichment
 At 20 % 235U critical mass for reflected
device in around 100 kg
Low enriched/high enriched uranium
boundary
3-8
Uranium enrichment
• Need to exploit different
nuclear properties
between uranium isotopes
to achieve enrichment
 Mass
 Size
 Shape
 Nuclear magnetic
moment
 Angular momentum
• Massed based separations
utilize volatile UF6

UF6 formed from
reaction of U
compounds with F2 at
elevated temperature
3-9
Uranium hexafluoride
• Colorless, volatile solid at room temperature
 Density is 5.1 g/mL
 Sublimes at normal atmosphere
 Vapor pressure of 100 torr
One atmosphere at 56.5 ºC
 Triple point at 64.1 ºC and 1133 torr
• Oh point group
 U-F bond distance of 2.00 Å
3-10
Uranium Hexafluoride
• Very low viscosity
 7 mPoise
 Water =8.9 mPoise
 Useful property for
enrichment
• Self diffusion of 1.9E-5 cm2/s
• Reacts with water
 UF6 + 2H2O UO2F2 + 4HF
 Depleted 238UF6 stored in stainless
steel drums
 Leaks self sealing by formation
of UO2F2
• Also reactive with some metals
 Does not react with Ni, Cu and Al
if pure (no HF)
 Indicative of work with UF6
3-11
Uranium Enrichment: Electromagnetic
Separation
• Volatile U gas ionized
 Atomic ions with charge +1 produced
• Ions accelerated in potential of kV
 Provides equal kinetic energies
 Overcomes large distribution based on
thermal energies
• Ion in a magnetic field has circular path
mcv
r
 Radius (r)
qB
m mass, v velocity, q ion charge, B
magnetic field
2Vq
• For V acceleration potential v 
m
3-12
Uranium Enrichment: Electromagnetic
Separation
• Radius of an ion is proportional to square root
of mass
c
r
B
• For electromagnetic separation process
 Low beam intensities
High intensities have beam spreading
* Around 0.5 cm for 50 cm radius
 Limits rate of production
 Low ion efficiency
Loss of material
• Caltrons used during Manhattan project
2Vm
q
3-13
Calutron
• Developed by Lawrence
 Cal. U-tron
• High energy use
 Iraqi Calutrons required
about 1.5 MW each
 90 total
• Manhattan Project
 Alpha
 4.67 m magnet
 15% enrichment
 Some issues with heat
from beams
 Shimming of magnetic
fields to increase yield
 Beta
 Use alpha output as feed
* High recovery
3-14
Gaseous Diffusion
• High proportion of world’s enriched U
 95 % in 1978
 40 % in 2003
• Separation based on thermal equilibrium
 All molecules in a gas mixture have same average
kinetic energy
 lighter molecules have a higher velocity at
2
2
same energy
m352v352
 m349v349
* Ek=1/2 mv2
v349
m352
352


 1.00429
• For 235UF6 and 238UF6
v352
m349
349
 235UF6 and is 0.429 % faster on average
 why would UCl6 be much more complicated
3-15
for enrichment?
Gaseous Diffusion
•
•
235UF
6 impacts
barrier more often
Barrier properties
 Resistant to corrosion byUF6
 Ni and Al2O3
 Hole diameter smaller than mean free path
 Prevent gas collision within barrier
 Permit permeability at low gas pressure
 Thin material
• Film type barrier
 Pores created in non-porous membrane
 Dissolution or etching
• Aggregate barrier
 Pores are voids formed between particles in sintered
barrier
• Composite barrier from film and aggregate
3-16
Gaseous Diffusion Barrier
• Thin, porous filters
• Pore size of 100-1000 Å
• Thickness of 5 mm or less
 tubular forms, diameter of 25 mm
• Composed of metallic, polymer or ceramic materials
resistant to corrosion by UF6,
 Ni or alloys containing 60 percent or more nickel,
aluminum oxide
 Fully fluorinated hydrocarbon polymers
 purity greater than 99.9 percent
 particle size less than 10 microns
 high degree of particle size uniformity
3-17
Gaseous Diffusion
• Barrier usually in tubes
 UF6 introduced
• Gas control
 Heater, cooler, compressor
• Gas seals
• Operate at temperature above 70 °C and pressures below
0.5 atmosphere
• R=relative isotopic abundance (N235/N238)
• Quantifying behavior of an enrichment cell
 q=Rproduct /Rtail

Ideal barrier, Rproduct =Rtail(352/349)1/2; q= 1.00429
3-18
Gaseous Diffusion
• Small enrichment in any given cell
 q=1.00429 is best condition
 Real barrier efficiency (eB) (qobserved  1)  e B (qideal  1)
 eB can be used to determine total barrier area
for a given enrichment
 eB = 0.7 is an industry standard
 Can be influenced by conditions
 Pressure increase, mean free path decrease
 Increase in collision probability in pore
 Increase in temperature leads to increase velocity
 Increase UF6 reactivity
• Normal operation about 50 % of feed diffuses
• Gas compression releases heat that requires cooling
 Large source of energy consumption
3-19
Gaseous Diffusion
• Simple cascade
 Wasteful process
 High enrichment at
end discarded
• Countercurrent
 Equal atoms
condition, product
enrichment equal to
tails depletion
• Asymmetric
countercurrent
 Introduction of tails
or product into
nonconsecutive stage
 Bundle cells into
stages, decrease cells
at higher enrichment
3-20
Gaseous Diffusion
• Number of cells in each
stage and balance of tails
and product need to be
considered
• Stages can be added to
achieve changes in tailing
depletion

Generally small levels
of tails and product
removed
• Separative work unit (SWU)

Energy expended as a
function of amount of
U processed and
enriched degree per kg

3 % 235U
 3.8 SWU for 0.25
% tails
 5.0 SWU for 0.15
% tails
•
Determination of SWU






P product mass
W waste mass
F feedstock mass
xW waste assay
xP product assay
xF feedstock assay
3-21
Gaseous Diffusion
• Optimization of cells within
cascades influences behavior
of 234U
 q=1.00573 (352/348)1/2
 Higher amounts of 234U,
characteristic of feed
• US plants
 K-25 at ORNL 3000
stages
 90 % enrichment
 Paducah and
Portsmouth
 Reactor U was
enriched
* Np, Pu and Tc in
the cycle
3-22
Thermal Diffusion
• Difference in velocity of different
masses at a given temperature
 Thermal gradient established
 Heavier mass concentrates at
lower temperature
 Ni inner tube high T
* 286 °C
 Cu out tube water cooled
* 64 °C
 mm spacing
• Gas introduced at 200 atmosphere
• 15 m tubes
 Length provides enrichment level
 2100 columns, enrichment to 0.86
% at ORNL
3-23
Gas centrifuge
• Centrifuge pushes heavier 238UF6 against wall with center
having more 235UF6
 Separate areas for collecting heavy and light isotopes
• Density related to UF6 pressure
 Density minimum at center of centrifuge
 p(r): pressure at distance r
 p(0): pressure at center of centrifuge
* Pressure less at p(0)
p(r )
e
p (0)
mw 2 r 2
2 RT
 m molecular mass, r radius and w angular velocity
• With different masses for the isotopes, p can be solved for each
isotope
m xw 2 r 2
p x (r )
 e 2 RT
3-24
p ( 0)
Gas Centrifuge
• Total pressure is from
partial pressure of each
isotope
 Partial pressure
related to mass
 Mass based
variation, used for
separation
• Single stage separation
(q)
 Increase with mass
difference, angular
velocity, and radius
• For 10 cm r and 1000
Hz, for UF6
 q=1.26
p x (r )
e
p ( 0)
m xw 2 r 2
2 RT
Gas distribution in centrifuge
q e
( m2  m1 ) w 2 r 2
2 RT 3-25
Gas Centrifuge
• More complicated setup than diffusion
 Acceleration pressures, 4E5 atmosphere from
previous example
 High speed requires balance
induces stress on materials
 Limit resonance frequencies
 Need high tensile strength
* alloys of aluminum or titanium
* maraging steel
 Heat treated martensitic steel
* composites reinforced by certain glass,
aramid, or carbon fibers
3-26
Gas Centrifuge
•
•
Gas extracted from center post with 3
concentric tubes

Product removed by top scoop

Tails removed by bottom scoop

Feed introduced in center
Mass load limitations

UF6 needs to be in the gas phase

Low center pressure
 3.6E-4 atm for r = 10 cm
3-27
Gas Centrifuge
• Superior stage enrichment when compared to
gaseous diffusion
 Less power need compared to gaseous
diffusion
1000 MWe needs 120 K SWU/year
* Gas diffusion 9000 MJ/SWU,
centrifuge 180 MJ/SWU
• Newer installations compare to diffusion
 Tend to have no non-natural U isotopes
Earlier diffusion included 236U
3-28
Aerodynamic Enrichment (Becker
process)
• Separation by
centrifugal force
applied to a gas stream
deflected by a barrier
• Feed mainly H2
 Increase flow
velocity by
reduction of mean
molecular mass
 Increase velocity
• Improves q
• Heavier gas along wall
3-29
Aerodynamic Enrichment
• Gas split into two fractions by knife edge
 25 % of gas to light fraction
• Highest q with high velocity small r
 Increase centrifugal acceleration through cell
Radius 100 micron
Gas velocity 300 m/s
* Centrifugal acceleration 9E8 m/s2
* qideal 1.015
• H2 concentration must be constant throughout
cascade
 Countercurrent will add in H2
• Process can be 3 dimensional (Helikon process)
 Stacked plates of cells
3-30
Laser Isotope Separation
• Isotopic effect in atomic spectroscopy
 Mass, shape, nuclear spin
• Observed in visible part of spectra
• Mass difference in IR region
• Effect is small compared to transition energies
 1 in 1E5 for U species
• Use laser to tune to exact transition specie
 Produces molecule in excited state
• Doppler limitations with method
 Movement of molecules during excitation
• Signature from 234/238 ratio, both depleted
3-31
Laser Isotope Separation
• 3 classes of laser isotope separations
 Photochemical
Reaction of excited state molecule
 Atomic photoionization
Ionization of excited state molecule
 Photodissociation
Dissociation of excited state molecule
• AVLIS
 Atomic vapor laser isotope separation
• MLIS
 Molecular laser isotope separation
3-32
Laser isotope separation
• AVLIS
 U metal vapor
 High reactivity,
high temperature
 Uses electron beam
to produce vapor
from metal sample
• Ionization potential 6.2 eV
• Multiple step ionization
 238U absorption peak
502.74 nm
 235U absorption peak
502.73 nm
• Deflection of ionized U by
electromagnetic field
3-33
Laser Isotope Separation
• MLIS (LANL method) SILEX
(Separation of Isotopes by Laser
Excitation) in Australia

Absorption by UF6

Initial IR excitation at 16
micron
 235UF6 in excited state

Selective excitation of 235UF6

Ionization to 235UF5

Formation of solid UF5 (laser
snow)

Solid enriched and use as feed
to another excitation
• Process degraded by molecular
motion

Cool gas by dilution with H2
and nozzle expansion
3-34
Chemical Exchange Enrichment
• Equilibrium exchange between different uranium species
 235UX+238UZ 238UX+235UZ
 Equilibrium constant K
Nominally 1, but variations due to quantum
properties
* Vibration states make lighter isotopes
concentrate in less tightly bound molecules
• Chemex process based on oxidation state variation
• Exchange reaction takes place between U(III) and U(IV)
 In aqueous solution
• Enrichment of 238U in the trivalent state
• Extraction of U(IV) from concentration HCl
 TBP based extraction
• Cascade of separations involving extractions
3-35
Nuclear Reactors
• Products of fission
 Heat
 Neutrons
Further fission
Capture
* Actinides, fertile-fissile reactions
Number of neutron reactions possible
 Fission products
• Range of initial conditions
 Enrichment levels
 Reflector
Material with low neutron reaction cross
section
3-36
Nuclear reactors
• Probable neutron energy from fission is 0.7 MeV
 Fast reactors
 High Z reflector
 Thermal reactors need to slow neutrons
 Water, D2O, graphite
* Low Z and low cross section
• Power proportional to number of available neutrons
 Should be kept constant under changing
conditions
 Control elements and burnable poisons
 k=1 (multiplication factor)
 Ratio of fissions from one generation to the
next
3-37
* k>1 at startup
Nuclear reactors
• Control of fission
 0.1 msec for neutron from fission to react
Need to have tight control
0.1 % increase per generation
* 1.001^100, 10 % increase in 10 msec
• Delayed neutrons useful in control
 Longer than 0.1 msec
 0.65 % of neutrons delayed from 235U
0.26 % for 233U and 0.21 % for 239Pu
• Fission product poisons influence reactors
 135Xe capture cross section 3E6 barns
3-38
Nuclear Fission
•
•
•
•
•
Demonstrate by CP-1

Graphite purified from B

Cd control

Moisture control

2 kW, air cooled
Basic reactor components

Fuel
 For forensics fresh and used fuel considered

Cooling
 Air, water, CO2

Cladding

Moderator
 Protons
 Carbon
* Carbon needs more collision (114 versus 18), but has smaller reaction cross section
* Windscale and Chernobyl graphite moderated
Pu isotopes

For devices, minimize 240Pu
 240Pu:239Pu 0.05 to 0.07

Different ratio for power reactors
Submarine reactors

Water both moderator and coolant

Highly enriched U
 Relatively high levels of 238Pu
Commercial reactors

Light water reactors
3-39
Typical LWR Fuel Rod
Number of Fuel Assemblies in Core*
121 - 257
Assembly Lattice
14x14 – 17x17
Distance Between Fuel Rods
in an Assembly (cm)
3.1 mm
Fuel Rod Diameter (cm)
0.94
U-235 enrichment (%)
1.9 – 4.95
Core Total Fuel Weight (metric tons)
89 - 121
Burnup (MWd/MTIHM)
33,000 – 60,000
•
•
•
•
•
•
Materials Selection Criteria – LWR
Fuels Components
Compatibility with water coolant

High temperature (>300oC in
PWRs) operation

Resistant to corrosion
Minimal neutron absorption
Compatibility with burnable
poisons

In water and in fuel
High strength
Resistance to hydrogen3-40
embrittlement
Pressurized Water Reactor
Height of Vessel (m)
11.5-13.5
Wall Thickness (mm)
180-255
Inside Diameter (m)
3.4-5.2
Weight (metric tons)
240-590
Max. Pressure (psi)
2,500
Max. Temperature (C)
343
3-41
PWR Core Dimensions
MWe
600
900
1200
1500
Power (MWth)
1650
2652
3411
4451
Core
Equivalent
Diameter (m)
2.5
3.0
3.4
3.9
Core Active
Height (m)
3.7
3.7
3.7
3.7
Total Number
of Fuel
Assemblies
121
157
193
257
Hoover Dam: 2,080 MWe
- Typically 264 fuel rods per assembly
- Nominally 500 kg uranium
- 3.7 m fueled length
- Typical linear heat generation rate: 18 kW/m
- Fuel pellet diameter: 0.8 cm
- 121-257 fuel assemblies per core
Principal material zirconium-based alloy:
Zircaloy-4: Zr - 1.5 Sn – 0.2 Fe – 0.1 Cr
ZIRLO: Zr – 1 Nb – 1 Sn – 0.1 Fe
3-42
Boiling Water Reactor
Steam
Dryer
Steam
Separators
Fuel Assemblies
35 operating BWRs in the U.S
Control Rod
Drives
•BWRs typically operate at a pressure of about 1,100 psi, with water boiling at 285°C
•Cruciform control rods are located in the center of each four-assembly cluster of
fuel assemblies
•Reactors are normally operated on a three-year cycle, with about one-third of the
core fuel discharged annually
•Total core loading of a 1,000 MWe BWR is about 100 metric tons of heavy metal
•Reactivity control is usually with burnable poisons added to the ceramic fuel (e.g.,
3-43
Gd2O3)
•Reactor power can also be controlled by changing the flow of water through core
BWR Fuel Assembly
• Fuel pellets stacked in cladding tube
made of Zircaloy-2
• 3.66 m active fuel length
• Fuel rods assembled in square lattice
(8x8, 9x9)
• Rods housed in channel box (“duct”)
• Water flows up along fuel
• Zircaloy-2 alloy:
Zr-1.5Sn-0.15Fe-0.1Cr-0.06Ni
FUEL
RODS
DUCT
3-44
CANDU Reactor Characteristics
•
•
•
•
•
•
•
CANada Deuterium Uranium reactor
Cooled and moderated with heavy water; bulk of the moderator is
contained in a large vessel called a calandria
Outlet temperature 300oC
Linear heat rate 24 kW/m, Core height 6m, core diameter 6m
Typical core loading 6,240 assemblies
Fuel throughput 121 metric tons per year for 935MWe plant (>>
comparable LWR discharge)
Fuel: natural uranium oxide, Pellet length 15.3 mm, diameter 12.1 mm
3-45
CANDU Reactor: On-Line Refueling
3-46
CANDU Fuel
The fuel assemblies used in the reactor are ~ 1.5 feet (0.5 m) CANDU reactors can be refueled on-line.
long, consisting of individual rods. The cladding is Zircaloy This photo shows the refueling machine.
New fuel assemblies are added
and the fuel pellets are uranium dioxide.
horizontally and the spent fuel assemblies
are pushed out to the spent fuel storage
area.
37-pin bundle
3-47
Liquid Metal-cooled Fast Reactor Designs
3-48
LMR Fuel Pin (example shown is reference fuel
for the FFTF reactor)
3-49
Materials Selection Criteria – Liquid Metal
Cooled Fast Reactor Fuels Components
• Compatibility with sodium (or lead) coolant

High temperature operation

Resistant to corrosion
• Minimal neutron absorption
• Compatibility with fuel and fission products
• Resistant to swelling and irradiation creep
• High strength at operating temperatures (up to 650oC)
• Resistance to low-temperature (i.e., refueling temperature
or scram) embrittlement
3-50
Commercial Fast Reactor Fuel Assembly
3-51
Sodium-Cooled Fast Reactor Comparison
Reactor
BN600
SuperPhenix
FFTF
MONJU
Power, MWt
1470
3000
400
714
Generation, MWe
600
1200
-
280
Active core height, m
1.03
1.0
0.91
0.93
Fuel
UO2
UO2-PuO2
UO2-PuO2
UO2-PuO2
205+164
364+233
82
119+81
127
271
217
217
“316”SS
316SS
316SS
316SS
0.69
0.85
0.58
0.65
LHGR, kW/m (max)
48
46
48
36
Inlet temperature, C
377
395
395
397
Outlet temperature, C
550
545
550
529
Fuel assemblies
Pins per assembly
Cladding type (ref.)
Cladding OD, cm
3-52
Thermal Efficiency of a Reactor System
• Carnot Cycle

η = thermal efficiency ≈ 1 – (TL/TH)
• If TL = 395C and TH = 545C, η = 0.28 (SuperPhenix)
• Thermal efficiency becomes the Holy Grail for reactor
designers and utility operators



10% increase in thermal efficiency for a 1,000 MWe plant
amounts to increased revenues of ~$100 million per year
To achieve this, designers push for higher outlet
temperatures
Must then either develop new materials or accept shorter
operating cycles to shut down before end-of-life failures
occur
 ~ $1M per day in lost revenues for a refueling outage plus
purchase of replacement power
 Fuel cycle change: lower enrichment, but lower burnup
3-53
Gas-cooled Fast Reactor
(490oC)
(850oC)
3-54
Gas-Cooled Fast Reactor
•
•
•
•
•
Fast spectrum reactor, no moderator allowed
Helium-cooled
Outlet temperature ~850oC
Direct Brayton cycle gas turbine, high thermal efficiency
Possible fuel forms:

Composite ceramic

Dispersed-particle

Ceramic-clad
•
Possible core configurations based on:

Pin-based fuel assemblies

Plate-based fuel assemblies
3-55
Gas-Cooled Fast Reactor Dispersion Fuel
• Fuel Particle: (U,Pu)C, with two size distributions


1.64 mm diameter
480 μm diameter
• Fuel kernels coated with two SiC layers


Inner coating: porous SiC with low crush strength, 58μm
thick
Outer coating: dense CVD SiC, 61 μm thick
• Matrix: dense SiC
• Heavy metal density: 6 g HM/cm3
3-56
Plate Fuel Concept for GCFR Fuel
•
•
•
•
UC particles in a SiC
matrix
Moderate level of
enrichment to foster Pu
breeding
SiC duct
Challenges:

Maintaining
structural integrity

Achieving high
fuel volume
fraction with
adequate heat
removal

Uniformity of fuel
particle
distribution

Welding and
joining
3-57
Dispersion Fuel Fabrication
• Uranium carbide microspheres are coated with a thin layer
of SiC using methylsilane
• Coated particles are assembled in a packed bed and the
bed is infiltrated with propane which is then converted to
pyrolytic carbon by pyrolysis
• The fuel matrix is then infiltrated with molten silicon, which
reacts with the carbon to form SiC
• The amount of non-fuel material in the blocks must be
minimized in order to obtain the high fuel volume fraction
required in the reactor

Drives design toward higher enrichment and thus greater
fuel cost
3-58
Thorium Fuel
• Th breeds 233U
 Blanket or mixed fuels
• Side reactions
 Formation of 232U, high gamma, from 231Th
 231Th (25.4 hours) beta decay to 231Pa
 231Pa(n,g)232Pa, s=260 b
 232Pa (1.31 d) beta decay to 232U
• Reactions for 231Th production
 232Th(n,2n)
Need high energy neutrons
 230Th(n,g), 23 b
 Decay of 235U
3-59
Plutonium isotopics
• Pu formed in core or from blanket
 U blanket may have a range of isotopic
composition
Natural and anthropogenic
• Pu formation from neutron capture on 238U
 239Pu becomes a target for higher isotope
production
(n,g) up to 241Pu
Also some (n,2n) Emin =5.7 MeV
* 238Pu from 239Pu
3-60
Plutonium isotopics
•
238Pu
can also be produced through successive
neutron capture on 235U
 235U(n,g)236U
 236U(n,g)237U
 237U beta decay to 237Np
 237Np(n,g)238Np
 238Np beta decay to 238Pu
• Mixture of Pu isotopics from fuel or blanket
can act as a signature
 239Pu dominates at low burnups
3-61
240
Device Pu has 6 % Pu
Plutonium isotopics
• Neutron behavior and Pu isotopics coupled
 Pu isotopics influence by neutron fluence and
energy
 neutron kinetic energy effected by reactor
operating temperature
 Moderator influence due to rate of moderation
 Fuel size influences distance between neutron
generation and moderator
 Fuel composition can influence neutron
spectrum
Depletion of neutron in 235U resonance
region
3-62
Plutonium isotopics
• Fluence influence
 240Pu production
Capture on 239Np produces 240Np, which
decays into 240Pu
Competition between capture and decay
3-63
Plutonium isotopics
• Computation from ORIGEN2 (average
composition)
 http://www-rsicc.ornl.gov/codes/ccc/ccc3/ccc371.html
 d isotope/dt integrated for spectrum of
neutron energies
Relates to reactor type
 Net isotope formation for given conditions
Reactor type library
Decay constants
 Fuel composition input
Solves for time and reactor power
3-64
Plutonium isotopics
• ORIGEN does not consider variation of Pu production
in different fuel locations
 Actinides on edge of fuel experience softer neutron
spectra then center material
 Pu distribution varies with depth
* Isotopics and ratios with fission products
vary
 Location of control elements also an influence
• ORIGEN useful for bulk
• Influence of moderator, enrichment, and burnup
greater than reactor power and operating temperature
 No ORIGEN library for graphite moderated
reactors for weapons Pu production
3-65
Plutonium Isotopics
• Evaluate ratios
with 240Pu
 Mass 240:239
(MS)
 Activity
238/(239+240)
(alpha
spectroscopy)
• Varied reactor
types, 37.5
MW/ton
• Large Pu isotopic
variation
• Obvious variation for
blanket and CANDU
3-66
Plutonium Isotopics
• Comparison to
239Pu concentration
 Can provide
time since
discharge
241Pu is time
sensitive
• Change in measure
and expect ratio
can be used to
determine time
since discharge
3-67
Plutonium Isotopics
• Use for
evaluating
reactor
power
 Mass
242:239
ration
 Activity
239/(239
+240)
3-68
Transplutonium Isotopics
• Am (242m, 243) and Cm (242, 244) sensitive to reactor
power
 Due to relative decay to 241Pu and formation of
241Am
 More 241Am is available for reactions from
longer times with low flux
* Weapons grade Pu from 10 day to 2 year
irradiation
• From neutron reaction and beta decay of heavier Pu
• 243Pu beta decay to 243Am
 Capture and decay to 244Cm
• Neutron capture on 241Am
 242Am states
 Ground state decay to 242Cm
• Strong flux dependence on ratios
3-69
6% 240Pu
•
244Cm
241Pu
•
242Cm

and 243Am arise from multiple neutron capture on
(t1/2=163 d) can determine time since discharge 3-70
Change from expected value compared to other ratios
Reprocessing
•
•
•
•
Used nuclear fuel
238Pu production
233U production
In fuel, fission products build up fast than Pu isotopes
3-71
Reprocessing
• 510 gU/g Pu for used fuel
• Majority of U remains, enrichment level to 0.62%
 Depends upon the level of burnup
• Reprocessing limitations
 Remote handling
 Criticality issues
 Limit impurities
• Range of reprocessing techniques
 Precipitation
 Molten salt
 Ion exchange
 Fluoride volatility
 Solvent extraction
3-72
Reprocessing
• BUTEX
 Dibutyl carbitol solvent
Relatively poor separation from Ru
* 1E3 rather than 1E6
1 g Pu has 1E7 dpm 106Ru
• HEXONE
 Methylisobutylketone solvent
 Decontamination factors
1E4 Ru
1E5 for Zr, Nb, and Ce
* 1 g Pu has 1000 dpm 93Zr
3-73
Reprocessing
•
PUREX

Zr, Tc, and Ru observed

Ln, Np, and Th nonmegligible

Nd is a key isotope, nature levels in reprocessing materials
 Natural and reactor Nd radios are plant signatures
3-74
35
• Isotopes in Pu from incomplete separation and
decay
3-75
Reprocessing
• Organic analysis
 Radiation effects
Polymerization
TBP degradation
* Di, mono, and phosphoric acid
N-butonal and nitrobutane
• Changes extraction behavior, limits extraction efficiency
 Formation of carboxylic acid, esters, ketones,
nitroorganic compounds
• Signatures from processing
 Range of sample states, including gas
 GC-MS methods
3-76
3-77
Metals and Alloys
• Range of behavior
and phases for
actinide metals
 Due to felectron
behavior
• Similarities for Th,
U, and Pu
 Oxide coating
 Need of
complexing
agent for
dissolution in
nitric acid
• U and Pu pyrophoric
• Th limited solubility
in liquid U
3-78
Metals and alloys
• Formation of metals from reaction of metal halides
with reducing elements
 Na, Ca, Mg
Group 1 and 2 halide formation favored up
to 1500 ºC
Na has low boiling point relative to metal
melting point
Ca reaction extremely exothermic, melts
reaction products
MgF2 and CaF2 not soluble in molten U
* Mg needs to limit formation of U oxides
due to solubility
• Reaction vessel must be coated to be inert
 CaF2, MgF2, BeO, ThO2, graphite (Mg only) 3-79
Metals and alloys
• Can also be formed from molten salt processes
 UF3 of UF4 in KCl/NaCl
Reduction of U to metallic state
 Grain size determined by temperature and
current density
• Hydride/dehydriding results in high surface area U
 Too pyrophoric for use
• U metal purification by melting in vacuum under
Ar
 Graphite crucibles lined with CaO, MgO, or
ZrO2
3-80
Metals and alloys
• Th metal from Ca reaction
 Higher Th melting point make reaction less
energetic
 May require initiator
I2, S, ZnCl2
• For Th oxide with Ca, reaction in Ni boat
 Dissolve Ca product in water
Need to consider Th metal dissolution
kinetics
• Th metal casting in BeO
 alumina, MgO, or CaO dissolve in Th metal
• High purity Th from formation of volatile ThI4
 Condense on cool surface
3-81
Metals and alloy
• For Pu reaction with oxides rather than
fluorides limits neutron formation
 19F(a,n)
 CaO will coat Pu product, need to dissolve
dilute nitric acid
• Hydrogenation reaction
 Pu and PuH2+x
 Free from oxygen
3-82
Preparation of Pu metal
• Ca reduction
• Pyroprocessing

PuF4 and Ca metal
 Conversion of oxide to fluoride
 Start at 600 ºC goes to 2000 ºC
 Pu solidifies at bottom of crucible

Direct oxide reduction
 Direct reduction of oxide with Ca metal
 PuO2, Ca, and CaCl2

Molten salt extraction
 Separation of Pu from Am and lanthanides
 Oxidize Am to Am3+, remains in salt phase
 MgCl2 as oxidizing agent
* Oxidation of Pu and Am, formation of Mg
* Reduction of Pu by oxidation of Am metal
3-83
Pu metal
• Electrorefining
 Liquid Pu oxidizes from anode ingot into
 molten salt electrode
 740 ºC in NaCl/KCl with MgCl2 as oxidizing
agent
Oxidation to Pu(III)
Addition of current causes reduction of
Pu(III) at cathode
Pu drips off cathode
3-84
Pu metal
• Zone refining (700-1000 ºC)

Purification from trace impurities
 Fe, U, Mg, Ca, Ni, Al, K, Si, oxides and hydrides

Melt zone passes through Pu metal at a slow rate
 Impurities travel in same or opposite direction of melt
direction

Vacuum distillation removes Am

Application of magnetic field levitates Pu
3-85
http://arq.lanl.gov/source/orgs/nmt/nmtdo/AQarchive/98fall/magnetic_levitation.html
Metallic Uranium phases
 a-phase

Room temperature to 942 K

Orthorhombic

U-U distance 2.80 Å

Unique structure type
 b-phase

Exists between 668 and 775 ºC

Tetragonal unit cell
a‐phase U-U distances in layer
 g-phase
(2.80±0.05) Å and between layers

Formed above 775 ºC
3.26 Å

bcc structure
• Metal has plastic character

Gamma phase soft, difficult fabrication

Beta phase brittle and hard
• Paramagnetic
• Temperature dependence of resistivity
• Impurities impact phases

2.3 % Mo in U eliminates beta phase
• Alloyed with Mo, Nb, Nb-Zr, and Ti
b-phase
3-86
Th metal properties
• silvery-white metal which is air-stable

Oxide slowly forms, to gray and finally black.
• Changes structure with temperature

ffc to bcc at 1360 ºC
 High pressure forms body centered tetragonal
• Metal is paramagnetic (2 d electrons)
3-87
Metallic Pu
•
•
•
•
•
Interests in processing-structure-properties relationship
Reactions with water and oxygen
Impact of self-irradiation
Relatively poor heat and electricity conductor
Mixes with Np in alpha phase, Th and U in beta and gamma phase
−3
Density
Liquid density at m.p.
Melting point
19.816 g·cm
−3
16.63 g·cm
912.5 K
Boiling point
3505 K
Heat of fusion
Heat of vaporization
Heat capacity
2.82 kJ·mol
−1
333.5 kJ·mol
−1
−1
(25 °C) 35.5 J·mol ·K
−1
3-88
3-89
Metallic Pu
• Pu liquid is denser
that 3 highest
temperature solid
phases
 Liquid density at
16.65 g/mL
 Pu contracts 2.5
% upon melting
• Pu alloys and the d
phase
 Ga stabilizes
phase
 Complicated
phase diagram
3-90
Small chemical additions can
stabilize the high-volume phases
of plutonium. The Pu-Ga phase
diagram shows how gallium
additions of a few atomic
percent form a solid solution
(gallium atoms are incorporated
into the plutonium fcc δ-phase)
that is retained to room
temperature.
The rest of the diagram shows
the enormous
complexity and richness of
alloying behavior. The Pu-Ga
system exhibits 11 different
intermetallic compounds and
several new phases that are
different from those of elemental
plutonium or gallium (Peterson
and Kassner 1988).
3-91
Phase
never
observed,
slow
kinetics
3-92
3-93
3-94
Metallic Pu
•
•
•
•
•
Other elements that stabilize d phase

Al, Ga, Ce, Am, Sc, In, and Tl stabilize
phase at room temperature

Si, Zn, Zr, and Hf retain phase under
rapid cooling
Microstructure of d phase due to Ga
diffusion in cooling
Np expands the a and b phase region

b phase stabilized at room
temperature with Hf, Ti, and Zr
Pu eutectics

Pu melting point dramatically reduced
by Mn Fe, Co, or Ni
 With Fe, mp=410 °C, 10 % Fe
 Used in metallic fuel

Limit Pu usage (melting through
cladding
Interstitial compounds

Large difference in ionic radii (59 %)

O, C, N, and H form interstitial
compounds
3-95
Metallic Pu
•
•
Electronic structure shows
competition between itinerant and
localized behavior

Boundary between magnetic
and superconductivity

5f electrons 2 to 4 eV bands,
strong mixing
 Polymorphism
 Solid state instability
 Catalytic activity
Isolated Pu 7s25f6, metallic Pu
7s26d15f5

Lighter than Pu, addition f
electron goes into conducting
band

Starting at Am f electrons
become localized
 Increase in atomic
volume
3-96
Review
• Range of signature from
Th, U, and Pu metal

Ore deposits
• Engineering signatures

Separation

Enrichment
• Signatures from reactors

Flux and neutron
energy

Fuel composition

Reactor type
• Separation signatures

Chemical impurities
• Isotopic ratios

Pu

Transplutonium

Sm
3-97
Questions
• What are the different types of conditions used for separation of U
from ore
• What is the physical basis for enriching U by gas and laser
methods?
• What chemistry is exploited for solution based U enrichment
• Describe signatures from the enrichment process for different
methods.
• Describe 4 different types of reactors
• What signatures are available from

Pu isotopic ratios

Transplution isotopic ratios

Sm isotopic ratios
• Describe the basic chemistry for the production of Th, U, and Pu
metal
• Why is Pu alloyed with Ga?
3-98
Pop Quiz
• What information can be gained from isotopic
ratio signatures that include 241Pu and 242Cm?
3-99