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CHEM 312: Lecture 16 Radiochemistry in reactors Part 1

• Readings: Radiochemistry in Light Water Reactors, Chapter 3 (link on lecture webpage)

Outline • Overview of reactors

Basic concepts Reactor types

Light water Heavy Water Liquid Metal Cooled Gas Cooled Molten Salt Space reactor

• Speciation in irradiated fuel • Utilization of resulting isotopics • Fission Product Chemistry

Cherenkov radiation

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Nuclear Reactor • Basic reactor components

Fuel Oxide, metal, nitride, salt U and Pu isotopes as fissile

component Cooling

Air, water, CO2 Cladding

Separate fuel from coolant Moderator

Protons Carbon

* Carbon needs more collision (114 versus 18) than proton

* smaller reaction cross section Heavy water (D2O)

* Deuterium small neutron absorbance cross section than proton

• Demonstrate by CP-1 Graphite purified from B as moderator Cd control Moisture control 2 kW, air cooled

• Power reactor uses heat from fission to generate electricity

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Reactor material characteristics

• Compatibility with water coolant High temperature, >300 oC in PWRs

operation • Resistant to corrosion • Minimal neutron absorption • Compatibility with burnable poisons In water and in fuel

• High strength • Resistance to hydrogen embrittlement

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Reactor fuel

• Fuel confined in reactor to specific region Potential for interaction with

cladding material Initiate stress corrosion

cracking Chemical knowledge useful in

events where fuel is outside of cladding

Molten salt reactors differ • Some radionuclides generated in

structural material Neutron activation of material

Activation products (i.e., 60Co)

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Pressurized Water Reactor

69 operating PWRs in the U.S.

Height of Vessel (m)

Wall Thickness (mm)

Inside Diameter (m)

Weight (metric tons)

Max. Pressure (psi)

Max. Temperature (C)

11.5-13.5

180-255

3.4-5.2

240-590

2,500

343

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PWR Fuel Assembly

- 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 of construction is a zirconium-based alloy: Zircaloy-4: Zr - 1.5 Sn – 0.2 Fe – 0.1 Cr ZIRLO: Zr – 1 Nb – 1 Sn – 0.1 Fe

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

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Boiling Water Reactor

35 operating BWRs in the US

• BWRs typically operate at a pressure of about 1,100 psi, with water boiling in the core at about 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., Gd2O3)

• Reactor power can also be controlled by changing the flow of water through the core

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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

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CANDU Reactor Characteristics • CANada Deuterium Uranium reactor • Cooled and moderated with heavy water

Exploits differences in neutron absorption of hydrogen isotopes 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 • Online refueling

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CANDU Fuel

The fuel assemblies used in the reactor are ~ 1.5 feet (0.5 m) long, consisting of individual rods. The cladding is Zircaloy and the fuel

pellets are uranium dioxide. CANDU reactors can be refueled on-line.

This photo shows the refueling machine.

New fuel assemblies are added horizontally

and the spent fuel assemblies are pushed out to the spent fuel

storage area. The fuel design is shown below.

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Liquid Metal-cooled

Fast Reactor Designs

• 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 • Fuels

Oxides, nitrides, carbides, metal

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Synthesis of U-Zr Alloys

• alloys increase melting point

• Alloys have increased durability in reactor environment

Figure 1: Calculated phase diagram for U-Zr alloy

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Metallic Fuel Behavior • As opposed to oxide fuel, which is hard and brittle, metallic fuel is highly

plastic during irradiation • Metal fuel swells at low burnup due to the accumulation of fission

products Swelling is both axial and radial

• At ~2 at.% burnup, the fuel contacts the inner cladding wall Mechanical force is not great, but fuel-cladding chemical

interaction can occur • At slightly higher burnup, fission gas bubbles become interconnected and

there is a pathway for release of fission gases into the fuel pin plenum Hoop stress imposed on cladding: σ = pd/2t, where p is the fission

gas pressure, d is the inner diameter of the cladding, and t is the cladding thickness

Bond sodium enters the interconnected porosity (an issue only when it comes to reprocessing)

• Life-limiting factors: Cladding and duct distortion (creating hot spots in the pin bundle) Cladding strain due to pressurization by fission gases Fuel-cladding chemical interaction

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Swelling/Fission Gas Release

• Fission Products Two atoms replace every U (or Pu) atom that

fissions 25% of fission products are gas atoms (Kr, Xe)

• Fuel Swelling generation of fission products Gas atoms coalesce into bubbles, accelerating

swelling Fuel swelling tends to reduce or close gap

• Fission Gas Release Some fission gas escapes fuel Pressurizes plenum Percent of gas escaping fuel < 10% in LWR fuel > 50% in fast reactor fuel Bubbles in metallic fuel

Metallic Fuel at 0.9% Burnup Rapid release of fission gases and inter-linkage of porosity; large fuel volume change

U-20Pu-10Zr fuel

Metallic Fuel at 17% Burnup

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Gas-cooled Fast Reactor

This image cannot currently be displayed.

78 ft (24 m)

Generator

Compressor

Turbine Recuperator

Intercooler

Precooler

Annular Core

• 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

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All MHR concepts use TRISO-Coated Particle Fuel

PARTICLES COMPACTS FUEL ELEMENTS

TRISO: tri-isotropic coating

TRISO-coated fuel particles (left) are formed into fuel rods (center) and inserted into graphite fuel elements (right).

Courtesy of General Atomics

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Molten salt reactor • Fuel in liquid form • Fuels melt between 450 °C and 550

°C UF4 based fuel LiF, BeF2 Can also contain ThF4

Uranium chloride fuel UCl3-UCl4-NaCl Faster neutrons

• Use of molten salt coolant to transfer heat

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Molten salt reactor • Range of variables Online chemical

treatment Breeding of 233U

from 232Th 232Th(n,γ)233Th 233Th beta decay

to 233Pa, then beta decay to 233U

• Confinement of fuel • Variations in salt

compostion

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Thermal Propulsion Nuclear Engine Overview

1. Hydrogen is introduced into the reactor from a storage Dewar.

2. Hydrogen gas is heated and accelerated by fission heat.

3. The nozzle directs the kinetic energy.

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Fuel Pellet Points to Meet

• Stable and will not melt at core temperature

• Fully enriched uranium as uranium carbide

• Compatible with coolant and hot frit materials

Project chose IK (infiltrated kernel) fuel pellets

• Stable and will not melt at core

temperature • Fully enriched uranium as uranium

carbide • Compatible with coolant and hot frit

walls • Made of graphite with UC2

dissolved into the graphite matrix • Require a coating to retain fission

products NbC, TaC/PC/NbC

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Space Reactor Materials

• mass already in earth orbit • accelerate a light mass to high

velocities using nuclear fission reactions

• high exhaust velocities make for more effective use of a light mass propellant Coolant enters at 40K and

leaves at about 3000K

• Choices of Hot Frit Material Rhenium Boron Nitride Coated Carbon/Carbon

TaC, NbC, PC, and ZrC

• Choices of Cold Frit Material Stainless Steel Aluminum Beryllium

• Coolant is about 18% of total moderation

• Other choices are Al/Polyethylene/H,

Be/Polyethylene/H, Be/7Li/H

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CHEM 312: Lecture 16 Radiochemistry in reactors Part 1

• Readings: Radiochemistry in Light Water Reactors, Chapter 3 (link on lecture webpage)

Outline • Overview of reactors

Basic concepts Reactor types

Light water Heavy Water Liquid Metal Cooled Gas Cooled Molten Salt Space reactor

• Speciation in irradiated fuel • Utilization of resulting isotopics • Fission Product Chemistry

Cherenkov radiation

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CHEM 312: Lecture 16 Radiochemistry in reactors Part 2

• Readings: Radiochemistry in Light Water Reactors, Chapter 3 (link on lecture webpage)

Outline • Overview of reactors

Basic concepts Reactor types

Light water Heavy Water Liquid Metal Cooled Gas Cooled Molten Salt Space reactor

• Speciation in irradiated fuel • Utilization of resulting isotopics • Fission Product Chemistry

Cherenkov radiation

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Fission process • Fission drives process for

introducing elements into fuel • Recoil length about 10 microns,

diameter of 6 nm About size of UO2 crystal 95 % of energy into stopping

power Remainder into lattice

defects * Radiation induced

creep * Swelling

High local temperature from fission 3300 K in 10 nm

diameter • Delayed neutron fission products

0.75 % of total neutrons 137-139I and 87-90Br as

examples • Some neutron capture of fission

products influences effective decay

constant

σφλλ +=eff

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Burnup • Measure of extracted energy from fuel

Fraction of fuel atoms that underwent fission %FIMA (fissions per initial metal atom)

Actual energy released per mass of initial fuel Gigawatt-days/metric ton heavy metal (GWd/MTHM) Megawatt-days/kg heavy metal (MWd/kgHM) 1 MeV=4.45E-23 MW h

• Burnup relationship Plant thermal power times days dividing by the mass of the initial fuel loading Converting between percent and energy/mass by using energy released per fission event

typical value is 200 MeV/fission 100 % burnup around 1000 GWd/MTHM

• Determine burnup Find residual concentrations of fissile nuclides after irradiation

Burnup from difference between final and initial values Need to account for neutron capture on fissile nuclides

Find fission product concentration in fuel Need suitable half-life Need knowledge of nuclear data

* cumulative fission yield, neutron capture cross section Simple analytical procedure 137Cs (some migration issues) 142Nd(stable isotope), 152Eu are suitable fission products

Neutron detection also used Need to minimize 244Cm due to spontaneous fission of isotope

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Radionuclides in fuel

• Actual Pu isotopics in MOX fuel may vary Activity dominated

by other Pu isotopes Ingrowth of 241Am MOX fuel fabrication

in glove boxes

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Fuel variation during irradiation

• Chemical composition of fuel Higher concentrations

of Ru, Rh, and Pd in Pu fuel

• Radionuclide inventory • Pellet structure • Total activity of fuel

effected by saturation Tends to reach

maximum • Radionuclide fuel

distribution studied Fission gas release Axial distribution by

gamma scanning • Radial distribution to

evaluate flux

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Distribution in fuel • Axial fission product

distribution corresponds very closely to the time-averaged neutron flux distribution PWR activity level in the

middle Activity minima from

neutron shielding effect of spacer grids local decrease in

fission rates Fuel density effects

Dishing at end of fuel Disappear due to fuel

swelling BWR shows asymmetric

distribution Control rod positions

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Distribution in Fuel • Radial distribution of

fission products mainly governed by thermal neutron flux profile

• Higher Pu concentration in outer zone of fuel Epithermal neutron

capture on 238U Small influence of thermal

migration on Cs Gaseous and volatile fission

products Influence by fuel initial

composition (O to M ratio)

Transuranics on fuel rim

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Distribution in Fuel

• Increased Pu concentration on rim leads to increased fission product density Xe behavior

influenced by bubble gas location

• Consumption of burnable poison Gd isotopes 157

and 155 depleted in outer zone

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Distribution in fuel: Thermal behavior • Mainly affects gaseous and volatile fission products linear heat rating pellet temperatures during reactor operation stoichiometry of fuel

• Halogens and alkali elements Cs and I volatility

High fission yields Enhanced mobility

Can be treated similarly different chemical behavior limited in fuel

behavior

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Iodine and Cs • CsI added to UO2

Both elements have same maximum location at 1000 °C

Behavior as CsI • UO2+x

Iodine property changes, mobility to lower temperature regions Elemental I2

rather than I-

• Formation in range of x to 0.02

• No change in Cs chemistry remains monovalent

• release of cesium and iodine from fuel at 1100 to 1300 K Increases with

temperature

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Iodine and Cs • Cs and iodine release

rates increase with increasing temperature 2100 K largest

fraction released after 60 seconds

• Both elements released at significantly faster rate from higher-burnup fuel Different release

mechanism • Attributed to fission

product atoms which already migrated to grain boundaries UO2 lattice

difficulty in incorporating large atomic radii ions

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Perovskite phase (A2+B4+O3) • Most fission products homogeneously

distributed in UO2 matrix Solid solution formation or

relatively low concentration of fission products

• With increasing fission product concentration formation of secondary phases possible Exceed solubility limits in UO2

• Perovskite identified oxide phase B site: U, Pu, Zr, Mo, and

Lanthanides Mono- and divalent elements at A

Ba, Sr, Cs • Mechanism of formation Sr and Zr form phases initially Lanthanides added at high

burnup

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Epsilon phase • Metallic phase of fission products in fuel Mo (24-43 wt %) Tc (8-16 wt %) Ru (27-52 wt %) Rh (4-10 wt %) Pd (4-10 wt %)

Metal above tend to not forms oxides

• Grain sizes around 1 micron

• Concentration nearly linear with fuel burnup 5 g/kg at 10 MWd/kg

U 15 g/kg at 40 MWd/kg

U

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Epsilon Phase

• Formation of metallic phase promoted by higher linear heat high Pd concentrations

(20 wt %) indicate a relatively low fuel temperature

Mo behavior controlled by oxygen potential High metallic Mo

indicates O:M of 2 O:M above 2, more

Mo in UO2 lattice

Relative partial molar Gibbs free energy of oxygen of fission product oxides and UO2

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Grouping fission product and actinide behavior • Experiments performed between 1450 °C and

1825 °C trace-irradiated UO2 fuel material Limit formation of fission products

compounds • 4 categories • Elements with highest electronegativities have

highest mobilities Te, I

• Low valent cations and low fuel solubility Cs, Ba

• Neutral species with low solubility Xe, Ru, Tc Similar behavior to low valent cations (xenon, ruthenium,

• polyvalent elements were not released from fuel Nd, La, Zr, Np

• Ions with high charges remain in UO2 • Neutral atoms or monovalent fission products

are mobile Evident at higher temperatures

higher fuel rod heat ratings accident conditions

Orange: volatile fission products Grey: metallic precipitates Blue: oxide precipitates Green: solid solution

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Review

• How is uranium chemistry linked with chemistry in fuel

• What are the main oxidation states of the fission products and actinides in fuel

• What drives the speciation of actinides and fission products in fuel

• How is volatility linked with fission product chemistry

• What are general trends in fission product chemistry

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Questions • What drives the

speciation of actinides and fission products in spent nuclear fuel?

• What would be the difference between oxide and metallic fuel for fission product speciation?

• Why do the metallic phases form in oxide fuel

• How is the behavior of Tc in fuel related to the U:O stoichiometry?

Amount of fission Fuel type and composition

Metal fuel is elemental Only interactions with fuel or fission product

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Question

• Comment on blog • Respond to PDF Quiz 16