Nuclear Fuels and Materials - 151-2017-00L · 2016-03-13 · Nuclear Fuels and Materials - 151-2017-00L ⌸Lecture 4 - Page 13/86 • Purpose: Enclosure of fuel pellets • Mechanical

WIR SCHAFFEN WISSEN – HEUTE FÜR MORGEN

Nuclear Fuels and Materials - 151-2017-00L

Manuel A. Pouchon :: Head of LNM :: Paul Scherrer Institut

Master of Nuclear Engineering ‐ Spring Semester 2016

Lecture 4: Cladding, ….

Nuclear Fuels and Materials - 151-2017-00L ⌸ Lecture 4 - Page 2/86

o INTRODUCTION TO CLADDING MATERIALS REACTOR / ASSEMBLY / FUEL PIN / CLADDING / PELLET – ADVANCED SYST. REQUIREMENTS

MATERIAL

• ZIRCONIUM / OTHERS

HISTORY OF ZIRCALOY

ZIRCONIUM PROPERTIES

PRODUCTION OF ZIRCONIUM AND CLADDING

TEXTURE, ALLOYING ELEMENTS, ZIRCALOY, …

o POTENTIAL FAILURE MECHANISMS FOR CLADDING MATERIALS PCMI AND PCI/SCC FRETTING

HYDRIDES

IRRADIATION EMBRITTLEMENT

GROWTH AND CREEP

TOC


Pellet

FuelCladding

Fuel tube

Fuel rod Assembly

Introduction – From the pellet to the assembly


PWR(pressurized water reactor)

BWR(boiling water reactor)

Introduction – Thermal reactors (selection)


Rods• Zircaloy tube filled with U(Pu)O2 tablets and helium • cylindrical rods put into bundles

PWR• 179-264 fuel rods per fuel bundle (14×14 to 17×17)• 121 to 193 fuel bundles• bundles are about 4 meters long

BWR• Similar to above, but …• canned• 91, 92, or 96 fuel rods per assembly• 368-800 assemblies

Introduction – Thermal reactors – assembly II


BWR

PWR

Introduction – Thermal reactors – assembly


BWR PWR

Cladding Thickness (mm) 0.813 0.57-0.65

Cladding Zircaloy 2 Zircaloy-4

Advanced Cladding Barrier Zircaloy 2 ZIRLO, M5, low-Sn Zry-4

Grid (#fuel locations) 8 x 8 square (62) 17 x 17 (264) or 16 x 16 (236)

Pitch (mm) 16.2 12.6

Cladding outer diameter (mm) 12.7 9.5-9.7

Height (m) 4.1 4.0-4.1

Active Fuel height (m) 3.81 3.6-3.8

Fuel enrichment (wt. %) 2.8 3.3 (now almost 5%)

Pellet diameter (mm) 10.4 8.2

Pellet Height (mm) 10.4 9.5-13.0

Inlet Temperature (ºC) 278 292-301

Outlet Temperature (ºC) 288 325-332

Linear Power (kW/m) average 19 16-17.8

Linear Power (kW/m) max 44 41-42.7

Number of assemblies 748 193-241

Reactivity Control Rods Cruciform Rod cluster

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Fuel Assembly Characteristics for BWR and PWR


Two CANDU fuel bundles: Each about 50 cm in length and 10 cm in diameter, and generating about 1 GWh of electricity during its time in the reactor. Photo courtesy of Atomic Energy of Canada Limited.

CANDU

Introduction – Thermal Reactors –Special systems


(V)HTR

Cladding• porous carbon buffer layer • pyrolytic carbon (PyC), • SiC • PyC

Introduction – Thermal Reactors –Special systems


Liquid metal (sodium) cooled fast breeder reactor (LMFBR / SFR)

Introduction – Fast reactors (selection)


Joyo – Oarai / Japan

Introduction – Fast reactors – SFR Assembly


GFR

http://www.sckcen.be/en/Media/Images/Our-Research/ANS/IrrDemo-Fuel

Cladding: SiCf/SiC

MT Aerospace AG, Augsburg

Introduction – Fast reactors – Special Systems


• Purpose: Enclosure of fuel pellets

• Mechanical stability(under temperature, pressure and irradiation)

• Chemical stability in the special environment(water, pressure, temperature, radiation)

• Small capture cross section for neutrons (especially thermal neutrons)

• Available in large quantities, comparable cheap(BWR80t, PWR 40t)

Requirements



MATERIAL


HISTORY OF ZIRCALOY





HYDRIDES


GROWTH AND CREEP

TOC


• Zircaloy: Zirconium based alloys (thermal reactors)

• Steel / ODS Steel (fast reactors, thermal reactors)

• Aluminum and beryllium alloys(low power pool type reactors)

• Magnesium with small amount of aluminum and other metals (Magnox: Magnesium non-oxidizing)

• SiC: Fiber-reinforced:(Gas cooled fast reactor, today also in discussion for LWR)

Materials for Cladding

westinghouse


• First test reactors (mild environment, pool-type, i.e. water < 100°C ): Aluminum and beryllium alloys were used, due to their low thermal neutron capture cross section.

• First nuclear power reactors (submarine propulsion): higher efficiency. Zirconium as possible candidate (very low thermal neutron capture cross section): poor ductility and corrosion resistance, Hf to be separated. Therefore stainless steel

• Strong development after WW II fro nuclear submarine program: Improved Hf separation process, industrial production procedures. Tests with binary and ternary alloys

History of Zircaloy - 1


Base Metal UTS (MPa) Macroscopic neutron capture Cross Section (cm¯¹)

Relative neutron absorption for a given design stress

Zirconium 900 0.01 1

Iron 1100 0.17 4

Nickel 1100 0.31 25

Titanium 1000 0.26 28

Aluminum 90 0.014 14

Magnesium 90 0.005 5

Beryllium 180-350 0.001 0.25-5

Properties of metals


• Tests with binary and ternary alloys:

• Zircaloy-1: Zr - 2.5%Sn:increased corrosion resistance, mitigated nitrogen content

• Zircaloy-2: Zr - 1.45%Sn, 0.1%Cr, 0.135%Fe, 0.055%Ni, 0.01%Hf [*]:accidental contamination of Zircalloy-1 with stainless steel lead to an alloy with good corrosion resistance

• Zircaloy-3: was a low tin variant of Zircaloy-2

• Zircaloy-4: Zr - 1.45%Sn, 0.1%Cr, 0.21%Fe, 0.01%Hf [*] :a Ni-free variant fo Zircaloy-2

• Former Soviet Union & Canada:Zr - Nb binary system:Today:

Zr-2.5 in CANDU (HWR) and RBMK Zr-1Nb/(Zirlo)/M5 in PWR

[*] http://iriaxp.iri.tudelft.nl/~leege/SCALE44/stdcmp.pdf

History of Zircaloy - 2


• Earth‘s crust: 0.002 wt% Zr (more than Cu or Zn)

• Zircon (ZrSiO4) is most important source Tetragonal ditetragonal dipyramidal Hyacinth (reddish gem, already known in

antiquity) is a zircon with impurities Fe2O3 / CaO / Hf impurities

• Baddeleyite (ZrO2) is another important source Monoclinic zirconoxide with impurities and

only 90-95% ZrO2

Zirconium Ressources


• Name, symbol, number: zirconium, Zr, 40• Element category: transition metal• Melting point 2128 K / 1855 °C• Crystal lattice: < 862 °C: hexagonal, hpd (α) ─ density: 6.5g·cm−3

> 862 °C: cubic body centered, cbc (β) ─ density: 6.5g·cm−3

• Thermal expansion• α-Zr parallel to c-axis: 6.4·10-6 K-1

• α-Zr perpendicular to c-axis: 5.6·10-6 K-1

• β-Zr parallel to c-axis: 9.7·10-6 K-1c-axisα

Zirconium – Physical Properties


• Forms dense layers of stable oxides with high melting point and well bond to the metal

• In most environments more stable than titanium and stainless steels. Examples

Sulphuric acid: in problematic concentration range of 40-60% one of the best structural materials

Hydrochloric acid: stable at boiling temperature up to an acid concentration of some 37 %

Nitric acid: Outstanding stability fo all concentrations and temperatures

Zirconium – Chemical Properties


• Oxidation behavior : Zr + 2H2O = ZrO2 + 2H2 + 140 kCal/mol

Time to consume a 0 .087’’ (2.2 mm) thick plate as a function of temperature

Bostrom, W.A., THE HIGH TEMPERATURE OXIDATION OF ZIRCALOY IN WATER, (1957) OSTI ID: 4360187

(140 kCal = 585.76 kJ)



• What is the oxidation energy in a• BWR (80t Zircaloy)• PWR (40t Zircaloy) • How much hydrogen is produced?

• What is thermal power of the oxidation reaction of a 1 mm cladding at a temperature of 1450 °C ?

• BWR

Zr: 91.224 g·mol−1

H: 1.00794 g·mol−10.08988 g/L

R = 13.9·P1/6·exp(−1.47/kBT)

solution



Remark

Use as metal: only 5% of the zirconium ores worldwide (rest is used as oxides, e.g. ZrO2 as high temperature ceramic, …..)

• Nuclear application

• Chemical industry (in corrosive environment)

• As alloying element in steels

• Application for medical technique and surgery

• Supra conducting magnets (in combination with niobium)

Zirconium – Use


Hexagonal structure• Less gliding planes than cubic lattice• Distinct formation of

twin boundaries Complex behavior under

plastic deformation Strong anisotropy Strong formation

of structure

Thermo-mechanical fabrication process is essential for the evolution the microstructure, the mechanical properties and the corrosion resistance

Hexagonal lattice..ABABCBCBCBCBABA..(= Stacking fault)

= Twin

Zirconium – Mechanical Properties



MATERIAL


HISTORY OF ZIRCALOY





HYDRIDES


GROWTH AND CREEP

TOC


Chloration:Zircon (ZrSiO4, ZrO2, HfO2) + carbon + chlorine: Chloration (@ 1200 °C)

ZrO2(+ SiO2+HfO2) + 2C + 2Cl → ZrCl4(+SiCl4↑+HfCl4)+2CO

Zr/Hf separation by chloride distillations process (@ 350 °C in vacuum)

Hf-free ZrCl4 is compacted and subjected to the Kroll reduction:

ZrCl4+Mg → Zr, Mg + MgCl2 (@ 850 °C in vacuum)

Excess Mg is separated by vacuum distillation 1000 °C → zirconium sponge

Zirconium/Cladding – Fabrication I


Zirconium sponge:

Compaction to electrode-part:

K.Baur, 2002

Zirconium/Cladding – Fabrication II


Welding of electrode parts to full electrode – mixing with scrap and alloying elements

Electrode of company ABSS

K.Baur, 2002

Zirconium/Cladding – Fabrication III


K.Baur, 2002

consumable electrode

water cooled copper chill

electric arc

ingot

Melting 3 times in a consumable electrode vacuum arc furnace

Ingot of CEZUS

Zirconium/Cladding – Fabrication IV


http://www2.mne.psu.edu/motta/Chapters/Book%20Aug%202011/Chapter17_ZrAlloys.pdf

Zircaloy Tubing Fabrication Process


Forging to a rod (in β and α-region) then temperature treatment for microstructure.

CEZUS

K. Baur, 2002

Zirconium/Cladding – Fabrication V


K. Baur, 2002

Drilling of central hole, then hot extrusion:

Zirconium/Cladding – Fabrication VI


http://www.smst-tubes.com/

Zirconium/Cladding – Fabrication VII


K. Baur, 2002Zircaloy tubes before final

pilgering step

Zirconium/Cladding – Fabrication VIII


The extruded tube is then cold pilgered

K. Baur, 2002

Zirconium/Cladding – Fabrication XIV


http://www.smst-tubes.com/en/products-services/production/cold-pilgering-and-cold-drawing/

Zirconium/Cladding – Fabrication XV



MATERIAL


HISTORY OF ZIRCALOY





HYDRIDES


GROWTH AND CREEP

TOC


Radial orientation of c-axis due to high reduction in diameter (beneficial for later hydride precipitation):

K. Baur, 2002

Zircaloy Cladding – Texture I


Radial orientation of c-axis due to high reduction in diameter (beneficial for later hydride precipitation):

K. Baur, 2002

Zircaloy Cladding – Texture II


• Oxygen (800-1600 ppm): stabilizes the α-phase (solid solution), increases yield strength

• Tin (0.5-1.8 wt%) stabilizes the α-phase (solid solution), mitigates nitrogen impurities, increases slightly yield strength

• Iron, chromium and nickel: stabilizes the β-phase (solid solution in β-phase, precipitates in α-phase), increase corrosion resistance by formation of second phase particles (SPP‘s)

• Niobium (0-2.5 wt%): stabilizes the β-phase precipitates in α-phase), increases yield strength and corrosion resistance

Zircaloy – Alloying elements


wt% Zr Sn Fe Cr Ni Nb O

Zircaloy-2 Bal. 1.2-1.7 0.07-0.2 0.05-0.15 0.03-0.08 - 0.1-0.14

Zircaloy-4 Bal. 1.2-1.7 0.18-0.24 0.07-0.13 0.007 max - 0.1-0.14

ZIRLO Bal. 1.0 0.1 - - 1 ca 0.1

M5 Bal. - 0.015-0.006 - - 0.8-1.2 0.09-0.12

Zr-2.5Nb Bal. 0.05 max 0.15 max 0.02 max 0.007 max 2.4-2.8 0.09-0.13

Zircaloy – Some Alloys

Zirconium alloys used in nuclear reactor


Zircaloy-2 Zircaloy-4 Zr -2.5Nb

Sn 1.20 - 1.70 1.20 - 1.70

Fe 0.07 - 0.20 0.18 - 0.24

Cr 0.05 - 0.15 0.07 - 0.13

Ni 0.03 - 0.08 -

Nb 2.40 - 2.80

O 0.09 - 0.13

Zircaloy – Some Alloys


Classical:• Zircaloy-2: for BWR application, an alloy with about 1.5% Sn and 0.3% Fe+Cr+Ni)

• Zircaloy-4: with a composition similar to Zircaloy-2 but free of Ni and with lower tendency of hydrogen pick-up, for PWR application

• USSR binary Zr-Nb alloy 1%Nb adn 2.5% Nb were selected for the Soviet PWR and HPWRs

Recent developments:• Zircaloy-2 types with optimized microstructure in BWRs

• Optimized Zircaloy-4 in microstructure and material chemistry)

• Modified Zircaloy-4 a low Sn Zircaloy with increased Fe+Cr)

• ZIRLO (1% Sn, 1%Nb and 0.1% Fe)

• M5 (1% Nb) are used for fuel elements in western PWR‘s

• DUPLEX-ELS – Zircaloy tube having an outer layer with increased Fe+Cr and decreased Sn

Zircaloy – Important Alloys



MATERIAL


HISTORY OF ZIRCALOY





HYDRIDES


GROWTH AND CREEP

TOC


• The ceramic fuel has a lower thermal expansion coefficient but runs at higher temperatures.

• At elevated burn-ups the fuel might give a load on the cladding axially and in circumference.

Radial temperature distribution in a pin

PCMI (Pellet Cladding Mechanical Interaction) I


•Which are the relevant processes for the PCMI?

•Try to draw the pellet-cladding gap evolution as function of time. When the gap closes try to represent the influence to the cladding.

solution

PCMI - Question


Radial expansion of pellets might lead to long axial cracks

PCMI II


PCMI III – Pellet cracking during irradiation


BeforeIrradiation

AfterIrradiation

• Pellet swelling• Pellet bambooing• Pellet cracking• PCMI• Iodine release

B.R.T. Frost (Ed), 1994, Nuclear Materials, Vol. 10B, Materials Science and Technology, VCH, p.43

PCMI IV – Pellet cracking during irradiation


PCMI V – Increase of diameter


Aggressive environment:

Iodine

Tensile stress:Power Ramp

Sensitive Material:Zry-2, Zry-4

SCC

ANT international

SCC:stress corrosion cracking

PCI (Pellet Cladding (chemical) Interaction) I: SCC


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PCI II: Stress concentration at cracks


Sofer G.A. et al. Proc. ANS Top Meeting LWR Fuel Performance, Williamsburg VA, p. 41-53, Apr. 17-20, 1988

PCI III: Example


More cracks in the pellet (new development)

PCI IV: Countermeasure



MATERIAL


HISTORY OF ZIRCALOY





HYDRIDES


GROWTH AND CREEP

TOC


• Debris fretting is the most common failure cause for leakers • Debris are caught in the spacer

Westinghouse Atom

Debris fretting I


Taken from filter of feed water pump

Debris fretting II: Typical Debris


Debris fretting mark in spacer position 7

Hydride plunge between spacer 1 and 2

Debris fretting often cause secondary failures ….

Westinghouse A

tomDebris fretting III: Secondary Failures


Severe circumferential break above

… sometimes ending with the total destruction of the fuel rod.

Westinghouse A

tom

Debris fretting mark in spacer position 6

Debris fretting III: Secondary Failures


Water ingress causes the oxidation of the inner cladding surface. The oxide protects the surface against H-uptake. At a certain distance to the leak, the protection is diminishing and hydride lenses can be developed due to massive hydrogen uptake

Debris fretting IV: Consequences of Fretting Leaks



MATERIAL


HISTORY OF ZIRCALOY





HYDRIDES


GROWTH AND CREEP

TOC


• Hydrogen is taken up during normal corrosion of the cladding (pick-up fraction).

• When the solubility limit is exceeded, hydrogen precipitated as zirconium hydrides

• Zirconium hydride has a considerable high yield strength, but is brittle.

• Zirconium hydrides are normally distributed over the whole cladding thickness (PWR) and form lines due to the texture

• In BWRs we can often observe a “hydride rim“ of 5 to 10 μm in thickness at the outer cladding surface

• Hydrogen can diffuse to the cold spots (sites of oxide spalling) and form hydride lenses

Hydrides I


• terminal solid solubility at dissolution (TSSD)

• terminal solid solubility at precipitation (TSSP)

McMinn et al., 12th Symp. On Zirconium in the Nuclear Industry, 2000

Hydrides II: solubility limits. Irradiation effects irradiated

non‐irradiated


Distribution due to texture

Hydrides III: “Normal” hydride distribution


• Lately cracks have been observed in high burnup rods going from outside to inside, i.e. no PCI

• The crack propagation is connected with the formation of hydrides in front of the crack tip

The mechanism is described as follows: Hydride is diffusing to lattice under high strain, i.e. to the crack tip. At the crack tip, the solubility limit is exceeded and the hydrides are formed perpendicular to the maximum stress. The crack can now propagate along the hydride. New hydride will diffuse to the new crack tip.

Hydrides IV: Delayed Hydride Cracking - DHC


The crack propagation is connected with the formation of hydrides in front of the crack tip

D. Rogers, AECL/Chalk River

Hydrides V: Delayed Hydride Cracking - DHC


A. Alam / PSI

Hydrides VI: Delayed Hydride Cracking - DHC


• Operational transients can support DHC.

• During a higher power period, more hydrides are dissolved. Additionally higher cladding strain might be present due to PCMI.

• A larger international research projects.

Hydrides VII: DHC and operational transients


• Modern alloys with high burnup have taken up 300 ppm (BWR), resp. some 600 ppm (PWR) hydrogen, i.e. far above the solubility limit.

• When transported to the dry storage, the cladding can reach up to 400 °C. At this temperature, hydrides are dissolved. During the subsequent cooling, the hydrides precipitate. If the (circumferential) stress is high enough, they will be radially oriented. If cracks are present, this might cause DHC and lead to rod failure

• This has to be excluded by limitation in stress and max. temperature. example from the lab

Hydrides VIII: DHC and dry storage


Neutron irradiation effects the atom-scale arrangement of the elements in each zirconium alloy

For 50 MWd/kgU, each atom in the lattice has been displaced 20 times (20 dpa) in average.

Consequences:

• Enhances diffusion rates

• Creation of small dislocations, loop dislocation

• Dissolution of SPP‘s

• Solubility changed

• Composition of alloy matrix changed

Irradiation damage I: Embrittlement



MATERIAL


HISTORY OF ZIRCALOY





HYDRIDES


GROWTH AND CREEP

TOC


Every neutron “shot into” the metal matrix cause local amorphization of the crystal structure. Upon recrystallization, interstitials and vacancies are produced.

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Loss of ductilityHardening

Irradiation damage II: Formation of dislocations


• Irradiation growth in the change in shape of a solid at constant volume that occurs during irradiation in absence of stress

• Growth occurs in solids with crystallographic anisotropy; e.g. in orthorhombic α-uranium and hcp α-zirconium; it does not occur in cubic materials such as Fe of steel

• In addition to crystal anisotropy, macroscopic anisotropy (texture) is necessary for growth. The orientation of grains(crystallites) in a polycrystalline specimen can not be random but must be anisotropic as well

• Texture results from the fabrication process

Irradiation damage III: Growth


Crystallographic anisotropy:

fcc structures:(isotropic, no growth)

α-uranium:• elongates along b• shrinks along a• unchanged along c

α-Zr:• elongates along a• shrinks along ca

Irradiation damage IV: Growth


Irradiation damage V: Growth


For Zircaloy: • expansion along <a>-direction• contraction along <c>-direction

For Zircaloy cladding

with a radial basal plane

texture:

• wall thickness decreasing,

• axial length (and diameter) increasing.

For Zircaloy cladding

with a radial basal plane

texture:

• wall thickness decreasing,

• axial length (and diameter) increasing.

Lem

aign

anan

Mot

ta,

1994

CW: cold-worked stress-relieved

RX: recristallized alloy

CW: cold-worked stress-relieved

RX: recristallized alloy

Irradiation damage VI: Growth


• Rod length growth is limited due to the assembly design

• Asymmetric irradiation might lead to asymmetric length growth might lead to bowing

• Also relevant for fuel channels (BWR)

Irradiation damage VII: Growth


• Creep occurs at elevated temperatures. It is a diffusion controlled deformation under stresses lower than the yield stress

• Irradiation creep is the additional contribution of irritation to the creep behavior

• The irradiation affects both the primary and the secondary creep; creep rate is increased in proportion to the dose rate

Lemaignan and Motta, 1994

Irradiation damage VIII: Creep


U.S. LWR Fuel Failure Rates / Total power:101 GWe

Failures


Appendix: Phase Diagram: Zr-Sn


Appendix: Phase Diagram: Zr-Cr


Appendix: Phase Diagram: Zr-Fe


Appendix: Phase Diagram: Ni-Zr


Appendix: Phase Diagram: Zr-O


Appendix: Phase diagram: Zr-H

Documents

Nuclear Fuels and Materials - 151-2017-00L · 2016-03-13 · Nuclear Fuels and Materials - 151-2017-00L ⌸Lecture 4 - Page 13/86 • Purpose: Enclosure of fuel pellets • Mechanical