Light Water Reactor Fuel Performance Review of Degradation ...€¦ · ¾Secondary degradation can induce different modes of failure like “guillotine” rupture, perforation with

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Light Water Reactor Fuel PerformanceReview of Degradation Phenomena affecting Fuel

Rod Cladding

P.Bouffioux (EDF R&D) – B.Cheng (EPRI)

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Introducing LWR Fuel (1)

PWR

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Introducing LWR Fuel (2)

BWR

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Cladding Material for LWR Fuel (1)

Zirconium alloys are commonly used as material for cladding tubes in LWR because of their inherent resistance to a wide variety of environmental conditions and their neutron transparency

The most used alloys are Zircaloy 2 in BWR and Zircaloy 4 in PWR

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According to the evolution of PWR operating conditions (load follow, high burn up), the fuel vendors have developed new Zr alloys with improved performances

Nb based binary and quaternary alloys

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In order to mitigate the risk of failure during power transients in BWR, the fuel vendors have developed a barrier cladding by co-extrusion of Zircaloy 2 bulk with an alloyed inner liner

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Irradiation damage (1)

Interaction between high energy neutrons (E > 1 MeV) and cladding bulk cause irradiation damage by producing point defects (vacancies, interstitials) and dislocations loops

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Irradiation damage has a significant effect on burst and tensile properties by

inducing hardening (For CWSR material, the initial high dislocation loop density attenuates the irradiation effect)reducing the ductility

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Irradiation damage has a significant effect on viscoplastic behavior

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Irradiation damage increases the creep resistance

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Irradiation growth

Dimensional changes of the cladding at constant volume in the absence of stress application

Expansion along the <a> axis of the hexagonal crystallite concomitant with contraction along <c> axisAccording to texture, irradiation growth evinces axial length increasing

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Primary Environment Corrosion

In primary environment (water or steam), Zr alloy cladding undergoes corrosion according to following chemical reaction

Zr + 2H2O → ZrO2 + 2(1 – w)H2(coolant) + 4w H(metal)

w : fraction of reaction produced hydrogen absorbed by the metal

Progressive formation of a ZrO2 layer

Hydriding of the cladding metal bulk

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Uniform Corrosion (1)

Uniform corrosion is the dominant mechanism observed in LWR

PWR – CWSR Zy 4 after 1, 2 & 4 cycles PWR – RXA Zr – 1% Nb after 1 & 3 cycles

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Oxide layer grows homogeneously overall outer cladding surface

Schematic scenario of corrosion

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Uniform corrosion is very sensitive to temperature

In-Reactor From Laboratory

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Uniform corrosion resistance is strongly dependent upon the chemical composition and the microstructure of the Zr alloy

M5 (Binary alloy with fully recrystallized microstructure) appears the most corrosion resistant while CWSR Zircaloy is the less resistant

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Laboratory studies have shown cyclic kinetics for uniform corrosion with a ZrO2 layer periodicity of about 2 µm

This phenomenon has been validated for any Zr alloy

Long term corrosion resistance is governed by a critical thickness concomitant to hydride precipitation in the vicinity of outer diameter cladding

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Maximum oxide layer have not to exceed technological limit in order

To minimize loss of mechanical resistance of the cladding concomitant to the loss of metal (≈ 0.6 µm for 1 µm of oxide)To minimize hydrogen pick-upTo avoid local oxide delamination

Outer surface of the cladding experience an excess of cooling (local cold spot)Impact on hydride precipitation

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Nodular Corrosion (1)

Nodular corrosion is typically observed in BWR

In steam environment, the oxide growth is localized to small nodules

Second phase particles, with a large size, seem to be the sites for nodule nucleation

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Schematic approach of nodular corrosion mechanism

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Severe risk of oxide spallation and cladding perforation

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Degradation by Crud Induced Localized Corrosion (CILC)

Phenomenon observed in BWR with brass condensersInteraction between Cu in the coolant and the nodulesDeposits in layers with oxides, forming steam pockets Temperature is locally rising Enhanced corrosion concomitant with pickling Significant risk of failure

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Shadow Corrosion

Phenomenon observed in BWR on cladding in close vicinity of other alloys like stainless steel or Inconel

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Hydriding : Phenomenology (1)

As Zirconium is corroding in primary environment (water or steam), the cladding metal bulk is absorbing a fraction of the hydrogen released by the oxidation reaction

This hydrogen precipitates as hydrides in the bulk when the solubility limit is exceeded

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Hydriding : Phenomenology (2)

Fraction of hydrogen absorbed by the metal is directly correlated to the ZrO2 layer thickness

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Hydride Morphology

Zr hydrides are normally distributed over the whole cladding thickness, precipitating under the form of platelets, preferentially aligned along the hoop direction according to texture

PWR (High Burn up)CWSR Zy4 (ZrO2 > 50 µm), a hydride rim of about 30 µm to 60 µm is observed close to the colder outer surface of claddingM5 cladding (ZrO2 ≈ 20 µm), according to low hydrogen pick-up, no hydride rim is observed

BWRBarrier Zy2 cladding (ZrO2 ≈ 20 µm), most of the hydrides tend to precipitate in the liner

PWR – CWSR Zy4 - ZrO2 > 50 µm PWR – RXA M5 – ZrO2 ≈ 20 µm BWR – Zy2 with liner - ZrO2 ≈ 20 µm

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Impact of Oxide Spalling on Hydride Precipitation

Impact of oxide spalling at high burn-upIf the ZrO2 layer starts to delaminate and to spall, the outer surface of the cladding will locally experience an excess of coolingHydrogen diffuse to that cold spot and precipitate to form a hydride lens

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Impact of Hydrides on Mechanical Properties (1)

Hydrides have no impact on yield and ultimate stresses

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Uniformly distributed hoop hydrides reduces the ductility of unirradiated material

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On irradiated material, irradiation damage has a dominant effect on the ductility

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Deleterious Impact of Hydride Rim or Lens

Hydride rim has significant deleterious effect on the risk of cladding failure under severe power transient

Crack initiates in the rim and propagates in the bulkDuctility is drastically reduced

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Long Term Hydrogen Assisted Cladding Failure

Fracture proceeds by Delayed Hydride Cracking (DHC) mechanism

Phenomenon might be activated under decreasing temperature, for instance during Dry StorageThe pre-existence of a crack is required (For instance, initiation in hydride rim under reactor operation)Propagation of the crack is assisted by hydrogen diffusion and hydride precipitation at the crack tip

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Re-orientation of Hydrides

Hydride reorientation (HRO) from hoop to radial direction may occur when the cladding is cooled down under the tensile hoop stress generated by the internal pressure of the fuel rod

The precipitation of radial hydrides is observed whether the cladding hoop stress exceeds some critical value

The radial hydrides have a deleterious impact on the mechanical properties of the cladding i.e Ductility drop

HRO phenomenon is a key issue with respect to the Spent Nuclear Fuel (SNF) integrity during dry transportation and storage

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Re-orientation of Hydrides : Quantitative Analysis (1)

Hydride reorientation (HRO) conditions have been assessed from laboratory testsFor CWSR Zy4, the analysis have been performed through the Deformation Energy

Density dissipated at failure in burst or ring tensile testsThe results have revealed a ductile to brittle transition

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Re-orientation of Hydrides : Quantitative Analysis (2)

Hydride reorientation (HRO) conditions have been assessed from laboratory testsFor Zy2 with Liner , the interpretation have been based on an index describing, from

image analysis, the fraction of radial hydrides The results of ring compressive tests do not reveal a so clear ductile to brittle

transition as observed on CWSR Zy4

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Pellet Clad Interaction (1)

Pellet – Cladding gap closes progressively from the concomitant effect of cladding creep down and fuel pellet swelling

Cladding is strained under tensile conditions, the hoop stress saturating once the pellet swelling rate balances the cladding creep rate

PCI results from the thermal expansion of the pellet during power transients

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PCI failure risk requires three simultaneous conditions : chemical aggressive environment (release of fission product like iodine), sensitive cladding material & tensile hoop stress

PCI defects tends to initiate on cladding ridge, in front of pellet cracks where local stress concentration is maximum

ghRigde

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PCI failure generally appears as a pinholePCI might generate incipient defects without through wall cracking

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PCI might lead to significant cladding opening when the failure process is assisted by hydrides

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Fretting (1)

Debris and Grid to Rod Fretting is a major cause of cladding degradation in LWRRod vibrations induced by flow turbulence can lead to wear of the cladding by Inconel spring in the grid spacer

PWR Spacer Grid BWR Spacer Grid

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Fretting (2)

Debris and Grid to Rod Fretting is a major cause of cladding degradation in LWRDebris can be trapped in the fuel rod bundle and wear the cladding

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Fretting (3)

Fretting does not systematically lead to through wall crack

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Secondary degradation (1)

Secondary degradation results from water ingress in the fuel rod after primary failure (PCI, fretting, CILC)

Water causes oxidation of the inner surface of the cladding as well as of the fuel pelletSignificant amount of hydrogen is produced which is picked up by the cladding at

some distance away from the primary defectHydride Sun Burst can be formed leading to the perforation of the cladding

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Secondary degradation (2)

Secondary degradation can induce different modes of failure like “guillotine” rupture, perforation with loss of metal and axial splitting

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Summary (1)

The Zr alloys are commonly used as material for cladding of fuel rods in LWR

The reactor experience feedback shows excellent operating reliability

However, fuel performance is affected by number of phenomena which might induce significant degradation i.e.

Irradiation damage

Hardening of the mechanical properties ( YS, UTS) and loss of ductility

Excessive irradiation growth Impact on assembly dimensional stability

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Summary (2)

Primary environment corrosion

Uniform corrosion

o ZrO2 layer formation loss of metal weakening of the cladding

mechanical resistance

o Hydrogen pickup hydriding

o Excessive corrosion beyond design limit and oxide spalling

Nodular corrosion

o CILC risk of localized cladding perforation

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Summary (3)

Hydriding

Cladding embrittlement, especially in case of hydride lens formation due to oxide spalling or of radial hydride precipitation

Pellet Clad interaction

PCI is a reality in BWR while it is more a potentiality in PWR

Risk of failure (pinhole) or incipient defect formation without through wall cracking

Fretting

Wear by debris or by cladding –spring contact in spacer grid cell risk of perforation or defect initiation without failure

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Summary (4)

In-reactor cladding degradation might lead to major deleterious impact on fuel behavior during the different phases of the back end fuel cycle (wet storage, transportation, dry storage)

Fuel vendors are developing and designin new products to minimize in-reactor degradation i.e.

Zr alloy with enhanced water corrosion resistance in PWR (M5, Optimized Zirlo)

Barrier cladding (Zircaloy 2 with Liner) to mitigate PCI in BWR

Optimized assembly structure to eradicate risk of fretting by reducing thermal-hydraulic vibrations in spacer grid cells

Optimization of reactor operating conditions might also contribute to minimize fuel degradation

Documents

Light Water Reactor Fuel Performance Review of Degradation ...€¦ · ¾Secondary degradation can induce different modes of failure like “guillotine” rupture, perforation with