Comprehensive nuclear materials 5 07 performance of aluminum in research reactors

Comprehensive nuclear materials 5 07 performance of aluminum in research reactors Comprehensive nuclear materials 5 07 performance of aluminum in research reactors Comprehensive nuclear materials 5 07 performance of aluminum in research reactors Comprehensive nuclear materials 5 07 performance of aluminum in research reactors Comprehensive nuclear materials 5 07 performance of aluminum in research reactors v

K Farrell

Formerly of Oak Ridge National Laboratory, Oak Ridge, TN, USA

ß 2012 Elsevier Ltd All rights reserved.

5.07.7.4 Radiation Softening, Creep, and Stress Relaxation 170

Abbreviations

AIME American Institute of Mining,

Metallurgical, and Petroleum Engineers

ANL Argonne National Laboratory

ANSI American National Standards Institute

ASM American Society for Metals

ASTM American Society for Testing Materials

ATR Advanced Test Reactor

CRC Chemical Rubber Company

CTE Coefficient of thermal expansion

EBR-II Experimental Breeder Reactor-II

E mod Modulus of elasticity

ETR Experimental test reactor

GR Graphite Reactor

HEU Highly enriched uranium

HFIR High Flux Isotope Reactor

HPRR High performance research reactor

IAEA International Atomic Energy Authority

INL Idaho National Laboratory

IRV-M2 Acronym for a recent Russian research

reactor

LANL Los Alamos National Laboratory LEU Low enriched uranium

MTR Specifically, MTR is the Materials Testing

Reactor at Idaho National Laboratory Also used generically for materials test reactors

OPAL Open Pool Australian Light water reactor ORNL Oak Ridge National Laboratory

ORR Oak Ridge Research Reactor PIE Post irradiation examination PIREX Proton Irradiation Experiment facility RERTR Reduced enrichment for research and

test reactors

RR Research reactor SNF Spent nuclear fuel STP Special Technical Publication TRIGA Test, research, isotopes, general atomic TEM Transmission electron microscopy UTS Ultimate tensile stress

VPH Vickers pyramid hardness

YS Yield stress

143

5.07.1 Introduction

Aluminum alloys are generally too weak or have

temperature limitations that preclude their use in

reactors built to produce electricity, high-temperature

process heat, or marine propulsion But in the milder

conditions in most research reactors (RRs) where

bulk water coolant temperatures are usually<100
C,

aluminum alloys are quite comfortable and are

uni-versally employed and have greatly contributed to

the success and longevity of the reactors RRs are

those whose principal function is to generate

neu-trons for purposes of nuclear education and training,

production of medical and industrial isotopes,

neu-tron activation analyses, neuneu-tron scattering studies,

and even semiconductor doping, neutron

radiogra-phy, and food preservation treatments RRs are also

employed to study basic radiation effects in materials

and as test beds for evaluating candidate

struc-tural materials and fuels/assemblies for power

reac-tors RRs come in many shapes, sizes, and types

For descriptions of the various classes of RRs, see

http://www.world-nuclear.org/and West.1They are

generally low power, typically about a few kilowatts,

thermal, but range up to about 250 MW According to

the recently updated list2of worldwide RRs published

by the IAEA, a total of 674 RRs have been built in

57 countries, of which 234 are still operational, and

7 are planned or under construction Two new ones

are OPAL, the 20 MW Open Pool Australian Light

water-cooled reactor, which opened at Lucas Heights,

Sydney, in April 2007, and the Russian 4 MW

pool-type IRV-M2 commissioned in 2008

This chapter is a review, more a tutorial, of the

behavior of aluminum alloys in water-cooled RRs

It is a somewhat personal view, based on American

experience in the area Because that experience

has been adopted in many countries and is still

influencing the current state of the art, this chapter

should be of interest outside the borders of the

United States

Aluminum is the material of choice for

con-struction of many components in low-temperature

water-cooled-and-moderated RRs Typical

applica-tions are the reactor tanks in open-pool reactors;

containment vessels in some sealed reactors; core

grids; pedestals; neutron beam tubes; cold neutron

source moderator vessels; shrouds to direct and

separate water flows; shuttles (‘rabbits’) and num filler powder used to convey isotope targetmaterials and test materials rapidly in and out ofthe reactor via aluminum hydraulic and pneumatictubes; sheaths and finned tubing for stationary long-term isotope target rods; cladding for control plates/rods; cladding and liners for reflector materials; clad-ding and thermal conduction filler for fuel rods/plates; and temporary plugs for closing idle irradiationfacilities in and around the core Applications outsidethe reactor per se are in-pool tool extension arms;transfer gates between pool sections; restraint baskets

alumi-in some shippalumi-ing casks; support beams for pool covers;and hot cells manipulator arms

5.07.2.1 History of AluminumApplications in Research ReactorsAluminum was at the forefront of the development

of nuclear technology It has the distinction of beingthe first nonfissile, non-neutron absorber class metalused in the world’s first continuously operatingnuclear reactor, the X-10 Graphite Reactor at OakRidge, TN The Graphite Reactor became critical

on 4 November 1943,<1 year after Fermi’s stration of a self-sustaining nuclear fission chain inthe graphite pile at the University of Chicago on

demon-2 December 194demon-2 In Fermi’s experiment, the onlymetals in the pile were natural uranium and thecadmium-coated control rods The pieces of naturaluranium (238U containing about 0.7 at.% 235U) anduranium oxide were bare, placed in shallow depres-sions carved into the upper faces of the graphite slabs,and cooled by convection of ambient air The powerlevel was about 2 kW The X-10 Graphite Reactorpile3 was much bigger than the Chicago pile andwas designed to operate at 1 MW thermal power,later upgraded to 4 MW It was built to producepilot plant quantities of plutonium isotopes TheChicago pile had no shielding; the Graphite Reactorwas shielded by a 2.2-m thickness of high-densityconcrete Aluminum made its debut in the Graph-ite Reactor as fuel cladding to protect the highlychemically reactive uranium from contamination

by air and graphite during the higher power andlonger fissioning periods and to safeguard it fromattack by water during subsequent radioactive decay

in underwater storage In addition, it trapped themore copious volatile radiation products resultingfrom the longer irradiation exposures Thesealuminum–clad pieces of natural uranium, called

‘slugs,’ were the forerunners of metal–clad fuel

elements A slug was made by placing a solid cylinder

of uranium in a thimble-shaped aluminum can

25 mm diameter 100 mm long with a 0.75 mm

wall A flat Al end cap was added, and the assembly

was passed through a die to force the can walls tightly

around the fuel Surplus wall material was cut off

above the cap, and the cap was welded all around its

edge These slugs were pushed end to end into the

reactor via round horizontal holes through the

con-crete face, which were aligned with 44 mm square

holes cut through the full 7.3 m width of the cubic

array of graphite blocks The square holes were

ori-ented on edge such that the slugs occupied the lower

corner, allowing cooling spaces around the slugs

Cooling was simple: two large fans at the rear of the

pile sucked ambient air through the holes around the

slugs and discharged it up a tall chimney The slugs

exited the pile at the rear face and were channeled into

a deep water canal where they were held until shipped

to hot cells for processing to extract the plutonium

Some early problems4were encountered in the slugs,

including faulty welds and blisters and formation of

an intermetallic UAl3phase by interdiffusion at the

U–Al interface, especially in the high-temperature

regions in the center of the reactor The blistering

was attributed to fast-growing gas bubbles in the UAl3

phase These problems were overcome by better

welding practice and the development of bonded

slugs as described next

The next phase of exploitation of aluminum was

in the B reactor at the Hanford site in Washington

State, which went critical on 27 September 1944

The B reactor was a scaled-up production model

of the Graphite Reactor designed to operate at

250 MW At such power, forced air cooling would

have been inadequate So the horizontal holes

were replaced with aluminum tubes in which

aluminum–clad uranium slugs were cooled with

flowing water from the Columbia River To improve

the transfer of heat from the uranium to the

clad-ding, the spaces between them were filled with a low

melting Al–12% Si eutectic alloy by melting the

eutectic in situ A bonus of this treatment was that

it killed the formation of the UAl3 phase and

asso-ciated blistering, presumably due to an inhibiting

effect of the silicon The successes of these upgrades

established aluminum as a suitable material for use

in combined conditions of intensive irradiation and

a flowing aqueous environment Aluminum became

more firmly entrenched in RRs with the

develop-ment of advanced fuel eledevelop-ments, as described in

of a permanent magnet Sufficient electricity wasgenerated to light a flashlight bulb The thermalefficiency was estimated to be 2% The GraphiteReactor is now a National Historic Landmark and isopen to the public A commemorative plaque and areplica of the steam engine and coupled dynamofrom Ramsey and Cagle’s pioneering boiling waterpower reactor are displayed in a small showcase

in the reactor lobby The ‘official’ first production

of nuclear electricity is credited to the lighting ofanother bulb on December 1951 at the ExperimentalBreeder Reactor-I, Arco, Idaho, now the IdahoNational Laboratory

Heat removal and reduced generation of heatare major considerations in the popularity of alumi-num in RRs Most of the energy released from con-trolled nuclear fission appears as heat Much of theheat, >80%, arises in the fuel from nuclear fissions.However, a significant portion, 5–20%, is produced

in the nonfissile materials in the core and its roundings by bombardment with particles emanatingfrom the fission reactions and from decay of fissionproducts For power reactors, the heat is essential togenerate the electrical output In the case of RRs, theheat is a nuisance product; and the goals are tominimize heat generation from the nonfissile materi-als in the system and to get rid of it from thosematerials and from the fuel as fast as possible.Hence, structural materials that create the least heatand/or conduct it away the fastest are the most

sur-favored for RRs In this regard, aluminum is

out-standing Generally, heat production is greater with

increasing material density and with decreasing

specific heat It is increased by high cross-sections

for neutron absorption and scattering, which also

reduce reactor efficiency by stealing neutrons from

participation in fission processes Heat removal rate

is larger with higher thermal conductivity

There-fore, construction materials with low density, high

specific heat, high thermal conductivity, and low

nuclear cross-sections offer the best prospects for

minimizing heat generation and maximizing heat

removal In Table 1, the relevant properties for

aluminum are compared with those of other

clad-ding and structural materials used in power reactors

and for uranium All values are for room

tempera-ture or 100
C The scatter in values for a given

parameter and material is due in part to sensitivity

to chemical composition and heat treatment, etc

These variations do not mask the large differences

between Al and the other materials The density of

Al is 1/2–1/3 of those of the other cladding

materi-als, and only 1/7 that of U Its specific heat capacity

is twice as high as the other materials And its

thermal conductivity is 5–10 times greater than

the values for the other materials Additionally,

its neutron capture and scattering cross-sections are

much smaller than those of the other materials, except

for nuclear-grade Zr In that respect, it should

be remembered that in the early days when Al was

establishing its foothold in nuclear technology

com-mercial Zr was contaminated with up to 3% of the

strong neutron absorber Hf It was also inordinately

expensive

5.07.3.1 Practical CharacteristicsHaving attractive physical properties for reactor use

is of no merit if those properties cannot be exploited

in a practical manner The commercial and economicattributes of aluminum that encourage its deploy-ment in RRs are: It is ductile, plentiful, cheap, andlight weight It is castable, machineable, and weld-able, and it can be shaped readily by conventionalprocesses of rolling, forging, extrusion, drawing, andcupping It has good aqueous corrosion resistance due

to near-insolubility in water and formation of a sive, self-restoring surface film of hydrated aluminumoxide It is nonmagnetic and nonsparking Althoughaluminum is inherently weak, it can be strengthened

pas-by cold work, solid solution hardening, and tation treatments It has an fcc crystal structure and

precipi-no crystallographic phase changes Its crystal ture is near isotropic, ensuring that it will not sufferdamaging directional thermal expansion and radia-tion growth like those exhibited by graphite and thehexagonal metals Mg and Zr It does not form stableembrittling hydride phase(s) as Ti and Zr do At lowtemperatures, it has no ductile-to-brittle transition

struc-On the contrary, it is somewhat special in that atcryogenic temperatures, where it gains strength, itoften gains ductility too This combination of nohydride phase, outstanding low temperature proper-ties, and low neutron cross-sections make aluminumthe prime material for building cold neutron sources.Another attractive feature is that pure aluminum has

no long-lived radioisotopes The major source ofimmediate radioactivity is from decay of 24Na pro-duced via 27Al(n,a)24Na, decaying by g-emission

Table 1 Relevant properties of reactor materials

Material Density

(kg m3)

Specific heat (J kg1K1)

Thermal conductivity (W m1K1)

Melting point (
C)

E mod (GPa) CTE, lin.

( 10 6 K1)

Nuclear cross-section (barns)

with a half-life of 15 h In alloys, long-lived

radioac-tivity arises from decay of isotopes produced from

alloying elements and residual impurity elements

present in the aluminum, primarily 65Zn,51Cr, 59Fe,

with half-lives of 250, 28, and 45 days respectively

So if low residual radioactivity is an objective it can

be met to a large extent by avoiding alloys containing

significant quantities of Zn, Cr, and Fe

Aluminum is not without its shortcomings It has a

low elastic modulus and low melting temperature

The former means that in their annealed conditions

aluminum alloys have low strengths compared with

annealed austenitic steels, Zr, and bcc metals

How-ever, aluminum can be hardened by various treatments

as described in Section 5.07.3.2 However, the low

melting temperature of 660
C imposes operating

temperature limits of 100–150
C, which are

homol-ogous temperatures of 0.4–0.45Tm where lattice

vacancies are mobile and can invoke susceptibility

to creep and stress relaxation Even without imposed

stresses, the strength condition of prehardened alloys

can become compromised at temperatures above

150
C because of the possibility of thermal overaging

as described inSection 5.07.3.2Aluminum has poor

abrasion resistance It can be sensitive to localized

galvanic and pitting corrosion It is prone to liquid

metal embrittlement, particularly Hg Difficulties

may be encountered in obtaining leak-tight fusion

welded joints for hi-tech applications, mainly due

to porosities resulting from solidification shrinkage

(volumetric change) and dissolved gases, in particular,

hydrogen.5In addition, aluminum does not undergo a

color change on heating, and during manual welding

may melt abruptly without warning, allowing

over-heating that can cause excessive sagging and

drop-through of the weld bead The advent of a solid-state

joining process, namely friction-stir welding,6 has

largely overcome those welding troubles

5.07.3.2 Alloy Types, Temper

Designations, and Tensile Properties

There is no universally embraced international

stan-dard system for defining the types and conditions

of aluminum alloys The International Standards

Organization does have classifications for aluminum

and its alloys, but most countries adhere to their

own systems The system followed in the United

States of America is ANSI H35.1-1990, instituted

by the American National Standards Institute and

supported by the Aluminum Association and ASM

International The ANSI system and the US alloys

covered by it are described in reference,7which is anexcellent source of aluminum data; it includes a shortlist of alloys for other nations and their nationaldesignations The ANSI system is used herein In itsentirety, it is a morass Here, it is outlined just to theextent that is necessary to provide an uninformedreader with enough details to understand the nomen-clature and the various processing treatments and theupper service temperature limits those treatmentsimpose for maintaining stability of the processedmaterials

The system has two classifications, one forwrought alloys and one for cast alloys Only thewrought alloy classification is described here Briefly,

it is an eight-character code consisting of two groups

of four characters separated by a hyphen The firstfour characters are all numerals and they identify thealloy group by chemical composition There are eightaluminum alloy groups The first digit of the first alloygroup is 1, which represents alloys with a minimum of99.00 wt% aluminum In this group, the major foreignelements are Fe and Si, which are really residues fromthe aluminum extraction process and will be found tovarious degrees in all aluminum alloys The next threedigits in the group identify specific alloys in the sameseries, and the group as a whole is denoted the 1xxxseries, often vocalized as the one-thousand series.The other seven alloy series are 2xxx (major alloyingelement, Cu), 3xxx (Mn), 4xxx (Si), 5xxx (Mg), 6xxx(Mgþ Si), 7xxx (Zn), and 8xxx (other element)

An upper case X preceding the series identifiernumeral indicates an experimental alloy

The second group of four characters in the character designation represents the temper condi-tion, that is, the heat treatment or degree of coldwork The first character of the four-character tem-per condition is an upper case letter representing atype of treatment The other three characters are digitsindicating variations within the treatment There aremany temper treatments Only the three treatmentsmost likely to be encountered in RR materials aredescribed here They are ‘O’ for the fully annealedcondition, ‘H’ for a strain-hardened condition, and

eight-‘T’ for a precipitation-hardened condition The

O condition is attained by annealing the material atabout 340
C then slowly cooling it There are nospecified variations of the O condition The H temper

is more complex The first digit after the H is a 1, 2,

or 3 H1 signifies hardened only H2 is hardened and partially annealed H3 is strain-hardenedand stabilized by a low temperature heat treatment.The second digit, that is, the one following the H1,

strain-H2, or H3 designation is a number between 1 and

8 and is the degree of reduction in thickness or

cross-sectional area given to the alloy in its fully annealed

condition to bring it to the desired strength level

Level 8 corresponds to a maximum reduction of

about 75% Level 1 represents approximately

one-eighth of 75%, 2 is two-one-eighths, and so on The third

digit, if used, implies a variation of the two-digit

temper Partial annealing for the H2 condition is

applied to products that are strained beyond the

desired final amounts and are then brought back

to the needed strength level by the partial anneal

Stabilization heat treatment for the H3 condition

is applied to those products that, unless stabilized,

would gradually age-soften at room temperature

Partial annealing also inhibits age softening This

tendency for softening of some cold-worked

alumi-num alloys at room temperature is important

because such recovery requires the involvement of

mobile lattice vacancies and/or self-interstitial

atoms that promote climb and rearrangement of the

cold work dislocations It indicates the occurrence of

atomic movement at room temperature, which, as we

shall see shortly, is a factor affecting the

develop-ment of radiation damage in aluminum

In addition to hardening by cold work, aluminum

can be strengthened by solid solution treatment

and by precipitation hardening Only two alloying

elements, Mg and Li, have sufficient solubility

(several %) at room temperature to provide

signifi-cant solid solution strengthening Al–Li alloys are

not recommended for reactor use because natural

Li contains about 7.5%6Li, which has a large

cross-section for transmutation to 3H and 4He, both of

which can be highly detrimental to aluminum

So the only solid solution-hardened alloys available

for reactor use are the 5xxx (Al–Mg) series Other

metallic elements, principally Cu, Si, and Zn, have

little or no solubility in aluminum at room

temper-ature but are modestly soluble at higher tempertemper-atures

near the melting point This latitude permits

consid-erable strengthening of such alloys by

quenching-and-aging, also known as precipitation hardening The

ANSI designations for the precipitation-hardened

T conditions comprise ten subdivisions, T1–T10

For all T treatments, the alloy is heated to a

temper-ature of 500–540
C to dissolve segregated alloying

elements, followed by a rapid quench into cold water,

which gives an unstable supersaturated solid solution

Precipitation is achieved by allowing the material to

sit at room temperature for periods of weeks called

‘natural aging’ (tempers T1–T4) or by ‘artificial

aging’ at temperatures of 160–190
C for times of6–24 h (tempers T5–T10) Flattening or straighten-ing treatments may be applied before or after theaging treatment and are indicated by numbers inthe third and fourth character positions The temperconditions for aluminum alloys most often encoun-tered in RRs are T4, T6, and T651 A T651condition indicates a material that has been artifi-cially aged then subjected to a light stretchingoperation insufficient to change its mechanicalproperties from those of the T6 condition Of theprecipitation-hardened alloys, the 6xxx series hard-ened by precipitates of Mg2Si is by far the mostpopular for RR service The 6061 alloy in its T6and T651 conditions has been approved for ser-vice as a class 1 nuclear components material underthe Boiler and Pressure Vessel Code of the Ameri-can Society of Mechanical Engineers, Case N-519.8Two types of precipitation-hardenable wroughtaluminum alloys, the 2xxx series (Al–Cu) and the7xxx series (Al–Zn), both of which can be heattreated to greater strengths than the 6xxx alloys,are not usually found in nuclear reactors Some2xxx alloys are prone to aqueous pitting corrosion

or may release Cu ions to the coolant that could

be deleterious to other materials in the reactorsuch as stainless steel The 7xxx series alloyshave too low ductility and are the most difficult toweld Their high zinc contents will cause highradioactivity

Unlike the cold-worked 1xxx alloys that canundergo recovery at room temperature, theprecipitation-hardened alloys are thermally stable

at temperatures up to about 150
C provided theyhave been given appropriate natural or artificialaging treatments However, exposure to higher tem-peratures will cause overaging and associated re-duction in mechanical strength This softening isillustrated inFigure 1for 6061-T6 alloy after heating

to various temperatures for various times and testing

at room temperature.9It can be seen that softening ispromoted by both time and temperature For times

up to 1 h, softening commences at about 200
C and

is substantial but incomplete at about 370
C For

a longer exposure of 1000 h, the softening beginsaround the aging temperature, indicated by thedown-pointing arrow, and is essentially complete attemperatures between 260 and 300
C The data in

Figure 1 are for specimens reheated without load

If reheating occurs under loads sufficient to inducecreep and stress relaxation, the softening tempera-tures are pushed downward

Table 2gives typical tensile properties of various

Al alloys employed in RRs The weak 1100-O alloy

is simply annealed commercial purity aluminum with

no deliberate alloy additions; it is hardenable to an

H condition by cold work if so desired The 4032

alloy is a eutectic composition of Si in aluminum that

has been solution-treated and aged to create finely

divided precipitates of Si; this alloy is used

princi-pally as a filler wire to improve the weldability of

aluminum alloys The 5052 alloy is a solid solution

alloy of 2.5% Mg with a small amount of Cr added to

control grain size and strengthen the grain

bound-aries The particular 5052 alloy in the table has been

work hardened to a 4/8, or half-hard, condition

before stabilization The 6061-T651 alloy has been

solution treated and artificially hardened by

precipi-tates of Mg2Si phase and its precursors, then given a

mild stretching treatment

conduc-0

50 100 150 200 250 300 350

Elong., 6 and 30 min

Figure 1 Softening effects of reheating temperature and time on room temperature properties of 6061-T6 aluminum (originally aged 18 h at 160
C) Data from Structural Alloys Handbook, 1989 ed., Vol 3, Battelle Memorial Institute, Columbus,

OH, 1989; p 14.

Table 2 Example alloys and their room temperature tensile properties

common ones are U3O8, UAlx, wherex can be 2, 3, or 4

but is usually considered10to be a mixture of 3 and 4,

and U3Si2 There is also a hydride fuel, U–ZrH1.6,

which is used exclusively in the open-pool TRIGA

(test, research, isotopes, General Atomic) types where

the fuel is in the form of slugs comprised of particles of

U dispersed in the ZrH1.6 phase (seeChapter 3.12,

Uranium–Zirconium Hydride Fuel) Originally, the

TRIGA slugs were sheathed in aluminum, which has

now been replaced with stainless steel or nickel alloy

However, TRIGA reactors still contain other

alumi-num components

There is no outstandingly superior aluminum

cladding alloy The most common aluminum cladding

alloys are 1100 and the stronger 6061 Other alloys

have been investigated in neutron irradiations,11

namely 5052; X800N, where N is 1, 2, or 3 and

whose compositions are Al–1Ni–1Fe; and two

sin-tered aluminum powder alloys, M257 and M470,

which were fabricated by ball milling flake powder

of 1100Al in air until it contained a dispersion of 6%

and 10% Al2O3, respectively, then consolidating by

pressing, sintering, and hot rolling The Mxxx alloys

were deemed to be no better than 1100 and 6061

types They are more difficult and expensive to

make and harder to weld than regular melted-type

alloys In Europe, particularly in France, two preferred

alloys are AlFeNi, a relative of X8001 with the

com-position 1Fe–1Ni–1Mg, and AG3-NET, a 5xxx-type

with 2.5–3.0Mg and low residuals The greatest

concern for cladding is its corrosion behavior (see

Section 5.07.5)

A feature of RR fuels is that they are much more

highly enriched in235U than those in power reactors:

12–93% versus about 2.5% Drivers for raising the

235

U levels were extended fuel cycles; the growing

demands for industrial and medical isotopes,

partic-ularly99Mo the parent of the all-important medical

diagnostics tool99mTc; and the need for higher

neu-tron fluxes for increased production of the heavy,

transuranic isotopes The use of highly enriched

uranium (HEU) meant higher heat generation and

required improved means of removing the heat The

solution was the development of dispersion fuels

in which particles of the enriched fuel were distributed

in a matrix of thermal conductor material, all

com-pressed together in sealed aluminum cans The

thermal conductor is aluminum powder, usually a

1xxx-type, often atomized powder of better than

99.5% purity and particle size <100 mesh (150 mm

maximum, 23–48 mm mean) Atomized powder

par-ticles are denser, pour more easily than milled flake

powders, and have less low conductivity surface oxideper unit volume The aluminum matrix may occupymore than 50 vol% of the fuel/aluminum mixture

A huge advance in fuel element morphology andheat removal efficiency took place when EugeneWigner designed his thin, curved fuel plates for thehigh flux Materials Testing Reactor (MTR) built atArco, Idaho A thin plate has a number of advantagesover cylindrical slugs The rolling treatment used toproduce the plates from a fuel slab, or from a disper-sion of fuel particles in aluminum matrix powder,sandwiched between two aluminum cladding sheetsgives superior mutual contact of cladding, matrix, andfuel for improved heat transfer paths to the cladding.The much larger surface-to-volume ratio of platesprovides more efficient heat transfer to the coolant,thus permitting higher fuel loadings per unit volume.The benefit of a curved fuel plate is that any bucklingand bowing in the plate due to irradiation will befocused in the direction of the radius of curvature.Thus, in a fuel element comprised of a stack ofcurved plates restrained at their edges and separatedfrom each other by cooling channels of the samewidth as the thickness of the plates, any such distor-tions will be accommodated cooperatively in theradial direction without unacceptable narrowing ofthe cooling channels An MTR fuel element contained

18 plates each about 72 mm wide and about 727 mmlong bent to a curvature of 140 mm radius in the widthdirection The plate thickness was 1.27 mm including aminimum cladding thickness of 0.25 mm on each face.The plate edges were brazed into sturdy side panels

to seal the plate edges and impart rigidity to theassembly The water gap was 1.27 mm The claddingand side panels were made from 1100Al; the Albrazing alloy contained about 13% Si.12This assem-bly was then enclosed in a long, rectangular alumi-num box fitted with end fixtures for remote handling.The end fixtures were castings of Al–7% Si Thereactor core was built from groups of such elementsassembled upright in rectangular arrays held together

by aluminum grid plates Refueling was done fromthe top, and any element could be replaced by a box

of the same size containing a reactor experiment ormaterials for isotope production, or a berylliumreflector or a control rod These MTR-type boxedfuel elements in open grid core arrangements per-formed very well and became very common for RRs

To satisfy demands for higher power densities andmore sophisticated tailoring of local neutron fluxes,the next advancement in aluminum–clad fuel elementswas the development of upright, annular

elements using curved fuel plates in which the fuel

particles may be required to be graded in

concentra-tion across the thickness and width Beryllium

reflec-tors surrounding the annulus direct neutrons from

the fuel back to the hollow center, or ‘trap,’ of the

annulus where reactor experiments and

iso-tope targets are placed The Be also creates

addi-tional neutrons from (n, 2n) reactions Vertical holes

bored through the reflector allow passage of cooling

water and house irradiation experiments Two

high-performance beryllium-reflected reactors using

annu-lar fuel elements are the High Flux Isotope Reactor

(HFIR) at Oak Ridge National Laboratory (ORNL),

rated at 100 MW thermal and currently running at

85 MW, and the Advanced Test Reactor (ATR) at

Idaho National Laboratory, rated at 250 MW but

lately operating at 100–125 MW The cores of these

reactors are of uncommon designs and deserve

com-ment The ATR core13is 1.22 m diameter and 1.22 m

high It contains a continuous serpentine-like wall of

fuel elements looped around nine flux traps each

about 120 mm diameter arrayed in a square 3 3

grid In plan view, the wall forms the shape of a

four-leaf clover It fully embraces the central flux

trap and the four corner ones The other four traps

lie just outside the wall; each is tucked in between the

junctions of two leaves and is about half wrapped by

the wall At each corner lobe, there are four shim

control cylinders just outside the wall and six shim

rods at the neck of the wall inside the cloverleaf

These controls allow each of the four lobes to be

run at different power levels simultaneously, as

needed by the experiments in the traps The

remain-der of the space in the core is occupied by blocks of

Be reflector containing numerous experiment holes

The wall is built14,15 from 40 individual

wedge-shaped fuel elements, each containing 19 curved

fuel plates The cross-sectional area of an element is

a 45
sector of a circular annulus Its outer arc, plate

#19, has a radius of 137 mm and an arc length of

100.9 mm Its inner arc, plate #1, has a radius of

77 mm and an arc length of 54.1 mm The 19 fuel

plates are attached by roll-swaging to 6061-T6Al side

panels 64.6 mm wide 1257 mm long Within the

elements, the curved plates are concentric with the

circumferences of the traps The plates are 1.27 mm

thick except for #1 and #19, which are thicker The

water gap is 1.98 mm The ATR fuel is UAlxenriched

with235U to 93%, dispersed in a matrix of Al powder

and clad with 0.38 mm thick 6061-OAl

The HFIR core16is more compact, about the size

of a small trash can, into which are packed 540 fuel

plates in quite a different arrangement than in theATR The core diameter is 435 mm and it is 791 mmtall It has a single central flux trap, 129 mm diameter.The fuel is granules of U3O8enriched with 235U to93% and embedded in Al powder The cladding is6061Al The core consists of two concentric annulararrays of involute-curved fuel plates, as shown in thesketch of a radial segment in Figure 2 The blackregion in the fuel plates is the fuel dispersed in its

Al matrix; the white area is Al filler There are 369plates in the outer annulus and 171 in the innerannulus The plates are 610 mm high with widths forthe inner and outer annulus plates of 94 and 81 mm,respectively, before bending The plate thickness andcoolant gaps are 1.27 mm, as in the MTR-type ele-ments The two annuli are fabricated separately andare united when loaded into the reactor In addition tothe unique radial-like orientation of the fuel plates, thefuel particles are uniquely distributed in the plates Tominimize the radial peak-to-average power densityratio, the thickness of the compacted fuel mix is variedalong the arc of the involute curve as seen inFigure 2.This shaped region is backed by filler Al containing

no fuel particles For the inner annulus, the fillerpowder backing the shaped fuel region contains

1.27 mm Coolant channel

1.27 mm

Figure 2 Horizontal section through a small segment

of the HFIR core showing fuel plate curvatures and fuel distributions in the plates Modified from Binford, F.T.; Cramer, E N The High Flux Isotope Reactor; A Functional Description, Vol 1B, Illustrations; ORNL-3572 (Rev.2); Oak Ridge National Laboratory: Oak Ridge, TN, 1968.

particles of B4C burnable poison Two concentric

cylindrical control plates clad in Al are located

imme-diately surrounding the core Outside the control

plates are four concentric cylindrical Be reflectors

Because beryllium generates copious quantities of

helium and tritium from neutron irradiation, it tends

to swell and crack, particularly at the faces of its high

neutron flux regions To retain chips spalled from

these surfaces, the reflector and any penetrations in it

are clad with aluminum Four horizontal 6061Al beam

tubes and numerous vertical holes penetrate the

reflector

Like most dispersion-type fuel plates, the HFIR

and ATR plates are fabricated by what is called

a picture frame technique This utilizes powder

met-allurgy methods to disperse the fuel particles

uni-formly in the Al matrix and press the mixture into

a hard rectangular compact The rigid compact is

placed in a window of the same size cut in an Al

slab or frame, which is usually the same alloy as the

cladding Sheets of cladding material are welded to

the top and bottom faces of the filled frame and the

assembly is hot rolled through a large reduction in

thickness to ensure that the cladding is fully bonded

to the fuel charge and the frame After verifying the

location of the fuel charge, the rolled plate is cold

rolled to flatten it and bring it to the specification

thickness It is then given a final anneal at 500
C

to reveal any blisters and rolling defects in the

cladding surfaces After verifying the location of the

fuel region, the plates are blanked to finished size in

a press

Of course, it is not as simple as that Strict quality

assurance standards have to be met, and at every stage

in the operation, there are numerous inspections

and rigorous sizing and confirmation tests To

repro-ducibly obtain the graded fuel distributions in the

HIFR plates, a special procedure was developed.17,18

A custom-designed contoured auxiliary die plate

is mounted over the cavity of the powder press to

facilitate mounding of the fuel/matrix powder mix in

a semicylindrical hump Another auxiliary die plate is

added to allow filler powder to be leveled on top

of the humped fuel charge This duplex charge is

withdrawn into the press cavity, the auxiliary dies

are removed, the rectangular punch is inserted into

the die mouth and pressure applied, and the charge

is consolidated in a single cold pressing operation

The HFIR fuel plates are bent to the desired involute

shape in an elastomer-faced punch and die press

They are welded into the cylindrical inner and outer

sidewalls of the fuel elements The sidewalls are

machined from extruded-type 6061 aluminum tubing

in the T6511 temper Twenty-seven equally spacedcircumferential weld grooves are turned on one face

of each sidewall, and slots are milled at prescribeddepths and angles on the other face of the wall Theweld grooves intrude a short way into the slots.The fuel plates are slid into the slots and properlyspaced with the aid of temporary Teflon separators.The plates are machine welded in place through thegrooves A 4043Al weld filler wire and an argon shieldgas are used End fixtures machined from 6061Altubing are welded to the ends of the elements, andfinal machining and inspection are conducted.These multiplate fuel elements are a testimonial

to designer ingenuity and superb fabrication skills,and the versatility of aluminum Manufacturing thesefuel elements is not only painstaking but also expen-sive In year 2007, each HFIR element cost$1 M.19

It is replaced after its regular lifetime operatingcycle of 26 days With so much effort and costinvested in it, a rejected element is a severe financialloss The specifications and acceptance standards are

so high that the chances of producing a fuel elementcompletely free of specification violations are verylow The first 30 000 fuel plates suffered a rejectionrate of 12%, and of the first 45 fuel assemblies, only

4 inner elements passed the final inspection.20ever, the degrees of severity of the violations wereall minor or were correctable With waivers, all 45elements were accepted and gave exemplary service.After operation of the first 60 fuel cores at thefull design power level of 100 MW, 4 of them wereautopsied.21 No significant faults were found Thein-reactor performance of these complex ‘aluminum-based’ fuel elements has been incredible, surpassing allexpectations

How-Development of RR fuels and fuel plates iscontinuing Concerns over the possibilities of nuclearweapons proliferation and terrorism led to establish-ment of the Reduced Enrichment for Research andTest Reactor (RERTR) program at Argonne NationalLaboratory.22 The goal of RERTR is to eliminatethe use of highly enriched uranium (HEU) in RRs

by converting to the use of low enriched uranium(LEU) HEU is defined as uranium that has thefraction of the fissile isotope235U greater than 20%,LEU is less than 20% Historically, RRs have usedenrichment levels of 235U up to 93% RERTR

is intended to be achieved without impairing thesafety and performance of the reactors and/or jeo-pardizing the production of important isotopes,and at minimum cost for changes in fuel elements

In some RRs with modest uranium enrichment and

low power levels the RERTR LEU goal was met by

diluting the fuel with natural uranium For many of

the high performance RRs (HPRRs) that must retain

their 235U levels and cannot tolerate the burden of

added238U without excessive operational penalty, the

RERTR dilution can be achieved by replacing the

HEU fuel with LEU compounds or alloys containing

higher fractions of U To that end, the initial focus of

RERTR was on the development of uranium silicide

fuels, U3Si and U3Si2, dispersed in aluminum and

clad with aluminum.23,24 While this move has been

successful for many RRs it is not sufficient for the

most demanding HPRRs For them, attention has

turned away from dispersion fuels to monolithic alloy

fuels where higher U densities are attainable The goal

is to develop fuel plates built from foils of LEU alloy,

250–500 mm thick, clad with aluminum.19,25–27 In

order to prevent buckling and cracking of the foil

during multiple rolling and recrystallization

treat-ments and to inhibit radiation growth and warping,

there must be just enough alloying metal in the

U to stabilize it in its isotropic g-phase Several

alloying metals are suitable, but the field of

conten-ders has been reduced to the U–Mo system A 90%

LEU-10% Mo alloy currently holds the best

pro-spects Some serious hurdles are recognized

Interdif-fusion between the cladding and the fuel foil during

annealing and in-reactor exposure encourages the

formation of reaction layers of uranium–aluminum

compound(s) with low thermal conductivity and low

resistance to growth of fission gas bubbles Such

layers threaten the integrity of the fuel/cladding

interface Development of these layers is retarded

by additions of Zr or Ti to the fuel, or Si in the

cladding When Si is incorporated in the cladding, it

is found to segregate at the fuel/cladding interface,

acting like a diffusion barrier Thin film diffusion

barriers of Si, Zr, and ZrN applied directly to the

surfaces of the fuel foil by co-rolling and thermal

spraying have done well in reactor tests The current

hot roll bonding processes used for attaching

clad-ding to dispersion fuel plates may not be fully

adaptable to barrier-coated foil fuels Other bonding

methods such as hot isostatic pressing are under

investigation For HFIR plates, where the foils must

be tapered in both width and length and have

invo-lute shapes, fitting and bonding diffusion films and

cladding to the fuel foil on a mass production scale

is a challenge Hot roll bonding will not work

because the foil and the cladding will not deform

to the same extent and will result in nonuniformly

thick cladding, and shear deformation during rollingmay damage the diffusion barrier It is recom-mended19 that the tapered foil, bent to its involuteshape and with an adherent diffusion barrier, should

be prepared separately then sandwiched in shapedrecesses in two full-length clamshells of cladding

of appropriate thickness and bonded over all matingsurfaces Alternatively, if the clamshells can be madefrom a two-ply Al sheet, like the commercial One-Side Alclad™, the inner layer of, say, 1100Al, couldcontain the ingredients for a diffusion barrier Thehot isostatic pressing route may then allow bondingand barrier filming in a single operation and in batchmode If burnable poison cannot be incorporated inthe fuel foil, it may be possible to accommodate it

in the inner cladding layer with the diffusion barriercomponents

The corrosion behavior of the Al cladding on alloyfoil fuel elements will need to be explored thor-oughly A penetration of the cladding will probably

be more serious than one in current dispersion fuelplates because the alloy fuel will likely be more reac-tive and soluble in water than the dispersant-typeintermetallic and refractory fuels

Metallic corrosion, the removal of metal atoms fromthe metal surface by the electrochemical action ofthe environment, has many forms: uniform, galvanic,pitting, grain boundary, crevice, etc Uniform corro-sion and pitting are the types of most interest to RRs.The greatest worry is the aluminum fuel claddingwhere the environmental conditions are most aggres-sive and where an unexpectedly high corrosion ratemight breach the cladding and allow release of highlyradioactive fission products throughout the watersystem Pitting corrosion is the major form of attack

on the cladding of spent fuel elements during term storage in water basins.28Herein, the focus is onuniform corrosion of cladding

long-Aluminum is a very reactive metal In dry air,

it combines with oxygen to form an adhesive,self-healing Al2O3film that retards further oxidation

at the metal surface Such films are usually quite thin,tens of nanometers, usually described as amorphous.Films formed in moist air and water are much thicker,

1 mm or more The water-formed reaction films oped on aluminum cladding are variously described

devel-as ‘hydrated oxides’ and ‘hydroxides,’ and hydrates,’ and they are generically referred to as

‘oxide-‘oxide films.’ In HPRRs, the films grown on the fuel

cladding may be 20–50 mm thick.29,30The most

com-mon corrosion products28,30 reported on aluminum

cladding are boehmite, a crystalline monohydrated

aluminum oxide, Al2O3H2O, and bayerite, a

crystal-line trihydrated oxide, Al2O33H2O At temperatures

below about 77
C, the boehmite phase is formed

preferentially but may transform to bayerite with

continued immersion At temperatures above77
C

and below 100
C, a pseudoboehmite structure

grows, which may age to other hydrated oxide forms

or retain its pseudoboehmite structure Between100

and400
C, crystalline boehmite will form A

gelati-nous boehmite is the chemical precursor of both of the

crystalline hydroxides.30 The mature hydroxides are

normally white color but other hues have been

reported and may stem from absorption of Fe, Cr, Ni,

or other metal ions leached from steels in the reactors

or in the corrosion test loops

The corrosion film is both the reaction product

and the medium through which the corrosion process

occurs Whether corrosion is governed by ingress of

O and OH ions through the film to the metal surface

or by egress of Al ions to the film/water interface,

it is expected to be diffusion controlled Thus, all else

being equal, an increase in film thickness should

lower the corrosion rate by increasing the diffusion

length, and vice versa Therefore, the corrosion rate

should be parabolic with time and have an

Arrhenius-type dependence on temperature Moreover, ideally,

if all the corroded metal was retained in the

corro-sion film, if the chemical composition and physical

structure of the film were constant throughout

the thickness, and if all of the film was retained

on the metal, the film thickness would be proportional

to the amount of metal corroded Alas, such ideality

does not prevail The corrosion process is

con-founded by a number of interacting factors, including

the following: there is a one-sided heat flux on the

cladding; the corrosion film is a thermal insulator

compared with the Al cladding, so the temperature

of the film will increase with thickness; the film may

not be of uniform composition and/or structure;

the film is soluble to some extent in water, and its

solubility is strongly susceptible to the pH of the water,

which is related to water composition; the film is

subject to erosion in flowing water and to spontaneous

spallation above some uncertain thickness, about

50 mm in one case.31 And to further complicate the

situation, there is wide variation in the ways the

corro-sion tests are conducted and in the parameters that

are measured

The tests may be carried out in open cups, closedautoclaves, vented autoclaves, closed loops, bypassloops, or on used fuel plates Evaporation or con-sumption of the water may require that it will need

to be periodically replaced or its volume readjusted.Except in in-reactor tests and loop test systems withbypass monitoring and adjustment of the water, thechemistry of the water may change substantially dur-ing the test Few corrosion rates for cladding materi-als are measured directly They are usually derivedfrom measurements of the thickness of the corrosionfilm A thickness measurement gives the thickness ofthe film adhering to the substrate at the time of themeasurement It will not include film that has beendissolved and/or eroded away On a spent fuel ele-ment, it may include film that has formed in a storagepool over time periods much longer than it experi-enced in-reactor, and with no forced cooling Duringpreparation for post irradiation examination (PIE) in

a hot cell, the spent element is no longer fullyimmersed It gets hot and has to be periodicallysprayed with water to cool it It has been opined21that the resultant steaming and thermal cycling maycause more corrosion than in-reactor operation andunderwater storage There is no guarantee that thedensity and the composition of the film will be invari-ant through the film thickness On the contrary, mul-tilayer films are more common than not Almost allfilms have a thin, monolithic base in contact with the

Al surface, presumably associated with the ubiquitousair-formed Al2O3film On top of this base, there may

be one to three distinct layers Some films containpores or are cracked Only the films on irradiated fuelelements have been exposed to the effects of neutronirradiation and radiolysis of the water The way inwhich the film thickness is measured may be ques-tionable, too At least six different methods are used,viz.: (1) Scaled measurements by optical or scanningelectron microscopy of metallographically polishedand etched cross-sections of the corroded test piece;(2) micrometer measurements of the thickness of thetest piece before corrosion and after the corrosionproduct is removed by electrolytic polishing until theshiny metal is seen; (3) weight gains of coupons withfilm in place; (4) weight losses of coupons afterremoval of the film; (5) acoustic and eddy currentmeasurements with instruments calibrated againstaccepted standard films; and (6) temperature increasesmeasured with thermocouples attached to the noncor-roding back surface of the test piece during the test,and related to spot film thicknesses measured metal-lographically after the test

A neglected aspect of film measurements is that

almost all of the measurements have been made on

specimens that, deliberately or unavoidably, were

dried at room temperature or at 100
C32before the

measurement was attempted, or before the measuring

instrument was calibrated Until recently, nobody

seems to have determined whether such drying

treat-ments will shrink, spall, crack, or otherwise alter the

bulk film The gelatinous surface layer that precedes

the crystalline corrosion films will almost certainly

be altered during dehydration It is not uncommon

for test coupons to be dried, weighed, and placed

back in the test for the next exposure period, and

so on until the termination of the campaign That

was the method used in one seminal laboratory test

study.33 The first periods in the full exposure

se-quence were the shortest ones, 1 or 2 days, and they

always showed the largest weight gains, usually

60–90% of the total weight gained during the full

duration of the test, which was about 22 days Weight

gains after the first period were linear with time and

were relatively minor That is not parabolic corrosion

behavior The abrupt change in weight gain indicates

that something happened during the first

interrup-tion of the test that set the scene for a sudden switch

from an initial rapid corrosion rate to a subsequent

constant low rate Likely, the first drying treatment

irreversibly altered the structure and permeability of

the hydrated film Recent autoclave tests34on AlFeNi

alloy reinforce that suspicion It was demonstrated

that during a 34-day test, interruptions made every

7 days to remove, dry, weigh, descale, dry, reweigh,

and replace the test piece in the autoclave with

refreshed water for the next exposure period had

serious consequences to the corrosion kinetics

With-out interruptions, the inner and With-outer oxide layers

were twice as thick, the weight gain was 26% higher,

and the amount of metal removed from the substrate

was 23% higher

Some efforts have been made to correlate film

thicknesses with corrosion rates.31–33 Tests made

under controlled conditions in a corrosion loop31

found that the thickness of the boehmite film on

1100, 6061, and X8001 alloys was about 1.4 times

the depth of penetration into the aluminum

regard-less of changes in test parameters that changed the

film thickness, as long as there was no stripping or

spallation of the film Using a literature value for the

density of boehmite, it was estimated that about 70%

of the corroded Al remained in the adherent film and

about 30% was lost to the coolant When spallation

did occur, which was usually above a film thickness of

50 mm, the 1100 and 6061 alloys always showed lized attack of the aluminum under the spalled area,whereas the X8001 alloy showed only uniform attackunder all conditions This correlation was for aclosed, single set of data It should not be consideredrepresentative of all data and situations Other data

loca-by some of the same authors,32 where the principalvariables were temperature and flow rate, showedthat the ratio of corrosion product retained to theweight of metal corroded ranged from a high of 0.54

at a low temperature of 170
C and flow rate of6.1–9.5 m s1 to a low of 0.08 at 290
C and 29 at32.6 m s1 Another source29quotes a retention level

of 50–80% of the oxide on the cladding surface, but itmay be citing Griess et al.31

In general, the ship between film thickness and corrosion rate is notwell established

relation-Film thicknesses from laboratory tests31,35–38play power law growth with exposure time, but thetime exponents, preexponential factors, and activa-tion energies differ from one experimenter to anotherand may be applicable only to the particular set ofdata from which they were determined Nevertheless,the laboratory tests have established that the corro-sion films are sensitive to a number of interactingfactors They include the temperature and surfacecondition of the cladding; the heat flux density onthe cladding; and the temperature, pH, flow rate, andpurity of the water In RRs, water purity is controlled

dis-by filtration and ion exchange systems; it is alsolinked to pH With regard to pH, the films willdissolve if the water is strongly acidic (pH< 4.5)

or strongly alkaline (pH> 8.5); films are most ble in the range 5.0–6.5, the closer to 5.0 the better.The pH of reactor water and spent fuel storage poolwater tends to converge toward the desired range

sta-by carbonic and nitric acids formed from CO2 and

N absorbed from air It can be maintained close

to 5.0 by controlled additions of nitric acid Thestrongest increase in film growth is from increase intemperature, and the controlling temperature is that

at the hydroxide/water interface.31To lesser extents,increased heat flux density and water flow ratewill raise the film growth rate For the alloys 1100,

6061, and X8001, which all corroded alike until ation occurred,31 the rate of oxide formation at aheat flux of 1.58 MW m2was about half of that at3.13–6.31 MW m2, other conditions being the same

spall-At coolant flow rates in the range 7.6–13.7 m s1, therate of accumulation of the corrosion product wasthe same for all three alloys Corrosion rates measured

on the insides of 1100Al production tubes39 were

found to be unchanged by water velocities in the range

0.305–5.58 m s1 Reduced water temperatures will

reduce the film growth rate

Despite differences in strengths, compositions,

and microstructures, the alloys 1100, 6061, and X8001

all seem to have similar corrosion behavior under

similar conditions.31,32 Spalling tends to introduce

local attack in 1100 and 6061 but not in X8001.31

The AlFeNi alloy shows good performance at

temperatures up to 250
C in autoclave tests35 and

in-reactor exposures40at temperatures below 120
C,

but it has not been tested under high heat fluxes

AG3-NET cladding on U3Si2 dispersion fuel plates

undergoing in-reactor tests failed41at a heat flux of

5.5 MW m2 The cladding was swollen and breeched

by a combination of a very thick corrosion film and

subfilm intergranular corrosion Cross-section X-ray

spectroscopy analyses showed that oxygen had

pene-trated intergranularly all the way through the

clad-ding to the meat The corroded cladclad-ding was

interesting in other ways The outer oxide layer was

monolithic and was exceptionally thick, 100 mm

Directly beneath it was a region about 80 mm thick

containing many round 30 mm size pores Below the

porous region, the grain boundaries were enriched in

Mg and oxygen The plates were intended to reach a

temperature of about 180–200
C at the exterior

sur-face of the cladding and 220–240
C in the fuel

Temperatures estimated from the thick corrosion

layers were >300
C for the water/corrosion film

interface and>400
C for the fuel meat The

AG3-NET alloy has a history of intergranular cracking in

beam tubes and other structures in the Reacteur Haut

Flux at the Institut Laue-Langevin in France

Although that cracking occurred at high fluences,

the irradiation temperatures were low Such low

tem-perature intergranular cracking is a sign of pending

weakness in the alloy and does not bode well for

applications at higher temperatures as in fuel

cladding

The influences of neutron flux and radiolysis of

water are unclear These parameters are omnipresent

in RRs and we might imagine them to strongly affect

aqueous corrosion of fuel cladding by damaging

the cladding and its corrosion film and by altering

the activity of the water One researcher42writes that

reports of neutron flux effects on the hydroxide films

are few and there is disagreement; he claims that

the opinion of most (Russian) researchers is that

neutron irradiation decreases, rather than increases,

the corrosion rate Effects of radiolysis are uncertain

According to Golosov,42one Russian authority argues

that radiolysis may either accelerate corrosion byfacilitating cathodic processes or reduce corrosion

by promoting anodic passivation Data from laboratorycorrosion loop tests without radiation fields seem to befairly compatible with data from irradiated fuel ele-ments in terms of oxide thicknesses, compositions,and pH effects There are no outlandish differencesthat would immediately draw attention to radiationeffects At least, none that has been strong enough toinsist that loop tests should be repeated in irradiationfields A similar conclusion was reached for aqueouscorrosion of aluminum process tubes in productionreactors.39 Therefore, irradiation effects must bemodest at worst However, there are some troublingreports that seem to indicate large effects of irradia-tion fields in nonreactor conditions Sindelaret al.43studied 6061Al coupons exposed to moist air at 150and 200
C, with and without exposure to a60Co gsource at 1.8 106

R h1 Weight gains and filmthicknesses were measured The corrosion productwas patches of loosely aggregated, randomly oriented

1 mm size boehmite crystals sitting on a thin lithic base layer, even at 100% relative humiditywhere the product was permanently under a film ofwater g-Irradiation seemed to double the weightgains and increase the film thicknesses by a factor of

mono-10 There was substantial surface blistering of thebase layer, attributed to hydrogen gas The paperprovided no details of the experimental conditions.Enquiries to the authors produced a lengthier publi-cation44 with the missing details Those details castgrave doubt on the conclusions drawn from Sindelar

et al.43

In particular, the experiments with the g-fieldwere made under radically different conditionsthan those without the field Specimens for theg-irradiations were sealed in small stainless steelcans of just 78 ml and each can represented an unin-terrupted test for a given exposure period of 1, 4, 8,and 12 weeks The tests without the g-field were made

in stainless steel autoclaves of volume 37 850 ml for

15 unequal exposure periods totaling about 30 weeks

At the end of each period, the specimens wereremoved, dried, weighed, and replaced in the auto-clave with a new charge of water In light of theeffects of interruptions described in Wintergerst

com-(1010n m2s1) and, separately, g-rays from a 60Co

source at 15 Svh1 (1.6 104

R h1) Corrosion wasdetermined from weight losses It was not stated

whether the specimens were recycled from one period

to the next The neutrons and the g-rays had the same

effects and to the same degree; they promoted

forma-tion of a grayish layer on the specimen surfaces; they

reduced the weight losses by 25–30%; and they almost

eliminated severe pitting corrosion displayed by the

unirradiated specimens None of these three reports

mentioned whether radiation heating was a factor

The laboratory loop tests have verified the

expecta-tion that the corrosion film is a thermal insulator

com-pared with the Al cladding, and they have provided31

a thermal conductivity value of 2.25 W m1K for

boehmite, which is a factor of 70–100 less than Al

However, it is not always ascertained whether a

par-ticular film is boehmite or bayerite or a mix of both

No thermal conductivity value is available for

bayer-ite When insulating films build on the Al cladding

of heat sources like the fuel and long-term heavy

isotope targets, the temperatures of the sources and

their claddings or containers will rise This

tem-perature rise will increase the corrosion rate and

the growth rates and dissolution rates of the corrosion

films In HPRRs, a side effect of an increase in

clad-ding temperature by the adherent corrosion product

is the threat of plate buckling.31As described earlier,

the strengths of the cladding and Al fuel matrix can

be decreased significantly by tens of degrees increases

in temperature, and creep rates will increase If an

insulating corrosion film increases the temperature

gradients between the center thickness of the fuel

plate and the surface of the film, and between the

fuel-loaded portions of the fuel plates and their cooler

frames, the plates may distort If the distortion is not in

phase from one plate to the next, it might perturb the

coolant flow and accelerate the temperature changes

Griesset al.31

envisaged that the insulation provided by

the corrosion-product film might be more of a

limita-tion on the use of aluminum–clad fuel elements in

high flux reactors than is corrosion damageper se and,

in the worst case, may lead to burnout of parts of the

fuel plates Fortunately, that prophecy has not been

fulfilled Serious plate distortion has not been a

wide-spread issue One case of plate distortion is described

in Shaber and Hofman.30Plate buckling found in some

MTR elements12was blamed on new design changes

It is recommended30 that new fuel elements

should be prefilmed with a hydroxide film to reduce

the rate of in-reactor buildup of the corrosion layer

Tests32with 1100, 5154, 6061, and X8001 alloys at flow

rates in the range 6.1–20.4 m s1found that sure of the test pieces to water at 250–300
C for 24 h

preexpo-in an autoclave caused a significant improvement preexpo-incorrosion resistance, but not at higher flow rates TheATR elements are pretreated30by immersing them inwater for 48 h at 180
C and pH 5.0 In the early days

of HFIR operation, the new fuel elements were oftenstored in the reactor pool water for up to 3 monthsbefore being placed into service This immersionresulted in the formation of a rather thick, gelatinous,corrosion product film on the element surfaces.21In

an attempt to avoid that condition, some of the ments were pretreated by boiling them in deionizedwater for 24 h to produce a thin, boehmite film on thesurfaces of the elements before they were placed intoservice When the pretreated elements were used, thecoolant flow rate was found to gradually decreaseand the pressure drop across the elements graduallyincreased during the reactor fuel cycle No significantdamage was caused Changes in coolant flow rateand pressure drop were not observed when the reac-tor was operated with non-pretreated fuel elements.Metallographic examinations of cross-sections ofthe spent fuel plates revealed much thicker corrosionfilms on the pretreated plates Pretreatment of theHFIR fuel elements was discontinued Most RRs donot practice pretreatment of their fuel elements It isproposed here that because of the seemingly largeeffects of dehydration on retarding subsequent filmgrowth as discussed earlier, at least one in-reactortrial should be made of a prefilmed fuel plate with

ele-a dehydrele-ation step or ele-a low temperele-ature bele-akingtreatment added A drying treatment might also beworthwhile for a newly spent fuel element before itenters pond storage

What we really need to learn from corrosion surements and film thickness data is the thickness ofuncorroded Al cladding remaining on the fuel ele-ment at the end of reactor service, and whether thatthickness will be sufficient to continue to seal thespent fuel through further corrosion expected duringcool-down storage in water basins That is, we needreliable corrosion rates pertinent to the particularapplication Corrosion product thickness data areinvaluable in identifying and characterizing themajor factors governing corrosion and the interplaybetween them, but they are meaningless to corrosionrates if a reproducible relationship between filmthickness and corrosion rate is not established Weneed predictability To that end, efforts are underway

mea-to derive predictive models for film thicknesses40,46and corrosion rates.42 These models are in their

infancy Because they lack a large body of consistent

data to draw on, the authors must make many

assump-tions, fittings, and correlations to derive constants,

correction factors, adjustment factors, and

augmenta-tion factors With so much flexibility built into

the prediction equations, it is not surprising that the

authors can find good correlations with selected data

from measurements made on spent fuel cladding

This is not intended as a criticism of the modelers;

it is a reflection of the paucity of input data Reliable

modeling is essential But it needs reliable input data

Data obtained from recycled test coupons should

either be excluded from the models or modeled as a

separate category To be generically applicable, film

thickness models and corrosion rate models should

attempt to merge in a complementary manner

In low power RRs where convective flow is

sufficient to take care of cooling and water quality

is adequately controlled, problems from corrosion

films formed on the aluminum cladding and on other

aluminum components elsewhere in the reactor are

uncommon In HPRRs, the most prominent corrosion

problems were those in the early days of operation that

caused a milky turbidity of the coolant and a white

deposit and increase in surface radioactivity on all

surfaces exposed to the coolant The turbidity was

identified as a fine suspension of boehmite, and the

g-radioactivity was consistent with decay of24

Na, botheffects attributable to corrosion/erosion of the fuel

cladding The turbidity is created by increase in the

cladding temperature due to the warming effects of

the hydroxide film In turn, the temperature of the

coolant in immediate contact with the film is raised

This increases the solubility of aluminum oxide in the

immediate volume of coolant When this small volume

moves on and merges with the cooler bulk coolant,

the solubility falls and much of the dissolved film is

released as a particulate suspension Particles of film

washed directly into the coolant by erosion of the

cladding due to the high coolant flow rate contribute

to the turbidity Since turbidity ensues when the

con-centration of aluminum in the bulk water exceeds the

solubility of the aluminum oxide, turbidity problems

are brought under control by tuning demineralization

treatments to remove dissolved aluminum from the

bulk water and by reducing the degree of dissolution

through adjustments in pH to between 5.1 and 5.4

where aluminum oxides have minimum solubility

In-reactor pitting corrosion and galvanic

corro-sion have not been serious problems Pitting of Al,

which is encouraged by the presence of ions of Cu,

halides, and bicarbonates, is more serious in storage

pools where poorer water chemistry and nearly nant water conditions may exist, but diligent moni-toring and control of water chemistry can mitigatethese concerns Intergranular corrosion has not been

stag-a problem in RRs, but it could become stag-an issue stag-athigh irradiation temperatures as evidenced by theAG3-NET cladding described earlier

Overall, aluminum cladding has given very goodservice in water-cooled RRs and continues to do so.The major variables influencing the corrosion process(es) and corrosion products are fairly well identifiedexcept for effects of irradiation More data from spentfuel elements are needed to guide and refine modelsfor predicting film thicknesses and corrosion rates

5.07.6.1 Basics

As in other metals, irradiation of Al with neutrons orcharged particles introduces lattice vacancies, self-interstitial atoms, and transmutation products thatevolve into radiation damage microstructure, whichcauses swelling, radiation hardening, and loss of duc-tility Radiation damage effects in aluminum differfrom those in most other metals in two respects One

is that the radiation damage is affected strongly by asolid transmutation product, silicon, discussed more

in Section 5.07.6.2.3 The other is that Al is muchmore tolerant of radiation effects than most othermetals At least, it is for irradiations conducted atambient temperatures Neutron irradiation of Al

at temperatures between 25 and 100
C does notinduce detectable radiation hardening until the fastneutron fluence exceeds about 1 1024

n m2, whereas

in Fe and Zr, radiation hardening is detectable atfluences two to three orders of magnitude less thanthat.47Moreover, even when Al is radiation hardened

at 25–100
C, it still retains significant ductility whencompared with considerably reduced ductilities in Feand Zr This delayed display of radiation hardeningexists despite the fact that the number of atomicdisplacements per atom in Al are about twice asmany as in other metals at the same fast fluence,which is brought about by the lower displacementthreshold energy for Al The larger part of Al’s bettertolerance of radiation damage is owed to its lowmelting temperature, which makes its homologoustemperature high compared with those for Fe and Zr

At room temperature, the homologous temperature

of aluminum is 0.32Tm, versus 0.175 for austeniticsteel, 0.17 for ferritic steel, and 0.26 for a-Zr if