MATERIALSRESEARCH

in AECL FALL 1970

lit***

AECL - 3778

The reviews in this document are intended to give the reader a resume ofmaterials research at Atomic Energy of Canada Limited. Those who wish toobtain more detail of the work reported in this document can request copies ofthe literature cited and the published progress reports of the appropriateDivisions, Application stiould be made to: The Scientific Document DistributionOffice, Atomic Energy of Canada Limited, Chalk River, Ontario, Canada .

The cover shows a rectilinear growth spiral of USi in 3 2Specimen by L,C, Berthiaume; metallography by R.J. Dudzik,who was awarded a gold medal by the Canadian Council of theAmerican Society of Metals for this exhibit at the TorontoExposition, 1969.

(Magnification XlOOO approx.)

MATERIALS RESEARCHATOMIC ENERGY OF CANADA LIMITED

The control of corrosion is vitally important in atomic energy, lust as in manyother industries. This issue of "Materials Research in AECL" is therefore devotedto corrosion studies throughout AECL. The first five items, reviewing areas ofapplied technology, indicate the diversity of these studies while the sixth itemillustrates some of the more fundamental work undorUken to elucidate thecontrolling mechanisms.

FALL 1970

CONTENTSPage

1. CORROSION OF IN-CORE MATERIALS FOR WATER-COOLED REACTORS 2

J.E. LeSurfFuels and Materials Division, CRNL

2. CORROSION OF ZIRCONIUM ALLOYS IN ORGANIC COOLANTS 6

J. BoultonApplied Science Division, WNRE

3. LIQUID METAL TECHNOLOGY AT WNRE 10

S. BanerjeeChemistry and Materials Science Division, WNRE

4. HYDROGEN MIGRATION AND SEGREGATION AS A COROLLARY OF CORROSION.. 14

A. SawatzkyChemistry and Materials Science Division, WNRE

5. CORROSION OF OUT-OF-CORE COMPONENTS OF WATER-COOLED REACTORS . . . . 18

J.E. LeSurfFuels and Materials Division, CRNL

6. PLASMA ANODIZATION 22

N . RamasubramanianChemistry and Materials Division, CNHL

CRNL Chalk River Nuclear Laboratories,Chalk River, Ontario

WNRE Whitehall Nuclear Resuarch Establishment,Pinawa, Manitoba

November, 1970

CORROSION OF IN-CORE MATERIALSFOR WATER-COOLED REACTORS

Zirconium alloys are used for most in-core struc-tural components of CAIMDU power reactors. Thechoice of alloy, and its metallurgical condition, dependlargely on the corrosion behaviou..

For reasons of neutron economy and corrosionresistance, most structural components used in thecore of CANDU* reactors are made of zirconium-based alloys. A single fuel channel consists of thefuel cooled by pressurized water in a pressure tubewhich is centred in a calandria tube. A full CANDUreactor core consists of several hundreds of such chan-nels fitted horizontally into a tank of heavy water (thecalandria)(O.

Evidence for the effects of irradiation on the cor-rosion of zirconium alloys has been summarized byCox(2), who also discussed mechanisms for the ir-radiation-induced enhancement.

The fuel sheathsare made of Zircaloy-2 (Zr-1.5wt% Sn-0il5 wt% Fe-0.1 wt% Cr-0.05 wt% Ni) orZircaioy-4 (Zr-1.5 wt%Sn-0.2 wt% Fe-0.1 wt% Cr)(3).In Zircaloy4 the amount of nickel is reduced (0.007%maximum) because nickel has been shown to increasethe propensity of the alloy to absorb hydrogen pro-duced by the corrosion reaction with water. Theoxidation behaviour in 300°C water of the two alloysis virtually identical.

The oxidation of the Zircaioys under reactorradiation is increased when oxygen or oxidising free-radicals are present in the water (Figure 1). To preventthis, hydrogen (or deuterium) gas is maintained in so-lution in the water at a concentration greater than 5cm3 D2/kgD2O. An examination W of the oxidationof fuel sheaths that had been in the NPD reactor wK.tthe dissolved deuterium concentration in the waterwas generally below specified levels, showed patchesof zirconium oxide much thicker than would beexpected from out-reactor studies. The general cor-rosion resistance of Zircaloy is such that a thickpatch' of oxide is only a few microns thick, and it isunwu»l to find a patch as thick as 20 jim (0.02 mm)on a CANDU fuel sheath. Since the nominal sheaththkkivsss is 0.38 mm, the associated hydriding of thesheath is more significant than any small reduction Inwall thickness due to the oxidation per se.

* CANadian Deuterium Uranium

Linear extrapolation of early analyses for deu-terium in the NPD sheaths would have predictedquite high concentrations for fuel with greater than1000 days' exposure in the reactor. Improvements incoolant chemistry control, changes in fuel manu-facturing procedures and conversion from Zircaloy-2to Zircaloy4 have resulted in less than 50 ppm D2(equivalent to 25 ppm H2) being found in NPD orDouglas Point fuel sheathing at the end of normalexposure, which varies from 2 to 4 years dependingon the location of the fuel bundle in the reactor core(Figure 2).

The pressure tubes in NPD and Douglas Pointreactors were made from Zircaloy-2, but developmentof a stronger zirconium alloy (Zr-2.5 wt% Nb) haspermitted thinner tubes to be specified for subsequentreactors. Zr-2.5 wt%Nb is a heat-treatable alloy whichpossesses maximum strength when quenched fromhigh in the (o + 0) phase temperature region (850-900°C), cold-worked to 10-20% reduction in cross-section, then aged at a temperature high in the ci-pher (50O-55O°C) for about 24 hours(5). Thistreatment also produces excellent corrosion resistance(Figure 3) equal to that of the Zircaioys irs-reactor andless sensitive than the Zircaioys to the presence ofoxygen in the irradiated environment^).

TJje1 Boiling JJgtrt Vfeter (PANDU-BLW) reactorat Gentilly has heat-treated Zr-2.5 wt% Nb pressuretubes because oxygen or oxidising species are moredifficult to suppress, in boiling water than in pressur-ized water (Figure 1)('').

The corrosion assistance of cold-worked Zr-2.5wt% Nb (i.e., extruded between 700 and 800°C, thencold-worked to a total reduction in area of 1*5-20%with intermediate anneals at about 700°C) is not asgood as that of heat-treated material (i.e., quenched,cold-worked and aged), However, the corrosionresistance in pressurized water is adequate. Whenresistance to in-reactor creep and ease of fabricationare also considered cold-worked Zr-2.5 wt% 14b ismore attractive for pressure tubes in Pressurized HeavyWater (CANDU-PHW) reactors than either Zircaloy or

1000

Figure 1 Effect of oxygen on the in-reac-tor corrosion of Ziicaloy, 270-300°C.

r 100

1000

- r

200-

160

T " T "

LEOEND

CROSS HATCHED AREA - ZIRCALOY-S - EARLY MANUFACTURE - IRRADIATED WITHHIGH DgO CONCENTRATION IN NPD COOLANT.

SOLID LINE - ZIRCALOY-S - RECENT MANUFACTURE - IRRADIATEDWITH HIGH Dg CONCENTRATION IN NPD COOLANT.

BROKEN LINE - ZIRCALOY-2 AND tfl-FREE ZIRCALOY-S - RECENT MANU-FACTURE - IRRAWATED WITH LOW Eg CONCENTRATIONIN NPD COOLAttf.

X - ZIRCALOY-S - EARLY MANUFACTURE - IRRADIATED WITHLOW Dg CONCENTRATION IN NPD COOLANT.

100 300 30Q 400 300 600 700 MO 900O».VI IN HOT COOLANT

Figure 2 Deuterium pick-up of NPD sheathing (Courtesy of R.D. Page, AECL-2949)

heat-treated Zr-2.5 wt%Nb. The latest CANDU-PHWreactors (Pickering 3 and 4, Bruce J, 2, 3 and 4) willhave cold-worked Zr-2.5 wt% Nb pressure tubes. Thepredicted behaviours of the different alloys have beenconfirmed by examination of full-size pressure tubesafter operation under power reactor conditions.

The calandria tubes for all reactors after NPD(which used aluminum afty tubes)aremade ofZircaloy. Since these generally operate below 8Q°Cand at low pressure (just the head of heavy water inthe calandria tank plus about 0,4 bars pressuri7<ation)they aij already very thin (less than 1.25 mm) andthere is little incentive to change to a stronger alloy.Oxidation or hydriding of the calandria tubes is nota problem at these operating conditions.

The annular space between the pressure tube andcaiandria tube is filled with gas at atmospheric pressureto prevent heat losses from the primary coolant to themoderator. At NPD and Douglas Point the annulus isopen to the reactor vault air, but later reactors havesealed annuli which can be tilled with any gas (e.g.,

CO2 or N2) found suitable to minimise corrosion. Indry air or CO2 the gaseous oxidation on the outsideof the pressure tube is much less than the aqueousoxidation on the inside.

The pressure and calandria tubes are separated byspirally wound spacers (garter springs), one side ofwhich is close to primary coolant temperature andthe other side is close to moderator temperature.For NPD the garter springs were made of Inconel-X, ahigh nickel alloy with correspondingly high neutroncapture cross-section. Replacement with a zirconiumalloy was desirable but the fabrication process for thesprings resulted in zirconium alloys having a structurefavouring radial hydride precipitation which weakensthe material^), Heat-treating Zircaloy after fabrics-tion to improve the structure would impair the mecha-nical properties of the unhydrided spring; heat-treat-ment of Zr-2.5 wt% Nb after spring fabrication wouldimprove the mechanical properties but some doubtremained on the oxidation resistance of the binaryalloy under irradiation in air. A new alloy was devel-

140

AGING6HRSAT 24HRSAT

500 C 500 C

200 300 40C 500DAYS

600 700 800 900 1000

Figure 3 Effect of cold work and aping time on the corrosion of Zr-2.5% Nb exposed to water st 316°C,

oped containing copper (Zr-2.5 wt% Nb-0.5 wt% Cu)which had the heat-treatable characteristics of Zr-2.5wt% Nb but improved conosion resistance in air(9).Douglas Point and subsequent reactors use this alloyfor garter springs. Further research has shown that analloy containing a trace of beryllium (Zr-2.5 wt% Nb-0.02 wt% Be) has even better corrosion resistancethan the copper alloy, and its use as a garter spring iscurrently being considered.

Future developments in CANDU reactors are likelyto require zirconium alloys with higher strength forpressure tubes (possibly with a corrosion resistantcladding if used at higher temperatures) and otherswith good corrosion resistance at high temperatures(up to SOQ°C) for fuel sheathing. Zirconium-niobiumand zirconium-chromium alloys are being investigatedfor use in aqueous coolants at these higher tempera-tures.

J.E. LeSurf

REFERENCES

1. "Douglas Point Nuclear Generating Station"AECL-1596

2. B. Cox "Effects of Irradiation on the Oxidationof Zirconium Alloys in High TemperatureAqueous Environments", J. Nuc. Mat., 28(1968) p 1.

3. S. Kass "The Development of the Zircaloys" inProceedings of the USAEC Symposium on Zir-conium Alloy Development GEAP-4089 (Nov.1962)

4. A.S. Bain, J.E. LeSurf "Oxidation and Hydridingof Zircaloy in NPD Reactor" AECL-3065 (Jan1969)

5. W. Evans, J.E. LeSurf, W.R. Thomas "Heat-treated Zr-2.5 wt%Nb Pressure Tubes for Water-cooled Reactors" AECL-2890 (May 1967).

6. J.E. LeSurf "The Corrosion Behaviour of 2.S NbZirconium Alloy" in ''Applications Related Phe-nomena for Zirconium and its Alloys".ASTM-3TP458 (19(39)

7. J.E. LeSurf P.E.C. Bryant, M.C. Tanner "Useof Ammonia to Suppress Oxygen Productionand Corrosion in Boiling Water Reactors" Cor-rosion 23 (3)pp 57-64 (March 1967).

8. C.E. Ells "Hydride Precipitates in ZirconiumAlloys"J. Nucl. Materials 28 pp 129-151 (1968)

9. C.E. Ells, S.B. Dalgaard, W. Evans, W.R. Thomas"Development of Zirconium-Niobium Alloys".Proc. 3rd Int. Conf. on Peaceful Uses of AtomicEnergy (1964)9 pp91-99. (also as AECL-2022).

CORROSION OF ZIRCONIUM ALLOYSIN

The present state of the knowledge on the corrosionbehaviour of zirconium alloys in organic coolants isreviewed. Provided good coolant chemistry is main-tained, the hydriding of zirconium alloys can be con-trolled, Two alloys, Zr-2,5%IMb and OzhannUe 0.5,give the best performance and are currently used inthe organic cooled, heavy water moderated reactor,WR.1,

Early studies on the corrosion behaviour of variousmaterials in organic coolants suggested that zirconiumalleys would deteriorate quickly due to the absorptionof large amounts of hydVogen. Hydrogen is one ofthe decomposition products of organic coolants at thetemperatures ofintsrest, 400-500°C.

Tests made by AECL and also by Canadian West-inghouse did show; however, that the hydriding ofzirconium alloys could be controlled provided thatcertain precautions were taken. These v,"5re that aminimum concentration of wuter, >50 ppm, must bemaintained in the coolant and that chlorine com-pounds should be rigorously excluded. The presenceof water in the coolant provides a means of maintain-ing ah oxide film on the surface of the zirconium alloyand in this case the behaviour of the material is thesame as one would expect in low pressure steam at thesame temperature. The mechanism for the effect ofchlorine is not clear but it is known that clilorine isabsorbed in the oxide film and in some way destroysits protective properties.

Long term, out-reactor tests in Santowax OM orHB-40, two organic coolants of interest, have beendone at temperatures in the range 365-425°C. Elec-trically heated specimens have also been used to obtaininformation on the behaviour at temperatures of about500°C. In the tests, the chemistry has been carefullycontrolled with water concentration at 80-150 ppm,dissolved hydrogt.i at 60-150 cm3/kg and chlorine<0.5 ppm. The commonly available zirconium alloysZircaloy-2, Zircaloy-4 and cold worked Zr-2.5%Nbhave been examined and in more recent times Ozhen-nite 0.5 (an alloy containing 0.2%Sn, 0.1 %Fe, 0.1 %Niand 0.1%Nb)has been of great interest.

The nature of the organic coolant, whether Santo-wax OM or HB-40 is not' important and in fact,serves only as a vehicle for the water "impurity"which controls the tehayiviutof thtalloys. Corrosionand hydriding behaviourLit 406°C arti comparedbelow and a more complete summary o" the hydridingdata obtained, which is of the most concern, is givenin Figure 1.

ALLOY

Zircaloy-4

Zr-2.5%Nb

Ozhennite

EXPOSURE TIMEdays

490

490

440

CORROSION RATEmg/dm^day

HYPRIDING RATE

0.28

0.27

O.10

14.5

9.3

3.05

10

tilt-

Figure 1 Out-reactor hydriding rates in San- o:towax OM or HB-40. 2

io h

• Zircatoy - 2• Zircaloy - 4*Zr-Z.5%Nb •T Ozhennite 0 5

13 1.4 1.5jopo

".6 1.7

As a result of tests of this kind, Zr-2.S%Nb waschosen as cladding material for the standard fuel ofthe WR-1 organic cooled, heavy water moderatedreactor at WNRE. (At that time the information onOzhennite 0.5 was not available.) This fuel hasbeen used in the reactor since its startup in 1965 withHB-40 coolant and its performance has been vcygood. More recently; as part of the developmentprogram on Ozhennite 0,5 some fuel clad in thisalloy has been introduced. Some early fuel experi-ments conducted in the reactor were clad in Zircaloy-4.

A wire wrapped around the fuel cladding andspot-welded to it at intervals maintains inter-elementspacing and also acts as a hydrogen sink. Hydrogenwltich is absorbed by ths cladding migrates to thecolder wire-wrap; Under these conditions the hydro-gen concentration of the cladding itself never exceedsthe terminal solid solubility for hydrogen at theoperating temperature. Since quite high concen-trations of hydrogen can be tolerated in the wire-wrap,over 8000 ppm, extended lifetimes for the fuel can beachieved even when the cladding operates at tempera-tures of up to 485°C as it has in WR-1, An example

of the hydrogen migration to the wire-wrap from anearly test is shown in Figure 2.

Results of examination of cladding irradiated inWR-1 indicate that the hydriding of Zr-2.5%Nb issomewhat lower in-reactor than outaeactor at thesame temperature. Although the Ozhennite 0.5 alloyis clearly superior in terms of hydriding resistance Inout-reactor tests, preliminary results on irradiatedmaterial suggest that in-reactor the hydriding rates ofOzhennite OiS and Zr-2.5%Nb are about the same.Irradiation is apparently increasing the hydriding rateof Ozhennite 0.5. Results on Zircaloy4 also indicatea slight enhancement of the hydriding rate underirradiation.

A number of experimental zirconium alloy chan-nels that contain the fuel bundles and coolant havebeen installed in WR-1 and all are performing well. Bythe end of this year, all the remaining stainless steelfuel channels in WR-1 will be removed and replacedwith Ozhennite 0,5 fuel channels. These will provideinformation on corrosion, hydriding and creep underirradiation With coolant temperatures up to 4Q?0C.

The study of corrosion and hydriding in organics is

WIRE WRAP

FUEL CLADDING

Figure 2 Hydrogen concentration in the wire wrap of a zirconium clad fuel element.

continuing with the emphasis placed on in-reactor Sawatzky, A. "The Behaviour of Zirconium Alloysperformance. Both Ozhennite 0.5 and Zr-2.5%Nb in Santowax OM Organic Coolantare being considered as fuel cladding and fuel channel at High Temperature" AECL-2118,materials for organic cooled reactors. (i 964.)

Boulton, J. "The Use of Zirconium Alloys inJ.Boulton Otjpxao C o o l a n t s » AECL-2619,

(1966)

For further information see: Boulton J. and Wright, M.G."Ozhennite 0.5 - Its Potential and

Troutner, V.H. "Hydrogen Corrosion in Organic Development" Applications RelatedReactor Coolants" Corrosion, 16, Phenomena in Zirconium and Its2816,(1960) — Alloys, ASTM STP 458, p 325 (1969)

METAL TECHNOLOGY

The small experimental program instituted atWNRE with the object of evaluating liquid metals ashigh temperature thermal reactor coolants fs outlined.

INTRODUCTION

One of the main objectives at WNRE is to studymaterials and coolants for advanced CANDU reactors.It appears that major improvements in the Efficiencyof conversion of thermal energy to electricity and sub-stantially lower capital costs could be achieved byemploying liquid reactor coolants capable of serviceat about 600°C. The liquid metals satisfy this criter-ion and offer the further possibilities of operatingreactor systems at low pressure and with very highpower densities. A preliminary survey of liquid metalsindicated that lithium-7 and the euteetie alloys ofPb-Bi end Pb-Mg would be attractive high temperature,neutron economical coolants for use in thermalreactors. A small program to study these liquid metalsystems has been instituted at WNRE to provideinformation on th« following points:

(i) The compatibility of iton-based container ma-terials and neutron-transparent refractory alloysand ceramics with Pb-Bi, Pb-Mg and Li.

(ii) The effect of impurities and additives in thecoolant on corrosion rates.

(iii) The possibility of controlling the chemistryof the coolant system to minimise corrosion ofmaterials envisaged for use both in and out ofreactor,

<W) The behaviour of small liquid metal pumps,flow meters, heat exchangers, trace heatingsystems and the other components which mayhave to be incorporated into a liquid metalcirculating system.

(v) The transport of radioactive impurities by liquidmetals.

Studies are concurrently being carried out on theeconomic incentives for high temperature liquid-cooledthermal reactors.

EXPERIMENTAL FACILITIES

Experiments have been planned with the objectivespreviously outlined in mind. The program for Pb-Biwas the first to get started at WNRE because of thetechnology available from the ING (Intense NeutronGenerator) program.

In the development work for the liquid-metal-cooled ING target several studies were made on cor-rosion, chemistry, compatibility, mass transfer anddiffusion at-GRNL^and, under AECL contracts, atindustrial and university laboratories. The effect ofdissolved oxygen, magnesium and zirconium on masstransfer was investigated in small loops at Dilworth,Secord and Meagher, fue solubility of various metalsin liquid PtKBi was studied at Queen's University andthe diffusion of corrosion products and self-diffusionwas studied at Simon Fraser University.

At WNRE, the corrosion resistance of materials inPb-Bi has been studied by the following methods:

(0 Capsule tests in a furnace, usually at I000°C.About one hundred specimens have been ex-posed in this manner to Pb-Bi with variousoxygen concentrations in order to qualitativelydetermine their compatibility.

00 Specimens in the form of rods or tubes have

10

been rotated in an isothermal bath of Pb-Biunder an inert gas atmosphere. The Pb-Bi inthis apparatus is normally held ai 800°C in atantalum container. This apparatus has alsobeen utilised for developing a dry mechanicalseal which will operate satisfactorily about 20cm above 800°C Pb-Bi,

(jii) By operating a small non-isothermal pumpedloop - GP-5 (see Figure 1)-between 600°C and500°C. In addition to the corrosion experi-ments at low velocities (0.3 m/s) this loop hasbeen used to demonstrate the possibility ofmaintaining control of coolant chemistry andthe reliability of components.

Several small facilities are in the final stages of con-struction. These include two Pb-Bi loops (RD-7 andRD-8) and a lithium loop (RD-9). RD-7 is a 500 kWPb-Bi loop which will allow specimens to be exposedto Pb-Bi at velocity in the range 2-3 m/s. RD-8 is a13 kW loop constructed with tantalum. It contains apump with a graphite impeller and is designed tooperate at 800°C. RD-8 is especially suitable for cor-rosion testing of ceramics and refractory metals. RD-9is a stainless steel lithium-containing loop with an

electromagnetic pump. It is primarily designed forcorrosion testing of the austenitic stainless steels inliquid lithium. Attempts will also be made to controlthe level of impurities, especially nitrogen, by hottrapping with zirconium or yttrium.

Work with Pb-Mg is planned in a facility which willreplace GP-5 some time earlj' in 1971. The technolo-gies involved in handling Pb-Bi and PbrMg are expectsto be quite similar and the results from trie Pb-Bi studywill be relevant in all phases of the Pb-Mg work.

The contract with Queen's University for studieson the solubilities of various metals in Pb-Bi has beenextended to include Pb-Mg over a wide range of tem-perature, it is expected that this solubility work atQueen's will be followed up with research an cor-rosion inhibition in Pb-Bi and Pb-Mg. The solubilitystud;ss are considered important at this stage becauseone of the main corrosion mechanisms in the heavyliquid metals is mass transfer of container materialsinduced by temperature gradients in the circulatingsystem.

RESULTS

(i) The Pb-Bi capsule tests indicate that POCO gra-phite and Lucalox alumina are very resistant to

Figure 1 S. Banerjee (left) and D.M. Sol-berg discuss sampling techniques for the hightemperature lead-bismuth locp.

11

Bi RICHAREAS

Pb-BI

Figure 2 A photomicrograph of a piece of reactionbonded SiC which has been exposed to 800°C Pb-Bi.Note that Bi appears to have leached out the Si phase.

I I I I I I I I I 1' I I I I I ! I I

• -Zr» -Ojo -r*

Figure 3 Concentration of select-ed impurities in the lead-bismuth ofloon GP-5 from December 1969 toJune 1970.

'-,,

attack by Pb-Bi even when it is contaminatedwith ISO /Lig/g of oxygen. Further testing inRD-8 is planned to verify this conclusion in avelocity field.

(ii) The results of capsule tests with hot pressed andreaction-bonded SiG have not been fully assessedand several more experiments are planned.Reaction-bonded SiC appears to be resistant toPb-Bi if the oxygen concentration is very low(about 20 jig/g) but the silicon phase sufferssome attack at higheir concentrations. Thisis illustrated in Figure 2 which shows a photo-micrograph of a reaction-bonded SiC specimenwhich had been rotated at 200 rpm in 800°CPb-Bi for 162 hours. The average oxygenconcentration during the run was about 70 fig/g.The dark areas show high bismuth concentra-tions when examined by electron beam micro-probe.

(iil) Several corrosion specimens of 2 1/4% Cr • 1%Mo steel (A335.P22) have been exposed to'Pb-Bi at 600°C in GP-5. Small quantities ofzirconium, which appear to inhibit corrosionof ferritic steels, and of magnesium which actsas a deoxidant, have been added to the Pb-Siin GP-5. the 2 1/4% Cr -1% Mo steel specimenssuffered negligible (about 0.5 jttm/yr) corrosionin 3000 hours. In addition the loop is con-structed entirely of 2 1 /4% Cr -1 % Mo steel andthere appear to be no signs of internal corrosion,i.e., the cold leg specimens have maintained thesame weight for 4200 hours and the flow velo-city is exactly the same for the same valve settingas when it was commissioned.

(iv) The concentration of 14 impurities has beenmonitored in GP-5 and some of the more im-portant ones are shown in Figure 3. It has beenpossible to maintain all impurities and additivnsat acceptable levels. The loop components havealso shown a high degree of reliability since nofailures have been experienced in 6036 hours ofvirtually continuous operation. Several shortshutdowns have been necessitated by failures

in the building power and water supply but nomajor difficulty has been experienced in un-freezing the loop.

(v) Some experiments have been done to determinethe manner in which zirconium inhibits cor-rosioniqfrlow appy fteels. Previous work atBfookhavenW suggested that an imperviouslayer of ZrC/ZrN is formed on the surface of thesteel. 2 1/4% Ct H%Mo steel specimens fromGP-5 have been examined by X-ray reflectancediffraction at WNRE, The results show a verythin surface layer on the steel but its patterndoes not correspond to any known ceil structure.It is possible that the layer is so thin thnt thestructure is distorted by the metallic substrate.

FUTURE EXPERIMENTS

In addition to the experiments already committed,serious consideration is being given to:

(i) a small unfuelltd Pb-Bi or Pb-Mg loop to be putin-reactor to study activity transport, corrosionand coolant chemistry;

(ii) a components development loop for Pb-Bi orPb-Mg which will test pumps, heat exchangers,valves, etc.

Some iho'ight is also being given to a series ofexperiments to clarify the mechanism of corrosionInhibition of low alloy steels exposed to heavy liquidmetals.

S. Banerjee

REFERENCE

1. C.J. Klarnut et al, "Material and Fuel Tech-nology for an LMFR", in Progress in NuclearEnergy, Series IV, Vol 2, p 433471 (I960)

13

HYDROGEN MIGRATION ANDSEGREGATION AS A COROLLARYOF CORROSION

The solubility ef hydrogen in zirconium alloys,the precipitation of the hydride phase and its effecton the mechanical properties are discussed. The dif-fusion of hydrogen in zirconium and the resultinghydrogen distribution "ire described for several caseswith different boundary conditions,

Hydrogen is a product of the corrosion of zirco-nium in nuclear reactor coolants such as water ororganics, Some of this hydrogen enters the metal atthe surface from where it diffuses throughout thezirconium. Hydrogen is soluble in zirconium alloysbut When it exceeds a concentration known as theterminal solid solubility (TSS) zirconium nydride(approximate composition ZrH2) is precipitated.Under certain conditions the hydride may embrittlethe zirconium and is therefore undesirable in load-bearing members.

The TSS is relatively independent of alloyingelements but increases with temperature as shown inFigure 1. Its experimental values generally fall on oneof two lines depending on the experimental methodusedO>2). EricksonCO has suggested that the twoterminal solid solubilities correspond to metastableand stable hydride structures; the former reverting tolatter on reheating after cooling.

It is obviously desirable to be able to predict thehydrogen distribution for given conditions. Thisrequires the solution of a differential equation whichfor the one-dimensional case is given by

dc dT7 0)

where J is the hydrogen flux, D the diffusion co-efficient, c the hydrogen concentration, R the gasconstant, T the absolute temperature and Q* theheat of transport. The hydrogen flux is seen to bedependent on both concentration and temperaturegradients.

The steady state solution ofequation(l)(i.e. J=0)is shown to be c = co exp Q*/RT. Q* is positive sothe hydrogen concentration is seen to decrease withincreasing temperature. This solution applies approx-

imately when the rate of hydrogen absorption is veryslow compared with its rate of diffusion in the metaland may help in giving at least a qualitative hydrogendistribution as shown in Figure 2. This shows a micro-graph of a portion of sheath and end plug from anexperimental fuel element held in WR-1, the organic-cooled reactor, for 8 months at an average sheathsurface temperature of460°C. The dark arcs in thesheath indicates massive hydrides. The two diagramshelp to explain the hydriding in this region. Thetemperature profile along the sheath, as representedby the upper curve, drops sharply as the end plug isapproached and then levels off. The lower diagramshows the corresponding terminal solid solubilitycurve and hypothetical hydrogen concentration pro-files for different times where ti < t2 < t3. When theTSS is exceeded hydride is precipitated.

Quantitative solutions of equation (1) have beenobtained for i number of different cases. The circlesin Figure 3 give the experimental hydrogen distri-bution in parts per million by weight (ppm) of a zir-conium cylinder initially uniformly hydrided and thenexposed to a uniform temperature gradient for 34days(4). The solid line is the theoretical hydrogendistribution and the broken line the TSS.

Wilkins and Wasylyshyn(5) did an experiment inwhich they hydrided a 2.5 cm long Zircaloy-2 cylinderto 2000 ppm and friction welded it to an unhydridedcylinder 13 cm long. This was then exposed to thetemperature profile shown by the up^er curve ofFigure 4 for SOI 6 hours. The circles of the lowercurve give the experimental hydrogen distributionand the solid line the theoretical one. The good agree-ment between experiment and theory in Figures 3 and4 suggests that, provided quantities such as tempera-ture distribution and hydriding rates are accuratelyknown, reliable predictions of the hydrogen distri-bution can be made.

14

00 490 400 390 300TEMPERATURE 1*51

Figure 1 The terminal solid solubility ofhydrogen in zirconium as a function of tempera-ture.

TEMPERATURE PC).130 150 SOP 850 300 350 400 450 477

720 -

0.4 OB 12 16 2 0

POSITION ALONG SPECIMEN (cm) .

Figure 3 The hydrogen distribution in Zirca-loy-2 initially hydrided to 130 ppm and exposedto a constant temperature gradient over theregion 130 to 477°C for 34 days.

, H-

I REGION II I

Figure 2 Hydride precipitation in a zirco-nium alloy fuel sheath exposed to organic cool-ant in WR-1 for 8 months at an average surfacetemperature of 460°C. Precipitated zirconiumhydride appears black, (Micrograph courtesy J.Walker.)

DltTAHCC M.0M DM) (Iml

Figure 4 Hydrogen distribution along a Zir-caloy-2 rod after 5016 hours due to an imposedtemperature profile. The a-phase is zirconiumcontaining digralved hydrogen and the >phase iszirconium hydride.

16

SEAM WELD

Zr ALLOYPRESSURE TUBE

I f ALLOY FINS

iODERATOR

NSULATINGGAS ANNULUS

CALANORIATUBE

Figure 5 Typical cross-sactlon of finned pressure tube in reactor

Figure 6 Hydride orientation due to strain in Zr-2.5 wt% Nb-0.5 wi% Cu wire.Hydrogen concentration 160 ppm. Specimens hydrlded, quenched from 825°C. heated6 h. at 535°C, bent to 10% strain, heated to 400°C and slowly cooled. Longitudinalsection.

(Magnification x 50 approx.)

16

The fact that hydrogen in zirconium tends tomigrate towards the colder regions makes it possibleto prevent hydride precipitation in load bearing com-ponents by attaching appendages to serve as hydrogensinks. An example of this is shown in Figure 5. Thefins, being at a lftwer temrjerature.than the pressuretubes to which they aretseam^weldedi servelas^thehydrogen sinks. Another example Is'found in Figure 2of the contribution "CorrosionirofZirconium AUoysin Organic Coolants" by J. Boulton.

The mechanical properties of zirconium can begreatly influenced by the orientation of the precipi-tated hydride platelets which has been shown to bedetermined by factors such as the applied stress anddeformation history(6) as illustrated in Figure *(7.8).Thus fabrication techniques play an important role inimparting the optimum mechanical properties to fuel-sheathing and pressure tubes that are subject tohyddding.

A. Sawatzky

REFERENCES

1. A. Sawntzky, "The Diffusion and Solubility ofHydrogen in the Alpha-Phase of Zircaloy-2", J.Nucl. Mat. 2 (I960) p 62.

2. A. Sawatzky and B.J.S. Wilkins, "Hydrogen Solu-bility in Zirconium Alloys Determined by ThermalDiffusion", J. Nuc. Mat. 22 (1967) p 304.

3. W.H. Erickson, "Hydrogen Solubility in ZirconiumAlloys", Eiectrochem, Tech. 4 (1966) p 205,

4. A, Sawatzky, "Hydrogen in Zircaloy-2: ItsDistribution and Heat of Transport", J. Nucl, Mat.2 (1960) p 235.

5. BJ.S. Wilkins and A. Wasylyshyn, "Diffusion ofHydrogen up a Thermal Gradient", J. Nucl. Mat.29 (1969) p 235,

6. C.E. Ells, "Hydride Precipitates in ZirconiumAlloys", J. Nucl. Mat. 28 (1968) p 129.

7. G.W. Parry, "Strain Induced Directionality ofZirconium Hydride Precipitates in Zirconium-2.5wt% Niobium • 0.5 wt% Copper". Report AECL-1888(1963).

8. S.A. Aldridge, "Hydride Banding in ExperimentalZirconium-2.5 wt% Niobium Pressure Tubes," J,Nucl. Mat. 34 (1970) p 209.

17

CORROSION OFOUT-OF-CORE COMPONENTSOF WATER-COOLED REACTORS

Conventional materials are used for the out-of-corecomponents of a nuclear station. To reduce capitalcosts while improving reliability and maintainabilityspecial attention must be paid to controlling corros-ion.

The out-of-core components of water-cooled powerreactors are similar to those used in a conventionalpower station, but two distinct characteristics of thenuclear station necessitate much more exacting stan-dards of performance from conventional equipmentfor nuclear applications:

1) Some corrosion products suspended or dissolvedin the primary cooling water become activatedon transport through the neutron flux, andsubsequently deposit on out-of-core surfaces.The resulting radiation field makes maintenancework on equipment much more difficult andcostly to perform. This problem is common toall water-cooled reactors, whether they use pres-surized or boiling, heavy or natural water ascaloporteurO "5 ).

2) Heavy water is expensive, providing a strongincentive to prevent leaks, including those dueto corrosion.

Corrosion research on out-of-core components aimsto minimize radiation fields from corrosion products,maintenance work and leakage of heavy water.

In pressurized heavy water reactors, the out-of-core components with the largest surface area are thesteam generators (boilers). The corrosion rates ofsome materials which could be considered for boilertubing are shown in Figure 1. These samples hadbeen exposed in an out-of-flux portion of an in-reactor loop in the NRX reactor to pressurized waterdosed to pHIO with NH4OH and greater than 5cm3 H2/kg H2O. Under these conditions, oxygenand oxidising radiolytic species are completely absentfrom the out-reactor coolant, and the corrosion rateof Monei-400 alloy (nominal composition Ni-28 wt%Cu-2 wt% Fe-2 wt% Ma) is very low and similar to thatof Inconel-600 alloy (nominal composition Ni-15 wt%

Cr-8 wt% Fe-1 wt% Mn). However, the corrosion rateof Monel-400 alloy is increased considerably by thepresence of oxygen in the water, and corrosionproduct release rates approach that of mild steelwhen the oxygen concentration in the water is only0.05-0.1 ppm. Irradiation of the corrosion productsfrom Monel-400 produces Co58 (JI/J = 71.3d;fromNi58), C06O (ri/2 - 5.26y; from Co59 which is anaturally occurring impurity in nickel) and Cu64(n /2 = 12.9h; from Cu63). Douglas Point andPickering reactors use Monel-400 boiler tubes, andcareful control of water chemistry is required at alltimes to avoid high release rates of corrosion products.

Inconel-600, is less sensitive to changes in waterchemistry than Monel-400. However, Inconel is moreexpensive than Monel, and may not be entirely freeof problems. Like other alloys which rely on apassive film for protection (such as austsnitic stainlesssteel) it can suffer localised corrosion when the filmbreaks down through chemical or nu chanical attack(6).Extensive programs are in progress at many labora-tories studying the localised corrosion of Inconel-600and related high nickel alloys.

An alternative approach is to use mild steel as boilertubing. Its corrosion rate is high compared to that ofInconel-600 but the principle activation productFe^9 fop = 45d; from Fe58) ^ | e s s objectionablethan those from the nickel alloys. Ccokni chemistrycontrol is of prime importance with mild steel, but soit is with the other alloys (Monel is affected byoxygen; Inconel is attacked by local high concen-trations of caustic). Thrfi: experimental boilers usingmild steel tubes expose a in the NPD reactor primarycircuit were in good condition when examined afterin-service exposure (Figure 2). Successful develop-ment of mild steel boilers could result in appreciablecapital cost savings for future PHWs. The NPD unitsare now being re-exposed.

18

The major problem with corrosion products in allboiling water reactors has been from dense depositson the fuel sheaths, restricting flow or causing localoverheating. CRNL work(?) has shown that ironoxide deposits are porous and do not interfere withboiling heat transfer unless they are consolidated bysolids such as copper oxide or calcium and magnesiumsalts which deposit from solution. Since the feed-water heaters were the main source of copper oxide,the high-pressure feedwater heaters in neutral boilingwater reactors operating outside Canada have beenchanged from copper-nickel alloy tubes to stainlesssteel tubes. In the Gentilly BLW reactor, we useammonia to suppress oxygen production^) and cor-sequently corrosion of out-reactor materials. Coppercontaining alloys have been rigourously excludedfrom the Gentilly circuit, wlu'ch has stainless steeltubes in the turbine cendenser and first two lowpressure feedwater heater, with steel tubes in therest of the feedtrain, The primary piping is made ofmild steel. This combination of materials is the result

of extensive corrosion testing at AECL establish-ments, and contributes to the low capital cost of theBLW reactor compared with a PHW of similar size.

The variations in corrosion product solubility withtemperature, pH and oxygen concentration controlthe movement of activity around the system. Byimposing controlled changes in these -parameters itshould be possible to move corrosion products fromcertain areas and redeposit them in others where theyare less troublesome. Demonstration of this facilityin loops at CRNL, both in« and out-reactor, suggeststhat 'on-stream cleaning' may be a regular mode ofactivity control in future reactor operations.

Another source of cobalt radioactivity is the hard-facing alloys used on valve seats, pump shafts andother wear-resistant surfaces. These alloys commonlycontain high concentrations of cobalt, and whilsttheir corrosion rate in water is not high, the combina-tion of mechanical damage (erosion, wear) corrosion,and high cobalt content make their use undesirable.Developments at WNREare directed towards replace-

o

o

2.6

2.4

2.2

2.0

1.6

1.6

1.4

1 -2

1.0

0,8

0.6

0.4

0.2

0

2% Ct 1 Mo

STEEL

MILD STEEL

_

410 STAINLESS

304 STAINLESS

MONEL 400

INCQNEL100 200

EXPOSURE DAYS

Figure 1 CORROSION RATES OF METALS IN PRESSURIZED WATER

Samples were exposed out-of-flux in an in-reactor loop to irradiated 27CC water dosedwith ammonia and dissolved hydrogen. Under th^c^di t ionsof inhibition corrosionrates are no greater than with unirradiated water.

Figure 2 Carbon steel tubed model boilerexposed in the NPD reactor for 218 days. Threemodel boilers tubed with carbon steel andthree others with Monet-400 alloy were in-stalled in NPD in parallel with primary andsecondary flows to the main boiler. Expo-sures have since exceeded 700 days.

Figure 3 Region of contact on the faceof a nickel seal ring, showing erosion damage.Leakage of water across the seal face hascaused deep erosion channel? in the nickel

(Magnification x 20)

ment of these alloys with others containing no cobalt.Two examples illustrate how leakage presents

specific problems dependent on the location of theleak:

1) The 410 stainless steel stems on certain valvesfor the Pickering reactor were found to bepitted in storage, as was the mild steel stuffingbox around the valve packing. Chemical analysisat CRNL.of a selection of commercially availablevalve packings and corrosion tests at both Cana-dian Westinghouse Laboratories and OntarioHydro Research Laboratories in several environ-ments and geometrical configurations showedthat some packings were very much more aggres-sive than otlters. Use of the correct valve pack-ing should reduce the incidence of leakage upthe valve stem.

2) The end of each fuel channel in a PHW reactoris closed with a mechanical plug in which anickel seal ring electroplated on a deformabledisc is pressed against a stainless steel end-fitting.The quality of the seal depends upon the defor-mation obtained in the nickel seal ring. Figure 3shows erosion paths across a nickel seal causedby water leakage. Development of softer gradesof nickel by controlling the plating techniqueand annealing the disc after the nickel is appliedhave improved the seal quality.

In combatting corrosion the combined efforts ofseveral Canadian laboratories are directed towardsincreasing the reliability of CANDU reactors, reducingthe requirements for - mechanical maintenance andheavy water upkeep, and reducing the capital cost offuture stations.

J.E. LeSurf

REFERENCES

1.

2.

3.

4.

6.

7.

8.

"Decontamination of Nuclear Reactors andEquipment" Edited by J.A. Ayres, The RonaldPress Company (1970)

"Single-pass Through-reactor Decontaminationof N-reactor Carbon Steel Piping", by W.K.Kratzner, 25th NACE Conference Paper 32(1969)

"Decontamination of N-reactor Stainless SteelSteam Generators", by W.D. Bainard, 25thNACE Conference Paper 33 (1969).

"Decontamination of the Shippingport AtomicPower Plant", by C.S. Abrams and E.A. Salterc-lli,WAPD-299(1966)

"Chemical Cleaning of Boiling Water Reactorand Steam Water System at the Dresden NuclearPower Plant", by M.F. Obrecht et al, presentedat the 21st Annual Water Conference, Eng. Soc.of Western Pennsylvania (I960)

"Sensitivity to Stress Corrosion and Intergran-ular Attack of High Nickel Austenitic Alloys",by H. Coriou, L Grail, C. Mahieu, M. Pelas,Corrosion^22,10,280-290 (1966)

"The Deposition of Corrosion Products inBoiling Water Systems", by D.H. Charlcsworth,presented at the A.I.Ch.E. Annual Meeting,Cleveland, (1969)To be published in the Chemical EngineeringProgress Symposium Series- Nuclear EngineeringSection.

"Use of Ammonia to Suppress Oxygen Pro-duction and Corrosion in Boiling Water Reac-tors", by J,E. LeSurf, P.E.C, Bryant and M,C.Tanner, Corrosion 23,57-64 (1967)

21

PLASMA ANODIZATIONPlasma anodization is a convenient method for

growing oxide films on a number of metals and semi-conductors. Thin film capacitors and metal-oxide-semiconductors (MOS) structures can be preparedentirely in the vacuum system by this method. It isa much cleaner method than the conventional wetanodization and may be used for metals which can notbe anodized in liquid electrolytes. The mechanism ofthe process is as yet incompletely understood.

The term "plasma anodization" describes a tech-nique for growing oxide films on metals arid semi-conductors in an oxygen discharge in the presenceof an applied field. In a dc glow discharge in oxygenthe potential gradient is almost entirely in the cathoderegion and there is very little change in the potentialalong the length of the[negative glow and the positivecolumn. These latter regions in th<.> discharge arefairly conductive and therefore meet the requirementsessential for anodic oxide growth i.e., an oxidizingmedium conductive enough to maintain field strengthsof the order of 106 to 10? V/cm across an insulatingfilm immersed in it. The oxygen discharge can there-for be considered as a gaseous electrolyte.

Plasma or gaseous anodization possesses a numberof advantages over conventional wet anodization inelectrolyte solutions. To mention a few: the problemof contamination or anion incorporation from so-lutions during oxide growth is absent; the density ofpin holes and other flaws usually associated withaqueous anodization is greatly reduced in plasmaanodized films; thin fdm capacitors and MOS devicescan be prepared entirety in the vacuum chamber; andoxides soluble; in many solutions (e.g. GeO2)may beformed easily in the oxygen discharge. Work on theplasma anodization of A!, Nb, Ta, Zr, Ge, Si and GaAshas been published by a number of workers active inthis field.

At. Chalk River our interest in plasma anodizationis directed towards understanding the mechanisminvolved in the oxide growth, its potential use in thepreparation of thin film devices and the developingof techniques for studying thermal oxidation ofmetals and alloys. Various types of discharges (e.g.low pressure dc, induced rf and microwave discharges)and different electrode geometries (e.g. parallel plates,concentric cylinders and a ring cathode parallel to a

plate anode) have been used in these studies.The most frequently used technique has been the

dc low pressure discharge with a parallel plate elec-trode geometry. Such a system is shown schematicallyin Figure 1. The system contains a ring cathode,preferably of the same material as the sample to beanodized and a stainless steel anode. Followingevacuation to about 10-6 Tort, oxygen at a pressureof approximately 5 x 10-2 Jorr is admitted into thebell jar and the glow discharge is struck with a poten-tial of about 1000 volts. The optimum oxygen pres-sure is determined by the spacing between the elec-trodes;0.15 to 0.05 Torr being suitable for spacings inthe range of 15 to 30 cm, respectivdv. The current-voltage characteristics of the plasma are first tracedout using a gold probe; analysis of these data usingLangtnuir's probe theory enables the plasma pro-perties to be determined. The potential of the speci-men is then measured with respect to the anode baseplate. This potential is called the floating or wallpotential of the specimens. At this potential thecurrent drawn by the specimen is zero, since positiveions and negative carriers (electrons and negative ionsif present} rrach the surface in equal numbers. Thisis the potential the specimen will tend to return toduring anodization at a fixed applied voltage.

Anodizing is then carried out by applying either afixed voltage or a constant current density to thespecimen. In the former method a fixed voltage(positive with respect to the specimen's floatingpotential by about S volts) is applied between thesample and the anode base plate. The current decayswith time as in wet anodization; the applied voltagemay be increased in steps to form thicker films. Inthe constant current density method the specimenvoltage relative to the anode base plate becomes morepositive with time at an approximately linear rate

22

when a constant current is passed. On readingthe required potential the current may be allowed todecay as in the fixed voltage mode. The anodizingvoltage is the difference between the specimen voltagefollowing oxidation and its floating potential, bothmeasured relative to the anode base, plate. The growthconstant, the ratio of oxide thickness to the anodizingvoltage is 25% more than that achieved by wet anodi-zation. Though the reasons for this difference are notclear at present, there are indications that the opticaland electrical properties of plasma and wet anodizedfilms are not the same.

Some examples of current-time curves obtainedduring plasma anodization of different metals areshown in Figure 2,

The mechanism of the process is not fully under-stood. The observation that the anodization proceedsin the negative glow as well as in the positive columnof the discharge has given rite to doubts regarding thegenerally held belief that negative oxygen ions areextracted from the discharge during anodization, Inoxygen discharges electron attachment processes lead-ing to 0* ion formation occur in the positive column;thus in the negative glow mainly electrons and Q+ ions

~2KV

Figure 1 PlasmaAmsdizationSystem;a)Stainlesssteslanodebaseplate,b) Rinscathode,c) Specimen, d) Gold probe and e) Quartz shields for electrical insulation.

23

are present. Electron attachment at the specimensurface is a possible process and in this connectionthe importance of understanding the plasma character-istics is evident.

N, Ramasubramanian

For more detailed information, reference may be

made to the following:

1. J.L. Miles and P.H. Smith, J. Electrochem. Soc.110, 1240 (1963)

2. J.F. O'Hanlon, J. Vac. Sci. and Tech, 1 330(1970)

3. N. Ramasubramanian, JF. Electrochem. Soc., H7947 and 950 (1970)

Figure 2 Current-Time Curves Obtain-ed During Plasma Anodization of Tan-talum, Zirconium and Zircalov-2; Ox-ygon Discharge at 700 Volts Carrying20mA; a) Ta-80A, 3.1V and Zr-170A,4.7V b) Z M 9 0 A , 5.0V and Zircaloy-2-160&, 4.5V.

4 -" 0T(Mltmlo)

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