Saimm 201510 oct

VOLUME 115 NO. 10 OCTOBER 2015

28–30 October 2015

our future through science

advanced metals initiative

Lighter,functional alloymaterials for theautomotive and

aerospaceindustries.

Beneficiation fornuclear

materials usedin nuclearreactors.

Beneficiation ofresources and

materialssolutions for thetransportation,

energy andpetrochemical

industries.

LMDN NMDN FMDN

Value-addedPGM products(Autocatalysts,PGM coatings,

etc.).

PMDN

Enabling Platform of Projects and Partnerships


ADVANCED METALS INITIATIVE

The Advanced Metals Initiative (AMI) wasestablished by the Department of Scienceand Technology to facilitate research,development and innovation across theadvanced metals value chain.

GOAL

To target significant export income and newindustries for South Africa by 2020 throughthe country becoming a world leader insustainable metals production and manu-facturing via technological competence andoptimal, sustainable local manufacturing ofvalue-added products, while reducingenvironmental impact.

STRATEGY

The AMI takes an integrated approachacross the entire value chain from resourcedevelopment to metal production and themanufacture of end-products, to achieve itsgoal, through:

• Reducing the energy required toproduce metals by 30%;

• Increasing asset productivity by 30%;

• Developing technologies that canenable new industries for South Africa;and,

• Reducing the full life cycleenvironmental impact of metalsproducts by 50%.

TECHNOLOGY NETWORKS

The AMI’s technology networks include:

• the Light Metals Development Network(LMDN) for titanium and aluminium co-ordinated by the Council for Scientificand Industrial Research (CSIR);

• the Precious Metals DevelopmentNetwork (PMDN) for gold and theplatinum group metals (PGMs), co-ordinated by Mintek;

• the Nuclear Materials DevelopmentNetwork (NMDN) for hafnium,zirconium and monazite co-ordinatedby the Nuclear Energy Corporation ofSouth Africa (Necsa).

• the Ferrous Metals DevelopmentNetwork (FMDN) for ferrous and basemetals, co-ordinated by Mintek.

To lead a global revolution in advancedmetals generating significant exportincome and new industries for South Africawhile reducing environmental impact.

The AMI promotes collaborative researchbetween the science councils, highereducation institutions, and industry.

Human resource development is critical forthe networks to:

- expand the country’s technicalcapacity;

- develop the use of metals in newapplications and more diverseindustries; and

- develop industrial localisation.

LIGHT METALS DEVELOPMENTNETWORK

• Global demand for ultralight,ultrastrong, recyclable metals isgrowing as the world switches to low-emission vehicles, energy-savingdevices and sustainable products.

• Aluminium demand is forecast toincrease by 30% in the near future,while for the emerging industrial lightmetal, titanium, the sky is the limit.

• For its part, South Africa has a maturealuminium industry, which is among thecountry’s top exporters, and one of theworld’s richest titanium resources onwhich to build a new industry.

The LMDN sees South Africa becoming aworld leader in light metals.

The LMDN conducts scientific researchactivities along the entire value chain, fromresource development to primary metalproduction, fabrication, casting, joiningtechnologies and manufactured products.

The aim is to create a globally competitiveintegrated light metals industry, to developsuperior cost-effective technologies andmanufacturing systems, and to reduceenergy use, greenhouse emissions andenvironmental impact.


NUCLEAR MATERIALSDEVELOPMENT NETWORK• The global upsurge in energy demand has

led to a renewed focus on nuclear energyand related nuclear materials likezirconium and hafnium.

• Zirconium is used as cladding in nuclearreactors and zirconium carbide hasapplications in future nuclear reactors.

• South Africa has a vast resource of zirconand supplies 30% of the world market.

The NMDN seeks to beneficiate zirconiumand hafnium across the value chain throughthe preparation and purification ofintermediate metal salts, metalmanufacturing and optimum zirconium-alloys.The NMDN targets alternative, novel,economic and environmentally friendlymanufacturing processes for the metal pairzirconium/hafnium via existing plasma andfluorochemical expertise.

PRECIOUS METALS DEVELOPMENTNETWORK

Precious metals are characterised by theirhigh density and cost, which make themless attractive for use in bulk componentsand more viable in coating/depositiontechnologies (chemical and physical) forvarious applications in which the uniqueproperties of these high-value metals arebeneficial.The PMDN assists South Africa in retainingthe precious metals value matrix through theidentification, research and promotion ofnew technologies and applications tosupport the long-term development of themining industry.

FERROUS METALS DEVELOPMENTNETWORK

The FMDN presents a unique opportunity tosimultaneously add value to severalminerals that South Africa possess in largequantities such as iron, chromium,manganese, vanadium, etc. whileaddressing key material challengesexperienced by strategic sectors of theeconomy such as the transportation, energyand petrochemical industries. The FMDNR&D programmes are done within atripartite collaborative framework involvingindustry, academia and science councils.The broad objectives of the FMDN can thusbe summarised as follows:•

Beneficiation of South Africa’s ferrousresources to stages 3 and 4 undefined.

• Improvement of the country’s capability to

produce high-end ferrous products,especially those that are needed by othercritical sectors of the economy, such aspetrochemical, energy generation,transportation, etc.

• Generation of local know-how (innovation).

• Human Capital Development which willalleviate the shortage in scientific andtechnological qualifications and skills inthese sectors and thereby ensuring thesustainablity and the competitiveness ofthe local industry. This will also improveSA’s attractiveness as an investmentdestination.

• Promotion of local and internationalcollaboration in the field of ferrousmetallurgy.


CONTACT DETAILSLlanley Simpson

Tel: +27 12 843 6436Fax: +27 86 681 0242Cell: +27 83 408 6910

Email: [email protected]

The key focus is on:

• creating new industries;

• supporting existing industries;

• localisation of technology.

supported by

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OFFICE BEARERS AND COUNCIL FOR THE2015/2016 SESSION

Honorary PresidentMike TekePresident, Chamber of Mines of South Africa

Honorary Vice-PresidentsMosebenzi ZwaneMinister of Mineral Resources, South Africa

Rob DaviesMinister of Trade and Industry, South Africa

Naledi PandorMinister of Science and Technology, South Africa

PresidentR.T. Jones

President ElectC. Musingwini

Vice-Presidents

S. NdlovuA.S. Macfarlane

Immediate Past PresidentJ.L. Porter

Honorary TreasurerC. Musingwini

Ordinary Members on Council

Z. Botha G. NjowaV.G. Duke A.G. SmithI.J. Geldenhuys M.H. SolomonM.F. Handley J.D. SteenkampW.C. Joughin M.R. TlalaM. Motuku D. TudorD.D. Munro D.J. van Niekerk

Past Presidents Serving on CouncilN.A. Barcza G.V.R. Landman R.D. Beck J.C. Ngoma J.R. Dixon S.J. Ramokgopa M. Dworzanowski M.H. Rogers F.M.G. Egerton G.L. Smith H.E. James W.H. van Niekerk

Branch ChairmenBotswana L.E. DimbunguDRC S. MalebaJohannesburg I. AshmoleNamibia N.M. NamateNorthern Cape C.A. van WykPretoria P. BredellWestern Cape A. MainzaZambia D. MumaZimbabwe S. NdiyambaZululand C.W. Mienie

Corresponding Members of CouncilAustralia: I.J. Corrans, R.J. Dippenaar, A. Croll,

C. Workman-DaviesAustria: H. WagnerBotswana: S.D. WilliamsUnited Kingdom: J.J.L. Cilliers, N.A. BarczaUSA: J-M.M. Rendu, P.C. Pistorius

The Southern African Institute of Mining and Metallurgy

PAST PRESIDENTS

*Deceased

* W. Bettel (1894–1895)* A.F. Crosse (1895–1896)* W.R. Feldtmann (1896–1897)* C. Butters (1897–1898)* J. Loevy (1898–1899)* J.R. Williams (1899–1903)* S.H. Pearce (1903–1904)* W.A. Caldecott (1904–1905)* W. Cullen (1905–1906)* E.H. Johnson (1906–1907)* J. Yates (1907–1908)* R.G. Bevington (1908–1909)* A. McA. Johnston (1909–1910)* J. Moir (1910–1911)* C.B. Saner (1911–1912)* W.R. Dowling (1912–1913)* A. Richardson (1913–1914)* G.H. Stanley (1914–1915)* J.E. Thomas (1915–1916)* J.A. Wilkinson (1916–1917)* G. Hildick-Smith (1917–1918)* H.S. Meyer (1918–1919)* J. Gray (1919–1920)* J. Chilton (1920–1921)* F. Wartenweiler (1921–1922)* G.A. Watermeyer (1922–1923)* F.W. Watson (1923–1924)* C.J. Gray (1924–1925)* H.A. White (1925–1926)* H.R. Adam (1926–1927)* Sir Robert Kotze (1927–1928)* J.A. Woodburn (1928–1929)* H. Pirow (1929–1930)* J. Henderson (1930–1931)* A. King (1931–1932)* V. Nimmo-Dewar (1932–1933)* P.N. Lategan (1933–1934)* E.C. Ranson (1934–1935)* R.A. Flugge-De-Smidt (1935–

1936)* T.K. Prentice (1936–1937)* R.S.G. Stokes (1937–1938)* P.E. Hall (1938–1939)* E.H.A. Joseph (1939–1940)* J.H. Dobson (1940–1941)* Theo Meyer (1941–1942)* John V. Muller (1942–1943)* C. Biccard Jeppe (1943–1944)* P.J. Louis Bok (1944–1945)* J.T. McIntyre (1945–1946)* M. Falcon (1946–1947)* A. Clemens (1947–1948)* F.G. Hill (1948–1949)* O.A.E. Jackson (1949–1950)* W.E. Gooday (1950–1951)* C.J. Irving (1951–1952)* D.D. Stitt (1952–1953)* M.C.G. Meyer (1953–1954)* L.A. Bushell (1954–1955)* H. Britten (1955–1956)* Wm. Bleloch (1956–1957)

* H. Simon (1957–1958)* M. Barcza (1958–1959)* R.J. Adamson (1959–1960)* W.S. Findlay (1960–1961)

D.G. Maxwell (1961–1962)* J. de V. Lambrechts (1962–1963)* J.F. Reid (1963–1964)* D.M. Jamieson (1964–1965)* H.E. Cross (1965–1966)* D. Gordon Jones (1966–1967)* P. Lambooy (1967–1968)* R.C.J. Goode (1968–1969)* J.K.E. Douglas (1969–1970)* V.C. Robinson (1970–1971)* D.D. Howat (1971–1972)

J.P. Hugo (1972–1973)* P.W.J. van Rensburg (1973–

1974)* R.P. Plewman (1974–1975)

R.E. Robinson (1975–1976)* M.D.G. Salamon (1976–1977)* P.A. Von Wielligh (1977–1978)* M.G. Atmore (1978–1979)* D.A. Viljoen (1979–1980)* P.R. Jochens (1980–1981)

G.Y. Nisbet (1981–1982)A.N. Brown (1982–1983)

* R.P. King (1983–1984)J.D. Austin (1984–1985)H.E. James (1985–1986)H. Wagner (1986–1987)

* B.C. Alberts (1987–1988)C.E. Fivaz (1988–1989)O.K.H. Steffen (1989–1990)

* H.G. Mosenthal (1990–1991)R.D. Beck (1991–1992)J.P. Hoffman (1992–1993)

* H. Scott-Russell (1993–1994)J.A. Cruise (1994–1995)D.A.J. Ross-Watt (1995–1996)N.A. Barcza (1996–1997)R.P. Mohring (1997–1998)J.R. Dixon (1998–1999)M.H. Rogers (1999–2000)L.A. Cramer (2000–2001)

* A.A.B. Douglas (2001–2002)S.J. Ramokgopa (2002-2003)T.R. Stacey (2003–2004)F.M.G. Egerton (2004–2005)W.H. van Niekerk (2005–2006)R.P.H. Willis (2006–2007)R.G.B. Pickering (2007–2008)A.M. Garbers-Craig (2008–2009)J.C. Ngoma (2009–2010)G.V.R. Landman (2010–2011)J.N. van der Merwe (2011–2012)G.L. Smith (2012–2013)M. Dworzanowski (2013–2014)J.L. Porter (2014–2015)

Honorary Legal Advisers

Van Hulsteyns Attorneys

Auditors

Messrs R.H. Kitching

Secretaries

The Southern African Institute of Mining and Metallurgy

Fifth Floor, Chamber of Mines Building

5 Hollard Street, Johannesburg 2001 • P.O. Box 61127, Marshalltown 2107

Telephone (011) 834-1273/7 • Fax (011) 838-5923 or (011) 833-8156

E-mail: [email protected]

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CContentsJournal Commentby J.T. Nel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv–v

President’s Cornerby R.T. Jones . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii

Where should the national R&D in materials science fit into South Africa’s future nuclear power programme?by W.E. Stumpf . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 893

Friction processing as an alternative joining technology for the nuclear industry by D.G. Hattingh, L. von Wielligh, W. Thomas and M.N. James . . . . . . . . . . . . . . . . . . . . . . . . . 903

Neutron- and X-ray radiography/ tomography: non-destructive analytical tools for the characterization of nuclear materialsby F.C. de Beer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 913

Non-destructive characterization of materials and components with neutron and X-ray diffraction methodsby A.M. Venter. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 925

Fluorine: a key enabling element in the nuclear fuel cycle by P.L. Crouse . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 931

Titanium and zirconium metal powder spheroidization by thermal plasma processesby H. Bissett, I.J. van der Walt, J.L. Havenga and J.T. Nel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 937

Plasma technology for the manufacturing of nuclear materials at Necsaby I.J. van der Walt, J.T. Nel and J.L. Havenga . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 943

Synthesis and deposition of silicon carbide nanopowders in a microwave-induced plasma operating at low to atmospheric pressuresby J.H. van Laar, I.J. van der Walt, H. Bissett, G.J. Puts and P.L. Crouse. . . . . . . . . . . . . . . . . . . 949

A redetermination of the structure of tetraethylammonium mer-oxidotrichlorido(thenoyltrifluoroacetylacetonato- 2-O,O')niobate(V)by R. Koen, A. Roodt and H. Visser . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 957

A theoretical approach to the sublimation separation of zirconium and hafnium in the tetrafluoride form by C.J. Postma, H.F. Niemand and P.L. Crouse . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 961

Glow discharge optical emission spectroscopy: a general overview with regard to nuclear materialsby S.J. Lötter, W. Purcell and J.T. Nel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 967

The influence of niobium content on austenite grain growth in microalloyed steels by K.A. Annan, C.W. Siyasiya and W.E. Stumpf. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 973

The influence of thermomechanical processing on the surface quality of an AISI 436 ferritic stainless steelby H.J. Uananisa, C.W. Siyasiya, W.E. Stumpf and M.J. Papo . . . . . . . . . . . . . . . . . . . . . . . . . . . 981

International Advisory Board

R. Dimitrakopoulos, McGill University, CanadaD. Dreisinger, University of British Columbia, CanadaE. Esterhuizen, NIOSH Research Organization, USAH. Mitri, McGill University, CanadaM.J. Nicol, Murdoch University, AustraliaE. Topal, Curtin University, Australia


All papers featured in this edition were presented at the

Nuclear Materials Development Network Conference

28–30 October 2015



Editorial BoardR.D. BeckJ. Beukes

P. den HoedM. Dworzanowski

B. GencM.F. Handley

R.T. JonesW.C. Joughin

J.A. LuckmannC. MusingwiniJ.H. PotgieterR.E. Robinson

T.R. Stacey

Editorial ConsultantD. Tudor

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THE INSTITUTE, AS A BODY, ISNOT RESPONSIBLE FOR THESTATEMENTS AND OPINIONSADVANCED IN ANY OF ITSPUBLICATIONS.Copyright© 1978 by The Southern AfricanInstitute of Mining and Metallurgy. All rightsreserved. Multiple copying of the contents ofthis publication or parts thereof withoutpermission is in breach of copyright, butpermission is hereby given for the copying oftitles and abstracts of papers and names ofauthors. Permission to copy illustrations andshort extracts from the text of individualcontributions is usually given upon writtenapplication to the Institute, provided that thesource (and where appropriate, the copyright)is acknowledged. Apart from any fair dealingfor the purposes of review or criticism underThe Copyright Act no. 98, 1978, Section 12, ofthe Republic of South Africa, a single copy ofan article may be supplied by a library for thepurposes of research or private study. No partof this publication may be reproduced, stored ina retrieval system, or transmitted in any form orby any means without the prior permission ofthe publishers. Multiple copying of thecontents of the publication without permissionis always illegal.

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This special edition of the Journal of the Southern AfricanInstitute of Mining and Metallurgy is dedicated to the NuclearMaterials Development Network (NMDN) of the AdvancedMetals Initiative (AMI) of South Africa’s Department ofScience and Technology (DST). The AMI consists of fournetworks: the Light Metals Development Network (LMDN)which is coordinated by the CSIR, the Precious MetalsDevelopment Network (PMDN) and the Ferrous MetalsDevelopment Network, both coordinated by Mintek, and theNMDN which is coordinated by the South African NuclearEnergy Corporation SOC Ltd (Necsa). The NMDN focuses onthe development and improvement of nuclear materials inorder to enhance the safety of nuclear reactors, thecharacterization of nuclear materials by various analyticaltechniques including nuclear techniques, and the beneficiationof South African minerals across the whole value chain withthe aim to manufacture nuclear components in South Africa inthe future.

The Integrated Resource Plan for Electricity for SouthAfrica (generally referred to as the IRP2010) identifiednuclear power as an important contributor to the totalelectricity mix of the future. It is planned to add 9.6 GW ofelectricity generated by nuclear power to the national grid by2030. This implies that six to eight new nuclear power plants(NPPs) need to be built in South Africa. Nuclear technologylocalization and local content will be very important factors inthe realization of the nuclear new-build programme.

Zirconium alloys are used as cladding material of thenuclear fuel in NPPs. All the zirconium metal that is used inthis application is extracted from the mineral zircon, of whichSouth Africa is the second-largest producer in the world. TheNMDN has developed a unique plasma and fluoridebeneficiation process to manufacture nuclear-grade zirconiummetal powder from locally mined zircon. Hafnium is alwaysassociated with zirconium in nature, but it has to be separatedfrom zirconium due to its high thermal neutron absorptioncross-section. This property of hafnium is, however, beingexploited by the nuclear industry in applications whereabsorbance of neutrons is required. Apart from its nuclearapplications, hafnium is also used in electronics, optics, high-temperature-resistant ceramic materials, and in the aerospaceindustry. The NMDN has consequently also embarked on thedevelopment of hafnium products for nuclear and non-nuclearapplications.

The unfortunate Fukushima incident is still fresh in theminds of everybody. High temperatures and rapid oxidation ofthe zirconium cladding material in contact with the coolantwater led to the generation of hydrogen gas and thespectacular hydrogen explosions that were witnessed byeveryone on television. Since then, there has been a majordrive in the nuclear industry to make zirconium alloys moreresistant to oxidation at high temperatures. The application ofzirconium carbide and silicon carbide layers on zirconiumnuclear fuel tubes is but one of the research programmes thatthe NMDN is pursuing in this regard.

Thorium is envisaged to be an important nuclear fuel forthe future Generation IV high-temperature gas-cooled nuclearreactors or molten salt nuclear reactors. The mineral monazitecontains a significant concentration of thorium along withvaluable rare earth elements (REEs) such as neodymium(Nd), cerium (Ce), praseodymium (Pr) and yttrium (Y). Nd,for example, is a crucial ingredient in permanent magnets thatare used in wind turbines to generate electricity. Monazite is aby-product from the heavy mineral sand industry in SouthAfrica, and is also found in hard-rock orebodies, for exampleat Steenkampskraal and Zandkopsdrift in the Western CapeProvince. Monazite, a rare earth phosphate, is extremelychemically inert. Conventional chemical extraction proceduresare very harsh, environmentally unfriendly, and produceradioactive waste. The NMDN is investigating new plasmafluoride beneficiation methods to recover the REEs and toseparate the contained thorium and uranium values.

Human capital development forms a fundamental pillar ofthe AMI. In 2015, the NMDN has 28 postgraduate studentsenrolled at various universities in South Africa to contributeto the building of a sound local knowledge base in nucleartechnology. The participation of these universities in theNMDN projects is highly appreciated. The postgraduatestudents of the NMDN express their extreme gratitude towardsthe DST for creating this programme as a launching platformfor their careers.

The papers that are published in this special journal of theJournal are a selection from the papers that will be presentedat the 2015 academic peer-reviewed process of the AMI, ofwhich this conference is the grand finale for postgraduatestudents. This year it is being hosted by the NMDN and willtake place at the Nelson Mandela Metropolitan University inPort Elizabeth from 28–30 October 2015. It is noteworthy thatthis is the first time in the history of the AMI that the specialedition of the Journal that is dedicated to the annual AMIconference will be published ahead of the conference. In thisregard the Chairman thanks the organizing committee of theconference, the referees, and the SAIMM publications team fortheir hard work towards achieving this historic goal.

The NMDN and Necsa express their sincere gratitudetowards the Honourable Minister Naledi Pandor and theHonourable Director General of the Department of Science andTechnology, Dr Phil Mjwara, for their continuing support ofthe AMI and the NMDN over many years.

I hope that everyone will find the conference an enjoyableevent and that this special edition of the SAIMM Journal willconstitute a major contribution to the nuclear science andengineering fraternity of South Africa.

J.T. NelNMDN Coordinator and Conference Chairman

Journal Comment

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It has been just over fifty years since the concept of the ‘global village’was introduced by Marshall McLuhan in 1964, yet it remains relevant toour everyday experiences. Modern communication technologies have

seemingly shrunk the world even further since then. Ray Tomlinson, in 1971, was the first person to send mail from one computer to another over

a network (and also initiated the practice of using the @ sign to direct the networked electronic mail messageto a particular user at a particular computer). In 1997, e-mail volume overtook postal mail volume, as moreand more people recognized the convenience of this almost immediate, yet still asynchronous, mode ofcommunication. That same year, 1997, saw the registration of the google.com domain name, and searching forinformation was transformed for ever, as people came to rely on the Google search engine to navigate theWorld Wide Web (invented by Tim Berners-Lee in 1989). It seems hard to believe that it’s been only about 20years since the mass popularization of the World Wide Web (arguably one of the world’s greatest inventionssince the wheel). Nowadays, we can almost instantly read about (or watch) events happening anywhere in theworld.

The interconnectedness of today’s world has led to a direct link between the slowing down of the rate ofgrowth in urbanization in China and the state of the economy in Rustenburg, for example. There is also muchmobility of people between countries and continents. Many engineers trained in South Africa work inAustralia, and many Australian engineers work in the USA, and so on.

The SAIMM maintains strong links with similar societies in other countries. In November 2011, aninaugural meeting was held in London between several leading international mining and metallurgicalsocieties – AusIMM (Australasian Institute of Mining and Metallurgy), CIM (Canadian Institute of Mining,Metallurgy and Petroleum), IOM3 (Institute of Materials, Minerals and Mining), SAIMM (Southern AfricanInstitute of Mining and Metallurgy), and SME (Society for Mining, Metallurgy and Exploration). The meetingwas intended to foster cooperation between the various organizations, to discuss opportunities for improvingand sharing benefits to members, and to benchmark the institutions against each other. Further meetingsbetween these societies were held in September 2012 in Las Vegas (SME), in February 2013 in Denver (SME),in February 2014 in Cape Town (SAIMM), in October 2014 in Vancouver (CIM), and in March 2015 in HongKong (AusIMM). Agreements have been signed between these societies, resulting in the formation of what isknown as the Global Mineral Professionals Alliance (GMPA). Discussions were held about the state of themining industry in the various countries, as well as the structure and strategies of the societies represented.There was broad agreement that the societies would offer services to each other's members at member rates.This is a significant benefit to SAIMM members, as they can attend international conferences held by AusIMM,CIM, IOM3, and SME at the same cost as members of those societies. Calendars of events are circulatedbetween the organizations to coordinate major events and minimize clashes.

The flagship project of the GMPA is OneMine.org, a database of over 100 000 technical papers that isfreely available to be used by the members of GMPA societies. Support of this project – both financially and bysharing technical papers – is a necessary precondition for a society to belong to the GMPA. Participatingsocieties also agree to publicize their GMPA affiliation on their websites, and to share meeting calendars andinformation about each other’s international events. Representatives of each society meet once a year toexchange information, to maintain a common set of standards for technical events, and to look for furtherways to increase member benefits with reciprocal arrangements. This also provides an opportunity to shareapproaches and resources to deal with global problems shared by all.

Until asteroid mining becomes accepted practice, we will have to settle for this global approach on PlanetEarth.

R.T. JonesPresident, SAIMM

President’s

Corner

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IntroductionSince 2008, and more particularly in2014/2015, South Africa has woken up to thefact that significant steps need to be taken toensure sufficient electricity generating capacityfor the future, even beyond the coal-firedstations at Medupi and Kusile currently underconstruction. It is, therefore, encouraging tosee some active large-scale wind farms in theEastern Cape near Jeffrey’s Bay, in the Cougaarea, and others in the Western Cape alreadyin operation. In addition, many solar energyprojects are also progressing from the smalllocalized scale to larger programmes in theNorthern Cape, which may contribute somecapacity on a national basis. South Africaneeds to tap into its renewable resources ofwind and solar much more, but will theseprojects solve the country’s long-termindustrial needs? Unfortunately not. Onecannot run mines and trains on solar cells.Industry needs reliable baseload capacity, andwith very limited easily accessible hydro-capacity this leaves really only coal, possiblynatural gas, and nuclear power as options.South Africa’s current over-reliance of about90% on coal–fired power, however, places it inan internationally vulnerable position, anddiversification into a more equitable energymix should be a national priority for themedium to long term. South Africa cannotsimply ignore the mounting evidence of

significant climate change confronting thehuman race, as the IPCC cautioned in 2013,and will have to adjust its future energyreliance to a more balanced combination ofsources.

After many years of internationalconferences, meetings, and working groupsessions, the world is no nearer to finding anequitable and binding international agreementon measures to curb climate change. It is,therefore, highly unlikely that the moreacceptable low-emission scenarios such as theRCP2.6 (Figure 1) are realistic, and currenttrends appear to indicate that the world isfacing a more pessimistic climate changefuture, such as the RCP8.5 scenario.

Does this mean that South Africa will needto completely phase out coal-fired power in themedium to long term? No, that would beimpossible, and even irresponsible, but it doesmean that a future energy mix of about 50%coal-fired, 25% nuclear-based, and 10%imported gas-fired power, with the remaining15% consisting of renewable energy sources,would be a typical future to plan for. Such ascenario would constitute a baseload capacityof about 80–85% with the remaindercomprising renewable energy sources, mainlywind and solar.

Such a turnaround from a very high to amore reasonable dependence on coal plus astill limited nuclear dependence will placeheavy demands on South Africa’s technicalexpertise to select, evaluate, and later tosupply the materials that are ‘fit for purpose’in the planned nuclear power programme.

Broad classification of nuclear materialsin a pressurized water reactor Although a modern nuclear power reactor suchas a pressurized water reactor (PWR) consists

Where should the national R&D inmaterials science fit into South Africa’sfuture nuclear power programme?by W.E. Stumpf*

SynopsisSouth Africa recently announced a resurgence in its commercial nuclearpower programme. The implications for the development of the necessaryhigh-level manpower within South Africa’s tertiary educational systemand its national research and development (R&D) capacity in materialsscience and engineering, as well as in other engineering disciplines, areplaced into perspective. An organized national process of developing thismanpower by moving away from the previously high-risk and costly ’largeprogrammes’ to rather a selection of ‘small and better’ research projectsand a redefinition of what constitutes ‘nuclear materials’ are proposed asparts of this strategy.

Keywordsmaterials science, pressurised water reactor, zirconium based claddingmaterials, uranium enrichment, steam generator.

* Professor in Physical Metallurgy, University ofPretoria.

© The Southern African Institute of Mining andMetallurgy, 2015. ISSN 2225-6253. Paper receivedAug. 2015 and revised paper received Aug. 2015.

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http://dx.doi.org/10.17159/2411-9717/2015/v115n10a1

Where should the national R&D in materials science fit into South Africa’s future nuclear power programme?

in essence of the same main components as those for a coal-fired plant, i.e. a heat source, a steam generating system, anda steam-driven turbine/generator combination, the operatingand safety requirements make a typical PWR a far morecomplex system that requires specialized materials. Figure 2shows a broad overview of the typical materials currently inuse in a modern PWR. Note the wide range, from low-alloysteel to more sophisticated ferrous and stainless steel alloys,from nickel-based creep-resistant alloys to corrosion-resistant titanium condenser tubes, from zirconium-based

fuel cladding to boron-based control rod materials, fromelectrically conductive copper to cathodically protected tubesheet and, last but not least, oxide fuel pellets. Productionand manufacturing processes for these materials range fromcast components to wrought and welded tubes and sheet,from passivated surfaces to corrosion-resistant weld-cladding, from sophisticated to more conventional heattreatments, from high purity to standard material purities,from solid to porous sintered items, and so on.

Such a wide range of materials of construction poses atremendous challenge to South Africa’s materials engineersand scientists if they wish to grow into and activelyparticipate in an expanding nuclear power programme. Tosimply sit back while all of the know-how is imported, evenin the long term, is not an option. On the other hand, toconsider actively mastering the know-how for all of the abovematerials is also unrealistic. Some hard choices, therefore,need to be taken to rather focus on those areas wheremaximum benefit can be gained within the limited researchresources at the country’s disposal.

The need for materials science in PWR technologyIn assessing the broad research focus areas of South Africa’sscience, engineering, and technology (SET) sector inpreparation for a future resurgence of nuclear power, oneneeds to firstly recognize the somewhat onerous process ofdevelopment, testing, evaluation, and safety assessmentbefore adoption, as described so elegantly by Hoeffelner(2011) (Figure 3).

The entire cycle of materials development, fromconceptual definition until final introduction in practice, canin essence be separated into two main focus areas: firstly, the

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894 OCTOBER 2015 VOLUME 115 The Journal of The Southern African Institute of Mining and Metallurgy

Figure 2 – Typical structural materials in use in a modern PWR (Zinkle and Was, 2013)

Figure 1 – Estimated IPCC global surface temperature changes forvarious models of climate control through curbing CO2 emissions(IPCC, 2013)

upper issues of technology, and secondly, the lower issues ofdesign and safety assessment. The two focus areas go hand-in-hand, and South Africa’s endeavours in nucleartechnology over the past three or four decades have taughtsome hard lessons of the consequences of focusing primarilyonly on the development of the technology, without planningfor the resources to bring the technology into safe, reliable,and cost-effective commercial fruition, which placed theentire process at risk of termination. This was a classictechnology push instead of a technology pull approach.

The demands of the entire development cycle as depictedin Figure 3 can partly be recognized in the unfortunateterminations of the uranium enrichment programmes (boththe Vortex and the Molecular Laser Isotope Separation (MLIS)systems) and the development of high-temperature gas-cooled reactor technology in the pebble bed modular reactor(PBMR) programme. In all of these programmes thetechnology developed by South Africa’s scientists andengineers was on an equal level with international norms,leading even to international participation in the MLISprogram. During a visit in the early 1990s to the Vortexuranium enrichment Z-plant for low-level enrichment, agroup of international engineers from a leading country inthe area of enrichment just shook their heads andcommented: ‘we would never have been able to design andbuild such a plant’.

Why, then, did all three of these programmes falter in theend? The exact reasons are, of course, different in each case,but overall all three really faltered due to one common factor:the lack of sustainability of resources needed to take eachone through the entire life-cycle of development shown inFigure 3, and then into full commercial viability. It was assimple as that! South Africa must recognize that it is arelatively small country with very limited resources, andfurthermore, new technology always carries a high risk offailure even if the R&D is on a par with international norms.

The crucial question regarding South Africa’s futurenuclear power programme, is therefore a fundamental one:

‘Should South Africa’s materials scientists and engineersattempt to be technology leaders as in the past, or shouldwe rather aim to be technology followers, but in theprocess improve on an incremental basis what othershave already done, i.e. a ‘small and better’ focus?The answer to this basic question can be sought inter alia

in the path taken by Japan after the end of the Second WorldWar, when a shattered country with limited own resourceshad to ’climb out of its ashes’ by emulating what others haddone, but doing it incrementally better. Within a decade ortwo, Japan had become internationally renowned for its high-quality cameras, binoculars, television sets, and many otherelectronic and engineering goods.

There is a lesson to be learnt here. South Africa shouldavoid ‘large and new’ high-risk technology programmes andfocus rather on the ‘small and better’ technical areas that willincrementally draw South African R&D, together with localindustry, into growing participation in the future nuclearpower programme.

A second strong argument for ‘small and better’ lies inthe inherent safety and performance guarantees that have tobe provided by the reactor vendor. Local participation in thesupply of key components, particularly those associateddirectly with the so-called ‘nuclear island’, will most likely bevery limited for many years to come, at least until SouthAfrica’s industrial base has reached a level of sophisticationequal to those of the reactor vendor countries. Does this,therefore, mean that no local R&D resources should befocused on these materials? No, not at all!

For purposes of design evaluation, operationaloptimization, and safety evaluation as depicted in the lowerhalf of Figure 3, decision-makers need to have a clearunderstanding of the limits of structural materials and afeeling for the behaviour of these materials under severeoperational conditions, which often requires much more than‘literature knowledge’. This route will be called‘understanding better’.

In designing a roadmap for South Africa’s materials R&Dcapacity in the future nuclear power programme, one cantherefore once more return to the model of Hoeffelner inFigure 3 and identify two main areas that need to beaddressed:

➤ Research aimed at the incremental development orimprovement of the upstream technology of materialsaimed at future supply into a growing nuclear powergenerating capacity, i.e. the ‘small and better’ route

➤ Developing an understanding of the positive andnegative limits of those materials for design, lifeassessment, and safety evaluation purposes, i.e. the‘understanding better’ route.

Each of these will be explored in some detail, withspecific examples from the nuclear industry.

Research focus area: technology of nuclear materials

Zirconium-based cladding materials A very visible illustration of the resources required for a‘large and new’ programme is provided by the developmentof improved zirconium-based cladding materials for PWRtechnology.



Figure 3 – The process of new materials development, testing andevaluation, design, and safety assessment in nuclear materials beforeacceptance and introduction (Hoeffelner, 2011)


Firstly, where would one have to focus in selecting a newdevelopment area? The leading nuclear countries of the worldare committing very substantial human and researchresources to the development of new alloys.

The difficulty in choosing a new alloy towards whichSouth Africa could make a meaningful contribution isimmediately apparent. The development of any new alloyinvolves a wide range of difficult technological challenges,which often require a compromise in final properties. There istherefore an inherent risk in any choice of R&D on advancedcladding materials.

Finally, the new alloy needs to be proven by means ofnumerous costly and lengthy in-reactor tests to evaluate itssafety performance.

Considering all of the above, it is therefore quite clear thatfor South Africa to embark on such a ‘large and new’programme of cladding alloy development would be illogical,particularly since a country such as France has seeminglymore than 100 engineers and scientists working on such aprogramme alone. Does this mean that South Africa shouldwithdraw completely from any zirconium-based research?The answer is, clearly, ‘no!’

South Africa, together with Australia, supplies most ofthe world’s zirconium-based minerals, and herein lies aparticular opportunity in the ‘small and better’ route. Thecurrent three major processes followed by companies in theUSA, India, and France all start off with zircon (ZrO2.SiO2),which always contains small amounts of hafnium (Hf)substituted for Zr, and use various refining processes toarrive at ZrCl4 followed by reduction to hafnium-freezirconium metal in the magnesium-based Kroll process. All

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Figure 4 – The development of potential advanced zirconium-based cladding materials (CEA-INSTN, 2008)

Figure 5 – Technical challenges to be addressed in the development ofany advanced zirconium-based cladding materials (CEA-INSTN, 2008)

Figure 6 – In-reactor oxide thickness measurements of Zircaloy-4 (redand yellow data points) and the new Zr-Nb alloy M5 (bottom black andgreen data points) (CEA-INSTN, 2008)

three processes are batch operations, and all of them haveenvironmental and safety implications.

South Africa has the zirconium-based mineral resources,and is recognized internationally for its pyrometallurgicalprocess technology.

Here is a prime example of a ‘small and better’ strategy to‘re-invent’ the whole, or only the upstream steps, of thecurrent zirconium production routes with better technology.

The incremental improvement in the three somewhatsimilar production routes for nuclear-grade zirconiumcertainly falls within the capabilities of South Africanscientists and engineers, and can benefit both the countryand the world’s nuclear industry in the long term. Necsa,with its internationally competitive capability in fluorine andhigh-temperature plasma technology, is uniquely placed toaccept such a challenge.

Uranium enrichmentIn any discussion of South Africa’s future nuclearprogramme, the question of uranium enrichment willinevitably arise. Should South Africa once more embark onsuch a ‘large and new’ venture for its future nuclearprogramme? This question is probably somewhat easier toanswer today than some decades ago. This is due to thefollowing reasons.

➤ South Africa’s uranium resources are found primarilyin gold-bearing ores. When South Africa was one ofthe world’s major gold producers, it simply made senseto also beneficiate the uranium to so-called‘yellowcake’ as a by-product. South Africa has sincelost its place in the ranking of the top few goldproducers. Furthermore, the fall in the price ofyellowcake has made it quite uneconomical to extractthe uranium from the gold recovery processes. SouthAfrica currently (2013) accounts for only 0.9% ofworld uranium production (World Nuclear Association,

2013), and is superseded by African countries such asNiger (7.6%), Namibia (7.3%), and even Malawi(1.9%). Because of this unfortunate co-existence ofgold and uranium, South Africa had only 5.5% of theworld’s ‘Reasonably Assured Reserves’ (RAR, i.e.recoverable at a cost of less than US$130 per kilogramU), in 2009 (TradeTech, 2010), a decline from 6.5% ofhistorical production up to 2008

➤ Countries to the north of South Africa, however, doproduce uranium as a primary product and thesecountries contain some noteworthy reserves. CouldSouth Africa serve as a regional uranium enrichmentcentre for Africa? A select working group that wastasked by the Director-General of the IAEA in Viennain 2007, and of which the author was a member,defined the boundaries of such regional nuclear fuelcentres with one of the main criteria being ‘majoritymultinational control’ from outside the region, mostlikely with the participation of one or more of the fivepermanent members of the UN Security Council. Tobring such a possibility into fruition, however, isfraught with a number of difficult questions, bothpolitically and technically:• Politically, nuclear non-proliferation and uranium

enrichment will always be very sensitive topics. Thisis not made easier by concerns about the realintentions of Iran and North Korea (both signatoriesof the Non-Proliferation Treaty or NPT) and those stilloutside the NPT, notably India, Pakistan, and Israel.With memories of the Cold War in the previouscentury still fresh in people’s minds and the currentinstabilities in many parts of the world, it is to beexpected that the five permanent members of the UNSecurity Council, the so-called ‘haves of nuclearweapons’ within the NPT, would strongly resist anymeasures to spread uranium enrichment technology.


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Figure 7 – Process flow sheets for the production of nuclear-grade zirconium of (left) Wah Chang in the USA, (middle) NFC of India, and (right) Cesuz ofFrance (CEA-INSTN, 2008)


• On a regional basis there is, of course thedeclaration of Africa as a Nuclear Weapon-freeZone, the so-called ‘Pelindaba Treaty’ originallyestablished in 1996 and finally ratified by the 28thMember State of the African Union on 15 July 2009.This could be offered as a guarantee of the peacefulintentions of such a regional uranium enrichmentcentre situated in South Africa. Internationally,however, it is to be expected that serious questionswill be raised as to whether such a regional treaty is‘watertight’ against proliferation.

• The next stumbling block arises from the question:‘What happens to the depleted UF6 from inter-African ‘imported’ uranium after enrichment?’Technically, natural uranium imported into SouthAfrica from any other country would be a tradeablecommodity, but not the depleted UF6, which now ismost likely labelled as ‘hazardous waste’. As asignatory to the Bamako Convention to ‘Control theBan of Imports into Africa and the Control ofTransboundary Movement of Hazardous Wasteswithin Africa’, which was signed on 30 January1991 in Bamako, Mali and came into force on 10March 1999, South Africa must abide by itsundertaking not to allow the movement of anyhazardous waste materials across internationalborders within Africa. South Africa, in its role ofsuch a regional nuclear fuel centre, would thereforehave to remain the host of all of the depleteduranium after the feed material had been convertedto UF6 and then enriched to low enriched uranium(LEU).

➤ Finally, should all of the above political questions besomehow resolved, the technical question of ‘interna-tional cooperation’ versus ‘go-it-alone’ in selecting thetechnology for enrichment, needs to be understood.Here the lessons from the past should once again be

recognized. Yes, South Africa’s SET capacity could, inprinciple, once more go down that road with centrifugetechnology, but it will always be a high-risk strategywith the very real possibility of another failure oncommercial grounds. The low-risk strategy of using thecurrently most competitive uranium enrichmentcentrifuge technology of URENCO under a licenceagreement, as France has done for some years, needsto be considered.

Considering all of the above, it seems that a possibleSouth African re-entry into the area of uranium enrichmentwould be faced with almost insurmountable hurdles, andwould need to be very carefully analysed before it is evenconsidered.

The ‘small and better’ approach for South AfricanSET in materials research

Recognize development trends in commercial powerreactor trends The development of commercial power reactor technology hasprogressed a long way towards the ‘Generation IV’ (GEN IV)light water reactor (LWR), with enhanced economy andsafety, minimal waste generation, and, last but not least,increased proof against proliferation. South Africa is one ofthe participating countries in GEN IV and is, therefore, wellplaced to use this association in planning its nuclear powerreactor programme for the future.

Note the key targets of better economy, enhanced safety,less waste, and proliferation resistance. In each of theseareas, South Africa’s SET capacity can certainly make a‘smaller and better’ contribution.

Should South Africa focus its nuclear materialsresearch on in-core or ex-core components?In considering this question for LWR technology, oneinevitably focuses on UO2 and zirconium-based cladding

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Figure 8 – -The Generation IV LWR development path (US Department of Energy, 2002)

material, but any modifications, even those that are onlymodest, without ultimate proof of performance underirradiation conditions, would be almost a futile exercise(Figure 6). Is the commercial transfer of in-core technologyfrom a reactor vendor into South Africa an option? Hereagain, lessons from the past need to be recognized, as in thetransfer in the 1980s of Koeberg fuel manufacturingtechnology from France in Necsa’s former BEVA programme.Although the transfer was technically successful (the BEVAfuel elements in Koeberg performed on par with those inFrance,) the transfer of technology incurred high costs over aperiod of about four years. At that point, however, the reactorvendor had already moved on with its next, more advancedfuel element design, which would have required anothersignificant investment for South Africa to ‘stay in the game’,a situation that would recur every few years. The messagehere is ‘don’t even consider investing heavily in the localmanufacture of critical items in the nuclear island unless asignificant percentage of South Africa’s electricity generatingcapacity will be nuclear-based, thus warranting such aninvestment’.

This raises the obvious question of whether South Africareally needs a reactor such as SAFARI I at all. The answer tothis question is an overwhelming YES. South Africa, as oneof the top three medical isotope producers in the world,should not relinquish its position. This was achieved underdifficult conditions and the replacement of the ageing SAFARII reactor needs to take that into account. Should ‘SAFARI II’then be only an isotope-producing reactor? This would be avery unfortunate retrograde step, as the growing use ofSAFARI I for non-nuclear industrial tests such as residualstresses and texture formation in metals, as well as neutronresearch, constitutes a powerful training and researchinstrument for South Africa’s national SET institutions.

In the ‘small and better’ strategy, South Africa’s SETcapacity should, therefore, rather focus on ex-corecomponents with the general aim of assisting local industryto participate in a meaningful manner in the future nuclearconstruction programme, but always focusing on theoverriding aims of the Generation IV LWR.

Some typical incremental advances with a highimpact in the LWR nuclear industry

The development of advanced steels for pressurevessels and steam pipingThe steels used for high-temperature steam piping forpressure vessels in both conventional coal- and nuclear-powered generating stations still require better understandingin terms of properties such as creep behaviour, corrosion,weldability, and end-of-life assessment. For instance, theweldability of P91 (nominally a 9%Cr-1%Mo-V-Nb steel) inthe Medupi power station required detailed attention to meetthe design requirements, and for application in a nuclearpower station the welding codes will even be stricter.

The steady improvements in the service performance ofP91 (Figure 10) shown below is evidence of ’small andbetter’ improvements over time, achieved only throughdedicated research.

Innovative materials science in solving a stresscorrosion cracking (SCC) problem through grainboundary engineering (GBE)Watanabe (1974) once had an ‘eureka’ moment when hechanged the relationship defined by the well-known Hall-Petch equation, that a reduction in grain size leads to the‘holy grail’ of higher strength with improved ductility, byasking what would happen if we were to change not the size,but the nature of the grain boundaries. This opened up manystudies towards understanding so-called coincident sitelattice (CSL) boundaries and how to increase their percentagein a mixture of low- and high-angle grain boundaries(LAGBs and HAGBs), twin boundaries (TBs), and then CSLboundaries. CSL boundaries are high-angle boundaries butpossess the special characteristics of LAGBs. The percentageof CSL boundaries can be measured with little difficulty bymany modern scanning electron microscopes fitted with anelectron backscatter diffraction (EBSD) capability. Thisinnovation is now being applied to the vexing problem ofgeneral intergranular stress corrosion cracking (IGSCC) in thetubing of steam generators of PWR stations.



Figure 10 – 100 000-hour creep rupture strength and temperature of9–12%Cr steels in general over time with the introduction of P91 steelby the late 1980s (Klueh and Harries, 2001)

Figure 9 – The allowable operating stresses for a design life of at least300 000 hours for three steels typically used in LWR technology. SA 508Grade 3 is a low-carbon manganese-molybdenum steel ’optimized forthe nuclear industry. X20 (2.25Cr-Mo steel) and P91 (X10CrMoVNb9-1)are both used for high-temperature steam piping (Buckthorpe, 2002)


In a number of groundbreaking patents and publications(Palumbo, 1993; Aust, Erb, and Palumbo,1994; Palumbo,Lehockey, and Lin, 1998; Was, Thaveeprungsriporn, andCrawford, 1998), Palumbo and others have found the meansto increase the density of CSL boundaries in Alloy 600 (a Ni-16Cr-9Fe alloy) through iterative strain-annealing or iterativestrain-recrystallization, typically from less than 10% CSLboundaries in the unprocessed alloy to as high as almost50% in the iteratively strain-annealed or strain-recrystallizedform. Note the very clear micrographic evidence of less grainboundary penetration from the surface for a grain-boundary-engineered Alloy 600 that has been subjected to laboratory-simulated stress corrosion cracking.

Note that in the coarse-grained microstructure in Figure14, the increase in the CSL boundary density from 20% to34% was sufficient to dramatically lower the alloy’s creep rateat 360°C, while in a fine-grained Alloy 600 the improvementin creep strength appears to be even greater. In both cases,

the improvements in creep strength were achieved with arelatively modest increase in CSL boundary density, with nofurther improvements at higher CSL densities.

Shifting the focus from ‘technology’ to ’technologyplus design and safety’

Up till now, the focus was very much on the technology ofnuclear materials, but as Hoeffelner (2011) has shown(Figure 3), this is barely half the story necessary for aresurgence of a national nuclear power programme. The otherhalf will require a significant body of high-level SET capacitywith better understanding to technically evaluate offers fromvendors, ask the right questions, retrieve the required designand performance data, and critically compare differences indesign approach from the point of view of:

➤ Expected performance➤ Life assessment

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Figure 11 – A schematic cutaway showing the internals of a PWR steam generator (Staehle, 2007) and general corrosion problems experienced in LWRsystems with the IGSCC Alloy 600 used in the U-tubes in the PWR steam generators (top of the figure) since the inception of commercial PWR technology(Palumbo, 1993)

Figure 12 – Alloy 600 (Ni –15.74Cr – 9.1Fe) sensitized for 1 hour at 600°C followed by 120 hours of corrosion testing according to ASTM G28 with (i) itsconventional microstructure and (ii) after grain boundary engineering (Lin et al., 1995)

➤ The demonstrated safety of the system being offeredunder all possible external or internal scenarios

before an operating licensing by the NNR (NationalNuclear Regulator) can even be considered. In the short termthis can be achieved by hiring in nuclear consultants(preferably with no national affiliation with the vendor-country), but this cannot be sustained in the long term forreasons of cost. This approach would also represent a lostopportunity for developing that high-level manpowerrequired for firstly the pre-operational evaluation phase ofeach offer, then the construction and operational stage of anumber of power stations, and finally waste managementand the eventual decommissioning of the power stations.

Building up such a body of high-level manpower mayappear to be a daunting task, but it is no accident thatHoeffelner (Figure 3) places the technology focus at the pointof entry for design and safety assessment. For instance, a

postgraduate engineer or scientist who has gained experiencein one aspect of the arc welding of zirconium-based claddingmaterial for a Masters or a PhD degree will have absorbed theintricate nature of this alloy far more than in a few months ofreading in a library or attending postgraduate courses,without going through the demanding process of an in-depthliterature review, research planning and execution, reportingof results, and finally modelling these results in aninformative discussion; with all of this being criticallyreviewed at the end by one or more external examiners fromthe nuclear industry. In addition, international publicationsthat are peer-reviewed by experts in the field add todemonstrating ‘better understanding’.

However, a basic change will, have to occur in thedefinition of which materials are nuclear and which are not,as Figure 2 has shown. Nuclear materials in a resurgentnuclear power programme are not simply limited tozirconium, uranium, and a very few others while the rest areseen as ‘conventional’ and therefore technically outside thecurrent narrow definition of ‘nuclear materials’. A steamgenerator on a PWR can never be viewed as simply anothertype of heat exchanger, and thereby not warranting the levelof regulatory supervision that a nuclear reactor pressurevessel does (and even the latter material is traditionally notviewed, in South Africa at least, as a ‘nuclear material’). Forreasons of public safety and acceptance/assurance, theprocess of obtaining an ASME Section III N-certificate onequipment used in a nuclear reactor are far more onerousthan if the same equipment is used in a conventional powerstation. This places most of the materials listed in Figure 2 inan entirely new class, that should be dealt with appropriately.

Some typical areas in the category of ‘betterunderstanding’ of an advanced PWR to consider in researchcould include:

➤ Projects that entail the welding of various components,such as fuel cladding, pressure vessel steel, steampiping, steam generator items, etc.

➤ Projects that entail the strength/ductility/creep/fracturerelationships of the above materials at roomtemperature as well as at typical steam operatingconditions

➤ Corrosion properties, including IGSS, pitting corrosion,



Figure 14 – Constant load creep data on Alloy 600 (Ni-16Cr-9Fe) of a (i) coarse-grained and (ii) a fine-grained microstructure in solution annealed and grainboundary engineered conditions. The creep tests were carried out under argon at 360°C with creep stresses of 300 and 450 MPa respectively. In (iii), thedependence of the steady-state creep rate and the percentage of cracked boundaries on the fraction of CSL boundaries on the coarse-grained material isshown (Was, Thaveeprungsriporn, and Crawford, 1998)

Figure 13 – Dependence of total grain boundary cracked fraction onCSL boundary fraction in Alloy 600 (Ni-16Cr-9Fe) (with varying amountsof carbon in solution and for the case of grain boundary carbides) forstrains of 15% and 20% with testing in 360°C PWR primary circuit waterat a strain rate of 3 ×10-7 per second (Alexandreanu, Capell, B., andWas, 2001)


hydriding of zirconium-based cladding, iodine SCC ofzirconium alloys, SCC of steam piping and steamgenerator components, etc.

ConclusionsThe following aspects need to be incorporated into a nationalmanpower training programme that should be well under waylong before the first reactor vendor’s technical offer isreceived.

➤ Set up a Governmental Advisory Board consisting ofthe main role-players in an expanded nuclear powerprogramme. These would include Eskom, the NNR,Necsa, senior representatives from industry, publicrepresentatives, and the like. The constitution of such abody would, however, have to be very carefully draftedto totally exclude items not associated purely withtraining of high-level manpower, as any other issueraised in such a forum could compromise theindependence of the NNR in ruling on safety issues

➤ Establish initially at least one (and later possibly asecond or third) research centre at a South Africanuniversity that will undertake the postgraduate trainingand development necessary for scientists, engineers,and technologists required by the programme. Thisneeds to be done in close association with Necsa

➤ Ensure that the local R&D effort eschews end-usenuclear materials, but rather focuses on upstreamprocesses in a ‘small and better’ focus

➤ Expand Necsa’s mandate to include research anddevelopment on materials covered by the broaderdefinition of nuclear material. The Advanced MetalsInitiative of DST can play a deciding role in overseeinghealthy cooperation between the Ferrous MetalsDevelopment Network managed on behalf of the AMIby Mintek and the Nuclear Metals DevelopmentNetwork managed on behalf of the AMI by Necsa.

AcknowledgmentsThe views expressed in this presentation are the author’sown and do not necessarily reflect official policy of theUniversity of Pretoria or of the South African Government.The author would also like thank many colleagues at Necsaand the University of Pretoria for valuable input andcomments. This contribution is published with the permissionof the University of Pretoria.

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IntroductionAssessing the potential for using alternativejoining and repair techniques for specificapplications in a given industrial sector isinformed by an appreciation of the current sizeand potential growth in that sector. Powergeneration by nuclear fission has been out offavour over the past few years, due to severalhigh-profile nuclear incidents, e.g. theFukushima Daiichi nuclear disaster, as well asconcerns over disposal of nuclear waste andspent fuel rods. However, due to rapidincreases in the demand for electricity, coupled

with many fossil-fuel fired plants reachingend-of-life, there is renewed interest ininstalling additional nuclear capacity.Proposing friction processing as an alternative(and relatively untried) joining technology forthe nuclear industry might be viewed aspotentially perilous because of stringentrequirements for validation of weld and repairprocedures. The intention in this paper is tointroduce some modern developments in thefriction processing arena and to outline thepotential that these processes hold formanufacturing of new components and formaintenance and life extension of ageingnuclear power plant.

The World Nuclear Association (WNA,2015) states that as at April 2015, there were437 nuclear reactors in operation, with another65 under construction and a further 481 eitherin planning or proposed. Clearly, there issignificant potential for the application offriction processing techniques in this powersector. The age of reactors varies, from first-generation reactors developed in the 1950s tothe current generation III reactors introducedin 1996 in Japan, with generation IV reactorsstill under development with a proposedintroduction date of 2020. The main reactortypes currently operational are pressurizedwater reactors (PWRs) – 273 units, boiling-water reactors (BWRs) – 81 units, pressurizedheavy-water reactors (CANDU-PHWR) – 48units, gas-cooled reactors (AGR and Magnox)– 15 units, light-water graphite reactors(RBMK and EGP) –15 units, and fast neutronbreeder reactors (FBR) – 2 units (WNA, 2015).

The majority of reactors currently inoperation are relatively elderly in terms of theirdesign life and would require decommissioning

Friction processing as an alternativejoining technology for the nuclearindustry by D.G. Hattingh*‡, L. von Wielligh*, W. Thomas† and M.N. James‡

SynopsisThe process of joining materials by friction is based on generating the heatnecessary to create a solid-state mechanical bond between two fayingsurfaces to be joined. In simple terms, the components to be joined aresubjected to frictional heating between rubbing surfaces, causing anincrease in interface temperature and leading to localized softening ofinterface material, creating what is described as a ’third body’ plasticizedlayer. This plasticized zone reduces the energy input rate from frictionalheating and hence prevents macroscopic melting. The plasticized layer canno longer transmit sufficient stress as it effectively behaves as a lubricant(Boldyrev and Voinov, 1980; Godet, 1984; Singer, 1998; Suery, Blandin,and Dendievel, 1994). The potential for this solid-state frictional joiningprocess to create high-performance joints between, for example, dissimilarmaterials with limited detrimental metallurgical impact, and reduceddefect population and residual stress level, has had a very significantimpact on fabrication and repair in industrial sectors such as transport.This paper presents a brief overview of the advances made within thefamily of friction processing technologies that could potentially beexploited in the nuclear industry as alternative joining and repairtechniques to fusion welding.

Modern friction processing technologies can be placed into two maincategories: those that make use of a consumable tool to achieve theintended repair or joint (friction stud and friction hydro-pillar processing)and those making use of a non-consumable tool (friction stir welding).The most mature friction joining technology is friction rotary welding,where a joint is formed between original parent materials only. A newaddition in this category is linear friction welding, which opens thepotential for joining complex near-net-shape geometries by frictionheating. The continuous innovation in friction processing over the last 25years has led to the development of a number of unique processes andapplications, highlighting the adaptability of friction processes forspecialized applications for high-value engineering components.

Keywordsfriction processing, joining, leak sealing, weld repair technology.

* Nelson Mandela Metropolitan University.† Co-Tropic Ltd., TWI.‡ University of Plymouth.© The Southern African Institute of Mining and

Metallurgy, 2015. ISSN 2225-6253. Paper receivedAug. 2015.



Friction processing as an alternative joining technology for the nuclear industry

in the short term if life extension methodologies are notadopted. The high capital cost and extensive lead timeinvolved in bringing new nuclear plants online make thewhole idea of accurately determining the remaining life ofexisting plants very topical.

The graphs in Figure 1 and Figure 2 were populated fromdata published by the WNA (2015) in 2015 and give anindication of the current operational plants, plants underconstruction, and proposed plants. Considering Figure 1,which shows the percentage contribution from nuclearfission, over the decade 2003 to 2013, to the electric powerneeds of both the ‘BRICS’ nations and of developed countries,it is clear that most of these countries made very little capitalinvestment in additional nuclear generation capacity, withmost countries showing a moderate decline in nuclear contri-bution. With the exception of France, and to a lesser extentBrazil, most countries currently obtain less than 10% of theirelectricity needs from nuclear. The future for nuclear powerlooks much more promising when one considers the potentialadditional capacity from the current build programme, andthe planned and proposed additions to the nuclear fleet overthe next 20 years (Figure 2).

Therefore, taking an overview of the installed, planned,and proposed future nuclear capacity, and factoring indevelopments in respect of the safer, more standardized andcost-effective generation IV reactors, there is considerableopportunity to plan for the implementation of alternativejoining and repair technologies in the nuclear industry.

The basic principles of generating electricity from nuclearfission are very similar to those used in fossil-fired thermalpower plants, if viewed from the secondary circuit (steam

generation) point onwards. Steam generated by controllednuclear fission is used to drive turbines, which are intercon-nected with electrical generators. A number of the mainmaterials used in the construction of nuclear power plants,including Zircaloy, can be joined or repaired by friction-basedprocesses. The joining of thin-wall sections intended for usein fuel rod application by rotary friction processes is currentlybeing studied at the Nelson Mandela Metropolitan University(NMMU) Work is also being done on Cr-Mo-V alloy steelsand various grades of stainless steel.

Friction processing technologiesOne of the simplest ways to warm up your hands is to rubthem together. This action generates heat in proportion to therapidity of the rubbing motion, the pressure applied on therubbing surfaces, and the duration. Similar concepts are usedto provide the primary thermal energy source in frictionprocessing (FP), allowing lower temperature solid-state jointsand repair processes to be applied to a wide variety ofmaterials (particularly the magnesium and aluminium lightalloys) used in manufacturing industry (Thomas andDuncan, 2010). In summary, friction processing comprises aset of solid-state joining techniques that are carried out bygenerating heat at a contact interface. This softens thematerial and allows plastic flow to occur which, incombination with an applied pressure, can be used to createstrong, low-defect, and low tensile residual stress bonds(Maalekian , 2007; Gibson, 1997; ASM International, 1993;Serva-Tech Systems, n.d. (a); Thomas and Dolby, 2001).

There are various types of friction processes, asillustrated in Figure 3.

The main friction processing joining technologiesconsidered in this paper are rotary friction welding (RFW),friction stir welding (FSW), friction hydro-pillar processing(FHPP, which can employ either tapered or parallel sidedconfigurations), and linear friction welding (LFW). LFW is a

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Figure 2 – Nuclear power generation landscape expressed in MWe(WNA, 2015)

Figure 1 – Nuclear power share for the period 2003 to 2013 (WNA, 2015)

Figure 3 – Friction process technologies (TWI) (Thomas and Dolby,2002)

relatively new and versatile process that permits the joiningof irregular shapes, allowing for near-net-shape fabrication.In this range of friction technologies, there are autogenousprocesses that form a joint by using parent material only(RFW and LFW), i.e. without the addition of any filler metal,processes that make use of a consumable tool (FHPP), andprocesses that use non-consumable tools (FSW).

There are four basic classifications for relative motion infriction welding (BSI Group, 2000; Serva-Tech Systems, n.d.(b)), namely: rotary, angular oscillation, linear reciprocation,and orbital motion. Figure 4 illustrates these variousmotions.

Fabrication by frictional heating is a well-establishedtechnology. The first use of friction knowledge to process andshape materials dates back to 1891, when Bevington (1891)used heat generated by friction to form and join tubes.

Rotary friction welding (RFW) RFW is a well-established ‘workhorse’ friction joiningtechnology that has been widely used for over a century innumerous applications ranging across the automotive,construction, aerospace, and medical sectors. Two alternativeRFW techniques, known as continuous drive and inertiafriction welding, are currently in use; the latter is the mostwidely adopted process. In both techniques joints are madeby placing a rotating component in contact with a stationarycomponent while applying a load perpendicularly to thecontact interface. Once the interface reaches the appropriatewelding temperature to plastically displace and fuse thematerials by forging, rotation is stopped and the weld isallowed to solidify while maintaining the forging force(Thomas and Duncan, 2010; Thomas, Nicholas, and Kallee,2001) as illustrated in Figure 5. The differences between thetwo methods are mainly in process control, with continuous

drive welding being done with a constant rotational speedthat may be varied at different stages of the weld cycle, whileduring inertia welding the process begins at a relatively highrotation speed, which gradually reduces to zero (Thomas andDuncan, 2010).

Friction hydro-pillar processing (FHPP)Like most of the modern friction technologies, FHPP (Figure 6)was invented and patented by TWI (World Centre for MaterialsJoining Technology) in the UK in the early 1990s. Adescription of the process, as given by Thomas and Nicholas(1992), states that friction hydro-pillar bonding is achieved byrotating a consumable tool coaxially in a hole while under load.The frictional heat generated results in the production of apillar of continuous plasticized layers until the hole is filled.The plasticized material consists of a series of shear layers orinterfaces which solidifies under the influence of the appliedforce. One of the main original observations was that theplasticized material advances more rapidly than the axial feedrate of the consumable tool, which results in the rising of thefrictional interface along the consumable tool to form thedynamically recrystallized deposit material. During the processa good degree of plasticization must be maintained at theinterface, allowing hydrostatic forces to be transmitted axiallyand radially to the inside of the hole in order to achieve a goodmetallurgical bond (Thomas and Nicholas, 1992).

A large body of reported work from various laboratorieshas shown that good mechanical integrity can be achieved inFHPP welds, with metallographic examination indicating afine-grained microstructure in the welded zone (TWI Bulletin,1997). The original geometry proposed for the FHPP processwas based on a horizontal cylindrical arrangement, withspecific clearance between the tool and the sides of the hole.However, in more recent applications tapered holes andconsumables are becoming more common, in which the angleof the two tapers is slightly different (Thomas and Temple-Smith, 1996; Bulbring et al., 2012; Meyer, 2001).Wedderburn (2013) reported that the tapered arrangementallows for an additional reactive force to develop horizontallyfrom the sidewall in addition to the vertical hydrodynamicforce, which can be utilized for heat generation when makingthe joint, and that this is beneficial for materials with poorextrusion properties.



Figure 5 – Schematic of the basic rotary friction welding process(Pinheiro, 2008)

Figure 4 – Relative motion classification in friction welding(Wedderburn, 2013)

Figure 6 – Schematic representation of the friction hydro-pillar process(Thomas, 1997)


A number of research institutes have claimed that goodquality FHPP welds can be made in steel and certain non-ferrous materials using a parallel hole geometry. In thiscontext a ‘good quality’ weld is characterized by high impact,tensile, and bend properties (TWI Bulletin, 1997). Thissentiment is supported by Wedderburn (2013) Bullbring, etal., (2012) who reported that good mechanical propertieswere achieved with a taper configuration on most thick-wallsteam pipe materials evaluated. Where welds were madeusing parallel tool and holes, periodic changes in themicrostructure were observed, where regions within thedeposit material were not fully transformed. This effect wasmore noticeable when a high tool rotation speed andconsumable displacement (upset) rate were used.Microstructural variations were also exacerbated when thehole depth to hole diameter exceeded a ratio of about 3.5:1(Meyer, 2003).

There is limited published work on the influence ofprocess parameters on FHPP welds. From the work by Meyer(2003), Wedderburn (2013), and Bulbring et al., (2012) onthe use of FHPP for high-strength low-alloy steel, it isapparent that hole geometry is more critical than tool shape.Meyer (2003) also noted that the influence on heatgeneration and bond quality is similar to that observed inconventional friction welding, although the rotational speedrequired for a good FHPP weld is significantly lower than inconventional friction welding. The forging force, which has amajor influence on weld properties in conventional frictionwelding, appears to have little or no influence on most FHPPwelds and its effect is limited to the upper area near thesurface region (Wedderburn, 2013; Bullbring, et al., 2012;Meyer, 2003).

Friction stir welding (FSW)FSW must be considered as one of the leading innovations insolid-state joining in the last 50 years. Invented and patentedby TWI (1991), FSW is a joining technique that allows bothlow and high melting point metals, including aluminium,lead, magnesium, steel, titanium, zirconium, and copper, tobe continuously welded with a non-consumable tool(Nicholas, 1998; Reynolds, Seidel, and Simonsen, 1999;Dawes, 2000). The wear rate and cost of refractory tooling isa limiting factor in welding steel alloys. One of the keybenefits of this solid-state friction welding technique is the

capability of welding dissimilar metals as well as joiningcertain metals that are difficult to weld by fusion processes,e.g. 2xxx series aluminium alloys. The process can best bedescribed as a continuous ‘hot-shear’ process involving anon-consumable tool that is responsible for mixing andforging the metals across the joint line (Delany et al., 2005).Selection of the tool geometry and material are importantconsiderations in achieving a sound joint. The mostimportant aspects of tool design are the pin and shouldergeometry, as these are responsible for forming the joint.

The basic principle of the FSW process involves plunginga rotating tool between abutting faces of the work piece andthen traversing it along the joint line (Figure 7). Rotarymotion of the tool generates frictional heat at the contactsurfaces, softening the material and creating a plasticisedzone around the tool pin and below the shoulder.

Apart from generating heat, the shoulder also containsthe plasticized material in the joint zone, which provides aforging action that assists with the formation of a defect-freesolid-state joint. In the nuclear industry, FSW has beenapplied to the manufacture of copper canisters for encapsu-lating nuclear waste (Andersson and Andrews, 1999) as wellas for leak sealing or material reprocessing techniques usedin the repair of surface-breaking or near-surface defects (VonWielligh, 2012).

Linear friction welding (LFW)(LFW is a joining technology that is ideal for welding non-symmetrical components. This solid-state joining processgenerates frictional heat through the linear forced interactionbetween a stationary surface and a reciprocating surface(Nentwig, 1995; Nicholas, 2003). The process requires alarge force applied perpendicular to the weld interface and, asthere is no containment of the hot plasticized material, acontinuous layer of flash is expelled during the rapid linearmovement. Since this softened ’third-body’ material is notcontained, surface oxides and other impurities are ejectedtogether with the plasticized ‘flash’ (Bhamji et al., 2011).

Unfortunately, LFW requires a major capital investmentdue to the complexity and size of the welding platforms. Thehigh process force must be contained within the platform

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Figure 8 – Schematic diagram of the linear friction welding process(Bhamji et al., 2011)

Figure 7 – Principle of friction stir welding (Thomas, Nicholas, andKallee, 2001)

structure, while achieving the required level of displacementand process control necessary to ensure joint integrity andgeometrical accuracy means that tight specifications are seton all system components and control algorithms. Hence LFWis considered feasible for the production of high-valueengineering components, e.g. joining of aero-enginecompressor blades to compressor disks (Wanjara and Jahazi,2005). LFW applications are predicated on the suitability ofthe process for joining materials with good high-temperatureproperties, low thermal conductivity, and acceptablecompressive yield and shear strength properties (Wanjaraand Jahazi, 2005). Low thermal conductivity helps confineheat to the interface region, while appropriate high-temperature properties (high yield strength and meltingpoint) allow a high level of frictional heat generation to occurbefore plastic collapse. Materials with good high-temperaturemechanical properties and low thermal conductivities, e.g.titanium, zirconium, and nickel alloys, are particularlysuitable for LFW applications.

Applications of friction technologies in the nuclearindustryA number of studies have shown that friction welding can beused to make sound joints and cost-effective repairs in thenuclear industry. These include the work done in a Swedishprogramme that evaluated FSW for encapsulating nuclear fuelwaste (Andersson and Andrews, 1999), while TWIdemonstrated that thermocouple probes can be fitted to boilerheader domes forged from 316 austenitic stainless steel forthe Hartlepool and Heysham nuclear power stations (TWIGlobal, 2014). Hy-Ten, an independent rebar and accessorysupplier, has gained approval for supplying friction weldedrebar couplers to be used in concrete construction by thenuclear industry (Maguire, 2012). Locally, the NMMU hasdeveloped two technologies with potential applications in thenuclear industry; one relating to leak sealing by FSW thatemploys a low-force partial-penetration methodology, and theother the registered Weldcore® process, which employs FHPPprinciples to core and repair sites used for creep assessmentas part of remaining life evaluation.

Nuclear waste storage encapsulation using FSWThis project formed part of a Swedish programme thatevaluated new ways of encapsulating nuclear fuel wastebefore storing it underground. The capsules weremanufactured from copper and consisted of a copper cylinderwith a base and lid (Andersson and Andrews, 1999). Thecapsule demand was estimated at 200 per year. Importantconsiderations in fabrication included inspection of the weldsand guarantees on achieving defect-free joints that wouldprevent water entering the canister during long-term storageunderground (Swedish Nuclear Fuel and Waste ManagementCo. (SKB), 2001). Typical canister dimensions were diameter1050 mm, length 4830 mm, and a total weight of 27 t. Tosatisfy the design requirements for environmental andchemical resistance, a copper alloy equivalent to EN 133/63with a 50 mm wall thickness was selected for the canisters.To improve creep resistance of the copper, 50 ppm ofphosphorus was specified in the alloy (Swedish Nuclear Fueland Waste Management Co. (2001). The project developed anelectron-beam welding solution for sealing the capsules and

also considered the use of FSW. The main consideration inselecting FSW as a potential alternative process was thesolid-state nature of the joining process, together with theability to make high-quality welds with a fully automatedplatform and standardized tool technology.

The investigation of FSW as an alternative joiningprocess started by considering the welding of 10 mm thickcopper plate strips about 0.5 m in length. This study wasused to find a suitable tool material as well as evaluatedifferent tool geometries. The only reference study at the timeinvolved FSW of thick-section aluminium plate with meltingtemperatures (659°C) well below that of copper (1083°C).However, copper tends to soften at lower a temperature,which allows the use of FSW on thick copper sections(Swedish Nuclear Fuel and Waste Management Co., 2001).This study led to the specification of process forces andpower requirements for welding 50 mm copper plate. Thiswas done by progressively increasing the plate thickness to50 mm while checking that the weld microstructure remainedacceptable for the identified application. The weld zoneshowed a fine equiaxed grain structure in the weld nuggetregion. Based on the success of these initial trials, SKBproceeded with the development of a FSW platform to weldlids to short (2 m) sections of 50 mm thick tubes. The tubewas placed in a vertical position with the welding head andtool in a horizontal position while the welding head rotatedaround the tube. For each weld, specific weld parameterswere recorded, providing a complete record of all the essentialvariables and their control during the welding process. Theseresults, in conjunction with non-destructive testing, form animportant resource for interpreting weld parameters and data.

Once adequate circumferential friction stir-welded jointswere achieved, attention was focused on eliminating the exithole left at the end of the weld run. This project has clearlydemonstrated the feasibility of fabricating capsules by FSW toenclose nuclear waste for long-term storage.

Leak sealing by FSWNuclear facilities situated close to the ocean could experiencematerial degradation, especially of stainless steel vessels andpipework, arising from environmentally-induced stresscorrosion cracking (SCC). This prompted an investigation intoidentifying a suitable refurbishment technique that wouldprovide a cost-effective and permanent leak-sealing/SCCrepair technique. Implementation of the repair process wasrequired to have no influence on the operation of the plant,and also to fulfil the stringent nuclear safety regulations. TheNMMU was therefore requested to investigate the feasibilityof using FSW as a potential leak-sealing technique for 304Lstainless steel. This study focused on establishing atechnology procedure for the identified application, whichentailed reviewing the process, tool material, and toolgeometry designs to develop a solution suitable for high-temperature FSW of unsupported, partial penetration leaksealing of 304L stainless steel with water backing.

The approach taken in the investigation was initially toestablish a theoretical processing parameter window frompublished literature. Subsequently, a suitable tool alloy andtool profile were identified before preliminary experimentalwork led to a low-force processing window for 304L stainlesssteel. Various experiments were carried out to determine theeffect of process parameters on weld quality and the




processing forces required to achieve high quality. Thegoverning factor was found to be the downwards forgingforce required to achieve a fully consolidated, partialpenetration weld that would seal a leak in stainless steelsheet resulting from, for example, SCC. Finally, a completeweld procedure specification was proposed and evaluated bywelding on a mock-up of the real application with waterseeping through the simulated crack during the weldingprocess. In principle, this is a similar concept to FSW underwater, which has been reported by Ambroziak and Gul(2007).

Such applications require refractory tool materials, andW-25wt%Re was selected for the leak-sealing application asit provides good resistance under conditions of unevensurfaces, high vibratory loads, and plate deflection (VonWielligh, 2012). Additional benefits include the ability to re-profile the tool using standard machining processes, henceincreasing tool life. Process development focused onestablishing a low-force processing window for the 304L

stainless steel that minimized plate deflection during weldingin an unsupported condition. The investigation led to thedevelopment of a new dual-control tool plunge strategy, inwhich the operator controls the maximum downwards Z-forceduring the plunge stage while simultaneously controlling theplunge rate and depth (see Figure 9). Plunge rate is, however,also a function of the Z-force feedback, which could be seenas a significant advantage because the maximum platedeflection at the start of the weld can therefore be limited(Von Wielligh, 2012).

The minimum Z-force required for repeatable weldconsolidation using tools with a 14 mm, 12 mm, or 10 mmdiameter shoulder was found to be 14.25 kN, 10.5 kN, andand 9 kN respectively (Von Wielligh, 2012). This investi-gation showed that water cooling and water seepage throughthe simulated SCC had only a minor effect on weld qualityand that the dominant factor influencing weld quality wasplate deflection. A full preliminary welding procedure specifi-cation was developed for both the 14 mm and 12 mmshoulder diameter tools, but only the procedure for the 14mm tool was deemed robust enough for industrialapplication.

Un-etched images of typical weld cross-sections obtainedduring the repair trials of SCC in stainless steel are shown inFigure 10. This figure shows the defect population observedat the start, middle, and end positions in two welds, A and B,made using different weld parameters while water wasseeping through the SCC, using two different sets of processparameters.

On the etched specimen in the macrograph in Figure 11,the lighter etching material clearly shows a well-defined flowpattern in the weld cross-section. This material was found tobe tungsten, a by-product of tool wear that has diffused intothe weld. The darker line extending beyond the flaw intendedto simulate a stress corrosion crack at the bottom of weldcross-section is termed a ‘lazy S’ defect in FSW and is notpart of the simulated crack. It is generally ascribed to the

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Figure 10 – The defect population found in weld cross-sections along the weld length representing the start, middle, and end of welds for two differentwelding parameter sets (Von Wielligh, 2012)

Figure 9 – Difference in the Z-force feedback for different controlstrategies (Von Wielligh, 2012)

presence of oxidation and corrosion by-products of the wirecutting process, and serves as a good indication of howactual corrosion products are likely to be distributed acrossthe weld.

Vickers microhardness tests did not show any significantchange in the hardness of the weld area or heat-affected zonecompared with the parent material. This is to be expected, asaustenitic stainless steels are not hardenable by heattreatment (Von Wielligh, 2012).

The feasibility of applying this FSW process to crackrepair in industry was tested using a mock-up of theindustrial application of a tank containing a stress-corrosioncrack. A simulated crack was successfully welded in anunsupported 304L stainless steel plate surface in theannealed condition with water backing up to a depth of 2 mm.

The weld repair process required making a number ofoverlapping friction stir welds, with a step distance betweenthem equal to the pin diameter. The step direction wastowards the retreating side of the weld, because flashformation arising from plate deflection made it impractical tostep towards the advancing side.

The typical weld repair pattern is shown in Figure 13. Thefirst weld, W1, is carried out from left to right with aclockwise spindle rotation. The retreating side of the weld isthus located at the bottom. The second weld, W2, is carriedout from right to left with an anticlockwise spindle rotationsuch that the retreating side remains at the bottom of theweld. This process of alternating welds is continued until thecracked area has been covered.

Figure 13 shows an example of such a weld repair withfive overlapping welds, covering an area of 120 mm × 20mm. In this figure, all the loosely attached flash has beenremoved.

This investigation has shown that leak sealing of stress-corrosion cracks in an unsupported tank surface is possibleby using overlapping, partial penetration, friction stirwelding. The investigation also highlighted the fact that dueto material thinning during the FSW process, the total areathan can be repaired has to be limited, based on allowableplate thinning. A reduction in the original plate thicknessoccurs with each successive overlapping weld. The decreasein plate thickness is likely to be a function of plate thickness;in other words, dependent on constraint. Furthermore, FSWtool exit holes can be partially eliminated by using tools witha retractable pin. In summary, FSW is a potential repairtechnique for SCC, although for unsupported welding carefulconsideration must be given to the influence of platethickness, surface condition of the area to be processed andits size, and deflection during welding, as well as thepermanent deformation (thinning) induced by the process.FSW is also likely to alter the metallurgical and mechanicalproperties in the processed area, and this aspect also needsconsideration. Associated friction processes such as frictionsurface dressing or friction cladding could also be consideredfor such repair procedures if it is accepted that the mainpurpose of the process is not to recover structural integrity,but rather top-seal seepage of the liquid inside the tankthrough stress corrosion cracks. Friction cladding is anadditive process, which has the potential to reduce platedeflection and permanent deformation imposed by the frictionprocess. Figure 14 shows the heat generation in the toolwhile traversing along a repair zone with water leakingthrough the simulated crack. The installation of thedeveloped platform is shown in Figure 15.



Figure 12 – The welding pattern of five consecutive welds used tofriction stir process an area of 120 × 20 mm (Von Wielligh, 2012).(Welds W1, W3, and W5 were welded from left to right and welds W2and W4 from right to left)

Figure 11 – Macrograph of a typical friction stir processed SCC, etchedwith Marbles reagent [30]

Figure 14 – FSW in process as water leaks through the simulated crack(Von Wielligh, 2012)

Figure 13 – The visual appearance of a friction stir processed tanksurface area with the loosely connected flash removed on theadvancing side of the weld (Von Wielligh, 2012)


Weldcore® sample removal and repair technologyOngoing assessment of the remaining life of high-temperature and -pressure (HTP) components in nuclearpower stations is of paramount importance in ensuring theirsafe and cost-effective operation. Although failure can havesevere consequences, economic considerations are requiringoperators to extend the operating lives of plants currently inoperation. To ensure safe operation, more direct experimentalassessments have to be made of remaining life tocomplement and underpin prediction models. Neutronirradiation embrittlement is one of the concerns for thenuclear industry, as this could limit the service life of reactorpressure vessels. Improved understanding of the underlyingcauses of embrittlement and better calibration of predictionmodels will provide both regulators and power plantoperators with improved estimates of vessel remaining life(Odette and Lucas, 2001). The NMMU, in collaboration withEskom, has developed the patented Weldcore® process,which is a sampling and repair process that is currently usedin fossil fuel power stations to assist in sample removal forcreep damage assessment.

The process involves removal of a cylindrical

metallurgical sample and hole preparation for repair in foursteps, as illustrated in Figure 16. The hole is then repairedusing friction taper hydro-pillar processing (FTHPP) asillustrated in Figure 17.

Current Weldcore® applications are primarily in a coringand repair procedure for local metallurgical sample removalsites. The advantages of the process are that the metallurgicalcore can extend much closer to the centre of a thick-walledpressure pipe, and that the coring and repair can beaccomplished during plant operation. The core hole may beeither tapered or vertical-sided, depending upon the thicknessof the material to be welded and the depth of the hole. Fordeeper repair welds, a cylindrical configuration for the hole

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Figure 15 – Set-up for the repair process feasibility investigation, showing the FSW platform attached to a curved stainless steel water-filled test coupon(Von Wielligh, 2012)

Figure 16 – Illustration of four Weldcore® coring stages (Hattingh et al.,2012)

Figure 17 – Schematic of the FTHPP process with macro appearance ofa cross-section of a weld (Hattingh et al., 2012)

and rotating tool is recommended to ensure that the processforce and torque are limited. The consumable rotating tool istypically made of the same alloy as the base metal, althoughit is possible to achieve a matched, over-matched, or under-matched weld zone. On completion of the weld, the remainingportion of the consumable rotating tool can be cut off andground flush with the metal surface. Since the ’weld’ does notproduce a liquid weld puddle, the orientation of the weld isnot affected by gravity, hence making the process position-independent.

The main factors making the Weldcore® processattractive to the nuclear fraternity relate to platform size(which is compact and hence suitable for on-site work) anda high degree of process automation. The Weldcore®

platforms were initially developed for high-temperaturesteam pipes used in power stations. However, the success ofthe platform has led to further developments for applicationsthat include turbine disc and reducer sections.

Figure 18 shows the modular arrangement andadaptability of the welding platforms that were developed atthe NMMU for specific applications of the Weldcore® processin the power generation industry.

The in-situ application on steam pipes required thedesign and development of a portable reconfigurable platformthat allowed operation at various locations throughout apower station. The modular arrangement of the weldingplatform, involving a frame, spindle, and drive motor,facilitates ease of transport, positioning, and mounting.Electrical and hydraulic power and control connections areoperated via a remote control unit.

The process and portable platform allow for the in-situremoval of a representative sample of metal from a structure(pipes, turbine rotors, etc.) that can be used for accuratemetallurgical assessment of the remaining life of thestructure. It provides superior data to other current methodsas the excised core is a far more representative sample of thethrough-thickness than other techniques, with a shorter

acquisition downtime. The process has been tested on steampipes and turbine components in the power generationindustry and is in the final phases of being tested onstructures on a petrochemical plant. The main benefit of theWeldcore® process is a reduction in the risk of failure andhence better management of safe life in high-valueengineering components.

DiscussionThe introduction of a new or alternative joining and repairprocesses into the nuclear industry entails a number ofchallenges from the engineering and regulatory points ofview. Nuclear power, for obvious reasons, is a tightlyregulated industry, but the authors propose that byintroducing improved life assessment standardization, therisk and complexity of safely operating these types offacilities could be further reduced. Cost-effective conditionmonitoring is an implicit part of life extension for ageingplant, and of the through-life cost and performanceoptimization of new installations.

Friction processing presents a viable alternativemethodology for some aspects of condition monitoring andrepair of thermal power plant, with significant advantagesaccruing from its solid-state nature (which opens the door towelding of dissimilar metals). A solid-state joining processlimits microstructural phase transformations, while lowerprocess temperatures result in a significant reduction inthermal stresses and associated distortion compared withconventional fusion welding processes. The lower thermalgradient results in a narrow heat-affected zone with a fineequiaxed weld nugget and limited grain growth in thethermomechanically transformed region. Thesemicrostructures are normally associated with good strengthand toughness. As illustrated in the applications presented inthis paper, friction processing lends itself to a high degree ofautomation and process control, giving repeatable welds andallowing sophisticated monitoring of weld quality. Friction



Figure 18 – Developmental progression of Weldcore® equipment (Hattingh et al., 2012)


welding offers the power generation industry a clean,reconfigurable welding process suitable for complex andbespoke applications and which can greatly reduce processtime and cost of condition monitoring and repair.

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IntroductionDuring the development of new materials forthe nuclear industry, materials testing andcharacterization are of the outmost importanceto maintain safety standards and reliability. Nocompromise on safety in the workplace in anyarea within the total nuclear fuel cycle can betolerated, and therefore most in-situ materialcharacterization and testing is conducted bycertified and qualified personnel schooled indestructive and non-destructive testing (DTand NDT) methods. Certification and qualifi-cation in NDT can be obtained through manytraining centres in South Africa in accordancewith European, American, and other interna-tional standards (SGS, n.d.; SAIW, n.d.;African NTD Centre, n.d.).

The testing of new methods and materialsrelated to the nuclear fuel cycle is essential forthe continuing development and safety ofnuclear-related materials and processes. Thesefundamental research initiatives are mostlyperformed at the laboratory scale by materialand instrument scientists, researchers, andmost likely postgraduate students.

Davies (2000) describes the role and valueof NDT during maintenance and in-serviceinspection of nuclear power plants duringoutages, and particularly the monitoring ofmaterial degradation to prevent failure.Ultrasonic testing (UT), magnetic testing (MT,)and electrical testing (ET) play a major role asNDT methods for monitoring materialsdegradation in-situ, while atomic- and nuclearphysics-based methods such as positronannihilation, neutron diffraction, as well as X-ray and neutron tomography are limited tolaboratory-scale experimentation. However,conventional film-based X-ray and gamma-rayradiography (RT) techniques are being appliedthroughout many areas of material testing inthe nuclear fuel cycle.

The ‘nuclear fuel cycle’ refers to the entirerange of activities associated with theproduction of electricity from nuclear fission,entailing (International Atomic EnergyAgency, n.d.):

➤ Mining and milling: from mined uraniumto yellowcake

➤ Conversion: from yellowcake to gas➤ Enrichment: increases the proportion of

the fissile Isotope➤ Deconversion: depleted uranium.➤ Fuel fabrication: UO2 pellets – fuel pins –

fuel elements➤ Electricity generation: fuel burn-up➤ Storage: spent fuel➤ Reprocessing: spent fuel➤ Radioactive waste: safe storage.

The nuclear fuel cycle includes the ‘frontend’, i.e. preparation of the fuel, the ‘serviceperiod’ in which fuel is used in the reactor togenerate electricity, and the ‘back end’, i.e. thesafe management of spent fuel, includingreprocessing and reuse and disposal.

Neutron- and X-ray radiography/tomography: non-destructive analyticaltools for the characterization of nuclearmaterialsby F.C. de Beer*†

SynopsisA number of important areas in nuclear fuel cycle, at both the front endand back end, offer ideal opportunities for the application of non-destructive evaluation techniques. These techniques do not only provideopportunities for non-invasive testing of e.g. irradiated materials, but alsoplay an important role in the development of new materials in the nuclearsector. The advantage of penetrating radiation used as probe in theinvestigation and testing of nuclear materials makes X-ray and neutronradiography (2D) and tomography (3D) suitable for various applicationsin the total nuclear fuel cycle. The unique and different interaction modesof the two radiation probes with materials provide several opportunities.Their complementary nature and non-destructive character makes themmost suitable for nuclear material analyses, analytical methoddevelopment, and the evaluation of the performance of existing nuclearmaterial compositions. This article gives an overview of the X-ray andneutron radiography/tomography applications in the field of nuclearmaterial testing, and highlights a few of the success stories. Severalselected areas of application in the nuclear fuel cycle are discussed toillustrate the complementary nature of these techniques as applied tonuclear materials.

Keywordsneutron radiography, X-ray radiography, SAFARI-1; non-destructivetesting.

* Radiation Science Department, Necsa, Pelindaba.† School of Chemical and Minerals Engineering,

North-West University, Potchefstroom.© The Southern African Institute of Mining and

Metallurgy, 2015. ISSN 2225-6253. Paper receivedAug. 2015 and revised paper received Aug. 2015.



Neutron- and X-ray radiography/tomography

The general nuclear fuel cycle is schematically depicted inFigure 1, showing the various activities in the production ofenergy through the nuclear fission process. Every activityrequires conventional NDT techniques to be conducted tomaintain the safe working and operation of the plants andfacilities. The standard NDT methods applied to e.g.inspection of welds in piping, are (Willcox and Downes, n.d):

➤ Radiography testing (RT)➤ Magnetic particle crack detection (MT)➤ Dye penetrant testing (PT)➤ Ultrasonic flaw detection (UT)➤ Eddy current and electromagnetic testing (ET).

This paper does not focus on the so-called conventionalNDT techniques and their application in the nuclear sector,but rather on the non-conventional NDT techniques that areused as needed, and which constitute important researchtools. In particular, penetrating radiation probes as realized inradiography/tomography are described with specificapplications in material research. Quantitative and/orqualitative data obtained through applying these noveltechniques in a laboratory environment adds value to manyareas within the nuclear sector. The following specificactivities, ranging from mining the ore to security of thewaste generated, and where radiography and tomography areapplied, are highlighted in this paper:

➤ Mining and geosciences: quantification of ore deposits➤ Fuel fabrication: development and testing of new

materials ➤ Electricity generation: fuel rod performance ; post-

irradiation examination (PIE) ➤ Radioactive waste: safe storage, civil engineering.

Analytical methods based on penetrating radiation Information about the internal structures of objects, forexample the hydrogen content of Zr cladding, can be obtainedby destructive analytical methods, e.g. cutting a the fuel rodin a 2D plane for analysis by electron microscopy, or sieveanalysis for particle size distribution of a soil sample. In mostcases, once the sample has been destroyed, no otheranalytical tests are possible and the larger picture (volumetrichydrogen distribution and particle size distribution in thesoil) is lost.

More valuable, unique, and in some cases more accurateresults can be obtained only when three-dimensionalinformation is available. For research purposes the mostacceptable way to obtain information while maintaining thesample integrity is to apply a non-destructive test usingpenetrating radiation (either with X-rays, gamma rays, orneutrons). It is worthwhile to mention that the neutron, thefission product of the nuclear fuel cycle, can be used as aprobe to investigate the integrity of the nuclear fuel itself.After irradiation the physical condition of fuel pellets, whilestill intact in the fuel pin, can be obtained only by means ofradiography. This manner of non-invasive investigationkeeps the sample intact, leaves the sample in its originalform, and it is possible for other tests to be conductedsubsequently on the real sample as if it was not touched. Thenon-invasive process allows for the generation of valuablequalitative information. However, when digital data istransformed into three-dimensional tomographic data(Banhard, 2008), it is possible to obtain high-resolutionquantitative information of the internal structures andproperties of the object. An example is the volumetric poresize distribution of voids within nuclear-encapsulatingconcrete matrixes, as well as their physical distributionthroughout the sample (McGlinn et al., 2010).

X-ray, gamma-ray, and neutron radiation are attenuated(absorbed and scattered by the sample) according to anexponential law (De Beer, Middleton, and Hilson, 2004):

[1]

where I is the intensity of the transmitted radiation beam, I0the intensity of the incident radiation beam, μ the attenuationcoefficient (cm2/g) of the material under investigation for thespecific radiation type, ρ the density of the sample (g/cm3),and x the thickness of the sample (cm).

The attenuation coefficient μ expresses the totalattenuation, due to both the scattering and capture processesfor the incident radiation. The term μρ is also called the totalabsorption coefficient of the sample. We assume that thequantity μρ is linearly related to its constituents:

[2]

where μi is the radiation attenuation coefficient of constituenti, ρi the density of constituent i, Vi the volume fraction ofconstituent i, and ∑ the summation symbol for the ithcomponents. Composites of materials will thus have adifferent radiation attenuation property from the individualelements.

The parameters I(E), I0(E), and μ(E) reflect radiationenergy dependency. This dependency, in a radiography

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Figure 1 – Schematic depiction of the nuclear fuel cycle (Pixshark, n.d.)

context, means that materials will attenuate differentradiation types by different magnitudes and thus will yielddifferent radiological images, and also that an element hasdifferent attenuation properties at different energy levels forthe same radiation type. This implies that an element can betransparent to fast neutrons (MeV energies) but can bedetected easily using thermal neutrons (eV energies). Anexample is the thermal neutron scattering and absorption ofhydrogen and boron (Domanus, 1992).

Although the basic interaction of X-rays and neutronswith the elements differs, the principle of conductingradiography to obtain a two-dimensional radiograph/imageof the sample is the same. Figure 2 schematically illustratesthe basic components and layout of a radiography facility.

A source of radiation emits penetrating radiation towardsa sample. For example, a sample contains either a defect, aninclusion of another material, or a void that is abnormal forthe sample, or an area that differs completely in terms ofcomposition from the basic matrix of the sample, whichresults in a lower or higher density at the location within thesample. The incident radiation will be attenuated (scatteredand/or absorbed) differently due to the abnormality. Asensitive area detector, with a high quantum efficiency forthe detection of the specific type of penetrating radiation,registers the difference in attenuated radiation that haspassed through the sample. The 2D data (image) obtained by

the detector is called a radiograph and contains the integratedradiation transmitted information for the total sample in acertain orientation with respect to the source and detectorconfiguration.

The information captured in the radiograph differs inprinciple for X-ray and neutron radiography/tomography.The two probes are mostly utilized within the nuclear fuelcycle as non-destructive techniques in research and fornuclear material qualification and quantification. Theseprinciples are discussed in more detail in the followingparagraphs.

X-ray radiographyX-ray interaction with materials depends on the density ofthe sample – i.e. the electron cloud density (Banhard, 2008).The area detector registers a two-dimensional image(radiograph) of the object representing the internal structuredensity. Elements with low electron densities are not easy toresolve in a radiograph, but they are easily penetrated toreveal denser materials embedded within the sample matrix.Figure 3 presents the different X-ray attenuation coefficients(cm-1) for 125 kV X-rays for the full spectrum of elements inthe periodic table. It clearly shows the increase in absorptionof X-rays (darker shading) at higher atomic numbers.

Neutron radiographyThe interaction of neutrons with materials is totally differentto that of X-rays, since neutrons, being neutral particles,interact only with the nucleus of the atom. Neutrons are notaffected by even a dense electron cloud, e.g. of a lead atom(μρn = 0.38 cm-1). The thermal neutron attenuation coeffi-cients depicted in Figure 4 shows a totally different, and insome instances an opposite attenuation capability (greyscale) to that for X-rays (Figure 3). Hydrogen, as a highlyattenuating material (μρn = 3.44 cm-1), will be easy to detectand clearly visible on a neutron radiograph when embeddedin e.g. a ZrTM (μρn = 0.29 cm-1) fuel pin, which is nearlytransparent to neutrons. A radiograph with low- tointermediate-energy X-rays is possible as the ZrTM tube (μρX= 2.47 cm-1) attenuates most of the X-ray radiation and withno photons remaining, the H (μρX = 0.02 cm-1) cannot beregistered/detected on the X-ray radiograph.



Figure 3 – Periodic table with X-ray attenuation coefficients of the elements for 125 kV X-ray energies (Grünauer, 2005)

Figure 2 – Principle and layout of a 2D radiography set-up. A similarset-up is used for tomography, with the sample rotating in the radiationbeam (Domanus, 1992)


TomographyThe word ‘tomography’ comes from the Greek words ‘to cutor section’ (tomos) and ‘to write’ (graphein) (Banhard,2008). Tomography is also known as computer tomography(CT) or computer assisted tomography (CAT) as in diagnosticinvestigations in the medical field. For the purpose of thisarticle, the following semantics are adopted: CT in generaldescription, XCT for X-ray computer tomography, and NT forneutron tomography. CT is a radiographic inspection methodthat uses a computer to reconstruct an image of a cross-sectional plane (slice) through an object (ISO 15708-1). Theresulting cross-sectional image is a quantitative map of thelinear radiation attenuation coefficient, μ, at each point in theplane. The linear attenuation coefficient characterizes thelocal instantaneous rate at which the incident radiation isattenuated during the scan, by scatter or absorption, as itpropagates through the object.

To obtain this ‘map’, the sample is radiographed and thusprojection data is gathered from multiple directions throughmany angles of the sample. For the purpose of this article, nodetailed description of the 3D reconstruction process of thesample is presented. To put it simply, multiple 2D-projectionsare fed into a dedicated computer with a specialized computeralgorithm to create cross-sectional planes of the sample.When these cross-sectional planes are stacked together, a fullvirtual three-dimensional image (tomogram) of the samplecan be viewed and analysed.

Application of radiography and tomography withinthe nuclear fuel cycleIn each of the following sectors of the nuclear fuel cycle,radiography and/or tomography have been applied usingeither X-rays or neutrons. In some areas the application waspioneered in the early second half of the 20th century bymeans of film techniques. The techniques applied and thedescription thereof is not within the scope of this article.However, the outcomes of these film-based investigationsand the results obtained will be described, together with therecent digital methods in this field.

MiningX-ray-, gamma-ray, and neutron tomography havedemonstrated their potential in the earth sciences asimportant diagnostic tools to generate volumetric data ongeological compositions, especially advances in the areaborehole core investigations, as depicted in Figure 5. Thisaspect is being explored further with optimum resolutionobtained through the application of micro-focus X-raytomography, as CT complements conventional destructiveanalytical thin-sectioning of drill core samples (De Beer andAmeglio, 2011).

The raw material for nuclear fuel is uranium, which is arelatively common element that can be found throughout theworld. Uranium is present in most rocks and soils, in manyrivers, and in seawater. Uranium is about 500 times moreabundant than gold and about as common as tin.

The largest producers of uranium are currently Australia,Canada, and Kazakhstan, with Namibia rated 5th and SouthAfrica 11th globally (World Nuclear Association, 2015). The

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Figure 4 – Periodic table with thermal neutron energy attenuation coefficients of the elements (Grünauer, 2005)

Figure 5 – Left: the grouped pyroxene mineral (coloured red) can beclearly seen in norite (top), and different concentration of feldspar areobserved in anorthosite (bottom) in a thin slice. Right: transparentcorresponding neutron tomograms. The minerals shown are onlypyroxenes that are present in both norite and anorthosite but atdifferent concentrations [archives from Necsa’s Nrad and MIXRADsystem]

concentration of uranium in the ore can range from 0.03 to20%. Conventional mining is by open cut or undergroundmethods. Uranium ore can be produced from a mine specif-ically for uranium, or as a by-product from mines with adifferent main product such as copper, phosphate, or gold(International Atomic Energy Agency, n.d.).

Using micro-focus X-ray CT with 100 kV potential todistinguish between gold (μρX = 358 cm-1) and uraninite(μρX = 283 cm-1), both the minerals are easy to distinguishfrom the matrix minerals such as pyrite (μρX = 18.4 cm-1),zircon (μρX = 45.9 cm-1), and brannerite (μρX = 89.6 cm-1)due to their much higher elemental densities. The use of CTas a sorting method is still a challenge, as it is difficult, at the100 kV X-ray energy CT capability available, to distinguishbetween gold and uraninite (Chetty et al., 2011). In a follow-up 3D micro-focus X-ray computer tomography (μXCT)study using 120 kV, the contrast and resolution of theminerals were well defined and individual minerals could beseparated and distinguished from other minerals (Sebola,2014). For the detection of uranium, 3D-CT wasbenchmarked against 2D mineralogical results from opticalmicroscopy and scanning electron microscopy (SEM).Uraninite, brannerite, and uraniferous leucoxene are theuranium-bearing minerals present in the samples (from theVaal Reef) and were quantified by μXCT-3D analysis for theirsizes, shapes, and distribution with respect to other mineralcomponents in the samples. Uraninite was found to be themajor mineral, occurring mainly in the quartz matrices andalso associated with carbonaceous matter as depicted inFigure 6. The uraninite and gold in the matrix occurred asrounded grains up to 200 μm in size, as observed by 2Dmineralogical techniques. CT allows for 3D grain sizeanalyses of the uraninite grains in the total volume of thesample, and in this study also their association with matrixminerals, as depicted in Figure 7. The CT observationssupported the results acquired by conventional mineralogicaltechniques, suggesting that 3D μXCT can be used tocomplement other mineralogical techniques in obtaining 3Dinformation. However, 3D μXCT has limitations such asspatial resolution, partial volume effect, and overlapping ofmineral grey-scale values. It is therefore suggested that the

technique cannot be used as an independent tool for mineralcharacterization, but rather in support of existingmineralogical techniques. However, valuable information isadded into the nuclear value chain via 3D-CT. No samplepreparation is needed other than cutting a small piece of rockfor analysis. The sample integrity is maintained due to thenon-destructive nature of the technique, as it provides 3Dinformation for the total volume of the sample, includinginternal components. Results can be obtained within about 1hour scan at high resolution with up to 2000 projections,providing resolution down to 6 μm (includingreconstruction).

EnrichmentEnriched uranium is uranium in which the concentration ofU-235 has been increased through the process of isotopeseparation. U-235 is the only nuclide existing in nature (inany appreciable amount) that is fissile with thermal neutrons(OECD Nuclear Energy Agency, 2003). Natural uranium is99.284% U-238 isotope, with the U-235 isotope constitutingonly about 0.711%.

The very first application of neutron radiography was inthe early 1960s for nuclear fuel characterization using thefilm technique. The use of neutron radiography in themonitoring of isotopic enrichment in fuel pellets loaded into afuel pin has been demonstrated by Frajtag (n.d.) (Figure 8).

Gosh, Panakkal, and Roy (1983) investigated thepossibility of monitoring plutonium enrichment in mixed-oxide fuel pellets inside fuel pins using neutron radiographyas early as 1983. Recently, Tremsin et al., (2013)investigated the very large difference in the absorption cross-sections of U-235 and U-238 isotopes, as shown in Figure 9,and deduced that very accurate non-destructive spatialmapping of the enrichment level of fuel pellets, as well as ofthe distribution of other isotopes in the spent fuel elements(Nd, Gd, Pu, etc.), can be achieved.

Additionally, information on the distribution of isotopescan then be used for the investigation of fuel burn-up ratesfor fuel elements placed at different rod positions in thereactor core.

Large differences in transmission spectra allow veryaccurate mapping of isotopic distributions in the samplesusing either transmission radiography or neutron resonanceabsorption characteristics of the respective isotopes (Tremsinet al., 2013). One of the attractive features of energy-



Figure 6 – X-ray CT-tomogram slice showing the distribution ofuraninite grains in the quartz-pyrite matrix and the carbonaceousmatter (Sebola, 2014)

Figure 7 – Grain size distribution of uraninite in the matrix and thecarbonaceous matter (Sebola, 2014)


resolved neutron radiography is the ability to enhancecontrast, and in some cases enable quantification, as shownin Figure 10. It was observed that the contrast between thepellets of different density depends strongly on the range ofneutron energies used. The more thermal part of the beamspectrum (neutron energies above 19.7 MeV) reveals thepellet with the lowest density as an object with the highesttransmission. The coldest part of the neutron spectrum(neutron energies even below 6 MeV) shows the least densepellet as a darkest in the assembly.

Nuclear fuel fabrication and testing (I & PIE)Nuclear fuel types range from isotopic sources in a form ofsalt or disks to pressurized water reactor (PWR) fuel in theform of UO2 fuel pellets inside ZrTM -cladded fuel pins.Nuclear fuel is subjected to stringent manufacturing and

performance criteria which have to be verified. Non-destructive testing of the fuel ensures that other tests can beperformed subsequently and that the material can still beapplied in its specific environment. Some of the NDT tests areapplied to characterize and/or quantify the integrity of thefuel or as quality assurance tests. X-ray radiography cannotbe used for irradiated fuel inspection, whereas neutronradiography becomes possible due to the following reasons(Lehmann, Vontobel, and Hermann, 2003):

➤ Uranium has a very high attenuation coefficient for X-rays (about 50 cm-1 at 150 keV). The diameter of fuelpellets is in the order of 10 mm and penetration by X-rays is impossible. High-energy gamma radiation (>1MeV) is, however, suitable for quality control of freshfuel pellets

➤ The neutron attenuation coefficient for the natural

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Figure 8 – Neutron radiograph (film) of fresh fuel pellets in a fuel pin with varying degrees of isotopic U-235 enrichment. Pellets with higher enrichmentappear darker (Frajtag, n.d.)

Figure 9 – Neutron attenuation of 100 μm thick U-235 and U-238 isotopes calculated from the tabulated data on the total cross-sections as a function ofneutron energy (Tremsin et al., 2013)

Figure 10 – Thermal neutron transmission radiographs obtained by grouping the energy-resolved images of different neutron spectra. The ranges ofneutron energies used to build each radiograph are shown in the respective legends (Tremsin et al., 2013)

composition of uranium is low (0.8 cm-1) and it is easyto transmit neutrons through thicker assemblies

➤ U-235 and U-238 have very different interactions withthermal neutron beams. Due to the 60 times highercross-section of U-235, it is very easy to distinguishbetween the two isotopes and to quantify the amountof the fissile isotope U-235

➤ Lead is used as shielding material around fuel samplesfor radiation protection purposes, and thus X-rayradiography fails in transmission experiments.Neutrons, on the other hand, penetrate lead shieldingwith a thickness of about 15 cm and allow neutronradiography investigations

➤ Additional substitutes in fuel compositions, which arein use as burnable poisons but are strong neutronabsorbers (e.g. B, Li-6, Dy, or Gd), are easily identifiedwith neutron methods

➤ After long-term exposure, hydrogen can be found inthe cladding outer region of fuel rods under somecircumstances. X-ray radiography fails to visualizethese material modifications because of the very lowcontrasts obtained for elements with with low atomicnumbers. Neutrons, on the other hand, have a highsensitivity for hydrogen, thus allowing quantificationof the hydrogen content in cladding.

IsotopesCharacterization of isotopic sources is a demanding anddifficult task due to their physical size and natural radioac-tivity, which makes visual inspection impossible. Hoffman(2012) investigated a small radioactive radium source usinghigh-resolution micro-focus X-ray tomography to determinewhether the sample contained a powdered form of radioactivematerial or whether it was solid. The source contained 20 mgof radium and was in a form of a needle with a diameter ofabout 1 mm and about 8 mm in length – all sealed within aglass tube. Figure 11 is an XCT tomogram of the ampouleshowing its serial number clearly on the outside, while Figure12 is a slice from the 3D tomogram revealing the position ofthe radioactive material inside the needle.

Valuable metrology quantitative information could bededucted from the 3D tomogram of the isotope (Table I). Themost important aspect for further processing of the isotope isthe quantification of the volume of radioactive salt present inthe needle.

Nuclear fuelA major field of neutron radiography application is theinspection of nuclear fuel and control rods, reactor materialsand components, and of irradiation devices for the testing ofnuclear fuels and materials. The fuel rods are used underextreme conditions such as very high power density,

temperature, pressure, and radiation level. Thermal neutronradiography investigations were conducted with the conven-tional film technique due to the radioactivity of the objects.The following issues are addressed through the use ofneutron radiography: (a) condition of the fuel assembly,including fuel rod condition, (b) detection of leaks such asingress of water, and (c) quality control, including functionaland dimensional evaluation and inspection of irradiationdevices and components. Figure 13 shows a neutronradiograph of a fuel pin with pelletized fuel as fabricated(Domanus, 1992).

Due to the high radioactivity of nuclear fuel afterirradiation, X-ray radiography cannot be used as an investi-gation technique. Investigations are done within a hot-celllaboratory set-up, which allows for the remote handling ofthe radioactive fuel (Klopper, De Beer, and Van Greunen,1998). A research reactor is normally an extension of a hot-cell laboratory, as neutron radiography is one of the majoranalytical probes in the post-irradiation examination (PIE) ofnuclear fuel. Typical findings using neutron radiography asan analytical probe on irradiated fuel pins are the condition ofthe fuel pellets and of the ZrTM -cladding material.

Fuel pellet investigations reveal fabricated conditionssuch as cracks, chips, change of shape or location, voids,inclusions, corrosion, nuclear properties, and coolant. Figure14 shows examples of these findings on film neutronradiographs.



Figure 11 – μXCT of a Ra isotopic source ampule (Hoffman, 2012)

Table I

Quantitative information deducted from the XCT

Description: Information Value Unit

Needle inscription 100 µgOfficial source activity 20 mCiContact dose rate 1.1 mSv/hDiameter of tube 1.67 mmDiameter of internal void 1.03 mmLength of tube 9.72 mmLength of internal void 7.57 mmWall thickness 0.34 mmLength of salt 3.01 mmInner volume 6.31 mm3

Volume of salt 2.51 mm3

Figure 12 – Slice of μXCT tomogram with enlarged end pugs revealingRa isotope internal information (Hoffman, 2012)


Domanus (1992) describes a number of other fuel pelletproperties revealed by neutron radiography, including centralvoids and the accumulation of Pu in the central void. Fuel rodinspections include deformed cladding, hydrides in cladding,plenum and spring, dislocated disks, condition of the bottomplug, and a picture of a melted thermocouple inside the fuelrod. Lehmann, Vontobel, and Hermann (2003) report on theextensive utilization of the NEUTRA and ICON neutronradiography (Nrad) facilities at Paul Sherrer Institute inSwitzerland, where a dedicated detection station is availablefor the inspection of irradiated fuel assemblies. Aspects suchas fuel enrichment, fuel poisoning, and hydrogen content inthe fuel cladding are being addressed and investigated byneutron radiography. Due to the importance of fuel claddinginvestigations, the utilization and function of neutronradiography is addressed in the following paragraph.

Hydrogen embrittlementIt is well known that hydrogen agglomeration is deleteriousin any material. More than a few hundred ppm hydrogen inthe cladding surface of fuel rods at compromises thestructural stability of the cladding tube significantly, with theconsequence of possible failure, especially when mechanicalloading is also involved. The ability of neutrons to penetrateuranium is considerably higher than for X-rays and allowsfor the structures of the nuclear fuel rods to be inspected.Furthermore, the probability of neutron interaction withhydrogen is very high, while for X-rays it is effectively zero.This allows neutron radiography to be effectively utilized for

the study of even small quantities of hydrogen ingress in thecladding, which is an important mechanism for claddingembrittlement, as depicted in Figure 15.

Furthermore, through proper characterization and withthe aid of digital radiographs, Nrad allows for the investi-gation of the absolute hydrogen content and its distribution.in situ investigations provide new information about thekinetics of hydrogen uptake during steam oxidation and ofhydrogen diffusion in zirconium alloys. Nrad-studies are theonly way to investigate and understand the phenomenon ofhydrogen ingress in the ZrTM cladding. A linear dependenceof the total macroscopic neutron cross-section on the H/ZrTM

atomic ratio, as well as on the oxygen concentration, wasfound, while no significant temperature dependence of thetotal macroscopic neutron cross-sections of hydrogen andoxygen was found, depending that zirconium and oxygen donot change their structures. Additionally, it was found thatrapid hydrogen absorption takes place in the absence of theoxide layer covering the metallic surface of the ZrTM cladding(Grosse et al., 2011). Figure 16 displays the results of in-situNrad investigations of hydrogen uptake during steamoxidation with the time dependence of hydrogen concen-tration of ZrTM-4 materials at 1273 K and higher, where avery rapid hydrogen uptake was found in the first couple ofseconds after the steam flow was switched on. At temper-atures of about 1273 K a phase transformation occurs and isaccompanied by a volume change and the formation of apronounced crack structure. When the cracks are formed, the

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Figure 13 – Neutron radiograph of nuclear fuel prior to irradiation (Domanus, 1992)

Figure 14 – Neutron radiographs (film prints) of irradiated nuclear fuel and their conditions (Domanus, 1992). Top: random cracks in pellets, middle: typicallongitudinal cracks in pellets

hydrogen uptake increases by nearly an order of magnitude(Grosse et al., 2008; Grosse, 2010). The decrease inhydrogen concentration is due to the consumption of the ß-ZrTM phase, which contains most of the absorbedhydrogen.

Pebble bed modular reactor (PBMR) fuelFigure 17 shows the composition of a 60 mm outer diameterhigh-temperature reactor (HTR) fuel pebble consisting ofthousands of 0.5 mm diameter low-enriched uranium oxidefuel particles with a tri-structural isotropic (TRISO) coating,embedded in a graphitic matrix. The pebble was analysedusing X-ray tomography technique prior to irradiation at the

SANRAD facility located at the SAFARI-1 nuclear researchreactor in South Africa. The aim of the investigation was toobserve the homogeneity of the TRISO particles within thecarbon matrix and to direct the manufacturing process toensure the centralization of the fuel within the carbon matrix

(Necsa, 2006). Figure 18 shows the misalignment of the fuelwithin the carbon matrix of the fuel pebble as well as thelocation and identification of a TRISO particle within the fuel-free zone. The inhomogeneous distribution of the TRISOparticles at the top of the fuel pebble can be clearly seen.Three-dimensional quantitative data of the misalignment of thefuel particles becomes available in the tomograms and ispresented in Figure 19, showing the extent of correction in X-,Y- and Z-directions to be introduced in the manufacturingprocess of the fuel pebble.

Lehmann, Vontobel, and Hermann (2003) reported thesuccessful application of neutron tomography to the 3Dscanning of PBMR fuel pebbles at the NEUTRA facility of theSINQ spallation source at PSI in Switzerland (see Figure 20). Asphere-type fuel element from the high-temperature reactor(HTR) programme was studied with neutron tomography. Thissample is 6 cm in diameter and contains about 8500 individualfuel pebbles (diameter 0.5 mm). No shielding of the fresh fuelelement was necessary for the tomographic inspection. Theinvestigation was aimed at the visualization of the 3D distri-bution of the fuel particles in the graphite matrix in order todetermine its uniformity and the fuel sphere’s content of fissilematerial.

TRISO fuel particles are an integral part of the fuel designfor current and future HTRs. A TRISO particle comprises fourconcentric spherical layers encasing a fuel kernel, namely thebuffer (porous carbon), inner pyrolytic carbon (IPyC), siliconcarbide (SiC), and outer pyrolytic carbon (OPYC) layers (seeFigure 17). Each layer performs specific functions. The fuelkernel, consisting of uranium or uranium carbide, provides thefissile material and retains some of the fission products. Thebuffer layer, a highly porous carbon structure, provides somefree volume for gaseous fission products, and protects the SiClayer from damage by high-energy fission products. The IPyClayer provides structural support for the subsequent SiC layerand prevents the chlorine compounds required for SiCdeposition interacting with the fuel kernel. The SiC layer forms



Figure 15 – Neutron radiograph of nuclear fuel and cladding material showing (black spots) hydrogen accumulation within the ZrTM tubing (Frajtag, n.d.)

Figure 16 – Neutron radiography (Nrad) investigation: kinetics ofhydrogen uptake and release during steam oxidation (Frajtag, n.d.)

Figure 17 – Composition of PBMR fuel (Weil, 2001)


the main diffusion barrier for fission products. It acts as apressure vessel, providing mechanical strength for the particleduring manufacture of the nuclear fuel compact or pebble bed.The OPyC layer protects the SiC layer during fuel fabrication asthe TRISO particle is pressed into a larger fuel compact orpebble.

Lowe et al., (2015) examined the applicability of multi-scale X-ray computer tomography (CT) for the non-destructivequantification of porosity and thickness of the various layers ofTRISO particles (see Figure 21) in three dimensions, andcompared this to the current destructive method involvinghigh-resolution SEM imaging of prepared cross-sections.

An understanding of the thermal performance andmechanical properties of TRISO fuel requires a detailedknowledge of pore sizes, their distribution, and interconnec-tivity. Pore size quantification (false color coding) and distri-bution in an X-ray tomogram of the SiC (D) and OPyC (E)layers within a TRISO particle is shown in Figure 22.

Direct comparison with SEM sections indicates thatdestructive sectioning can introduce significant levels of coarsedamage, especially in the pyrolytic carbon layers. Since it isnon-destructive, multi-scale time-lapse X-ray CT opens thepossibility of intermittently tracking the degradation of TRISOstructure under thermal cycles or radiation conditions in orderto validate models of degradation such as kernel movement. X-ray CT in situ experimentation on TRISO particles under loadand temperature could also be used to understand the internalchanges that occur in the particles under accident conditions.

Research reactor control rod verificationNrad is being applied as a verification and analytic techniqueat the SAFARI-1 nuclear research reactor on the control rods

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Figure 18 – X-ray tomography of a PBMR fuel pebble. Left: the non-centralized fuel sphere within the carbon matrix. Right: Location andidentification of a TRISO particle inside the fuel-free zone (Necsa, 2006)

Figure 19 – Graphical presentation of the deviation of the fuel zone of aBPMR pebble from the centre of the carbon matrix in three dimensions(Necsa, 2006)

Figure 20 – Neutron tomogram generated at PSI, Switzerland showingthe exact location and homogeneity of the TRISO (Lehmann, Vontobel,and Hermann, 2003)

Figure 21 – Micro-focus X-ray radiograph of a TRISO particle (Necsa)

Figure 22 – Xradia VersaXRM orthoslice showing the SiC (D) and OPyC(E) layers with the pore volumes superimposed. Pore diameter is colour-coded to identify large pore clustering (Lowe et al., 2015)

prior to their installation in the core of the reactor. Thequality control assurance test entails the verification of theneutron attenuation cross-section of the control rod against astandard consisting of Cd. The inspection entails a visualclarification of the attenuation of the thermal neutrons byinspection of the neutron radiograph of the control rod.Additionally, due to the digital radiography capability of theneutron camera detection system, the first-order neutrontransmission calculation can be made using the pixelgreyscale values on the radiographs of both the standard andcontrol rod sample. Pixel greyscale values represent a linearrelationship in the neutron attenuation of materials. In thisinstance a dramatic decline in greyscale pixel values is seendue to the high thermal neutron absorption by the Cd section(μρn =115.11cm-1) of the control rod. Neutron radiographs ofthe Cd standard and a control rod are depicted in Figure 23.

Radioactive wasteLow- and intermediate-level nuclear waste is normallyencapsulated in in some form of barrier to protect the wastefrom the environment and vice versa. Intermediate-levelnuclear waste is firstly encapsulated in a steel drum,compressed, and finally embedded normally in a concrete drumand safely stored underground in a remote location such asVaalputs in the Karoo region in South Africa (Necsa, n.d. (b))(see Figure 24). A site is normally chosen with low rainfall andsuitable surface and groundwater conditions.

Concrete is a porous medium and the characterization oftransport of water through concrete structures is well describedby De Beer, Strydom and Griesel (2004) and De Beer, Le Rouxand Kearsley (2005). It is especially important to understandthe transport of water through concrete because nearly allconcrete structures contain steel reinforcing, and in the case ofnuclear waste, intermediate-level nuclear waste in compressedsteel drums. When cracks in the concrete, caused by thetransport of liquid through it, reach the reinforcing, anenvironment conductive to the corrosion of steel is created.Corrosion affects the strength of the structural members, as thesteel is a major contributor to the tensile and compressivestrength of the members. Severe leakage of radioactivematerials into the surrounding environment is thus possible ifthe integrity of the concrete barrier is compromised.

Neutron radiography studies of concrete and mortarsenable the direct physical visualization and quantitativedetection of water inside concrete structures. The physical

properties of concrete such as porosity, permeability, andsorping characteristics are obtained through applying neutronradiography as a non-destructive analytic tool. The aim ofthese investigations is to maximize the properties to preventwater sorption and leaching of concrete structures and optimizeone of the physical properties which is sometimes neglected inthe criteria to develop structures for nuclear waste encapsu-lation (De Beer, Strydom and Griesel, 2004 ; De Beer, Le Rouxand Kearsley, 2005). To improve the durability of concrete, thecapillary and pore size within the concrete matrix must berestricted to a minimum. This is why hydration as well as W/Cratio properties is of great importance, and creates thus anideal opportunity for neutron radiography to play a role inobtaining the needed information in a non-destructive mannerto optimize these parameters. The visualization of the sorptionof water by means of neutron tomography of a laboratory-sizeconcrete structure is depicted in Figure 25.

ConclusionsX-ray and neutron radiography in two or three dimensionsplay an important role in many dedicated areas within thenuclear fuel cycle. The advantage of these methods is theircompletely non-destructive nature. Visualization of thestructure of samples, as well as quantitative description, areimportant aspects in materials research. The important roles ofX-ray and neutron radiography/tomography as non-invasiveanalytic techniques within specific areas within the nuclear fuel



Figure 23 – Neutron radiographs of a Cd standard (left) and control rod(right) showing Cd (black) indicating high thermal neutron absorbingmaterials (Necsa, 2010 (a))

Figure 25 – Neutron radiographs showing the effect of 70%, 60%, and50% W/C ratio on the sorptivity of water into a concrete slab (De Beer,Strydom and Griesel, 2004)

Figure 24 – Vaalputs intermediate-level waste storage site, South Africa(Necsa)


cycle should not be underestimated. X-rays and neutrons areproduced by very different methods, and also interact withmaterials in different manners. In the nuclear environment,each type of radiation has its own field of utilization due totheir different characteristics, but in some instances theirapplications complement each other to reveal comprehensiveinformation.

Neutron transmission analysis is a very helpful tool toobtain information on the properties of, and changes in,nuclear fuel material. Scientists and researchers in thegeosciences in South Africa have, in the availability of thetomography facilities at Necsa, the capabilities to conductquantitative analytical measurements at state-of-the-artradiation imaging facilities that compare to similar facilitieselsewhere in the world.

Within the mining area, 3D computer tomography showspotential for further development, and can be already used tocomplement and add value to current conventional 2Dmineralogical techniques. Neutron radiography analysis is ableto derive the hydrogen content in fuel cladding both qualita-tively and quantitatively, with high sensitivity and precision.

The results presented here illustrate how recent advancesin laboratory-based X-ray CT instruments allow theexamination of TRISO particles at the nano- and micro-scalesin 3D. In this case study, high-resolution X-ray CT has beenshown to be a viable tool for profiling the TRISO particles intwo important aspects; to characterize the individual TRISOlayers with variations in thickness and their subsequentinteractions, thus allowing manufacturing validation as well asassisting in working towards a mechanistic understanding offabrication and in-service issues.

The availability of these techniques in South Africa opensnew possibilities for research, quantitative analysis, and non-destructive evaluation. National capacity as well as interna-tional trends shows the ability for non-destructive testing ofnuclear materials utilizing penetrating X- ray- and neutronradiation in more comprehensive and unique ways than before.

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R.B., and DE BEER, F.C. 2011. The use of 3D X-ray computed tomographyfor gold location in exploration drill cores. Proceedings of the 10thInternational Congress for Applied Mineralogy (ICAM), Trondheim, Norway,1-5 August 2011. pp.129–136. http://www.mintek.co.za/wp-content/uploads/2011/11/ch-17.pdf

DAVIES, L.M. 2000. Role of NDT in condition based maintenance of nuclear powerplant components. 15th World Conference on Non-Destructive Testing,Rome, 15-21 October 2000.http://www.ndt.net/article/wcndt00/papers/idn078/idn078.htm

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DE BEER, F.C., MIDDLETON, M.F., and HILSON, J. 2004. Neutron radiography ofporous rocks and iron ore. Applied Radiation and Isotopes, vol. 61. pp. 487–495.

DE BEER, F.C., STRYDOM, W.J., and GRIESEL, E.J. 2004. The drying process ofconcrete: a neutron radiography study. Applied Radiation and Isotopes, vol.61, no. 4. pp. 617–623.

DE BEER, F.C., LE ROUX, J.J., and KEARSLEY, E.P. 2005. Testing the durability ofconcrete with neutron radiography. Nuclear Instruments and Methods inPhysics Research A, vol. 542. pp. 226– 231.

DOMANUS, J.C. 1992. Practical Neutron Radiography. Kluwer Academic Publishers. FRAJTAG, P. Not dated. Radiation Protection and Radiation Applications: Gamma

and Neutron Radiography.http://moodle.epfl.ch/pluginfile.php/1593971/mod_resource/content/2/RRA-EPFL- FS2014-Week14a.pdf [Accessed 1 May 2015].

GHOSH, J.K., PANAKKAL, J.P., and ROY, P.R., 1983. Monitoring plutoniumenrichment in mixed-oxide fuel pellets inside sealed nuclear fuel pins byneutron radiography. NDT International, vol. 16, no. 5, October 1983. pp.275–276. DOI: 10.1016/0308- 9126(83)90127-X

GROSSE, M. 2010. Neutron radiography—a powerful tool for fast, quantitative andnon- destructive determination of the hydrogen concentration and distri-bution in zirconium alloys. Proceedings of the 6th International ASTMSymposium on zirconium in the Nuclear Industry, Chengdu, China, 9–13May 2010.

GROSSE, M., VAN DEN BERG, M., GOULET, C., LEHMANN, E., and SCHILLINGER, B. 2011.in situ neutron radiography investigations of hydrogen diffusion andabsorption in zirconium alloys. Nuclear Instruments and Methods in PhysicsResearch, vol. A651. pp. 253–257. doi: 10.1016/j.nima.2010.12.070

GROSSE, M., STEINBRUECK, M., LEHMANN, E., and VONTOBEL, P. 2008. Kinetics ofhydrogen absorption and release in zirconium alloys during steamoxidation. Oxidation of Metals, vol. 70, no. 3. pp. 149–162.

GRÜNAUER, F. 2005. Design, optimization, and implementation of the new neutronradiography facility at FRM-II. Dr. Rer. Nat. dissertation, Faculty of Physics,Technischen Universität München.

HOFFMAN, J.W. 2012. Process description for micro-focus X-ray investigation ofsource at MIXRAD facility. Necsa internal report, DOC NO: RS-MFX-PRO-12002.

INTERNATIONAL ATOMIC ENERGY AGENCY. Not dated. Getting to the Core of the NuclearFuel Cycle. Department of Nuclear Energy, Vienna, Austria.https://www.iaea.org/OurWork/ST/NE/NEFW/_nefw-documents/NuclearFuelCycle.pdf

ISO 15708-1:2002(E). Non-destructive testing — Radiation methods —Computed tomography — Part 1: Principles.

KLOPPER, W., DE BEER, F.C., and VAN GREUNEN, W.S.P. 1998. Overview of hot cellfacilities in South Africa. Proceedings of the European Working Group ‘HotLaboratories and Remote Handling’.

LEHMANN, E.H., VONTOBEL, P., and HERMANN, A. 2003. Non-destructive analysis ofnuclear fuel by means of thermal and cold neutrons. Nuclear Instrumentsand Methods in Physics Research A, vol. 515, no. 3. pp. 745–759. doi:10.1016/j.nima.2003.07.059

LOWE, T., BRADLEY, R.S., YUE, S., BARII, K., GELB, J., ROHBECK, N., TURNER, J., andWITHERS, P.J. 2015. Microstructural analysis of TRISO particles using multi-scale X-ray computed tomography. Journal of Nuclear Materials, vol. 461.pp. 29–36. http://dx.doi.org/10.1016/j.jnucmat.2015.02.034

MCGLINN, P.J., DE BEER, F.C., ALDRIDGE, L.P., RADEBE, M.J., NSHIMIRIMANA, R.B.,BREW, D.R.M., PAYNE, T.E., and OLUFSON, K.P. 2010. Appraisal of acementitious material for waste disposal: neutron imaging studies of porestructure and sorptivity. Cement and Concrete Research, vol. 40. pp. 1320–1326

NECSA. 2006. RT-TVG-06/05: Test Report: X-Ray Radiography of PBMR-FuelSpheres with Zirconium Particles. Internal report, SAMPLE NO: DFS-T-F03G04.

NECSA. 2010 (a). RS-TECH-REP-10004: Neutron Radiography Quality Assurancetest report of neutron absorbing material in Control Rods of the SAFARI-1Nuclear Research Reactor (Lot/Batch No: RT-LOT-10/03). Internal report.

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IntroductionMaterials characterization is central inunderstanding the relationship between thestructure, properties, and performance in orderto engineer materials that fit the performancecriteria for specific applications. This isconveniently represented in the form of atetrahedron, generally known as the materialscience paradigm (Figure 1).

The availability of advanced characteri-zation techniques is integral to the developmentof advanced metals, not only duringdevelopment phases, but in the form ofmanufactured components. At Necsa (SouthAfrican Nuclear Energy Corporation), modernX-ray and neutron diffraction instruments arenow accessible to the South African researchand industrial communities. These facilitiesenable nondestructive investigations ofmaterials and components that could addsignificant downstream value to the anticipatednuclear industry development activities.

BackgroundApproximately 95% of solid materials can bedescribed as crystalline. By applying the waveproperties of X-rays (keV energies asgenerated in laboratory-based units, or high-power synchrotron facilities) and neutrons(meV energies) having wavelengths of thesame order as atomic spacings, i.e. 10-10 m,unique information can be obtained. Theirinteraction with the crystalline ordering leadsto constructive interference described byBragg’s law of diffraction. This leads tocharacteristic diffraction patterns for eachphase that essentially are fingerprints of thechemical phase content (Fourier transform ofthe atomic arrangement). Both the peakpositions (corresponding to lattice planespacings) and the relative intensities(corresponding to the atomic species andarrangement in the unit cell) in the diffractionpatterns are indicative of specific phases. X-ray photons scatter through an electromagneticinteraction with the electron charge cloud ofthe material, while neutrons are scattered byinteraction with the nuclei. In general theinteraction strength of X-rays with matter isdirectly related to the atomic number of thematerials being investigated, whereas neutronscattering lengths are approximately equal inmagnitude for most atoms. Apart from theinteraction strength differences, theirpenetration depths are also dependent on theatomic species, as summarized in Table I for anumber of technologically important materials.

The different interaction mechanisms ofneutrons and X-rays with matter offer comple-mentary techniques for investigatingcrystalline materials at the microstructural

Non-destructive characterization ofmaterials and components with neutronand X-ray diffraction methodsby A.M. Venter*

SynopsisThe availability of advanced characterization techniques is integral tothe development of advanced materials, not only during developmentphases, but in the manufactured components as well. At Necsa, twomodern neutron diffractometers equipped with in-situ sampleenvironments, as well as complementary X-ray diffraction instruments,are now available as User Facilities within the National System ofInnovation in support of the South African research and industrialcommunities. Neutrons and X-rays, owing to their different interactionmechanisms with matter, offer complementary techniques for probingcrystalline materials. Both techniques enable nondestructive investigationof phenomena such as chemical phase composition, residual stress, andtexture (preferred crystallite orientation). More specifically, the superiorpenetration capabilities of thermal neutrons into most materials allows forthe analysis of bulk or localized depth-resolved properties in a widevariety of materials and components. Materials that can be investigatedinclude metals, alloys, composites, ceramics, and coated systems. Inparticular, depth-resolved analyses using neutron diffraction complementssurface investigations using laboratory X-rays in many scientific andengineering topics. The diffraction techniques can add significantdownstream value to the anticipated nuclear industry developmentactivities.

Keywordsresidual stress; crystallographic texture; chemical phase identification,neutron and X-ray diffraction.

* Research and Development Division, The SouthAfrican Nuclear Energy Corporation SOC Ltd.,(Necsa), Pretoria, South Africa.




Non-destructive characterization of materials and components

level. Both techniques enable nondestructive investigation ofphenomena such as chemical phase composition, residualstress, and texture (preferred crystallite orientation).Materials include metals, alloys, composites, ceramics, andcoated systems. In particular, depth-resolved analysis usingneutron diffraction complements surface investigations usinglaboratory X-rays in many scientific and engineeringdisciplines.

Neutrons complement X-rays due to the followingproperties:

➤ Neutrons are electrically neutral. This enables orders ofmagnitude deeper penetration of bulkmaterial/components:• Allows nondestructive bulk analysis• Ease of in situ experiments, e.g. variable

temperature, pressure, magnetic field, chemicalreaction, etc.

➤ Neutrons detect light atoms even in the presence ofheavy atoms (organic crystallography) – especiallyhydrogen. This property has been decisive in theinvestigation of high-temperature superconductors

➤ Neutrons distinguish atoms adjacent in the periodictable, and even isotopes of the same element (changingscattering picture without changing chemistry). This isparticularly applicable to the transition metal series

➤ Neutrons have a magnetic moment. This enablesnondestructive investigation of magnetic phenomenafrom direct observation of the reciprocal lattice.

Neutrons of adequate flux for neutron diffractionapplications are produced as a by-product of the fission of235U in neutron research reactors (such as SAFARI-1operated by Necsa in South Africa), or in accelerator-basedfacilities when very high-energy protons strike a targetproducing ‘spallation’. These high-energy (MeV) neutronproducts are then thermalized and filtered to the thermalenergy range (meV).

The neutron diffraction instruments at SAFARI-1 operatein constant wavelength mode where a highly monochromaticthermal neutron beam (<1% wavelength spread withwavelengths selectable in the range 1.0–2.0 Å, typically 25meV energies) with a flux in the order of 106 neutrons persquare centimetre per second is extracted from the fissionenergy spectrum and directed to a sample. Bragg’s law ofelastic diffraction

nλ = 2dhklsinθhkl

describes the geometrical condition for coherent diffraction thatcan be measured to high precision on a diffractometer. In thisequation λ is the monochromatic wavelength in Å (10-10 m),dhkl is the interatomic spacing between the parallel crystalliteplanes of Å dimension (hkl refers to the Miller notation of thecrystal planes) and θhkl is the angle at which the diffracted peakis measured. In the angular dispersive operational mode of theSAFARI-1 neutron diffraction instruments, the wavelength isselected with the instrumental setup and thus accuratelyknown, with dhkl and θhk being the only variables. The latter ismeasured experimentally to high precision with the diffractioninstrument, which comprises a goniometer for samplepositioning, and a detector that is precisely rotated around thegoniometer axis in the horizontal plane. The diffracted intensityis measured with a high-sensitivity neutron detector that uses3He as ionization medium for the accurate measurement of thediffraction angles from all coherently scattered Bragg peaksfrom which respective values can be calculated.

As is the case with X-rays, all crystalline materials (evenchemically multi-phased materials) placed in the neutron beamproduces a diffraction pattern. This document provides anoverview of a subset of diffraction techniques that may be ofrelevance to the materials beneficiation aims of the AdvancedMaterials Initiative.

▲


Figure 1 – Material science paradigm

Table I

Neutron and X-ray scattering parameters for a number of elements that comprise the common engineeringalloys. Half attenuation lengths correspond to the thickness of material that reduces the intensity by 50%. 1.5 ÅX-rays correspond to Cu radiation (8 keV). 0.15 Å X- rays correspond to synchrotron radiation (100 keV)

Element Neutron coherent X-ray scattering Half attenuation lengths scattering length (bc) length 1.8 Å neutrons 1.5 Å X-rays 0.15 Å X-rays

(f0 at sin θ/λ = 0.2 [mm] [μm] [mm ][fm = 10-15 m ] [fm = 10-15 m]

Mg 5.375 0.246 43 100 24Al 3.449 0.258 66 52 15Ti -3.37 0.451 12 11 6Fe 9.54 0.565 20 20 2.4Co 2.49 0.594 2 14 2Ni 10.3 0.624 3 16 1.8Zr 7.16 0.884 24 8 1.1Hf 7.77 1.715 1 3 0.1W 4.86 1.762 5 2 0.1U 8.417 2.172 9 1 0.2

ApplicationsResults from selected typical applications pursued at theNecsa facilities are reported to provide an indication of thepotential value adding that these techniques could offer.

Chemical phase identificationThe most widespread use of powder (polycrystalline)diffraction is in chemical phase analysis. This encompassesphase identification (search/match), investigation of high-and low-temperature phases, solid solutions, and determi-nations of unit cell parameters of new materials. A multi-phase mixture, e.g. a soil sample, will show more than onepattern superposed, allowing for determination of the relativeconcentrations of phases in the mixture.

The powder diffraction method is well suited for charac-terization and identification of polycrystalline phases. This isdone against the International Center Diffraction Data (ICDD)database, which contains in excess of 50 000 inorganic and25 000 organic phases. Figure 2 shows a neutron diffractionpattern of a multi-phased sintered iron powder sample thatwas investigated to quantify the constituent phases. Thequantification was done with the Rietveld profile refinementtechnique (Rietveld, 1969; Taylor, 2001), in which atheoretical line profile (a combination of the crystal systemand instrumental model) is matched in a least-squaresrefinement approach to the measured data. Such quantitativephase analysis is now routinely used in industries rangingfrom cement manufacture to the oil industry and can providedetection limits of 1 weight per cent (wt.%).

Residual stressThe total stress that a component experiences in practical useis the vector sum of the applied and residual stresses. Theapplied stresses result from the loading forces in use and canbe calculated to high precision. Residual stresses, consistingof locked-in stress that remains in a material or componentafter the external forces that caused the stress have beenremoved, are mostly only approximated qualitatively. Thesestresses can be introduced by any mechanical, chemical, orthermal process, such as machining, plating, and welding.Stress analysis by diffraction techniques is based on accuratemeasurement of lattice strain distributions (variations ininteratomic spacing). Using Hooke’s law, stresses are

calculated from the strain distributions. Residual stresses area double-edged sword in material science applications.Compressive residual stresses are beneficial in applicationswhere fatigue performance is required, being able to mitigatecrack initiation and propagation. Tensile residual stresses aregenerally considered to be detrimental as they can lead tocrack initiation and propagation. Specifically, residual stresstailoring can render substantial improvements in theoptimization of component design.

X-ray and neutron radiation enable nondestructiveprobing at different penetration depths (Hutchings et al.,2005; Fitzpatrick and Lodini, 2003; Reimers et al., 2008;Webster, 2000; Ohms et al., 2008; Engler and Randle, 2010)(Table I). The stress tensors are obtained from the measuredstrain tensors and application of Hooke’s law of elasticity

where S1 and 1/2S2 are the diffraction elastic constants.Examples from recent investigations performed using

neutrons are presented to illustrate potential applications.

Welded mild steel plateAs part of an investigation into the influence of differentialhardness between the weld metal and base metal on residualstress and susceptibility to stress corrosion cracking, acomprehensive two-dimensional (2D) mapping of theresidual stress field, through the 17 mm plate thickness andacross the weld was completed. Figure 3a indicates the 2Dmap of the longitudinal stress component, which ismaximally influenced due the differential longitudinalcontraction. Values can be as large as the yield stress of thematerial. Figure 3b indicates the 2D map of this stresscomponent after the sample had undergone a post-weld heattreatment. This treatment has been very successful in havingcompletely relaxed the tensile stress values. Such analysesare only possible owing to the penetrating capabilities ofthermal neutrons and the nondestructive nature of themeasurement technique.

Laser shock-peened aluminium plateLaser shock-peen (LSP) is an emerging cold work processused to induce compressive residual stresses in metallic



Figure 2 – Neutron powder diffraction results (measured diffracted intensities as function of diffraction angle) of a mixed-phase sintered iron sample,shown as dots (underneath the red solid line at positive intensities). The multiphase quantification results from a Rietveld analytical approach (red solid lineat positive intensities) are indicated in the legend. The curve along the zero intensity level depicts the difference between the measured and analysed dataand gives the goodness-of-fit. Vertical lines at negative intensities represent the Bragg peaks (hkl) corresponding to each phase

components to depths up to 1 mm. The purpose of thistreatment is to reduce or remove tensile stresses in thesurface region to improve the fatigue performance. Figure 4summarizes the through-thickness in-plane residual stressdistribution in an as-rolled aluminium AA6056-T4 plate,taken as the reference, and the stress distribution after lasershock-peen using a 3 GW laser beam. The net effect, purelydue to the peen action, was determined by subtracting thereference values.

Laser-welded aluminium with laser shock peentreatmentLaser beam welding is a newly established joining technologyusing no filler material. Similar to other welding techniques,

it introduces adverse tensile residual stresses as well ascomponent distortions. Figure 5 indicates 2D maps of theresidual stress field measured through a 3.3 mm thickAA6056-T4 aluminium plate and across the weld.

Here again, the longitudinal stress component is mostdominant, as shown in Figure 5a. To improve this adversestress condition, the weld region has been subjected to lasershock-peen treatment using a 3 GW laser pulse. Compared tothe post-weld heat treatment results given previously, thelaser treatment has completely altered the tensile stress so asto be substantially compressive.

Additive-manufactured Ti-6Al-4VIn the selective laser beam melt (SLM) additive manufac-


▲


Figure 3 – Two-dimensional maps of the longitudinal residual stress component in SA 516 Gr 70 pressure vessel steel welded plates: (a) as-weldedcondition, (b) post-weld heat-treated condition. The weld centre-line is at 0 mm

Figure 4 – Through-thickness stress profile in aluminium plate: (a) parent material, (b) laser shock-peened plate, (c) net contribution by laser shock-peen

Figure 5 – 2D map of the longitudinal residual stress component in aluminium laser beam welded plates: (a) as-welded condition, (b) post-weld-treatedcondition after laser shock-peen treatment. The weld centre-line is at 0 mm

turing process, a high-power laser beam is used to melt anddeposit successive layers of powder to form complex three-dimensional metal parts. The highly localized heat inputleads to large thermal gradients. This in turn producescomplex residual stress states inside the part. Figure 6 showsa stress map measured in the centre of the sample perpen-dicular to the build direction in a Ti-6Al-4V sample. Thisreveals the existence of tensile residual stresses in the near-surface regions, while the central region of the sample is incompression.

Texture analysisMost solid-state matter has a polycrystalline structurecomposed of a multitude of individual crystallites or grains.The crystallographic orientation can have variousarrangements, ranging from completely random to thedevelopment of preferred alignment. The significance of thistexture lies in the corresponding anisotropy of many materialproperties. The influence could be as much as 20–50% of theproperty value. X-ray and neutron diffraction analysismethods are well established, rendering results referred to as



Figure 6 – Measured 2D stress component mapped perpendicular to the build direction in a SLM produces Ti-6Al-4V sample: (a) sample geometry showingthe ‘internal plane’ investigated nondestructively, (b) 2D residual stress map

Figure 7 – Pole figure representations of selected reflections from two ferritic stainless steel samples subjected to different heat treatments, measuredwith neutron diffraction


macro or bulk texture (Engler and Randle, 2010). Analysescan also be supplemented by methods whereby individualorientations are measured by transmission or scanningelectron microscopy and are directly related to themicrostructure, which has given rise to the term‘microtexture’.

The preferred orientation is usually described in terms ofpole figures. The inverse pole figure gives the probability offinding a given specimen direction parallel to crystal (unitcell) directions. By collecting data for several reflections andcombining several pole figures, the complete orientationdistribution function (ODF) of the crystallites within a singlepolycrystalline phase that makes up the material can bedetermined (Engler and Randle, 2010).

Shown in Figure 7 are bulk pole figures measured withneutron diffraction on ferritic stainless steel specimens thathave been subjected to different heat-treatment parametersafter rolling. Significant texture development is evident,which will contribute to anisotropic properties that could leadto physical phenomena such as earing, warping, and generaldimensional changes and distortions.

ConclusionsThese examples serve to demonstrate the capabilities of theX-ray and neutron diffraction techniques available at Necsafor the characterization of new materials and advancedbeneficiation techniques to enhance component performancein various applications. The vast field of applications can bereviewed in the extensive literature references provided, aswell as recent publications from Necsa (Taran et al., 2014;Troiano et al., 2012; Venter, Luzin, and Hattingh, 2014;Venter et al., 2008, 2013, 2014a, 2014b; Zhang et al., 2008).Due the nondestructive nature of the techniques, samples canbe investigated at various stages of their manufacture orutilization lifetimes.

AcknowledgmentsSpecific acknowledgment is attributed to the various co-workers from Necsa and academia and their willingness thatthe results may be used in this document: Professor Johan deVilliers (University of Pretoria), Deon de Beer (University ofPretoria), Victoria Cain (University of Cape Town), DanielGlaser (University of the Witwatersrand), and Deon Maraisfrom Necsa.

References

ENGLER, O. and RANDLE, V. (eds). 2010. Introduction to Texture Analysis,Macrotexture, Microtexture, and Orientation Mapping. CRC Press, Taylorand Francis.

FITZPATRICK, M.E. and LODINI, A. 2003. Analysis of Residual Stress byDiffraction using Neutron and Synchrotron Radiation. Taylor and Francis.

HUTCHINGS, M.T., WITHERS, P.J., HOLDEN, T.M., and LORENTZEN, T. (eds). 2005Introduction to the Characterization of Residual Stress by NeutronDiffraction. Taylor and Francis.

OHMS, C., MARTINS, R.V., UCA, O, YOUTSOS, A.G., BOUCHARD, P.J., SMITH, M.,KEAVEY, M., BATE, S.K., GILLES, P., WIMPORY, R.C., and EDWARDS, L. 2008.Proceedings of 2008 ASME Pressure Vessels and Piping Conference (PVP2008), Chicago, Illinois, 27–31 July 2008. (PVP2008-61913).

REIMERS, W., PYZALLA, A.R., SCHREYER, A., and CLEMENS, H. (eds). 2008.

Neutrons and Synchrotron Radiation in Engineering Materials Science.

Wiley-Vch Verlag GmbH.

RIETVELD, H.M. 1969. A profile refinement method for nuclear and magnetic

structures. Journal of Applied Crystallography, vol. 2, no. 2. pp. 65–71.

doi:10.1107/S0021889869006558

TARAN, Y., BALAGUROV, A., SABIROV, B., DAVYDOV, V., and VENTER, A. 2014.

Neutron diffraction investigation of residual stresses induced in niobium-

steel bilayer pipe manufactured by explosive welding. Materials Science

Forum, vol. 768-769. pp. 697-704.

doi:10.4028/www.scientific.net/MSF.768-769.697

TAYLOR, J.C. 2001. Rietveld made easy: a practical guide to the understanding of

the method and successful phase quantifications. Sietronics Pty. Ltd.,

Canberra.

TROIANO, E., UNDERWOOD, J.H., VENTER, A.M., IZZO, J.H., and NORRAY, J.M. 2012.

Finite element model to predict the reverse loading behavior of

autofrettaged A723 and Hb7 cylinders, Journal of Pressure Vessels and

Piping, PVT-11-1204, 041012-1.

VENTER, A.M., LUZIN V., and HATTINGH, D.G. 2014. Residual stresses associated

with the production of coiled automotive springs. Material Science Forum,

vol. 777. pp. 78–83. doi:10.4028/www.scientific.net/MSF.777.78

VENTER, A.M., OLADIJO, O.P., CORNISH, L.A., and SACKS, N. 2014a.

Characterisation of the residual stresses in HVOF WC-Co coatings and

Substrates. Material Science Forum, vol. 768-769. pp. 280–285.

doi:10.4028/www.scientific.net/MSF.768-769.280

VENTER, A.M., LUZIN, V., OLADIJO, O.P., CORNISH, L.A., and SACKS, N. 2014b. Study

of interactive stresses in thin WC-Co coating of thick mild steel substrate

using high-precision neutron diffraction. Materials Science Forum, vol.

772. pp. 161–165. doi:10.4028/www.scientific.net/MSF.772.161

VENTER, A.M., OLADIJO, O.P., LUZIN, V., CORNISH, L.A., and SACKS, N. 2013.

Performance characterization of metallic substrates coated by HVOF

WC–Co. Thin Solid Films, vol. 549. pp. 330–339.

VENTER, A.M., VAN DER WATT, M.W., WIMPORY, R.C., SCHNEIDER, R., MCGRATH, P.J.,

and TOPIC, M. 2008. Neutron strain investigations of laser bent samples.

Materials Science Forum, vol. 571–572. pp. 63-68.

WEBSTER, G.A. (ed.). 2000. Neutron Diffraction Measurements of Residual

Stress in a Shrink-fit Ring and Plug. VAMAS Report No. 38.

ZHANG, S.Y., VENTER, A.M., VORSTER, W.J.J., and KORSUNSKY, A.M. 2008. High

energy synchrotron X-ray analysis of residual plastic strains induced in

shot peened steel plates. Journal of Strain Analysis, vol. 43, no. 4. pp.

229-241. ◆

▲


Introduction

Fluorine chemistry has always had a very closerelationship with the nuclear industry.Although the first mention of fluorspar wasrecorded in the sixteenth century, and thefundamentals had been well-developed by theSecond World War, it was only during theManhattan Project (Banks et al., 1994), whenalmost unlimited funds were made availablefor the large-scale industrialization of fluorineand related compounds, that the technologytook off. Expertise in both the chemistry andthe engineering aspects of fluorine is essentialfor the design and running of several of theunit processes in the nuclear fuel cycle.

The aim of this paper is to give a briefoverview of the role of fluorine in the nuclearfuel cycle, and to outline the current SouthAfrican fluorochemical capability.

Hydrogen fluoride and elemental fluorineThe major barrier to entry into the fluoro-chemical industry is the ability to manufactureand, in general, to handle fluorine and itsprecursor, hydrogen fluoride. Both are extraor-dinarily difficult and dangerous substanceswith which to work.

Hydrogen fluoride (HF) is produced by thereaction of fluorspar (calcium difluoride) withsulphuric acid:

[1]

The reaction is endothermic and reactorsare generally run at temperatures above 200°C.HF is a clear liquid, with a boiling point of19.6°C. It readily dissolves in water and, in itsaqueous form, is known as hydrofluoric acid.Unlike the other common mineral acids, it isweak acid, thus does not readily deprotonate.Hydrofluoric acid is distinguished by its abilityto dissolve glass, and as a consequence cannotbe used in ordinary laboratory glassware.

Both hydrogen fluoride and hydrofluoricacid are enormously hazardous substances(Bertolini, 1992; Smith, 2004). Upon contact,human skin is not immediately burnt by theaction of the hydronium ion; rather, because ofthe small size of the HF molecule, it diffusesthrough the skin and precipitates andinactivates biological calcium and magnesiumsubcutaneously, causing tissue necrosis. Thewounds are extremely painful, and difficult totreat. In general the dead flesh has to be

Fluorine: a key enabling element in thenuclear fuel cycle by P.L. Crouse

SynopsisFluorine – in the form of hydrofluoric acid, anhydrous hydrogen fluoride,elemental gaseous fluorine, fluoropolymers, volatile inorganic fluorides,and more – has played, and still plays, a major role in the nuclear industry.In order to enrich uranium, the metal has to be in the gaseous state. Whilemore exotic methods are known, the standard and most cost-competitiveway of achieving this is by means of uranium hexafluoride (UF6). Thiscompound sublimates at low temperatures, and the vapour is enrichedusing centrifugal processes. The industrial preparation of uraniumhexafluoride requires both elemental fluorine gas and anhydroushydrogen fluoride (HF). HF is prepared by the reaction of sulphuric acidwith fluorspar (CaF2). Fluorine gas in turn is prepared by the electrolysisof HF. Yellowcake is first converted to uranium tetrafluoride (UF4), usingHF, after which the compound is treated with fluorine to yield UF6. Afterenrichment, the UF6 is reduced to UO2 for use in fuel elements in pelletform.

South Africa has the largest reserves of fluorspar internationally, andis the third largest producer after Mexico and China. Fluorine technologyhas many associated difficulties, because of the reactivity of fluorine andthe toxicity of HF. The main barriers to entry into the fluorochemicalindustry are thus the abilities to produce both HF and F2. Both thesesubstances are produced locally, at the industrial scale, at Pelchem SOCLtd. Should South Africa contemplate developing its own nuclear fuelcycle as part of the awaited new-build nuclear project, it will be imperativeto leverage the existing skills with respect to fluorine technology, residentat both Pelchem and Necsa, for this purpose.

This paper summarizes the fluorochemical skills developed locally overthe past several decades, and suggests strategies for maintaining thetechnology base and developing it for the next generation of scientists andengineers.

Keywordsflourine, hydrogen flouride, nuclear fuel cycle.

* Fluoro-Materials Group, Department of ChemicalEngineering, University of Pretoria.




Fluorine: a key enabling element in the nuclear fuel cycle

surgically removed, and a calcium gluconate solution injectedinto the tissue underneath the wound to prevent deeperdiffusion of the HF. A typical HF wound is shown in Figure 1.

HF is also very corrosive, and materials selection iscritically important in the development of an HF-relatedindustrial process.

Although HF was first synthesized in the eighteenthcentury, and was known to contain an element, it was only inthe late nineteenth century that fluorine was first isolated.This was accomplished by Henri Moissan (Argawal, 2007).Because it is a weak acid, anhydrous HF (AHF) does notconduct electrical current. Moissan’s discovery was that themolten salt KF.xHF does indeed conduct electrical current andcan be electrolysed. As a melt, e.g. with x=2, it dissociatesand forms the equilibrium:

[2]

Moissan (Figure 2) was awarded the Nobel Prize inChemistry in 1906. Since the first isolation of fluorine, theelectrolysis process has undergone a few technical changes,but has remained more or less static throughout the past fewdecades. Comprehensive descriptions of the technology canbe found in Slesser and Schram (1951), Rudge (1971), andShia (2004).

Fluorine itself is the most reactive element in the periodictable, and reacts with all other elements, excluding only the

two noble gases helium and neon (Cotton et al., 2007). Ingeneral the reactions of fluorine are highly exothermic, andbecause of its reactivity, materials of construction are ofcritical importance for safe operation. In general, expensivenickel-containing alloys are required. It should be noted thatHF sells for US$1–3 per kilogram, while F2 sells forUS$15–20 per kilogram. The high cost of fluorine can beattributed to electrical requirements and the highmaintenance costs of the electrolysis cells.

South Africa is richly endowed with fluorspar (Roskill,2009). Relative production and reserve figures are given inTable I. At present South Africa is the third largest producerof the ore, and has the largest reserves. China and Mexico,being closer to the larger international markets, are the twotop producers.

A brief history of fluorine chemistry and technology(Ameduri, 2011)A timeline for the major discoveries and developments influorine chemistry is listed below.

➤ Georg Bauer first describes the use of fluorspar (CaF2)as a flux in 1530 – as a flux aiding the smelting of oresby German miners

➤ Heinrich Schwanhard finds, in 1670, that fluorspardissolved in acid and the solution could be used to etchglass

➤ From the 1720s, the effect on glass by adding sulphuricacid to fluorspar is studied

➤ Scheele (a Swedish scientist) ‘discovers’ fluoric acid(HF) in 1771

➤ Several chemists try unsuccessfully to isolate fluorine,and several die of HF poisoning during separationexperiments

➤ The French chemist Moissan is the first to isolateelemental fluorine gas. He is awarded the Nobel Prizein 1906

➤ Swarts discovers the Cl/F exchange chemistry of SbF3➤ Midgley discovers Freons in 1928➤ In the 1930s General Motors begins using Freon-type

fluorocarbons (CFCs) as replacement for hazardousmaterials e.g. NH3. CFCs also finds use in propellantsand fire extinguishers

➤ 1938 Plunkett of DuPont discovers Teflon®

➤ WW II and uranium enrichment ➤ In 1947 Fowler discovers the CoF3 method of perfluori-

nation

▲


Figure 2 – Henri Moissan at his bench (left), and his original fluorine cell (right)

Figure 1 – A hydrogen fluoride wound being treated

➤ 1949 sees Simons’ discovery of electrochemical fluori-nation

➤ In the mid-1950s 3M invents ScotchguardTM

➤ Fried’s initial pioneering work in ‘medicinal fluorinechemistry’ commences in 1954

➤ Neil Bartlett’s discovery of noble gas chemistry in 1962(XePtF6)

➤ Rowland and Molina’s model for ozone depletion ispublished in 1974

➤ Hargreave’s ‘direct’ perfluorination discoveries in 1979➤ Fluorocarbon gases start finding application in the

semiconductor industry in late 1980s➤ 2003 sees O’Hagan’s isolation of first fluorinating

enzyme➤ Fluorine has become ubiquitous in pharmaceuticals,

and is essential in medicinal chemistry. At present there are more than 50 industrial producers of

HF (Roskill, 2009), the source precursor for all industrialfluorochemicals. Because of cost considerations, it is alwayspreferable in practice to employ a synthesis route that usesHF rather than F2.

The nuclear fuel cycleAs indicated in Figure 3, fluorine plays a critical role inseveral of the unit processes in the nuclear fuel cycle.Generally, the uranium arrives at the conversion plant in theform of U3O8. In order for isotope separation to be effected,uranium is required in the form of a gaseous compound. Thiscompound is UF6. U3O8 is converted to UF6 in a three-stepprocess, each requiring its own plant. The oxide is firstconverted to UO2 in a hydrogen atmosphere, according to thereaction

[3]

Note that U3O8 is a mixed valence oxide, thus reductionof a single U+6 to U+4 takes place. Subsequent to this, theuranium dioxide is converted into uranium tetrafluoride inthe substitution reaction:

[4]

Finally, the uranium tetrafluoride is fluorinated to thehexafluoride, using elemental fluorine gas:

[5]

Enrichment now takes place, with fissionable U235

separated from U238. This is normally done by centrifugetechnology. The enriched uranium is used as solid uraniumdioxide. There is more than one way of carrying out thereduction. A standard method is in a hydrogen-fluorine flamereactor. The high temperature is needed to initiate thereaction, given as

[6]

Solid uranium dioxide powder is pressed into pelletswhich are housed in Zircalloy tubes. These are bundled intofuel elements (in the case of pressurized-water reactors)ready for use.

UF6 is itself a powerful oxidant and fluorinating gas.Materials of construction for plants handling UF6 are thussimilar to those for plants that have fluorine as reactant orproduct. Seals, filters, bearings, etc., are machined fromvarious fluoropolymers, predominantly polytetrafluoroethylene(PTFE). Being fully fluorinated, PTFE is resistant to attack byfluorinating agents (Drobny, 2009; Ebnesajjad, 2013).

For comprehensive information about the nuclear fuelcycle, the reader is referred to Barré and Bauquis (2007), Kok(2009), Konings (2012), Tsoulfanidis (1996), Wilson(1996), and Yemelyanov (2011).

South Africa’s fluorochemical capabilityHighlights in the history of South African fluorochemicaltechnology platform (Naidoo, 2015) are listed below.

➤ Nuclear conversion starts at the Atomic EnergyCorporation (AEC) (now the South African NuclearEnergy Corporation, Necsa) in the 1960s

➤ Anhydrous hydrogen fluoride (AHF) and fluorine (F2)are required for uranium hexafluoride (UF6)production. AECI acquires the technology in the 1970s

➤ AECI stops producing AHF in 1984➤ Necsa commissions an HF plant in 1985



Table I

International fluorspar reserves and production (Roskill, 2009)

World mine production (1000 t/a) World reserves (Mt)2002 2003 Reserve Reserve base

China 2450 2450 China 21 110Mexico 630 650 South Africa 41 80South Africa 227 240 Mexico 32 40Mongolia 200 190 Mongolia 12 16Russia 200 200 Russia - 18France 105 110 France 10 14Kenya 98 100 Kenya 2 3Morocco 95 95 Morocco NA NANamibia 81 85 Namibia 3 5Spain 130 135 Spain 6 8USA 0 0 USA 0 0Other countries 310 320 Other 110 180Total 4550 4540 Spain 230 480


➤ Industrialization and commercialization – 1992➤ The Fluorochemical Expansion Initiative (FEI) is

adopted as a national initiative – 2006➤ Pelchem mandated to champion FEI by Necsa Board of

Directors – 2007➤ Two research chairs are founded via the National

Research Foundation’s SA Research Chairs Initiative(SARChI), one at the University of KwaZulu-Natal(UKZN) and one at the University of Pretoria (UP).

At present Necsa and its wholly-owned subsidiaryPelchem SOC Ltd are the main centres of South Africanfluorochemical expertise, along with two university chairs,one at the University of KwaZulu-Natal (UKZN) and theother at the University of Pretoria (UP). Pelchem runs acommercial 5000 t/a HF plant. Figure 4 shows a photographof the rotary kiln HF reactor.

The company also operates some 20 fluorine electrolysiscells (Figure 5).

Pelchem supplies a range of fluorine products to the localand international markets. These include xenon difluoride,nitrogen trifluoride, various organofluorine compounds,perfluorinated alkanes, and a variety of inorganic fluoridesalts.

Although a full fuel cycle existed on the Pelindaba site, itwas abandoned in 1995. The technology in effect does notexist anymore, and if a new fuel cycle is to be established inSouth Africa, it will have to be a start-up from scratch ratherthan resuscitation of the old technology. Should this come topass, our fluorochemical expertise, both existing and underdevelopment, will be invaluable if not essential. This will be

▲


Figure 3 – The nuclear fuel cycle (World Nuclear Association, 2015)

Figure 4 – HF rotary kiln at Pelchem SOC Ltd

the case whether the conversion is purchased off the shelf ordeveloped locally. The operation of a conversion plantrequires detailed and extensive fluorochemical expertise.

The road aheadSince the inception of the Fluorochemical Expansion Initiative(FEI), South Africa has made considerable inroads into thedevelopment of its fluorochemical capability. The Necsaresearch effort has been strengthened and expanded, theSARChI Chair at UKZN has been extremely productiveregarding research into various thermodynamic aspects ofindustrial fluorochemical processes, UP is developing afluoropolymer capability, and Pelchem SOC Ltd has commis-sioned a new pilot plant, known as the MultipurposeFluorination Pilot Plant (MFPP). The next few years arecritical. A number of things need to happen for the SouthAfrican research and development effort to continueprogressing, and for the technology to be leveraged for thenuclear build plan. These are:

➤ The next phase of FEI needs to be commerciallysuccessful, with visible new products

➤ The new cohort of scientists and engineers, trained viafunding by FEI, SARChI, and the Advanced MetalsInitiative (AMI), have to find employment in thefluorine/nuclear industry

➤ The decision about the nuclear new-build programmehas to be taken sooner, rather than later. Within thenext 5–7 years the majority of the last generation ofNecsa senior scientists and engineers will have retired

➤ The current postgraduate training programme has to beaccelerated, with Necsa senior scientists retained forco-supervision of dissertations and theses.

AcknowledgementsThe author acknowledges the South African NationalResearch Foundation for financial support via the SARChIprogramme, and Department of Science and Technology forfunding through their Fluorochemical Expansion Initiative.Rajen Naidoo, current acting CEO of Pelchem, is thanked theplant photographs, and Dr Johann Nel and Gerard Puts areacknowledged for comment and assistance with graphics,respectively.

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COTTON, F.A., WILKINSON, G., MURILLO, C.A., and BOCHMANN, M. 2007. AdvancedInorganic Chemistry, Wiley, New Delhi, India.

DROBNY, J.G. 2009. The Technology of Fluoropolymers, CRC Press, Boca Raton,FL.

EBNESAJJAD, S. 2013. Introduction to Fluoropolymers: Materials, Properties,Applications. Elsevier, Amsterdam.

KOK, K.D. 2009. Nuclear Engineering Handbook. CRC Press, Boca Raton, FL.

KONINGS, J.M. 2012. Comprehensive Nuclear Materials Vol 2. Elsevier,Amsterdam.

NAIDOO, R. 2015. History of fluorine and HF at Necsa. Personal communication.Pelchem SOC Ltd.

ROSKILL INFORMATION SERVICES. 2009. The Economics of Fluorspar. London.

RUDGE, A.T. 1971. Preparation of elemental fluorine by electrolysis,Introduction to Electrochemical Processes. Kuhn, T. (ed.). Elsevier,Amsterdam. Chapter 5.

SHIA, G. 2004. Fluorine. Kirk-Othmer Encyclopedia of Chemical Technology.Hoboken, NJ.

SLESSER, C. and SCHRAM, S.B. (eds). 1951. Preparation, Properties, andTechnology of Fluorine and Organic Fluoro Compounds. McGraw-Hill,New York.

SMITH, R.A. 2004. Hydrogen fluoride. Kirk-Othmer Encyclopedia of ChemicalTechnology. Hoboken, NJ.

TSOULFANIDIS, N. 2010. The Nuclear Fuel Cycle. American Nuclear Society,Scientific Publications, La Grange Park, IL.

WILSON, P.D. 1996. The Nuclear Fuel Cycle: From Ore to Waste. OxfordScientific Publications, Oxford.

WORLD NUCLEAR ASSOCIATION. 2015. The nuclear fuel cycle.. http://www.world-nuclear.org/info/Nuclear-Fuel-Cycle/ [Accessed July 2015].

YEMELYANOV, V.S. and YESVSTYUKHEN, A.I. 2011. The Metallurgy of Nuclear Fuel:Properties and Principles of the Technology of Uranium, Thorium andPlutonium. Pergamon Press, Oxford. ◆



Figure 5 – Pelchem fluorine cells

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Introduction Contemporary product design processes havebeen affected by new technologies that assistthe manufacturing sector to meet the specifi-cations of specialized components used in theaerospace and medical industries (Williams

and Revington, 2010). Direct selective lasersintering (SLS) or, more specifically whenconsidering the processing of metal, directmetal laser sintering (DMLS), are some ofthese technologies. Implants used in themedical industry to repair or replace bonestructure must be strong, ductile, and biocom-patible. Materials used for this purpose arestainless steel, cobalt, chromium alloys, orpure titanium, with titanium (Ti) being thematerial of choice (Taylor, 2004; Bertol et al.,2010). Titanium is also used for applicationsin the aerospace industry due to its lightweight, excellent corrosion resistance, highstrength, and attractive fracture behaviour (Liet al., 1997). In order to manufacture a high-quality component such as an implant, thetitanium powder used as feed material shouldbe dense and spherical. Powder particle sizecan affect the material spreading and thesintering rate due to the fact that the shape,density, and size of the particles have an effecton their packing density, sintering mechanism,and the flowability of the powder duringfeeding (Gignard, 1998, p. 34; Despa et al.,2011).

The chemical purity of the titanium powderduring processing is also very important.Surface oxidation should be prevented as thisincreases the surface tension, hinderingmaterial from flowing during sintering.Oxidation also results in poor bondingbetween sintered lines affecting themanufactured structures, while nitrides reducethe material’s corrosion resistance (Gignard,1998, p.34).

South Africa has an opportunity to gainbenefit from its abundant titanium-bearingmineral reserves through value beneficiation.The potential applications and markets fortitanium are in aerospace, the armaments

Titanium and zirconium metal powderspheroidization by thermal plasmaprocessesby H. Bissett, I.J. van der Walt, J.L. Havenga and J.T. Nel

SynopsisNew technologies used to manufacture high-quality components, such asdirect laser sintering, require spherical powders of a narrow particle sizedistribution as this affects the packing density and sintering mechanism.The powder also has to be chemically pure as impurities such as H, O, C, N,and S causes brittleness, influence metal properties such as tensilestrength, hardness, and ductility, and also increase surface tension duringprocessing.

Two new metal powder processes have been developed over the pastfew years. Necsa produces zirconium powders via a plasma process for usein the nuclear industry, and the CSIR produces titanium particles for use inthe aerospace industry.

Spheroidization and densification of these metal powders require re-melting of irregular shaped particles at high temperature and solidifyingthe resulting droplets by rapid quenching. Spherical metal powders can beobtained by various energy-intensive methods such as atomization ofmolten metal at high temperatures or rotating electrode methods. Rapidheating and cooling, which prevents contamination of the powder byimpurities, is, however, difficult when using these methods for high-melting-point metals. For this reason plasma methods should beconsidered.

Thermal plasmas, characterized by their extremely high temperatures(3000–10 000 K) and rapid heating and cooling rates (approx. 106 K/s)under oxidizing, reducing, or inert conditions, are suitable forspheroidization of metal powders with relatively high melting points.Thermal plasmas for this purpose can be produced by direct current (DC)plasma arc torches or radio frequency (RF) inductively coupled discharges.In order to obtain chemically pure spheroidized powder, plasma gases suchas N2, H2, O2, and CH4 cannot be considered, while Ar, Ne, and He aresuitable. Neon is, however, expensive, while helium ionizes easily and it istherefore difficult to obtain a thermal helium plasma at temperatureshigher than 3000 K. Therefore argon should be used as plasma gas.Residence times of particles in the plasma region range from 5–20 ms, butthis is usually sufficient as 7–8 ms is required for heating and melting oftitanium or zirconium metal particles in the 30 μm size range at 3500 K. In this study the melting and spheriodization of titanium powders wasinvestigated by DC non-transferred arc and RF induction plasma methods.The powders were characterized before and after plasma treatment byoptical microscopy and scanning electron microscopy (SEM) to observe ifany melting or spheroidization had occurred.

Keywordsplasma, zirconium, titanium, spheroidization.

* The South African Nuclear Energy Corporation SOCLtd., (Necsa).




Titanium and Zirconium metal powder spheroidization by thermal plasma processes

industries, naval applications, offshore oil industries,architecture, and medicine. By establishing a local titaniumindustry in South Africa, enterprise development across allsegments is expected (Van Vuuren, 2009).

A convenient way to produce spherical particles, and inparticular titanium particles, is to re melt irregularly shapedtitanium particles at high temperatures and solidify thedroplets by rapid quenching. This can be done by usingthermal plasmas, which are characterized by their extremelyhigh temperatures (3 000–10 000 K) and rapid heating andcooling rates (approx. 106 K/s) under oxidizing, reducing, orinert conditions. Plasma is a partially or fully ionized gas in acondition of quasi-neutrality. Thermal plasmas used forparticle spheroidization are generally produced by devicessuch as direct current (DC) arc plasma torches and radiofrequency (RF) inductively coupled discharges. Thermalplasmas with temperatures as high as 2 × 104 K can beobtained (Li and Ishigaki, 2001).

This paper presents a short background on the DC plasmaand RF induction plasma systems used for spheroidization,as well as some background on the properties of powderbefore and after plasma treatment. A few experimental resultsrelating to the spheroidisation of titanium powder using DCplasmas and RF induction plasmas at the South AfricanNuclear Energy Corporation (Necsa) are discussed.

Thermal plasmas for spheroidizationAlthough various plasma methods can be used, RF inductionplasmas are the preferred method for the spheroidization anddensification of particles. This is due to the longer residencetime in the plasma and also the lower possibility of contami-nation caused by electrode erosion. The residence timescommonly employed for particle melting in RF plasmas rangefrom 5 to 20 ms (Gignard. 1998, p. 24). DC arc plasmatorches, although not regularly used for spheroidization, arealso an option. Necsa has expertise in the design andoperation of these plasma torches. A short description of theRF and DC plasma torch methods is given, but the completesystem, including as particle feeding, quenching, andcollection methods, will not be discussed. Information on thespheroidization of titanium metal powders by thermalplasmas has not been found in the public domain, andliterature available on spheroidization by DC thermal plasmasis limited to in house designed plasma torches.

RF induction plasmaIn an induction torch, the energy coupling between theelectric generator and the plasma itself is done by acylindrical coil (Gignard, 1998, p. 6). A typical induction set-up is illustrated in Figure 1.

The ‘flame’ properties for a RF induction plasma aredependent on the power input, frequency, and the gaspressure and composition. Less power is required at higherfrequencies and lower gas pressures to maintain a plasma.Comparing argon and hydrogen, less power is required tomaintain an argon plasma due to the fact that argon ionizesmore easily than hydrogen (Gignard, 1998, p. 9), whilehigher plasma gas temperatures can be obtained at higheroperating pressures.

Although various RF induction plasma systems areavailable, several research studies have utilized thoseavailable from Tekna Systems Incorporated. Most of thesestudies have used the Tekna PL 50 torch, which can beoperated at various frequencies (0.3, 2, or 3 MHz) and powerinputs (30, 40, 50 and 100 kW). For spheroidization, Ar orAr/H2 is used as plasma gas (Jiang and Boulos, 2006; Li andIshigaki, 2001; Gignard, 1998, pp. 35-42). The use of theTekna PL 035LS torch has also been reported (Károly andSzépvölgyi, 2005).

DC plasmaThere are two types of DC torches, these being non-transferred and transferred arc torches. In non-transferred arcDC torches, the electrodes between which the arc is createdare inside the body of the torch itself. The plasma gas movesthrough the arc and is ionized to form the plasma jet. InFigure 2 a typical configuration for a non-transferred arcplasma torch is shown.

Non-transferred arc plasma torches are also used inplasma spraying. In this method a powder is introduced intothe plasma jet. In this instance, however, spherical powderswith a narrow size distribution are used to spray dense, evencoatings onto substrates. The purpose of this method is not tospheroidize the powder during the process, but the principleremains similar to spheroidization. In both instances thepowder will be quickly melted followed by rapid quenching.

Powder propertiesProperties such as particle size, bulk density, morphology,impurities levels, etc. are pivotal in selecting a suitablepowder to be used in SLS. Particle size affects the materialspreading and the sintering rate. The morphology, density,and size of the particles have an effect on their packing

▲


Figure 1 – RF induction plasma within the confinement tubes with aninduction coil Figure 2 – A typical non-transferred arc plasma torch

density, sintering mechanism, and the flowability of thepowder during feeding (Gignard, 1998, p. 34; Despa andGheorghe, 2011). Thermal treatment of a powder results inspherical particles with an increased bulk density anddecreased in impurity levels, while improving the flowabilityof the powder.

Density and flowability of powdersBulk density refers to the density of an uncompacted mass ofpowder taking into account the interparticular voids. The tapdensity refers to density of a compacted mass of powderwhere efforts have been made to eliminate the inter-particular voids by repeated tapping of the powder. Astandardized Hall flow method is used to investigate theapparent density and flow characteristics of free-flowingmetal powders (ASTM. 2006). In Table I the increases in tapdensity and Hall flow rate are indicated for titanium powdertreated by a Tekna torch operated at 50 kW (Boulos et al.,2011).

Impurities levelsTitanium becomes brittle due to its affinity for oxygen,nitrogen, and hydrogen. Contamination of titanium by air(more specifically oxygen) and hydrogen is thus a problemduring welding or sintering. This contamination causes anincrease in tensile strength and hardness but reducesductility, resulting in crack formation. For welding, 0.3%oxygen, 0.15% nitrogen, and 150 ppm hydrogen are seen asthe maximum tolerable limits. Surface discolouration gives agood indication of the degree of atmospheric contamination.The colour of the metal changes from silver to a light strawcolour (shades of yellow), then dark straw, dark blue, lightblue, grey and finally white (TiO2) as contaminationincreases. The light and dark straw colours indicate lightcontamination, which is usually acceptable. Dark blueindicates more contamination, while light blue, grey andwhite indicate high levels of contamination (TWI, 2014).

Impurity levels of elements such as O, C, H, N, and S canbe determined by using combustion methods or instrumentalgas analysis (IGA).

In Table I the decrease in impurity levels is clearlyindicated for titanium powder treated by a Tekna torchoperated at 50 kW using an argon plasma.

Visual characterization of powdersParticle morphology can be investigated by using opticalmicroscopy or scanning electron microscopy (SEM) todetermine whether melting or spheroidization of the powder

occurred during thermal plasma treatment. Figure 3 showsSEM images of titanium powder before and after treatment bya Tekna torch operated at 50 kW using an argon plasma. Theimages clearly indicate that the rapid heating followed bycooling resulted in spheroidization of the particles.

ExperimentalThe titanium powder used in this study was obtained fromthe Council for Scientific and Industrial Research (CSIR). Dueto the limited quantity of powder available, the as receivedpowder was characterized using SEM and an in-housemanufactured Hall flow meter according to ASTM (ASTM.2014). The amount of powder available for experimentationwas not sufficient for post-plasma measurement of theflowability using the Hall flow funnel, and for this reason,SEM and optical microscope images of the powder before andafter plasma treatment were used to assess the success of thespheroidization. Similarly, no assessment of the impuritylevels of the powder before and after plasma treatment waspossible.

As-received powder characterizationThe as-received and plasma-treated powders were charac-terized by SEM using a Quanta FEI 200 D instrument. Opticalimages of the powders were also obtained where possible. AZeiss Discovery V20 stereo microscope was used for thispurpose, and the images recorded utilizing Zeiss Axiovisionsoftware.

Owing to the large variation in the sizes of the particlesobserved, the as-received powder was separated into four



Table I

Spheroidization of titanium powder in a Teknaargon plasma torch

Properties Before plasma treatment After plasma treatment

Tap density 1.0 g/cm3 2.6 g/cm3

Production rate - 10 kg/h at 100 kWParticle size -140 + 400 mesh / -140 + 400 mesh /

40 100 μm 40 100 μmHall flow test No flow 37s/50 g[O] ppm 2050 1800[C] ppm 160 120[H] ppm 166 114[N] ppm 90 90[Cl] ppm 1200 360[Al] ppm 80 58

Figure 3 – SEM images of titanium powder (A) before and (B) after plasma treatment


fractions using sieve mesh sizes of 75, 125, 250, and 425μm. The four fractions were classified as < 75 μm, 75-125μm, 125-250 μm and 250-425 μm.

Spheroidization by thermal plasmasTwo high-temperature plasma systems were available for theexperiments, namely a RF induction plasma and a non-transferred arc DC plasma.

RF thermal plasma systemThe RF plasma system is shown in Figure 4. This system wasused for conceptual work and therefore the exact conditions(gas flow rates, powder flow rates, and input power) of theexperiments are not given at this stage. The RF induction coilwas connected to a 3 MHz power supply with a maximumpower input of 10 kW. Inside the induction coil, two quartztubes of different diameters were mounted in such a manneras to allow deionized cooling water to flow between them inorder to cool of the quartz tube. Above the quartz tube, ahopper loaded with titanium powder was mounted andconnected to a 1 mm orifice. Once the system was leak-tight,the system was evacuated and a non-thermal plasma-initiated. Argon gas was slowly introduced. The plasma wasgradually changed from a non-thermal to a thermal plasmaby increasing the operating pressure (increasing argon flow)and input power until a stable thermal plasma wasmaintained. At this stage the titanium powder was fedthrough the thermal plasma. Once a sufficient amount of agiven powder fraction had been treated, the plasma wasextinguished and the reactor allowed to cool. The powder wasthen removed for imaging by SEM and optical microscopy.The temperature of the plasma gas was not estimated due tothe complexity of the temperature determination method. Allfour powder fractions were treated.

DC thermal plasma systemA schematic of the non-transferred arc DC plasma system isshown in Figure 2. The system utilized a 30 kW DC powersupply with a maximum current input of 150 A. The systemwas operated at pressures slightly higher than atmospheric.In order to obtain temperatures high enough forspheroidization of titanium (approx. 3300 K) the system wasoperated using a nitrogen plasma. The reason for this is thatargon or helium ionize easily and therefore the voltage

(resistance) cannot be increased sufficiently. In order toobtain high enough temperatures using argon or helium, thepower supply would need to be operated at currents between400 and 600 A, which fall outside the window of operation ofthe power supply used in this study. The plasma torch wasmounted in a water-cooled reactor and operated at 150 A and200 V. The plasma gas was fed through the torch at a rate of1.43 g/s. The plasma gas temperature was estimated bycalculating the plasma gas enthalpy (kJ/kg) and relating thisto temperature. Once the plasma was stable, the as receivedpowder was fed near the tail flame (plasma jet) of the plasmaat a rate of 5 g/min.

Results and discussion

As-received powder characterizationFigure 5 shows an SEM image of the as-received powder. It isclear that the powder consisted of various particles sizes andmorphologies. Some particles appeared to be crystalline whileothers appeared amorphous. The particles had rough needle-like structures, porous round structures, and porous irregularstructures.

The SEM images of the four size fractions of the powderare shown in Figure 6. Even after size classification, largevariations in particle morphology and structure wereobserved.

RF thermal plasma treatmentThe temperature of the RF thermal argon plasma could not becalculated or determined at this stage. It is estimated that thetemperature of the plasma was near 3300 K (enthalpy of 1.6MJ/kg for argon) due to the fact that particle melting wasachieved.

Figure 7 shows an SEM micrograph of the as-receivedpowder after RF thermal plasma treatment. It can be seen thatthe smaller particles were affected by the thermal treatment.Some spherical particles were observed, as indicated by thearrows.

Figure 8 shows a SEM image of the < 75 μm powderfraction after plasma treatment. In most instances it appearedas though spherical particles smaller than 150 μm wereobtained, although some agglomeration occurred resulting inlarge irregularly-shaped particles.

▲


Figure 4 – The RF induction plasma system Figure 5 – SEM image of the CSIR Ti metal powder

Similar results were obtained using the 75 125 μmfraction, although slightly fewer spherical particles wereobserved. The large fractions (125-250 μm and 250-425 μm)were not affected by the plasma treatment. These resultssuggest that particles smaller than 125 μm can bespheroidized. Agglomeration of melted particles is a concern,however, and for this reason a sheath gas is required.Agglomerate formation is possibly due to the collection offine particles near the quartz tube. The particles ‘roll’ downthe tube surface, forming relatively large agglomerates. Thesheath gas is not only necessary to prevent agglomerationfrom occurring, but will also prevent the collection of fineparticles near the coil region of the quartz tube, a regularlyobserved occurrence. Once titanium metal particles collect onthe surface of the quartz tube, induction heating of theparticles occurs, resulting in a reduction of the plasma gastemperature.

DC thermal plasma treatmentThe DC plasma torch efficiency was calculated to be 61%.Therefore 18.3 kW of the rated 30 kW was available for theplasma gas, which at a nitrogen gas feed rate of 1.43 g/srelates to a gas enthalpy of 12.8 MJ/kg for nitrogen. Fromthermodynamic data it is estimated that the gas temperaturenear the arc was approximately 6200 K. Further from the arcthe temperature was approximately 3300 K (enthalpy of 3.7MJ/kg for nitrogen). For the DC plasma treatment, only theas-received powder was used and not the separate fractions.The estimated enthalpy cost per unit mass titanium forspheroidization was 60 kWh/kg (216 MJ/kg).

Optical micrographs of the treated powder are shown inFigure 9. A high degree of spheroidization is evident. Due tothe fact that nitrogen was used as plasma gas, surfacecontamination occurred, which is evident from the colours



Figure 6 – SEM images of the < 75 μm, 75-125 μm, 125-250 μm and 250-425 μm powder fractions

Figure 7 – SEM image of the as-received powder after RF thermalplasma treatment. The arrows indicate spheroidized particles

Figure 8 – SEM image of the < 75 μm fraction after RF thermal plasmatreatment


observed in the images. The light and dark straw coloursindicate light contamination, which is usually acceptable.Dark blue indicates more contamination, while light blueindicate high levels of contamination (TWI, 2014).

A SEM image of the treated powder is shown in Figure10. The variation in the contrast of the particles indicatesvarious degrees of contamination by either nitrogen oroxygen. Again, a high degree of spheroidization is evident. Anitrogen plasma was used in this instance, and it is expectedthat an argon plasma will yield similar results with respect tospheroidization of the particles, but without surface contami-nation occurring.

ConclusionsAvailable literature indicated that titanium metal particles canbe spheroidized in a thermal plasma by re melting ofirregularly shaped particles at high temperatures andsolidifying the droplets by a rapid quench. Thermal plasmatreatment of powders results in improved powder flow charac-teristics and an increased density of individual particles, pivotalin selecting a suitable powder to be used in SLS or DMLS.

The titanium powder received from the CSIR consisted ofvarious particle sizes and morphologies, where some particlesappeared to be crystalline and others amorphous.

In this study SEM images indicated that the powder couldbe spheroidized by an RF thermal plasma using argon gaswithout increasing impurity levels significantly. Variouspowder fractions were tested. Only the < 75 μm and 75-125μm fractions could be spheroidized; the larger fractions (125-250 μm and 250-425 μm) were not affected by the plasmatreatment.

SEM and optical microscope images also showed thattitanium powder could be effectively spheroidized by a DCthermal plasma using nitrogen gas. In this instance the powderwas used as-received and not sieved into fractions. Theparticles changed colour, indicating surface contamination bynitrogen. It is, however, expected that spheroidization withoutcontamination will be possible by DC thermal plasma treatmentusing argon gas at a gas temperature near 6200 K.

ReferencesASTM INTERNATIONAL. 2006. Standard test method for apparent density of free-

flowing metal powders using the Hall flowmeter funnel. Designation:B212 99. West Conshohocken, PA.

BERTOL, L.S., JÚNIOR, W.K., DA SILVA, F.P., and AUMUND, K.C. 2010. Technicalreport Medical design: Direct metal laser sintering of Ti 6Al 4V. Materialsand Design, vol. 31. pp. 3982 – 3988.

BOULOS, M., HEBERLEIN, J., and FAUCHAIS, P. 2011. Thermal plasma processes,fundamentals and applications. Short course presented at the University ofPretoria, 28 29 October. 2011. Pretoria, South Africa.

DESPA, V. and GHEORGHE, I.G. 2011. Study of selective laser sintering: aqualitative and objective approach. Scientific Bulletin of ValahiaUniversity Materials and Mechanics, vol. 6. pp. 150 – 155.

GIGNARD, N.M. 1998. Experimental optimization of the spheroidization ofmetallic and ceramic powders with induction plasma. Thesis, NationalLibrary of Canada, Sherbrooke, Quebec, Canada. .

JIANG, X. and BOULOS, M. 2006. Induction plasma spheroidization of tungstenand molybdenum powder. Transactions of Nonferrous Metal Society ofChina, vol. 16. pp. 13 – 17.

LI, Z. GOBBI, S.L. NORRIS, I. ZOLOTVSKY, S., and RICHTER, K.H. 1997. Laser weldingtechniques for titanium alloy sheet. Journal of Materials ProcessingTechnology, vol. 65. pp. 203 – 208.

LI, Y. and ISHIGAKI, T. 2001. Spheroidization of titanium carbide powders byinduction thermal plasma processing. Journal of the American CeramicSociety, vol. 84, no. 9. pp. 1929 – 1936.

TAYLOR, C.M. 2004. Direct laser sintering of stainless steels: thermalexperiments and numerical modelling. PhD thesis, School of MechanicalEngineering, University of Leeds, UK. Chapter 2, p. 5.

TWI. http://www.twi-global.com/technical-knowledge/job-knowledge/welding-of-titanium-and-its-alloys-part-1-109 [Accessed 18 July 2014].

VAN VUUREN, D.S. 2009. Titanium – an opportunity and challenge for SouthAfrica. Keynote address: 7th International Heavy Minerals Conference‘What Next’, Champagne Sports Resort, Drakensberg, South Africa, 20–23September 2009. Southern African Institute of Mining and Metallurgy,Johannesburg. pp. 1 – 8.

WILLIAMS, J.V. and REVINGTON, P.J. 2010. Novel use of an aerospace selectivelaser sintering machine for rapid prototyping of an orbital blowoutfracture. International Association of Oral and Maxillofacial Surgeons, vol.39. pp.182 – 184. ◆

▲


Figure 9 – Optical micrographs of the DC thermal plasma treated powder

Figure 10 – A SEM image of the DC thermal plasma treated powder

Introduction During the 1980s, a plasma research anddevelopment programme was launched at theerstwhile Uranium Enrichment Corporation ofSouth Africa (UCOR) at Valindaba outsidePretoria. The main purpose of this programmewas to develop alternative and moreeconomical processes for the processing andmanufacturing of uranium compounds used inthe nuclear fuel cycle. The conversion of UO2(as received from uranium mines) to UF6,which is needed for enrichment, conven-tionally involves several laborious, expensive,chemical processing steps. Plasma processescan eliminate several of the intermediate steps.

Enriched UF6 is converted to UO2, which ispressed into pellets that are used in nuclearpower plants. Plasma technology could also beused in a modified and optimized conversionprocess.

Zirconium alloys are used as nuclear fuelcladding material. In the 1980s, UCORdeveloped plasma process to producezirconium metal from ZrCl4, using hydrogen asa reductant (Nel et al., 2011).

With the termination of South Africa’snuclear programme in the 1990s, the SouthAfrican government asked the then AtomicEnergy Corporation (AEC) to investigate thepossible re-aligning of already establishedplasma technology, expertise and equipment toinvestigate other applications for plasmatechnology in the non-nuclear industry. Fromthis the so-called Metox (Metal Oxide)programme was born. This programmeinvestigated alternative means of zirconbeneficiation, as well as the production ofhigh-temperature-resistant ceramiccompounds that are used in the nuclearindustry, such as alumina, zirconia, silica,silicon carbide, zirconium carbide, and boroncarbide. By changing the process parameters,nanoparticles of the abovementioned productswere also synthesized in a plasma system.

Polytetrafluoroethylene (PTFE) wasextensively used in nuclear plants as seals, invalves and pipes, as containers, filters, etc.PTFE filters used in the enrichment plant thatare contaminated with uranium products canbe destroyed with a plasma process andsimultaneously, in the same process, theuranium values can be recovered.

There is no PTFE or fluoropolymerproduction facility in South Africa and all theseproducts are imported. In 1994, Necsa started

Plasma technology for themanufacturing of nuclear materials atNecsaby I.J. van der Walt, J.T. Nel and J.L. Havenga

SynopsisThe development of plasma technology at Necsa started in the early 1980s,when the applicability of high-temperature plasmas in the nuclear fuelcycle was investigated. Since 1995, this plasma expertise has expanded toother industrial applications, for example mineral beneficiation, nanotech-nology, fluorocarbon production and waste treatment, all of which are alsoof relevance to the nuclear industry.

Necsa has demonstrated the manufacture of plasma-dissociatedzircon, zirconium metal powder, carbon nanotubes, silicon carbide (SiC),zirconium carbide (ZrC) and boron carbide (B4C) at the laboratory andpilot plant scale. These materials are commonly used in the nuclearindustry. Zirconium alloys are used as fuel cladding material for nuclearfuel assemblies.

Necsa manufactured the monomer tetrafluoroethylene (TFE), using150 and 450 kW DC plasma systems, from which the polymer polytetraflu-oroethylene (PTFE) was synthesized for use in filters and as seals innuclear plants. With the nuclear renaissance at hand, it was demonstratedthat plasma technology can be used to produce hydrofluoric acid (HF),which is used in the manufacture of fluorine gas (F2) for the production ofuranium hexafluoride (UF6) directly from the mineral calcium fluoride(CaF2) without the use of sulphuric acid as in the conventional process.The recovery of valuable uranium from nuclear waste such as filters, oils,and solids with plasma processes will also be discussed. The destruction oflow-level nuclear waste by a plasma gasification system can reduce thevolume of this waste by several orders in magnitude, resulting in hugesavings in the storage costs. Another product of plasma technology is theencapsulation process for nuclear waste and the production of vitrifiedproduct, which could be used as filler material for medium-level nuclearwaste.

Keywordsplasma, zirconium, fluorocarbons, nuclear waste, nanomaterials.

* The South African Nuclear Energy Corporation SOCLtd. (Necsa), Pretoria.

© The Southern African Institute of Mining andMetallurgy, 2015. ISSN 2225-6253. Paper receivedAug. 2015.



Plasma technology for the manufacturing of nuclear materials at Necsa

a project on the production of TFE (C2F4), the precursor forPTFE, using a plasma process.

South Africa is the world’s fourth-largest producer ofcalcium fluoride (fluorspar or CaF2), but very little or no localbeneficiation takes place. In the nuclear fuel cycle,hydrofluoric acid is produced from CaF2 to manufacturefluorine gas, which is used to make UF6 for the enrichmentpurposes. It was proved that HF can be manufactured directlyin a plasma system instead of using the conventionalsulphuric acid route. The destruction, encapsulation, andvitrification of nuclear waste can also be accomplished with aplasma process.

Under the Advanced Metals Initiative (AMI) programmeof the Department of Science and Technology (DST), a newprocess was developed to make nuclear-grade zirconiummetal in a continuous plasma process.

The purpose of this paper is to discuss the abovemen-tioned processes in more detail.

Plasma technology in the nuclear fuel cycle The nuclear fuel cycle is schematically presented in Figure 1.The red arrows indicate where plasma technology can beused to replace conventional processes.

UF6 is a gas that is used for the enrichment of uranium.Processed uranium ore as received from the mines mayconsist of various chemical compounds, depending on thespecific process that the mine uses. The conventionalconversion route for UF6 production is a complex processinvolving multiple steps, each of which requires a separate,complete chemical plant. There are many minor variations onthis process, but in general it can be described in as follows.

Uranium oxide, which can be UO3 or U3O8, (also referredto as yellowcake) or ammonium diuranate (ADU or(NH4)2U2O7) is calcined to UO2 with hydrogen. The UO2 is

hydrofluorinated with hydrofluoric acid (HF(g)) to form UF4,a solid green compound, which is then fluorinated withfluorine gas to form UF6(g). This process is schematicallypresented in Figure 2. Hydrofluoric acid is produced by thereaction of CaF2 with H2SO4 according to Equation [1]. F2 gasis produced by electrolysis of KF.HF.

CaF2 + H2SO4 → CaSO4 + 2HF(g) [1]

The AEC proved in concept that a uranium oxide could bedirectly fluorinated to UF6 with capacitive and inductivecoupled non-thermal plasma according to Equation [2](Jones, Barcza, and Curr, 1993). This will eliminate the multi-step process presented in Figure 2 with a major economicadvantage.

2UO3 + 3CF4 → 2UF6 + 3CO2 [2]

Uranium metal can also be fluorinated by F2 or CF4 ormixtures thereof in an inductive coupled non-thermal plasma(Equation [3]).

U(m) + F2/CF4 → UF6 [3]

The AEC (now Necsa), demonstrated the use of a directcurrent (DC) plasma process to convert UF6 to UF4 usinghydrogen or cracked ammonia as reductant. This process wasscaled to a production rate of 1 kg UF4 per hour in acontinuous operation. It was estimated that the process couldbe 30% more economical than the conventional process. Themanufacture of depleted uranium metal from UF4 was alsodemonstrated, and a total of 13 kg of depleted uranium wasmanufactured using this process by the AEC using a scaled-up laboratory system. Processes have been proposed byToumanov (2003) and Fridman (2008, p. 449) for the directconversion of UF6 to uranium metal (Equation [4]).

UF6(gas) → Usolid + 3F2(g) [4]

▲


Figure 1 – A schematic of the nuclear fuel cycle. (World Nuclear University, n.d.) Red arrows indicate where plasma technology can be used

This process is highly endothermic and takes place attemperatures above 5000 K. High quenching rates in theorder of 10-7 seconds are required to prevent the reversereaction from taking place.

Beta-UF5(s) is an unwanted reaction product that formsunder certain conditions during the enrichment process. Ahigh-frequency non-thermal plasma process was utilized forthe in situ back-fluorination of beta-UF5 to UF6 using amixture of CF4/argon or fluorine/argon.

Unfortunately, further developments regarding theabovementioned nuclear plasma processes were halted whenSouth Africa’s nuclear programme was terminated in 1993.

Nuclear nano-materials Plasma technology is an extremely useful and effectivetechnique for the manufacture of nanoparticles, especiallywith regard to advanced ceramics (Fridman, 2008, pp417–498, 566; Van der Walt, 2012; Boulos, Fauchais andPfender, 1994). A typical plasma system for the manufac-turing of nanopowders is presented in Figure 3. The processinvolves the evaporation of reactants in the high-temperatureenvironment of a plasma arc, followed by rapid quenching ofthe vapour in order to nucleate the particles. Ceramic powderssuch as carbides, nitrides and oxides have been synthesizedin this way. Ceramic materials like SiC, ZrC, B4C and ZrO2,which are stable at high temperatures, have numerousapplications in the nuclear industry, and will become evenmore prominent in Generation IV high-temperature gas-cooled nuclear reactors (HTGRs) (Konings et al., 2012; Kok,2009).

Necsa has produced several of these nanomaterials.Nano-zirconia can be produced by the plasma dissociation ofzircon, followed by selective removal of the formedamorphous silica by fluorinating agents. Necsa commissionedand operated a pilot plant with a capacity of 100 kg/h toproduce plasma-dissociated zircon and a 10 kg/h plant forthe production of nano-silica.

Boron carbide is often applied in nuclear technology as aneutron shielding material and as control rods inside nuclearreactors. It has a high neutron absorbance cross-section forthermal neutrons of 755 barns and a high melting point of2723 K (Kirk-Othmer Encyclopedia of Chemical Technology;Lipp, 1965). Necsa produced nuclear-grade B4C powder in a30 kW non-transferred arc DC plasma system according toEquation [5]. BCl3 was evaporated at 60°C and fed into theplasma reactor at a rate of 120 litre per hour. The reactiontook place at about 2200°C, and nitrogen was used as aquench gas. A particle size of between 80 nm and 100 nmwas obtained, depending on the plasma parameters. Nano-sized B4C powder has enhanced compaction properties for theproduction of B4C pellets.

4BCl3 + CH4 + 4H2 → B4C + 12HCl [5]

Nano-sized ZrC can also be manufactured by a plasmaprocess.

Fluorocarbons in the nuclear industryPTFE is one of the few fluorocarbons used in the nuclearindustry. A sintered filter was produced from PTFE and usedin the enrichment stages.

A thermal chemical process was developed to produce CF4gas by reacting solid carbon with pure fluorine. A 99 %conversion was achieved, and the gas was used in a plasmaprocess to produce PTFE monomer. The chemical formulaegoverning these reactions are presented in Equations [6] and[7]:

C(s) + 2F2(g) → CF4 [6]

CF4 + C(s) + Plasma → C2F4 [7]

Various CF4 plasma plants for the production of TFE(C2F4) were constructed and commissioned at Necsa. A small30 kW laboratory system and 100, 150, and 450 kW pilotsystems were successfully developed and operated. A typicalplasma TFE production system is presented in Figure 4.

As part of a DST-funded Fluorochemical ExpansionInitiative, the C2F4 (TFE) was used in a suspension polymer-ization process to produce PTFE. After PTFE had beenproduced successfully, the second stage of the project toproduce FEP was developed and proven. A schematic diagramof the system is presented in Figure 5. Since neither of thesepolymers are produced in South Africa, this an opportunityfor a new business.



Figure 2 – The conversion of uranium oxide to UF6

Figure 3 – Schematic of a typical plasma system for the manufacturingof nano-powders


Plasma treatment of nuclear waste

VitrificationVitrification is a recognized standard method of reducing theenvironmental impact of harmful waste. In vitrification, thesolid component of the treated waste is encapsulated by high-temperature treatment. In recent years a glass formulationhas been added to the vitrification process that allows thetailoring of the waste matrix. Nuclear waste containment inglass has several advantages: it guarantees the stability ofthe waste package over very long time periods (50–300years) and could reduce the waste volume. Disposal safetycriteria can be more easily met by adaptation of the waste

matrix. This aspect is crucial in the treatment and disposal ofnuclear waste due to the long periods of storage required.With the addition of a glass/ceramic mixture thatencapsulates the waste, the resistance to leaching andchemical attack can be optimized (Petijean et al., 2002). Thematrix developed for the containment of radionuclides shouldtake into account the following issues (Luckscheiter andNesovic, 1996; Bisset and van der Walt, 2009):

➤ The waste must be accommodated within the matrixcomposition

➤ Waste loading➤ Characteristics such as resistance to solubility, phase

separation, and devitrification➤ Long-term behaviour such as thermal stability,

irradiation resistance and chemical and mechanicalstability

➤ Process considerations such as melting temperature,viscosity and electrical conductivity.

A thermal plasma is able to decompose various types ofwaste into a gas and a residue by exposing it to a very hightemperature. Because of the very high temperature in theplasma, no sorting is necessary, thus human exposure isminimized. This system also has the capability of containingthe uranium contamination in the molten pool of metal (if thewaste is in a drum) as well as in ceramics and non-combustible materials.

The cost implications will include a capital installationcost and an operational cost. If designed properly, a plasmavolume reduction system for nuclear waste can generateelectricity to feed back into the process, reducing theoperating cost to a minimum. There is a case to be madewhether the long-term cost of disposal at a waste repositorylike Vaalputs will be comparable to the costs of volumereduction of nuclear waste by a plasma process. Theadditional benefits of the plasma system are the production ofa vitrified product and the reduction of the waste volume(van der Walt and Rampersadh, 2011).

▲


Figure 4 – Schematic of a plasma TFE production process

Figure 5 – Schematic presentation of the polymerization reactor

Uranium recoveryThe recovery of valuable constituents such as uranium fromhistorical and current waste streams is important and mayalso be economically viable. Plasma technology could replaceother competing processes for this purpose, and may haveimportant advantages over the conventional processes.Conceptual research was done on one such waste stream,namely the coated uranium kernels used in the pebble bedmodular reactor (PBMR) project.

An alternative recovery method to the mechanicalcrushing of off-specification tri-structural-isotropic (TRISO)coated fuel microspheres was demonstrated. It was shownthat the inert SiC layer can be completely removed by etchingwith the active fluorine species by an inductively coupledradio-frequency CF4 glow-discharge impinging a static bedfrom the top, at a working pressure of 1 kPa (van der Walt etal., 2011). The apparatus is shown in Figure 6.

The recovery of uranium from contaminated PTFE filterswas also investigated by Necsa. A process was developed thatdepolymerizes PTFE and thereby separates the matrix fromthe extremely valuable uranium. This system makes use ofradio frequency (RF) induction heating to heat a reactor up tothe depolymerization temperature of PTFE. Inside the reactorthe PTFE is depolymerized and the product gas is cooled,separated from residual uranium particles, scrubbed andevacuated to a destruction facility where fluorocarbons areconverted into CO2 and HF by means of a DC thermal plasma

system using O2 and water gas. The HF is scrubbed bymeans of a KOH scrubber. A schematic diagram of thissystem is presented in Figure 7.

The AMI zirconium processIn 2005, the DST initiated the Advanced Metals Initiative(AMI) programme and asked Necsa to coordinate the NewMetals Development Network, with the specific focus on themanufacturing of nuclear-grade zirconium metal from themineral zircon (Zr(Hf)SiO4). The request from DST wasspecifically to develop a new and more economical way tomanufacture Zr metal using Necsa’s plasma and fluoro-chemical expertise and existing facilities. The conventionalmethods are described in detail in the scientific and patentliterature dating back to the early 1920s (Blumenthal, 1958).Significant industrial manufacturing of nuclear-gradezirconium started in the late 1940s after World War II, whenthe advantages of using zirconium in nuclear power plantswere realized. In general, most industrial manufacturingprocesses for nuclear-grade zirconium use high-temperaturecarbochlorination of zircon, selective separation of ZrCl4 andSiCl4, the separation of Zr and Hf by means of solventextraction, distillation, selective crystallization, etc., andeventually the reduction of purified ZrCl4 with magnesiumvia the Kroll process (Equation [2]). These conventionalprocesses can consist of between 16 and 18 individual steps,making the manufacturing of nuclear-grade zirconium one ofthe most expensive operations in the nuclear fuel cycle.

ZrCl4 + 2Mg → Zr + 2MgCl2 [8]

Zircon is an extremely chemical inert mineral. However,activating zircon with a plasma process to produce plasma-dissociated zircon (PDZ) makes it very reactive, especiallytowards fluoride-containing compounds such as HF andammonium bifluoride (ABF, NH4F.HF) (Nel et al., 2011a,2011b). Necsa operated a semi-commercial PDZ plant from1995 to 2003, which had a capacity of 100 kg/h using a 3 ×150 kW DC plasma torch configuration (Havenga and Nel,2011). Necsa developed a plasma process for the manufac-turing of Zr metal powder from PDZ, making use of thereactivity of PDZ with ABF to produce ZrF4 (Makhofane etal., 2011). ZrF4 or ZrCl4 is reduced with magnesium in a in a30 kW DC non-transferred arc plasma plasma process



Figure 6 – Non-thermal CF4 plasma system Figure 7 – Uranium recovery system with plasma waste destruction


(similar to the Kroll process) (Nel et al., 2012). The wholeprocess is generally referred to as the AMI zirconium metalprocess (Nel et al., 2013). The plasma process has, however,the additional advantage that it can be modified to make it acontinuous process, unlike the conventional Kroll process.The AMI zirconium metal process making use ofplasma/fluoride technology consists of only six steps, andoffers the potential of huge cost savings in comparison withthe conventional processes.

This process is schematically presented in Figure 8.

ConclusionsPlasma processing has numerous potential applications innuclear science and technology and for the manufacturing ofnuclear materials. Over a period of three decades, Necsa has

developed plasma applications in the nuclear fuel cycle,including the manufacture of nuclear ceramics and nanopar-ticles that are being (or can be) used in nuclear reactors,especially in high-temperature gas-cooled reactors.Fluoromonomers, which are used as precursors for manyfluoropolymers, can also been made via plasma processes.Fluoropolymers are extensively used as filters, seals, andcontainers in the nuclear industry. Nuclear waste destructioncan also be accomplished by plasma processes. Many of theseprocesses have been developed to pilot scale, while otherswere developed only to the laboratory scale and proof-of-concept. Necsa has patented plasma and fluoride processesfor the manufacturing of nuclear-grade zirconium metalpowder using the mineral zircon as starting material. ◆

▲


Figure 8 –Block flow diagram of the AMI zirconium metal process

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IntroductionIt is well known that methyltrichlorosilane(CH3SiCl3 or MTS) decomposes to form siliconcarbide (SiC) and hydrogen chloride (HCl) asshown in Equation [1]. The reaction kineticsand mechanisms for this reaction arethoroughly reported in the literature (Sone etal., 2000; Papasouliotis and Sotirchos, 1999;Kaneko et al., 2002; Wang et al., 2011).

CH3SiCl3(l) →SiC(s) + 3HCl(g) [1]

Owing to its physical and mechanicalproperties, SiC has found application in severalareas of high-power, high-frequency, andhigh-temperature technology (Harris, 1995;Saddow and Agarwal, 2004). Amongst itsseemingly limitless applications, SiC is gainingincreased attention as a nuclear ceramic due toits excellent mechanical properties anddimensional stability under irradiation (Katohet al., 2012). SiC composites have also beenproposed to act as fuel cladding in light waterreactors as a direct replacement for zirconiumalloy cladding (Katoh et al., 2012).Conventional ceramics exhibit certaindrawbacks such as low ductility and highbrittleness; however, the fact thatnanopowders can overcome theseshortcomings has encouraged the use of SiC invarious fields. The application of SiC

nanopowders in the nuclear industry, forexample, allows for fast recovery ofirradiation-induced defects (Vaßen and Stöver,2001).

Various synthesis methods for SiCnanoparticles have been reported in theliterature. These include carbothermicreduction (Dhage et al., 2009), pulsed laserdeposition (Kamlag et al., 2001), sol-gelprocesses (Ahmed and El-Sheikh, 2009),microwave heating (Satapathy et al., 2005;Moshtaghioun et al., 2012) as well asnumerous plasma techniques ranging frominductive radio-frequency (RF) (Károly et al.,2011; Sachdev and Scheid, 2001) tomicrowave plasma-assisted chemical vapourdeposition (Tang et al., 2008; Honda et al.,2003; Vennekamp et al., 2011). Previous workby the authors (Van Laar et al., 2015) reportedthe empirical study of the synthesis of SiCpowders using a microwave-induced plasma atatmospheric pressure only.

In this work a microwave plasma systemsimilar to that reported in the earlier paper(Van Laar et al., 2015) was used. The study isexpanded here to include initial results of SiCdeposition at low pressure (15 kPa) onto aquartz substrate. MTS was chosen as precursordue to its advantageous stoichiometric silicon-to-carbon ratio of unity, allowing it to act asboth a carbon and a silicon source. Acombination of its liquid state, high vapourpressure at standard conditions, and highvolatility also allows for easier feeding of theMTS into the system at both low and highpressures. Hydrogen was fed into the reactorto serve as a reductant for driving the

Synthesis and deposition of siliconcarbide nanopowders in a microwave-induced plasma operating at low toatmospheric pressuresby J.H. van Laar*†,I.J. van der Walt*, H. Bissett*, G.J. Puts†

and P.L. Crouse†

SynopsisSilicon carbide nanopowders were produced using a microwave-inducedplasma process operating at 15 kPa absolute and at atmospheric pressure.Methyltrichlorosilane (MTS) served as precursor, due to its advantageousstoichiometric silicon-to-carbon ratio of unity, allowing it to act as bothcarbon and silicon source. Argon served as carrier gas, and an additionalhydrogen feed helped ensure a fully reducing reaction environment. Theparameters under investigation were the H2:MTS molar ratio and the totalenthalpy. The particle size distribution ranged from 20 nm upwards, asdetermined by SEM and TEM micrographs. It was found that an increase inenthalpy and a higher H2:MTS ratio resulted in smaller SiC particle sizes.The adhesion of particles was a common occurrence during the process,resulting in larger agglomerate sizes. SiC layers were deposited at 15 kPawith thicknesses ranging from 5.8 to 15 μm.

Keywordssilicon carbide, microwave plasma, methyltrichlorosilane, nanoparticles.

* Applied Chemistry Division, South African NuclearEnergy Corporation (NECSA) SOC Ltd., Pelindaba,South Africa.

† Fluoro-Materials Group, University of Pretoria,South Africa.




conversion reaction. Argon served as the carrier gas. Changesin total enthalpy and the H2:MTS molar ratio were studied atatmospheric pressure.

Experimental

ApparatusThe experimental set-up consisted of a 1.5 kW power supplywith a MOS-FET amplifier, a microwave generator operatingat 2.45 GHz, a water-cooled magnetron head, and rectangularwaveguide with sliding short and stub tuners. The quartztube, with internal diameter of 2 cm and a length of 30 cm,was positioned through the middle and perpendicular to the

waveguide and served as the reaction zone. Support flangesat the top and bottom of the tube also served as gas inlets.Argon and hydrogen flow rates were controlled usingcalibrated Aalborg rotameters. This set-up is similar to theone used previously (Van Laar et al., 2015). An illustrationof the physical layout of the reactor assembly is shown inFigure 1, and a schematic representation of the flow path isshown in Figure 2.

At atmospheric pressures, argon was bubbled through theMTS in order to help vapourize and carry the MTS throughthe system. The MTS-rich argon stream was then mixed withhydrogen and argon streams in predetermined ratios beforeentering the reactor. At low pressures, the MTS wasvapourized under vacuum, and the flow rate controlled usinga valve placed before the entrance to the reactor. Calibrationcurves were reported in previous work (Van Laar et al.,2015).

The exiting gas was passed through a CaCO3 scrubber toremove HCl and any unreacted MTS before entering theextraction system. A T-connection and valve assemblyimmediately following the scrubber allowed for shiftingbetween vacuum and atmospheric operating pressures.

Characterization of the particles was performed using theavailable equipment at the University of Pretoria. Particle sizedistribution was determined with a ZEN 3600 MalvernZetasizer Nano System. Scanning electron microscopy (SEM)was performed on the particles using a high-resolution (6 Å)JEOL 6000 system, and transmission electron microscopy(TEM) was performed using the JEOL JEM2100F TEM (JEOLJapan). Powder X-ray diffraction was conducted with aPANalyticalX`pert Pro diffractometer using Co Kα radiation.The peaks were assigned using the databases supplied by theinstrument manufacturer.

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Synthesis and deposition of silicon carbide nanopowders in a microwave-induced plasma

Figure 2 – Schematic diagram of the experimental set-up

Figure 1 – Physical layout of the reactor assembly

MethodAt the start of each experimental run the argon plasma wasinitiated under vacuum at approximately 15 kPa (Figure 3b),using an Alcatel 2010I duel-stage rotary vane pump. Thedeposition experiments were performed at this pressure.Atmospheric pressure was reached by gradually increasingthe operating pressure, at which point filamentation of theplasma structure occurred (Cardoso et al., 2009) as shown inFigure 3a.

Hydrogen and MTS were then fed into the system.Depending on the H2:MTS ratio, stable plasmas were possibleat applied powers between 200 and 1500 W. High hydrogenconcentrations tended to extinguish the plasma, presumablydue to increased energy demand for dissociation of the H2bonds. During experiments at atmospheric pressure, a blackpowder deposited on the inner walls of the quartz tubes.These tubes were removed after each experimental run andflushed using distilled water. The water was collected andevaporated in a drying oven at 80°C, after which the blackpowders were collected. Quartz tubes with smaller diameters(15 mm) were used as substrates for the depositionexperiments. These smaller tubes were placed inside thelarger tubes, and held in place using zirconium wool.


Synthesis of SiC at atmospheric pressureThe results of the synthesis experiments are reportedelsewhere (Van Laar et al., 2015). The enthalpy values arethose of the system enthalpy, HT, which combines all thechemical species (argon, hydrogen, and MTS). Theseenthalpy values were determined by Equation [2]:

[2]

where Pf is the forwarded power, Pr is the reflected power,and mT is the total mass flow rate. The enthalpy of the MTS,HMTS, was calculated from Equation [3]:

[3]

The average particle sizes, as determined from SEM andZetasizer results, are also listed elsewhere (Van Laar et al.,2015). Particle agglomerates were a common occurrence.Based on the results, the best-fitted model includedquadratic and 2-factor interaction terms. Analysis ofvariance (ANOVA) results for the agglomerate sizesindicated that enthalpy had the greatest effect on theagglomerate sizes, whereas the H2:MTS ratio was found tobe least significant (Van Laar et al., 2015).

Zetasizer and SEM results were analysed using responsesurface analysis (RSA). The resulting surface contour plotsare shown in Figure 4 and Figure 5 respectively. Figure 4suggests that particle size decreases with increasingenthalpy and H2:MTS ratio. Considering the effect of thetwo studied parameters separately, higher enthalpy valuespresumably allow de-agglomeration to occur more readilydue to more energetic particle collisions. This results insmaller particle sizes. Higher H2:MTS ratios result in a morereducing environment, leading to smaller particle sizes.

When considering the effect of both parameters inconjunction, two contrasting trends are seen. At highenthalpy values (195–220 MJ/kg), particle sizes seem toincrease with increasing H2:MTS ratios. This trend couldpossibly be attributed to the increasing energy demand forhydrogen dissociation with increasing H2:MTS ratio. At lowenthalpy values (70 –120 MJ/kg), particle size decreaseswith increasing H2:MTS ratio. This is in contrast to trendsseen at high enthalpy values. It is speculated that this trendat low enthalpy values occurs because the energy supply isnot adequate to allow for hydrogen dissociation, increasingthe reducing environment and allowing more availableenergy for de-agglomeration.

The particle size distributions determined by theZetasizer show that lower enthalpies produce largeragglomerates, but that the agglomeration process is muchmore sensitive to H2:MTS ratios, with higher ratiosnegatively influencing the agglomerate size.

Figure 6 and Figure 7 present SEM micrographsshowing agglomerate particle sizes down to approximately50 nm.



Figure 3 – Operation of the reactor at (a) atmospheric pressure and (b) low pressure

•


Nanoparticles tend to form strongly bound agglomerates.Ultrasound sonification is often used to de-agglomeratenanoparticles; however, there is insufficient knowledge onthe de-agglomeration of nanoparticulate systems (Sauter et

al., 2008), making it extremely difficult to obtain a reliableparticle size distribution. From the images in Figure 6 andFigure 7, it can be seen that agglomerates exist in a widevariety of sizes. The smallest particle sizes to be confidentlyidentified from SEM images were approximately 50 nm.

The TEM micrographs are presented in Figure 8 andFigure 9. Figure 8 shows particle sizes down to approxi-mately 20 nm, and Figure 9 shows larger structures,presumably those of the agglomerates.

X-ray diffraction results show diffraction peaks atpositions indicative of beta (β) SiC (also referred to as cubic)shown in Figure 10. Also present are peaks indicative ofsilicon, although in much lower amounts than SiC. This couldimply the presence of silicon in the SEM and TEM images.

Other elements present within the plasma were verifiedusing an optical emission spectrometer, reported previously(Van Laar et al., 2015). The majority of peaks were in goodagreement with experimental values of elemental silicon,carbon, and argon (Kramida et al., 2014), suggestive of MTSdecomposition in the plasma. The presence of elementalsilicon in the gas phase as well as silicon in the productmaterial suggests that the addition of hydrogen to the plasmadrives the conversion reactions too far into the reductiveregime.

The equilibrium thermodynamics and formationmechanisms of Equation [1] have been reported in theliterature (Deng et al., 2009). The thermodynamics softwarepackage TERRA (Trusov, 2006) was used to confirm theoptimum conditions for the formation of β-SiC. The resultspredict that optimum yield for β-SiC formation is achieved attemperatures of around 1400 K. Microwave plasmas areknown to achieve temperatures in the region of 1000–10 000K (Tendero et al., 2006). The temperatures of theexperiments reported in this paper were measured using apyrometer, giving an indication of the SiC temperature insidethe reactor. SiC is chemically inert, with excellent microwaveabsorption and heat-conducting properties (Isfort et al.,2011). This enabled a rough estimation of the temperaturesinside the reactor, which were measured to range from 1100to 1400 K.

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Figure 4 – Effect of total enthalpy and H2:MTS molar ratio on individualparticle size

Figure 5 – Effect of total enthalpy and H2:MTS molar ratio onagglomerate size

Figure 6 – SEM image of SiC (1)

Figure 7 – SEM image of SiC (2)

Deposition of SiC at 15 kPaThe experimental results of the initial deposition experimentsare shown in Table I. All five runs were performed at thesame conditions, namely P = 500 W, mMTS = 0.11 g/min, andQAr = 100 sccm. The run times varied randomly, as theplasma extinguished at unpredictable times. The efficiency isan indication of the percentage of MTS mass converted anddeposited onto the quartz tubes. The layer thickness wascalculated using deposition rate, run time, and assumed SiCdensity of 3.21 g/cm3 (Harris, 1995).

SEM images of deposited SiC layers are shown in Figure11 and Figure 12. These images were taken from experimentnumber B, and show what appear to be layered structures.

Figure 12 shows agglomerates and smaller structuresdown to approximately 50 nm, supporting the presence ofnanoparticles.

Figure 13 shows the XRD analysis of the SiC layers. NoSiC peaks could be detected, indicating the presence ofamorphous SiC. The absence of peaks could also be attributedto the small layer thickness and curvature of the quartzsubstrates. The small peak broadening at 25° is believed tobe that of the amorphous quartz substrate.

Typical EDAX results are shown in Table II, indicating therelative amounts of carbon, oxygen, silicon, and chlorinefound in the SiC layers. The presence of oxygen and chlorineis indicative of SiO2 and HCl.

Conclusions and recommendationsA microwave-induced plasma operating at atmosphericpressure was used to decompose methyltrichlorosilane toform SiC nanoparticle agglomerates with sizes down to 20 nmas determined from SEM and TEM micrographs.



Figure 9 – TEM image of silicone carbide (2)Figure 8 – TEM image of silicone carbide (1)

Figure 10 – XRD spectrum of a product sample synthesized at an H2:MTS ratio of 4. SiC is found in the β phase

•

Agglomerates were a common occurrence, resulting in largerparticle size distributions when measured by the Zetasizer.The presence of β-SiC and silicon was confirmed using X-raydiffraction studies as well as optical spectroscopy. The firstpart of the investigation focused on the effect on particle sizeby varying the H2:MTS molar ratio and the total enthalpy.SEM, TEM, and Zetasizer results showed that higher enthalpyvalues and higher H2:MTS ratios produced smaller particlesizes. Furthermore, RSA results indicated that at high

▲



Table I

Experimental results of SiC deposition

Exp no. Run time (s) Deposition rate (g/min) Efficiency (%) Layer thickness (μm)

A 320 0.013 12.16 8.352B 900 0.009 18.48 15.12C 360 0.008 9.692 5.927D 160 0.018 9.898 5.854E 180 0.016 12.08 5.776

Figure 11 – SEM image of SiC layers Figure 12 – SEM image of SiC layer showing smaller nanostructures

Figure 13 – XRD analysis of SiC layers, indicating an amorphous structure

Table II

EDAX results indicating the relative elementalabundances

Element Weight % Atomic % Net int. Net int. error

C 12.16 21.37 6.55 0.1O 25.22 33.27 74.05 0.02Si 51.65 38.82 638.18 0.01Cl 10.97 6.53 77.49 0.03

enthalpy values (195–220 MJ/kg), particle size increases withincreasing H2:MTS ratio. This trend could possibly beattributed to the increasing energy demand for hydrogendissociation with increasing hydrogen amounts.

The second part of the investigation included initialexperiments on SiC layer deposition. Operating and synthesisconditions remained similar to those in the first part of thestudy, except the process was run at low pressure (15 kPa).Initial results indicate the successful deposition of SiC layerswith thicknesses ranging between 5.8 and 15 µm. XRDresults indicate amorphous crystalline structures, althoughfurther analysis is needed.

The use of microwave-induced plasma shows promise forthe deposition of SiC layers. Further experimental work isneeded, however, in order to determine the quality andstructure of these deposited layers. Pending these results, itis recommended that the use of microwave plasma systemsbe considered for the encapsulation of nuclear waste.

AcknowledgementsThe authors acknowledge the South African NationalResearch Foundation for financial support, and the SouthAfrican Nuclear Energy Corporation for use of theirequipment.

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IntroductionComplexes containing organometallic type β-diketone ligands with O,O and O,N donoratoms are used widely in coordinationchemistry and have applications in catalysis,radiopharmaceuticals, etc. (Schutte et al.,2011; Roodt, Visser and Brink, 2011; Brink etal., 2010; Otto et al., 1998). These ligandsystems are very useful because of their highlycoordinative nature, high solubility, and alsodue to their ability to be functionalized withvarious substituents on the carbonyl carbonatoms (Schutte et al., 2011).

Only a small number of β-diketonateligands have successfully been coordinated toa Nb(V) metal centre, with only a select fewbeing characterized by X-ray crystallography(Viljoen, 2009; Bullen, Mason and Pauling,1965; Preuss, Lamding and Mueller-Becker,1994; Funk, 1934; Davies, Leedlam and Jones,1999; Allen, 2002. The synthesis and crystalstructure determination at room temperature ofmer-oxidotrichlorido (thenoyltrifluoroacety-lacetonato-κ2O,O’)niobate(V){(NEt4)[NbOCl3(ttfa)]} was first reported byDaran et al. in 1979. Accordingly, for thiscurrent investigation, (NEt4)[NbOCl3(ttfa)]was re-evaluated at 100 K to determine if thestructural features might change withtemperature.

This study of this structure forms part ofan AMI-funded programme to betterunderstand the solid-state characteristics ofTa(V) and Nb(V) complexes and the influences

of the electron-donating and -withdrawinggroups of the β-diketone on the activityinduced and reaction mechanisms at thesemetal centres.

Experimental

Materials and instrumentsAll chemicals used for the synthesis andpreparation of the complexes were of analyticalgrade and were purchased from Sigma-Aldrich, South Africa.

The 1H-, 13C-, and 19F FT-NMR solution-state spectra were recorded on a BrukerAVANCE II 600 MHz (1H: 600.28 MHz; 13C:150.96 MHz; 19F: 564.83 MHz) or Bruker DPX300 MHz (1H: 300.13 MHz; 13C: 75.47 MHz;19F: 282.40 MHz) nuclear magnetic resonancespectrometer using the appropriate deuteratedsolvent. Chemical shifts, δ, are reported inppm. 1H NMR spectra were referencedinternally using residual protons in thedeuterated solvents, Acetonitrile-d3 [CD3CN =1.94(5) ppm]. 13C NMR spectra were similarlyreferenced internally to the solvent resonance[CD3CN = 1.39(4) ppm and 118.69(8) ppm]with values reported relative to tetramethyl-silane (δ 0.0 ppm).

The X-ray intensity data was collected on aBruker X8 ApexII 4K Kappa CCD area detectordiffractometer, equipped with a graphitemonochromator and MoKα fine-focus sealedtube (λ = 0.71069 Å, T = 100(2) K) operatedat 2.0 kW (50 kV, 40 mA). The initial unit celldeterminations and data collection were doneby the SMART (Bruker, 1998a) softwarepackage. The collected frames were integratedusing a narrow-frame integration algorithmand reduced with the Bruker SAINT-Plus andXPREP software package (Bruker, 1999)

A redetermination of the structure oftetraethylammonium mer-oxidotrichlorido(thenoyltrifluoroacetylacetonato-κ2-O,O')niobate(V)by R. Koen, A. Roodt and H. Visser

SynopsisThe tetraethylammonium salt of the mono-anionic coordination compoundmer-oxidotrichlorido(thenoyltrifluoroacetylacetonato-κ2O,O’)niobate(V)(NEt4)[NbOCl3](ttfa)], has been prepared under aerobic conditions andcharacterized by single-crystal X-ray diffraction. (NEt4)[NbOCl3](ttfa)]crystallized in the monoclinic P21/c space group, with a = 11.483 (5), b =12.563 (5), c = 17.110(5) Å, and β = 100.838 (5)º. The complex structureexists in a 50.0% (NbA) : 50.0% (NbB) positional disorder ratio.

KeywordsBidentate, niobium(V), disorder.

* Department of Chemistry, University of the FreeState, South Africa.




Tetraethylammonium mer-oxidotrichlorido(thenoyltrifluoroacetylacetonato-κ2-O,O')niobate(V)

respectively. Analysis of the data showed no significantdecay during the data collection. The data was corrected forabsorption effects using the multi-scan technique SADABSBruker, 1998b) and the structure was solved by the directmethods package SIR97 (Altomare et al., 1999) and refinedusing the WinGX (Farrugia, 1999) software incorporatingSHELXL (Sheldrick, 1997). The final anisotropic full-matrixleast-squares refinement was done on F2. The aromaticprotons were placed in geometrically idealized positions (C–H= 0.93 – 0.98 Å) and constrained to ride on their parentatoms with Uiso(H) = 1.2Ueq(C). Non-hydrogen atoms wererefined with anisotropic displacement parameters. Thegraphics were obtained with the DIAMOND program(Brandenburg, 2006) with 50% probability ellipsoids for allnon-hydrogen atoms.

Synthesis of (NEt4)[NbOCl3(ttfa)] (1) (Et4N)[NbCl6] (0.500 g, 1.147 mmol) was added to 4,4,4-trifluoro-1(2-thienyl)-1,3-butanedione (ttfaH) (0.327 g,1.147 mmol) in acetonitrile (20 cm3). The resulting solutionwas heated to 50ºC and stirred for 30 minutes. The excesssolvent was evaporated and dark yellow plate-like crystals ofthe title compound (1), suitable for X-ray diffraction, wereobtained (0.565 g, yield 89 %). IR (ATR, cm-1): ν(Nb=O) =952. 1H NMR (300.13 MHz, Acetonitrile-d3, ppm): δ = 5.88(s, 1H), 6.83 (m, 1H), 6.93 (dd, 1H), 7.40 (dd, 1H). 13C NMR(75.47 MHz, Acetonitrile-d3, ppm): δ = 30.1, 118.8, 123.9,130.0, 137.4, 142.0, 182.3. 19F NMR (564.83 MHz,Acetonitrile-d3, ppm): -73.37.

Results and discussionThe title compound was previously prepared by Daran et al.(1979), with X-ray diffraction data collected at roomtemperature. For this study, the reaction was modified asdescribed above and the data collected at 100 K.

The compound crystallizes in the monoclinic space group,P21/c, with four molecules in the unit cell (Z = 4). Theasymmetric unit consists of a Nb(V) metal centre surroundedby three crystallographically independent chlorido groups(Cl1A – Cl3A), an oxido (O3A) and one O,O’-bonded thenoyl-trifluoroacetonato ligand and a tetraethylammonium cation. Agraphic illustration is shown in Figure 1. The complex moleculeand the counter-ion are disordered over two positions in a 50NbA: 50 NbB ratio as shown in Figure 2. General crystallo-graphic details are presented in Table I, while selected bondlengths and bond angles are listed in Tables II and III respec-tively.

A distorted octahedral geometry is displayed for NbA andNbB. The Nb-Claxial distances for NbA vary between 2.428(1)and 2.507(1) Å and the Nb1A-Cl1A and Nb1A-O3A havedistances of 2.390(1) and 1.733(1) Å, respectively. Whencomparing Nb1A-O1A and Nb1A-O2A bond lengths, distancesof 2.357(1) vs. 2.037(1) Å are observed. This difference is dueto the effects of the electron withdrawing, CF3 substituent onthe bidentate ligand backbone causing a longer NbA-O1A bondlength. The trans Cl2-Nb1-Cl3 angle is 168.11(1)º, while theO1-Nb1-O2 bite angle is 79.65(1)º. A similar distortion isobserved for NbB, with bondlengths and angles in accordancewith NbA.

The coordination plane constructed through Cl1A, Cl2A,Cl3A, and O2A, as indicated in Figure 3, shows the niobiummetal centre is slightly shifted out of this plane by 0.2751(3)Å.

The molecular packing within the unit cell is illustrated inFigure 4. The packing illustrates a ‘head-to-tail’ arrangement

▲


Table I

Crystallographic and refinement details of the titlecompound

Crystallographic data (NEt4)[NbOCl3(ttfa)]

Empirical formula C16H24C13F3N1Nb1O3SFormula weight 566.68

Crystal system, space group Monoclinic, P21/ca, b, c (Å) 11.483(5), 12.563(5), 17.110(5)α, β, γ (°) 90.000, 100.838(5), 90.000

Volume (Å3), Z 2424.3(16), 4Density (calculated) (mg/m3) 1.553

Crystal colour, crystal size (mm3) Yellow, 0.99 × 0.79 × 0.31Absorption coefficient μ (mm-1) 0.951

Theta range, F(000) 2.664 – 27.99°, 1144Index ranges -16<=h<=15,

-15<=k<=15,-22<=l<=22

Reflections collected, 5834, 5209, 0.0574independent reflections, Rint

Completeness (%) 99.6Max. and min. transmission 0.750 and 0.412Data, restraints, parameters 5834, 894, 501

Goodness-of-fit on F2 1.0980Final R indices [I>2sigma(I)] R1 = 0.0260

wR2 = 0.0740R indices (all data) R1 = 0.0303

wR2 = 0.0796Largest diff. peak and hole (e.Å-3) 0.58, -0.58

Figure 2 – Graphic illustration of the mer-[NbOCl3(ttfa)] anion illustratingthe disorder in an overlay. (red) NbA = 50.0%; (blue) NbB = 50.0%.Hydrogen atoms and counter-ion omitted for clarity

Figure 1 – Graphic illustration of the mer-[NbOCl3(ttfa)] anion showinggeneral numbering of atoms. Numbering of the disordered complexdenoted by A = 50.0%. The displacement ellipsoids are drawn at 50%probability displacement level. Hydrogen atoms and counter-ionomitted for clarity

when viewed along the c-axis. There are no classicalhydrogen bonds or interactions observed in this structure.

In Figure 5, two coordination planes are illustrated; thefirst one constructed through O1A, O2A, and Nb1A, and thesecond through O1A, C2A, C3A, C4A, and O2. The anglebetween planes revealed the slight, 1.677º out-of-plane bendof the coordinated O,O’-bonded thenoyltrifluoroacetonatoligand. This also contributes to the distorted octahedralgeometry.

The crystal structure determination of the publishedcomplex was performed at room temperature (298 K), whilethe synthesized analogue (1) was determined at 100(2) K.Data obtained for the title compound (1) correlates well withthe previously published structure (Daran et al. 1979). Thedisordered part denoted by NbB differs less from thepublished structure and is probably a better representation ofthe anionic complex. Table IV illustrates a comparisonbetween bond angles and distances of the published structurevs. the structure collected at 100K. The greatest difference



Table II

Selected bond lengths of the two disordered parts in the title compound mer-[NbOCl3(ttfa)] anion, denoted by

NbA and NbB

NbA NbB

Atoms Bond length (Å) Atoms Bond length (Å)

Nb1A-Cl1A 2.390(1) Nb1B-Cl1B 2.389(1)Nb1A-Cl2A 2.507(1) Nb1B-Cl2B 2.373(1)Nb1A-Cl3A 2.428(1) Nb1B-Cl3B 2.339(1)Nb1A-O1A 2.357(1) Nb1B-O1B 2.254(1)Nb1A-O2A 2.037(1) Nb1B-O2B 2.095(1)Nb1A-O3A 1.733(1) Nb1B-O3B 1.745(1)

Table III

Selected bond angles of the two disordered parts in the title compound mer-[NbOCl3(ttfa)] anion, denoted by

NbA and NbB

NbA NbB

Atoms Bond angle (º) Atoms Bond angle (º)

O1A-Nb1A-O2A 79.65(1) O1B-Nb1B-O2B 76.82(1)Cl1A-Nb1A-O3A 96.28(1) Cl1B-Nb1B-O3B 103.50(2)O1A-Nb1A-O3A 171.69(2) O1B-Nb1B-O3B 166.36(2)Cl1A-Nb1A-O2A 163.58(1) Cl1B-Nb1B-O2B 166.38(1)Cl2A-Nb1A-Cl3A 168.11(1) Cl2B-Nb1B-Cl3B 160.12(1)C2A-C3A-C4A 120.60(2) C2B-C3B-C4B 120.99(2)

Figure 4 – Packing of (NEt4)[NbOCl3(ttfa)] (NbA) along the c-axis.Displacement ellipsoids are drawn at the 50% probability level

Figure 3 – Side view of the axial plane illustrating the out-of-planedistortion. Displacement ellipsoids are drawn at the 50% probabilitydisplacement level

Figure 5 – Illustration of the out-of-plane bend of the coordinated O,O'-bonded thenoyltrifluoroacetonato ligand. Displacement ellipsoids aredrawn at the 50% probability level


between the two complexes is the positional disorderobserved in the newly synthesized product.

ConclusionsA simplified method to obtain (NEt4)[NbOCl3(ttfa)] in aerobicconditions is reported. This highlights the misrepresentationof the ‘extreme sensitivity’ of niobium(V) complexes to airand water. Clearly, the exclusion of oxygen is not thatimportant, while the exclusion of water is, since it willincrease hydrolysis and thus the loss of chloride in favour ofaqua, hydroxide, or oxo coordination. The crystallographicinvestigation revealed that this structure exhibited a 50:50positional disorder. The electron withdrawing effects of theCF3 substituent on the bidentate ligand backbone isillustrated by the longer Nb-O bonds of Nb1A-O1A andNb1B-O1B vs. Nb1A-O2A and Nb1B-O2B. All bond lengthsand angles of the complex were found to be in accordancewith similar structures in the literature.

AcknowledgementsFinancial assistance from the Advanced Metals Initiative(AMI) of the Department of Science and Technology (DST) ofSouth Africa, through the New Metals Development Network(NMDN) managed by the South African Nuclear EnergyCorporation Limited (Necsa) is gratefully acknowledged.Gratitude is also expressed towards SASOL, PETLabsPharmaceuticals, and the University of the Free State forfinancial support of this research initiative outputs.Furthermore, this work is based on research supported inpart by the National Research Foundation of South Africa(UIDs 71836 and 84913).

ReferencesALLEN, F.H. 2002. Cambridge Structural Database (CSD) Version 5.35,

November 2013 update. Acta Crystallographica, vol. B58. pp. 380–388.ALTOMARE, A., BURLA, M.C., CAMALLI, M., CASCARANO, G.L., GIACOVAZZO, C.,

GUAGLIARDI, A., MOLITERNI, A.G.G., POLIDORI, G., and SPAGNA, R. 1999.SIR97: a new tool for crystal structure determination and refinementJournal of Applied Crystallography, vol. 32. pp. 115-119.

BRANDENBURG, K. 2006. DIAMOND, Release 3.0e. Crystal Impact GbR, Germany.BRINK, A., ROODT, A., STEYL, G., and VISSER, H.G. 2010. Steric vs. electronic

anomaly observed from iodomethane oxidative addition to tertiaryphosphine modified rhodium(I) acetylacetonato complexes followingprogressive phenyl replacement by cyclohexyl [PR3 = PPh3, PPh2Cy,PPhCy2 and PCy3]. Dalton Transactions, vol. 39. pp. 5572–5578.

BRUKER AXS Inc. 1998a. Bruker SMART-NT Version 5.050, Area-DetectorSoftware Package. Madison, WI, USA.

BRUKER AXS Inc. 1998b. Bruker SADABS Version 2004/1, Area DetectorAbsorption Correction Software. Madison, WI, USA.

BRUKER AXS Inc. 1999. Bruker SAINT-Plus Version 6.02 (including XPREP),.Area-Detector Integration Software. Madison, WI, USA.

BULLEN, G.J., MASON, R., and PAULING, P. 1965. The crystal and molecularstructure of bis(acetylacetonato)nickel (II). Inorganic Chemistry, vol. 4.pp. 456–462.

DARAN, J., JEANIN, Y., GUERCHAIS, J.E., and KERGOAT, R. 1979. The crystalstructure of tetraethylammonium trichlorooxo(1,1,1-trifluoro-4-thenoyl-2,4-butanedionato)niobate(V). Inorganica Chimica Acta, vol. 33. pp. 81–86.

DAVIES, H.O., LEEDLAM, T.J., and JONES, A.C. 1999. Some tantalum(V) β-diketonate and tantalum(V) aminoalcoholate derivatives potentiallyimportant in the deposition of tantalum-containing materials. Polyhedron,vol. 18. pp. 3165–3174.

FARRUGIA, L.J. 1999. WinGX suite for small-molecule single-crystal crystal-lography. Journal of Applied Crystallography, vol. 32. pp. 837–838.

FUNK, H. 1934. Über die einwirkung von niob- und tantalpentachlorid auforganische verbindungen (IV. Mitteil.). Berichte der DeutschenChemischen Gesellschaft, vol. 62. pp. 1801–1808.

OTTO, S., ROODT, A., SWARTS, J.C., and ERASMUS, J.C. 1998. Electron densitymanipulation in rhodium(I) phosphine complexes: structure of acetylacet-onatocarbonylferrocenyl diphenylphosphinerhodium(I). Polyhedron, vol.17. pp. 2447–2453.

PREUSS, F., LAMDING, G., and MUELLER-BECKER, S. 1994. Oxo- und thiotan-tantal(V)-verbindungen: Syntese von TaOX, und TaSX (X = OR, SR), Z.Zeitschrift für Anorganische und allgemeine Chemie, vol. 620. pp. 1812–1820.

ROODT, A., VISSER, H.G., and BRINK, A. 2011. Structure/reactivity relationshipsand mechanisms from X-ray data and spectroscopic kinetic analysis.Crystallography Reviews, vol. 17. pp. 241–280.

SCHUTTE, M., KEMP, G., VISSER, H.G., and ROODT, A. 2011. Tuning the reactivityin classic low-spin d(6) rhenium(I) tricarbonyl radiopharmaceuticalsynthon by selective bidentate ligand variation (L,L'-Bid, L,L ' = N,N', N,O& O,O' donor atom sets) in fac-[Re(CO)3(L,L'-Bid)(MeOH)]n complexes.Inorganic Chemistry, vol. 50. pp. 12486–12498.

SHELDRICK, G.M. 1997. SHELXL97. Program for crystal structure refinement.University of Göttingen, Germany.

VILJOEN, J.A. 2009. Speciation and interconversion mechanism of mixed haloO,O’- and N,O-bidentate ligand complexes of hafnium. MSc dissertation,Universtry of the Free State. 132 pp. ◆

▲


Table IV

Comparison of bond lengths and bond angles of (NEt4)[NbOCl3(ttfa)] (Bullen, Mason, and Pauling, 1965) collectedat room temperature vs. (NEt4)[NbOCl3(ttfa)] at 100 K

(NEt4)[NbOCl3(ttfa)]-NbA (100 K) (NEt4)[NbOCl3(ttfa)] (298 K)

Atoms Bond length (Å) Atoms Bond length (Å)

NbA NbB

Nb1-Cl1 2.390(1) 2.389(1) Nb1-Cl1 2.367(1)Nb1-Cl2 2.507(1) 2.373(1) Nb1-Cl2 2.365(2)Nb1-Cl3 2.428(1) 2.339(1) Nb1-Cl3 2.422(2)Nb1-O1 2.357(1) 2.254(1) Nb1-O1 2.285(3)Nb1-O2 2.037(1) 2.095(1) Nb1-O2 2.044(3)Nb1-O3 1.733(1) 1.745(1) Nb1-O3 1.704(3)Atoms Bond angle (º) Atoms Bond angle (º)

NbA NbB

O1-Nb1-O2 79.65(1) 76.82(1) O1-Nb1-O2 78.7(1)Cl2-Nb1-Cl3 168.11(1) 160.12(1) Cl2-Nb1-Cl3 165.0(1)C2-C3-C4 120.60(2) 120.99(2) C2-C3-C4 122.9(1)

Introduction Zirconium requires several purification steps toconform to nuclear-grade specifications. Littleinformation is available on the sublimationseparation of Zr and Hf compounds, especiallyin the fluoride form, the majority of whichdeals with sublimation under vacuumconditions. On the industrial scale, onlyvacuum sublimation of ZrF4 has beenreported. No records were found for thesublimation of ZrF4 in an inert atmosphere.Information on the sublimation rate of ZrF4 orHfF4 in an inert atmosphere is also limited.The rate is assumed to be dependent onseveral factors, of which temperature, area,and composition are considered the mostimportant.

In the process currently under investi-gation, the separation technique envisaged isby sublimation/desublimation in thetetrafluoride form. The separation involves thesublimation of the metal tetrafluorides in aninert atmosphere under controlled conditions,followed by desublimation (i.e. condensation)of the one metal fluoride in preference to theother.

In this paper, a sublimation model isdeveloped to predict the sublimation rates ofboth ZrF4 and HfF4 in an inert gas. These ratesare used to calculate the partial pressures ofthe two fluorides exiting a sublimer andentering a desublimer where the onetetrafluoride is selectively removed from thegas stream, thereby separating the twotetrafluorides.

Literature survey and theoreticaldiscussionSublimation methods used for the separationof Zr and Hf are reported in literature, butthese methods are all under vacuumconditions (Monnahela et al., 2013; Solov’evand Malyutina, 2002a).

Sublimation is, however, a general methodused for the purification of ZrF4 by removingmost trace elements, e.g. Fe, Co, Ni, and Cu(Abate and Wilhelm, 1951; Dai et al., 1992;Kotsar’ et al., 2001; MacFarlane et al., 2002;Pastor and Robinson, 1986; Solov’ev andMalyutina, 2002b; Yeatts and Rainey, 1965).

The addition of baffles (Abate andWilhelm, 1951; Kotsar’ et al., 2001; Yeatts andRainey, 1965) is used quite frequently to helpreduce the mechanical carry-over of impurities.These baffles are merely plates positionedbetween the charge and the cold finger. Theseimpurities impart a greyish colour to ZrF4,whereas pure ZrF4 is much whiter.

The literature also describes the use of agettering agent (Monnahela et al., 2013;Solov’ev and Malyutina, 2002a), which seemsto reduce the number of steps required toproduce nuclear-grade ZrF4. Getters usedinclude NiF2, zirconium oxides, and/orzirconium oxyfluorides.

A theoretical approach to thesublimation separation of zirconium andhafnium in the tetrafluoride form by C.J. Postma*, H.F. Niemand* and P.L. Crouse†

SynopsisThe separation of zirconium and hafnium is essential in the nuclearindustry, since zirconium alloys for this application require hafniumconcentrations of less than 100 ppm. The separation is, however, verydifficult due to the numerous similarities in the chemical and physicalproperties of these two elements.

Traditional methods for separation of zirconium and hafnium relypredominantly on wet chemical techniques, e.g. solvent extraction. Incontrast to the traditional aqueous chloride systems, the AMI zirconiummetal process developed by Necsa focuses on dry fluoride-based processes.Dry processes have the advantage of producing much less hazardouschemical waste.

In the proposed AMI process, separation is effected by selectivesublimation of the two tetrafluorides in an inert atmosphere undercontrolled conditions, and subsequent selective desublimation. Estimatesare made for the sublimation rates of the two tetrafluorides based on theequilibrium vapour pressures. A sublimation model, based on thesublimation rates, was developed to determine if the concept of separationby sublimation and subsequent desublimation is theoretically possible.

Keywordssublimation separation, zirconium tetrafluoride, hafnium tetrafluoride.

* The South African Nuclear Energy Corporation SOCLtd. (Necsa).

† Department of Chemical Engineering, University ofPretoria.

© The Southern African Institute of Mining andMetallurgy, 2015. ISSN 2225-6253. Paper receivedMar. 2015 and revised paper received July 2015.



A theoretical approach to the sublimation separation of zirconium and hafnium in the tetrafluoride form

Area-dependent sublimation rateMacFarlane et al. (2002) calculated the area-dependent rateof sublimation of ZrF4 and obtained a value of approximately1.87 g/m2/s at 850–875°C.

Product compositionTi, Esyutin, and Scherbinin (1990a, 1990b) found that pureZrF4 has a higher sublimation rate than industrial-gradeZrF4, which contains a degree of impurities. They concludedthat this might be due to the accumulation of low-volatilecomponents in the near-surface layer of the sample, makingdiffusion and evaporation increasingly difficult and resultingin decreased sublimation flux.

Layer heightIn a study on the influence of layer height on the vacuumsublimation rate of ZrF4, Ti, Esyutin, and Scherbinin (1990c)concluded that the sublimation rate does not necessarilydepend on the height of the sample in the sublimator.

Vapour pressure of ZrF4

Figure 1 gives a range of vapour pressures from the literaturefor both ZrF4 and HfF4 at temperatures above 600°C(Benedict et al., 1981; Cantor et al., 1958; Koreneo et al.,1972; Sense et al., 1954, 1953).

The data in Figure 1 was combined and can be expressedas Antoine constants for both ZrF4 and HfF4. These twoexpressions are given in Table I.

Experimental concept to be modelledThe flow diagram for the proposed process is presented in

Figure 2. The concept under consideration is to pass a streamof pre-heated dry nitrogen as a carrier gas over a bed ofsubliming ZrF4 and HfF4 (Figure 3); the gas exiting thesublimer then enters a desublimer operating at a slightlylower temperature. The difference between the vapourpressures as function of temperature is used to determine anoptimum temperature for the desublimer. In the desublimer,one of the tetrafluorides is desublimed in preference to theother, thus effecting separation. The remainder of the gasmixture exits the desublimer and enters a water-cooled coldfinger for collection of the remaining ZrF4 and HfF4, whichcan be subjected to a further separation step.

The sublimer (Figure 3) consists of two rectangularsections; a bottom section containing the bulk mass to besublimed, and a top section that facilitates the movement ofthe nitrogen gas and carries the sublimed tetrafluorides in thegas phase to the desublimer.

The desublimer is a long cylindrical pipe that is heated toa predetermined temperature, depending on the partialpressures of the ZrF4 and HfF4 entering the desublimer.

▲


Figure 2 – Block flow diagram for the sublimation separation of ZrF4 and HfF4

Figure 1 – Literature vapour pressures for ZrF4 (Benedict et al., 1981;Cantor et al., 1958; Koreneo et al., 1972; Sense et al., 1954, 1953)

Table I

Combined vapour pressure correlations for ZrF4

and HfF4

Component Vapour pressure Temperature range [kPa] [°C]

ZrF4 log(P) = 12.096−11879/T 600 – 920HfF4 log(P) = 12.391−12649/T 600 – 950

Figure 3 – Sketch of the sublimer

Modelling

Flux modelThe rate model for the sublimation of ZrF4 and HfF4 is basedon the work of Smith (2001), who predicted evaporationrates for liquid spills of chemical mixtures by employingvapour-liquid equilibrium. As sublimation progresses, thebed height decreases with time, which causes changes in themass transfer coefficient resulting in a change in thesublimation rate of the two tetrafluorides. The rate model forboth ZrF4 and HfF4 is given in Equation [1]:

[1]

where ri is the sublimation rate of ZrF4 (or HfF4) in mol perunit sublimation area per unit time, ki,t is the mass transfercoefficient in m/s at time t, Pi

* is the vapour pressure in kPa,pi

’ is the partial pressure in the bulk gas, xi,t is the molfraction of the respective tetrafluoride in the unsublimed bulkmass, R is the ideal gas constant (8.314 kPa.m3/kmol/K),and T is the temperature in K.

In order to calculate the total flux along the length of thesublimation pan (nj,t), the pan is divided into segments each oflength ΔL. The flux in each successive segment is calculated byadding the flux in the previous segment to the sublimedmasses of ZrF4 and HfF4 in segment j (Equation [2]).

[2]

where Zt is the height of the headspace above the solid bed atany given time t.

Model parameters

Mass transfer coefficientThe mass transfer coefficient (ki) is required for calculatingthe rate of sublimation and is a function of the Sherwoodnumber (Shi), the diffusion coefficient (DAB), and theequivalent flow diameter (De) (Equation [3]):

[3]

Sherwood numbers differ for each experimental set-up. Inthe case of convective mass transfer for forced convectionover a flat plate (in this case a sublimation pan), and forlaminar flow conditions with Reynolds number < 5 × 105,Prandtl number > 0.6, and Shmidt number (Sci) > 0.5, theSherwood number can be calculated using Equation [4](Çengel, 2006).

[4]

Diffusion coefficientsThe diffusion coefficient can be estimated using the Lennard-Jones potential to evaluate the influence of the molecularforces between the molecules. This correlation (Equation[5]), also known as the Chapman-Enskog equation, holds forbinary gas mixtures of non-polar, non-reacting species(Green, 2008; Welty, 2001), which is the case for ZrF4 andHfF4 in nitrogen.

[5]

σAB is the collision diameter, a Lennard-Jones parameterin Å, where A refers to nitrogen and B to either ZrF4 or HfF4.Since σ is denoted as the Lennard-Jones diameter of therespective spherical molecule (Welty, 2001), an estimationwas made for the diameter of a ZrF4 and a HfF4 moleculeassuming sphericity. The sizes of the respective moleculeswere calculated at room temperature with the use ofSpartanTM software. The equilibrium geometry was calculatedusing the Hartree-Fock method with the 6 31* basis set.Estimated values for the collision diameters of ZrF4 and HfF4with N2 were calculated as 4.146 and 4.127 Å respectively.

The collision integral (ΩD) is a dimensionless parameterand a function of the Boltzmann constant Κ (1.38 × 10-16

ergs/K), the temperature, and the energy of molecularinteraction ∈AB. The boiling points (Tb) for ZrF4 (912°C) andHfF4 (970°C) (Lide, 2007) were used to calculate the valuesfor ∈i with the use of an empirical correlation, given byEquation [6]:

[6]

Estimated values for the collision integrals for ZrF4 andHfF4 in N2 were calculated as 0.980 and 0.987, respectively.

The diffusion coefficients were calculated at severaltemperatures and are listed in Table II.

Model results

Limitations One practical problem encountered in the design of theexperimental set-up is the working temperature of the valves(650°C). At this temperature, the vapour pressures arerelatively low, which results in a very long sublimation time.Since all of the components of the valves are metallic, and itis general knowledge that these valves have a built-in safetyfactor, the decision was made to operate slightly above themaximum specification temperature of the valves, i.e. at700°C.

Sublimation ratesThe initial load to be sublimed was taken as 80 g, whichincludes the HfF4 and other impurities. The full sublimationarea is 0.0075 m2.

The average area-dependant rates of sublimation atseveral temperatures calculated from the model are given in



Table II

Diffusion coefficients for ZrF4 and HfF4 in nitrogen

at a pressure of 1 atmosphere

Temperature (°C) (cm2/s) (cm2/s)

700 0.693 0.677750 0.757 0.740800 0.824 0.806850 0.893 0.874


Figure 4, which is a sum of the average rates of ZrF4 andHfF4 at several time intervals. It can be seen that the rate is apower function of the temperature, illustrating the effect of ahigher temperature on the rate.

The total sublimation time is dependent on the area ofsublimation and the temperature of sublimation, the lattercontrols the two vapour pressures. The dependence of thetotal sublimation time on temperature is shown in Figure 5.At 850°C, the total sublimation time equals approximately 33minutes, whereas at 700°C the total sublimation time wascalculated to be approximately 24.5 hours. The vapourpressure ratio between 850 and 700°C is 42.7, whichindicates that the rate at 700°C should be lower by at leastthis factor, which amounts to a total sublimation time of 23.5hours. The other factor influencing the rate is the partialpressure in the gas stream, which is also higher at the highertemperature, which may contribute to the difference in thetotal sublimation time at 700°C.

The mass sublimed with respect to time at 700°C ispresented in Figure 6. Here it is evident that, according to themodel calculations, some HfF4 will remain in the sublimationpan once all the ZrF4 has sublimed. Theoretically this impliesthat the sublimation can be stopped after a certain time,thereby separating most of the HfF4 from the ZrF4 in the coldfinger.

The rates of both HfF4 and ZrF4 sublimation are given inFigure 7. From this figure it is evident that the sublimationrate of HfF4 becomes increasingly significant as thesublimation progresses, i.e. as the bed height lowers withtime. This is probably due to the mass fraction HfF4

increasing, since the ZrF4 has a higher rate of sublimation.The sublimed ZrF4 and HfF4 diffuse into the nitrogen

stream, exit the sublimer, and enter a desublimer operating ata slightly lower temperature than that of sublimation. Thedesublimation temperatures for the two tetrafluorides can becalculated from the vapour pressure correlations based on thepartial pressure of the respective fluorides entering thedesublimer. Figure 8 indicates the desublimation temper-atures obtained from the model at a sublimation temperatureof 700°C for ZrF4 and HfF4. It is evident that a desublimeroperating temperature of between 540°C (maximumtemperature for HfF4) and 610°C (minimum temperature forZrF4 desublimation) is required to ensure that, according tothe model calculations, no HfF4 will desublime in thedesublimer.

▲


Figure 5 – Total sublimation time calculated at different sublimationtemperatures

Figure 4 – Rate of sublimation of impure ZrF4 at different temperatures

Figure 8 – Desublimation temperatures of ZrF4 and HfF4 at asublimation temperature of 700°C

Figure 7 – Rate of sublimation for ZrF4 and HfF4 at 700°C as a functionof the sublimation time

Figure 6 – Mass of ZrF4 and HfF4 sublimed with time at 700°C

ZrF4 lost to cold fingerIn the case of ZrF4, the desublimer operating temperature isstill relatively high for all the ZrF4 to desublime while passingthrough the desublimer.

A desublimer operating temperature of 30°C higher thanthat of the maximum temperature for HfF4 (i.e. 570°C)results in a vapour pressure of ZrF4 which is still relativelylarge, and some of the ZrF4 will therefore not desublime andwill be collected on the cold -finger.

Comparison to literature dataThe area-dependent rates of sublimation for both ZrF4 andHfF4 at 850°C were calculated. The average rate over theentire duration of sublimation of the total sublimated massamounts to approximately 5.36 g/m2/s, which is 2.87 timeshigher than the value estimated from literature data (1.87g/m2/s). The difference between the literature and modelrates may be attributable to the impurities present in thesample used by Macfarlane et al., (2002), since the presenceof impurities can have a direct influence on the rate ofsublimation.

ConclusionsA sublimation model has been developed to predict thesublimation rates and the partial pressures of ZrF4 and HfF4in the tetrafluoride form and in an inert gas. The gas exits asublimer and enters a desublimer where the one tetrafluorideis desublimed in preference to the other, separating the twotetrafluorides.

The model revealed an area-dependent sublimation ratethat is 2.87 times higher than the value estimated fromliterature data (1.87 g/m2/s) at 850°C. This indicates that therates obtained from the model are within an acceptable range.The difference between the literature and model rates may beattributable to the impurities present in the sample used byMacfarlane et al., (2002), since impurities can have a directinfluence on the rate of sublimation.

Due to experimental/equipment constraints, the operatingtemperature of the sublimer should be in the range of 700°C.Optimal temperature selection is imperative, since lowtemperatures result in a low sublimation rate and hightemperatures increase the level of impurities in the sublimedproduct.

At a sublimer operating temperature of 700°C, the modelindicates that an operating temperature for the desublimer ofbetween 540°C (maximum temperature for HfF4) and 610°C(minimum temperature for ZrF4 desublimation) is required toensure that, according to the model calculations, only ZrF4and no HfF4 will desublime. Selection of an optimaltemperature of the desublimer is also critical, since too high atemperature will result in more ZrF4 lost to the cold finger.

It is recommended that the model results be comparedwith experimental values, including if necessary thesublimation kinetics to account for any possible timedependencies of the vapour pressures.

References

ABATE, L.J. and WILHELM, H.A. 1951. Sublimation of zirconium tetrafluoride. USAtomic Energy Commission, Ames Laboratory.

BENEDICT, M., PIGFORD, T.H., and LEVI, H.W. 1981. Nuclear ChemicalEngineering, McGraw-Hill, New York.

ÇENGEL, Y.A. 2006. Heat and Mass Transfer. McGraw-Hill, New York.

DAI, G., HUANG, J., CHENG, J., ZHANG, C., DONG, G., and WANG, K. 1992. A newpreparation route for high-purity ZrF4. Journal of Non-Crystalline Solids,vol. 140, no. 1-3. pp. 229–232.

CANTOR, S., NEWTON, R., GRIMES, W., and BLANKENSHIP, F. 1958. Vapor pressuresand derived thermodynamic information for the system RbF-ZrF4. Journalof Physical Chemistry vol. 62, no. 1. pp. 96–99.

GREEN, J.D. 2000. II / distillation / sublimation. Encyclopedia of SeparationScience. 2nd edn). Wilson, I.D. (ed.). Academic Press, Oxford.

KORENEO, Y., SOROKIN, I., CHIRINA, N., and NOVOSELO, A.V. 1972. Vapor-pressureof hafnium tetrafluoride. Journal of Inorganic Chemistry, vol. 17, no. 5.pp. 1195.

KOTSAR’, M.L,. BATEEV, V.B., BASKOV, P.B., SAKHAROV, V.V., FEDOROV, V.D., andSHATALOV, V.V. 2001. Preparation of high-purity ZrF4 and HfF4 for opticalfibers and radiation-resistant glasses. Inorganic Materials, vol. 37, no. 10.pp. 1085–1091.

LIDE, D.R. 2008. CRC Handbook of Chemistry and Physics. CRC, London.

MACFARLANE, D.R., NEWMAN, P.J., and VOELKEL, A. 2002. Methods of purificationof zirconium tetrafluoride for fluorozirconate glass. Journal of theAmerican Ceramic Society, vol. 85, no. 6. pp. 1610–1612.

MONNAHELA, O.S., AUGUSTYN, W.G., NEL, J.T., PRETORIUS, C.J., and WAGENER, J.B.2013. The vacuum sublimation separation of zirconium and hafniumtetrafluoride. AMI Precious Metals 2013 Conference, Protea Hotel, CapeTown, 14-16 October 2013.

NEL, J.T., DU PLESSIS, W., CROUS, P.L., and RETIEF, W.L. 2011. Treatment ofzirconia-based material with ammonium bi-fluoride. PatentWO2011013085 A1.

PASTOR, R.C. and ROBINSON, M. 1986. Method for preparing ultra-pure zirconiumand hafnium tetrafluorides. US patent 4,578,252 A.

SENSE, K.A., SNYDER, M.J. and CLEGG, J.W. 1953. Vapor pressures of berylliumfluoride and zirconium fluoride. US Atomic Energy Commission TechnicalInformation Services, Oak Ridge, Tennessee.

SENSE, K.A., SNYDER, M.J., and FILBERT, R.B.J. 1954. The vapor pressure ofzirconium fluoride. Journal of Physical Chemistry, vol. 58, no. 11. pp. 995–996.

SMITH, R.L. 2001. Predicting evaporation rates and times for spills of chemicalmixtures. Annals of Occupational Hygiene, vol. 45, no. 6. pp. 437–445.

SOLOV’EV, A.I. and MALYUTINA, V.M. 2002a. Production of metallic zirconiumtetrafluoride purified from hafnium to reactor purity. Russian Journal ofNon-Ferrous Metals, vol. 43, no. 9. pp. 14–18.

SOLOV’EV, A.I. and MALYUTINA, V.M. 2002b. Metallurgy of less-common andprecious metals. Production of metallurgical semiproduct from zirconconcentrate for use in production of plastic metallic zirconium. RussianJournal of Non-Ferrous Metals, vol. 43, no. 9. pp. 9–13.

TI, V.A., ESYUTIN, V.S., and SCHERBININ, V.P. 1990a. The dependence of thezirconium tetrafluoride sublimation rate in vacuum upon the processtemperature and product composition. Kompleksnoe Ispol'zovanieMineral'nogo Syr'ya, vol. 9. pp. 63–64.

TI, V.A., ESYUTIN, V.S., and SCHERBININ, V.P. 1990b. The influence of the sampleheight on the zirconium tetrafluoride sublimation process. Kompleksn.Ispol’z. Miner. Syr’ya, vol. pp. 60–61.

TI, V.A., ESYUTIN, V.S., and SCHERBININ, V.P. 1990c. The dependence of thezirconium tetrafluoride sublimation rate upon the process pressure.Kompleksnoe Ispol'zovanie Mineral'nogo Syr'ya, vol. 10. pp. 61–63.

WELTY, J.R., WICKS, C.E., WILSON, R.E. and RORRER, G. 2001. Fundamentals ofMomentum, Heat and Mass Transfer. Wiley, New York.

YEATTS, L.B. and RAINEY, W.T. 1965. Purification of zirconium tetrafluoride.Technical report ORNL-TM-1292, US Atomic Energy Commission. ◆



Ferrous 2016FERROUS AND BASE METALS

DEVELOPMENT NETWORK CONFERENCE 201615–17 October 2016

BACKGROUNDThrough its Advanced Metals Initiative (AMI) the South African Department of Science and Technology (DST) promotes research, developmentand innovation across the entire value chain of the advanced metals field. The goal of this initiative is to achieve sustainable local mineralbeneficiation and to increase the downstream value addition of advanced metals in a sustainable manner. The achievement of this is envisionedto be through human capital development on post-graduate and post-doctoral level, technology transfer, localization and ultimately,commercialisation.The AMI comprises four networks, each focussing on a different group of metals. These are Light Metals, Precious Metals, Nuclear Materialsand Ferrous and Base Metals (i.e. iron, steel, stainless steels, superalloys, copper, brass, etc.).The AMI FMDN 2015 Conference aims to bring together stakeholders from the mineral sector, academia, steel industry, international researchinstitutions and government in order to share and debate the latest trends, research and innovations, specifically in the areas of energy,petrochemical, corrosion, materials for extreme environments and transport, local mineral beneficiation and advanced manufacturing relatedto these materials. Keynote speakers to be invited include international specialists in the fields of ferrous metals, computational materials science, high temperaturecorrosion and mineral beneficiation.The Ferrous and Base Metals Development Network (FMDN) of the DST’s Advanced Metals Initiative (AMI) programme will host the AMI’sannual conference in 2016. The conference seeks to share insight into the state of R&D under the AMI-FMDN programmes and explore anddebate the following broad themes:

� Development of high performance ferrous and base metal alloys for application in the energy and petrochemical industries

� Development of corrosion resistant ferrous and base metal alloys

� Development of lightweight and/or durable steels for cost-effective transportation and infrastructure, and

� Panel discussions on possible future value-adding R&D programmes under FMDN within the South African National Imperatives.

WHO SHOULD ATTENDStakeholders from the energy, petrochemical, corrosion andtransportation industries where ferrous (i.e. iron, steel, stainless steels,superalloys, etc.) and base (i.e. copper, brass, etc.) metals are used intheir infrastructure. Also included in this invitation are local andinternational Higher Education Institutions (HEIs), GovernmentDepartments and Science Councils that are involved and/or have interestin R&D in these areas.


our future throughscience

For further information contact:Head of Conferencing, Raymond van der BergSAIMM, P O Box 61127, Marshalltown 2107

Tel: +27 11 834-1273/7 ·Fax: +27 11 833-8156 or +27 11 838-5923E-mail: [email protected] · Website: http://www.saimm.co.za

Conference

Announcement

OBJECTIVESInsight into ferrous and base metal materials R&D for applicationin the areas of energy, petrochemical, corrosion, extremeenvironments, improved processing technologies and advancedalloys for the transport industry in South Africa and globally.

EXHIBITION/SPONSORSHIPSponsorship opportunities are available. Companies wishing tosponsor or exhibit should contact the Conference Co-ordinator.

IntroductionGlow discharge optical emission spectroscopy(GD-OES) combines the detection range,speed, and lack of interferences of aninductively coupled plasma optical emissionspectroscopy (ICP-OES) instrument with solidsampling reminiscent of X-ray fluorescence(XRF) techniques. Its ability to perform depth

profiles of materials that contain layers ofvarying composition and thickness allows notonly for the quantification of a sample, butalso the evaluation of its composition in termsof homogeneity. Unfortunately the method isnot yet as well established as other techniques.In order to be at its most effective it requiresconductive samples. GD-OES was developed byGrimm in 1967 (Payling, Jones, and Bengtson,1997). It was initially used to analyse solidmetallic samples (bulk analysis) where allsurface effects were eliminated by a pre-burn.In the 1970s, Roger Berneron, among others,began investigating the phenomena thatoccurred during this pre-burn period anddeveloped GD-OES as a surface analyticaltechnique. In 1988 M. Chevrier and RobertPassetemps first successfully applied an RFpotential to a conventional Grimm-type source(GD) to produce an RF device that was able toalso analyse non-conductive samples.

Currently, GD-OES is still used mainly inthe analysis of solid metallic samples(Azom.com, 2004). Its most useful applicationis as a depth profiling technique on sampleswith varying layers of both conducting andnon-conducting materials (Shimizu et al.,2003), but it is also capable of rapid bulkanalysis of homogenous samples (bothconducting and non-conducting). A typicalanalysis takes only a few minutes, which isone of the main advantages of GD-OES,especially where quick results are required, forexample at a smelter. In GD-OES analysis,when using a Grimm-type direct current (DC)lamp source, a potential difference of the orderof 1 kV is applied between two electrodes in acell containing a gas, usually argon, at lowpressure (around 1 Torr) (Bogaerts and

Glow discharge optical emissionspectroscopy: a general overview withregard to nuclear materialsby S.J. Lötter*†, W. Purcell* and J.T. Nel†

SynopsisGlow discharge optical emission spectroscopy (GD-OES) is an analyticaltechnique used in the analysis of solid, conducting materials. Thoughprimarily of interest as a depth profiling technique on samples withvarying layers of both conducting and non-conducting materials, it is alsocapable of rapid bulk analysis of homogenous, solid samples. GD-OEScombines the advantages of ICP-OES (wide detection range, speed, andlack of interferences) with the solid sampling of XRF techniques. Thisallows the analyst to not only quantify the elemental composition of asample, but to evaluate it in terms of homogeneity with depth, a field inwhich auger electron spectroscopy (AES) and secondary ion massspectrometry (SIMS) have traditionally been the primary techniques.Although GD-OES does not replace these useful techniques, it does offervarious advantages over them, making it an excellent complementaryanalytical tool.

In GD-OES analysis, a low-pressure glow discharge plasma isgenerated with the sample material acting as a cathode, accelerating thecations in the plasma towards the sample surface. This bombardmentcauses the sample material to ‘sputter’, essentially knocking free atoms ormolecules of analyte material, which are then drawn into the plasmawhere they are excited. The light emitted by this excitation is thendiffracted to separate the wavelengths emitted by the specific elementsand detected by a spectrophotometer. The intensity of the signal is directlyproportional to the quantity of analyte element present in the sample,allowing simple calibration and quantitative determination of theelements.

Technological improvements made in the past twenty years or so havesignificantly increased the usefulness of GD-OES for surface analysis.Faster plasma stabilization and start-up allow for the quantification ofsurface layers as thin as 1 nm. Sputter rates have been accuratelymeasured for many common materials, allowing them to be built intosoftware libraries in the instrument’s control software. This dramaticallyexpands the usefulness of this software and the ease of performinganalyses.

GD-OES analysis of nuclear materials allows for the rapid determi-nation of the elemental composition without the requirement of initialdissolution. The thickness of any corrosion layers on nuclear materials canalso be determined.

Keywordsglow discharge optical emission spectrometry, GD-OES, nuclear materials.

* Department of Chemistry, University of the FreeState, South Africa.

† The South African Nuclear Energy Corporation SOCLtd.

© The Southern African Institute of Mining andMetallurgy, 2015. ISSN 2225-6253. Paper receivedJuly 2015 and revised paper received Aug. 2015.



Glow discharge optical emission spectroscopy

Gijbels, 1998). This causes the gas to ionize into positive andnegative ions, forming the plasma. The electrons areaccelerated towards the anode, sustaining the plasma andgenerating more positive ions. The positive ions are similarlyaccelerated towards the cathode, the sample material. Theresulting bombardment of the sample material causes it to’sputter’, essentially knocking free atoms or molecules ofanalyte material. These atoms are then drawn into the plasmawhere they too are excited. A diagram of this process can beseen in Figure 1. A radio frequency (RF) lamp may also beused for analysing non-conducting samples (Winchester andPayling, 2004), see Figure 2. The sputtered analyte is thendrawn into the high-energy plasma where it is excited into ahigher electronic energy state. In order to return to theirground state the analyte ions emit a photoelectron of awavelength characteristic of the element emitting it.

In GD-OES the characteristic light emitted by thisexcitation/de-excitation process is passed into a spectropho-tometer through an entrance slit, and is then diffracted by aconcave holographic mirror grating that contains 3600 linesper millimetre. This is done in order to separate the emittedlight into wavelengths that are specific for each element. The

individual wavelength intensities can then be detected usingeither a charge-coupled device (CCD) or a photomultipliertube (PMT). The intensity of the signal is directly propor-tional to the quantity of analyte element present in theplasma, allowing simple calibration and quantitative determi-nation of elements. In practice this is often achieved using aRowland circle (Figure 3) with PMTs positioned behind exitslits that are positioned at the wavelength corresponding tocharacteristic emission lines of various elements. This allowsthe instrument to read all of these lines simultaneously,greatly increasing sample throughput rate at the cost of theability to generate a complete spectrum. As there is only asingle PMT installed for most elements in this type ofspectrometer, detection of interferences is more difficult thanwith scanning sequential instruments. A sequentialinstrument would, however, not be capable of the most usefulaspect of the GD-OES, which is its ability to perform depthprofiling. The signals that are captured by the PMTs arerecorded by a computer at a typical rate of 10 000 to 30 000measurements per second.

Technological improvements made in the past twentyyears or so have significantly increased the usefulness of GD-OES for surface analysis (Azom.com, 2004). Faster plasmastabilization and start-up times allow for the quantification ofsurface layers as thin as 1 nm. Sputter rates have beenaccurately determined for many common materials, allowingthem to be built into the software libraries of the instrument’scontrol software. This significantly expands the usefulness ofthe software. Improvements in calibration systems allow forthe changing of instrumental conditions without the need tocompletely recalibrate, saving time and effort.

In glow discharge mass spectrometry (GD-MS), a massspectrometer is attached to the GD source rather than anoptical detector. This allows for a much lower limit ofdetection at the cost of linear dynamic range. The instrumentis otherwise essentially the same as a GD-OES instrument. Acollision cell can be used at the interface between the glowdischarge plasma and mass spectrometer in order to minimizeinterference from polyatomic species formed in the plasma.

▲


Figure 2 – Diagram of a Rowland circle used in the LECO GDS850A(Leco Corporation, 2007)

Figure 1 – Schematic of the various processes occurring during glowdischarge (GD) (Betti and de las Heras, 2004)

Figure 3 – Conceptual diagram of a radio frequency glow dischargedevice (RF GD) (Winchester and Payling, 2004)

As GD-MS still makes use of a glow discharge plasmasampling system it shares the GD-specific advantages andlimitations of the GD-OES. Due to the cost of a massspectrometer a GD-MS instrument is generally significantlymore expensive than an OES instrument.

Modern instruments are equipped with sophisticatedsoftware packages that use libraries containing the sputterrates of known materials and advanced algorithms. These areused to calculate both the depth and mass composition of thelayers, as seen in Figure 4 and Figure 5. Figure 4 shows thedepth profile of a hard-disk platter with seven distinct layersclearly visible, while Figure 5 shows the composition of thesurface of a photovoltaic cell.

InstrumentationA GD-OES instrument is a fairly simple piece of equipment tooperate, in comparison to other instruments capable ofsimilar analyses. Once a sample is properly prepared, sampleintroduction consists of simply placing a sample over the O-ring at the lamp opening and allowing the vacuum to hold itin place. A simple cross-section of a standard Grimm-type GDlamp can be seen in Figure 6.

When positioning a sample it is preferable to ensure thatno previous GD craters overlap with the surface to beanalysed. Care must be taken that samples are non-porousand smooth. If they are not, the vacuum will not be strongenough to form the GD plasma, the instrument will detectthis, and the sample will be ejected before analysis can begin.If the seal is tight the sample will then be held in place by thereamer, a pneumatically controlled ‘air jack’ with anintegrated drill bit. This reamer serves multiple functions,including holding the sample in place, cleaning the anodebetween analyses, and serving as a conductor to allow thesample to act as cathode (when using the DC lamp). Whenmoving into position the reamer impacts the sample withsome force, which can break brittle samples.

The LECO GDS850A (Figure 7) supports both a DC andan RF lamp. The RF lamp is intended for use with non-conducting materials and has a lower effective applied power.It was found that using this lamp with the copper anodeallowed only for the analysis of very thin, regular, non-conducting layers on a conducting substrate. The DC lamp isable to generate a greater applied power, which can increasesensitivity but is also more prone to short-circuits.



Figure 5 – Depth profile of the absorber layer of a copper indium galliumselenide (CIGS) thin film photovoltaic (PV) cell by pulsed RF GD-OES(Horiba Scientific, 2014)

Figure 4 – Three replicate quantitative depth profiles (QDPs) of thesurface of a hard disk exhibiting seven resolved layers at a depth of 100nm (Leco Corporation, 2007)

Figure 7 –The LECO GDS850A (Leco Corporation, 2007)

Figure 6 – Cross-section of a glow discharge lamp showing vacuum,cooling, and gas flow lines (Payling, Jones, and Bengtson, 1997)


In order to ensure that the Rowland circle is in alignment,the instrument is equipped with two PMTs mounted atdiffraction angles for detecting the characteristic wavelengthsof iron. These two iron ‘lines’ are used when profiling theinstrument. This is done by analysing a sample containingiron. Detecting these two lines allows the instrument toconfirm the position of the spectrometer’s entry slits. As theinstrument stores calibration data it is necessary to drift-correct a method’s calibration before quantitative analysis.This is performed by analysing standard samples, pre-selected in the method, and adjusting the calibrationaccording to the intensities detected. This drift must be donewhenever a different method is used but the warm-up andprofiling need to be performed only once per working period.

The use of GD-OES in the nuclear industryAs previously mentioned, the most striking capability of GD-OES is its ability to accurately and rapidly determine a depthprofile of multiple layers of varying composition on asubstrate. The raw data from a GD-OES analysis is given intime versus intensity of emission, as can be seen in Figure 8,where layers of UO2F2·1.5H2O, and UF4 are seen bonded to anickel substrate. This data, along with known sputter ratesfor each material, allows for the characterization of thevarious uranium species present on the surface of thematerial.

In the 1980s the erstwhile Atomic Energy Corporation ofSouth Africa (AEC) used GD-OES intensively to study surfaceand interface phenomena like corrosion, surface cleanness,passivation of surfaces, decontamination efficiency, plating,etc. (Nel, 1991). Valuable information could be obtained byanalysing surfaces of materials that were exposed to UF6, asindicated in Figure 8. The types of uranium compounds thatwere found on a surface were used by nuclear engineers andscientists to choose the best materials for construction or todiagnose chemical and material problems that occurredduring plant operation.

GD-OES was also routinely used for the bulk analysis ofmaterials that are used in nuclear and uranium enrichment

plants. Examples of these are high nickel-containing alloys,normal and special stainless steels (e.g. 316, 304, and 17-PH), Monel, Inconel, aluminium alloys, copper, and brass,etc. The impurity specifications for nuclear materials areusually extremely strict and the low limit of detection of GD-OES makes this technique extremely powerful in the nuclearindustry.

In the current programme of the Advanced MetalsInitiative (AMI), and especially the Nuclear MaterialsDevelopment Network (NMDN), one of the main focuses isthe development of nuclear-grade zirconium metal and thestudy of corrosion of zirconium alloys.

GD-MS has been used in the analysis of spent uraniumoxide fuel (Robinson and Hall, 1987) prior to recycling. It isnecessary to confirm that the reprocessing plant is workingefficiently and that fission products, activation products, andcorrosion products are successfully removed. GD-MS offers afaster and more quantitative approach than traditionalmethods, such as spark source mass spectrometry (SSMS),without the need for time-consuming interpretation of resultsby highly skilled operators.

ExperimentalSamples of Zircaloy 2 and Zircaloy 4, obtained from ATI WahChang, were exposed to air at 600°C for periods of either 16or 32 hours. The samples were allowed to cool rapidly in airand then analysed by GD-OES to determine the thickness ofthe oxide layer formed. A sample of zirconium metal,obtained from Alfa Aesar, was analysed as well in order toprovide a baseline from which to compare the degree of thegrowth of the oxide layer. The elemental composition of theZircaloy materials can be seen in Table I.

The GD-OES instrument was calibrated using NISTstandards 1212a, 1234, 1235, 1236, 1237, 1238, and 1239,obtained from Pelindaba Analytical Laboratories. Thephotomultiplier tube power settings were determined usingthe automated software tool built into the control software,with applied voltages ranging between 400 and 900 V.Instrument settings can be seen in Table II. As the RF lampwas used the Teflon isolation puck was utilized in alldeterminations and was kept at a constant temperature of12°C by an external chiller. This puck is similar to the coolingpuck used with the DC lamp but, in addition to cooling, alsoelectrically isolates the sample from the reamer.

Results and discussionOxygen readings in all cases were higher than expected, butthis was likely a calibration issue. As the oxygen calibrationvalues were for dissolved oxygen in the metal standardsrather than the stoichiometric levels found in the oxide

▲


Figure 8 – Depth profile of a thin layer containing UO2F2·1.5H2O and UF4

on a nickel substrate (Nel, 1991)

Table I

Elemental composition of Zircaloy 2 and Zircaloy 4(Roskill Information Services, 2011, p. 322)

Sample Zr % Sn % Fe % Ni % Cr %

Zircaloy 2 98.23 1.5 0.12 0.05 0.1Zircaloy 4 98.28 1.5 0.12 - 0.1

layers, the intensities detected were several orders ofmagnitude higher than the calibration range. Oxygen valuesmust thus be described as semi-quantitative. As the objectiveof the study was to determine the thickness of the layersrather than their elemental composition, this was notconsidered too great of a difficulty.

The oxide layer formed naturally at room temperature byzirconium metal was found to be approximately 100 nm thickwith approximately another 200 nm required for thetransition to mostly pure zirconium, as seen in Figure 9. Thethickness of the corrosion layers varied between 31 and 100µm, depending on the duration of and temperature at whichthe corrosion test was performed (Table III). The least



Table II

GDS850A settings

Setting Value

Voltage (V) 1200True plasma power (W) 20

Vacuum (Torr) 6 - 12Analysis time (min) 10

Figure 10 – Quantitative depth profiles of Zircaloy 2 and 4 after 16hours’ exposure in air at 600°C

Figure 9 – Quantitative depth profile of a clean plate of zirconium metalbefore corrosion

Figure 11 – Quantitative depth profiles of Zircaloy 2 and 4 after 32hours’ exposure in air at 600°C

Figure 12 – Quantitative depth profiles of Zircaloy 2 after 16 and 32hours’ exposure in air at 600°C

Figure 13 – Quantitative depth profiles of Zircaloy 4 after 16 and 32hours’ exposure in air at 600°C

Table III

Measured thicknesses of Zircaloy corrosion layers

Sample Time at 600°C (h) Layer thickness (μm)

Zircaloy 2 16 35Zircaloy 2 32 100Zircaloy 4 16 31Zircaloy 4 32 60


corrosion was exhibited by Zircaloy 4, with layer thicknessmarginally thinner than those seen with Zircaloy 2 at 16hours but significantly thinner after both had been corrodedfor 32 hours. These results are clearly visible in Figure 10and Figure 11. The increase in oxide layer thickness of 285.7% for Zircaloy 2 can clearly be seen in Figure 12, while theZircaloy 4 showed an increase of only 193.5 % (Figure 13).As expected, Zircaloy 4 exhibited a significantly betterresistance to corrosion than Zircaloy 2.

These results clearly show the obvious utility of GD-OESfor the rapid determination of oxide layers on nuclear fuelcladding. It can be highly useful as a method to quantify thecorrosion of known and experimental materials for nuclearapplication. Its usefulness extends to the determination oflayer thickness and composition for known and potentialcoatings on nuclear and non-nuclear materials.

ReferencesAZOM.COM. The A to Z of Materials. 2004.

http://www.azom.com/article.aspx?ArticleID=2449 [Accessed 9 July 2013].

BETTI, M. and DE LAS HERAS, L.A. 2004. Glow discharge spectrometry for thecharacterization of nuclear and radioactively contaminated environmentalsamples. Spectrochimica Acta Part B, vol. 59. pp. 1359–1376.

BOGAERTS, A. and GIJBELS, R. 1998. Fundamental aspects and applications ofglow discharge spectrometric techniques. Spectrochimica Acta Part B, vol.53. pp. 1–42.

HORIBA SCIENTIFIC. 2014. Ultra fast elemental depth profiling.http://www.horiba.com/fileadmin/uploads/Scientific/Documents/Emission/GDPROFILER_Series.pdf [Accessed 18 April 2015].

LECO CORPORATION. 2007. GDS850A Glow Discharge Spectrometer.http://www.leco.com/component/edocman/?task=document.viewdoc&id=49&Itemid=0 [Accessed 4 September 2013].

NEL, J.T. 1991. Oppervlakanalises van uranielfluoriedlagies met 'n gloeiontlad-ingslamp. DPhil thesis, University of Pretoria.

PAYLING, R., JONES, D. and BENGTSON, A. 1997. Glow Discharge Optical EmissionSpectrometry. Wiley, Hoboken, NJ.

ROBINSON, K. and HALL, E.F.H. 1987. Glow discharge mass spectrometry fornuclear materials. Journal of Metals, vol. 39. pp. 14–16.

ROSKILL INFORMATION SERVICES. 2011. Zirconium: Global Industry Markets andOutlook. 13th edn. London.

SHIMIZU, K., HABAZAKI, H., SKELDON, P. and THOMPSON, G.E. 2003. Impact of RF-GD-OES in practical surface analysis. Spectrochimica Acta Part B, vol. 58.pp. 1573–1583.

WINCHESTER, R.M. and PAYLING, R. 2004. Radio-frequency glow dischargespectrometry: a critical review. Spectrochimica Acta Part B, vol. 59. pp. 607–666. ◆

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IntroductionGrain refinement has been found to increaseboth the strength and toughness of steels (Gaoand Baker, 1998; Sharma, Lakshmanan, andKirkaldy, 1984; Seok et al., 2014; Maalekianet al., 2012). It is also known that theaustenite grain size directly influences themicrostructure, and thus the mechanicalproperties, of the steel (Maalekian et al., 2012;Yue et al., 2010). Effective grain growthcontrol is reported to be achieved throughaddition of precipitate-forming elements, suchas Nb that, slow down the grain boundarymigration through pinning and solute dragmechanisms (Yu et al., 2010; Olasolo et al.,2011; Alogab et al., 2007). Much work hasbeen carried out on austenite grain lsizecontrol by the addition of precipitate-formingelements that have a strong affinity forinterstitial elements, such as carbon andnitrogen, which form the dispersed pinningparticles to inhibit the austenite grain growth(Alogab et al. 2007; Hodgson and Gibbs,1992; Nanba et al., 2003; Flores and Martinez,1997; Rollett, Srolovitz and Anderson, 1989).Austenite grain growth can be described usinga conventional or a modelling approach. Withconventional approaches (e.g. metallography),

in-situ monitoring of austenite grain growth athigh temperatures is impossible. Modelling thegrain growth behaviour based on availabledata is the ideal option. To quantitativelydescribe austenite grain growth thereforerequires development of a sound mathematicalaustenite grain growth equation that accountsfor the effects of the varying microalloyingelements in inhibiting austenite grain growth.Numerous attempts have been made todevelop an empirical model based on thegeneral equation developed by Sellars andWhiteman (1979). Many of these models donot account for the direct effects of themicroalloying elements such as Nb in austenitegrain growth control [Fu et al., 2011; Wangand Wang, 2008; Wang et al., 2006;Shanmugama et al., 2005; Banerjee et al.,2010; Pous-Romeroa et al., 2013). The currentwork has considered this limitation, takinginto account the direct effect of niobium ingrain growth control during thermalprocessing. This is done by incorporating theinitial austenite grain size Do and the microal-loying element niobium in the development ofa constitutive equation for grain growthprediction in Nb-containing microalloyedsteels.

Materials and techniques Table I shows the chemical compositions of thefive microalloyed steels used in the study,which were cast by vacuum induction meltinginto ingots of 16 kg. The compositions werechosen to test the effect of the Nb content onthe austenite grain growth while the otherelements were kept approximately constant.Note, however, that the melting and castingprocedure used for these laboratory steels

The influence of niobium content onaustenite grain growth in microalloyedsteels by K.A. Annan, C.W. Siyasiya and W.E. Stumpf

SynopsisThe relationship between niobium content and austenite grain growth hasbeen investigated through hot rolling simulation on a Bähr dilatometer.The effect of delay time between passes during rough rolling in Nb-microalloyed steels with nitrogen contents typical for electric arc furnace(EAF) melting was studied. The results indicate that the grain growthconstants n, Q, and A increase with an increase in Nb content. Theactivation energy for austenite grain growth Q was found to be in therange of 239 to 572 kJ/mol, the exponential constant n ranged from 2.8 to6.2, and the material and processing condition constant A from 4.24 × 1012

to 4.96 × 1028, for steels with niobium contents ranging from 0.002% Nb to0.1% Nb. A general constitutive equation for the prediction of austenitegrain growth in these Nb-microalloyed steels under rough rollingconditions has been developed. Good agreement between the experimentaland the predicted values was achieved with this constitutive equation.

Keywordsconstitutive equation, austenite grain growth, microalloying, deformation.

* Department of Materials Science and MetallurgicalEngineering, University of Pretoria, South Africa.




The influence of niobium content on austenite grain growth in microalloyed steels

resulted in somewhat higher nitrogen contents, which arenearer to those expected from electric arc furnace (EAF) thanfrom basic oxygen furnace (BOF) steelmaking.

Compression test on a Bähr dilatometer Cylindrical samples (5 mm diameter × 10 mm length)machined from the laboratory cast slabs were heated byinduction in the Bähr dilatometer to 1150°C at a rate of 5°C s-1

and held at 1150°C for 5 minutes to achieve homogenization.The samples were then cooled at 5°C s-1 to the testingtemperature (1000, 1050, 1100, and 1150°C) where a single-hit compression was applied after holding at the testtemperature for 20 seconds. A strain of 0.4 was used at astrain rate of 0.1 s-1. After compression, the samples werethen held at the deformation temperature in the Bährdilatometer for times of 0, 10, 30, 60, 90, or 120 minutes tosimulate the delay time after rough rolling, before rapidcooling at a rate of 600°C s-1 to room temperature. Oxidationof the specimens during compression was prevented bypassing a continuous flow of high-purity helium through thesystem. The cooled samples were then tempered at 490°C for72 hours to improve the response of prior austenite grainboundaries to etching. In this study, the grain size of thesamples held for zero (0) minutes after deformation was usedas the initial grain size D0 for the steel at that temperature.The scheduled profile followed in the single hit compressiontests is shown in Figure 1.

The samples were mounted, ground, polished, and thenetched with picric acid solution containing 100 ml saturatedaqueous picric acid, 100 ml distilled water, 4 g sodium

tridecylbenze sulphonate, and 2-3 drops of triton. Etchingwas carried out at a temperature of 60–70°C for times rangingfrom 3 to 14 minutes. The samples were observed under anOlympus BX51MTM optical microscope to reveal the prioraustenite grain boundaries. The austenite grain size wasmeasured using the average linear intercept methodaccording to ASTM E112 (Van der Voort, 1984). To obtain astatistically acceptable grain size distribution, more than 300intercepts were measured on each sample.

Results

The solubility behavior of precipitates in microalloyedsteels predicted by Thermo-CalcTMThe volume fractions of precipitates as a function of thetemperature predicted by Thermo-CalcTM for the studiedsteels are shown in Figure 2. These show that the reheatingtemperature range of 800–1250°C will lead to dissolution of asubstantial amount of Ti,Nb(C,N) carbonitrides but not to acomplete dissolution of all of these precipitates. NbN precip-itates within the temperature region of 1000–1200°C, whileTiN does not go into solution at temperatures considered inthis study. The higher dissolution temperatures shown forNbN in these microalloyed steels are due to their higher Ncontent of about 0.019 wt%, which is more typical for steelproduced in an EAF (N in solute-rich regions preferentiallycombines with Nb). The presence of precipitates means that aunimodal grain structure can be predicted if these precipitatesare effective in pinning the grain boundaries (Fernández,Illescas and Guilemany, 2007; Van der Voort, 1984; Zener,1948 (as cited by Ringer, Li and Easterling, 1989).

The volume fraction of Nb precipitates with respect totemperature as predicted by Thermo-CalcTM is shown inFigure 3. This shows that the volume fraction at a giventemperature is highly dependent on the Nb content in thesteel, thus the higher the Nb content, the higher the volumefraction of the precipitate at any given temperature.

Austenite grain growth behaviour in the steels afterdeformationFigure 4 shows the grain size distribution in the Nb-bearingsteels after high-temperature deformation followed bytempering at 490°C for 72 hours. The steels showed aunimodal grain structure during deformation. These distrib-utions and structures attest to the fact that no abnormal graingrowth, i.e. no bimodal distribution, was visible in thesesteels at the thermomechanical controlled processing (TMCP)conditions employed in this work, confirming the presence ofpinning particles in the steels as predicted by Thermo-CalcTM.

▲


Table I

Chemical composition of the steels used in the study

Steel Chemical composition (wt %)C Mn Si Nb Ti V Al N Ni

0.002% Nb 0.17 1.13 0.11 0.002 0.001 0.007 0.05 0.018 0.150.03% Nb 0.16 1.15 0.15 0.03 0.02 0.008 0.02 0.018 0.150.05% Nb 0.17 1.15 0.15 0.05 0.02 0.009 0.03 0.019 0.150.07% Nb 0.14 1.10 0.15 0.07 0.02 0.006 0.04 0.020 0.170.10% Nb 0.18 1.27 0.15 0.10 0.02 0.009 0.04 0.020 0.17

Figure 1 – Schematic representation of the deformation process on theBähr dilatometer

Increased Nb content and austenite grain growth inthe microalloyed steelsThe observed effect of an increased Nb content in themicroalloyed steels is shown in Figure 5. It is evident thatthere was limited grain growth in the steel containing 0.1wt% Nb, as it produced the smallest final grain sizes, whilethe highest grain growth was seen in the 0.002 wt% Nb steel,thereby following the volume fraction of Nb precipitates inthe steel. The observed results are in agreement with Zener’s

basic relationship for the pinning force of particles on grainboundaries

(Zener, 1948, as cited by Ringer, Li, and Easterling, 1989),which predicts a stronger effect of pinning with increasedvolume fractions of small precipitates at constant particlesizes. The higher Nb content accounts for a greater volumefraction of NbN precipitates as shown in Figure 3, providing agreater area fraction of solute-rich regions compared to thesteels containing less Nb (Xu and Thomas, 2011; Brewer,Erven and Krauss, 1991; Deus et al., 2002; Akamatsu,Senuma and Hasebe, 1992). The quantitative analysis of theoptical measurements of austenite grain growth, is shown inFigure 6, which confirms the substantial influence of anincreased Nb content on austenite grain growth inhibition inthese steels.

Quantitative evaluations of the average grain size as afunction of austenitizing time and temperature in the 0.03wt% Nb microalloyed steel are shown in Figures 7 and 8respectively.A comparative analysis of Figures 7 and 8indicates the expected greater effect of temperature comparedwith time on grain growth, as recorded by numerous studies(Seok et al., 2014; Yue et al., 2010; Nanba et al., 1992;Akamatsu, Senuma and Hasebe, 1992; Sha and Sun, 2009;Zhao et al., 2011)



Figure 2 – Thermo-CalcTM predictions of the volume fraction of precipitates in the high-N laboratory-produced steels with (a) 0.03 wt%, (b) 0.05 wt%, (c)0.07 wt%, and (d) 0.1 wt% Nb

Figure 3 – Effect of Nb content on volume fraction of niobiumprecipitate as predicted by Thermo-calcTM


▲


Figure 4 – Optical micrographs and grain size distribution for (a) an 0.07 wt% Nb deformed at 1100°C at a strain rate of 0.1 s-1 to a strain of 0.4, (b) 0.05 wt%Nb deformed at 1150°C at a strain rate of 0.1 s-1 to a strain of 0.4

Figure 5 – Optical micrographs of (a) 0.002 wt%, (b) 0.03 wt%, (c) 0.05 wt%, (d) 0.07 wt%, and (e) 0.1 wt% Nb steels deformed at a temperature of 1100°Cand held for 120 minutes

Development of a constitutive equationIt is generally accepted that the grain boundary migrationvelocity (v) is proportional to the driving force (Seok et al.,2014; Maalekian et al., 2012; Yue et al., 2010; Olasolo et al.,2011; Alogab et al., 2007; Hodgson and Gibbs, 1992; Nanbaet al., 1992), an approximation justified at the relativelysmall driving forces involved in grain growth. This velocity isgiven by the expression (Stumpf, 2010)

[1]

where M is the grain boundary mobility, Fd is the drivingforce for grain growth, and Fp is the pinning force induced byprecipitates.

From the rate of movement:

[2]

The grain boundary mobility M is given by the expression

and the grain growth rate equation therefore becomes:

[3]

Under isothermal conditions

[4]

The differential solution to Equation [4] is given by(Sellars and Whiteman, 1979)

[5]

where T is the absolute temperature, t is the time, Q is theactivation energy for grain boundary migration, A is aconstant parameter dependent on the material and processingconditions, and R is the universal gas constant. A generalsolution to Equation [4] in a linear form is expressed inEquation [6]:

[6]

Equation [5] has been used by many authors (Seok et al.,2014; Maalekian et al., 2012; Nanba et al., 1992; Florez andMartinez, 1997; Rollett, Srolovitz and Anderson, 1989;Sellars and Whiteman, 1979; Fu et al., 2011; Wang andWang, 2008; Wang et al., 2006) by assuming the value of Doto be zero or constant. This simplifies the plot of lnD as afunction of 1/T. In the current study Do, which was experi-mentally measured, was not discarded or kept constant butwas included and used in Matlab programming and an Excelsolver to plot

as a function of 1/T in generating the constants used in thedevelopment of the constitutive equations. This was donethrough regression analyses of the first and second derivativeorders of the experimentally derived regression value R2.



Figure 6 – Influence of Nb content on the average austenite grain size(AGS), showing a strong inhibition in Nb-containing steels

Figure 7 – Isothermal grain growth behaviour of austenite in steels,showing an increase in the AGS with time in the 0.03 wt% Nb steel afterhot deformation

Figure 8 – Isothermal grain growth behaviour of austenite in steels,showing an increase in the AGS with temperature in the 0.03 wt% Nbsteel after initial hot deformation at the same temperature


The grain growth constants n, Q, and AThe constant n was obtained by regression analysis ofexperimentally measured grain sizes through iterative tests ofthe straight-line regression coefficients R2, from a plot ofEquation [6] where a second differential of the R2,(dR2)2/dn2 with respect to n gave a maximum correlationvalue of n at (dR2)2/dn2 = 0.

Figure 9 shows a plot of ln(−) as a function of the inversetemperature 1/T, which was used in determining theconstants. Thus, a plot of Equation [6] provided a linearrelationship between

and 1/T. Comparison of Equation [6] with the equation of astraight line made it possible for the constants Q and A to bedetermined from the plots shown in Figure 9. Thus,

(while the slope is negative) and A = ec/t where, c is theintercept from the graph and t is the time.

The dependence of the grain growth constants n, Q, andA on the Nb content was quantitatively analysed by multipleregression to obtain the following reliance equations, where[Nb] is the wt% in the steel:

The statistical analysis resulting from the multipleregression analysis presented in Table II, showing that theactivation energy Q has a strong linear relationship with theNb content, while there is no linear relationship between Aand the Nb content. There is, however, a weak linearrelationship between the grain growth constant n and the Nbcontent.

The constants derived from experimentally measuredgrain sizes for the microalloyed steels under deformationconditions with and without taking Do into consideration arepresented in Table III.

General constitutive equation for austenite graingrowth prediction in Nb-bearing microalloyed steelswith high N contentsBased on the analysis of the isothermal grain growth kineticsin Nb-microalloyed steels, the following isothermal grain

growth equation for austenite grain growth prediction in0.002–0.1% Nb microalloyed steels during deformationwithin the temperature range of 1000–1150°C and delaytimes ranging from 0–120 minutes is constituted for DDef inμm:

where Do is the initial austenite grain size in μm and [Nb] isin wt %.

▲


Figure 9 – Plot of the natural log of the average grain size as a functionof the inverse of the absolute temperature for the 0.07 wt% Nb steel

Table III

Grain growth constants generated from deformation data of Nb-microalloyed steels

Constants determined from deformed Nb-microalloyed steels

Steel (wt % Nb) Constants using measured values of Do Constants with Do= 0 Difference in constants from neglecting D0

n A Q n A Q Δn ΔA ΔQ

0.002 2.8 4.24E+12 276 2.8 3.00E+11 239 0 3.94E+12 37.270.03 5.5 1.33E+22 417 5.5 6.60E+21 384 0 6.70E+21 33.100.05 5.6 7.94E+23 474 5.5 5.00E+23 428 0.1 2.94E+23 46.310.07 6.0 9.63E+24 489 6.1 4.62E+24 459 -0.1 5.01E+24 30.410.1 6.2 4.96E+28 572 6.2 1.22E+28 559 0 3.74E+28 12.67

Table II

Statistical results of the multiple regressionanalysis of the dependence of the austenite graingrowth constants on varied Nb content

Regression statistics Grain growth constants

Q n AAdjusted R square 0.91 0.64 0.40Standard error 32.31 0.83 1.7 × 1028

Number of observations 5 5 5ANOVA 0.0072*** 0.066* 0.152*** p<0.01 (this means there is a very strong relationship with probability <1%)

* p<0.1 (this indicates a weak relationship with probability <10%)

Discussion

Austenite grain growth behaviour of Nb-bearingmicroalloyed steelsThe grain growth rate was significantly slower inmicroalloyed steels with higher contents of Nb. Grain growthcontrol in microalloyed steels has been shown to bedependent on the amount of microalloying additions (Aloganet al., 2007; Hodgson and Gibbs, 1992). The greatestinhibition here was recorded in the steel containing 0.1 wt%Nb. It was also found that the average austenite grain sizedecreased with increasing Nb content in the steels. Generally,it has been reported that increasing the N content in the steelleads to a beneficial increase in grain refinement (Alogan etal., 2007; Hodgson and Gibbs, 1992; Nanba et al., 1992),which without doubt results from the larger volume fractionof the precipitates that act as pinning particles. It should benoted, however, that while increased N content is beneficialfor grain refinement, excessive levels of N dissolved inaustenite may be detrimental to other properties of the steel,such as hardenability, through the decrease of the Nb fractiondissolved in the austenite, which is a key in improving mostof the desired mechanical properties of the steel (Alogan etal., 2007; Hodgson and Gibbs, 1992; Nanba et al., 1992).

Predictive potential of the equation for grain growthin microalloyed steels in the current studyA logical degree of precision in predicting austenite graingrowth in Nb bearing steels has been achieved by comparisonof experimentally measured grain sizes with predicted grainsizes using the general constitutive equations developed in

this study, as shown in Figure 10. The correlation R2 valuesobtained from the comparison of the experimentallymeasured grain sizes with the predicted grain sizes are 0.94,0.91, and 0.90 for 0.03 wt% Nb, 0.05 wt% Nb, and 0.07 wt%Nb steels respectively. The 0.1 wt% Nb steel, however,recorded a poor correlation value of R2 = 0.63, showing theinability of the equation to predict grain sizes for steelscontaining 0.1 wt% Nb and also most likely for those withhigher Nb contents. This is due to the inability of thedeveloped equation to account for the excess precipitates thatform owing to the high addition of the microalloying element.

Conclusions➤ A general constitutive equation for predicting austenite

grain growth in Nb-bearing steels that incorporates theNb content and then the starting grain size Do aftersimulated rough rolling has been developed

➤ There is a linear relationship between Nb content andthe activation energy for grain growth

➤ The austenite grain growth exponent n, the activationenergy Q, and the material and processing conditionsconstant A all increase with an increase in Nb content,most likely from increased effectiveness of grain-boundary pinning in the microalloyed steel withincreased Nb content

➤ The constitutive equation shows a poor correlation forthe steel containing 0.10 % Nb, with an R2 value of0.63 recorded in the plot of measured against predictedgrain sizes for the 0.1 wt% Nb steel. The equation ismost likely not applicable at such high values of Nb.



Figure 10 – Comparison of predicted and measured austenite grain sizes in Nb-bearing steels


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▲


IntroductionOwing to the price volatility of nickel overrecent years, the need to cut down on the useof nickel in the steelmaking industry hasbecome a matter of keen interest, hence therising interest in ferritic stainless steels toreplace their austenitic counterparts forvarious industrial applications. Ferritic grades,containing chromium and possibly otherstabilizing elements (Ti, Mo, Nb, etc.) are wellknown as cost-saving materials as they do nothave the expensive nickel additions (Charles et

al., 2008), while at the same time presentinggood mechanical properties similar to thoseoffered by austenitic steel grades. Moreover,some standard ferritic grades such as 409,410, and 430 are readily available all over theworld, and are already very successfully usedfor various applications, such as washing-machine drums and exhaust systems in theautomotive industry, and actually have a muchbroader application potential in many fields(Charles et al., 2008). Since new designs inthe automotive industry are driven by safetyand environmental concerns, the need toreduce the weight of the exhaust systemswhile maintaining good resistance to thermalfatigue is continuing to see the exploitation ofthe ferritic stainless steel family for alternativesolutions. Ferritic stainless steels have arelatively low coefficient of thermal expansionand, therefore, some efforts have been made tocreate new ferritic stainless steels with highyield strengths at elevated temperatures,particularly by the addition of niobium (Nb),which increases the initial high-temperaturestrength through solid solution hardening(Sello and Stumpf, 2010). The ferritic typeAISI 436 is increasingly used for automotivetrims, with a major application being for lugnuts or wheel nuts on trucks. It has also seenextensive use in the production of both thecentral and rear mufflers of exhaust systems,due to its good hot and wet corrosionresistance properties (Charles et al., 2008).

However, during the deep drawing processassociated with the manufacture of most ofthese products, tearing or cracking cansometimes occur. Deep drawing is one of themost common processes for forming metalparts from sheet metal plates, and it is widelyused for the mass production of parts in the

The influence of thermomechanicalprocessing on the surface quality of anAISI 436 ferritic stainless steelby H.J. Uananisa*, C.W. Siyasiya*, W.E. Stumpf* and M.J. Papo†

SynopsisThe need to reduce weight while maintaining good mechanical propertiesin materials used in the automotive industry has over the years seen anincreased exploitation of various steels to meet this new demand. In linewith this development, the ferritic stainless steel family has seen a wideapplication in this industry, with the AISI 436 type increasingly being usedfor automotive trims and mufflers for exhaust systems, as well as asignificant part of this steel’s application being for the manufacture ofwheel nuts and wheel nut caps in trucks, mainly through the deep drawingprocess. However, there have been reports of some poor surfaceroughening of this material during deep drawing, with tearing and/orcracking also reported in some instances. This has been suspected topossibly be associated with some local differences in localized mechanicalproperties between grains and grain clusters of the rolled and annealedmaterial.In order to investigate the poor surface roughness exhibited by AISI 436ferritic stainless steel (FSS) during deep-drawing, Lankford values (R-mean and Δr), grain size, and microtextures of various sheet samples fromthis steel were studied. The chemical composition range for the sampleswas 0.013–0.017% C, 17–17.4% Cr, 0.9–1% Mo, and 0.4–0.5% Nb. Thesteels were subjected to various hot and cold rolling processing routes i.e.involving industrial direct rolling (DR) or intermediate annealing rolling(IR), and the drawability and final surface qualities of the steels werecompared. It was found that the DR route gave an average R-mean and Δrvalue of 1.9 and -1.4 respectively, while the IR route yielded an average R-mean and Δr value of 1.6 and 0.52 respectively. The high Δr value for theDR processing route had a substantial adverse effect on the drawability. IRsamples exhibited a smoother surface finish on visual inspection, whileclear flow lines were visible on the DR samples, despite the fact that DR isthe preferred industrial processing route due to the reduced productioncosts it offers. This observation was also confirmed through SEMexaminations. The difference in the surface quality was attributed tomicrotexture. However, the mechanism responsible for this difference stillneeds to be identified.

Keywordsferritic stainless steel, formability, microtexture, EBSD.

* University of Pretoria, Department of MaterialScience & Metallurgical Engineering, South Africa.

† Advanced Materials Division, Mintek.© The Southern African Institute of Mining and

Metallurgy, 2015. ISSN 2225-6253. Paper receivedAug. 2015.



The influence of thermomechanical processing on the surface quality of an AISI 436 ferritic stainless steel

packaging industry, automotive industry, and many others. Atypical example of this process is illustrated in Figure 1,which shows some wheel-nut covers manufactured through asix-stage deep-drawing process.

In the reported cases of poor surface roughness, tearing,and cracking, the surface of the drawn products appears tohave roughened, possibly due to local differences inmechanical properties between grains or grain clusters of theraw material. This behaviour has been suspected to be acrystallographic texture effect similar to ridging and roping,which is the very common susceptibility of ferritic stainlesssteel to develop narrow ridges on the sheet surface duringforming (Knutsen and Wittridge, 2002; Raabe et al., 2003;Shin et al., 2003). The ridges or undulations result in a verydull surface appearance, which in turn reduces the surfaceshine and quality of the formed product (Knutsen andWittridge, 2002). The purpose of this study is to investigatehow various thermomechanical processes, with particular

emphasis on the processing route, influence this poor surfacequality behaviour of AISI 436. This was achieved by studyingsome crystallographic parameters of the steel, and correlatingthose parameters to surface topography, in order to helpunderstand the mechanism behind the surface roughening ofthe steel in subsequent studies. The ultimate objective is todevelop possible solutions to the problems experienced in thedeep drawing of AISI 436.

Experimental procedureVarious samples of commercial AISI 436 were received andcharacterized. These were sampled from a production trialoperation, as shown in Figure 2, in which two processingroutes, direct rolling (DR) and intermediate rolling (IR), wereinvestigated. The emphasis in this study was limited to thefinal cold-rolled and annealed samples marked F1, F2, F3,and F4 in Figure 2, with thicknesses ranging from 0.46 mmto 0.50 mm.

All the samples were processed from the same ‘productionheat’ and hence had the same chemical composition, shownin Table I.

Each sample was first mechanically polished, etched in anaqua regia solution, and examined under an opticalmicroscope to determine the grain structure morphology.Grain-size measurements were determined by the ‘Image J’line-intercept method. The samples were then furthermechanically prepared to a final fine polish in a colloidalsilica medium (OP-S) for electron backscatter diffraction(EBSD) analysis, with the scans being performed using a JeolJSM-IT300LV scanning electron microscope (SEM).

▲


Figure 1 – A multi-stage deep-drawing process for wheel-nut covers

Figure 2 – Block diagram showing the thermomechanical trial processing routes for the AISI 436 samples used in this work

Orientation distribution functions (ODFs) were furthercalculated by the series expansion method (lmax=22) usingthe SALSA post-processing software of the CHANNEL5package. The mean R-values (Rm) and planar anisotropyvalues (Δr) for the various samples were measured after 10%strain along the longitudinal (0°), transverse (90°), anddiagonal (45°) directions. The Rm and Δr values were thencalculated using the following standard equations:

[1]

[2]

The subscripts 0°, 45°, and 90° refer to the longitudinal,diagonal, and transverse directions with respect to the rollingdirection. The planar anisotropy (Δr) gives an indication ofthe amount of necking or earing that will occur on the edges

of the deep-drawn items (Maruma et al., 2013). Samples of the steel from different production heats, with

the compositions shown in Table II, were then deep-drawninto lug-nut covers through the multi-stage forming processshown in Figure 1, and analysed for surface roughness.Sample D1 was intermediate rolled (IR) prior to the deep-drawing process, while D2 was directly rolled (DR). Thisanalysis was used to investigate the effect of the productionprocessing route on the surface roughness of the steel.


Microstructural analysisFigure 3 shows the microstructures of the cold-rolled andannealed samples by optical microscopy. Grain sizemeasurements showed an average grain size value of about25 μm for the four final cold-rolled and annealed samples (F1to F4), with the respective measured values shown in TableII. It is evident from these results that the processing route



Table I

Chemical composition of the AISI 436, wt.%

C Mn Si P S Cr Mo Ni Al Cu Nb Ti Sn N

0.013 0.42 0.43 0.022 0.002 17.14 0.94 0.25 0.003 0.09 0.41 0.001 0.012 0.017

Table II

Chemical composition of the deep-drawn AISI 436 samples, wt.%

Sample C Mn Si P S Cr Mo Ni V Cu Nb Ti Co N

D1 0.015 0.46 0.55 0.02 0.0005 17.04 0.94 0.15 0.12 0.08 0.50 0.002 0.02 0.02D2 0.017 0.56 0.38 0.02 0.0005 17.23 0.90 0.24 0.10 0.10 0.37 0.003 0.02 0.02

Figure 3 – Microstructures of the samples after a 30-second deep etching in aqua regia solution. (a) F1 sample, (b) F2 sample, (c) F3 sample, and (d) F4sample


had no significant effect on grain size as the average grainsizes measured were within an average ±3 μm standarddeviation from each other.

The optical microstructures also clearly reveal variousgrains that are strongly etched (darker) for all samples, whileothers are lightly etched. This disparity in grain morphologyis caused by differences in grain orientations (Raabe andLucke, 1992). In order to evaluate the drawability of thesheet samples, measurements of the R-mean (Lankfordparameters – Rm) values for each sheet were determinedthrough tensile tests. The results, also shown in Table III,indicated a lower average Rm value for the IR route at about1.6 compared to an average Rm value of 1.9 for the DR route.It is of particular interest that the planar anisotropy parameter(Δr) for both DR samples is significantly higher in magnitudeand negative (average value of -1.4) compared to that of theIR samples (0.52). This behaviour is indicative of a rotationof earing from an angle of 0° and 90° with respect to therolling direction (RD) to a 45° or 135° direction with respectto RD, which could be attributed to the poor resultingformability of the DR samples despite their higher Rm value.

The severity of the surface roughness of the DR sampleswas further illustrated by examining one sample from eachroute by SEM. The micrographs are shown in Figure 4. Bothsamples were obtained from the ‘wall area’ of a stage twosample (as shown in Figure 1) from each route, with D1being from the IR process and D2 from the DR process. It isevident, even by visual inspection of this early forming stageof the two samples, with scans at the same magnification,that the DR sample shows severe surface roughnesscompared to the flatter surface for the IR sample. Moreover,

as may be seen from the distribution of the ‘hills and valleys’on the micrographs, some areas on the IR sample seem not tohave undergone any deformation (circled regions on ‘D1’ inFigure 4). This observation calls attention to a phenomenonmentioned by Knutsen and Wittridge (2002), who emphasizethe influence of grain size banding in the light of the Hall-Petch effect, hence possible resultant differential yieldingunder tension. That is to say, clusters consisting of differentgrain sizes would obviously respond to deformationdifferently, thereby resulting in distortion of the surface.

Texture evolutionMany studies have shown that ferritic stainless steels, whichhave a body-centred cubic (bcc) structure, have a tendency toform a preferred orientation of their grains (fibre texture)during rolling and annealing. Siqueira et al. (2011), Huh etal., (2005), and Maruma Siyasiya, and Stumpf (2013) haveelaborated on the fact that rolling deformation in most casesleads to a texture characterized by two orientation fibres – anα-fibre texture typically comprising orientations with acommon <110> direction parallel to RD (RD//<110>, and a γ-fibre texture comprising orientations of the {111} planeparallel to the RD. Normally, subsequent annealing of cold-rolled sheets increases the γ-fibre component at the expenseof the α-fibre, with possible improvements in the formabilityof the sheet metal (Yazawa et al., 2003). Figure 5 shows theorientation distribution function (ODF) maps of the foursamples in this work alongside a Bunge notation texturediagram showing the main texture fibres in bcc taken atΦ=45°. The F1 and F2 IR sample textures are predominantlycharacterized by a strong γ-fibre, notably {111}<011>,{111}<123>, and {111}<112>, as well as {554}<225>. The

▲


Table III

R-mean, Δr, and measured grain sizes

Sample RHF Dis. temp (°C) Rolling route Annealing after Steckel Rm delta-r Grain size (μm)

F1 1079 IR No 1.55 0.53 24.0 ± 3.7F2 1083 IR Yes 1.55 0.51 27.6 ± 3.8F3 1079 DR No 1.65 -1.38 24.2 ± 2.1F4 1160 DR No 2.07 -1.39 22.0 ± 2.5

Figure 4 – SEM micrographs of deep-drawn AISI 436 sheets after the second stage of the wheel-nut cover forming process. D1: smoother surface of IRsample, and D2: severe surface roughness of DR sample

striking difference between the two samples is observed inthe intensities of these γ-fibres, with the F1 sample showinga stronger intensity of a maximum of about 11 compared to amaximum of about 5 for the F2 sample. This observationclearly illustrates the insignificance of the AP1 annealingprocess, which is clearly an additional industry cost.

The F3 DR sample, however, is characterized by a strongintensity near or between {322}<258> and {322}<236>.Thus, it is evident that a strong texture component in the{100} plane would always have an adverse influence on the

annealing texture (Maruma et al., 2013). This is possiblycompounded by the absence of subsequent ‘intermediateannealing’ in the DR process, which hindered an increase ofthe γ-fibre evolution. Similarly, the F4 DR sample shows aweak α-fibre near {322}<258> as well as a weak γ-fibre inthe vicinity of {554}<225>, despite it having the highestLankford parameter value (Rm = 2.1). The high negativeearing parameter (Δr), which has already been alluded to, issuspected of adversely affecting the formability of thesesheets (F3 and F4), despite their good ductility parameters.



Figure 5 – Orientation distribution function (ODF) maps for the four samples studied. The adjacent diagram shows the texture fibre positions in Euler space(Raabe and Lucke, 1992)


Effect of thermomechanical processing on grain sizedistributionThe grain size clustering effect mentioned above is clearlyillustrated by the inverse pole figure (IPF) maps in Figure 6,particularly on the map for sample D2 (the DR sample), inwhich there is evident clustering of smaller grains separatelyin a particular region, almost forming a band of smallergrains, and clustering of larger grains in other regions withinthe microstructures. However, despite this grain sizeclustering, which may be responsible for differences inyielding across the sheet metal plates and hence the poorsurface roughness, no clear texture banding along aparticular direction was observed in the microstructures orthe SEM images, as is normally the case in ridging androping. This leads to the conclusion that another localizedgrain-related mechanism, rather than ridging, is at play inthis case.

Conclusions➤ Grain size measurements suggested that the processing

route had no significant influence on average grainsize, but had an effect on grain size distribution

➤ The intermediate rolling (IR) route resulted in thedesired γ-fibre texture, and hence superior surfacequalities (surface roughness) in comparison to thedirect rolling (DR) route

➤ In addition, the IR route also resulted in lower planaranisotropy (average Δr = 0.52) compared to the DRroute (Δr = -1.4), suggesting superior deep-drawabilityas observed

➤ Grain-size clustering as opposed to texture banding issuspected as a possible factor responsible for surfaceroughening.

AcknowledgementsThe authors gratefully acknowledge the financial contributionprovided by the Advanced Metal Initiative (AMI) of theDepartment of Science and Technology (DST) through theFerrous Metals Development Network (FMDN), and ColumbusStainless (Middelburg, South Africa). Special thanks are alsodue to Mr Dave Smith and Mr Jaco Kruger (ColumbusStainless) for their immense contribution and input into thisstudy.

References

CHARLES, J., MITHIEUX, J.D., SANTACREU, P.O. and PEGUET, L. 2008. The ferritic

stainless steel family: the appropriate answer to nickel volatility. 6thEuropean Stainless Steel Conference, Helsinki, 10–13 June 2008.

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through-thickness macro and micro-texture gradients on ridging of 17%

Cr ferritic stainless steel sheet. Materials Technology - Stainless Steels,

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KNUTSEN, R.D. and WITTRIDGE, N.J. 2002. Modelling surface ridging in ferritic

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and annealing temperature on texture evolution of AISI 441 ferritic

stainless steel. Journal of the Southern African Institute of Mining andMetallurgy, vol. 113, no. 2. pp. 115–120.

RAABE, D. and LUCKE, K. 1992. Influence of particles on recrystallization

textures of ferritic stainless steels. Steel Research, vol. 63, no. 10.

pp. 457–462.

RAABE, D., SACHTLEBER, M., WEILAND, H., SCHEELE, G. and ZHAO, Z., 2003. Grain-

scale micromechanics of polycrystal surfaces during plastic straining. ActaMaterialia, vol. 51. pp. 1539–1560.

SELLO, M.P. and STUMPF, W.E. 2010. Laves phase embrittlement of the ferritic

stainless steel type AISI 441. Materials Science and Engineering Section A,

vol. 527. pp. 5194–5202.

SHIN, H.-J., AN, J.-K., PARK, S.H. and LEE, D.N. 2003. The effect of texture on

ridging of ferritic stainless steel. Acta Materialia, vol. 51. pp. 4693–4706.

SIQUEIRA, R.P., SANDIM, H.R.Z., OLIVEIRA, T.R. and RAABE, D. 2011. Composition

and orientation effects on the final recrystallization texture of coarse-

grained Nb-containing AISI 430 ferritic stainless steels. Materials Scienceand Engineering Section A, vol. 528, no. 9. pp. 3513–3519.

YAZAWA, Y., OZAKI, Y., KATO, Y. and OSAMU, F. 2003. Development of ferritic

stainless steel sheets with excellent deep drawability by {111} recrystal-

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vol. 24. pp. 483–488. ◆

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Figure 6 – Inverse pole figure maps //ND for the first stages of the deep-drawn samples, showing clear indications of grain size clustering

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The objective of the conference will be to provide aforum for the dissemination of information relatingto the latest mining methods and technologies

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Being the s ix th conference in the ser ies, the Diamonds—st i l l Spark l ing Conferencetargets the fu l l spectrum of the diamond pipel ine f rom explorat ion through to salesand market ing. The last conference was held in 2013 at Misty Hi l ls , Muldersdr i f t : the

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14–17 March 2016, Gaborone International Convention Centre

Workshop: 14 March 2016Conference: 15 –16 March 2016Technical Visit: 17 March 1026

23rdINTERNATIONAL SYMPOSIUM

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BACKGROUNDThe Southern African Institute of Mining and Metallurgy (SAIMM) will be hosting the 23rd InternationalSymposium on Mine Planning and Equipment Selection (MPES) for the first time in South Africa in 2015.The symposium has successfully been organized annually for the past 25 years with intermittent 1-yearrecess as necessary. In 2013 the symposium was held in Germany. It will be in recess in 2014. MPES haspreviously been held in Turkey, Greece, Canada, Kazakhstan, Australia, Czech Republic, Brazil, India, China,Ukraine, Poland and Italy. Other venues for future MPES are Czech Republic, Sweden, and Australia.

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SHELL PECTEN

201512–14 October 2015 — Slope Stability 2015:International Symposium on slope stability in open pitmining and civil engineeringIn association with the Surface Blasting School15–16 October 2015Cape Town Convention Centre, Cape TownContact: Raymond van der BergTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail: [email protected]: http://www.saimm.co.za

20 October 2015 — 13th Annual Southern African StudentColloquiumMintek, Randburg, JohannesburgContact: Yolanda RamokgadiTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail: [email protected]: http://www.saimm.co.za

21–22 October 2015 — Young Professionals 2015ConferenceMaking your own way in the minerals industryMintek, Randburg, JohannesburgContact: Camielah JardineTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail:[email protected]: http://www.saimm.co.za

28–30 October 2015 — AMI: Nuclear MaterialsDevelopment Network ConferenceNelson Mandela Metropolitan University, North CampusConference Centre, Port ElizabethContact: Raymond van der BergTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail: [email protected]: http://www.saimm.co.za

8–12 November 2015 — MPES 2015: Twenty ThirdInternational Symposium on Mine Planning & EquipmentSelection Sandton Convention Centre, Johannesburg, South AfricaContact: Raj SinghalE-mail: [email protected] or E-mail: [email protected]: http://www.saimm.co.za

201614–17 March 2016 — Diamonds still Sparkle 2016Conference Gaborone International Convention CentreContact: Yolanda RamokgadiTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail: [email protected]: http://www.saimm.co.za

13–14 April 2016 — Mine to Market Conference 2016South AfricaContact: Yolanda RamokgadiTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail: [email protected]: http://www.saimm.co.za

17–18 May 2016 — The SAMREC/SAMVAL CompanionVolume ConferenceJohannesburgContact: Raymond van der BergTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156E-mail: [email protected]: http://www.saimm.co.za

21–28 May 2016 — ALTA 2016Perth, Western AustraliaContact: Allison TaylorTel: +61 (0) 411 692 442E-mail: [email protected]: http://www.altamet.com.au

May 2016 — PASTE 2016 International Seminar on Pasteand Thickened TailingsKwa-Zulu Natal, South AfricaContact: Raymond van der BergTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156E-mail: [email protected]: http://www.saimm.co.za

9 –10 June 2016 — 1st International Conference on SolidsHandling and Processing A Mineral Processing PerspectiveSouth AfricaContact: Raymond van der BergTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156E-mail: [email protected]: http://www.saimm.co.za

1–3 August 2016 — Hydrometallurgy Conference 2016‘Sustainability and the Environment’in collaboration with MinProc and the Western Cape BranchCape TownContact: Yolanda RamokgadiTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail: [email protected]: http://www.saimm.co.za

16–19 August 2016 — The Tenth InternationalHeavy Minerals Conference ‘Expanding the horizon’Sun City, South AfricaContact: Camielah JardineTel: +27 11 834-1273/7, Fax: +27 11 838-5923/833-8156 E-mail: [email protected]: http://www.saimm.co.za

INTERNATIONAL ACTIVITIES

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Company AffiliatesThe following organizations have been admitted to the Institute as Company Affiliates

AECOM SA (Pty) Ltd

AEL Mining Services Limited

Air Liquide (PTY) Ltd

AMEC Mining and Metals

AMIRA International Africa (Pty) Ltd

ANDRITZ Delkor(Pty) Ltd

Anglo Operations Ltd

Anglo Platinum Management Services (Pty) Ltd

Anglogold Ashanti Ltd

Atlas Copco Holdings South Africa (Pty) Limited

Aurecon South Africa (Pty) Ltd

Aveng Moolmans (Pty) Ltd

Axis House (Pty) Ltd

Bafokeng Rasimone Platinum Mine

Barloworld Equipment -Mining

BASF Holdings SA (Pty) Ltd

Bateman Minerals and Metals (Pty) Ltd

BCL Limited

Becker Mining (Pty) Ltd

BedRock Mining Support (Pty) Ltd

Bell Equipment Company (Pty) Ltd

Blue Cube Systems (Pty) Ltd

Bluhm Burton Engineering (Pty) Ltd

Blyvooruitzicht Gold Mining Company Ltd

BSC Resources

CAE Mining (Pty) Limited

Caledonia Mining Corporation

CDM Group

CGG Services SA

Chamber of Mines

Concor Mining

Concor Technicrete

Council for Geoscience Library

CSIR-Natural Resources and theEnvironment

Department of Water Affairs and Forestry

Deutsche Securities (Pty) Ltd

Digby Wells and Associates

Downer EDI Mining

DRA Mineral Projects (Pty) Ltd

DTP Mining

Duraset

Elbroc Mining Products (Pty) Ltd

Engineering and Project Company Ltd

eThekwini Municipality

Exxaro Coal (Pty) Ltd

Exxaro Resources Limited

Fasken Martineau

FLSmidth Minerals (Pty) Ltd

Fluor Daniel SA (Pty) Ltd

Franki Africa (Pty) Ltd Johannesburg

Fraser Alexander Group

Glencore

Goba (Pty) Ltd

Hall Core Drilling (Pty) Ltd

Hatch (Pty) Ltd

Herrenknecht AG

HPE Hydro Power Equipment (Pty) Ltd

Impala Platinum Limited

IMS Engineering (Pty) Ltd

JENNMAR South Africa

Joy Global Inc. (Africa)

Leco Africa (Pty) Limited

Longyear South Africa (Pty) Ltd

Lonmin Plc

Ludowici Africa

Lull Storm Trading (PTY)Ltd T/A WekabaEngineering

Magnetech (Pty) Ltd

Magotteaux(PTY) LTD

MBE Minerals SA Pty Ltd

MCC Contracts (Pty) Ltd

MDM Technical Africa (Pty) Ltd

Metalock Industrial Services Africa (Pty)Ltd

Metorex Limited

Metso Minerals (South Africa) (Pty) Ltd

Minerals Operations Executive (Pty) Ltd

MineRP Holding (Pty) Ltd

Mintek

MIP Process Technologies

Modular Mining Systems Africa (Pty) Ltd

MSA Group (Pty) Ltd

Multotec (Pty) Ltd

Murray and Roberts Cementation

Nalco Africa (Pty) Ltd

Namakwa Sands (Pty) Ltd

New Concept Mining (Pty) Limited

Northam Platinum Ltd - Zondereinde

Osborn Engineered Products SA (Pty) Ltd

Outotec (RSA) (Proprietary) Limited

PANalytical (Pty) Ltd

Paterson and Cooke Consulting Engineers (Pty) Ltd

Polysius A Division Of ThyssenkruppIndustrial Solutions (Pty) Ltd

Precious Metals Refiners

Rand Refinery Limited

Redpath Mining (South Africa) (Pty) Ltd

Rosond (Pty) Ltd

Royal Bafokeng Platinum

Roymec Tecvhnologies (Pty) Ltd

Runge Pincock Minarco Limited

Rustenburg Platinum Mines Limited

SAIEG

Salene Mining (Pty) Ltd

Sandvik Mining and Construction Delmas(Pty) Ltd

Sandvik Mining and Construction RSA(Pty) Ltd

SANIRE

Sasol Mining(Pty) Ltd

Scanmin Africa (Pty) Ltd

Sebilo Resources (Pty) Ltd

SENET

Senmin International (Pty) Ltd

Shaft Sinkers (Pty) Limited

Sibanye Gold (Pty) Ltd

Smec SA

SMS Siemag South Africa (Pty) Ltd

SNC Lavalin (Pty) Ltd

Sound Mining Solutions (Pty) Ltd

South 32

SRK Consulting SA (Pty) Ltd

Technology Innovation Agency

Time Mining and Processing (Pty) Ltd

Tomra Sorting Solutions Mining (Pty) Ltd

Ukwazi Mining Solutions (Pty) Ltd

Umgeni Water

VBKOM Consulting Engineers

Webber Wentzel

Weir Minerals Africa

WorleyParsons (Pty) Ltd

2015� SYMPOSIUM

International Symposium on slope stability in open pit mining and civil engineering12–14– October 2015In association with the Surface Blasting School15–16 October 2015, Cape Town Convention Centre, Cape Town

� COLLOQUIUM13th Annual Southern African Student Colloquim 201520 October 2015, Mintek, Randburg, Johannesburg

� CONFERENCEYoung Professionals 2015 Conference21–22 October 2015, Mintek, Randburg, Johannesburg

� CONFERENCEAMI: Nuclear Materials Development Network Conference28–30 October 2015, Nelson Mandela Metropolitan University, North Campus Conference Centre, Port Elizabeth

� SYMPOSIUMMPES 2015: Twenty Third International Symposium on MinePlanning & Equipment Selection8–12 November 2015, Sandton Convention Centre, Johannesburg, South Africa

2016� CONFERENCE

Diamonds still Sparkle 2016 Conference 14–17 March 2016, Gaborone International Convention Centre

� CONFERENCEMine to Market Conference 201613–14 April 2016, South Africa

� CONFERENCEThe SAMREC/SAMVAL Companion Volume Conference17–18 May 2016, Johannesburg

� SEMINARPASTE 2016 International Seminar on Paste and Thickened TailingsMay 2016, Kwa-Zulu Natal, South Africa

� CONFERENCE1st International Conference on Solids Handling and ProcessingA Mineral Processing Perspective9 –10 June 2016, South Africa

� CONFERENCEHydrometallurgy Conference 20161–3 August 2016, Cape Town

� CONFERENCEThe Tenth InternationalHeavy Minerals Conference ‘Expanding the horizon’16–19 August 2016, Sun City, South Africa

SAIMM DIARY

For further information contact:Conferencing, SAIMM

P O Box 61127, Marshalltown 2107Tel: (011) 834-1273/7

Fax: (011) 833-8156 or (011) 838-5923E-mail: [email protected]

For the past 120 years, theSouthern African Institute ofMining and Metallurgy, has

promoted technical excellence inthe minerals industry. We striveto continuously stay at the cuttingedge of new developments in themining and metallurgy industry.The SAIMM acts as thecorporate voice for the miningand metallurgy industry in theSouth African economy. Weactively encourage contact andnetworking between membersand the strengthening of ties.The SAIMM offers a variety ofconferences that are designed tobring you technical knowledgeand information of interest for thegood of the industry. Here is aglimpse of the events we havelined up for 2015. Visit ourwebsite for more information.

Website: http://www.saimm.co.za

EXHIBITS/SPONSORSHIP

Companies wishing to sponsor

and/or exhibit at any of these

events should contact the

conference co-ordinator

as soon as possible