Tribology in nuclear power generation

Tribology in nuclear power generation W.H. Roberts*

A prerequisite for the efficient production of electricity from nuclear power reactors is a high degree of reliability and predictability of performance in engineering components and plant. This can only be achieved if engineering designs are based on sound tribological principles and incorporate the best available data from well planned back-up research and development programmes. The approaches being adopted to tackle the specifically nuclear orientated (ie non-conventional) engineering problems in nuclear power generation are discussed and examples given of successful solutions at all of the various stages of the nuclear fuel cycle ranging from uranium enrichment to power station operation, spent fuel reprocessing and, finally, disposal of waste products. The paper is not intended to be a comprehensive review and does not cover all the areas in great detail

Tribologists have had, and will continue to have, an important role to play in all aspects of nuclear power generation, ranging from the enrichment of uranium fuel to the operation of the reactor plant, through to the reprocessing of spent fuel and the disposal of highly active fission product wastes. In the development programme of reactor types from Magnox through Advanced Gas-Cooled Reactors (AGRs) to the Prototype Fast Reactor (PFR), fuel power ratings and coolant temperatures have both significantly increased and for the fast reactors coolant technology has switched from gas to liquid metal (sodium), all of which have set difficult and challenging tribological problems for the engineer and materials scientist. Good tribological design is concerned with predictable and reliable performance of components, based on established friction and wear parameters, achieved with the minimum possible maintenance. Tribology requirements for nuclear power systems differ only in degree from conventional engineering plant in that the achievement of the (usually stringent) specification requirements for reliability and predictability of performance are paramount insofar as they impinge on the safe working of the plant; and the high capital cost of nuclear plant puts a high premium on a high availability factor. It is therefore a feature of the nuclear power industry that engineering development is backed by extensive R and D facilities; and that attention is given to assessing and predicting the reliability of mechanisms"2. There is a large measure of co-ordination of the basic tribology and associated corrosion and materials compatibility programmes in the UK, as between the AEA, the Electricity Generating Boards and the Nuclear Power Company (NPC). The basic philosophy underlying tribology programmes is to try and understand the mechanisms of various wear and contact processes 3 (so as to establish the relevance of small scale experimental data to large scale plant problems); *National Centre of Tribology. UKAEA. Risley, Warrington WA3 6A T, UK

0301-679X/81/010017-12 $02.00 © 1981 IPC Business Press

and the appropriate modelling parameters so that data can be applied to life or safety assessments. An operational life free of maintenance in the context of a nuclear power reactor could mean, typically, 40 000 to 200 000 h. On nuclear engineering development, the UK approach is to test complete components only where this is found to be the most effective means for deriving the information 4 or where, as in the development of the large mechanical circulating pumps for the PFR, operational experience with the prototype was particularly desirable s . Otherwise, specific mechanical features of complex mechanisms are individually studied. Thus in the PFR fuel handling machine, ball nut and screw, ball bearings, spline and couplings were all separately examined 6 , and resulted in a highly successful piece of hardware which has given trouble-free service. The problems peculiar to nuclear power reactors arise because of the high temperatures and the hostile (reactor coolant) environments in which mechanisms must operate. Added to this, there is the presence of nuclear radiation which inhibits maintenance and generally precludes the use of conventional mineral oil-based lubricants, although some special radiation-resistant lubricants have been developed capable of operating to moderately high temperatures ( ~ 200°C). Even so, experiments have shown that low concentrations of moisture (at the I00 vpm level) can significantly reduce the fatigue life of greased ball bearings in pressurised CO: (Magnox reactor) environments 7 . Operating conditions for nuclear reactor mechanisms are such, especially the wide variations in temperature, that changes in unlubricated wear rates by 2, or even 3, orders of magnitude are not unusual; and for friction coefficients to approach unity, and even higher, is the rule rather than the exception (a factor which must be taken into account when designing mechanisms, such as fuel element grabs, for example, to ensure that they do not become self-locking for high breakaway friction). The failure mode of rolling element bearings in the absence of conventional lubrication, even though the load capacity is drastically de-rated, is

TRIBOLOGY international February 1981

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Roberts - Tribology in nuclear p o w e r generation

generally by wear 6'~-12 (although very hard cermet bearings operating in sodium have been observed to fail by fatigue 13 ).

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The fact that wear is the predominant mode of failure of components under unlubricated reactor coolant environments means that the concept of specific wear rate (the volume of material removed per unit sliding distance per unit load) should in principle give a reasonably accurate prediction of the behaviour of plant components over their lifetimes. The reliability of such predictions does, of course, depend upon the quality of the data obtained from laboratory short-term/accelerated tests. There are a wide variety of rigs available but these will give meaningful results only if careful attention is paid to simulating as closely as possible the real-life situation in terms of all salient parameters such as temperature, load, speed, etc. and also, most importantly, the composition of the environment since all reactor coolants are chemically reactive to a greater or lesser degree and can give rise to the formation of complex compounds, some of which can be beneficial, between interacting surfaces. The experimental data produced by such rigs must always be regarded as having relative, as opposed to absolute significance ]4, and the only reliable approach adopted is to use the information thus obtained to narrow the field of contending materials and thereafter to carry out proving trials on a prototype system. The UK nuclear industry, broadly following this philosophy, has a record of well

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A feature of unlubricated sliding in gaseous, water and sodium environments is the wide variation of specific wear rates, both as a function of temperature in a given environment (Fig 1) and as between different environments for the same materials combination (Fig 2). Nitrided steels, for example, found application in many dry sliding applications in Magnox reactors based on the low wear rates they exhibited in C02 over the temperature range 200 ° 400°C (Fig 3). Even for self-lubricating graphite compositions, which have been widely studied as promising bearing materials ~8'~'6-2a, similar large variations with temperature (Fig 4) and environment (Fig 5) are observed. Molybdenum disulphide (MoS:) has been successfully used in both helium 24 and CO: environments :9, as also has tungsten disulphide where higher temperatures ( ~ 400°C) are involved, as, for example, in Magnox reactor refuelling machines 3° (which have given generally reliable service 3a ).

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The selection of materials for tribological applications, while not unimportant in conventional lubrication practice, assumes predominant significance in systems where conventional lubricants cannot be used. It is therefore not surprising that attention is being paid to the collation of the vast amount of friction wear and self-welding data which exists in order to enable the nuclear design/development engineer to make rational decisions on the information available to him. Figs 6 and 7 illustrate the manner in which typical wear and friction results for Stellite alloys, arising from the UK's supporting programme on fast reactors, are presented (and also, incidentally, give an indication of the reproducibility of wear data for a given combination of engineering alloys). The occurrence of high friction coefficients will also be noted. The Americans, likewise, are producing a Nuclear Systems Materials Handbook which has a tribology section (from which the example in Fig 8 has been taken) a3 .

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There is still much tribology activity in support of CO2, and high temperature helium, cooled reactor systems (the latter more especially in the USA ~ , W. Germany 3s and Japan ~'37, but also in Brazi138). However, the main emphasis of the R and D effort appears to be orientating towards sodium-cooled fast reactors 4-6'1°-14'22'23'39-62 . A great deal of work in this area has been carried out over the last five years, or so, in several American laboratories as part of a co-ordinated National Friction, Wear

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Roberts - Tribology in nuclear power generation

and Self-Welding Program under the aegis of the United States Energy Research and Development Administration (USERDA). Much of this study has been directed at evaluating the performance of coatings and hardfacings, applied by welding and by plasma and detonation

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The wide span of friction coefficients observed in high temperature liquid sodium has already been commented upon. The inter-relation of oxygen level and sodium temperature in controlling the effective boundary between high and low friction regimes has been reported and is the subject of current study at Risley 64. Fig 10 shows the results of sliding experiments with two chromium-containing alloys over the temperature range 5 0 0 ° - 6 5 0 ° C . The effective tribological boundary, which differed slightly for the two combinations, showed a very similar trend to the experimentally observed boundary based on Harwell SEM and X-ray examination of AISI 316; both sets of curves fell at significantly higher oxygen levels than the thermodynamically predicted boundary. This trend is not altogether surprising since for good tribological performance more than the minimum amount of sodium chromite will need to be continuously regenerated. An important implication of such studies

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Fretting Low amplitude wear phenomena covering both fretting wear, fretting corrosion and fretting fatigue have been widely investigated in all reactor coolant envirionments of interest ss-sS'~-ra. Of particular interest in the context of heat exchangers is the performance of ferritic steels s9-62 which in high temperature sodium give considerable mutual transfer and surface damage with comparatively little production of wear debris. Little operational experience is available of the performance of tubes in liquid sodium heat exchangers in reactors or large size loops but it is clearly important that excessive wear rates must be avoided where thin-walled

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TRIBOLOGY international February 1981

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Roberts - Tribology in nuclear power generation lubricant film, probably sodium aluminate (NiAI02), formed by reaction of the nickel aluminide layer with oxygen-containing sodium. The aluminate is thermodynamically more stable than sodium and chromium oxides, and also sodium chromite. Attempts to observe aluminate f'flms directly have, so far anyway, been unsuccessful because the (very thin) films are lost during the sodium decontamination process. Tests in high temperature sodium are currently underway to check on long-term corrosion rates of aluminised nickel-based alloys; data so far obtained (in the USA) covering 8000 h in 625°C sodium are extremely promising.

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is part). The philosophy of identifying potential tribological problems in component parts of a complex reactor mechanism has already been referred to. A good example of the successful execution of this approach was the fuel handling mechanism of the PFR charge machine. This had several ball bearing pivot points, a spline, two unusual joints, gripper mechanism, thrust race and a ball nut and screw mechanism (Fig 14). All of these features were individually tested and led to the production of a prototype mechanism whose performance has been faultless.

Process fluid lubrication With much of the emphasis in the support R and D pro-

Fig 12 As for Fig 11 after a total o f lO0 traverses, showing a wider wear track with smearing and flaking o f the chromite film 6s (x 400)

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TRIBOLOGY international February 1981

grammes for nuclear power reactor mechanisms centred on the intrinsic non-lubricated friction and wear properties of materials, it must nevertheless not be overlooked that in certain special circumstances it is possible to use the process fluid being handled as the lubricant; gas, water and molten sodium (reactor coolants) have all been successfully employed as bearing lubricants. In each of the primary pumps in the PFR, for example (Fig 15), the shaft bottom bearing, which is below the sodium surface, is sodium lubricated, and a hydrostatic/jetted bearing fed from the pump delivery was chosen for this application, primarily because it permitted a larger radial clearance than the

Roberts - Tribology in nuclear p o w e r generation

alternative hydrodynamic type of bearing vs. The theory of hydrostatic bearings under the operating conditions of interest (including both laminar and turbulent flow regimes in the bearing clearance) was developed and verified by experiments 76, using water and paraffin as the working fluid with a 200 mm diameter x 200 mm long bearing (Fig 16). A prototype 6000 gal/min (1620 ma/h) pump was tested in sodium at Risley and the confidence in the bearing was such that it was not considered necessary to test the full-scale prototype pumps (18 000 gal/min; 500 mm diameter x 500 mm bearings) in sodium (Fig 17). There are, of course, no difficulties in principle in the design of bearings to operate with non-conventional lubricants. The choice between self-acting (hydrodynamic) and externally pressurised configurations will depend on the particular performance specification. Low viscosity liquids like sodium (and water) can, however, present problems; for a self-acting configuration the bearing load capacity will be limited and even to achieve this result it may be necessary to go to extremely fine dimensional (bearing clearance) tolerances. And for externally pressurised bearings turbulent/superlaminar flow conditions will almost certainly predominate 77'7s.

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The earliest successful industrial-scale application of gas bearings in the UK was on the gaseous diffusion plant for uranium enrichment at Capenhurst. Such plants contain several thousand gas circulators which handle U F 6 gas which is both radioactive and corrosive. There was therefore a great incentive to try to utilise this gas as the bearing lubricant or to use a gas-bearing seal and allow a small amount of an inert gas to leak across the seal into the plant. In the several thousand compressors operated over many years at Capenhurst it has become clear that gas-bearings and gas-bearing seals are inherently more reliable than conventional bearings and are virtually wear-free. The significant factor for obtaining the undoubted benefit from self-acting gas lubricated bearings lies in continuous, uninterrupted operation. They were successfully employed in gas circulators over several years in prototype reactors at Harweli, Dounreay and Windscale. However, where there is frequent stopping and starting (in the absence of auxiliary pressurised gas supplies) metallic bearing surfaces in self-acting bearings become irrevocably damaged and the bearing eventually unserviceable. Such, it must be recorded, was the general experience in the 1950s when an attempt was made to introduce gas-lubricated bearings into circulators in gas-cooled reactors (although Sultzer designs in Switzerland were more successful) with the result that this type of bearing has not found wide application in the power generating industry. Work at NCT has led to the manufacture of gas bearings in thermosetting plastic which are capable of sustaining many thousand stop/start operations without any surface damage and deterioration of bearing performance (but they do have a temperature limitation of about 125°C and limited radiation resistance).

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Roberts - T r i b o l o g y in nuclear p o w e r generation

The extension of uranium enrichment facilities in the UK in recent years has been by the recently developed gas centrifuge process. And whereas the diffusion process depends on the differential passage of uranium hexafluoride gas molecules through a membrane, in the gas centrifuge process separation is achieved as the heavier uranium-238 hexafluoride molecules tend to spin towards the outside wall, allowing the lighter uranium-235 hexafluoride molecules to be scooped up and collected. Commercial operation of this process requires a very large number of machines tens of thousands for a sensible output. And the reliability of the lower pivot bearing which supports the centrifuge bowl rotating at very high speeds is a critical tribological feature (Fig 18). Fuel reprocessing

A computer literature search failed to throw up any significant references in the fields of tribology in reprocessing (or in waste-product disposal). However, since reliability and durability of plant under active (and hence, remote) operational and maintenance conditions is paramount and this inevitably includes components and equipment incorporating moving parts, one can only conclude that the dearth of published tribology information reflects the commercial confidentiality of such work.

Fuel element cutting is one of the main technological operations in spent nuclear fuel reprocessing. Stainless steel fuel cans containing irradiated ceramic (uranium oxide) fuel pose a severe abrasive wear problem. Again designs must cater for the need for remote handling at all times. Clearly there is a very high premit, m on maintaining blade cutting edges since a change of blade is a time-consuming and expensive operation. Diamond impregnated blades have been successfully used to slit stainless steel cladding of irradiated fuel elements ~3, as also, incidentally, have lase,s. Development of a prototype double-roll fuel particle crusher to fracture the silicon cabide coatings on high-temperature gas-cooled reactor (HTGR) fuel particles included, as well as the crushing roll materials, drives, gearing, bearings and seals consideration ~ . The hostile, radioactive, environment in-cave generally favours the adoption of dry lubricant solutions to tribology problems. Not only do conventional lubricants deteriorate (due to irradiation) but there is also a tendency for radioactive particles to become attached to wet surfaces which then present a decontamination problem.

The reprocessing of spent fast reactor fuel involves solution in acid to produce a highly radioactive liquor, followed by separation of the residual solids by centrifuging. The design of a suitable chemical plant centrifuge to work reliably in such a hostile environment gave rise to many interesting tribological problems 82. Apart from achieving stability of operation over the full speed range, up to, typically 36 000 rev/min, detailed consideration had to be given to overcoming the problems associated with remote handling in "cave" facilities.

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TRtBOLOGY international February 1981

Fig 17 Prototype mechanical sodium pump

Roberts - Tribology in nuclear p o w e r generation

Problems exist with pneumatic rotary or linear drives. Conventionally, oil particles suspended in the air feed are used to lubricate moving parts. The selection of an appropriate (dry) low friction and wear material for the rubbing components eliminates the widespread deposition of oil from the exhaust air and the associated fire hazard. Air bearings are useful devices when employed in the active handling environment, providing no contamination and serving to "'blow" abrasive debris away from bearing surfaces. A typical example is the support of lead spheres through which handling tongs pass in lead brick walls. Characteristically they provide low friction support to aid sensitivity of manipulation and additionally the exhaust air forms a barrier to migration of active particles through the joints.

Waste products disposal A process for the vitrification of active fission product waste, the HARVEST project, is being developed with BNFL towards an industrial scale. An interesting tribological aspect of this work, which is currently underway at Risley, is the development of pumps for handling the radioactive, corrosive and abrasive slurries used in the glass making process. Where flow control is not critical, the transport of slurries around a vitrification plant can generally, be effected by simple gravity feed, air lift and suction lift systems. With such systems the tribological problems are confined to abrasive/corrosive wear of nozzles and pipework, etc. When accurate flow control is required the problem areas extend to valves, mechanical pumps and measuring devices. Solutions are then required for lubrication of drive mechanisms, wear of seats, seals and bearings, and fatigue of flexing elements. The selection of materials and finesse of designs become important if components are to withstand the duty and be capable of servicing in the hostile conditions of corrosion, abrasion and irradiation over long periods of time.

behaviour of surface coatings and treatments which enhance the tribological performance of structural materials. Data is also required which can be applied to high temperature operation over extended periods of many years which means parallel studies of surface chemistry and corrosion aspects if we are to be able to extrapolate experimental data to anticipated reactor component operational times which could be as long as 40 000 to 200 000 h.

Acknowledgements The author would like to thank Dr T.N. Marsham, Managing Director of the Northern Division, UKAEA, for permission to publish this paper, lte is indebted to many colleagues in the UKAEA, CEGB, BNFL and GEC(REL) for their help in the preparation of the paper. This paper was presented at the "Patterns of Tribology" conference held at Paisley, UK, 1 0 13 September, 1979.

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Consideration must also be given to the ultimate disposal of all the items used in the irradiation environment because they in their turn become active waste. Simplicity, ease of decontamination, mobility and low volume are essential features. Radioactive materials converted to mobile slurries can ease the problems of transportation, but increase the bulk for storage of ultimate disposal. "De-watering" by means of hydro-cyclones, centrifuges and self-clean filters are further areas for development to increase their reliability and durability. Since maintenance will be extremely difficult, the specification calls for reliable operation of components over long periods.

Conclusion An attempt has been made to indicate some of the fields of tribological study in relation to nuclear power generation, ranging from fuel enrichment and reactor plant operation through to fuel reprocessing and disposal of waste products. In all these areas, interesting and challenging tribology problems are yet to be solved. There is still scope, in all the reactor power systems, gas, sodium and water, for improved tribological solutions, some of them having considerable economic implications. We need a better understanding of fretting and impact wear phenomena. More comprehensive basic friction, wear and self-welding data are required to establish the

b'ig 18 Diagram showing uranium hexafluoride gas flow withh7 an enrichment centrifuge (BNFL)

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Roberts -- Tribology in nuclear power generation

References 1.

2.

3.

Mitchell L.A. and Osgood C. Prediction of the Reliability of Mechanisms from Friction Measurements. Paper G265/73,

23. Roberts W.H. Friction and Wear Behaviour of Sliding Bearing Materials in Sodium Environments at Temperatures up to 600°C. UKAEA TRG Ret~ort 1269(R). 1966

l.Mech.E. Proceedings, 1st European Tribology Conference, London, 19 73

24. Frieker H.W. Bearings and Gears for Operation in Inert Gases.

Syrup. on Reliability of Nuclear Power Plants, Innsbruck, April 1975

25. Heath H.H. Bearings for Nuclear Engineering. 1. Mech. E.

Mitchell L.A. Applications for Contact Theories in Nuclear Reactor Technology. CEGB/BNL Report RD/B/N3177, 1974

26. Manjoine MJ. Friction Characteristics of Graphite and Graphite-Metal Combinations at Room and Elevated Temperatures. Bearing and Seal Design hz Nuclear Power

4.

Delves P.H. and Rodwell W. Mechanisms in Sodium.

5.

IAEA/SM-130/14, Proc. Syrup. on Progress in Sodium-Cooled Past Reactor Engineering, Monaco, March 1970 Eickhoff K.G., Allen J. and Boorrnan Engineering Development for Sodium Systems in Fast Breeder Reactors. Proc. British Nuclear Energy Society Conference on Fast Breeder Reactors, London, 1966, Pergamon Press, 1967, 873-895

6.

Proceedings 1. Mech. E. Lubrication and Wear Com,ention, Bournemouth, May 1963

Mitchell L.A., Osgood C. and Radcliffe S.J. Reliability Studies in a Developing Technology. Paper IAEA/SM-195/43, Proc.

Delves P.H. and Hill G.D. Development Tests in Sodium on Two Types of Gripper Drive. Proc. Int. Conf. on Liquid Metal

Technology in Energy Production, Seven Springs, 1976. (National Technology Information .Service (USA)), CONF- 76053, Part 1

Proceedings, 196 7- 68, 182(3/1), 449- 459

Machinery {ed R.A. Burton), ASME, 1968 (Proc. Syrup. on Lubrication in Nuclear Applications, Miami Beach, June 1967; 27.

Lipp L.C. and Stern it. Eftect of Radiation and Additives on Graphite Lubricants for High Temperature Applications.

ibMem 28. Stansfield O.M. ['riction and Wear of Graphite in Dry Helium at 25,400 and 800 ° C. Nuclear Applications, April 1969, 6,

313-.320

London, 1962, 176(2), 27-59

29. Parry A.A. and Mitchell L.A. Research on Solid Lubricants for use within Gas-Cooled Power Reactors. Proc. 1. Mech. E., 1967-68, 182, Part 1, No 20, 489..-494

8.

Radcliffe S.J. The l.atigue Performance of Grease Lubricated Ball Bearings in High Pressure Carbon Dioxide. Proc. 1. Mech. E., Tribology Convention, Durham, 1976, 37-40

30. Mitchell L.A. and Woodail D. Failure of Helically Wound Cables in High Friction Environments. 1. Mech. E. Proceedings, 1972, 186, Part 43/72, 491-498

9.

Eickhoff K.G. and White A. The Performance of Ball Bearings in Nitrogen and Carbon Dioxide at Elevated Temperatures.

31. Proc. Symposium on Refuelling of Gas-Cooled Reactors.

7.

Vaile P.E.B. Lubricants for Nuclear Reactors. Proc. I. Mech. E.,

LMech. E., London, December 1968

ASLE Trans, 1961, 4, 39-49 10. lwal S., lshii M., Nozawa M., Tamwa T., Abe Y., Teramoto A. and Akiyama M. Testing of Feasibility of Ball Bearings in Liquid Sodium Environments. ANL- 7520 (Pt 2), 1968, 121-129 11. Kleefeldt K. On the Utilisation of Roiling Bearings in Sodium Engineering. KFK.Nachr, 1975, 7(2), 5 - 1 3

32. Campbell C.S. NCT l:riction and Wear Data for Stainless Steels, Stellites and Tool Steel in Liquid Sodium. Risle.v

Nuclear Laboratories, Internal Report, 1978 33.

Proc. Int. Conf. on Liquid Metal Technology in Energy Pro duction, Seven Springs, 1976. (National Technology In.tormation Service (USA)), CONF-76053, Part I 13. Kleefeldt K.W. Experimental Investigations of the Friction and Wear Behaviour of Tungsten Carbide Ball Bearings Under Axial Load in Liquid Sodium. ibidem 14. Wild E. and Mack K J . Friction and Wear in Liquid-Metal Systems; Comparability Problems of Test Results Obtained from Different Test Facilities. ibidem 15. Cornelius D.F. and Roberts W.H. l:riction and Wear of Metals in Gases up to 600°C. ASLE Trans, 1961, 4, 20--32 16. Heath H.H. and Phillips K.F. An Investigation into Sliding Bearings Suitable for a COa -cooled Nuclear Reactor. Wear,

1960, 3(5), 3 5 8 3 7 3 17. Clark W.T., Pritchard C. and Midgley J.W. Mild Wear of Unlubricated Hard Steels in Air and Carbon Dioxide. 1. Mech. E. Proceedings, 1967-68, Vol 182. Part 3N, 97-106 18. Cornelius D.F., Garnham A . L and Green S.M. Graphite Wear Behaviour in Argon, Carbon Dioxide, Steam and Water Environmerits. L Mech. E. Proceedings, 1964-65, Vol 179, Part 3T,

34. Lai (;.Y. Evaluation of Sprayed Chromium Carbide Coatings for Gas-Cooled Reactor Applications. Thin Solid Films,

1978, 53(3), 343-351 35.

36. Nemoto M. and Okamoto Y. A Survey on Gas Bearing Circulators of High-Temperature Gas Cooled Reactors. Report

,IAERI-M- 7214, 1977 37. Nomura S. Recent Trends and Problems of Development for tligh Temperature Reactor Materials. Nippon Kinzcku Gakkai

Kaiho, 1977, 16(11), 779- 88 38. Shukla V., Padilha A.F. and Watanabe S. Wear of Materials for Turbine and Cladding in the HTGR and GCFR. Cienc. Cult. (Sao

Paulo) 1975, 27(7), 509, (in Portuguese) 39.

Metal Technology in Energy Production, Seven Springs, 1976. (National Technology Information Service/USA/), CONF- 760503, Part 1, 122 130

tligh-Temperature, High-Pressure Water Environments.

L Mech. b.: Proceedings, 1965 -66, Vol 180, Part 3K, 37-48

lIostile Environments, 1968- 69, Vol 183, Part 31.28 34 22.

26

Roberts W.H. The l:riction and Wear Behaviour of Molybdenum-Tungsten-Chromium Alloys in High Temperature Sodium Environments. ASLb." Trans, 1965, 8. 109-122

TRIBOLOGY international February 1981

Johnson R.N., Aungst R.C., Hoffman N.J., Cowgiil M.G., Whitlow G.A. and Wilson W.L. Development of Low Friction Materials for LMFBR Components. Proc. Int. Conf. on Liquid

19. Roberts W.H. Measurement of Sliding Friction and Wear in

21. Phillips K. and Pritchard C. Adhesion of Metals in Hot Oxidising Gases. 1. Mech. E. Proc. Syrup. on I.ubrication in

Engel R. et al Studies on Ceramic Coatings - a Possible Means to Prevent Friction and Diffusion Welding of Metallic Materials in Itelium-Cooled Reactors. Proc. Reactor Congress, Hannover,

ZAED 1978 (in German)

123-130

20. Rowe M.D. The Static Adhesion of En58B Stainless Steel in Carhon Dioxide Between 4500( ` and 750°C. CE(;B/BNI. Report RD/B/N2321, 1972

Nuclear Systems Materials Handbook, USAEC USDOE Report

No. TID-26666, 1975

12. Campbell C.S. Wear of Rolling Element Bearings in Sodium.

40. Whitlow G.A., Wilson W.L., Galioto T.A., Miller R.L, Schrock S.L., Hoffman N.J., Droher J.J. and Johnson R.N. Corrosion and Tribological Investigation of Chromium Carbide Coatings for Sodium-Cooled Reactor Applications. ibidem, 138-144 41.

Hoffman N.J., IMoher JJ., Chang J.Y., Galioto T.A., Miller R.L., Schrock S.L., Whitlaw G.A., Wilson W.L. and Johnson R.N. Evaluation of Cobalt and Nickel Base Materials for Sliding and Static Contact Applications in a Liquid Metal Fast Breeder Reactor. ibidem. 16 7-1 76

Roberts -- Tribology in nuclear power generation

42. IOmoh S., Mizobuchi S. and Atsumo H. Friction and Wear Behaviour of Colmonoy and Stellite Alloys in Sodium Environments. ibidem, 1 5 3 - 9 43.

Huber F. and Mattes K. Investigation into the Self-Welding Behaviour of Metallic Materials Exposed to Sodium. ibidem

44. Chang J.Y., Flagella P.N. and Schrock S.I.. Self-Welding Evaluation of Type 304 and A286 Stainless Steel in the Temperature Range 800 ° - 1140°F in Flowing Sodium. Microstructural Science, 19 77, 5, 9 9 - 1 1 3 45. Fedorchenko I.M., Shevchuk J.F., Miroshnikov V.N., Tuznikov A.F., Markov V.G., Gorynin I.V., Krasnoshchekov M.M. and Mostovskij D.M. Study of Some Properties of Ironbase Materials with Calcium Fluoride Additions for Friction Nodes in Nuclear Plant. Poroshk. Metall, 19 77 {Engligh trans in Sot,. Powder Metall. Met. Ceram, No 7, 6 4 - 6 9 ) 46. Wilson W.L., Galioto T.A., Miller R.L. and Whitlow G.A. Tribological and Corrosion Behaviour of Materials for High Temperature Sodium Service. Proc. Intl. Metall. Sot., Ann. Mtg., Montreal, July 1979, 265--274 47. Johnson R.N. and Farwick D.G. Friction, Wear and Corrosion of Laves-hardened Nickel Alloy Hardsurfacing in Sodium. Thin Solid bThns, 1978, 53{3), 3 6 5 - 3 7 3 48.

Johnson R.N., Schrock S.L. and Whitlow G.A. Wear Resistant Coatings for Reactor Components in Liquid Sodium Environments:./. Vac. Sci. Technol, 1974, !1(4), 759--764

49. Wolfla T.A. and Johnson R.N. Refractory Metal Carbide Coatings for LMFBR Applications - A Systems Approach. J. Vac. Sci. Technology, 197.5, 12(4), 7 7 7 783 50. Kissel J.W., Glaeser W.A. and Allen C.M. Sliding Contact Frictional Behaviour in Sodium Environments. A S L E Trans, 1962, 5(1), 39--44 51.

Kissel J.W., Glaeser W.A. and Allen C.M. Friction Behaviour of Sodium-Lubricated Materials in a Controlled High-temperature Environment. Wear, 1962, 5(6), 4 4 6 - 4 5 7

52. Baumgartner A.J. Mechanical Element Design for tligh Temperature Sodium Environment, (also bibliography), Bearing and Seal Design in Nuclear Power Machinery (ed R.A. Burton) - ASME, 1968 (Proc. Syrup. on Lubrication ht Nuclear Applications, Miami Beach, June 196 7) 53. Wild E. and Mack K.J. Tribology in the Core of a Sodiumcooled Fast Breeder Reactor. Wear, 1975, 34(3}, 3 3 1 - 3 4 0 54. Barrau P., Fontaine J.P. and Morin B. Friction and Wear in Liquid Metals. Proc. Int. Conf. on Sodium Technology and Large Fast Reactor Design ANL-7520 (Part 1), 1968, 1 1 0 - 1 2 4 55. Brunori G., Fatineili U. and Tomasetli G. General Problems of Fretting Corrosion. Work in support of the nuclear fast reactors developed at CSN Casaccia, Presented at First Italian Tribolog.v Convention, San Donato Milanese, June 1973 56. D'Agraives B.C., Voican A. and Bacchilega A. Experimental Evaluation of the Wear of the Sub-Assembly of a PEC Type Fuel Element. Tribological experimental studies in Na at high temperature, CNEN/EURA TOM, EUR.580 7, 19 78 /in ItalimO 57. Fukuda S. Experimental Confirmation of the Design to Minimise Vibration and Wear in 61-pin Wire-spaced EBR-II Sub-assemblies. Trans, Am. Nuc. Soc., June 1978, 28,222 - 223 58.

Fontaine J.P. Quelques Aspects des Phenomenes de Frottemerit et de Fretting-Corrosion en Sodium Liquide. Corrosion: Traitements Protection et Finition, 1971, 19(7), 3 7 9 - 3 9 3

61. Hattie R.T. Tribological Behaviour of Materials for Use in Sodium Cooled Fast Breeder Steam Generator Applications. ibidem 62. Van Westenbrugge J.IC, Belier N.H.H., Schinkel J.W. and de Gee A.W.$. On the Wear Behaviour of Nb-stabilised 2¼Cr IMo Steel Tubes in Sodium. ibidem, 3 3 8 - 3 4 1 63. Midgley J.W. Communication: Proc. Symp. Non-conventional Lubricants and Bearing Materials Such as Used in Nuclear Engineering. 1. Mech. E., London, 1962 64. Campbell C.S. and Lewis M.WJ. Some Aspects of the Tribological Behaviour of Materials in Sodium. Proc. 2nd lntl. Conf. on Liquid Metal Technology in Energy Production, Richland, Wa., 1980 65. Cavell l.W. and Nicholas M.G. The Formation of Sodium Chromite on AISI-316 Stainless Steel Exposed to Oxygenated Liquid Sodium. A E R E Report (to be published) 66. Chivers T.C. and Gordelier S.C. Vibration Data and its Employment for Component Life Projections in AGR circulators. Paper 6/6, Proc. British Nuclear Engineering Society Conf. on Vibration in Nuclear Plant, Keswick, May 1978 67. Ko P.L. Wear of Zirconium Alloys due to Fretting and Periodic Impacting. Wear, 1979, 5 5 . 3 6 9 - 3 8 4 68. Corrosion and Wear Handbook for Water-Cooled Reactors. DePaul (Ed), 7"11)-7006, 1957 {USAEC) 69. Schmugar K.L. Vibration',d Characteristics and Wear of Fuel Rods. l'rans. 4th Intl. Conf. on Structural Mechanics in Reactor Technolog.v. Salt Francisco, August 1977 70. Gadda F. and Ronchetti C. The Fretting Corrosion Problem in Water Cooled Nuclear Reactors. Energy Nucl., (Milan), 1977, 24(6) 71. Fricker H.W. Fretting in Tube Supports of Heat Exchangers. Proe. 1. Mech. E. Con]i (Nuel. En. Gp.), Components Design hi HTRs Using Itelhlm as Coolant. 1972 72. Pettigrew M.J. Flow-Induced Vibration of Nuclear Power Sta'tion components, Especially CANDU Reactors. 19 77, Atomic Energy o f Canada Ltd, Report No AECL-5852 73. Kalashnikov V.V. and Solyanyj V.I. Material Aspects of Operating Reliability of Water-Cooled Power Reactor Fuel Elements. 19 78, A t. LTtnerg (in Russian),,. trans in J. Soy. A t. Energy, 44(6), 499.--505 74. Eickhoff K.G., Dickinson K. and Hamer A.N. Development and Operation of Sodium Test Rigs for PFR Components. I A E A / SM-130/13, Proe. Syntp. on Progress in Sodiunt-Cooled Fast Reactor Engflteering, Monaco, March 1970, 41 54 75.

Laithwaite J.M., Bowles L. and Delves P.H. Sodium Pumps for Fast Reactors. IAEA/SM-130/lO, ibident, 319--330

76. Belts C. and Roberts W.H. A Theoretical and Experimental Study of a Liquid-Lubricated Hydrostatic Journal Bearing. l.Mech.E. Proceedings, 1968-69, 183{1), 6 4 7 - 5 7 77. Ketola H.N. and McHugh H.D. Experimental Investigation of Water Lubricated Bearings in the Turbulent Regime, Bearing and Seal Design in Nuclear Power Machhwry (ed R.A. Burton) - ASME. 1968 (proc. ,~vmp. on Lubrication in Nuclear Applications, Miami Beach. June 196 7) 78. Pan C.H.T. and Vohr J.H. Super-Laminar Flow in Bearings and Seals. ibidem

59. Campbell C.S. and Lewis M.WJ. Tribology of Ferritic Steels in Sodium. Proc. BHtish Nuclear Energy Society Conference on Ferritic Steels for bast Reactor Steam Generators, London, 1977

79. Whitley S. Review of Research on Gas Bearings in the United Kingdom Atomic Energy Authority. Proc. 1st lntL Syrup. on (;as-Luhricated Bearings. ONR-A CR49. 1959

60. Wilde E. and Mack K.J. Ferritic Steels in Reactor Tribology. ibidem

80. Whitley S. Nuclear Applications of (;as Bearings. Univ. Sottthampton (;as Bearing Symposh~m, Paper 5. 1965

TRIBOLOGY international February 1981

27

Roberts - Tribology in nuclear power generation 81. Sternlicht B., Schwarz H. and Luchter S. Process Fluid Lubricated "l-urbomachinery in Nuclear Applications, Bearing and Seal Design in Nuclear Power Machinery (ed. R.A. Burton). ASME, 1968 (Proc. Syrup. on Lubrication in Nuclear Applications. Miami Beach, June 196 7), 1 - 20 82. Chemical Plant Centrifuge. NC'/'Newsletter No 26.

Lubrication fundamentals J. George Willis

This book is essentially from Mobil and is a very good example of highclass company literature. It covers in 16 very readable chapters the whole of the subject of automotive and industrial lubrication from the viewpoint of probably the oldest company in the business. Its style is clear, simple and descriptive, but, perhaps necessarily, at an almost elementary level. To illustrate its width of scope it begins with a fascinating chapter on the history of petroleum, ends with a very useful section on the disposal of used lubricants and covers just about every up-to-date topic in between.

9R

December 1977 83.

Berger J.D. High-speed Slitter for Irradiated Stainless Steel Tooling. Proc. 25th ConJ~ on Remote Svstems Technolo~(v. 1973, Am. Nuc. Soc. Inc., 123 129

84.

Baer J.W. HTGR Fuel Particle Crusher: Mark 2 Design. 1979, General Atomic Co, San Diego, Ca, Report No. GA-A.15447

There is a large chapter on the refining processes used to produce lubricants from petroleum, on the additives which improve their performance in specific respects, on the physical and chemical characteristics of lubricants and on their evaluation and performance. But the largest chapter is, quite properly, on machine elements: bearings, gears, slideways, couplings, chains, cams, cylinders and wire ropes. This is an excellent way of tackling the problem of covering the wide range of machinery as it allows the author to go more deeply into the common 'denominator' problems. This idea is, however, rather lost in tile case of cylinders as the reader is referred to the individual machines. It would have been a good opportunity to point out the general necessity of carefully running-in pistons and cylin-

T R I R N I CI~Y int~rnatinnal F~hruarv 1981

ders so that the piston rings are smoothly worn to the delicate double-taper profde and the cylinder acquires its smooth glaze necessary for hydrodynamic lubrication during full load operation. The onlysubject which seems not to have been covered is industrial hydraulic systems, though some hydrostatic systems are described in the chapter on automotive transmissions. This omission seems all the more surprising in view of the space and depth of treatment given to the somewhat esoteric effect of nuclear radiation on lubricants. It is altogether an excellent, wideranging survey of lubricants and lubrication. T.L Fowle Published, price Swiss Francs 92, b v Marcel Dekker lnc, 2 70 Madison A renue, New York, N Y 10016, USA