Wear, 64 (1980) 151 - 162 0 Elsevier Sequoia S.A., Lausanne - Printed in the Netherlands

151

TRIBOLOGICAL PROBLEMS IN SODIUM-COOLED NUCLEAR POWER PLANTS*

EUGEN WILD and KLAUS J. MACK

Kernforechungezentrum Karleruhe GmbH, 7500 Karleruhe. Postfach 3640 (F.R.G.)

(Received November 7,1979)

summary

Owing to a possible future shortage of uranium sodium-cooled fast breeder reactors are becoming important. Tribological processes in these facilities, especially in their nuclear components, are of special interest to reactor designers. However, reactor-specific operating conditions do not provide any parameters suitable for establishing reliable working tribo- systems between the mating interfaces of contacting components. The fric- tion and wear behaviour of materials has been extensively investigated under special reactor conditions. Work is being carried out on the equipment crite- ria characteristics of the frictional systems, the surface macroroughness and the chemical reactivity of liquid sodium. The current state of knowledge concerning the selection of materials, the interpretation of test data and engineering design is surveyed.

1. Objectives

The design of tribological systems in sodium-cooled reactors cannot be based on the data and past experience obtained with conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, the operating conditions specific to nuclear reactors and prevailing especially in the nuclear components of such facilities pose special problems [l] .

-4s shown in Fig. 1 a distinction is made between the conventional and non-conventional plant sections from the tribological point of view. The conventional section (right-hand side) includes the turbine and generator, and the nonconventional section (left-hand side) includes the nuclear reactor core and the primary and secondary cooling systems. Friction, wear

*Paper presented at the 4th International Conference on Tribology, Paisley, September 10 - 15,1979.

152

,NNERCONTA,~~~~T~.~.‘.‘.‘.‘.‘.‘.’.’.’.’.’.’ .,.,.,.,.,.,.,.,.,. ., ..:.I. ‘.‘.HEAT

.‘.‘.‘,‘. .~.~.~.~:~:‘.‘.‘.‘.‘:~:~:~EXCHANGEFi

SODIUM LOOP FEED WATER PUMP

non- conventional I conventional

Fig. 1. Schematic diagram of a sodium-cooled fast reactor.

TABLE 1

Operating conditions

Movement Oscillating, intermittent Pressure (MPa) ~60 Velocity (mm min-l) 0 - 1000 Sodium temperature (K) Q 900 Viscosity = vHzO 0, content (ppm) < 10

and lubrication problems in the conventional part have largely been solved, or at least their nature is known and understood. Therefore the precondi- tions for establishing reliable and stable lubricating systems can be deter- mined. In this part of the plant it is possible to use oils or greases with suf- ficiently high viscosity in’ friction or roller bearings and sufficiently high circumferential speeds will generally be encountered.

The reactor core, coolant pipes and heat exchangers can, apart from the sodium pump, be regarded as relatively rigid structures. However, the struc- ture consists of many tens of thousands of different components which in operation move relative to each other owing to differential thermal expan- sion or flow-induced vibration or during loading and unloading events [ 21.

In all these cases reactor-specific operating conditions are extremely unfavourable from the tribological point of view. Table 1 shows these para- meters which are known as the collective stress pattern in more detail. The most important points of contact are summarized in Table 2.

Figure 2 shows a section of core structure which in the sodium-cooled fast reactor SNR 300 consists of some 500 core elements. They are all centred in the bottom Glate and supported against each other by distance pads in support or restraint planes above and below the fuel zone indicated. The whole structure is restrained by an upper and a lower ferritic clamping ring. The fundamental hexagonal form of the fuel elements results in a honeycomb system in which each element must be able to self-orient. This

153

F@r SECTION A-A ww

SECTION B-B

Fig. 2. Schematic diagram of a core restraint system.

TABLE 2

Locations of rubbing contact

Fuel cladding/spacer Fuel element load pads Control rod guides Flowmeters Heat exchanger tube supports Centring knobs Valve rods, valve seats

is only possible if the coefficient of friction ~1 of the system between the adjacent elements is not greater than 0.57. A relative radial movement of this type caused by rod bowing can be superimposed on an axial movement brought about by differential temperatures. Fuel elements must be able to move axially even after prolonged periods of operation, especially for load- ing and unloading.

There are various friction contacts inside each fuel element, both in the area of each fuel rod suspension and in the spacer region as shown in Fig. 3. There may be relative movement between the individual spacers and between the spacers and the cladding tubes. If the coefficients of friction are too high (following swelling) grid-type spacers may be deformed or even destroyed. In adhesive wear with materig displacement there may be local blockages in the cooling channels. Abrasive wear may weaken the wall of the cladding tube involved [3] .

SPACERS

FUEL ROD

GRID WIRE SPACERS SPACERS

Fig. 3. Inner fuel rod arrangement.

WIRE SPACERS

CentrIng knob ConcentrIc poper

Ferritlc steel

(a) Ferritic steel HTX -tubes

Colmonoy 5

tube plate

(b)

Fig. 4. Different contact geometries.

Figure 4(a) shows the application of centring systems in various compo- nents (a thermoshock shield or concentric heat exchanger tubes); Fig. 4(b) shows heat exchanger pipe penetrations with unfinned and finned contact areas respectively. In typical straight tube heat exchangers the mounting of the grid spacers is comparable with that in fuel rod bundles (Fig. 5).

Despite all these adverse preconditions it is necessary to guarantee safe functioning of these complicated tribological systems over prolonged periods of time. For this purpose the tribological behaviour of all components in contact and moving relative to each other must be known. Tribological ex- periments have been carried out over a number of years to elucidate the

166

STRAIGHT HTX TUBES

GRID SPACER ’

Fig. 5. Schematic diagram of a straight tube heat exchanger.

associated problems. First, a number of commercially available materials were selected for screening tests. These had metallurgical, physical and mechanical properties which indicated good tribological behaviour and above average wear resistance. The most important aspects guiding this selection were grid parameters, compatability with liquid sodium, thermal expansion, heat resistance, cost and availability.

Screening studies were carried out under standard conditions and the materials with the best tribological properties in sodium were determined [4] . These materials were then subjected to parameter studies in which the tribological load limits were determined. The influence of the collective stresses specific to reactor conditions were determined. Wear determination has always been a special problem in this respect. This is understandable if one considers that experimental results on pairs of specimens with contact areas of the order of 1 cm2 must be extrapolated, in the specific case of distance pads, to a total area of 10 m2 over several years of operating time. Sophisticated experimental facilities were built to carry out the necessary experiments under extreme operating conditions.

2. Experimental facilities and execution of experiments

Various experimental facilities which could be connected to a sodium flow supply and clean-up system with four test vessels [ 5,6] were used to move pairs of materials relative to each other under loads in various geome- tries. A pin-on-disk configuration (Fig. 6(a)) was constructed for rotating or intermittent relative motion, and a plate-on-plate system (Fig. 6(b)) was con- structed for reciprocating relative motion. During the experiments the fric- tion forces were measured as the parameters were varied. After the speci- mens had been dismantled changes in length or shape caused by the wear were measured and the wear rate was determined.

(4 (b)

Fig. 6. Tribological test sections.

TABLE 3

Materials under investigation

Steels

Superalloys

1.4961, 1.4981 (austenitic) DTOz, 1.6770 (ferritic)

Inconel7 18 hone1 750 Hastelloy C Nimonic 90

Coatings

Coated carbides

Special task

Stellite 6 Stellite 1 Colmonoy 5 Colmonoy 6 Akrit Co50 Tribaloy 700

WC-Ni (85-15) Ferro Tic U Ferro TiC T Cr,C2 (LC-lc, LC-1H)

Armco Iron TZM (MO 99.4) Tantal

3. Materials

The steels and alloys studied are summarized in Table 3. As well as their tribological properties an important criterion of materials for use in compo- nents subject to friction wear in reactor facilities is their compatibility with liquid sodium.

157

I I / blcm 716 !

i Slellit 6

Fig. 7. Wear rates of materials.

4. Experimental results

During the first few years of the study wear resistance was the main criterion by which material pairs were classified. When the extent of wear expected in sodium-cooling systems for core design work still under way was established from available test results increased importance was attached first to the coefficient of friction and then to the problem of self-welding. Figure 7 shows the mean comparative wear resistance of different pairs of materials obtained using results from large numbers of experiments. In the evaluation of these data pairs with high wear rates, low compatibility with liquid metal or insufficient adhesive strengths between the coatings and the substrates were excluded. Only the cobalt alloy (Stellite 6) and the nickel alloy (Inconel 718) were considered to be satisfactory for use in reactor systems.

These alloys were used for a large number of parameter studies to establish coefficients of friction. The difficult operating conditions exist- ing in reactors were taken into account in determining test parameters [ 7, 81. Stellite 6 and I&one1 718 exhibited coefficients of friction well below the limit required by the reactor manufacturer. However, abrasion due to the wear of cob~t-cont~~g Stellite 6 (60% cobalt content) may result in circuit contamination. The coefficients of friction of Inconel 718 sometimes exceed the limit, especially at higher temperatures.

It was therefore found to be necessary to seek low cobalt or cobalt- free alternative materials with the same or better tribological properties.

158

RUBBING DISTANCE [ml -

1 Iv1 1 __-__---~-_____ _: _~f

0.25 2.5 AVERACE DRIVE VELOCITY [mm/mm] -

Fig. 8. Friction coefficients of CraC2 as a function of drive velocity.

Alternatives studied recently are CrC coatings and alitized and molybdenum- containing nickel alloys.

The results of the experiments on CrC were first compared with those performed by an American research establishment in order to obtain indica- tions of their potential influence on the different test section systems. There was some disagreement between the experimental results regarding the test criteria for the equipment. For this reason and also because the studies of various combinations of materials had not eventually led to satisfactory re- sults, it was particularly important to study the parameters which determine friction behaviour independently of the material involved so as to provide in- formation concerning the optimum tribological design of components subject to wear stresses. The parameters studied include the equipment criteria of the test section systems (elasticity, mass distribution and relaxa- tion vibrations) and the macrostructure of the surfaces.

It was found that relatively elastic mechanical systems produce higher friction coefficients than less elastic systems for the same experimental con- ditions. This occurs mainly at high temperatures or low friction rates. A fric- tional movement no longer proceeds continuously but in several steps inter- rupted by short holding times (the so-called stick-slip type of friction). The evaluation of numerous stick-slip experiments has shown that the levels of the coefficients of friction encountered are mainly determined by the period of time for which the specimens are kept under static loads during the fric- tional movement [9] . For this reason the coefficient of friction is defined as the static coefficient of friction & in the following explanations, while C(d represents the mean dynamic coefficient of friction of a friction pair.

Examples selected from many experimental results are the static and mean coefficients of friction of Stellite 6 on Stellite 6 plotted in Fig. 8 as a function of the driving speed and in Fig. 9 as a function of the elasticity of the test section.

The influence of surface macroroughness on the frictional behaviour of material pairs in sodium was studied using CrC specimens finished in differ- ent ways [lo]. A wear-resistant coating (LC-1H) was applied by detonation gun techniques and finished by brushing (R, < 30 pm), grinding (R, < 5 pm)

159

L

t

RUBBING

1 7,7q,yT; I- 0,5.4@ ------

t, ti

::q,,, I ‘. , --r-3*

h -

y w

8

g 0,3- - ~!zEE~~E

8MPo 873 K

MOVEMENT + 5mm !

834 !

LL.2 39.2 ’ 2.6

TEST SECTION ELi;T,C,TY6-hml N 1 -

Fig. 9. Friction coefficient of Cr,Cz as a function of the elasticity of the test section.

20 40 60 100 cumulatk rubbq distance [ml -_)

Fig. 10. Coefficient of friction of Cr& as a function of surface macroroughness.

or lapping (R, < 0.5 pm) where Rt is the depth of roughness. The results ob- tained in these experiments are shown in Fig. 10. This diagram is a summary of the friction coefficients taken from several series of tests of the three sur- face qualities studied. The coefficients of friction pd for the lapped and ground specimens are generally above 0.5. These values are not acceptable for reactor core components subject to friction loads. Only the coefficients of friction of the brushed specimens (Fig. 11) meet the requirements im- posed by the reactor manufacturer.

The different coefficients of friction for various surface roughnesses in sodium are due to the different volumes of liquid sodium stored between the contact areas. The sodium is also responsible for chemical reaction on the surface and for adsorption effects accompanied by a reduction in surface energy.

Chemical reactions between sodium and the sliding surfaces are par- ticularly important because the generation of corrosion layers generally favours friction processes. However, these special tribochemical processes are

160

Fig. 11. Brushed CraCs-coated surface (D-gun process).

t 0,s

i 1

0 0 20 LO 60 80 100

RUBBING DISTANCE [ml

H SODIUM. m ARG0N.m ARGON +NA.

Fig. 12. Coefficient of friction of Cr3C2 as a function of environment.

hindered by the oxygen content of sodium which is kept low particularly be- cause of the corrosion effects in all other components [ 11,121.

Different frictional behaviour can also be expected when components are in contact either in liquid sodium or in the cover gas atmosphere. There- fore experiments were carried out with CrsC, specimens (brushed surfaces) in liquid sodium, in sodium-saturated argon and for comparison in dry argon. As shown in Fig. 12 the lowest coefficients of friction were measured in sodium-saturated argon. In liquid sodium the friction coefficients increased over the rubbing distance (0.12 < 1-1 < 0.7) with a slight drop after each dwell time.

The experiments described were mainly performed on test specimens of simplified geometries (pin-on-disk and plate-on-plate geometries). In a

161

number of other experiments the specimens were of the original component geometries.

For the KNK II compact sodium-cooled reactor facility it was necessary to determine whether the tank wall would be weakened if self-welding, fric- tion and wear occurred alternately between the wall and the centring knobs of the thermoshock shield. Experimental results showed permanent mutual transfer of material between the contacting surfaces without an appreciable wear rate.

An investigation was carried out to determine whether the tribological criteria between the flowmeters and the instrument plate of the SNR-300 core would suffer deterioration if the plate was made of Inconel 718 and not of Stellite 6 as originally planned. Improved friction and self-welding data were found for Inconel 718.

Coefficients of friction between the cladding tubes and various spacers were determined for various fuel element designs (SNR-300, KNK carbide element and fuel element with ferritic cladding tube). The coefficients of friction measured with different spacer geometries were almost the same. For austenitic steels (cl < 0.9) and for ferritic steels (cc < 1.2) the coefficients of friction were clearly functions of temperature and load.

5. Conclusions

General information was obtained concerning the behaviour of friction systems under extreme reactor conditions. Materials were found which could overcome current problems even if they do not constitute satisfactory long- term solutions.

It has been clearly shown that there is not just one coefficient of fric- tion for a specific material, but that there are a number of coefficients of friction for a complex friction system. The test results provide information concerning the design of components subject to friction loads and their surface structures.

Specific experiments have contributed to the designs of fuel elements, fuel element duct load pads, control rod guides, centring systems (KNK II) and heat exchanger tube supports.

All the experimental data influence the design work of the reactor manufacturer. They are used by the responsible authorities (Technical In- spectorate) in licensing procedures to clarify special points of detail and to provide essential information to materials industries regarding the develop- ment of suitable materials with high wear and welding resistance and favour- able slip properties.

Future research efforts will mainly be directed towards,long-term materials selection and the influence of superimposed movements and extremely low relative speeds (in the stick-slip region) on the behaviour of tribological systems.

162

Acknowledgments

The authors thank Dr. I-3. Hoffrnann for helpful discussions, Dr. H. LJ. Borgstedt and Dr. Helga Schneider for considerable assistance with sodium purity control, material characterization and surface analysis, and Mr. M. Gegenheimer for the major part he played in running the sodium test facil- ities.

References

1 W. A. Glaeser, Survey and analysis of materials wear and friction in sodium, Rep. BMI-1907, Battelle Columbus Laboratories, 197 1.

2 R. N. Johnson et al., Development of low friction materials for LMFBR components, Znt. Conf. on Liquid Metal Technolo~ in Energy Production, Champion, U.S.A., 1976.

3 R. J. Jackson et al., Experimental fuel subassembly irradiation experience in EBR-II, Int. Conf. on Fast Breeder Fuel Performance, Monterey, U.S.A., 1979.

4 E. Wild, K. J. Mack and H. Hoffmann, Experimentelle Untersuchungen des Ver- schleissverhaltens von Stiihlen und Legierungen in Natrium, Rep. KFK 1251/1970, Kemforschungszentrum Karlsruhe GmbH, 1970.

5 E. Wild, K. J. Mack and G. Drechsler, Das Ver~hlei~verh~ten von Werkstoffen in fliissigem Natrium-Ver~chse~~ch~ngen und experiment&e Ergebnisse, Rep. KFK 1659/1972, Kernforschungszentrum Karlsruhe GmbH, 1970.

6 E. Wild and K. J. Mack, Tribology in the core of a sodium-cooled fast breeder reactor, Wear, 34 (1975) 331.

7 W. Dietz, H. Weber and E. Wild, Friction behavior of Inconel 718 and Stellite 6 as fuel assembly duct pad material, IAEA/SM-I 73, VIZ-18, BrusEetE, IAEA, Vienna, 1973.

8 E. Wild and K. J. Mack, Experimental parameter investigations on the tribological behavior of Stellite 6 in liquid sodium, Nucl. Technol., (Feb. 1979).

9 E, Wild and K. J. Mack, Friction and wear in liquid metal systems: comparability problems of test results obtained from different test facilities, Znt. Conf. on Liqtiid Metal Technology in Eneigy Production, Champion, U.S.A., 1976.

10 E. Wild and K. J. Mack, Lubrication in nuclear reactor components. Frictional sys- tems in liquid sodium and argon atmosphere, In& Conf. on Frontiers of Lubricant Technology, Naples, 1978.

11 H. U. Borgstedt, Natriumkorrosion von Verschleissfesten Kobaltlegierungen (Stellite), Werket. Korroe., 29 (1978).

12 G. A. Whitlow et al., Sodium combatibllity studies of low friction carbide coatings for reactor application, Int. Corrosion Forum, Chicago, 1974.