A review of fretting study on nuclear power equipment

Journal Pre-proof A review of fretting study on nuclear power equipment Zhen-bing Cai, Zheng-yang Li, Mei-gui Yin, Min-hao Zhu, Zhong-rong Zhou PII:

S0301-679X(19)30609-7

DOI:

https://doi.org/10.1016/j.triboint.2019.106095

Reference:

JTRI 106095

To appear in:

Tribology International

Received Date: 9 August 2019 Revised Date:

13 November 2019

Accepted Date: 29 November 2019

Please cite this article as: Cai Z-b, Li Z-y, Yin M-g, Zhu M-h, Zhou Z-r, A review of fretting study on nuclear power equipment, Tribology International (2020), doi: https://doi.org/10.1016/ j.triboint.2019.106095. This is a PDF file of an article that has undergone enhancements after acceptance, such as the addition of a cover page and metadata, and formatting for readability, but it is not yet the definitive version of record. This version will undergo additional copyediting, typesetting and review before it is published in its final form, but we are providing this version to give early visibility of the article. Please note that, during the production process, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain. © 2019 Published by Elsevier Ltd.

Graphical Abstract

Typical positions of fretting wear in steam generator.

A review of fretting study on nuclear power equipment Zhen-bing Cai, Zheng-yang Li, Mei-gui Yin ,Min-hao Zhu, Zhong-rong Zhou Tribology Research Institute, Southwest Jiaotong University,610031,Chengdu, China Abstract: Nuclear power plants, working in extremely harsh environment, primarily in the form of high-speed fluid flow, circulate through complex systems. Serious tribological problems can occur when a small amount of nuclear energy is converted into mechanical energy in the components (e.g., fuel-rod cladding, tube of the heat exchange systems). With the increase of the service life of nuclear power equipment, a considerable number of nuclear power equipment or structure failures occur one after another. Although the influencing factors are different, fretting damage is one of the important factors. Fretting damage has strong concealment and high risk, and it is often the main cause of component failure. Thus, improving the reliability of nuclear power equipment, extending their life, and optimizing their structure are important. In recent decades, many scholars have studied fretting wear, fretting fatigue, and fretting corrosion behavior in nuclear power equipment. Accordingly, they have solved many problems, accumulated a lot of experience, and put forward many criteria. In this article, the research status of fretting damage in key equipment and structures of nuclear power plant is reviewed. Keyword: nuclear powder, fretting, SG tube, fuel rod cladding 1. Introduction Electric power generation usage is a key factor of advances in the industry, agriculture, and socioeconomic level of living [1]. Electrical energy can be generated from burning mined and refined energy sources, such as coal (37.9%), natural gas (26.6%), oil (3.0%), and nuclear (10.1%), and from harnessing energy sources, such as hydro(15.8%), biomass (2.2%), wind (5.5%), solar (2.4%), and geothermal and wave power (0.4%). According to the report of the World Nuclear Association in July 2019, the global nuclear power generation is steadily growing. In 2018, the nuclear power generation reached 2563 TWh, an increase of 1.7% compared with 2519 TWh in 2017, and the per capita nuclear power consumption was 10.3%. China has made rapid progress in nuclear power generation, ranking third in the world. However, in 2018, nuclear power generation accounted for only 4.2% of total power generation, higher than 2017, but far below the global average of 10.3%. On the development 1/32

speed of nuclear power generation, China has the fastest development, with 13 nuclear reactors under construction, 43 was proposed. Table 1 Nuclear power plant reactors in some countries (June 2019) [2] Total Nuclear Power in 2018

June,2019 (Reactors) In operation

Under

Proposed

construction TWh

%e

Num

MWe

Num

MWe

Num

MWe

USA

808.0

19.3

97

98699

4

5000

3

2550

France

395.9

71.7

58

63130

1

1750

0

0

China

277.1

4.2

46

44688

12

11091

43

50900

Russia

191.3

17.9

36

29139

6

4973

24

25810

Korea

127.1

23.7

24

23231

4

5600

0

0

Germany

71.9

11.7

7

9444

0

0

0

0

U.K

59.1

17.7

15

8883

1

1720

3

5060

World

2563

10.3

447

399224

55

58894

111

121829

The world’s first nuclear power plant was put into operation 70 years ago. Fig 1 shows the development of nuclear power plants. At present, nuclear reactors in commercial operations mainly belong to the second- and third-generation nuclear reactors. To enable secured, efficient, and environmentally friendly nuclear energy than the existing reactors, the fourth-generation nuclear reactors have been actively developed by various research institutes to be applied in commercial operations before 2030 [3, 4].

Fig. 1. Development of nuclear power plants In nuclear power equipment, several kinds of components or materials are subjected to wear phenomena [5]. Fuel, steam generators, pipelines, and other systems involve many connectors. Wear of materials and interface loosening or the gap caused by vibration, heating, and heat transfer are inevitable. Taking pressurized water reactor (PWR) system as an example, the service life of a steam generator, which relies on a tube heat exchanger, is the most significant issue in nuclear power plants. A great number of steam generators are removed from the equipment to be repaired or 2/32

replaced due to flow-induced vibration (FIV), which initiates fretting that occurs during a heat exchange. Fretting with small-amplitude oscillatory motion occurring between tubes and their supports (or antivibration bars) [6]. Another component that undergoes fretting is the fuel cladding. The service environment of fuel cladding is more complex, involving high temperature and high-pressure water, intense neutron irradiation, electrochemistry, complex mechanical stress, and fission gas of nuclear fuel. The FIV cause small relative motions between the fuel rod and spacer grids. Grid-to-rod fretting (GTRF) is the leading cause of the fuel failure of PWRs and is one of the problems addressed by the Consortium for Advanced Simulation of Light Water Reactors to guide its efforts and develop a virtual reactor environment [7]. GTRF may produce wear of the rod cladding that can lead to the exposure of fuel pellets if not caught early enough (such situations are called “leakers”). Leakers are a significant concern for PWR designers and plant operators [8, 9]. More than 70% of leakers result from GTRF wear in PWRs. In this paper, the research status of nuclear power materials, mainly heat transfer and fuel systems, has been reviewed in the recent 20 years. The test materials, test methods, and research hotspots are also analyzed and summarized. 2 Fretting test method and test device The actual nuclear power equipment service environment is complex and harsh. Hence, the simulation research and analysis in the laboratory cannot easily achieve actual working conditions. In addition, obtaining the parameters of evaporator or fuel chamber components, such as vibration displacement, vibration frequency, and actual contact conditions between the support interfaces, is difficult. Except failure analysis, most research on the fretting of nuclear power materials is based on material-level experimental or numerical analysis. The research on fretting can be divided into fretting wear, fretting fatigue, and fretting corrosion. At present, many studies in the field of fretting wear are available. In the literature, fretting wear models have historically taken several forms:
Mechanical eccentric cylinder or cam [10, 11],
Hydraulic pressure [12, 13],
Piezomotors [14, 15],
Electromagnetic fields [16–18],
Voice coil motor [19–20],
Improvement by fatigue tester [21]. Mechanical eccentric cylinders have lower displacement output accuracy and wider displacement range, hydraulic pressure testers have smaller displacements and lower frequencies, and piezomotors and voice coil motor can provide small 3/32

displacement and high frequencies. The tester improved from the fatigue machine’s accuracy of displacement control is poor, and the fixture installation is not convenient enough. Generally, the installation and operational performance of the clamp in the horizontal mode is better than those in the vertical mode in the material testing processing. Thus, the use of the horizontal mode in fitting liquid containers, arranging motors and sensors, and clamping samples during tests is convenient.

(a) Eccentric model[10]

(b) Hydraulic pressure[12]

(c) Piezomotor[14]

(d) Electromagnetic[16]

(e) Voice coil motor[18-19] (f) Improvement by fatigue tester[21] Fig. 2 Fretting wear test machine with various driving modes Compared with fretting wear, fretting fatigue and fretting corrosion are less studied. Fretting fatigue. Fig 3a [22, 23] shows a typical schematic diagram and photograph of fretting fatigue testing apparatus in the room temperature test. Base plates were fastened on the main frame of the system. Loading screws were placed through the proving ring contact on the back of the arm whose root was welded to the 4/32

base plate. The inner tip of the arm fits in to the conical groove on the back of the bridge pad. As the loading screw was tightened, the tip of the arm pushed the pad in the ward and the contact force on the fatigue specimen increased. Many fretting corrosion tests only add a solution immersion chamber and an electrochemical measurement system on the fretting wear test equipment (Fig. 3b). Research institutes have recently developed micro-motion research equipment with an enclosed high-temperature fluid environment, which is near the service environment. However, the sensor cannot be implanted into the test system because of high temperature and pressure, which makes the tribological parameters of the actual interface difficult to obtain [24–26]. Researchers have to rely on the physical and chemical properties of the tested material after the end of the experiment.

(a) Fretting fatigue tester [22] (b) Fretting corrosion [23] Fig. 3 Fretting wear and fretting fatigue test machine 3 Fretting research in steam generator tubes 3.1 SG tube materials In currently operating nuclear power plants, steam generators, which always include a tubular evaporator, are primary responsible for the heat exchange between the primary and secondary circuits [27]. The service environment for steam generator tubes is exceptionally harsh, mainly due to high temperature, strong stresses, chemical corrosion, and neutron irradiation [28–29]. The heat transfer area of the steam generator in most commercial nuclear power plants are always composed of a large number of small-diameter thin-walled seamless U-shaped tubes. The outer diameter of these heat transfer tubes are usually approximately 22, 19, and 15.8 mm, and the corresponding wall thickness are approximately 1.27, 1.09, and 0.86 mm (Fig. 4).

Fig. 4. Dimensions of varied steam generator tubes Excellent structural characteristics and mechanical properties of steam generator 5/32

tubes play a key role in the heat transfer safety, capacity, and efficiency of nuclear power plants. Since the nuclear power plants have been put into operation, a considerable amount of effort has been made to improve the mechanical properties and corrosion resistance of steam generator tubes. In the 1960s, the heat transfer tubes of many Western press water reactors were manufactured by 304 and 316 austenitic stainless steels. After a period of service, they found that these materials are very sensitive to intergranular and stress corrosion under liquid mediums that contain chloride. To remedy this situation, most of them were substituted by the Inconel-600 (I-600MA) alloy that is only treated by factory annealing. In the mid-1970s, researchers found that erosion, pitting, intergranular corrosion, and other failure phenomena sequentially happen on this kind of tubes. To resolve this problem, refined craftwork has been applied to improve the intrinsic resistance of Inconel-600 alloy, and then Inconel-600TT was used since the mid-1980s. After the 1990s, the Inconel-690 alloy and Incoloy-800 were derived and used to manufacture heat exchanger tube production. All these materials were treated by some special heat treatment processes [30–33]. Figure 5 exhibits the evolution of steam generator tubes with time.

Fig. 5. Evolution of steam generator tube materials and hot working parameters. Fourth-generation nuclear reactor includes three different thermal neutron reactors (supercritical water reactor, very-high-temperature reactor, and molten salt reactor) and three fast neutron reactors (gas-cooled, sodium-cooled, lead-cooled fast reactors). The definitions of these new nuclear reactors differ from one another in terms of their neutron spectrum, coolant, and moderator [34]. Table 2 shows the materials that have been widely studied and are likely to be applied in the practical 6/32

employment of fourth-generation nuclear reactors [35–37]. Table 2. Fourth-generation nuclear reactors and their materials in heat exchange

3.2 Fretting behaviors of steam generator tubes In practical service, the working environment for steam generator tubes is complex and harsh. The difference in pressure inside and outside the tubes, neutron irradiation, and various fretting wears and fatigue behaviors can easily be identified among heat exchange tubes because of FIV. Furthermore, high temperature and water pressure containing boric acid and lithium hydroxide have corrosive actions to steam generator tubes. Thus, the fretting corrosion is common among steam generator tubes, and such damages are difficult to overcome and avoid. Figure 6a shows some typical wear positions that often occur between heat transfer tubes and their support structures. To alleviate and avoid the fretting wear and corrosion of steam generator tubes, researchers have paid much attention to the thorough and detailed studies of these phenomena and problems under different conditions [38].

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(a)Typical positions of fretting wear in steam generator

Fig. (b)Fretting researchon the heat tube materials Fig.6 Fretting in steam generator and heat tube materials Considerable research works on the fretting behaviors and performance of steam generator tube materials, fretting wear, fretting fatigue, and fretting corrosion are available in the literature. To make the experimental studies similar to actual service conditions of heat exchange tubes, different experimental variables have been simulated and implemented for investigation. Figure 6 b exhibits the experimental variables that researchers have widely studied under their real research works. A lot of research works on the fretting behaviors of various steam generator tubes have been performed, and some remarkable results have been achieved. These research works can be roughly divided into five parts based on the test environment: 8/32

(a) Dry condition room temperature in air: The influence of displacement amplitude, sliding frequency, normal force, and wear cycles have been widely studied by researchers. Almost all results indicate that the friction force, wear volume, and fretting mode change with the amplitude [39–41]. Yun et al. concluded that the wear mode does not change with increasing sliding frequency [42]. Li et al. found that the wear degree and mechanism of alloy tubes slowly change with the increase of wear cycles, and all these will rapidly change after a particular number of cycles [43]. Xin’s results revealed that the normal force can affect the material transfer, which occurs with different slip regimes [44–45]. Soria et al. found that the dominant wear mechanism mainly depends on the particular composition of the mated wear materials [46]. (b) Dry condition with different air temperatures: Heat exchange tubes are serviced in high-temperature environments. Xin et al. studied the influence of displacement amplitude and wear cycles on the structure of wear scars [47, 48]. They found that the content of oxygen can influence the characteristics of nanostructured tribo-layers and therefore affect the wear degree and mechanism of materials [49]. Mi’s results found that the wear volume in 90 °C is more serious than that of 200 °C and 285 °C [13]. Zhang et al concluded that fretting running regimes are basically unaffected by temperature [50, 51]. Kwon et al. studied the fretting wear behaviors of Inconel-600 under room and 320 °C temperatures. The wear groove that formed in high temperature was much rougher than that formed in the room temperature, and the fatigue strength was not significantly different [22]. (c) Water condition room temperature: Xin et al. compared the wear behaviors of Inconel-600 alloy tubes under the conditions of dry air and deionized water. They found that the white etching layer is more difficult to form in deionized water than that in dry condition [52]. Li et al. studied how the polarization condition influence the wear behaviors of alloy tubes and found that the crack nucleation and propagation, microstructure evolution and damaged mechanisms, COF, and wear volume strongly depend on the polarization condition [53]. (d) Different water temperatures/air condition: In practical conditions, the steam generator tubes are always serviced in high-temperature and water pressure solutions. Ming and Xin investigated the effect of normal force on the fretting wear behaviors of alloy tubes under 288 °C pressed water. They all found that normal force significantly influences the degree of wear and failure mode [15, 54]. Guo found that test samples mainly exhibit abrasive wear when the sliding amplitude or normal load is low; however, delamination wear dominates when either of the element is increased [55]. Liao concluded that the fretting fatigue life of Inconel-600 can be in high temperature water [56]. Lai et al. found that the wear volume of test samples increased with the water temperature before 90 °C and then decreased beyond that temperature [24, 25]. Xin et al. found that sliding frequency has little effects on the 9/32

oxide phases of worn surfaces; however, the increasing frequency can enhance the thickness of oxide film, which can accelerate the fretting corrosion of test samples [57]. Guo found that the higher the chromium content of mated wear material, the better the wear resistance of the Inconel-690 alloy tube [24]. Guo separately compared the wear behaviors of alloy tubes under argon and water conditions and found that high temperature water can attribute to the form of oxide layer and transportation of wear particles in the fretting process and finally strengthen the wear resistance of materials [24]. (e) Other solutions: Wang investigated how the PH value of solution affects the wear behavior of alloy tubes and the results showed that the corrosion rate decreased as the PH values increase from 4 to 13. The Inconel-690 alloy became the active metal at a pH value of 13 [58]. Li et al. found that the layer of wear scar will be oxidized by the chloride ion [59]. Table 3 Test materials and parameters. Tubes

Friction

Lubrication

Temperature

Displacement

Frequency

pair

condition

℃

(µm)

(Hz)

Ref

In600

GCr 15

dry

RT

30-150

20

[40.45]

In690

409SUS

dry

RT

25-300

10-60

[44]

405SS

dry

RT,90,200,285

100,200

5

[12]

405SS

8.6 MP water

285

20-100

5

[59]

405SS

2 MPa water

RT, 90 °C

100

5

[26]

10 MPa water

285 °C

100

5

18 MPa water

350 °C

100

5

405SS

8.6MPa water/Agon

285 °C

100

5

[25]

304SS

PH1-4,PH7,PH10,13.

RT

30

20

[66]

304SS

3.5% NaCl solution

RT

100

20

[67]

304SS

Dry and water

RT

Impact

5

[18]

1Cr13

dry

RT,100,300

100,200

10,50

[54]

In690

304SS

dry

RT

15-75

-

[41]

TT

304SS

dry

RT

45

20

[46,47]

304SS

dry

220

90

20

[49]

304SS

dry

RT,320

15-75

20

[50]

304SS

5 vol%-21% oxygen

320

150

20

[51]

304ss

Water, 15 MPa

288

96,139.165

30

[15]

304ss

dry

RT

15-75

20

[55]

Inconel

AISI304

dry

RT

75

15

[48]

800

AISI1060

In600 MA

10/32

304SS

dry

RT,300.400

2-40

2

[53]

(a) Damage evolution with increasing cycles[43] (b) Cross sections of worn scar [47]

(c) Schematic of the typical worn surface [47]

(d) Damage evolution [49] Fig. 7 Results of the fretting wear scar

Research works about the impact or impact–sliding wear behaviors of various steam generator tubes have been carried out. Cai et al. used a self-made test rig to study the impact wear behaviors of alloy tubes under different test conditions or parameters [60–62]. Souilliart et al. investigated the influence of incidence energy and angle on the impact wear behaviors of steam generator tubes under dry and water environments [63–64]. For the impact–sliding wear, Cai et al. designed a test rig to study the effect of impact and sliding velocities on the wear behaviors of Inconel-690 alloy tubes [65]. Cheng et al. studied how the ratio of impact and sliding frequency and displacement amplitude affect the wear behaviors of heat exchange tubes based on an impact–sliding wear test rig [66, 67]. Figure 8 shows the variables that have 11/32

been investigated during the impact or impact–sliding wear test of steam generator tubes.

Fig. 8. Impact and impact–sliding wear[66-68] 3. 3 Surface strengthening processes for anti-fretting Researchers have proposed some technologies for metal surface intensity to further improve the wear performances and some mechanical properties of alloy tubes. Figure 9 summarizes the surface hardening process that has been applied to improve some mechanical properties of steam generator tubes. Hu et al. studied the influence of different reductions of cold rolling on the fretting wear behavior and mechanism of Inconel-690 alloy at 320 °C in air. The grains were horizontally elongated along the rolling direction as the degree of cold rolling increased (Fig. 10 a). They found that the damage mechanism of test samples changed from oxidation, delamination, and abrasive wear to adhesive wear with the increase of cold rolling reduction [68]. Lee et al. studied how the temperature and time of aging treatment affect the wear behaviors of alloy tubes. No carbide was found after aging treated for 1 h, and the carbides that precipitated along the grain boundary became coarse as the treated time increased to 10 h (Fig. 10b). They found that the wear behavior of test samples was significantly affected by the carbides that precipitated during the treatment process. The friction coefficient and wear volume increased with the carbide precipitation [69, 70]. Zhang et al. investigated that the grain size of the Inconel-600 alloy increased and its hardness decreased as the solution temperature increased. During the wear process, the bigger the grain size, the larger the wear volume; however, the bulk hardness played a little role in the wear volume [71]. From the cross section of wear scars, the length of the main crack distinctly increases at a solution temperature higher than 1000 °C (Fig. 10 c). After alloy tubes treated by ultrasonic surface rolling process 12/32

(USRP), a hardening of surface up to a certain depth is observed. Microstructural evolution and hardening are attributed to the increase in density of various dislocation structures [72]. When the same wear condition is suffered, the wear resistance of the USRP treated sample is obviously better than that of the untreated one (Fig. 10d).

Fig. 9. Processes for alloy tube surface strengthening

(a) Cold rolling [68]

(b) Aged treated [69]

13/32

(c) Cross section in high temperature [70] (d) Treated USRP [65] Fig. 10. Wear characters of alloy tubes 3.4 Development trend of the fretting steam generator tubes in the future The materials used for heat transfer tubes of nuclear power evaporators have gone through the evolution process of Inconel 600, Incoloy 800 and Inconel 690. Inconel 690 is widely used in the heat transfer tubes of nuclear power evaporators because of its outstanding oxidation resistance and stress corrosion resistance. As for the fretting research of heat transfer tubes, there may be the following trends in the future: (a) The complex fretting forms of 690 alloy were studied, such as fretting damage mechanism under impact-tangential coupling, fretting corrosion behavior and damage mechanism, fretting damage under abrasive action. (b) Heat transfer tube of the fourth generation reactor, fretting under the action of molten metal medium (such as metal sodium, lead bismuth environment) (c) The application of surface engineering technology in heat transfer tube, including fretting damage behavior, friction and wear resistance behavior.

. Fretting on fuel rod system Fuel rods are the key components of a light-water reactor’s fuel assembly. Fuel rods are composed of cladding tube in which uranium dioxide pellets are stacked [73, 74]. Cladding tube has a length of approximately 4 m, a diameter of approximately 9.5 mm, and wall thickness of approximately 0.5 mm. The fuel rods are supported by spacer grids, and the rods directly contact with the springs and dimples. The most important function of cladding tube is to prevent fission products from escaping and efficiently dissipate heat [75]. 14/32

The high temperature, high pressure, high radiation, and severe operating condition in the reactor cause the failure of cladding tube. According to the survey, fuel rod failure is mainly caused by the GTRF and debris-induced fretting wear in Fig. 11 [76].The GTRF was caused by excessive FIV, and this vibration is restrained by its contact with the springs and dimples of the spacer grid [77–78]. The excessive FIV may be divided into external and internal vibrations. The external vibrations were caused by reactor internals, inter fuel-assembly gaps, and fuel assembly-shroud gaps, whereas the internal vibrations were due to inadequate spacer grid mixing vane designs, spacer grid strap designs, water-to-fuel volume ratios, fuel rod and assembly stiffness, and bottom and top nozzle flow hole designs. GTRF will cause wear damage of the cladding tube in Fig. 12, which may lead to the wearing out of the cladding tube in serious cases. If such case happens, then radioactive fission gases in the fuel rods would be released to the coolant, which increases the radioactivity level of the coolant. Even if a perforation does not occur, thickness reduction of cladding tube caused by wear can decrease the strength of the fuel rods [79–80].

(a) 2002 to 2006 (b) 2007 to 2011 Fig. 11 Percentage of fuel failure in PWRs all over the world [76]

Fig. 12 Fretting damage of cladding tube in PWRs [79-80] 15/32

4.1 Evolution of cladding materials The preparation of cladding materials is important, and the materials must possess excellent corrosion, fretting wear, creep, and high-temperature thermal oxidation resistance property in the aggressive reactor environment. Zirconium alloys are widely used as a fuel cladding material in the world because of low thermal neutron capture cross section of 0.18×10-28 m2, good corrosion resistance, good compatibility of uranium fuel, and good machining performance [81–82]. Figure 13 shows the evolution of cladding materials developed by the United States, Korea, and China. Zirconium alloys have been developed to three types: Zr–Sn, Zr–Nb, and Zr–Sn– Nb. Table 4 shows the typical cladding materials that have been successfully used and tested in the world [83, 84]. In the 1950s, Zr-2 alloy was developed in the United States, which has a good high-temperature water and steam corrosion performance and successfully used as fuel cladding material [85]. However, the presence of Ni increases the hydrogen absorption of Zr-2 alloy [86]. Therefore, Zr-4 alloy was developed by decreasing Ni and increasing Fe content on the basis of Zr-2 alloy. Zr-2, Zr-4, and Zr-1Nb are called as first-generation zirconium alloys. With the development of smelting technology, Sn content in zirconium alloy can be controlled at a lower level. By adjusting the contents of Sn, Fe, and Cr and improving the microstructure of the alloy through heat treatment, an improved Zr-4, namely, low-tin Zr-4 alloy was developed, which is considered a second-generation zirconium alloy [87]. To achieve higher usage requirements, researchers developed Zr–Sn–Nb alloy after combining the advantages of Zr–Sn and Zr–Nb alloys. Zirlo, M5, HANA, and N36 alloys are called as third-generation zirconium alloys [76, 88, 89]. In the 2011 Fukushima nuclear power plant accident, zirconium alloy cladding in the reactor fuel was aggressively oxidized by high-temperature steam generated under accident conditions, resulting in a large amount of hydrogen and heat, which eventually led to a reactor core meltdown and hydrogen explosion [89]. From then on, the development programs of accident-tolerant fuels (ATFs) have been initiated worldwide. ATF program is aimed at the development of alternative cladding and fuel concepts. According to the research and evaluation of different researchers, ATF cladding materials can be classified into ceramic and metallic materials [84, 90, 91]. The metallic materials mainly include FeCrAl and Mo-based alloys. The ceramic materials are mainly represented by SiC/SiC composite materials.

16/32

Fig. 13 Evolution of cladding materials developed by the United States, South Korea, and China Table.4 Chemical element composition of the typical zirconium alloys Country

Alloy

Chemical element composition (w/%) Zr

Nb

Sn

Fe

Cr

Ni

Zr-2

Balance

-

1.2-1.7

0.07-0.20

0.05-0.15

0.03-0.08

Zr-4

Balance

-

1.2-1.7

0.18-0.24

0.07-0.13

0.007

Zr-4 (Low Sn)

Balance

-

1.2-1.5

0.18-0.24

0.07-0.13

-

Zirlo

Balance

1.0

1.0

0.1

-

-

X5A

Balance

0.3

0.5

0.35

0.25

Japan

NDA

Balance

1.0

1.0

0.27

0.16

0.01

France

M5

Balance

1.0

-

-

-

-

Russia

E635

Balance

1.0

1.3

0.4

-

-

Korea

HANA6

Balance

1.1

N18

Balance

0.3

1.0

0.3

0.1

-

N36

Balance

1.0

1.0

0.3

-

-

C7

Balance

1.0

-

0.01 (Fe)

-

-

U.S.A

China

0.05 (Cu)

4.2 Fretting wear of cladding tube in different test conditions The fretting wear mechanism of zirconium alloy has been studied worldwide. In the initial stage, considerable research has been conducted in the dry condition, and this condition can only partially reveal the fretting wear mechanism of the cladding tube. Kim [92–93] studied the fretting wear of Zr-4 alloy with Inconel-600 alloy in the air environment, indicating that wear volume increased with load, slip amplitude, and number of cycles and wear volume was mainly affected by slip amplitude. Attia [94–95] investigated the fretting wear mechanism of Zr-2.5Nb alloy in air at 265 ℃, using a specially designed fretting wear tribometer. A novel approach of the thermal 17/32

and electrical contact resistances was introduced to measure the oxide layer thickness in the fretting interface. Winter [96–97] reported that FeCrAl alloy is an excellent ATF cladding material compared with Zr-4 alloy and SiC/SiC composite material due to its good wear resistance performance. The formation of a protective aluminum oxide and chromium oxide layer can reduce the COF and wear at 350 ℃ under dry fretting conditions. With the development of testing technology, fretting wear test was conducted in an autoclave to simulate the high-pressure water environment in a reactor. Qu [98–100] investigated the effect of water temperature (22 °C, 150 °C, and 204 °C) on the fretting wear behavior of zirconium alloy used in actual PWR and found that the zirconium alloy wear volume rapidly increased with a high water temperature. Zhang [101–102] revealed that the dominant wear mechanisms of SZA-4 alloys in 300 °C pressurized B–Li water condition are delamination and oxidation. Gene [103] investigated the change in surface roughness of wear surface before and after fretting testing in 260 °C water condition and found evidence of plastic deformation and wear debris. Fretting wear test in high-pressure water environment can simulate the actual operating condition and reveal the synergetic action between wear and corrosion. However, sensors (such as force and displacement) are limited and difficult to be applied in to a high-pressure water environment to monitor the state data and fretting run region in real time. 4.3 Effect of spacer grid shape on the fretting wear of cladding tube GTRF behavior is also affected by the design of a spacer grid. Different grids decide the contact mode among the spring, dimple, and fuel rod. Lee and Kim [80, 104–110] designed springs and dimples with different geometric shapes (Fig. 14a) and investigated the effect of supporting structure shape on the fretting wear behavior of zirconium alloys. The results shows that the shape of supporting structures directly decides the contact form (point, line, and surface contacts) and further affect the contact stress, contact area, wear depth, wear area, wear evolution, debris dispersion, fretting region, and wear volume. Kim [111, 112] investigated two spacer grids that have been applied in PWR on the fretting wear behavior of Zr-4 alloy (Fig. 14b). Grid B with conformal contacts shows a better fretting wear resistance than Grid A with point contacts. Hu [113–114] reveals that the wear rate depends on the size of gap between the grid and fuel rod (Fig. 14c) and found that the existence of a critical gap size results in a maximum wear rate. Kovács [115] found that the springs with a larger radius of curvature has a larger contact points and larger contact surfaces (Fig. 14d), which leads to lower wear depth and better fretting resistance behavior. Shin [116– 117] used the homology constraint model and found that maintaining the elastic 18/32

deformation of the spring center can increase the contact area and decrease the fretting wear (Fig. 14e).

(a)

(b)

(c)

(d) (e) Fig. 14 Different shapes of space grid: (a) (b) [80,104–108], (c) [111–112], (d) [113], (e) [114] 19/32

4.4 Impact fretting wear of cladding tube In actual operating conditions, the fuel assembly is subjected to axial and transverse flows of high-speed coolants. Impact, sliding, and impact–sliding movement occur in the cladding tube with spring and dimple (Fig. 15) [110]. Pablo [118] proposed a nonlinear vibration model and calculated the impact force between the rod and supports as approximately 0.2 N. Ko [119] found that the impact fretting on zirconium alloys is correlated with motion parameters, test duration, material combination, and temperature. Kim [108] compared the sliding and sliding–impact motion mode on the wear property of Zr-4 alloy and found that the sliding–impact motion mode shows a larger wear volume. Jones [120] investigated the impact fretting wear behavior on the reactor component in the pressurized CO2 advanced gas-cooled reactor. The impact wear behavior shows a transition from a severe wear to a mild wear at a temperature in the range of 250 °C–450 °C, and the critical temperature depends on the material and test parameters. Fisher [121] found that the wear volume of zirconium alloy under the impact fretting wear condition at temperatures below 150 °C is approximately equal to that above 300 °C but is five to 10 times larger than that in 250 °C–286 °C. Cai [19] studied the effect of diameter– thickness ratio on the impact fretting wear of Zr-4 alloy in a novel low-velocity impact fretting wear test rig. The test rig can provide cycling constant kinetic energy at every impact cycle and measure and record the deformation, loss of kinetic energy, absorbed energy rate, and contact force in real time and each cycle during the impact process. The results showed that under the same tube outer diameter, an increase of tube wall thickness will lead to an increase in contact time and a reduction in peak force, energy absorption ratio, and damage degree. Many researchers just use the tangential fretting wear to evaluate the wear resistance of cladding tube and ignore the impact fretting wear and impact–sliding fretting wear[18-20,61-62,66-67]. However, many studies indicate that impact–sliding fretting wear has a larger wear damage than single impact fretting wear and tangential fretting wear. Hence, impact fretting wear and impact–sliding wear of cladding tube is needed to perform more research and evaluate reliability.

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(a)Impact and tangential motion[110]

(b) Impact fretting[19]

Fig. 15 Impact and tangential motion of the cladding tube and impact fretting 4.5 Finite element analysis and numerical model on GTRF Finite element analysis (FEA) and numerical simulations have been used in GTRF to analyze the contact state, optimize spacer grid design, simulate vibration of the fuel rod, and predict wear rate. Yan [122] developed a method based on computational fluid dynamics to predict the GTRF in the fuel inlet region and found that a P-grid in fuel assembly inlet regions can greatly reduce the fretting damage. Bakosi [123] build an implicit large-eddy simulations of the time-dependent single-phase turbulent flow to calculate the shear forces with time on the fuel rods in 3 × 3 and 5 × 5 rod bundles with a single grid spacer. Kim [111–112, 124] used ANSYS to calculate the fuel rod vibration modes and natural frequencies and proposed that large initial elastic spring deflection of spacer grid, small cladding creep down, low spacer grid spring force relaxation, and large grid-to-rod contact area will reduce the fretting wear damage of cladding tube. Song [117] used a FEA to check a newly designed spacer grid and found that the new spacer grid can reduce the fretting wear of cladding tube and improve the crush strength of the spacer grid. Jiang [100, 125] proposed three-dimensional FEA models with detailed geometries of the dimple, spring, and fuel rods to evaluate the flow-induced impact intensity between the fuel rods and spacer grids. Mohany [126] developed a numerical model to predict the vibration response of CANDU fuel bundle and the associated fretting wear of cladding tube. This model discusses two excitation mechanisms, namely, turbulence-induced excitation and seismic excitation, and reveals that seismic excitation has no effect on the long-term fretting wear of fuel bundle. Blau [127–128] constructed a GTRF model based on the concepts of classical fretting to accommodate the complexities of the PWR environment. This empirical model can input the friction coefficient, slip amplitude, load history, frequency of oscillation, and contact area to determine the overall wear 21/32

depth and wear volume with time. Dyk [129] presented a modelling concept for the GTRF analysis of fuel rods caused by the vibration of coolant. Based on the simulated vibration model of the fuel rod, the evolution of the GTRF during operation can be determined. Lu [130–131] presented a numerical model combined with creep and wear to analyze the gap formation during grid-to-rod fretting and discussed the effect of pressure fluctuations, friction coefficients, wear coefficients, and initial interference on the wear profile and time at which contact is lost between the rod and grid. FEA and numerical simulation are important to analyze and predict the contact stress, vibration frequency, displacement amplitude, wear rate, and service life of cladding tube in PWR because GTRF in high-pressure water environment can hardly be monitored in actual condition and simulated in experimental conditions. FEA and numerical simulation are helpful to life prediction with engineering data, and they are key research points in the future. 4.6 Surface engineering technology applied on cladding tube To further satisfy the severe operation condition and improve the ability of accident tolerance, surface engineering is one of the most promising and effective ways to improve the property of cladding materials [132]. Three main principles for the selection of coating on zirconium alloy was proposed: (a) satisfy the operating conditions in the reactor and has a good corrosion resistance, hydrogen absorption resistance, radiation resistance, low thermal neutron capture cross section, and high heat transfer efficiency under normal operating conditions; (b) has an excellent ability of accident tolerance and good resistance to steam oxidation and thermal shock; and (c) has a good compatibility with zirconium alloy, such as expansion coefficient and preparation process [133]. The preparation methods of surface coating materials for zirconium alloys have been mainly reported in vapor deposition [134], cold spraying [135], micro-arc oxidation [136–137], thermal spraying [138], arc ion plating [139], and magnetron sputtering [140]. The studied coatings include the MAX phases (Ti2AlC [135], Ti3SiC2 [141]), carbide (SiC [134], ZrC [142]), nitride (TiN [143– 144]), oxide (ZrO2 [82, 136, 137], Al2O3 [145]), metal coating (FeCrAl [7, 146], Cr [87,148–149], Zr [149], and composite coating [150]). The preparation methods and properties of zirconium alloy surface coatings are shown in Table 5. Cr coating is the most promising coating for zirconium alloys due to its excellent high-temperature oxidation resistance [151]. Researchers are working on new coatings for zirconium alloys. At present, the selection of coating materials and coating process on zirconium alloys is a complicated process, which needs to be repeatedly verified in accordance with the research results of the key application properties of the coating, and a lot of research work is needed. 22/32

Table 5. Types and preparation process of coatings on the cladding tube Coating

Coating/Substrate

Coating process

Ti2AlC/Zr-4

Cold spraying

90

[135]

Ti3SiC2/Zr-4

Hot press

50-100

[141]

SiC/Zr-4

Chemical vapor deposition

1

[134]

ZrC/Zr-4

Chemical vapor deposition

-

[142]

TiN/ Zirlo

Physical vapor deposition

10

[143]

ZrO2/Zr-4, Zirlo, Zr-1Nb

Micro-arc oxidation

5-10

Al2O3/Zirlo

Atomic layer deposition

0.75

[145]

FeCrAl/ Zr alloys

Hot isostatic pressing

2

[146]

Cr/Zr-4

arc ion plated

10

[139]

FeCrAl, Zr /Mo

Physical vapor deposition

50

[149]

Cr-Zr /Cr/ Cr-N/E110, Zr-1Nb

Vacuum-arc evaporation

2

[150]

thickness/µm

Ref.

[82,136] [137]

4.7 Development trend of the fretting wear of cladding tube in the future GTRF has been studied for more than half a century ago, and cladding materials are constantly changing to improve the ability of accident tolerance. At present, key research points of GTRF should be focused on the following: (a) The development of new ATF cladding materials (coatings and alloy) is needed urgently to reduce the wear damage of GTRF. (b) More advanced techlonogies should be applied to the fretting wear test rig to similate high-radiation, high-pressure, and high-temperature water condition in reactors and monitor the state data and run region in real time. (c) More attention should be given to the failure mode of impact and impact–sliding wear on the cladding tube, which shows a more severe damage than the single tangential fretting wear. (d) Theoretical analysis should be emphasized to analyze the vibration state and predict the service life of the cladding tube combined with engineering and experimental data.

Conclusion Efforts to improve the existing research on impact fretting in nuclear power plants have been extensively reviewed in the present paper. The highlighted points are listed below: (a) Fretting wear equipment of various modes has been developed, and the test environment is becoming more similar to engineering practice. The contents of 23/32

fretting wear (e.g., materials, test conditions) are more than those of fatigue and corrosion (b)The fretting characteristics of advanced surface engineering technology applied to nuclear power plants is a research hotspot in the future. (c)As the nuclear power service environment is very complex, studying the damage progress of related materials under micro-action, obtaining real-time data, and carrying out life prediction are worthwhile.

Acknowledgement Young Scientific Innovation Team of Science and Technology of Sichuan (Nos. 2017TD0017) and National Natural Science Foundation of China (Nos.51627806) .

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Table 2. Fourth-generation nuclear reactors and their materials in heat exchange Type

Thermal Neutron Reactors

Fast Neutron Reactors

Reactor

Coolant

Coolant temperature In/out( )

Materials for heat exchanger

Supercritical water reactor

Light-water

280 (inlet)/500 (o utlet)

Very high temperature reactor

Helium

640 (inlet)/≥950 ( outlet)

Martensitic/ Ferrite Stainless Steel/ Austenite Steel Nickel Base Alloy/ Ceramic

Liquid salt

565 (inlet)/≥700 ( outlet) 490 (inlet)/≥850 ( outlet)

Gas-cooled Fast reactor

Helium

Sodium-coole d fast reactor

Liquid sodium

98 (inlet)/≥550 (o utlet)

Lead-cooled fast reactor

Molten lead or lead-bismuth eutectic

420 (inlet)/≥540 ( outlet)

Nickel Base Alloy Martensitic/ Ferrite Stainless Steel/ Austenite Steel/ Oxide Dispersion-Strengthened Alloy/ Nickel Base Alloy/ Ceramic Martensitic/ Ferrite Stainless Steel/ Austenite Steel/ Oxide Dispersion-Strengthened Alloy Martensitic/ Ferrite Stainless Steel/ Austenite Steel/ Oxide Dispersion-Strengthened Alloy

Highlights

1.

The research status of fretting method, nuclear materials damage in key equipment and structures of nuclear power plant is reviewed.

2.

Fretting research of heat exchange materials for steam generators in nuclear power systems is reviewed.

3.

The research status of fretting wear of fuel cladding is analyzed, including the latest coating and surface strengthening technologies.