Nuclear fuel rod

Abstract

A nuclear fuel rod for a pressurised or boiling water reactor, the rod having a corrosion-resistant tube with an external surface comprising a textured pattern 100. The textured patter has first areas

Description

Field of the Invention The present invention relates to a nuclear fuel rod, and particularly to a nuclear fuel rod having an external surface with a textured pattern. Background Nuclear reactors are a useful addition to a power grid. Specifically, they provide base load stations, but are also considered a low-carbon source of electricity and are not dependent upon variable weather conditions, which is a limiting factor for other low-carbon sources. Two of the most common types of nuclear reactor are the pressurised water reactor (PWR) and the boiling water reactor (BWR). PWRs and BWRs are well-understood and are scalable, and thus are suitable for both large-scale power plants as well as for small modular reactors. Figure 1 is a schematic diagram of a PWR 20. A reactor pressure vessel (RPV) 22 enclosing a reactor core containing fuel assemblies is centrally located in the reactor. Clustered around the RPV are three steam generators 24 connected to the RPV by pipework 23 of the pressurised water primary coolant circuit. A pressuriser 28 maintains the water pressure in the primary coolant circuit. Coolant pumps suspended beneath the steam generators circulate pressurised water around the primary coolant circuit, taking heated water from the RPV to the steam generators, and cooled water from the steam generators to the RPV. In the steam generators, heat is transferred from the pressurised water to feed water circulating in pipework of a secondary coolant circuit 26, thereby producing steam which is used to drive turbines which in turn drive an electricity-generator. The steam is then condensed before returning to the steam generators. The pressuriser maintains a pressure of around 15.5 MPa in the primary circuit. The primary circuit of pressurised water is therefore used as coolant, moderator and heat transfer fluid for generating steam in the secondary circuit. Unlike a BWR, the pressurised water coolant is kept at a pressure so it cannot readily boil. However, an efficient PWR seeks to maximise the heat transferred from the nuclear fuel rods to the pressurised water, and paradoxically, localised boiling in the pressurised water can be a key energy-transfer mechanism. In particular, as the nuclear fission rate in the reactor increases, the temperature of the fuel rods increases, thereby increasing the pressurised water temperature. On the fuel rod external surfaces, the temperatures may be high enough to induce nucleate boiling. This is characterised by the formation of small steam bubbles on the fuel rod surface which then recondense promptly in the surrounding water. It is known that as the nucleation of bubbles increases so does the heat transfer coefficient (HTC), a measurement of the heat transfer efficiency between the fuel rod surface and the surrounding coolant. Thus, the HTC can be increased by improving the amount of nucleate boiling on the fuel rod surface. Increasing the bubble formation too much, however, can detrimentally result in film boiling in which bubbles coalesce, forming a layer of steam on the rod surface. This layer insulates the surface, significantly reducing the HTC and causing the surface temperature to rise dramatically, risking system failure or material degradation. The point at which the transition to film boiling begins is termed the critical heat flux (CHF). Ideally, for enhanced heat transfer between the fuel rod surface and the pressurised water, simultaneous improvements to both the HTC and CHF are required. It is also desirable to improve the CHF margin so that the reactor has improved safety characteristics in relation to faulted scenarios such as main steam line break, reactivity faults and loss of primary coolant flow. Unfortunately, attempts to increase the nucleation-site density to improve the HTC can promote the onset of bubble coalescence and film boiling. Figure 2 is a schematic diagram of a BWR 50. A reactor pressure vessel 53 contains a core 52 surrounded by a core shroud 56, a steam-water separator 54 and a steam dryer 55. A plurality of internal impeller or jet pumps 57 are installed at the bottom of the pressure vessel in the downcomer annulus formed between the pressure vessel and the core shroud. A main steam pipe 59 supplies the steam from the steam dryer to a steam turbine 58 for generation of electricity by a generator 61. The steam discharged from the turbine is condensed in a condenser 62, and the condensed water is guided back into the reactor pressure vessel via a feed water pump 63 and a feed water pipe 60. In operation, water enters the downcomer annulus of the pressure vessel 53 from the feed water pipe 60. It is pressurized by the pumps 57 and supplied to the reactor core 52 for heating by its fuel assemblies. A part of the water becomes steam, and the two-phase flow is guided from the reactor core 52 to the steamwater separator 54 for steam separation. The moisture level in the separated steam is further reduced by the steam dryer 55. The water separated by the steam-water separator 4 is mixed in the downcomer annulus with the water supplied through the feed water pipe. The pressurised water supplied to the reactor core is maintained at about 7.6 MPa. Like a PWR, the pressurized water is both a coolant and moderator, but unlike a PWR it is used, in the form of steam, directly in the turbine 58. Control rods enter the core 52 through guides 64 extending from the underside of the pressure vessel 53 in order to achieve a more uniform power distribution as the effect of the increasing the proportion of steam in the two-phase flow in the upper part of the core is to reduce the amount of moderation. Similarly to a PWR, operational efficiency of a BWR is improved by maximising the heat transferred from the nuclear fuel rods of the fuel assemblies to the water. Thus in a BWR it also usual to induce nucleate boiling on the fuel rod external surfaces, although in this case the steam generated persists in the two-phase flow. It is also generally desirable to avoid film boiling. Accordingly, in a BWR, just as in a PWR, simultaneous improvements to both the HTC and CHF are required, and it is desirable to improve the CHF margin. The present invention has been conceived in view of the above considerations. Summary of the Invention The present invention is at least partly based on a recognition that a successful approach to increasing both HTC and CHF should take account of bubble dynamics and water wetting at the external surfaces of the vertically extending fuel rods, as well as water pressure and temperature. Thus, in a first aspect, the present invention provides a nuclear fuel rod according to claim 1. Advantageously, the textured pattern provides hierarchical surface structures which promote nucleation and wetting on the vertical surfaces under the high temperature and pressure flow conditions of PWRs or BWRs. In particular, providing a fine scale pattern of the first and second areas makes it possible to increase both the HTC and the CHF, which are typically otherwise antagonistic properties. Increasing both these values in turn can enable an increased core duty, and thus a greater net-power output from the reactor. The textured pattern also allows the CHF to be controlled with a greater degree of certainty, allowing narrower safety margins in the reactor in relation to faulted conditions. The textured pattern is formed by ultrafast laser ablation. The term “ultrafast” typically refers to lasers with a pulse length of <10 ps. In general, the pulse length in ultrafast laser ablation is shorter than the electron-cooling and the lattice-heating time of the incident material, i.e. zirconium alloy. Ultrafast laser ablation provides a means to perform highly localised material removal from the fuel rod surface, without significant impact on the surrounding material. Using a picosecond or femtosecond pulsed laser, ultrafast laser ablation removes material via non-thermal ablation mechanisms. In contrast, the use non-ultrafast (e.g. nanosecond) pulsed lasers typically results in the formation of a melt zone around the ablated region, as well as stress fractures into the material and other imperfections. A major advantage of ultrafast laser ablation is that it is effective at machining materials with high resistance to thermal degradation, such as fuel rod cladding, which are typically highly corrosion-resistant and thermally stable and thus difficult to machine with more conventional laser technologies or by other micro-texturing methods such as photolithography. The external surface of the tube is formed by a zirconium alloy. Such alloys have high resistance to corrosion, low absorption cross-section of thermal neutrons, and good mechanical properties at the relatively high temperature and pressure conditions of a PWR or BWR reactor. The type of zirconium alloy is not limited and may be chosen to suit the specific reactor conditions. Typical alloys include Zircaloy-2, Zircaloy-4, ZIRLOTM, Optimised ZIRLOTM, M5TM and HANA-6TM. However, particularly when the fuel rod is a next-generation Accident Tolerant Fuel (ATF), the external surface may be formed of a different material, e.g. a cladding on a zirconium alloy substrate. Such a cladding may be formed, for example, of chromium or of ferritic FeCrAl alloy. Under normal power operation of a PWR or BWR, the reactor is critical and produces more than 5% of nominal power. In this mode of operation, in a PWR the external surface of the tube at the inlet to the reactor core may be in contact with pressurised water at an operating temperature of 250°C or higher, e.g. about 300°C, and an operating pressure of 14 MPa or higher, e.g. about 15.5 MPa. In a BWR, the external tube surface at the inlet to the reactor core may be maintained at a similar operating temperature but in contact with pressurised water an operating pressure of 7 MPa or higher, e.g. about 7.6 MPa. The textured pattern may be arranged such that no point in each first area is more than 2 mm from one of the second areas, or more preferably, more than 1 mm from one of the second areas. Arranging the first and second areas in this way helps to inhibit film boiling, as limiting the maximum distance a nucleated bubble in a first area can be from a second area helps to limit bubble coalescence in the first area. The nucleation structures of each first area is formed by an arrangement of protuberances or spikes extending perpendicularly to the external surface. Typically, the formation of these protuberances results in the protuberances being combined with cavities of a similar scale. The shape and size of the protuberances may be chosen to improve nucleation under forced convection, high pressure, high temperature and vertical surface conditions. Each protuberance, on a transverse cross-section relative to its direction of extension, may have an area of 100 pm2 or less, 50 pm2 or less, 20 pm2 or less, or 10 pm2 or less, the smallest cross-sectional area achievable being limited by the resolution of the process used to form the textured pattern. In general, the smaller cross-section for each protuberance, the higher the nucleation-site density, and therefore the greater the potential HTC improvement. The protuberances may have a surface roughness, Rz (characterising maximum peak-to-valley height), of 150 pm or less, 100 pm or less, or 50 pm or less, wherein Rz, defined according to ISO 21920, is measured by interferometry. The protuberances may have a surface roughness, Rz, of 5 pm or more, or 10 pm or more. Changing the surface roughness can modulate the nucleation-site density. The arrangement of protuberances may be an irregular arrangement, such that there is no long-range order within the first area or, alternatively, may be ordered in a regular array. Providing an ordered arrangement may aid in the countable modulation of nucleation-sites, whilst an irregular arrangement may be quicker to manufacture and produce a higher nucleation-site density. The first areas of the textured pattern may form isolated islands surrounded by the second areas. This helps to limit the maximum distance a nucleated bubble in a first area has to travel to arrive at a second area, and thus helps to limit bubble coalescence in the first area and increases the CHF by delaying the departure form nucleate boiling (DNB) point. The first areas may have a nearest-neighbour spacing from each other of 0.1 mm or more, 0.2 mm or more, 0.5 mm or more, or 1 mm or more. The first areas may have a nearest-neighbour spacing of 5 mm or less, 4 mm or less, 3 mm or less, or 2 mm or less. The second areas surrounding the first areas help to draw the pressurised water across the tube surface, supplying water to the first areas and inhibiting the formation of an insulating steam layer. The wetting structures of each second area may be formed by an arrangement of depressions formed in the external surface. These structures may be designed to enhance the effect of capillarity. For example, the arrangement of depressions may be an arrangement of parallel channels extending over the external surface. There may be more than one set of parallel channels, each set of parallel channels extending in a respective direction and overlapping with the other sets. For example, two overlapping sets of channels extending in orthogonal directions can produce a repeating square pattern covering the second areas. The channels draw water to the tube surface and may also be used to direct its flow, thereby directing movement of small bubbles away from their origins in the first areas. The depressions may have a nearest-neighbour interval spacing of 80pm or less, or 40pm or less. The depressions may have a nearest-neighbour spacing of 1 pm or more, 2 pm or more, or 5 pm or more. In a second aspect, the present invention provides a method of producing the nuclear fuel rod according to the first aspect, the method including: providing the corrosion-resistant tube; and using ultrafast laser ablation to form the textured pattern on the external surface of the tube. The method may additionally include filling the tube with nuclear fuel pellets. The ultrafast laser ablation may be performed with a 4-axis laser ablation apparatus. This facilitates the patterning of tubes by introducing tube rotational motion as well 3-axis laser movement. In a third aspect, the present invention provides a method of operating a pressurised water reactor having a core within which nuclear fuel rods according to the first aspect extend vertically, the nuclear fuel rods containing nuclear fuel pellets and the external surfaces of the tubes being in contact with pressurised water primary coolant, the method including: circulating the pressurised water in a primary coolant loop to attain a pressure of 14 MPa or more, while heating the pressurised water to a temperature of 250°C or more at the inlet to the reactor core by heat generated by nuclear fission in the nuclear fuel pellets. The pressurised water used in the third aspect may be boron-free pressurised water. Advantageously, boron-free pressurised water allows the core duty of the reactor to be increased. Furthermore, without boric acid as a soluble shim, less waste is produced by the PWR. Burnable poison or more control rods may be used in place of a soluble shim. In a fourth aspect, the present invention provides a method of operating a boiling water reactor having a core within which nuclear fuel rods according to the first aspect extend vertically, the nuclear fuel rods containing nuclear fuel pellets and the external surfaces of the tubes being in contact with pressurised water coolant, the method including: circulating the pressurised water in a coolant loop to attain a pressure of 7 MPa or more, while heating the pressurised water to a temperature of 250°C or more at the inlet to the reactor core by heat generated by nuclear fission in the nuclear fuel pellets. The invention includes the combination of the aspects and preferred features described except where such a combination is clearly impermissible or expressly avoided. Summary of the Figures Embodiments and experiments illustrating the principles of the invention will now be discussed with reference to the accompanying figures in which: Figure 1 shows a schematic diagram of a pressurised water reactor; Figure 2 shows a schematic diagram of a boiling water reactor; Figure 3 shows schematically a coolant flow through a reactor pressure vessel; Figure 4 shows schematically a boiling curve for water; and Figure 5 shows schematically an example of a textured pattern on a fuel rod surface comprising isolated islands containing nucleation-promoting structures surrounded by regions containing structures promoting wetting. Detailed Description of the Invention Aspects and embodiments of the present invention will now be discussed with reference to the accompanying figures. Further aspects and embodiments will be apparent to those skilled in the art. All documents mentioned in this text are incorporated herein by reference. Pressurised water flow Figure 3 schematically shows the flow path of the coolant flow through a lower portion of an RPV 2 of a PWR. Within the RPV 2, the flow inlet to the reactor core 5 is typically located at a lower core support plate 9 of a core barrel 6, the lower core support plate being spaced from a bottom wall 4 of the RPV 2 by a lower plenum volume 8. Coolant flow enters the lower plenum volume downwardly from an annular downcomer outlet 7. It then changes direction in the lower plenum volume to flow upwardly through the lower core support plate 9, where it accepts heat from and moderates the fuel assemblies 10 of the reactor core. Each fuel assembly comprises a plurality of vertically extending nuclear fuel rods, each rod being a corrosion-resistant tube containing nuclear fuel pellets. The flow path of the coolant flow in the lower portion of a BWR RPV is similar. The coolant water initially travels downwards in the RPV through the downcomer annulus. It is then pressurised by the internal pumps and turns upwards to pass through a core plate on which the fuel assemblies are mounted. The water accepts heat from and moderates the vertically extending nuclear fuel rods of the assemblies as it passes over them. Heat transfer and boiling Heat transfer mechanisms between the external surface of these tubes and the pressurised water depend on multiple factors and can form different regimes depending on the dominant mechanism. For example, a typical water boiling curve for a surface, such as shown schematically in Figure 4 plotting heating flux from a surface against excess temperature (i.e. the difference between the surface temperature and the boiling point of the water), can display the following four regimes as the surface temperature rises: • Stage 1: Natural Convection Initially, the heated surface simply heats the surrounding water, causing its temperature to rise relative to water further from the surface. In the absence of forced convection, the water near the surface would become hotter and less dense, and therefore rise, while cooler, denser water further from the surface would sink, creating natural convection currents. However, in the reactor core 5, the upward coolant flow over the vertically extending nuclear fuel rods tends to override natural convection. • Stage 2: Nucleate Boiling As the surface temperature increases, small steam bubbles begin to form at nucleation-sites on the surface. This is nucleate boiling. The small bubbles quickly form, rise and in a PWR collapse by condensation in the surrounding water. In a BWR, by contrast, the bubbles can coalesce away from the surface into larger bubbles, which eventually evolve into slugs and even an annular flow regime. In the nucleate boiling stage heat transfer becomes more efficient (i.e. the HTC rises) as bubbles form and rise, entraining the surrounding water and removing heat from the surface. With further surface temperature increases, more vigorous nucleate boiling occurs. Eventually, the CHF is reached. This is a peak point of heat transfer beyond which heat transfer efficiency starts to drop. • Stage 3: Transition Boiling As the surface temperature increases still further, boiling becomes increasingly unstable. Bubbles form more rapidly and coalesce adjacent the surface, initiating film boiling in localised areas in which unstable and transient steam layers form on the surface. This leads to a reduction in the overall efficiency of heat transfer. • Stage 4: Film Boiling Eventually, a continuous steam layer forms between the surface and the water, which is thus no longer in direct contact with the heat source. As heat must be transferred through this insulating steam layer, heat transfer is less efficient. Nonetheless, after the transition to film boiling the heat flux increases as the surface temperature increases. In a PWR or a BWR, it is desirable to operate in the nucleate boiling region where heat transfer from the fuel rods is efficient. In particular, it is desirable to operate in the region close to the CHF, as this is the most effective region for heat transfer. However, operating close to the CHF increases the potential for film boiling. This should be avoided as it can have deleterious effects on reactor performance. There is therefore a desire to increase and better control both the CHF and HTC to improve reactor efficiency and / or safety margins. Textured patterning of the fuel rod surface The present invention relates to a textured pattern 100 for a nuclear fuel rod surface that simultaneously promotes nucleation and wetting. It can thereby be optimised to increase / control HTC, as well as other key boiling characteristics such as CHF. An example of the textured pattern 100 on the surface of a nuclear fuel rod is illustrated schematically in Figure 5. The textured pattern 100 comprises first areas 101 that promote the nucleation of steam bubbles and second areas 102 that promote the wetting of the pressurised water to the surface of the fuel rod. Mechanistically, the first areas 101 comprise structures to encourage the formation of steam bubbles on the fuel rod surface and thus encourage nucleate boiling. The nucleate boiling resulting from the first areas 101 increases the HTC due to the efficient energy transport associated with bubble formation, growth and release. Bubble agitation attributed to the turbulence created as the bubble rises and collapses, results in liquid mixing. The steam bubbles are then drawn away from their originating first areas 101 by the wettability of the second areas 102. To help avoid bubble coalescence adjacent the surface and prevent a transition to localised film boiling, the first areas 101 are provided as a regular array of isolated islands surrounded by the second areas 102. This arrangement helps to ensure that water remains on the surface of the fuel rod between the first areas 101, thereby inhibiting the formation of a steam layer and, in a PWR, encouraging re-condensation of the detached steam bubbles. For any bubbles that do originate in the second areas 102, the enhanced wettability of these areas also reduces the contact area between the bubbles and the surface, as well as the bubble departure diameter. The volumes of these bubbles are therefore lower than those produced in the first areas 101, further reducing the chance for bubble coalescence adjacent the surface. Distinct areas for wetting and nucleation give greater control of the bubble dynamics and flow of water over the external vertical surface of the tube. In particular, adjustment of the shape, size and spacing of the first 101 and second 102 areas can influence the overall performance of the pattern 100 via modulation of the CHF, HTC, and other key boiling characteristics. In particular, these characteristics can be optimised for the normal operating point of a PWR (e.g. a temperature of about 300°C at the inlet to the reactor core, and a reactor operating pressure of about 15.5 MPa) or the normal operating point of a BWR (e.g. a temperature of about 300°C at the inlet to the reactor core, and a reactor operating pressure of about 7.6 MPa). For example, the nearest neighbour spacing, d, of the first areas 101 measures the shortest distance between a point on the edge of a first area 101 to the closest point on the edge of an adjacent first area 101. In general, an increased spacing d correlates with an improved CHF, whilst a smaller spacing correlates with an improved HTC. The nearest neighbour spacing may be 0.1 mm or more, 0.2 mm or more, 0.5 mm or more, or 1 mm or more. The nearest neighbour spacing may be 5 mm or less, 4 mm or less, 3 mm or less, or 2 mm or less. A pattern having a spacing of less than 0.1 mm may have too small a total surface area of enhanced surface wetting. Conversely, a pattern having a spacing of more than 5 mm may have too few nucleation-sites. To help prevent bubble coalescence near the surface and the onset of film boiling, no point in each first area 101 may be more than 2 mm from one of the second areas 102 and preferably more than 1 mm from one of the second areas 102. In the example of Figure 4, the first areas 101 are formed as squares, and thus to comply with this condition may have sides of length of no more than 4 mm, and preferably no more than 2 mm. In the example of Figure 5, the nucleation structures of each first area 101 may be formed by a regular or irregular arrangement of columnar protuberances (i.e. spikes) extending perpendicularly to the fuel rod surface, and associated cavities. These structures may conveniently be formed by ultrafast laser ablation (discussed in more detail below), and their average cross-sectional area can influence the HTC. For example, the average cross-sectional area may be 100 pm2 or less, 50 pm2 or less, or 20 pm2 or less. In general, a smaller cross-section provides a higher nucleation-site density, and therefore a higher potential HTC improvement. The maximum peak-to-valley height surface roughness (Rz) of the protruberances may be 150 pm or less, or 100 pm or less, and / or may be 5 pm or more, or 10 pm or more. Preferably Rz is defined according to ISO 21920 and measured by interferometry. The wetting structures may be provided as a regular arrangement of depressions on the fuel rod surface. The depressions may have a nearest neighbour interval spacing of 40 pm or less, or 20 pm or less, and / or 1 pm or more, 2 pm or more, or 5 pm or more. For example, the depressions preferably take the form of parallel channels or similar structures to encourage capillarity and thereby draw water across the fuel rod surface. The channels preferably run vertically up the external surface of the rods, i.e. parallel to the direction of fluid flow through the core. By tuning the characterising parameters of the first 101 and second 102 areas, such as those discussed above, the areas may be configured to produce a suitable CHF and a suitable HTC at the operating conditions of the RPV. For example, the ratio of total area of nucleation promoting first areas versus total area of increased wettability second areas can be optimised depending on PWR or BWR application. Specifically, in a BWR application the ratio is likely to be higher than in a PWR application as a higher nucleation rate is compatible with the desire in a BWR for the nucleate bubbles to coalesce (away from the surface) into larger bubbles. The textured pattern 100 of Figure 5 on the fuel rods can conveniently be produced by ultrafast laser ablation, e.g. using 4-axis equipment allowing highly flexible relative movement of the laser about a rotatable fuel rod surface. A suitable laser for performing the ablation is the Tangor 300TM from Amplitude SA. With such equipment it is possible to machine the pattern across each fuel rod in approximately 5 to 15 minutes. The tubes are typically formed from zirconium alloy for its corrosion resistance, low neutron absorption and good mechanical properties. However, these properties render surface modification by conventional methods, such as photolithography and etching, difficult to perform. Ultrafast laser ablation, by contrast, has the following attributes, which make it suitable for producing the textured pattern: • Removal of material with little or no impact on surrounding areas due to an ablation mechanism which rapidly removes material in the pathway of the laser without forming significant surrounding heat-affected zones. • High resolution and consistency, allowing reproducible, precision micro-processing of micron and sub-micron scale structures. • Flexibility of texturing, enabling iterative design optimisation of hierarchical surface structures. • Possibility to machine both the first 101 and second 102 areas with a single laser and optical hardware setup. • Higher speed of machining than other precision surface texturing methods such as photolithography and deep-reactive etching. *** The features disclosed in the foregoing description, or in the following claims, or in the accompanying drawings, expressed in their specific forms or in terms of a means for performing the disclosed function, or a method or process for obtaining the disclosed results, as appropriate, may, separately, or in any combination of such features, be utilised for realising the invention in diverse forms thereof. While the invention has been described in conjunction with the exemplary embodiments described above, many equivalent modifications and variations will be apparent to those skilled in the art when given this disclosure. Accordingly, the exemplary embodiments of the invention set forth above are considered to be illustrative and not limiting. Various changes to the described embodiments may be made without departing from the spirit and scope of the invention. For the avoidance of any doubt, any theoretical explanations provided herein are provided for the purposes of improving the understanding of a reader. The inventors do not wish to be bound by any of these theoretical explanations. Any section headings used herein are for organizational purposes only and are not to be construed as limiting the subject matter described. Throughout this specification, including the claims which follow, unless the context requires otherwise, the word “comprise” and “include”, and variations such as “comprises”, “comprising”, and “including” will be understood to imply the inclusion of a stated integer or step or group of integers or steps but not the exclusion of any other integer or step or group of integers or steps. It must be noted that, as used in the specification and the appended claims, the singular forms “a,” “an,” and “the” include plural referents unless the context clearly dictates otherwise. Ranges may be expressed herein as from “about” one particular value, and / or to “about” another particular value. When such a range is expressed, another embodiment includes from the one particular value and / or to the other particular value. Similarly, when values are expressed as approximations, by the use of the antecedent “about,” it will be understood that the particular value forms another embodiment. The term “about” in relation to a 5 numerical value is optional and means for example + / - 10%.

Claims

1. A nuclear fuel rod for a pressurised or boiling water reactor, the rod having a corrosion-resistant tube which forms a container for nuclear fuel pellets, the rod being configured to extend, in use, vertically within a core of the reactor with the external surface of its tube being in contact with pressurised water primary coolant, wherein:the external surface of the tube has a textured pattern (100) comprising first areas (101) containing nucleation structures and second areas (102) containing wetting structures, the second areas being different from the first areas; andthe nucleation structures, relative to the wetting structures, are configured to promote steam nucleation from the pressurised water on the external surface, whereas the wetting structures, relative to the nucleation structures, are configured to promote wetting of the external surface by the pressurised water.

2. The nuclear fuel rod of claim 1, wherein the textured pattern is formed by ultrafast laser ablation.

3. The nuclear fuel rod of claim 1 or 2, wherein the pressurised water is at a temperature of 250°C or higher at the inlet to the reactor core.

4. The nuclear fuel rod of any one of the previous claims, wherein the nucleation structures of each first area are formed by an arrangement of protuberances extending perpendicularly to the external surface.

5. The nuclear fuel rod of claim 4, wherein each protuberance, on a transverse cross-section relative to its direction of extension, has an area of 100 pm2 or less.

6. The nuclear fuel rod of claim 4, wherein the protuberances have a surface roughness, Rz, characterising maximum peak-to-valley height of 150 pm or less and 5 pm or more, wherein Rz is defined according to ISO 21920 and measured by interferometry.

7. The nuclear fuel rod of any one of the previous claims, wherein the first areas form isolated islands surrounded by the second areas.

8. The nuclear fuel rod of claim 7, wherein the first areas have a nearest-neighbour spacing (d) from each other of 0.1 mm or more and 5 mm or less.

9. The nuclear fuel rod of any one of the previous claims, wherein the wetting structures of each second area are formed by an arrangement of depressions formed in the external surface.

10. The nuclear fuel rod of claim 9, wherein the arrangement of depressions is an arrangement of parallel channels extending over the external surface.

11. The nuclear fuel rod of claim 9 or 10, wherein the depressions have a nearest-neighbour interval spacing of 40 pm or less.

12. A method of producing the nuclear fuel rod of any one of the previous claims, the method including: providing the corrosion-resistant tube; and using ultrafast laser ablation to form the textured pattern on the external surface of the tube.

13. A method of operating a pressurised water reactor having a core within which nuclear fuel rods according to any of claims 1 to 11 extend vertically, the nuclear fuel rods containing nuclear fuel pellets and the external surfaces of the tubes being in contact with pressurised water primary coolant, the method including:circulating the pressurised water in a primary coolant loop at a pressure of 14 MPa or more, while heating the pressurised water to a temperature of 250°C or more by heat generated by nuclear fission in the nuclear fuel pellets.

14. A method of operating a boiling water reactor having a core within which nuclear fuel rods according to the any of claims 1 to 11 extend vertically, the nuclear fuel rods containing nuclear fuel pellets and the external surfaces of the tubes being in contact with pressurised water coolant, the method including: circulating the pressurised water in a coolant loop to attain a pressure of 7 MPa or more, while heating the pressurised water to a temperature of 250°C or more at the inlet to the reactor core by heat generated by nuclear fission in the nuclear fuel pellets.Amendments to the claims has been filed as follows:Claims:

1. A nuclear fuel rod for a pressurised or boiling water reactor, the rod having a corrosion-resistant tube which forms a container for nuclear fuel pellets, the rod being configured to extend, in use, vertically within a core of the reactor with the external surface of its tube being in contact with pressurised water 5 primary coolant, wherein:the external surface of the tube is formed by ultrafast laser ablation incident on zirconium alloy and has a textured pattern (100), the external surface of the tube comprising first areas (101) containing nucleation structures formed by an arrangement of protuberances extending perpendicularly to the external surface and second areas (102) containing wetting structures, the second areas being different10 from the first areas; and the nucleation structures, relative to the wetting structures, are configuredto promote steam bubble nucleation from the pressurised water on the external surface, whereas the wetting structures, relative to the nucleation structures, are configured to promote wetting of the external surface by the pressurised water.

152. The nuclear fuel rod of claim 1, wherein each protuberance, on a transverse cross-section relative to its direction of extension, has an area of 100 pm2 or less.

3. The nuclear fuel rod of claim 1, wherein the protuberances have a surface roughness, Rz,20 characterising maximum peak-to-valley height of 150 pm or less and 5 pm or more, wherein Rz is defined according to ISO 21920 and measured by interferometry.

4. The nuclear fuel rod of any one of the previous claims, wherein the first areas form isolatedislands surrounded by the second areas.

255. The nuclear fuel rod of claim 4, wherein the first areas have a nearest-neighbour spacing (d)from each other of 0.1 mm or more and 5 mm or less.

6. The nuclear fuel rod of any one of the previous claims, wherein the wetting structures of eachsecond 30 area are formed by an arrangement of depressions formed in the external surface.

7. The nuclear fuel rod of claim 6, wherein the arrangement of depressions is an arrangement ofparallel channels extending over the external surface.35 8. The nuclear fuel rod of claim 6 or 7, wherein the depressions have a nearest-neighbour intervalspacing of 40 pm or less.

9. A method of producing the nuclear fuel rod of any one of the previous claims, the method including: providing the corrosion-resistant tube; and40 using ultrafast laser ablation to form the textured pattern on the external surface of the tube.

10. A method of operating a pressurised water reactor having a core within which nuclear fuel rods according to any of claims 1 to 8 extend vertically, the nuclear fuel rods containing nuclear fuel pellets and the external surfaces of the tubes being in contact with pressurised water primary coolant, the method including:5 circulating the pressurised water in a primary coolant loop at a pressure of 14 MPa or more, whileheating the pressurised water to a temperature of 250°C or more by heat generated by nuclear fission in the nuclear fuel pellets.

11. A method of operating a boiling water reactor having a core within which nuclear fuel rods according 10 to the any of claims 1 to 8 extend vertically, the nuclear fuel rods containing nuclear fuel pellets and the external surfaces of the tubes being in contact with pressurised water coolant, the method including:circulating the pressurised water in a coolant loop to attain a pressure of 7 MPa or more, while heating the pressurised water to a temperature of 250°C or more at the inlet to the reactor core by heat generated by nuclear fission in the nuclear fuel pellets.

Images