Abstract: A method of jacketing a uranium slug to an aluminum container comprising applying a coating to the exterior of the container, the coating consisting of colloidal graphite in water, permitting the coating to dry, applying an alloy of aluminum and silicon to the interior surface of the container at a temperature between 588° C. and 594° C., inserting the slug into the container in complete contact with the alloy, and quenching the assembly.

Abstract: A method of jacketing a body of fissionable material within a nonfissionable jacket having a cup-shaped body open at one end and a cap effecting closure at said end comprising the steps of inserting the cup-shaped body into a tightly fitting cup-shaped sleeve; completely submerging said assembly in a bonding bath of a molten metallic bonding material, allowing the inner body to fill with molten bonding material; dipping the fissionable body into the open end of said assembly while under the surface of the bonding bath; closing the open end of the assembly with the cap while under the surface of the bonding bath; removing the complete assembly from the bonding bath; quenching in cold water; and removing the sleeve.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith.

Abstract: The invention provides a method for making Zr alloy nuclear reactor fuel cladding having excellent corrosion resistance and creep properties. The method includes performing hot forging, solution heat treatment, hot extruding, and repeated cycles of annealing and cold rolling of a Zr alloy including, by weight, 0.2 to 1.7% Sn, 0.18 to 0.6% Fe, 0.07 to 0.4% Cr and 0.05 to 1.0% Nb, with the remainder being Zr and incidental impurities, and the incidental nitrogen impurity content being 60 ppm or less, and then performing final stress relief annealing thereon. The annealing is performed at a temperature of 550.degree. C. to 850.degree. C. for 1 to 4 hours such that the accumulated annealing parameter .SIGMA.Ai=.SIGMA.ti.multidot.exp(-40,000/Ti) satisfies relationships -20.ltoreq.log.SIGMA.Ai.ltoreq.-15, and -18-10.multidot.X.sub.Nb .ltoreq.log.SIGMA.Ai.ltoreq.-15-3.75.multidot.(X.sub.Nb -0.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A zirconium-based alloy with a reduced ahoy content is described that has resistance to both uniform and nodular corrosion comparable to present zirconium-based alloy compositions, such as Zircaloy-2. The alloy represents in essence a modified or diluted Zircaloy-2 or Zircaloy-4. The alloys of this invention are also expected to have improved uniform corrosion resistance at under high burn-up conditions The alloy comprises 0.05-0.09 weight percent of iron, 0.03-0.05 weight percent of chromium, 0.02-0.04 weight percent of nickel, 1.2-1.7 weight percent of tin and 0-0.15 weight percent oxygen, with a balance of zirconium. The iron chromium and nickel alloying elements form precipitates in the alloy matrix. The alloy is suitable for use as a cladding material for a fuel element housing fissionable nuclear materials in water cooled nuclear fission reactor.

Abstract: A method and an apparatus for producing microspherical ceramic particles such as particulate ceramic fuels for use in nuclear reactors. To produce gel particles by performing microwave heating on the small droplets of a stock solution or sol that contains uranium and other elements for nuclear reactor fuels, the small droplets are allowed to fall down through a cavity resonator using microwaves having an engineering frequency of 2.45 GHz and a quartz rod is inserted into or withdrawn out of the resonator to adjust the resonant frequency so that it will lie within the spectrum width of the microwave frequency.

Abstract: A method for mitigating crack initiation and propagation on the surface of metal components in a water-cooled nuclear reactor. An electrically insulating coating doped with a noble metal is applied on the surfaces of IGSCC-susceptible reactor components. The preferred electrically insulating material is yttria-stabilized zirconia doped with palladium or platinum. The presence of an electrically insulating coating on the surface of the metal components shifts the corrosion potential in the negative direction without the addition of hydrogen. Corrosion potentials .ltoreq.-0.5 V.sub.SHE are believed to be achievable even at high oxidant concentrations and in the absence of hydrogen, although the coatings are believed to be particularly suited to applications where a reductant, such as hydrogen, is present.

Abstract: The surface of the component (1, 2) to be protected is scanned, a jet of semitransferred arc plasma (17) into which a metal powder is introduced. The process can be applied to the production of a layer of coating (18) on the outer surface of a region of welding connection (3) between a nozzle (1) and a primary pipe (2) of a pressurized water nuclear reactor or to the coating of the inner surface or the connecting surface of an adapter passing through the reactor vessel head.

Abstract: Zircaloy 2 and zircaloy 4 are zirconium alloys which are permitted and tried and tested in nuclear engineering and which have constituents with fixed concentration ranges. The properties, especially corrosion resistance, mechanical stability and sensitivity to pellet-cladding interaction of those alloys are subject to pronounced spreads of unknown origin. According to the invention, the tin content is between 1.4 and 1.8% by weight, the Fe content between 0.1 and 0.25% by weight, the Cr content between 0.1 and 0.3% by weight, the Si content between 0.05 and 0.02% by weight, the O content between 0.05 and 0.11% by weight, the C content below 0.02% by weight and the Ni content below 0.08% by weight. This restriction of the permissible concentration ranges ensures that the material properties are spread only within a narrow favorable range. A liner made from zirconium with an iron constituent of between 0.2 and 0.8% by weight is proposed for the inner lining of a fuel-rod sheathing tube.

Abstract: A cladding tube having a cross-section and including (1) a zirconium alloy outer circumferential substrate having an inner surface and having one or more alloying elements, (2) a zirconium barrier layer bonded to the inner surface of the outer circumferential substrate and being alloyed with the one or more alloying elements, and (3) a zirconium alloy inner circumferential liner bonded to the inner surface of the zirconium barrier layer. The barrier layer will have a concentration profile including a diffusion layer extending from the barrier layer's inner surface (facing nuclear fuel) to the barrier layer's interior (between the barrier layer's inner and outer surfaces). At the interior edge of the diffusion layer, there will be substantially no alloying elements. At the outer edge of the diffusion layer (the barrier layer's inner surface), the maximum concentration of alloying elements will occur.

Abstract: A fuel rod has a cladding including a thicker inner layer and a thin outer layer being metallurgically bound thereto. In view of the conditions prevailing on the inside of the cladding tube and the mechanical properties of the entire cladding tube, the inner layer is formed of zircaloy having a comparatively high Sn content and a low Fe and Fe+Cr content. The outer layer also contains virtually only zircaloy constituents, but in view of corrosion, H2 take-up and sensitivity to Li dissolved in the cooling water, the Fe and Fe+Cr content is greater than or at most equal to that of the inner layer, the chosen Sn content is less than 1.3% and the chosen Sn+Fe+Cr content is more than 1.0%. Low failure rates of the cladding tube are thereby achieved even for long service lives.

Abstract: Zirconium or zirconium alloy components of a cylindrical shape are cleaned with an ice blasting process to produce a defect-free bondline in multi-layered tubing suitable for nuclear fuel cladding and the chemical processing industry. The ice blasting process improves the integrity of the metallurgical bond by removing foreign contamination that can initiate non-bonding defects.

Abstract: A method is provided for forming a three-layer cladding tube having an outer substrate, a zirconium barrier layer, and an inner liner having alloying elements, in which the zirconium barrier layer (located between an outer substrate and inner liner) is at least partially alloyed with alloying elements that impart resistance to corrosion. The barrier layer has a diffusion layer extending from its inner surface (facing the fuel) to the barrier layer's interior (the interior being defined between the barrier layer's inner and outer surfaces). At the interior edge of the diffusion layer, there will be substantially no alloying elements beyond those normally present in zirconium. The methods of forming such structure include a diffusion anneal of a three-layer cladding in the range of 650.degree.-1000.degree. C. for times between about 1 minute and 20 hours. This anneal drives some of the alloying elements from the inner liner into the zirconium barrier layer to form the diffusion layer.

Abstract: A method for mitigating crack initiation and propagation on the surface of metal components in a water-cooled nuclear reactor. An electrically insulating coating is applied on the surfaces of IGSCC-susceptible reactor components. The preferred electrically insulating material is yttria-stabilized zirconia. The presence of an electrically insulating coating on the surface of the metal components shifts the corrosion potential in the negative direction without the addition of hydrogen and in the absence of noble metal catalyst. Corrosion potentials .ltoreq.-0.5 V.sub.she can be achieved even at high oxidant concentrations and in the absence of hydrogen.

Abstract: A cladding is provided for use in housing fissionable material in water cooled nuclear fission reactors. The cladding has inner and outer surfaces and includes (1) a cross-section of a Zirconium-based alloyed matrix, and (2) alloying elements in sufficient concentration to form precipitates disposed in the matrix. The cladding includes no more than 20 parts per million nitrogen by weight and is typically a modified Zircaloy-2 or Zircaloy-4. Metallurgically bonded to the inner region of the cladding is a zirconium barrier layer.

Abstract: A neutron absorbent including a composite material is provided. The composite material is obtained by impregnating a carbon material with a boron oxide and/or a hydrate compound thereof, and baking such carbon material under pressure by inert gas at a temperature of not lower than 1500.degree. C.

Abstract: A cladding tube is provided having an axis and a cross-section perpendicular to the axis. The cross-section includes (1) an outer circumferential substrate having an inner surface, (2) a zirconium barrier layer bonded to the inner surface of the outer circumferential substrate, and (3) an inner circumferential liner bonded to the inner surface of the zirconium barrier layer. The inner circumferential liner includes a plurality of facets aligned substantially in parallel with the cladding tube axis. The facets--which define the geometry of the cladding interior--facilitate mixing among the gases in the cladding interior. Also provided is a fuel element including (1) a cladding tube having an faceted inner liner as described above, (2) nuclear fuel material disposed within said cladding tube in a fuel region, and (3) a plenum or plenums located in one or both end regions (beyond the nuclear fuel region) within the cladding tube.

Abstract: A wear resistant coating (50) for a component of a nuclear reactor fuel assembly or control assembly comprising a ceramic material (52) which is premixed with a glass (54). The coating is applied to components of the nuclear reactor vessel, or portions thereof, which are subject to the greatest wear due to friction or fretting. Also disclosed herein is a method for enhancing the wear resistance of a portion of a metal component of a nuclear reactor by coating a surface of the component with a wear-resistant coating formed from diamond, metal nitride, or a composite of a ceramic material and a glass.

Abstract: For the purpose of preventing reductions of wall thickness due to erosion-corrosion in parts (2, 3, 7) of a system made of carbon steel, and which constitute a wet steam system, a feedwater and condensate system, and a drain system of a thermal or nuclear power plant, a coating (17) of metal or ceramic which is chemically stable against a fluid flowing in the systems is formed by spraying on the surface of the devices which are exposed to the fluid.

Abstract: A fuel rod for a nuclear reactor is described. The fuel rod contains a zirconium alloy cladding tube, with a thin coating of a burnable poison which has been plasmaarc sprayed on the inside diameter of the cladding tube. Also described is a plasma-arc spraying device and a method for applying a coating to the inside of a small-diameter metal tube.

Abstract: A method for coating the inside surface (18) of tubular components (16) of a nuclear fuel assembly, including the steps of supporting the component within a vacuum chamber (12), positioning a source rod (20) having a field emitter structure (22) within the component, the structure being formed of material to be coated on the surface, and inducing an electrical current flow (26) through the rod sufficient to evaporate at least a portion of the emitter structure, whereby the evaporated material of the emitter structure is deposited on and adheres to the surface as a coating. Optionally, the vacuum chamber is backfilled with a reactive gas (38), and the material evaporated from the emitter structure chemically reacts with the gas before adhering to the surface. The reactive gas can be one of nitrogen, oxygen, or carbon plasma and the coating adhered to the component can be one of a nitride, oxide, or carbide, respectively.

Abstract: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing on a weight percentage basis tin in a range of 0.4 to 1.0 percent and typically 0.5; iron in a range of 0.3 to 0.6 percent, and typically 0.46 percent; chromium in a range of 0.2 to 0.4 percent, and typically 0.23 percent; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 1200 to 2500 ppm, typically 1800 to 2200 ppm. The high oxygen level assists in reducing hydrogen uptake of the alloy compared to Zircaloy-4, for example.

Abstract: A method for enhancing the wear and corrosion resistance of a tubular nuclear fuel assembly component (40), comprising the step of coating the component with a corrosion and wear resistant material by an anodic arc plasma deposition process (70). The coating is preferably a nitride reactively formed during the plasma deposition process. The component is preferably a nuclear fuel rod cladding tube and the coating material is one of ZrN or TiN.

Abstract: A wear resistant coating (50) for fuel rod cladding (20) comprises a ceramic material (52) which is premixed with a glass (54). The cladding tube is heated and the ceramic/glass mixture is flame sprayed onto the cladding tube. The coating is applied to lower portions of the fuel rods (10) in the area of the first support grid (12) where debris tends to fret the fuel rod.

Abstract: A method for enhancing the wear resistance of a tubular component (62) of a nuclear fuel assembly, including the first steps of supporting the component in an implantation chamber (16), removing ambient air from the chamber such as by a pump (56) and generating a plasma plume (28) of positively charged metal source material by establishing an electrical discharge arc which travels from a cathode (18) of said source material to an anode (24) of a different material. At least a portion of the plasma plume is passed through an electromagnetic duct (14) which filters constituents other than free, high energy source material ions out of the plume. The high energy source material positive ions are directed through the chamber onto the negatively charged component. The chamber can be backfilled with a reactive gas such as nitrogen which forms an ionic compound with the source material ions such as Zr or Ti, and the ionic compound such as ZrN or TiN implants in the component.

Abstract: Methods and apparatus for improving fretting resistance of zirconium alloy components formed into a shape for use in a nuclear reactor are disclosed in which at least a portion of the outer surface of a component is reacted with material selected from the group consisting of carbon, nitrogen, oxygen and combinations of the foregoing at a temperature below about 700.degree. C. to form a wear resistant layer on the surface of the component.

Abstract: A fuel element (10) for a nuclear reactor having a zirconium-tin alloy cladding tube (20), with a thin coating (30) of a burnable poison consisting of an enriched boron particles sealed in a boron-containing glass or glass compound deposited from a liquid sol-gel.

Abstract: A corrosion resistant metallic coating (60) of zirconium nitride is applied to the cladding tube (40) of a nuclear fuel rod (20). The zirconium nitride is reactively deposited on a zirconium-alloy cladding tube by a cathodic arc plasma deposition process. The zirconium nitride coating provides superior wear test results and enhances the corrosion resistance of the cladding tube.

Abstract: A fuel assembly for a nuclear reactor comprising a fuel cladding tube of three-layer structure having an outer surface in contact with reactor water of the nuclear reactor, an inner surface layer in contact with the nuclear fuel, and an intermediate layer interposed between the outer surface layer and the inner surface layer. the outer surface layer is made of a Zr-based alloy containing Nb, Sn and Mo. The inner surface layer is made of pure zirconium. The intermediate layer is made of a high ductility alloy which is higher in ductility than the outer surface layer and is higher in strength than the inner surface layer.

Abstract: Nuclear fuel compacts, containing thousands of individually coated fuel particles in a carbonaceous matrix, are provided with an overcoating of silicon carbide which can be readily inspected for defects. This overcoating helps retain fission products within the compact should any of the fuel particles have damaged or defective coatings, and provides an outer secondary barrier that can be 100% inspected.

Abstract: A fuel rod for a water cooled nuclear reactor comprises a cladding made of a continuous wound ceramic fiber tube. The wound fibers are impregnated with a matrix to achieve a required density and exposed to ammonia gas to create a solid gel. The gel is fired to form a ceramic composite having desirable thermal properties, strength, corrosion resistance, toughness and neutron capture cross-section.

Abstract: Particulates from a source fall into a tundish which discharges to a collection chamber. The build-up of particulates in the chamber is sensed, and a valve opened at the bottom of the chamber when a pre-determined level of the particulates is sensed in the chamber. A valve at the top of the chamber may be closed at the same time so that pressurized air can be injected into the chamber to expel the particulates through the bottom of the chamber through the valve. A filter circuit may be connected to the tundish to assist in removing particulates from the tundish.

Abstract: A wear resistant coating (50) for fuel rod cladding (20) comprises a ceramic material (52) which is premixed with a glass (54). The cladding tube is heated and the ceramic/glass mixture is flame sprayed onto the cladding tube. The coating is applied to lower portions of the fuel rods (10) in the area of the first support grid (12) where debris tends to fret the fuel rod.

Abstract: A nuclear reactor core has a first group of fuel rods containing fissionable material and no burnable absorber, and a second group of fuel rods containing fissionable material and two burnable absorber materials. The groups of fuel rods are arranged in the core for controlling power peaking and moderator temperature coefficient. The number of fuel rods in the first group are greater than the number in the second group. The two burnable absorber materials can be provided as separate coatings or a mixture. One burnable absorber material is an erbium-bearing material such as erbium oxide and the other is a boron-bearing material such as zirconium diboride. Alternatively, the erbium-bearing material can be interspersed or mixed with the fissionable material.

Abstract: An organometallic chromium compound in the gaseous phase is brought into contact with a substrate consisting of the inner surface of the zirconium alloy tubular cladding (5) of the fuel element, or the outer surface of the pellets of fuel material, the substrate being kept at a temperature between 300.degree. and 600.degree. C. The organo metallic compound may consist of chromium acetylacetonate. The process enables a chromium oxide coating to be obtained inside the tubular cladding (5) and/or on the outer surface of the nuclear fuel pellets. Pellet/cladding interaction is thus prevented or limited when this fuel element is used in the reactor.

Abstract: A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor.

Abstract: The present invention relates to the control of grain structure in unalloyed zirconium metal and, more particularly, to the control of grain structure in zirconium metals containing less than 300 parts per million Fe.

Abstract: Zirconium-based corrosion resistant alloys for use primarily as a cladding material for fuel rods in a boiling water nuclear reactor consist essentially of by weight percent about 0.5 to 2.0 percent tin, about 0.24 to 0.40 percent of a solute composed of copper, nickel and iron, wherein the copper is at least 0.05 percent, and the balance zirconium. Nuclear fuel elements for use in the core of a nuclear reactor have improved corrosion resistant cladding made from these zirconium alloys or composite claddings have a surface layer of the corrosion resistant zirconium alloys metallurgically bonded to the outside surface of a Zircaloy alloy tube. The claddings may contain an inner barrier layer of moderate purity zirconium metallurgically bonded on the inside surface of the cladding to provide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

Abstract: A nuclear fuel rod, in which a cladding tube contains fuel pellets therein and two plugs stop end openings of the cladding tube, in which a stopper member stops the fuel pellets from their moving through a spring member arranged between the stopper member and one of the plugs for biasing the stopper member onto the fuel pellets, and a getter tube is also arranged between the stopper member and the one of the plugs, and in which at least one of the spring member, the stopper member and the getter tube is made of a nonmagnetic steel.

Abstract: A nuclear power plant has a water or liquid sodium coolant and a metallic component having surfaces coming into contact with the coolant. A coating is disposed on the surfaces of the metallic component, such as fuel rod cladding tubes or spacer grids. The coating is formed of titanium carbide (TiC), titanium nitride (TiN), zirconium nitride (ZrN), chromium carbide (CrC), titanium aluminum vanadium nitride (TiAlVN), tantalum nitride (TaN), zirconium carbide (ZrC), or tungsten carbide (WC).

Abstract: A nuclear fuel element for use in the core of a nuclear reactor is disclosed having an improved corrosion resistant cladding. The cladding is comprised of zirconium alloys containing in weight percent 0.5 to 2.0 percent tin, or 0.5 to 2.5 percent bismuth, or 0.5 to 2.5 percent bismuth and tin, and about 0.5 to 1.0 percent of a solute composed of a member selected from the group consisting of molybdenum, niobium, tellurium and mixtures thereof, and the balance zirconium. Composite claddings are disclosed having a surface layer of one of the corrosion resistant zirconium alloys metallurgically bonded to a Zircaloy alloy tube. Claddings may contain an inner barrier layer of a moderate purity zirconium metallurgically bonded on the inside surface of the cladding to provide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

Abstract: Nuclear fuel particles having fission-product-retentive characteristics are disclosed which are particularly adapted for the production of nuclear fuel compacts or nuclear fuel elements by combination with a matrix material having an acceptable nuclear properties and good thermal conductivity. Preferably, the outermost shell of the fission-product-retentive layer is formed of a refractory carbide, such as silicon carbide or zirconium carbide, and a thin overcoating material is applied thereto. When aluminum is employed as the matrix material, an overcoating of elemental silicon or elemental zirconium is used which is wet by the aluminum matrix. Silicon also forms a low melting eutectic alloy with the aluminum of the matrix that provides good lubricating properties--a particularly valuable feature when high pressure extrusion is used to form fuel elements.

Abstract: Methods for making nuclear fuel compacts containing precise amounts of nuclear fuel material which exhibit low heavy metal contamination and fewer defective coatings following compact fabrication using a hardenable binder including petroleum pitch or the like. Nuclear fuel particles having a multiple layer fission-product-retentive barrier, with a dense outer layer thereof being surrounded by a protective overcoating, e.g., pyrocarbon having a density between about 1 and 1.2 g/cm.sup.3, that is encapsulated within a thin shell of pyrocarbon from about 1.7 to about 2 g/cm.sup.3 in density, can be precisely metered to create charges containing very precise amounts of nuclear fuel material that can be pre-compacted in molds under relatively high pressures and then combined with the fluid binder which is ultimately carbonized to produce carbonaceous compacts containing very precise nuclear fuel loadings.

Abstract: Nuclear fuel elements for use in the core of a nuclear reactor include an improved composite cladding having a zirconium barrier layer metallurgically bonded on the inside surface of a zirconium alloy tube, wherein the inside surface of the barrier is alloyed with preselected elemental impurities to improve oxidation resistance. The zirconium barrier layer forms a shield between the zirconium alloy tube and a core of nuclear fuel material enclosed in the composite cladding. The alloy layer formed on the barrier surface acts to inhibit cracking during the tube production fabrication step and limits oxidation in the event that the cladding is breached during operation of the reactor, allowing the entrance of water or steam into the fuel element.

Abstract: A nuclear fuel assembly comprising a plurality of fuel rods, each fuel rod comprising a cladding tube closed at both ends and filled with cylindrical nuclear fuel pellets such as sintered uranium dioxide or uranium plutonium dioxide. All or nearly all of the cylindrical pellets are coated with a refractory boride coating of sufficient thickness to prevent mechanical interaction between the pellets and the tube. The isotopic composition of the boron in the refractory boride is adjusted for each batch of fuel pellets, rods, or assemblage of rods so that the initial excess reactivity, k, of the batch, when assembled in a reactor core, is nearly zero.

Abstract: A process for forming a boron-containing coating on the internal surface of a zirconium or zirconium alloy hollow tube by heating the internal surface to a temperature of between 200.degree.-450.degree. C. and passing through the tube a mixture of a volatilized boron compound in helium or argon, such that the boron compound decomposes to form an integral boron containing coating on the internal surface.