Metal nuclear-fuel pin including a shell having threads or fibers made of silicon carbide (SiC)

Abstract

A nuclear-fuel pin including a linear element made of a metal nuclear-fuel material consisting of uranium and/or plutonium, and cladding including Fe and Cr or an alloy including at least both of said elements, comprises a main shell provided around the linear nuclear-fuel element, said shell including threads or fibers made of SiC. A method for producing a nuclear-fuel pin is also provided.

Claims

1. A nuclear fuel pin comprising a linear element defined by a bar or billet in a cylindrical shape of metal nuclear fuel material including uranium and/or plutonium and a cladding comprising Fe and Cr or an alloy comprising at least these two elements, further comprising: a main shell positioned around the linear element of metal nuclear fuel material, said main shell being disposed between an inside surface of the cladding and the linear element of metal nuclear fuel material, said main shell comprising yarns or fibers made of SiC; and a primary shell of silica or quartz fibers inserted between the linear element of metal nuclear fuel material and the main shell, the primary shell being disposed directly onto an outer surface of the cylinder of metal nuclear fuel material.

2. The nuclear fuel pin as claimed in claim 1, further comprising: a plenum to receive said discharge of fission gasses; and a reservoir, wherein the linear element is disposed at a first end of said nuclear fuel pin, the plenum is disposed at a second end of said nuclear fuel pin, and said reservoir is disposed between said plenum and said linear element, the cladding is configured to cover and contain said linear element, said reservoir, and said plenum.

3. The nuclear fuel pin as claimed in claim 2, wherein said reservoir comprises an annulus made of a material which is resistant to a corrosion of molten actinides.

4. The nuclear fuel pin as claimed in claim 3, wherein said annulus is made of tantalum.

5. The nuclear fuel pin as claimed in claim 1, wherein the SiC constituting the fibers is of cubic β allotropic variety.

6. The nuclear fuel pin as claimed in claim 1, wherein the main shell also comprises free Si fillers.

7. The nuclear fuel pin as claimed in claim 1, wherein said main shell comprises strips comprising SiC yarns or fibers.

8. The nuclear fuel pin as claimed in claim 1, wherein the main shell comprises a plurality of wrapped layers of SiC fibers wound around the linear element of metal nuclear fuel material.

9. The nuclear fuel pin as claimed in claim 8, wherein the primary shell comprises a plurality of wrapped layers of silica or quartz fibers wound around the linear element of metal nuclear fuel material.

10. A process for manufacturing a metal nuclear fuel pin as claimed in claim 1, further comprising the production of the main shell around the linear element of metal nuclear fuel material by weaving or braiding SiC fibers.

11. The process for manufacturing a metal nuclear fuel pin as claimed in claim 10, further comprising a prior step of surface oxidation of the linear element of metal nuclear fuel material.

12. The process for manufacturing a metal nuclear fuel pin as claimed in claim 10, further comprising a prior step of coating the linear element of metal nuclear fuel material with a binder comprising a soft brazing powder, the soft brazing powder including a conductive material.

13. The process for manufacturing a metal nuclear fuel pin as claimed in claim 10, further comprising the production of the primary shell between the linear element of metal nuclear fuel material and the main shell, said primary shell comprising silica or quartz fibers.

14. A process for manufacturing a metal nuclear fuel pin as claimed in claim 1, further comprising the production of the main shell with strips of SiC fibers wound around the linear element of metal nuclear fuel material.

15. The nuclear fuel pin as claimed in claim 1, wherein the primary shell comprises a plurality of wrapped layers of silica or quartz fibers wound around the linear element of metal nuclear fuel material.

Description

DETAILED DESCRIPTION

(1) Generally, and according to the present invention, the pin comprises, as illustrated in FIG. 1, a cladding Go made of stainless material based on Fe—Cr—Ni or Fe—Cr, a linear metal nuclear fuel element C.sub.Nu that may be present in the form of a bar or billet of cylindrical shape having a small diameter, typically 5 to 10 mm, it being possible for the fuel material to be of UPuZr or UPuX type with X possibly being, for example, molybdenum. The pin thus comprises a first portion comprising the fissile column of material C.sub.Nu and a second portion constituted by the plenum P.sub.Le for the gases.

(2) Advantageously, an annulus An or liner of tantalum Ta is provided constituting a reservoir that is resistant to the corrosion of molten actinides, under operational conditions.

(3) Advantageously, provision may be made to carry out a surface oxidation step, between 200 and 250° C., in air and for a few hours, in order to obtain a layer of a few micrometers of cubic MO.sub.2 type that adheres to its substrate, M being the constituent metal alloy of the metal nuclear fuel.

(4) For the manufacture of SiC fibers or yarns, various production processes may be envisaged and notably processes comprising steps of fiber weaving (simple and proven principle in the aeronautic field notably).

(5) Two exemplary embodiments of the main shell based on SiC fibers will be described below.

(6) First Exemplary Embodiment of a Pin According to the Invention:

(7) The SiC matrix that is in the form of yarns or fibers is woven or braided along the billet using a technical device well known to a person skilled in the art. The first turns, in contact with the billet, are advantageously produced with pre-oxidized fibers or with a weaving of quartz fibers or silica SiO.sub.2 fibers.

(8) It is also possible to use borosilicate glass if it is desired to have a reserve of neutron poison, determined by the boron in the glass. Indeed, for example, boron 10 (isotope 10 of boron) captures neutrons (it is a poison for the fission reactions), in order to be converted to boron 11 which gives He+Li. The isotope .sup.10B (natural boron is a mixture of the isotopes .sup.10B (19.8 mol %)+.sup.11B (80.2 mol %)), has, compared to other absorbents, an effectiveness over a very broad spectrum, from fast neutrons to thermal neutrons. The capture reaction, of (n, α) type, is given below: .sup.10B+1n=>7Li+4He+2.6 MeV.

(9) The thickness determined by the number of turns and the width of the overlap of each of the turns is adjustable data. The turns are loose (not tightened).

(10) A brazing powder (typically based on Ni to then improve the conductivity), with its liquid binder (which evaporates easily by drawing under vacuum), may advantageously be painted onto the billet before the operation for covering by weaving or braiding.

(11) The billet that is covered and optionally painted with braze is gradually introduced into the cladding, so that the cover thus fixed does not become slack.

(12) A schematic diagram illustrated in FIG. 2 demonstrates this example of a process for producing the main shell. A cladding Go is intended to receive a billet of nuclear fuel C.sub.Nu, around which yarns or fibers F.sub.SiC of SIC are wound and, not represented, quartz fibers or SiO.sub.2 fibers have been wound beforehand as a first thickness, after having been anchored in order to initiate the start of the weaving or braiding.

(13) In order to produce this main shell with the fibers F.sub.SiC, the fuel billet is rotated and a multilayer braiding machine comprising a flywheel and supports for braiding bobbins Bo.sub.FSiC is also rotatably mounted.

(14) Second Exemplary Embodiment of a Pin According to the Invention

(15) The SiC matrix in the form of woven fabric is draped along the billet using a pre-woven and not very dense strip based on standard SiC fibers. It is anchored for the covering operation at an anchorage point P.sub.A. The first layers, in contact with the billet, are advantageously produced with pre-oxidized strips or fibers or with a weaving of quartz fibers or silica SiO.sub.2 fibers or borosilicate glass fibers if it is desired to have a reservoir of neutron poison (determined by the boron in the glass). The first layers in contact with the preoxidized metal billet are constituted of SiO.sub.2.

(16) The thickness (number of layers or turns) and the width of overlap of each of the turns is adjustable data. The layers are loose (not tightened).

(17) It is optionally possible to fill the inter-turn space with pulverulent Si or very porous SiC foam, poured as a function of and on demand into the closure space of the strip undergoing the covering operation, during rotation.

(18) A brazing powder (typically based on Ni to then improve the conductivity), with its liquid binder (which evaporates easily by drawing under vacuum), may advantageously be painted onto the billet before the operation for covering by bondaging.

(19) The billet that is covered and optionally painted with braze is gradually introduced into the cladding, so that the cover thus fixed does not become slack.

(20) FIG. 3 illustrates the assembly of strips BF.sub.SiC around the billet of nuclear fuel C.sub.Nu, and also the zones Z.sub.inter for filling the inter-turn space with Si or Si foam for example and the inter-turn overlap zones Z.sub.Re, the horizontal arrow indicating the direction for introducing the thus wound billet into its main shell within the cladding Go.

(21) When the covering operation is finished, the system is adjusted to the correct length (by simple radial cutting) and optionally placed under vacuum, in order to evacuate the binder of the soft braze optionally affixed, before the closure by welding, the remainder of the operations for manufacturing the pin being well known to a person skilled in the art.

(22) Behavioral Validations of the Pin of the Present Invention under Nominal Operation at Static Temperatures of 500° C.

(23) In this temperature range, the applicant proposes to use SiC for its good thermal and physical properties, for its excellent behavior under irradiation, notably in these temperature ranges, around 500° C., where the swelling is typically of the order of 0.5 to 1%, at a given integrated dose. The solid-state interactions, between the metal fuel and the SiC (as soon as they are in contact), are not zero but are kinetically postponed or delayed by the presence of SiO.sub.2 (quartz or silica) on the weaving, in contact with the MO.sub.2 layer formed by oxidation pretreatment on the metal fuel.

(24) Thermodynamically, it is well known that the affinity of silicon Si for oxygen is less than that of U, or of Zr. Thus, the natural chemical evolution of the system is the displacement of oxygen from the quartz or the silica SiO.sub.2 toward the actinide alloy in order to favor the maintenance of its oxidation by means of the formation of an undefined layer M.sub.uSi.sub.vO.sub.w as illustrated in FIG. 4 showing the affinity of certain compounds for oxygen.

(25) The interaction between the woven cover and the actinide alloy may then form a complex interaction layer M.sub.xC.sub.ySi.sub.z, undefined a priori, but for which the growth (diffusion) kinetics are limited by the temperature (typically some 10 to 100 μm). It should be noted that there are no notable interactions, in the solid state, between SiC and the stainless alloys, in particular the austenitic alloys of 316L type such as the cladding, up to a temperature of more than 1200° C. These alloys may furthermore be filled with SiC during their processing in order to mechanically reinforce them as described in the article: Journal of Materials Science Letters 19 (2000), Vol. 7, pp 613-615; Materials Science and Engineering: A, 335 (2002), Vol. 1-2, pp 1-5.

(26) During operation, the mechanical properties of the SiC weaving (to start with loose at the implementation), make it possible, during the temperature rise, to contain the expansion of the actinide alloy (the expansion coefficient of which is typically three times greater than that of SiC), then to force the metal fuel to be plasticized for the most part in the longitudinal direction. When the fuel/cladding gap is taken up, this SiC makes it possible to ensure good removal of the heat toward the cladding and the coolant, via conductivity (fuel-woven SiC-cladding contact), with, optionally and advantageously, a soft braze of Ni, without overloading the Fe—Cr or Fe—Cr—Ni cladding with stresses (with the suitable sizing).

(27) The porous nature of this weaving of wound turns makes it possible to discharge the fission gases toward the plenum.

(28) Therefore under nominal conditions the interactions that should always be expected or feared at temperature between a metal nuclear fuel under flux and its cladding are here reduced and spread out on principle while by design, the fuel-cladding eutectic cannot be formed (no direct contact).

(29) Behavioral Validations of the Pin of the Present Invention under “Incidental” Operation at Temperatures above the Melting Point of the Metal Fuel Alloy Used

(30) At higher temperature, above the melting point of the UPuZr metal fuel, starting from 1000° C. typically, it is the high reactivity and the low corrosion resistance of the SiC by the molten actinides which is interesting. Unlike most other uses, applications or inventions using SiC, the applicant proposes to use SiC as a material of low chemical inertness, used as a consumable, in order to react with the alloy of molten actinides, and to give rise to refractory carbides and silicides that are less dense than the UPuZr metal alloy that gave rise to them.

(31) The SiC may not thermodynamically and kinetically withstand the corrosion of the alloys of molten actinides, and in particular of an alloy such as UPuZr, since the elements Pu, U and Zr have a very strong affinity for carbon, in order to form numerous carbides that are thermodynamically more stable than SiC, and for the Si, in order to form numerous silicides, and mixed carbosilicides M.sub.xC.sub.ySi.sub.z; FIG. 5 illustrates, in this regard, the diagram of the relative stability of carbides, silicides or carbodisilicides, showing the free enthalpy formation Δ.sub.fG (a.u.) for M=U, Pu, Zr.

(32) Thus the formation of these mixed carbosilicides M.sub.xC.sub.ySi.sub.z, illustrated in FIG. 6, follows, these mixed carbosilicides constituting a wall between the cladding and the unmelted nuclear fuel. These less dense refractory compounds reduce the local neutron reactivity since they are less dense, and therefore see their temperature decrease (versus the neutron reaction), since they heat less.

(33) From a simple thermodynamic viewpoint, it is possible to understand on the basis of the physical chemistry of the molten state/SiC interfaces [Survey on wetting of SiC by molten metals, G. W. Liu, M. L. Muolo, F. Valenza, A. Passerone, Ceramics International 36, 4 (May 2010) 1177-1188; Acta metall, mater. Vol. 43, No. 3, pp. 907-912, 1995] and also on reading U—Si binary systems [Journal of Nuclear Materials 389 (2009) 101-107], U—C binary systems [Journal of Nuclear Materials 288 (2001) 100-129], Pu—Si binary systems [Journal of Nuclear Materials, Volume 15, Issue 1, 1965, pages 23-32], Pu—C binary systems [Computer Coupling of Phase Diagrams and Thermochemistry 32 (2008) 371-377], that the thermodynamic activity of carbon or silicon in the molten PU or U increases very rapidly, and that the precipitation of the solids MSi.sub.x (U.sub.3Si.sub.2, at 1000° C. typically), MC.sub.x or even MC.sub.xSi.sub.y is imposed thermodynamically relative to the binary compounds: Fe—Si, U—Si, Pu—Si, Pu—C, U—C, Zr—Si.

(34) Therefore, the liquid metal will rapidly form a very complex interaction layer MSi.sub.xC.sub.y made of numerous carbides and silicides of these various elements, which interaction layer is known to be somewhat refractory (the melting points of the compounds are for the most part between 1000 and 1600° C. typically like plutonium silicides [Suppl. to IEEE Transactions on Aerospace, June 1965, Plutonium Compounds for Space Power Applications] and an effective conductor of heat, like most carbides and silicides as described notably in the article: Journal of Nuclear Materials, Volume 168, Issues 1-2, October-November 1989, pages 137-143). Suppl. to IEEE Transactions on Aerospace, June 1965, Plutonium Compounds for Space Power Applications.

(35) The MSi.sub.xC.sub.y compounds formed are less dense than the fuel metal (between 4.9 g.cm.sup.−3 for ZrSi.sub.2, between 7-8 and 10 g.cm.sup.−3 for silicides of U and of Pu) and, on average, less dense than the average of the densities of the volumes of SiC (advantageously filled with Si) and of molten actinide that gave rise to these compounds.

(36) Due to the dedensification, the molten metal, likened to a liquid tube that has not yet reacted is mechanically discharged upwards as indicated by the arrow in FIG. 6, carrying along possible less dense MSi.sub.xC.sub.y products or floating by de facto decreasing the density of fissile atoms locally in the fissile column (in situ creation of a composite nuclear fuel, for which the density of heavy atoms is lower) by the effect of radial compression and closure of said liquid tube, during the progression of the interaction.

(37) The expansion vessel of the molten actinide alloy (in the plenum) is protected from the corrosive nature of the molten fuel, and notably of plutonium, by the tantalum Ta coating of materials conventionally used for this type of problem (see for example patent FR 2 752 234 from 1998 describing a stainless steel/Ta/stainless steel composite cradle developed by CEA/DAM in order to contain the liquid alloys of Pu—Ga).

(38) In principle, any chemical reaction producing compounds that are solid and less dense at the temperature considered, and that are not very dense, may enable an operation such as that described.

(39) FIGS. 7a and 7b thus schematically show, with transverse cross sections, the evolution of the metal nuclear fuel C.sub.Nu/primary shell constituted of fibers F.sub.SiO2 of silica/main shell constituted of fibers F.sub.siC of SiC/cladding Go interfaces, during the reaction for melting the nuclear fuel with the appearance of an additional interface of molten metal nuclear fuel that has reacted with the SiC in order to give rise to compounds M.sub.xC.sub.yS.sub.iz and oxides M.sub.uSi.sub.vO.sub.w. The nuclear fuel C.sub.Nu is expanded and swollen, the constituent turns of the main shell made of F.sub.SiC being compressed against the cladding Go.