WO2000026921A2 - Rodlet absorbing members for use with spent fuel - Google Patents

Abstract

To attenuate heat generation due to the release of neutrons from remnant fissile fuel which remains in spent fuel rods, rod-like reaction attenuating members (RAM) are interspersed among the spent rods. When the rods contain boron such as B-10, most neutron capture occurs near the surface of the RAM. Accordingly, the formation of RAM cross sections having increasing surface area enables a reduction in the weight of the rod/unit length while maintaining the 'blackness' or neutron opacity of the member. RAM can be deployed in both the spaces provided in the fuel rod assemblies for control rods during combustion, as well as about their periphery when they are removed from the reactor pressure vessel and are immersed in 'on-site' storage pools. The rods provide flexibility in disposition and numbers and can thus be applied to a wide variety of different control rod assembly units and evaluated against a plurality of criteria. As the activity of the fuel decreases, rods can be removed, allowed to cool off for a few days and subsequently used in 'hotter' incoming assemblies if so desired.

Description

RODLET ABSORBING MEMBERS FOR USE WITH SPENT FUEL

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This application claims priority from Provisional Patent Application Serial No. 06/107,009 filed on November 4, 1998, entitled RODLET ABSORBING MEMBERS FOR USE WITH SPENT FUEL.

BACKGROUND OF THF INVFNTION

Field nf the Invention

The present invention relates generally to the storage of spent nuclear fuel and more specifically to a rod-like structure which is used in conjunction with the storage of the spent fuel, and which can, in accordance with certain embodiments, have a shape adapted to increase the amount of neutron absorption for the given mass of the rod.

npsrriptinn nf the la d Art

With the passing of time, the amount of spent fuel which is generated by the operation of commercial nuclear reactors is accumulating at a rate which is greater than it can be treated and/or placed in permanent storage facilities. This accumulation is therefore posing a problem in that it is increasing the amount of spent fuel which, after being removed from a reactor vessel, must be transferred to and temporarily stored immersed in an on-site pool or pools which are located adjacent and/or relatively close to the reactor.

This fuel, even though it is "spent", still contains small amounts of fissile/radioactive material and is therefore capable of releasing neutrons and accordingly promoting the generation of heat. It must therefore be carefully monitored and maintained constantly immersed in water and/or boronated water until such time as the amount of heat which is generated, becomes constant and/or lowers to an acceptable level. This process is quite long. The "hottest" of the spent material is stored in an area referred to as "region I" and is kept in this area for about 5 years. It is then shifted to a "region II" storage area until further decay/cooling has taken place. After region II, the spent fuel is moved to a region III area wherein it awaits removal from the wet storage and disposition in dry casks.

However, due to the current shortage of government operated permanent repositories for such material, the build-up of the spent fuel at these on-site locations is increasing. This, due to the limited space available in the on-site storage facilities, has lead to increased storage density of spent fuel assemblies. Accordingly, in addition to the shielding effect of the water/boronated water, neutron absorbing/shielding structures must be additionally interposed between the individual spent fuel rod assemblies to attenuate excessive heat generation and the possibility of a chance reaction or the like. These reaction attenuating structures often contain materials such as B¹⁰ or the like having a large neutron cross section.

In some support structures which are provided in the storage pools, the actual partitions which form the "egg-crate" like cell arrangements include materials which have a large neutron cross section and therefore tend to capture neutrons before a reaction can be initiated. However, it also desirable, as the clearance between units reduces, to be able to supplement this with additional reaction attenuating members.

Nevertheless, the fact that the commercial reactors which are currently in commercial use about the world, are built by different manufacturers and have been built at different times, leads to the situation wherein, with the increase in knowledge and the experience which is gleaned from each construction and the advancement in technology which occurs between constructions, no two facilities tend to be the same and no basic standards have therefore been established regarding the construction and arrangement of such structures. Accordingly, the shape and size of the fuel rods/fuel rod assembly constructions and the method of dealing with the spent fuel tend to vary on an almost case-by-case basis.

Storage arrangements which have been and/or are currently in use are exemplified by United States Patent No. 4, 1 1 5,700 issued on September 1 9, 1 978 in the name of Grooves; United States Patent No. 5,063,299 issued on November 5, 1 991 to Efferding; United States Patent No. 4,203,038 issued on May 1 3, 1 980, to Takahashi et al.; or United States Patent No. 4,225,467 to McMurty et al. and which was issued on September 30, 1 980. The disclosure of these documents is hereby incorporated by reference.

Figs. 1 -4 show basic features of the support structure 10 which is disclosed in the above mentioned United States Patent No. 4,203,038. In these arrangements, however, the use of additional flat sheets and/or chevron-shaped neutron shielding/capturing members can't always be conveniently dropped into place. That is to say, in some instance the cells 12 into which the fuel rod assemblies 1 4 are disposed are rectangular/square in configuration (see Fig. 2) while others are hexagonal or the like (see Fig. 4). Thus, while the flat sheets and the right angle chevron-shaped members may be lowered into the square chambers, they will not readily fit into the hexagonal ones. On the other hand, chevron-shaped members which are angled to fit into the hexagonal cells will not be readily deployed in the square cells. This problem is, of course, rendered all the more difficult by the fact that as the material is still highly radioactive, the work of introducing the reaction attenuating members in among the fuel rods/fuel rod assemblies must be carried out with the operator or operators well separated from the scene. This of course involves the use of a crane arrangement, remote control cameras and the like, and is accordingly tedious, especially when the reaction attenuating members tend to be oversized or so shaped as to not fit into place with ease.

Accordingly, there is a need for a reaction attenuating member which can be deployed in essentially all of the various types of fuel assemblies and the structures which are currently used to support same, and which are both light and economical, and can be removed when fuel becomes stable and can be moved from the storage pools and loaded into dry shielded containers or casks wherein it can be encased and transported to a permanent storage site such as an underground facility or the like.

SUMMARY OF THF INVENTION

The invention is based on the fact that it is possible to attenuate heat generation due to the release of neutrons from remnant fissile fuel which remains in spent fuel rods, through the use of rod-like reaction attenuating members (or RAM as they will be referred to hereinafter) which are interspersed among the spent rods. These rodlets not only fit in tight nooks and awkward spaces, but are also controllable in number.

One aspect of the present invention takes advantage of the fact that the fuel rod assemblies usually have provision for the movement of control rods between clusters of fuel rods that go to make up each of the fuel assemblies, and deploys RAM in the tubes normally occupied by "hot" control rods used during actual nuclear combustion.

Figs. 5, 6 and 7 show the layout of three different types of fuel cell or fuel assembly. The arrangement show in Fig. 5 has nine zirconium alloy tubes 16 which permit control rods 1 8 to be selectively moved therewithin during the operation of the reactor and reaction of the fuel contained in the surrounding clusters of fuel rods 20. The arrangement show in Fig. 6 on the other hand, is such as to feature a central channel 22 which extends along the very center of the unit in combination with four hollow wings 24. Normally, four of this type of fuel cell 25 are arranged together and a cruciform control rod carrier or support 26 is disposed therebetween in the illustrated manner. Thus, this arrangement does not in fact include guide tubes for the control rods. The arrangement depicted in Fig. 7 is somewhat similar to the arrangement shown in Fig. 5, however, it features a greater number of guide tubes 28 per unit. The units in this instance include an egg-crate support 29 which not only supports the guide tubes 28 but the fuel rods 30.

For further details concerning the structures shown in Figs. 5, 6 and 7, reference may be had to United States Patent No. 3,879,259 issued to Persson et al. on April 22, 1 975; United States Patent No. 5,483,565 to Magnus et al. issued on Jan. 9. 1 996; and United States Patent No. 5,249,210 issued to Nylund et al. on September 28, 1 993. The contents of these documents is hereby incorporated by reference thereto.

The RAM according to the present invention, can be made of a boron containing stainless steel, containing about 2% enriched boron. However, this less preferred embodiment of the invention is such that the rods tend to suffer from several drawbacks. They can contain only a limited amount of boron, accordingly must be relatively thick and therefore quite heavy. This lack of boron tends to limit their "blackness" and more importantly, adds directly to the total weight of the radioactive material during storage.

A further shortcoming of these storage control rods is that their exterior surfaces are constantly exposed to the heat from the decaying spent fuel in a manner which tends to prevent heat loss from their relatively high mass. Accordingly, they therefore tend to develop core temperatures which are in excess of desirable levels. These rods therefore can reach states wherein they themselves require cooling.

More preferred embodiments overcome these shortcomings by enabling the inclusion of a greater amount of boron and thus increase the number of sites where neutron capture can occur while obviating the weight problem mentioned above. In fact the most preferred embodiments from the view point of weight and effectiveness, are based on the fact that in boron containing RAM most neutron capture actually occurs near the surface of the RAM. Accordingly, the formation of RAM having surface area increasing cross sections is possible and enables a reduction in the weight of the rod/unit length while maintaining the "blackness" or neutron opacity of the member.

The RAM according to the above feature of the present invention can, due to their lower mass and adequate blackness, be deployed in both the spaces provided in the fuel rod assemblies for control rods during reaction, as well as about their periphery when they are removed from the reactor pressure vessel and are immersed in "on-site" storage pools, without excessive weight penalties. The rods provide flexibility in positional deployment and numbers and can thus be applied to a wide variety of different control rod assembly units. As the activity of the fuel decreases, rods can, if so desired, be selectively removed from "cooler" assemblies for use with "hotter" incoming assemblies, and thus be "recycled". As will be appreciated, an outstanding advantage of the invention resides in the ability to reduce the weight of the reaction attenuating members (RAMs) without loosing any opacity to neutrons (blackness), renders the stored spent fuel lighter and less susceptible to the effects of seismic events. That is to say, the reduced weight reduces the mass of the stored units and thus reduces the force needed to retain the stored units safely in place during an earthquake or the like.

Another advantage of the invention comes in the ability to reduce the amount of material which is used to store the waste fuel and thus reduces the parasitic costs involved therewith.

Therefore an object of the present invention is to provide a rod-shaped reaction attenuating member or RAM, which can be slipped with relative ease into position within either a spent fuel assembly and/or a cell in which a spent fuel rod assembly has been disposed.

It is a further object of the present invention to provide a rod-shaped reaction attenuating member which contains sufficient amount of boron, including the B¹⁰ isotope, to render the RAM sufficiently "black" (opaque to neutrons) and thus effective to capture neutrons which are emitted by the spent fuel which is being "guarded" by same.

It is a further object of the present invention to provide a rod-like RAM which can be extruded using an aluminum/boron carbide metal matrix alloy in a manner which provides a cross-sectional configuration that increases the surface area which is available for neutron capture and which simultaneously reduces the weight and cost of the member.

It is a further object of the present invention to provide a RAM which can be used in a fuel rod assembly in a manner which enables sufficient inter-rod shielding to be achieved that the possibility of a reaction establishing is attenuated to a zero level.

Thus, as will be appreciated from the forgoing, the present invention features embodiments wherein neutrons are captured by the B¹⁰ in the superficial portions of the rod attenuating structures and that the deeper core portions of the structures are not essential to the attenuating function thereof. The present invention further features, a number of embodiments, increasing the amount of surface area of an attenuating member allows the member to remain as "black" (opaque to neutrons) as required while enabling the member to be made of a reduced amount of material, and thus be reduced in both cost and weight.

In other words, the present invention is directed to rod-like reaction attenuating structures which are readily disposed in positions where a shadow shielding effect can be produced and which are, in certain embodiments, shaped in a manner which increases the amount of surface area which is available for neutron impingement and thus capture. Examples of such cross sections for solid members includes: circular, cruciform, four or five point stars and multi-lobed arrangements. These members are, in accordance with embodiments of the present invention, made of an aluminum boron carbide metal matrix and are unitary in construction. However, the invention is not so limited and even though there are some practical limitations involved with the extrusion of this type of material, it is within the scope of the invention to form hollow (tube) members or to "build" structures which have a basic support core member in the form of either an endo- or exoskeleton type arrangement. While more difficult to form, the hollow arrangements are such as to provide internal surfaces where a neutron which has by some chance made its way into the interior of the member will be captured as it attempts to exit therefrom.

A first specific aspect of the invention resides in a reaction attenuating member for use with spent nuclear fuel, comprising: an elongate structure which comprises a material having a having a large neutron cross section and a cross-sectional shape which provides a surface area via which neutrons, which are emitted from the spent fuel, are captured by the reaction attenuating member.

A second specific aspect of the invention resides in a reaction attenuating member for use with spent nuclear fuel rods which are immersed in a water filled pool, comprising: a rod having a configuration that increases the amount of surface area available for a given mass of material and which includes a material having a large neutron cross section, the configuration enabling neutrons to be absorbed for the given mass in a manner which attenuates neutron induced reaction between fissile material remaining in the spent fuel rods.

A third specific aspect of the invention resides in a reaction attenuating arrangement for use with a plurality of spent nuclear fuel rods comprising: a plurality of rod-like elements, each of said rod-like elements having a configuration which: a) increases the amount of surface area available for a mass of the rod, and which b) enables neutrons to be captured by a neutron capturing material included in each of the said rods in a manner wherein neutron induced reaction between fissile material remaining in the plurality of spent fuel rods, is attenuated.

A fourth specific aspect of the invention resides in a method of storing spent nuclear fuel, comprising the steps of: immersing a fuel rod assembly containing spent fuel rods, in a water filled storage structure; and disposing elongate reaction attenuating members in spaces which are provided in fuel rod assembly and/or located about the exterior of the fuel rod assembly, and which are adapted to receive control rods when disposed in a pressure vessel of nuclear reactor , to attenuate heat generation due to neutrons which are released by fissile fuel that remains in the spent fuel rods, during a storage period which follows combustion in a pressure vessel of a nuclear reactor. A fifth specific aspect of the invention resides in a method of storage comprising the steps of: disposing structures which define storage cells in a pool filled with water; disposing spent fuel rod assemblies in the cells; disposing elongate reaction attenuation rod members in one of: a) spaces defined between each of the fuel rod assemblies and walls of the cell in which it is disposed, and b) tubes in which control rods were slidably disposed when the fuel rod assembly was disposed in a reactor pressure vessel .

A sixth specific aspect of the invention resides in a spent nuclear fuel storage arrangement, comprising: a pool filled with water; a structure immersed in the pool; a plurality of spent fuel rods supported in the immersed structure; and a plurality of elongate reaction attenuating rodlets interspersed between the fuel rods to interact with and capture neutrons emitted from fuel remaining in the spent fuel rods and attenuate an undesirable heat generating reaction between the spent fuel rods.

RRIFF nF.ςrRIPTIDN DF THF DRAWINGS

The various features and advantages of the present invention will become more apparent as a detailed description of the preferred embodiments are given with reference to the appended drawings in which:

Fig. 1 is a perspective view showing an arrangement used to support spent fuel rod assembly units during the immersed storage following removal from the reactor pressure vessel;

Fig. 2 is a sectional view showing the disposition of a fuel cell or fuel assembly in the support structure shown in Fig. 1 ;

Fig. 3 is a plan view of the arrangement depicted in Fig. 2 wherein rectangular cells are used;

Fig. 4 is a plan view showing a variant of the arrangement depicted in Fig. 3 wherein hexagonal cells are used to store the spent fuel assemblies;

Fig. 5 is a plan showing the arrangement of fuel rods and control rod passages according to a first type of fuel rod assembly;

Fig. 6 is a plan view showing the arrangement of fuel rods and voids in accordance with a second type of fuel rod assembly;

Fig. 7 is a plan view showing the arrangement of a third type of fuel rod assembly; and

Figs. 8 to 10 are respectively charts showing an analysis of 1 1 different embodiments #1 - #1 1 , which were tested and evaluated against a plurality of criteria.

ΠFTΔII FΓ> nFSCRiPTiON DF TMF PRFFFRRFΠ FMRDDIMFNTS Before dealing with the specific shapes of the embodiments of the invention, it is deemed appropriate to consider the material from which it is possible to form the RAMs. Details of how a suitable material can be created and converted into suitable shapes and structures is basically outline in copending United States Patent Application Serial No. 09/059,389 filed on April 14, 1998 in the name of Haynes et al. This application is incorporated by reference thereto for disclosure relating to the use of a boron containing metal matrix.

Due to its unique combination of relatively low cost, low toxicity and high radiation absorption capacity, boron has long been used for shielding of neutron radiation in connection with nuclear reactors. Boron has a high absorption cross section, so that a neutron passing through material containing boron atoms will tend to interact with a boron nucleus and hence be prevented from passing through. Native boron contains both B¹⁰ and B^{1 1} isotopes, the former having a fivefold higher neutron absorption capacity than the latter. Through an enrichment process, the B¹⁰ content can be increased above the naturally occurring level.

The radiation shielding composition which is used in connection with the present invention is formed by a process which involves the compaction, sintering and extrusion of a mixture of a metal matrix material and a boron-containing particulate material dispersed in the matrix. One of the features of the composite structure of the invention is that the boron-containing particulates are sufficiently small and evenly distributed to preclude fracturing. This avoids fractured particles can smear during metal working, the formation of voids and associated corrosion.

Another feature of this extrusion process it that it provides strong metal particle-to- metal particle bonds without the need for binders or chelating agents. Further, the boron carbide particulates incorporated in this manner in the composition are readily available and meet ASTM 750 specification. In fact ASTM boron carbide has demonstrated outstanding performance in both hot section and spent fuel storage environments for over 40 years of service. The metal matrix incorporating the boron particulates is selected for its strength, radiation half-life characteristics, and corrosion resistance in both boiling water reactor (BWR) and pressurized water reactor (PWR) wet environments and all dry storage and/or spent fuel storage applications.

Aluminum alloys are highly preferred as the matrix material. Generally, aluminum alloys useful in the invention have a combined total of iron, cobalt, manganese, copper and zinc of less than about 0.5 percent by weight and preferably less than about 0.3 percent by weight. Further, the combined amounts of nickel, cobalt and manganese are less than about 50 ppm total and preferably less than about 30 ppm total. The amount of cobalt is generally less than about 10 ppm and preferably less than 5 ppm. Suitable aluminum alloys include the 6000 series alloys, and most preferably a 6061 alloy having a composition low in Fe, Ni, Co, Mn, Cu, and Zn. These aluminum alloys are metallurgically well understood and are very widely used with a long history of good corrosion resistance and mechanical properties. When exposed to neutron radiation, very little long term radioactivity of the material is generated due to the alloys' chemical composition. This in turn is due to the fact that the primary elements in the alloy (Al, Si and Mg) all have relatively low cross sections for neutrons and the isotopes formed from transmutation have short half-lives.

Conversely, materials having high Cu content (2000 series alloys), Zn (7000 series alloys), Mn and Fe (many 3000 series alloys), Ni and Co are quite radioactive after neutron irradiation because these elements transmutate into isotopes with longer half-lives. (For example, Co 60, etc.). It is therefore highly preferred that the aluminum alloys used in forming the radiation shield composition of the invention have a low percentage of these elements. The 6000 series alloys are also heat treatable with good strength and elevated temperature resistance. These 6000 series alloys can also be anodized and hard coated to improve corrosion resistance, abrasion resistance and emissivity, all of which are significant in dry cask storage applications. Dry casks are used to store spent nuclear reactor fuel prior to long term disposal. Further, because of their high strength, 6000 series alloys are capable of being used as structural elements. Not only can the material shield from neutron radiation, it can act as the supporting structure into which radioactive material is stored.

A preferred aluminum alloy is in the form of a pre-alloyed powder formed by subjecting an aluminum alloy melt to a powder metallurgy technique. The term "pre- alloyed" means that the molten aluminum alloy bath is of the desired chemistry prior to atomization into powder. In a highly preferred embodiment, the alloy melt is passed through a nozzle to form an atomized stream of the melt which is cooled at a rapid rate by an inert gas stream (e.g., argon or helium) impinging the atomized stream. Cooling takes place at a rate of about 1000°C (1832°F) per second, producing a spherical-shaped powder. The powder has an oxide layer, but the thickness of this layer is minimized due to selection of inert gas as the cooling fluid. It is possible to use water on air as the cooling fluid, but the oxide layer thickness is increased. Preferably no other low melting alloy addition is blended with the alloy composition.

The aluminum alloy powder may be characterized by particle size distribution ("D"). The term "D10," for example, would indicate that 10% of the alloy particles have a particle size less than or equal to the assigned value (e.g. 6 μm). Generally, the particle size distribution of the alloy powder has a D10 value of about 6.0 μm, a D50 of about 20 μm and a D90 value of about 38 μm. This particle size distribution may be measured by a Microtrac Analyzer (laser-based technology) or equivalent sedagraph. Desirably the D10, D50 and D90 values are about 4μm, 12 μm and 25 μm, respectively, and preferably, the values are 2 μm, 9 μm and 17 μm, respectively. It is to be understood that the above stated values are independent of one another. Thus, a particle size distribution within the scope of the invention includes, for example, particles having a D10 of 4 μm, a D50 of 20 μm and a D90 of 17 μm.

The aluminum alloy powder is blended with boron-containing particulates comprising from about 2 to about 45% by volume of the overall composition. Desirably the amount of boron is from about 10 to about 40% by volume and desirably from about 1 5% to about 35% by volume of the overall composition. Preferably, the boron is in the form of boron carbide particulates. Generally, the boron carbide has a particle size characterized by a minimum of 98% less than 40 μm, desirably less than 30 μm and preferably 98% less than 20μm. In a highly preferred embodiment, the boron containing particulates comprise nuclear grade boron carbide powder prepared according to ASTM C750-89 (Type 1 ). This boron carbide powder has the following composition:

Note: *B¹⁰ specified as atomic weight percent * *Specifying a maximum level of water soluble boron is important to limit leaching of free boron in certain environments.

The particle size of the boron carbide according to this standard is 98.0% min. less than 20 microns.

After the aluminum alloy powder and boron-containing particulates are uniformly mixed, they are subjected to a compacting step whereby the mixture is placed in a urethane elastomeric bag, tamped down and vibrated, and then subjected to vacuum to remove air and other gaseous materials. The vacuum is generally 10 torr or less absolute pressure, desirably about 1 torr or less and preferably about .50 torr or less absolute pressure. After vacuum is applied for a period of from about 2.5 to about 5 minutes, the compressed particulates are subjected to isostatic compression at a pressure of at least about 30,000 psi, desirably at least about 45,000 psi, and preferably at least about 60,000 psi. This isostatic compaction takes place at a temperature of less than about 212°F (100°C), desirably less than about 122°F (50°C), and preferably less than about 77°F (25°C), i.e. about room temperature.

The resulting "green" billet is then vacuum sintered at a temperature which is a function of the particular alloy composition, and is such that during the sintering process the particulate microstructure is left substantially unaffected. By the term "substantially unaffected" is meant that while the majority of the sinter bonds are formed by metallic diffusion, a small amount of melting can occur, however, this amount does not change the physical properties of the aluminum alloy powder to an extent that would affect the strength of the subsequently formed article. Generally, the sintering temperature is within 50°F (28°C) of the solidus of the particular composition, but may be higher or lower depending on the sintering characteristics desired. The term "solidus" refers to the point of the incipient melting of the alloy and is a function of the amount of alloying materials present, e.g. magnesium, silicon, etc. The vacuum under which sintering takes place is generally 100 torr or less absolute pressure, desirably 10 torr or less, and preferably about 1 torr or less absolute pressure.

The sintered billet may then be subjected to additional processing such as extrusion or other hot working processes. In a preferred embodiment, the sintered billet is extruded using the process described hereinbelow. This extrusion process has the advantage that strong metal particle-to-metal particle bonds are formed. As the sintered billet is extruded, the sintered particulates abrade against each other as they pass through the extrusion die. This abrading process removes the naturally occurring metal oxides on the outer surfaces of the aluminum particles, exposing the underlying metal and allowing a strong metal to metal bond to be formed. This phenomenon is in contrast to that occurring in the process described in U.S. Patent No. 5,700,962, which employs a chelating agent to bind the particulates.

A preferred extrusion process includes provisions for maintaining extrusion die temperature within close tolerances, i.e. within about ± 50°F (28°C) of a target temperature, desirably within about ± 30°F (1 7°C), and preferably within about ± 1 5°F (8°C) of a target temperature. The actual target temperature is itself a function of the particular alloy being extruded but is typically between about 930°F (499°C) and about 970°F (521 °C). It is highly preferred that the extrusion temperature not exceed the solidus temperature. The extrusion temperature is preferably measured at the exit of the die, thus accounting for temperature effects due to friction and working of the billet.

One or more, and preferably all of the components which are used with an extrusion die may be constructed of Inconel 718 or another alloy having a yield strength equivalent to or greater than that of Inconel 71 8 at 900-1000°F (482-538°C) to prevent deflection or mandrel "stretch" due to high temperature creep. This is particularly important at die face pressures greater than 95,000 psi at 900°F (482°C). At die pressures below this level, the extrusion die may typically be constructed of H13 tool steel.

A nonmetal insert which forms part of the die is preferably micrograined tungsten carbide (less than one micron diameter grain size) with a cobalt binder level between about 12% and 1 5%. This material exhibits a minimum transverse rupture strength of 600,000 psi. The use of Inconel 71 8 as the die insert holder with the tungsten carbide insert minimizes the possibility of cracking of the insert due to differences in coefficient of thermal expansion.

The extrusion container temperature is maintained within the same temperature limits as the extrusion die. In both cases, this may be accomplished by microprocessor controlled resistance band heaters or cartridge type heaters strategically placed on the extrusion container. Temperature is measured by multiple thermocouples imbedded in the die and container adjacent the container surface (generally within ^ΛA inch). Each portion of the extruder and die monitored by a thermocouple has independent temperature control.

The microstructural homogeneity of the boron carbide particulate in the aluminum matrix can be quantified by reference to the nearest neighbor particle spacing measurement. As discussed earlier, this measurement technique quantifies the distance between adjacent particles of the boron carbide within the aluminum matrix. The matrix is a 6061 alloy with 21 wt% B- . The 50 percentile nearest neighbor distance is 10.01 μm, while the 90 percentile value is 1 7.65. The standard deviation is 5.28 μm. This represents a very narrow range of values and is indicative of a highly uniform matrix. A measure of particle size distribution in the radiation shielding composition is the standard deviation of the particulate nearest neighbor distance. Generally, the standard deviation is less than about 1 5 microns at a 20 wt. % loading, desirably less than about 10 microns and preferably between about 4 and about 6 microns.

Both the particle size of the alloy powder and the boron containing material (refractory particulates) may be carefully controlled. The particle size relationship between the aluminum powder and the refractory particulates may be optimized for reproducibility of microstructure homogeneity. If too large a size between aluminum and boron containing material, the boron containing material will cluster together. The size and homogenous distribution of the boron carbide particulates in the matrix alloy is a factor in preventing neutron "streaming" and/or "channeling" through the composite cross section.

The following example illustrates the manner in which the alloy suitable for use in this invention, is produced:

FYAMPI F Rlenriing Operation

A pre-alloyed aluminum powder and boron carbide particulates are pre-weighed, added to a mixer, and mixed under a vacuum of 28 in. Hg for one hour. The aluminum powder is prealloyed 6061 alloy in the form of spherical particulates. This material has a melting point range of 1080-1 205°F (582-652°C), a density of 0.098 lb. /in³ (2.71 g/cm³) and a thermal conductivity of 1250 BTU/hr-ft²-°F/in (1 .55 kcal/hr-cm²-°C/cm) and has the following composition:

* Oxygen is noted as a reference information that a powder metallurgy process is used to manufacture the product.

* * Nickel, Cobalt, Iron, Manganese and Chromium are tramp elements not covered by ASTM B 221 Specification but may be significant to the performance of nuclear grade aluminum/BiC metal matrix composites in a radiation environment to prevent transmutation of elements with longer half lives, thereby maintaining the short half -life of aluminum material.

The particle size distribution is as follows:

Particle Size Range

With 100% of the particles less than 44 microns

The boron carbide particulate used conforms to ASTM C750-89 (Type 1 ). The mixing operation is done at room temperature and no organic binders or other additions are added to the batch. The mixing container is brought back to atmospheric pressure using nitrogen gas. The entire batch is processed through a Sweco Unit with a 200 mesh "supertaught plus" screen. This operation deagglomerates the AI/B4C mixture to prevent "cluster defects" in the compacted billet.

This screened mixture is placed back into the mixer and blended under vacuum for an additional 45-60 minutes. After this blending operation the mixing container is brought back to atmospheric pressure using nitrogen gas. Rillet Cnnsnliriatinn

A billet consolidation tooling consists of a stainless steel perforated basket, a urethane elastomeric bag, and a urethane top closure with a urethane deair tube. The perforated stainless basket is used to support the elastomeric mold during the mold fill operation and prevent bulging or distorting of the mold wall because of the hydrostatic pressure.

The elastomeric bag is filled with the IB C blend, vibrated, and tamped to obtain maximum packing and sealed. This tool assembly is evacuated to 28.5-29.0 in Hg to remove trapped air in the powder vacancies.

This evacuated elastomeric tooling is placed in a cold isostatic press (CIP) and a lightweight turbine oil is pressured to 55,000-60,000 psi pressure. The use of lightweight turbine oil eliminates the potential H2 gas generation if the elastomeric tooling is not pressure tight. The pressurized turbine oil goes through the perforated stainless steel basket and pushes against the urethane bag wall which consolidates the loose powder blend. Once the peak pressure is reached, the CIP automatically retains the full pressure for a dwell time of .75 to 1 .5 minutes.

After the dwell time is completed, the vessel is decompressed back to atmospheric pressure at a preset rate of 1 500 psi/second. The elastomeric tooling is run through a low pressure/high volume pass through washer to remove residual oil. The elastomeric tooling is removed from the billet and can be reused several hundred times. The billet is now solid and is 82-95 percent of theoretical density. .Sintering Operation

The resulting green billet is loaded into the hot zone of a vacuum sintering furnace. A two stage degassing operation prior to sintering is used. In the first stage, a major vacuum "surge" occurs between 250-340°F (1 21 -1 71 °C) and is the stage in which the free water vaporizes off the green billet. The second stage occurs between 800-880°F (427-471 °C), during which the chemically bonded water is removed from the hydrated oxide layer of the atomized aluminum powder. The vacuum level in the furnace recovers to a minimum vacuum level of 2.0 x 10-3 torr prior to proceeding to the final sintering temperature. The final sintering temperature depends on the matrix alloy. "Differential thermal analysis" is one of the techniques used to establish the final sintering temperature. Once the two stage degass is completed, the step is performed at the final set temperature. Upon completion of the sinter cycle, the hot zone is backfilled with low dew point nitrogen gas to cool down the furnace load. The billet is now in the "sintered condition". Fxtπisinn Operation

A 3000 ton Sutton extrusion press is used under the following conditions:

Billet Temp. - 980-995°F (527-535°C)

Container Temp. - 920-940°F (493-504°C)

Die Temperature - 900-920°F (482-493°C)

Extrusion exit speed - 7.5-8.0 feet/minute

The extrusion die has a fine grained tungsten carbide bearing insert that is interference fitted in an Inconel 718 insert holder to assure compression loading of the insert at the extrusion temperature. The die design also incorporates a 30-32 degree metal prework pocket prior to the final O.D. bearing to optimize particle to particle bond between the aluminum powder and the boron carbide particulate. All die components are interference fitted to assure compression loading at extrusion temperature to prevent die deflect.

During extrusion the die bearing area has a nitrogen gas "blanket" to optimize extrusion surface finish and prevent aluminum oxide buildup at the bearing land area. The extrusion is air cooled and stretch straighten prior to the cut-to-length operation.

RAM CONFIGURATIONS

The ideal rodlet RAM shape is readily extrudable, presents a high surface area for neutron capture and has a minimal cross-section so as to minimize overall weight and minimize material cost. Because boron carbide is roughly ten times more costly than aluminum, a rodlet design which maximizes surface area with minimum weight offers an economic benefit up to the point that increasingly difficult extrudability results in offsetting costs. Overall extrudability considers factors such as billet length, recovery, die face pressure, break pressure and extrusion speed. Rodlet designs can be evaluated by ranking important characteristics such as extrudability, surface area, raw material utilization (e.g., weight of alloy material per unit length of rodlet) and the net cost to convert from billet to a given rodlet shape including raw material cost and recovery of useable material. The following discussion is reflective of extrudability and conversion costs associated with extruding mutiples of a given rodlet shape in a 3000 ton direct extrusion press. Ratings can change as a function of press design, press capacity and die design. The discussion compares the characteristics of various RAM rodlet shapes in an extruded aluminum - boron carbide metal matrix material having a density of 0.098 pounds per cubic inch at a loading of 1 5 % boron carbide. These rodlets are designed to replace control rods in a nuclear fuel bundle. In concert with a specific commercial bundle, the rodlets were constrained to a diameter of 0.430 inches.

Figs. 8 to 10 contain tables which include depictions of eleven different cross- sectional shapes which were evaluated. The first of these (#1 ) has an essentially circular cross-section. The second #2 is tubular and has a wall thickness of 0.040" by way of example. The remaining cross sections #3 - #1 1 , will, for the sake of disclosure, be respectively referred to as a: four-pointed star, a tri-lobe, a curved tri-lobe, a thick five- pointed star, a thin five-pointed star, a thin cruciform, an intermediate cruciform, and a thick cruciform.

As will be appreciated, two compositions of each rodlet shape were considered. The first of each contained 1 5 - 25% B-tC, while the second contained higher contents of this material (i.e. 26 - 40% BiQ. It will be noted that some of the results are expressed as "NP". This denotes the instance that simple extrusion was nαLfound to be practicable. While, for example, in connection with the tube configuration, it is possible that a thicker wall thickness could be initially extruded and subsequently reduced to 0.040", using a secondary drawing process, the tests were focussed primarily on the ability to extrude the elements quickly, simply and therefore inexpensively.

It is therefore to be understood that the invention is such as to cover all of such possibilities and that the data which is presented in these charts is biased in the direction of cost efficient production via a single extrusion process.

The solid rod depicted as Item #1 in Fig. 8 can be taken as the baseline case against which to compare the characteristics of other rodlet shapes using a scale of one to eleven to rank individual characteristics of the various shapes. In this numerical ranking system, the lowest number represents the best performance and the highest number represents the worst performance. The solid rod has excellent extrudability (hence, the rating of " 1 ") because of its simple solid large cross-section. A rodlet of this shape has a surface area of 1 6.21 square inches per lineal foot. The rodlet weights 0.1 67 pounds per foot of length.

Fig. 8, Item #2 illustrates one method to greatly increase effective surface area while decreasing the mass of material needed to make a unit length. In this case the shape is a circular tube although other hollow shapes could be used. The tradeoff is in extrudability to the extent that at the higher volume loadings of boron carbide, the tube is not etrudable by economic means (hence, the rating of "NP", not practical).

Fig. 8, Item #3 depicts a solid, four-pointed star shape which has the same surface area as the baseline solid rod but owing to its diminished cross-section, requires only half as much alloy to produce a foot of rodlet length. At the lower particulate loadings, there is only a minor tradeoff in extrudability as compared with the solid rod shape of Item #1 . At the higher particulate loadings, Item #3 is not practical to fabricate.

Fig. 8, Item #4 depicts a solid, tri-lobe shape which has the same surface area as the solid rod and the four-pointed star. Improved material utilization results from a further reduced cross-section which greatly decreases the weight per unit length of rodlet. However, at the lower volume loadings of particulate, there is a significant penalty in extrudability. Interestingly, at the higher particulate loadings, the tri-lobe shape outperforms the four-pointed star in extrudability because the latter is not practical to fabricate because of billet length requirements and resulting recovery of usable product.

Fig. 9, Item #5 is a variation of the tri-lobe design which maintains the original surface area of 1 6.21 square inches. Extrudauility is improved but the cross-section is greater and material utilization suffers. There is more material per unit length of rodlet compared with the design of Fig. 8, Item #4.

Fig. 9, Item #6 is a modification of the tubular design of Fig. 8, Item #2 wherein the wall thickness has been increased to improve extrudability. Still, the tube is not extrudable at the higher volume loadings of boron carbide.

Fig. 9, Item #7 shows another lobed design, in this case, a five-pointed star shape. Compared with the tri-lobe shape of Fig. 8, Item #4, the cross-section is increased resulting in improved extrudabilty. Likewise surface area is increased but not by as much as the cross-section. Thus this five-pointed star design results in the consumption of more alloy material per unit length of rodlet.

Fig. 9, Item #8 is a modification of the five-pointed star design which reduces the cross-section to slightly less than that of the tri-lobe of Fig. 8, Item #4. Consequently, the amount of alloy material per unit length of rodlet is now slightly less than that of the tri- lobe. As compared with Item #7, extrudability is worse for the case of lower particulate loadings and slightly better for the case of higher particular loadings, reflective of billet length and recovery considerations.

Fig. 10, Items #9 through #1 1 depict a solid cruciform shape having increasingly thick legs and increasingly greater cross-section. As the thickness of the legs is increased, the extrudability improves while the extruded rodlet consumes increasingly more alloy material per unit length.

In the test data which is presented in the charts set forth in Figs. 8-10, the "Total value" is such as to represent an arbitrary value via which the various configurations and contents could be compared on an overall sense. As will be appreciated, the thick and thin five pointed star configurations actually scored the best overall results.

While the examples given in the charts of Figs. 8-10 disclose the use of BiC in the ranges of 1 5-25 % and 26-40%, the invention is in no way limited to these ranges and the content can be varied below 1 5% and above 40% although the amount of boron B¹⁰ which is present will be in approximately its minimum acceptable and maximum necessary amounts to provide the degree of shielding that is deemed necessary in the environment that the invention is necessary. The amounts of B C which are used of course will be effected by the addition of materials which contain the B¹⁰ isotope. In the event that enriched boron is used, then the amount of B*C particulate can be reduced to 8% reinforcement level. Mixtures of other materials which contain the appropriate isotope such as silicon hexaboride or alternatively other neutron capturing containing materials such as gallium or the like, can be used, in order to tailor the "blackness" or neutron opacity of the member to the desired level, if so required.

RAM DISPOSITION

In one deployment technique, the RAMs are suspended in a predetermined arrangement on a headpiece such that the rod-like members are located so that they can be lowered as a group, using a crane or the like, into preselected positions in, or around, one or more spent fuel rod assemblies. This procedure of course, reduces the amount of time and effort required to transfer the spent fuel rod assemblies from the reactor pressure vessel, dispose then in the storage pool and then place the RAMs in position to ensure safe storage over the 3-5 year cool-down period wherein the assemblies must remain immersed.

In the case of the type of fuel rod assembly depicted in Fig. 6, it is possible, even though there are no control rod guide tubes per se provided in the structure, to utilize the spaces within the structure in lieu thereof. For example, it is possible to deploy one or more RAMs in the central channel 22 or in the hollow wings 24.

Even in this position, water would be allowed to convect through these passages via the channels which are defined by the RAM configurations. The increased surface area of the RAM would allow for heat to be readily dissipated therefrom.

AI TFRNATIVF RAM CONFIGURATIONS

While the above described embodiments of the RAM are predominantly solid, and as extrusion of the above described material tends to be somewhat difficult, it is within the scope of the present invention to form flat members which are slotted and which can be interleaved in an interlocking manner and thus be assembled into a desired configuration. It is, as will be appreciated from the preceding, additionally not outside the scope of the present invention to form hollow members using a technique of forming an elongated flat sheet and winding same into helical form and welding or suitably joining the overlapping edges to define a tube. The various technologies which are available for producing tubes are known to those skilled in the art and therefore the methods of making tubular members from materials which contain amounts of boron or a similar type of neutron absorbing member will, in light of the preceding disclosure, be self-evident.

The various other techniques via which an economical RAM of the nature described hereinabove can be produced is therefore deemed to be self-evident to those skilled in the art to which the present invention pertains.

Claims

WHAT IS Cl AIMFD IS

1 . A reaction attenuating member for use with spent nuclear fuel, comprising: an elongate structure which comprises a material having a having a large neutron cross section and a cross-sectional shape which provides a surface area via which neutrons, which are emitted from the spent fuel, are captured by the reaction attenuating member.

2. A reaction attenuating member as set forth in claim 1 , wherein the material includes natural boron, boron enriched in a B¹⁰ or B¹¹ isotope, or a mixture thereof.

3. A reaction attenuating member as set forth in claim 1 , wherein the material is an aluminum boron carbide metal matrix alloy.

4. A reaction attenuating member as set forth in claim 1 , wherein the cross-sectional shape increases the surface/unit mass which is available for neutron capture.

5. A reaction attenuating member as set forth in claim 1 , wherein the elongate structure is solid.

6. A reaction attenuating member as set forth in claim 1 , wherein the elongate structure is hollow.

7. A reaction attenuating member as set forth in claim 1 , wherein the elongate structure has an essentially circular cross-section.

8. A reaction attenuating member as set forth in claim 1 , wherein the elongate structure has a non-circular cross-section and comprises a plurality of outwardly extending projection portions.

9. A reaction attenuating member as set forth in claim 8, wherein the plurality of projection portions are arranged essentially equidistantly about the periphery of the attenuating member.

10. A reaction attenuating member as set forth in claim 1 , wherein the elongate structure has a rod-like configuration which exhibits a predetermined surface area per unit mass ratio and which, when interposed between spent fuel rods from a nuclear reactor, attenuates heat generation due to neutrons which are emitted from the fuel which remains in the spent fuel rods.

1 1 . A reaction attenuating member as set forth in claim 10, wherein the cross-sectional configuration of the elongate structure is selected to increase the predetermined surface area per unit mass ratio and maximize the surface area per unit mass within constructional constraints.

12. A reaction attenuating member for use with spent nuclear fuel rods which are immersed in a water filled pool, comprising: a rod having a configuration that increases the amount of surface area available for a given mass of material and which includes a material having a large neutron cross section, the configuration enabling neutrons to be absorbed for the given mass in a manner which attenuates neutron induced reaction between fissile material remaining in the spent fuel rods.

1 3. A reaction attenuating member as set forth in claim 1 2, wherein the material includes natural boron, boron enriched in a B¹⁰ or B¹¹ isotope, or a mixture thereof.

14. A reaction attenuating member as set forth in claim 1 2, wherein the material is an aluminum boron carbide metal matrix alloy.

1 5. A reaction attenuating member as set forth in claim 1 2, wherein the cross-sectional shape of the rod increases the surface/unit mass which is available for neutron capture.

16. A reaction attenuating member as set forth in claim 1 2, wherein the elongate structure is solid.

1 7. A reaction attenuating member as set forth in claim 1 2, wherein the elongate structure is hollow.

18. A reaction attenuating member as set forth in claim 1 2, wherein the elongate structure has an essentially circular cross-section.

1 9. A reaction attenuating member as set forth in claim 1 2, wherein the elongate structure has a non-circular cross-section and a plurality of projection portions about a periphery thereof.

20. A reaction attenuating member as set forth in claim 1 9, wherein the plurality of projection portions are arranged essentially equidistantly about the periphery of the attenuating member.

21 . A reaction attenuating arrangement for use with a plurality of spent nuclear fuel rods comprising: a plurality of rod-like elements, each of said rod-like elements having a configuration which: a) increases the amount of surface area available for a mass of the rod, and which b) enables neutrons to be captured by a neutron capturing material included in each of the said rods in a manner wherein neutron induced reaction between fissile material remaining in the plurality of spent fuel rods, is attenuated.

22. A reaction attenuating arrangement as set forth in claim 21 , wherein each of said rod-like elements is disposed in a predetermined spatial arrangement with respect to a fuel rod assembly that has been removed from a pressure vessel of a nuclear reaction and contains fuel rods wherein most of the fuel has been spent but which contains residual amounts of fissile material.

23. A reaction attenuating arrangement as set forth in claim 21 wherein the fuel rod assembly contains spaces into which control rods are inserted during reactor operation, and into which the rod-like elements are selectively inserted.

24. A reaction attenuating arrangement as set forth in claim 21 , wherein said rod-like elements are made of a metal or a metal containing material.

25. A reaction attenuating arrangement as set forth in claim 24, wherein the metal is boronated.

26. A reaction attenuating arrangement as set forth in claim 24, wherein the metal contains boron, enriched boron containing a B¹⁰ isotope, or a mixture thereof.

27. A reaction attenuating arrangement as set forth in claim 24, wherein the metal is an aluminum boron carbide metal matrix.

28. A method of storing spent nuclear fuel, comprising the steps of: immersing a fuel rod assembly containing spent fuel rods, in a water filled storage structure; and disposing elongate reaction attenuating members in spaces which are provided in fuel rod assembly and/or located about the exterior of the fuel rod assembly, and which are adapted to receive control rods when disposed in a pressure vessel of nuclear reactor, to attenuate heat generation due to neutrons which are released by fissile fuel that remains in the spent fuel rods, during a storage period which follows combustion in a pressure vessel of a nuclear reactor.

29. A method as set forth in claim 28, further comprising the step of forming the rodlike reaction attenuating members so as to have a cross section wherein the surface area which is available for neutron capture, is selected to provide a predetermined surface area to mass ratio.

30. A method as set forth in claim 29, wherein the step of forming includes extruding a boron containing metal matrix.

31 . A method as set forth in claim 29, wherein the step of forming includes extruding an aluminum boron carbide metal matrix.

32. A method of storage comprising the steps of: disposing structures which define storage cells in a pool filled with water; disposing spent fuel rod assemblies in the cells; disposing elongate reaction attenuation rod members in one of: a) spaces defined between each of the fuel rod assemblies and walls of the cell in which it is disposed, and b) tubes in which control rods were slidably disposed when the fuel rod assembly was disposed in a reactor pressure vessel .

33. A method of storage as set forth in claim 32, further comprising the step of forming the elongate reaction attenuating members to have a rod-like configuration.

34. A method of storage as set forth in claim 33, wherein the elongate reaction attenuating members have a predetermined cross-sectional area/mass ratio.

35. A method of storage as set forth in claim 33, wherein the elongate reaction attenuating members have a cross section wherein the surface area which is available for neutron capture, is increased and surface area per unit mass is increased.

36. A method as set forth in claim 33, wherein the step of forming includes extruding a boron containing metal matrix.

37. A method as set forth in claim 33, wherein the step of forming includes extruding an aluminum boron carbide metal matrix.

38. A spent nuclear fuel storage arrangement, comprising: a pool filled with water; a structure immersed in the pool; a plurality of spent fuel rods supported in the immersed structure; and a plurality of elongate reaction attenuating rodlets interspersed between the fuel rods, each of the rodlets containing a material which interacts with neutrons emitted from fuel remaining in the spent fuel rods and thus attenuate a heat generating reaction between the spent fuel rods.

39. A spent nuclear fuel storage arrangement as set forth in claim 38, wherein each of the rodlets contains between 5 and 50% B4C.

40. A spent nuclear fuel storage arrangement as set forth in claim 38, wherein each of the rodlets contains between 1 5 and 40% B4C.