Abstract:
A neutron emitting assembly, which is useful in nuclear reactors and other industrial applications, is made of a major amount of beryllium encapsulating a minor amount of  252 Cf, which can be placed in a capsule having end plugs and a holding spring.

Description:
BACKGROUND OF THE INVENTION 
     1. Field of the Invention 
     This invention provides the means and mechanism by which to produce a steady source of high-energy neutrons which, in addition to the multiplication and efficient transformation of the radioactive energy of the primary driver isotope, can also be changed in strength through simple adjustments to the physical layout of the multiplier assembly. The resulting neutron source has many practical uses including, but not limited to: startup source for a nuclear reactor, non-destructive testing of materials, neutron activation analysis, sample moisture analysis, oil well logging, medical treatment of cancer, explosive detection, metal fatigue detection, and other real-time evaluations of chemical composition and moisture content of process streams such as combustion optimization in power plants and cement kilns. 
     2. Description of Related Art 
     Multiple neutron sources (emitters) are generally required in order to safely start up a nuclear reactor core. The reactor startup sources used for this purpose are referred to as “primary sources” and “secondary sources.” Primary sources are self-contained sources of neutrons that provide neutrons without the need for external power or irradiation from the reactor itself. Secondary reactor startup sources are universally made of initially non-radioactive driver materials uniformly mixed with beryllium. The secondary source driver material (typically antimony) is non-radioactive for manufacture. As a result, the secondary source does not produce a neutron source until the driver material is irradiated in a nuclear reactor. The secondary source produces neutrons as a result of the interaction of high energy gamma radiation from the radioactive decay of the driver material with the beryllium. Typical of the current art primary source driver materials, all used in combination with beryllium, are strong alpha particle emitting isotopes of polonium, radium, plutonium, americium or curium. The only material that is a practical primary source for commercial applications without the use of admixed beryllium is californium-252 or  252 Cf. 
     Descriptions of producing “secondary source” radio-isotopes within nuclear reactors is generally described by Ransohoff et al. and Bodnarescu (U.S. Pat. Nos. 3,269,915 and 3,396,077, respectively). A description of use of “primary sources” and the general use of neutron sources is described, in detail, by Impink, Jr. (U.S. Pat. No. 4,208,247—issued in June 1980, hereinafter “Impink”), where, preferably, plutonium-238 and beryllium are encapsulated in an alloy that does not allow transmission of thermal neutrons, that is, essentially “black” to thermal neutrons, such as pure cadmium; 65% silver/cadmium or 80% silver/15% indium/cadmium. 
     A reactor start-up neutron source is used to safely assist the initiation of nuclear chain reaction in the initial core loading of nuclear reactors. A reactor startup source is required for safe startup of an initial core containing only fresh unirradiated nuclear fuel because the neutron population density from all sources (e.g., spontaneous fission of the fuel, cosmic radiation, deuterium photoneutrons) is insufficient for reliable monitoring of the reactor neutron population to assure safe reactor start-up. Low neutron fluxes occur in nuclear reactors with initial cores with only mildly radioactive fuel or after prolonged shutdown periods in which the irradiated fuel has decayed thereby reducing the inherent neutron source of the reactor from the previously mentioned mechanisms. Fixed reactor primary and secondary startup neutron sources provide a population of neutrons in the reactor core that is sufficient for the plant instrumentation to reliably measure and therefore provide reactor power and reactivity information to the reactor operator to enable a safe reactor startup and also to the reactor protection system to override the operator and halt the reactor startup if an unsafe situation is detected. Without reactor startup neutron sources, the reactor could suffer a fast power excursion during start-up before the reactor protection system could intervene to terminate the startup. The start-up sources are typically inserted in regularly spaced positions inside the reactor core either in place of some of the fuel rods or within structures inside the reactor core. 
     In addition to the startup of nuclear reactors, neutron sources have many uses in other industrial applications. These industrial uses for neutron sources typically involve the use of the neutron source to create radioisotopes in the vicinity of the source after which the unique nuclear decay characteristics of the radioisotope(s) so created in the process being evaluated are measured and concentrations or compositions are inferred from the measurements in a process typically referred to in the art as neutron activation analysis. The resulting industrial applications include but are not limited to: non-destructive testing of materials, neutron activation analysis, sample moisture analysis, oil well logging, medical treatment of cancer, explosive detection, metal fatigue detection, and other real-time evaluations of chemical composition or moisture content in process streams such as combustion optimization in power plants and cement kilns. 
     Impink (cited previously) further teaches that (at the time of the patent), neutron sources for commercial reactors have been positioned within the nuclear core, and remained within the core, during at least one entire operating cycle. The sources maintained a fixed position. In reactors, sources are inserted in selected fuel assemblies and extend within fuel assembly guide thimbles designed to provide structure for the fuel assembly and provide guidance for the insertion of control elements into the reactor. The sources are also disposed in assemblies close to the core periphery so as to be positioned within the detection range of the detection and monitoring apparatus outside of the reactor vessel. 
     Beryllium is a light weight, strong but brittle, light grey alkaline earth metal. It is primarily used in non-nuclear applications as a hardening agent in alloys, notably beryllium copper. Structurally, beryllium&#39;s very low density (1.85 times that of water), high melting point (1287° C.), high temperature stability and low coefficient of thermal expansion, make it in many ways an ideal high-temperature material for aerospace and nuclear applications. Commercial use of beryllium metal presents technical challenges due to the toxicity (especially by inhalation) of beryllium-containing dusts. Beryllium produces a direct corrosive effect to tissue, and can cause a chronic life-threatening allergic disease called berylliosis in susceptible persons. 
     In the nuclear area, beryllium is an extremely unusual element in that essentially all naturally occurring beryllium is of the  9 Be isotope which has a very low binding energy (1.69 MeV) for its last neutron. The result of this peculiar aspect of the nuclear physics of beryllium is that, when excited by radiation more energetic than the threshold energy shown below, the  9 Be disintegrates as shown below by neutron emission and forms the much more stable helium or carbon atoms.
 
 9 Be 4 + 4 He 2 → 12 C 6 + 1   n   0 E α =0 (exothermic)
 
 9 Be 4 +γ→2· 4 He 2 + 1   n   0 E y ≧1.6 MeV
 
 9 Be 4 + 1   n   0 →2· 4 He 2 +2· 1   n   0 E n ≧1.6 MeV
 
     Californium (element 98) is a rare and exclusively man-made element that is synthesized by long term irradiation of other rare man-made isotopes such as plutonium or curium in specialized high flux reactors specifically designed to produce high-order actinide isotopes. Californium (Cf) is used exclusively for applications that take advantage of its strong neutron-emitting properties. The  252 Cf isotope is, by far, the most widely used isotope of californium for neutron sources due to its high source strength, production yield and relatively long half life. There are currently only two facilities in the world that currently synthesize and separate  252 Cf. At this time, ˜90% of the world&#39;s annual production of ˜200 milligrams is produced at the fifty year old High Flux Isotope Reactor at the Oak Ridge National Laboratory in Tennessee. The  252 Cf produced in the reactor is initially purified at the reactor site by separating the  252 Cf from all of the other actinides and fission products that result from the target irradiation in a complex radiochemical process that is performed remotely in a hot cell laboratory. The separation process is concluded by coating an inert material wire, foil or other form with the  252 Cf chemical compound from the separation process and placing the resulting form in a cask that shields the resulting  252 Cf source material, thereby allowing the material to be removed from the hot cell laboratory. The high neutron strength of  252 Cf makes it necessary for any source manufacturing subsequent to the separation of the Cf from all of the other actinides and fission products to be done remotely in a well shielded facility to protect the manufacturing staff. As a result, it is only practical to employ simple manufacturing processes in the manufacture of neutron sources using  252 Cf. Even in view of previous patents cited, there seems no logical reason to try to add anything to californium as a neutron source as it is already the strongest source of neutrons by weight of any available radioisotope. 
     Referring now to prior art  FIG. 1 , there is shown one embodiment of a typical thermal nuclear reactor including a sealed reactor vessel  10  housing a nuclear core  12  comprised of a plurality of fuel assemblies  14  (shown in  FIG. 2A ). A reactor coolant, such as one including water, enters the vessel through inlet nozzles  16 , passes downward in an annular region between the vessel and a core support structure, turns and flows upward through a perforated plate  20  and through the core  12  and is discharged through outlet nozzles  22 . 
     A fuel assembly  14  is shown in prior art  FIG. 2A  and includes a plurality of fuel pins  24 , containing nuclear fuel pellets  26 , arranged in a bundle. The assembly also includes a plurality of guide thimbles  28  which provide skeletal support for the assembly and which are sized to removably receive control rods  29  of control elements  30 , positionable above and within the core area by means such as electromagnets  32  which act upon shafts  34  ( FIG. 1 ) removably connected to the control elements  30 . 
     The neutron flux within the core is continuously monitored by detection apparatus such as the neutron detectors  36  ( FIG. 1 ) which are located at an elevation aligned with the elevation of the core  12 . The detectors, located external to the vessel, may be fixed or laterally movable by positioning bars  38 . 
     The guide thimbles  28  of the fuel assemblies  14 , in addition to receiving control rods  29 , shown in  FIG. 2A , are sized to receive neutron sources capsules shown in  FIG. 2B . The capsules contain a neutron emitting source  44 . 
     The source  44  includes a major mass of fast neutron emitting material, encapsulated and held in place by cladding  48 . The preferred source material, for current art reactor startup sources is  252 Cf due to a combination of factors including source strength. Nonetheless,  252 Cf source material is extremely expensive and only available in limited quantities, so minimizing the requirements of these materials is very important. The optimal solution for a primary source is one that minimizes the amount of  252 Cf required to accomplish the required function. Additionally, the lifetime of a neutron source is determined by the minimum source strength that achieves the required function. Therefore, it is one of the main objects of this invention to make more efficient use of the  252 Cf to either reduce the amount of  252 Cf required for a source or to extend the useful lifetime of a given amount of  252 Cf. 
     SUMMARY OF THE INVENTION 
     The above problems are solved and objects met by combining a  252 Cf driver source and a beryllium multiplier assembly in a manner that the large majority of the radioactive decay energy from the  252 Cf driver source can be transformed into neutrons by the beryllium multiplier (“multiplier assembly”) and the resulting neutrons can then be multiplied by the beryllium (n,2n) reaction. The invention involves a fast neutron emitting source multiplier assembly, consisting essentially of a driver source of  252 Cf deposited on a surface consisting essentially of foil and wire, and encapsulated and surrounded by a beryllium segment as a multiplier segment. The current art primary source designs utilize only the 3.1% of the decay events of  252 Cf that are spontaneous fission events. The remainder of the decay events are high-energy alpha decays whose energy is completely shielded by the source cladding ( 48 ) that surrounds the  252 Cf source ( 44 ) as shown in prior art  FIG. 2B . The preferred embodiment of the invention driver source is a  252 Cf coated wire or foil embedded in a recess within a simple machined beryllium multiplier. Preferably, the beryllium will be in two parts as shown in  FIGS. 3A and 3B  for ease of insertion of the driver source  68 . The dimensions of the beryllium multiplier are only critical to the extent that the energy of the alpha particle and spontaneous fission products is captured within the beryllium multiplier. Due to the massive, charged nature of these particles, the amount of beryllium necessary to absorb the energy is much less than that necessary to form a structurally adequate container for the driver source assembly. The capture of the energy of the  252 Cf alpha and spontaneous fission decay results in approximately nine-fold increase in neutron source strength per unit mass of  252 Cf driver material relative to current art  252 Cf primary sources. The strength of the invention neutron source can also be modulated by the inclusion of a shield curtain that can be imposed between the  252 Cf driver source and the beryllium multiplier. This shield curtain is capable of stopping alpha particles and interferes with transmission of alpha particles to the beryllium multiplier. 
     Increasing the mass of the multiplier assembly will further increase the neutron source strength by increasing the beryllium (n,2n) reaction resulting from the neutron produced directly from the  252 Cf by spontaneous fission as well as those produced in the beryllium as a result of interactions with the high-energy alpha particles and fission products resulting from the  252 Cf decay. The preferred embodiment encapsulates the multiplier assembly within a hermetically sealed source capsule which includes a means for holding the multiplier assembly together, preferably a spring and void volume to provide space to collect the helium gas that evolves from the beryllium disintegration reaction without over pressurizing the source capsule. In the multiplier of this invention, the neutrons produced directly by the  252 Cf and those produced by the transformation of alpha and fission products by the beryllium multiplier assembly are further multiplied by beryllium (n,2n) reactions before they are emitted from the source assembly. 
     The main innovation of this invention is the combination of  252 Cf, already a strong neutron source, with the heterogeneous beryllium multiplier to complete the transformation of the  252 Cf radioactive energy into neutrons. The manufacture of this invention requires that the  252 Cf driver source be inserted into the multiplier assembly prior to any structural encapsulation. Further, it requires machining and fabrication of metallic beryllium or beryllium oxides. Finally, all of the manufacturing must be performed remotely in the presence of an intense neutron source. 
     The  252 Cf and Be together provide a synergy, allowing weight reduction of  252 Cf from about 260 micrograms to about 30 micrograms, per multiplier assembly, an 8 + x reduction due to beryllium excitation neutron multiplication. 
    
    
     
       BRIEF DESCRIPTION OF THE DRAWINGS 
       The advantages, nature and additional features of this invention will be better understood from the following description, taken in connection with the accompanying drawings, in which: 
         FIG. 1  is a prior art elevation view, partially in section, through the reactor vessel of one embodiment of a typical nuclear reactor; 
         FIG. 2A  is a prior art perspective view of a fuel assembly having a control element inserted therein; 
         FIG. 2B  is a prior art neutron source insert into a fuel assembly; 
         FIG. 3A  is a cross-sectional view of the neutron source capsule of this invention, disposed in a reactor thimble tube; 
         FIG. 3B , which best illustrates the broadest embodiment of this invention as a reactor startup source, is a three-dimensional view of the neutron source showing the  252 Cf, wire and beryllium components. 
     
    
    
     DESCRIPTION OF THE PREFERRED EMBODIMENTS 
     In this invention, a major amount of beryllium will be used to encase/surround/encapsulate a minor amount of  252 Cf, as shown in  FIG. 3A  discussed below. Only  252 Cf and Be are used in the multiplier assembly of this invention. The multiplier assembly consists of  252 Cf coated onto wire or foil and Be. The preferred embodiment of the invention described herein utilizes all of the different types of radiation from the  252 Cf so that they are efficiently transformed into neutrons. Even though the  252 Cf is a very strong neutron source, neutrons are only directly produced as a result of the 3.1% of the decays that are spontaneous fission with an average of 3.77 neutrons emitted per fission. The current art  252 Cf neutron sources render the remaining 96.9% of the  252 Cf radioactive energy as alpha particles useless by dissipating the energy of this energy as heat in the standard source design stainless steel sheath. 
     The preferred embodiment does not use a source sheath, which is also an extremely effective shield for the alpha particle and fission product energy, but rather utilizes a bare wire, typically of palladium, onto which  252 Cf has been deposited after separation from the various irradiation products from the reactor. Instead of the wire being encapsulated in a shield, it is encapsulated in a simple beryllium multiplier assembly which then is directly illuminated with the alpha particles, fission products, prompt fission gammas and high energy neutrons that result from the decay of  252 Cf. As a result, the neutron source strength of the bare  252 Cf coated wire is multiplied by approximately a factor of eight to ten resulting in either a significantly stronger or longer lived source for the same amount of  252 Cf or a ninefold reduction in the amount of  252 Cf required for a constant source strength. Calculations have shown that the typical 600 MBq reactor startup primary source with the current art unmultiplied source requires nearly 260 μg of  252 Cf while the multiplied source requires only 29 μg. 
     Referring now to  FIG. 3A , a primary source capsule  60  is shown including the driver source of  252 Cf, shown as  68  coated onto a substrate wire  69 , and an encasing/surrounding/encapsulating beryllium segment  64 , to provide multiplier assembly  62 . This multiplier assembly  62  is better illustrated in  FIG. 3B . The multiplier assembly  62  can have a wide variety of uses in nuclear power plants, oil well logging and elsewhere. 
     Here, the multiplier assembly  62  consisting of  252 Cf shown as  68 , coated on a substrate/surface  69 , surrounded by Be, shown as  64 , can be inserted or be contained/encased by a surrounding hollow tube/rod  70 . The ends of the primary source capsule can be sealed by top end plug  84  and bottom end plug  84 ′, with a positioning element, most simply a spring  78  holding the contained/encased multiplier assembly  62  in place near or next to the bottom end plug  84 ′. The void volume within the primary source capsules is shown as  86 , and is capable of capturing helium gas released directly by the  252 Cf alpha decay as well as that generated by the beryllium decomposition reactions. 
     While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.