Patent Description:
The present invention relates to a method of fabricating a gamma radiation detector and a system for measuring energies and amplitudes of gamma radiation, and more particularly to an assembly of a plurality of Schottky diodes for measuring fission gamma radiation for power distribution measurements.

Gamma radiation is created by nuclear decay, such as the nuclear fission that occurs in nuclear reactors. Nuclear reactors are equipped with measurement systems to detect and measure gamma radiation. However, the increase in new nuclear fuel and reactor designs present challenges to the ability of existing measurement systems to confirm fuel design performance predictions and to measure operation performance in detail throughout reactor operation in a fuel cycle.

Nuclear reactor design relies on software simulations for core design and fuel performance projections, as well as performance projections for reactor components in both normal operating conditions and in a variety of accident scenarios. The software performance projections are eventually compared to actual measurements under normal operating conditions, or simulations thereof in test reactors, in a commercial or test reactor.

The early generations of light water reactor (LWR) designs utilized miniature fission chambers that continuously measured neutron flux along the length of roughly one third of the fuel assemblies in the reactor to benchmark the power distribution measurement and the core design software methods. This type of measurement system is referred to as a Movable In-core Detector System (MIDS). The finely-spaced axial neutron distribution measurement resolution (~<NUM> (~<NUM> inches)) provided by MIDS measurement systems allowed the identification and diagnosis of many fuel performance issues that caused significant reactor operation issues, such as fuel rod bow, debris deposits on the outside of fuel rods that affect heat transferor reactivity, and inlet flow distribution asymmetries. The primary negative operational issues associated with the use of a MIDS is the complexity, required size, and operation and maintenance costs associated with use of the associated piping and sensors.

The current generation of reactor power distribution measurement systems, referred to as fixed in-core detector (FID) systems, rely on a limited number of radiation sensors in fixed axial and radial locations inside the reactor core. <FIG> depicts aspects <NUM> of the use of FID detector systems compared to an MIDS detector system. As indicated on <FIG>, the signals from a FID sensor represent the average reactor power over the fixed axial region defined by the active length of the sensing element. One example of a typical Rh FID detector system configuration <NUM> illustrates axial spacing <NUM> of the detectors at about <NUM> (<NUM> inches). A second example of an OPARSSEL V detector system configuration <NUM>, illustrates axial spacings <NUM> of the detectors at about <NUM> (<NUM> inches). The radial distribution of fuel assembly FID measurements requires, and is constrained to, locations associated with penetrations in the reactor vessel. The necessary averaging and radial distribution location constraints of FID systems makes it difficult to observe the differences between highly localized and detailed measurements versus predicted power distribution and long term operating characteristic. For example, the overlaid flux distribution graph <NUM>, depicts a measured flux distribution versus tracking point or reactor core depth. For data obtained using a multiple MIDS detector configuration <NUM>, a finer resolution of depth versus flux measurement is more readily obtained than would be possible by either Rh FID configuration <NUM> or OPARSSEL V FIG configure <NUM>. The ability to detect these differences could be crucial qualifying the design of reactors and in the safe operation of both the fuel and the reactor. <CIT> discloses a silicon carbide Schottky diode solid state radiation detector that has an electron donor layer such as platinum placed over and spaced above the Schottky contact to contribute high energy Compton and photoelectrical electrons from the platinum layer to the active region of the detector to enhance charged particle collection from incident gamma radiation.

Further, the nuclides produced during reactor operation may be detected and determined by the use of various types of spectrographic measurement including neutron and gamma radiation spectroscopy. The accuracy of performing gamma radiation spectroscopy using existing methods and equipment is often limited by the interactions of different radiation energies in the active volume of the detector that produces a continuum of pulse interest. It is also useful to gamma radiation spectrographic sensors that are small enough to place them in a number of locations within a reactor. Disclosed herein is a method and system that will allow a much clearer representation of the gamma energies and intensities being emitted from the material being analyzed than is currently achievable with other solid- state gamma detector spectroscopy systems.

The following summary is provided to facilitate an understanding of some of the innovative features unique to the embodiments disclosed and is not intended to be a full description. A full appreciation of the various aspects of the embodiments can be gained by taking the entire specification, claims, abstract and drawings as a whole.

There is provided a method of fabricating a gamma radiation detector, as claimed in claim <NUM>. There is also provided a method of fabricating a gamma radiation detector array, as claimed in claim <NUM>. There is further provided a system for measuring energies and amplitudes of gamma radiation, as provided in claim <NUM>. The system may include an elongate housing for placement within a nuclear reactor. Each detector may be positioned axially within the housing in a radially spaced relationship relative to each adjacent detector.

Each gamma radiation detector may include a first lead extending upwardly from the radiation detector proximate the source material and a second lead extending upwardly from the radiation detector proximate the Ohmic contact layer. The detectors may be spaced within the housing such that the first and second leads of each detector are spaced away from the first and second leads of each of the other of the plurality of detectors.

Each detector may be covered by an intermediate layer and an outer layer. In various aspects, the intermediate layer is an aluminum oxide layer. In various aspects, the outer layer is a stainless steel layer.

The distance between the Schottky contact and the layer of the Compton and photoelectron source material may be adjusted to detect only the highest energy prompt fission gamma radiation. For example, the distance may be adjusted to detect fission gamma radiation greater than about <NUM> MeV.

The characteristics and advantages of the present disclosure may be better understood by reference to the accompanying figures.

As used herein "axially" means in the direction of or in alignment with an axis. With respect to two or more objects, axially means the objects are positioned along an axis, either in a co-axial alignment or parallel to an axis.

As used herein, "radially spaced" means two or more objects are positioned such that the objects are spaced from each other along an arc of a circle, or placed along a radius.

<FIG> illustrates an assembly <NUM> for in-core power distribution detection that is suitable for power distribution measurements for more recent nuclear fuel and reactor designs is described. The assembly <NUM> uses a plurality of gamma detectors <NUM>' (depicted in <FIG> and <FIG>), preferably positioned axially, one on top of another, within an elongate container, such as a tube <NUM>. The detectors <NUM>' each have two leads <NUM> and <NUM>. The axial stack of detectors <NUM>' are rotated relative to each other along the length of the elongate container <NUM> so that the leads of each detector <NUM>' do not interfere with the leads of the other detectors, and to maximize the density of measurements that can be obtained.

In various aspects, the detectors <NUM>' depicted in <FIG> and <FIG> may be modifications of those described in <CIT> entitled "Solid State Radiation detector with Enhanced Gamma Radiation Sensitivity". The previously disclosed gamma detector <NUM>, depicted in <FIG>, includes a Schottky diode having an active semiconductor region and a Schottky contact over at least a portion of the semiconductor region. The detector <NUM> described herein and in <CIT>, includes an Ohmic contact layer <NUM>, made for example of tungsten, positioned beneath a silicon carbide conducting be construed in light of the number of reported significant digits and by applying ordinary rounding techniques.

In various aspects, the detectors <NUM>' depicted in <FIG> and <FIG> may be modifications of those described in <CIT> entitled "Solid State Radiation detector with Enhanced Gamma Radiation Sensitivity" and which is incorporated herein by reference in its entirety and for all purposes. The previously disclosed gamma detector <NUM>, depicted in <FIG>, may include a Schottky diode having an active semiconductor region and a Schottky contact over at least a portion of the semiconductor region. The detector <NUM> described herein and in <CIT>, includes an Ohmic contact layer <NUM>, substrate <NUM>, which is covered by a layer of epitaxial silicon carbide <NUM>. The substrate <NUM> in various aspects is approximately <NUM> microns in thickness and the epitaxial layer <NUM> in various aspects is approximately from <NUM> to <NUM> microns in thickness. The epitaxial silicon carbide layer <NUM> is covered by a Schottky contact <NUM>. The Schottky contact <NUM> may be formed from any highly conductive metal such as platinum or gold, about <NUM> micron in thickness.

A thin layer of a Compton and photoelectron source material <NUM>, made for example, from platinum or another suitable high atomic donor material, such as lithium fluoride or tungsten, is positioned above and spaced from at least a portion of the Schottky contact <NUM>, defining a gap <NUM> that will, in response to incident gamma radiation, release electrons that will penetrate the active region <NUM> and contribute to the collection of charged particles in the region <NUM>. In response to incident gamma radiation, the source or electron radiator material <NUM> will release electrons that will penetrate the active region <NUM> and contribute to the collection of charged particles in the active region.

The distance defined by gap <NUM> between the Schottky contact <NUM> and added source layer <NUM> is adjustable, and preferably includes a fluid <NUM> with a low effective atomic number and negligible conductance, such as the properties of air at <NUM> atmosphere of pressure with a relative humidity less than or equal to <NUM> percent at <NUM> ° F (<NUM>), between the electron donor layer <NUM> and the Schottky contact <NUM>. Any such fluid with known density and electron attenuation properties may be used in gap <NUM>.

The gap <NUM> between the source layer <NUM> and the Schottky contact <NUM> ensures that only electrons produced by gamma radiation of a desired energy will contribute to the measured signal.

The material used and the thickness of the source layer <NUM> are selected based upon the energy range of the gamma radiation that is targeted to be detected by the end-user. The addition of an adjustable electron donor layer (symbolically represented by a telescoping sleeve surrounding the layer of fluid <NUM>), i.e., adjustable in thickness and distance from the Schottky contact <NUM>, allows the gamma radiation to interact with the electrons surrounding the source atoms in the donor material <NUM> to produce high energy Compton and photoelectrical electrons inside the donor layer that penetrate into the active region <NUM> of the silicon carbide detector <NUM>. The thickness of the intervening fluid <NUM> in the gap <NUM> controls the energy of the donor electrons so that they are collected in the active region.

The charge deposited over a fixed amount of time will be proportional to the energy of the gamma radiation incident upon the layer <NUM>, so both gamma energy and gamma radiation intensity can be determined from the proper analysis of the electrical outputs from the silicon carbide device.

As shown in <FIG> and <FIG>, the detector <NUM> described above may be modified (detector <NUM>') to include an insulating layer <NUM> between the Ohmic contact layer <NUM> and the conducting layer <NUM>. The insulating layer <NUM> prevents discharge of electrons and short circuiting of the tungsten Ohmic contact <NUM>. In various aspects, the active regions of the modified detector <NUM>' include the SiC layers <NUM> and <NUM>. Epitaxial layer <NUM> may be made of SiC that has been lightly doped with a source of additional electrons, n-. In this context, light doping may correspond to a concentration of electron donating elements on the order of about <NUM><NUM> cm-<NUM> to about <NUM><NUM> cm-<NUM>. Conducting layer <NUM> may be formed from SiC that has been more heavily doped with a source of additional electrons, n+. In this context, heavy doping may correspond to a concentration of electron donating elements on the order of about <NUM><NUM> cm-<NUM> to about <NUM><NUM> cm-<NUM>. The heavily doped region may insure better electron transport to the ohmic contact. A reverse bias depletes the electrons in the n- region <NUM> and ionizing radiation produces electron-hole pairs in the depleted region. The charge may then be collected at the ohmic contact under the influence of an applied voltage across the contacts.

In an alternative aspect, the active region may comprise epitaxial layer <NUM> comprised of two regions, one positively doped layer of about <NUM> micron in thickness and one negatively doped layer of about <NUM>-<NUM> microns in thickness. The conductive layer <NUM> may in various aspects include hydrogen ions.

The detectors <NUM>' are very small, less than about <NUM><NUM>. The SiC detectors are preferably configured to only detect the highest energy prompt fission gamma radiation(> ~<NUM> MeV) by adjusting the distance between the source material <NUM> and the n- region of the SiC, epitaxial layer <NUM>.

The exterior surface of detector <NUM>', referring again to <FIG> and <FIG>, is covered by an intermediate layer <NUM>, which in various aspects is made of aluminum oxide (Al<NUM>O<NUM>) or magnesium oxide (MgO), and acts as an electrical dielectric and isolates the conductive materials in the detector from each other in addition to the stainless steel enclosure. The intermediate layer <NUM> is preferably covered by an outer layer <NUM>, made of stainless steel or Iconel™, and which acts as an enclosure providing structural integrity to the detector <NUM>'.

A first lead <NUM>, also covered by the aluminum oxide layer <NUM> and the outer stainless steel layer <NUM>, extends from the top of detector <NUM>' near the source material layer <NUM> at point <NUM>. A second lead <NUM>, also covered by the aluminum oxide layer <NUM> and the outer stainless steel layer <NUM>, extends from the bottom of detector <NUM>' near the Ohmic contact layer <NUM> at point <NUM>. The first and second leads <NUM> and <NUM> are on opposite sides of the detector <NUM>'. In a radial configuration, the first and second leads <NUM> and <NUM> of a detector <NUM>' may, for example, be positioned <NUM>° apart from each other. The distance between the first and second leads is denoted in <FIG> and <FIG> by ΔV, representing the difference in voltage.

<FIG> and <FIG> illustrate the arrangement of detectors <NUM>' that comprise an embodiment of the assembly <NUM> for power distribution measurements. The detectors.

<NUM>' are shown as being housed in an elongate container, such as an instrument tube <NUM> that would be positioned in use in a reaction vessel, for example, adjacent the fuel rods. A distance between the interior surface of the tube <NUM> and the exterior surface of the detector <NUM>' defines an open space <NUM> filled with air, argon or another inert gas.

<FIG> is an axial schematic of assembly <NUM>, showing just three axially spaced detectors for illustrations purposes. <FIG> is a cross-section view of the assembly <NUM> shown in <FIG>, showing the tube <NUM> and detector <NUM>' arrangement, showing three sets of leads <NUM>/<NUM> separated by distances denoted ΔV1, ΔV2, and ΔV3, and another set denoted by ΔVn, to indicate that n may be any number of additional detectors <NUM>'. In use, there would be a plurality of detectors <NUM>'. For example, as many as <NUM> detectors <NUM>' can be placed in a typical instrument tube <NUM> used in a nuclear reactor (not shown). The number of detectors <NUM>' in any tube <NUM> will vary depending on the length of the tube and the measurement needs of the reactor. The detectors <NUM>' would be arranged such that the leads <NUM>/<NUM> of each detector <NUM>' are radially spaced from the leads <NUM>/<NUM> of the rest of the detectors <NUM>' in the tube <NUM>.

This approach will essentially eliminate the contribution of fission product gamma radiation in the measured signals. In various aspects, the SiC detectors <NUM>' are positioned at different closely spaced positions inside and along the length of a dry tube <NUM> that has an outer diameter small enough to fit, for example, inside the fuel assembly central instrument thimble or other strategic measurement location inside or around the reactor and/or inside the reactor vessel. As used herein, "closely spaced" means less than about <NUM> (twelve inches), and in various aspects, less than or equal to about <NUM> to <NUM> (two to three inches), and preferably about <NUM> (<NUM> inches) or less, equivalent to or less than the spacing in the finely-spaced axial neutron distribution measurement resolution (~<NUM> (~<NUM> inches)) provided by MIDS measurement systems.

There is a SiC signal response that will have been predicted by the core design software for each of the many SiC detectors <NUM>' located within or around the reactor core. The use of the small, closely spaced, SiC detectors <NUM>' essentially eliminates the effects of averaging flux measurements over long fuel assembly lengths, as shown in <NUM> of FFIG. As explained previously, the necessary averaging and radial distribution location constraints of the currently used FID systems makes it difficult to observe the differences between highly localized and detailed measurements versus predicted power distribution and long term operating characteristic. The configuration of SiC detectors <NUM>' described herein will closely mimic the distribution of the MIDS movable fission chamber axial measurement density (e. - one per <NUM> (<NUM> inches)), and will have the ability to resolve the impacts that the presence of things like grids, debris deposits on the outside of fuel rods that affect heat transfer or reactivity (e.g., CRUD), and local boiling have on the reactor core. This allows a more accurate synthesis of the axial flux distribution to be developed. The differences between the measured and predicted detector <NUM>' signals can be used to determine the accuracy of the core design modeling tools in high detail, and produce a highly detailed core power distribution measurement that can be used to identify the presence of fuel performance anomalies.

The detector <NUM> assemblies can be positioned permanently inside all of the fuel assembly instrument thimbles or inside prepared positions in the reactor fuel matrix.

The signal leads <NUM>/<NUM> used to output the detector signals in the detector tubes <NUM> are oriented as shown on <FIG> and <FIG> to allow the maximum possible SiC sensing element active volume surface area and axial density while using standard minerally insulated cable designs for the signal leads. The SiC signals will utilize a common reactor ground to simplify the configuration of the electrical connector that joins the measured voltage differences to the signal processing electronics.

The novel aspects of the detector assembly <NUM> design described and shown herein includes, for example:.

In principle the radiation detector assembly <NUM> design and configuration shown in <FIG> and <FIG> can displace all other radiation sensors used in nuclear reactor operations. Moreover, the effort to transition from current measurement systems to the radiation detector assembly <NUM> described herein should be very easy and cost effective.

In a nuclear reactor, a rate of flow of the coolant through the fuel channels will be known and can be continuously or periodically measured, as desired, by known techniques. A fission product that would be expected to be present in sufficient amounts in the event of a leak in a cladding tube to generate measurable gamma radiation may be chosen. An exemplary fission product is La<NUM> because it is one of the most prevalent fission products in a reactor using UO<NUM> as the fissile material. Further, it may be useful to monitor changes in the presence of other fission products in the coolant that produce relatively low energy gamma radiation, such as the prompt n-y emitted from Xe<NUM>. Since this product may be much more likely to escape from the fuel matrix through a fuel cladding defect, it may be present in higher concentrations in the coolant than the La<NUM> in the event a leak develops in the fuel cladding. Therefore, it would be useful to employ an array of gamma detectors able to identify the types of gamma-emitting products through spectrometry. Additionally, some neutron irradiated materials may form nuclides capable of emitting gamma radiation. Gamma energy and intensity measurements may provide the information that is needed to determine the composition of the neutron irradiated elements, along with their quantities, using Neutron Activation Analysis (NAA) techniques well known to those skilled in the art.

Some examples of elements present in a neutron irradiated material sample can be determined by the measurement of a gamma energy spectrum like that shown on <FIG>. The gamma energy and intensity measurements provide the information that is needed to determine the elements and quantities present in irradiated material using Neutron Activation Analysis (NAA) techniques well known to those skilled in the art. The suppression of the continuum noise observed in <FIG> may allow a more accurate determination of the intensity of a single gamma energy peak than is currently achievable using current gamma spectroscopy equipment. Continuum noise suppression may allow a more accurate measurement of fission product concentration changes in the fuel channels as described in <CIT>. Such measurements may be useful to identify the presence and axial location of a fuel defect in a fuel channel.

The use of a solid-state radiation detector configured to be particularly sensitive to gamma radiation may allow the creation of a gamma radiation spectrogram from a neutron irradiated material sample, such as exemplified in <FIG>. Such a detector is described in <CIT>, previously referenced and further shown schematically in <FIG> and <FIG>. Each SiC detector can be configured as a Single Channel Analyzer (SCA) (e. - ORTEC® 550A Single Channel Analyzer) input to allow each detector to cover a very narrow gamma energy range. The energy range may be determined according to the depth of gap <NUM> between the electron emitter <NUM> and the Schottky contact <NUM>, and the depth, Te, of the n- active region <NUM>.

The gap <NUM> between the electron emitter <NUM> and Schottkey contact <NUM> includes a fluid <NUM> interposed between the Schottky contact <NUM> and the layer of the Compton and photoelectron radiation material <NUM>, as disclosed above. Gamma radiation <NUM> impinging on the electron emitter <NUM>, will create electrons having an energy related to the energy of the impinging gamma radiation <NUM>. Low energy gamma radiation may result in corresponding low energy Compton or photoelectrically scattered electrons. Low energy may be defined as the energy resulting in electrons unable to transit through the entire thickness of gap <NUM>, and therefore will fail to enter into the active n-region <NUM>. As a result, such low energy electrons may not be detected by the SCA. An array of such detector elements may include individual elements, each having a differing thickness of gap <NUM> between the electron emitter <NUM> and the Schotkky contact <NUM>. Because the size of the gap <NUM> determines the lower energy cut-off for a detector, the array may include a number of detectors elements that differ in their lower limit energy detection capabilities.

The value of Te, the depth of the n- layer <NUM>, may be chosen to define an upper limit of the Compton or photoelectrically scattered electrons <NUM> that may be detected by the sensor. Te may be calculated according to following empirical formula given by <NPL>: <MAT> where Eβ is the maximum beta energy in MeV. The ability to stop beta radiation depends primarily on the number of electrons in the absorber (i.e., the areal density, which is the number of electrons per cm<NUM>). Hence, the range when expressed as a density thickness (g/cm<NUM>) of the material gives a generic quantifier by which various absorbers can be compared.

It may be understood that all electrons <NUM> having an energy that results in a transit distance greater than Te along the thickness of the active volume <NUM> and <NUM> will produce pulses in the SCA with essentially the same amplitude. These high energy electrons <NUM> may result from Compton Scattering and the photoelectric effect corresponding to high energy gamma radiation <NUM> impinging on the electron radiator layer <NUM>. The resultant high energy electrons <NUM> may have sufficient energy to completely penetrate the thickness, Te, of the active region <NUM> of the detector to impinge on the back ohmic contact <NUM>. The pulses associated with these events will all have essentially the same pulse amplitude and may be removed by the upper level discriminator component of the SCA. It may be understood that the thickness of Te therefore determines the upper range of energy detected by the sensor.

Claim 1:
A method of fabricating a gamma radiation detector, the method comprising:
providing a SiC radiation detector (<NUM>, <NUM>') comprising:
a Schottky diode having an active semiconductor region (<NUM>, <NUM>) and a Schottky contact (<NUM>) over at least a portion of the active semiconductor region;
a layer of a Compton and photoelectron source material (<NUM>) configured to react with incident gamma radiation to emit Compton and photo-electric electrons to penetrate into the active semiconductor region (<NUM>, <NUM>) of the Schottky diode through the Schottky contact (<NUM>), the layer of the Compton and photoelectron source material (<NUM>) being supported above the Schottky contact (<NUM>); and
a layer of fluid (<NUM>) interposed between the Schottky contact (<NUM>) and the layer of the Compton and photoelectron source material (<NUM>);
adjusting a distance between the Schottky contact (<NUM>) and the layer of the Compton and photoelectron source material (<NUM>) thereby determining a minimum detection energy of the SiC radiation detector (<NUM>, <NUM>');
fabricating the active semiconductor region (<NUM>, <NUM>) to have a specified thickness, thereby determining a maximum detection energy of the SiC radiation detector (<NUM>, <NUM>'); and
contacting the SiC radiation detector (<NUM>, <NUM>') to a charge input of a single channel analyzer;
characterized in that
adjusting the distance between the Schottky contact (<NUM>) and the layer of the Compton and photoelectron source material (<NUM>) comprises actuating a telescoping sleeve in contact with the Compton and photoelectron source material (<NUM>) and surrounding the layer of fluid (<NUM>).