Patent Number: 044938101
Section: summary

BACKGROUND OF THE INVENTION The present invention relates generally to passive methods and apparatus for measuring fissionable materials which emit neutrons and relates more particuarly to passive methods and apparatus for measuring spent fuel assemblies. Assaying (i.e., measuring the fissionable content of) spent fuel is important for nuclear safeguards in order to prevent unauthorized diversion of the nuclear material, to provide information necessary for criticality control of spent-fuel storage pools, and to provide process control for reprocessing and reactor operation. Measurement techniques for the assay of spent fuel have included passive and active neutron methods and passive gamma-ray methods. In active neutron methods, an external isotopic neutron source is used to interrogate the spent fuel, whereas in passive methods, no external neutron source is used. When an assay is performed, the goal is to determine the fissile content (i.e., the amounts of uranium and plutonium) in the material being assayed. It is well known in the art that the ratio of Pu/U and fissile content can be correlated to both burnup and reactivity, where burnup is a measure of the number of fissions which occurred in the fuel while the fuel was within the reactor and where reactivity is related to burnup. This is disclosed for example in S. T. Hsue et al., "Nondestructive Assay Methods for Irradiated Nuclear Fuels," Los Alamos Scientific Laboratory Report LA-6923, January 1978. Reactivity depends upon how long the fuel has been in the reactor and not upon cooling time (which is the length of time the fuel has been out of the reactor). Fissile content, reactivity, and burnup can be correlated. Therefore, these quantities can be considered equivalent because the measurement of one quantity can provide a means of determining the other quantities. Passive gamma-ray measurements and passive neutron measurements have been correlated with burnup and have provided a means of verifying the fissile material inventory of spent fuels. However, passive gamma-ray assay is not sensitive to the interior of the spent fuel assembly and therefore cannot truly verify the integrity of the interior fuel rods. Passive gamma-ray assay makes the assumption that the interior fuel rods are present in the fuel assembly and have not been tampered with. The technique of passive neutron assay relies on correlations between the neutron emission rate and declared burnup to determine the fissile content of the spent fuel assembly. The neutron emission rate is N passive=M.multidot.S, neutron rate where M is the multiplication of the assembly (which is defined as M.tbd.1/(1-k.sub.eff) where k.sub.eff is the effective multiplication constant of the fuel asembly) and S is the spontaneous fission rate of the isotopes of Pu and Cm and the emission rate due to the (.alpha.,n) reactions. For pressurized-water reactor (i.e., PWR) fuel assemblies, good correlation between the U and Pu content and the neutron rate has been observed, although a substantial fraction of the neutron emission rate is due to Cm isotopes. This good correlation is due primarily to the fact that the Cm production rate is a uniform function of burnup. However, good correlations do not always exist for boiling-water reactor (BWR) fuel assemblies because (as has recently been shown by T. Yokoyama et al. in "Measurement and Analysis of Neutron Emission Rate for Irradiated BWR Fuel," Journal of Nuclear Science and Technology, 18, pp. 249-260 (April 1981)) the relationship between the neutron emission rate and burnup for these assemblies is double-valued (rather than a single-valued functional relationship). The transuranic production chain depends on the thermal-to-epithermal neutron ratio in the irradiation environment; and in BWR reactors this ratio depends upon the void fraction. For the same burnup, the upper portion of the fuel assembly has more transuranium nuclides than the lower portion because the increased void fraction in the upper portion of the fuel assembly causes the neutron energy spectrum to be harder (i.e., have higher energy) than in the bottom portion of the fuel assembly. Active assay systems (wherein an external isotopic neutron interrogating source or external fissile material is used) have long been considered to give the best assay results. A determination of the amount of fissionable material and a determination of the reactivity can both be made by using active assay techniques. Here, an external neutron source or a neutron source and external fissile material induces the fissions in the U and Pu isotopes. The fission neutrons are detected and correlated to either burnup, the fissile content, or reactivity. The interrogating sources can be either isotopic sources (for example, .sup.252 Cf, .sup.124 Sb-Be, AmLi, etc.) or can be accelerator sources (for example, sealed-tube neutron generators). Systems based on prompt and delayed neutron counting using either .sup.252 Cf, .sup.124 Sb-Be, or neutron generators have been designed and tested; however, these systems require strong sources (e.g., 3Ci .sup.252 Cf and 800Ci .sup.124 Sb-Be) so that the induced signals can be measured in the presence of strong passive neutron emission rates (which are the noise in the system). For high burnup and freshly discharged reactor fuel, the strength of the interrogating source needed to overcome the passive neutron rate can be prohibitively large. Furthermore, when an active system is used to assay, either the fuel material being assayed must be scanned or multiple isotopic sources must be used. Additionally, with active assay systems the measurement geometry is limited because the neutron source is usually a point source. Therefore, despite the assay systems which have been available in the prior art, a need has existed until now for a method and apparatus which has the measurement capability of an active system for measuring both reactivity and content of fissionable material but which does not require scanning nor use of an external isotopic neutron source or external fissile material and which in particular does not require use of large isotopic sources. SUMMARY OF THE INVENTION Objects of this invention are an apparatus and method for measuring both reactivity and quantity of fissionable material without requiring scanning and without the use of any external isotopic source or external fissile material. Other objects of this invention are a method and apparatus which are capable of assaying any fissionable material which emits neutrons. Other objects of this invention are a method and apparatus for assaying any spent-fuel assembly, including spent fuel which has been freshly discharged from a reactor. Still further objects of this invention are a method and apparatus for assaying in which the geometry being used in the measurement can be chosen as desired. Other objects of this invention are a method and apparatus for measuring the fissile content of fuel rods or assemblies before they are placed into cooling pond storage arrays, without using large isotopic sources, as would be required in active systems of the prior art. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects, and in accordance with the purposes of the present invention, as embodied and broadly described herein, the method of nondestructively and non-invasively (i.e., using no internal probing) measuring any material which emits neutrons and has fissionable components, using no external neutron emitting interrogation source comprises: (a) making a first measurement and a second measurement of the neutron count rate of the fissile material being measured while the fissile material is located within a multiplying system (defined below) by using a suitable neutron or gamma detector, wherein the first measurement is made with a reflector material located adjacent to and in close proximity to the multiplying system being measured and the second measurement is made with a different reflector material located adjacent to the multiplying system, (where the adjacent reflector material can be a vacuum); and then (b) using the two measurements made in step (a) to correlate to either reactivity, burnup, or fissile content of the material being measured. The multiplying system is defined to be all fissile material within an arbitrarily chosen physical boundary, together with any moderator material or other material inside the physical boundary which may come in contact with the fissile material being measured. Thus, the multiplying system being measured includes the fissile material being measured and any moderator material (for example, water) or other material (e.g., NaK in an LMFBR fuel assembly, wherein the NaK is not a strongly moderating material) which comes in contact with the fissile material being measured and which is within the chosen arbitrary physical boundary. A critical feature of the method of the invention is that the multiplying system within the physical boundary is the same when the first and second measurements are made and that only the reflector material is changed. It is also to be understood that all measurements are made after the material being measured is neutronically isolated from other materials. The present invention also comprises, in accordance with its objects and purposes, an apparatus for nondestructively and non-invasively measuring at least one quantity selected from the group consisting of burnup, fissile content, and reactivity of any material which emits neutrons, which has fissionable components, and which is located within a multiplying system, without using any external neutron-emitting interrogation source, wherein the multiplying system is bounded by an arbitrarily chosen physical boundary and wherein the multiplying system comprises the fissile material and other material which may come in contact with the fissile material, the apparatus comprising: (a) a support means for supporting the multiplying system in a fixed, reproducible position; (b) a detection system comprising at least one detector which detects at least one type of particle selected from the group consisting of neutrons and gamma rays, wherein the detection system is located in a fixed position spaced apart from said multiplying system and wherein the detection system can be operably connected to a means for operating the detection system; (c) a first reflector material (i) to be located in a reproducible first position such that at least a portion of the first reflector material is located adjacent to but not necessarily surrounding the multiplying system while a first measurement is made with the detection system, and then (ii) to be next removed from the first position; (d) a second reflector material to be positioned in substantially the first position when the first reflector material is removed from the first position and while a second measurement is made with the detection system; and (e) a means for alternately positioning the first reflector material and the second reflector material adjacent to the multiplying system. It is emphasized that here only the reflector material is changed. Although in active assaying systems two measurements of count rates are made, an external isotopic source is used in active assaying for one of these measurements. This is quite different from the method and apparatus of the present invention wherein no external isotopic source is used and wherein only the self-interrogation of the system is used. In the present invention, instead of an external source, the system acts as an internal sink or source when the reflector material is changed because the reactivity has changed (thus changing the number of neutrons returning into the material being measured). Here, the fissile material being assayed is used as the neutron source itself. By using the method and apparatus of the invention, a much simpler, safer, and more elegant way of assaying the material being measured is possible. No external isotopic source is needed, and in particular no large isotopic source is needed to overcome any noise problem. Furthermore, no scanning is required and the measurement geometry can be chosen as desired. No cooling time correction is needed (as is required in prior art passive methods). Until the development of this method, active systems provided the best assay measurements because both reactivity and amount of fissionable material were capable of being measured with the same system. That same measurement capability is also an attribute of the method and apparatus of the present invention, as is the advantage over active systems that here there is no problem with large external sources and excessive passive neutron rates.