Characterization of a Portable Neutron Coincidence Counter
Neutron coincidence counting is a technique widely used for qualitative and quantitative analysis of nuclear material. Because different isotopes possess different coincident neutron characteristics, the coincident neutron signature can be used to identify and quantify a given material. In an effort to identify unknown nuclear samples in field inspections, a new portable neutron coincidence counter has been developed. An indepth analysis has been performed to establish whether the nuclear material in an unknown sample could be quantified with some confidence. The analysis was performed by comparing the true measurements of the system to the calculated output produced using MCNPX and the neutron coincidence point model. Based on the analysis, it is evident that this new portable system can play a useful role in identifying nuclear material for verification purposes. Introduction Neutron coincidence counting is an important passive, nondestructive-assay method that can be used to characterize nuclear material in bulk samples, namely plutonium. Because the number of neutrons emitted simultaneously during the fission process is particular to each isotope, the coincidence neutrons create a useful signature that can be used to describe the isotopes in a sample. Coincidence measurements can be done regardless of background or (α, n) neutrons making them less sensitive to external factors. For plutonium, Pu, Pu, and Pu all have dominant spontaneous fission yields producing coincidence neutrons. The measurement of these spontaneous fission neutrons, however, is complicated by the presence of induced fission neutrons. Induced fission is dependent on the fissile isotopes, primarily Pu, and is caused by the absorption of neutrons in the sample. Both spontaneous fission and induced fission produce coincidence neutrons. Because these neutrons are emitted practically simultaneously, they create a useful signature that can be used to quantify a particular nuclear material. The Portable Neutron Coincidence Counter was developed by Los Alamos National Laboratory to aid inspectors in field measurements of bulk samples. In order to benchmark and characterize the detector, a computer model was created using the Monte Carlo N-Particle eXtended (MCNPX) code and its embedded advanced multiplicity capabilities. The results determined using MCNPX were then compared to the results calculated using the traditional Neutron Coincidence Point Model. This paper discusses both these results and their implications. Background The PNCC consists of four individual high-density polyethylene slabs. The slabs are approximately 7-in. (length) x 10-in. (height) x 3-in. (width) and weigh approximately 8
SOURCES: a code for calculating (α,n), spontaneous fission, and delayed neutron sources and spectra
In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay {alpha}-particles in ({alpha},n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO{sub 2}, ThO{sub 2}, MOX, etc.), enrichment plant operations (UF{sub 6}, PuF{sub 4}, etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of {alpha}-emitting source material and low-Z target material) and in interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of targetmore » material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an analysis of the contributions to that source by each nuclide in the problem.« less
Microwave source with enhanced radiation power based on orbitron MASERs
We have proposed the design of high power microwave source in a vacuum tube based on the theory of negative mass instability. The proposed source does not operate based on a slow-wave structure but instead makes use of accelerated charged particles. In our structure, an electron emitter injects the charged particles into to the cavity. Emitted particles move around the anode, which leads to electron acceleration and emission of electromagnetic waves with a specific frequency. In this paper, we have enhanced the output power and frequency of microwave source by variation of particles initial energy and the number of anodes incorporating their specific arrangement and applied voltages. Working at room temperature, compactness, wide bandwidth, and tunability of output frequency are among the advantages of the proposed source.