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
Comparison of (alpha, n) thick-target neutron yields and spectra from ORINGEN-S and SOURCES
Both ORIGEN-S and SOURCES generate thick-target neutron yields and energy spectra from ({alpha},n) reactions in homogeneous materials. SOURCES calculates yield and spectra for any material containing {alpha}-emitting and ({alpha},n) target elements by simulating reaction physics, using {alpha}-emission energy spectra, elemental stopping cross sections, ({alpha},n) cross sections for target nuclei, and branching fractions to produce-nuclide energy levels. This methodology results in accurate yield and spectra. ORIGEN-S has two options for calculating yields and spectra. The UO{sub 2} option (default) estimates yields and spectra assuming the input {alpha}-emitters to be infinitely dilute in UO{sub 2}. The borosilicate-glass option estimates yields from the total input material composition and generates spectra purportedly representative of spectra generated by {sup 238}Pu, {sup 241}Am, {sup 242}Cm, and {sup 244}Cm infinitely dilute in borosilicate glass, even if none of these four {alpha}-emitters are present in the input material composition. Because yields from the borosilicate-glass option in ORIGEN-S are based on entire input material composition and are reasonably accurate, the same is often assumed to be true for spectra. The input/output functionality of the borosilicate-glass option, along with ambiguity in ORIGEN-S documentation, gives the incorrect impression that spectra representative of input compositions are generated. This impression is reinforced by wide usage of the SCALE code system and its ORIGEN-S module and their sponsorship by the US Nuclear Regulatory Commission.