Neutron Multiplicity Counting Research Articles

Using neutron angular anisotropy information to dynamically determine the ratio of the (α,n) rate to spontaneous fission rate for coincidence counting applications

Typically, when neutron coincidence or multiplicity counting is performed, there are three unknowns: the sample mass, the leakage self-multiplication, and the ratio of the (α,n) rate of the source to the spontaneous fission rate. For a given counting time, the strength of the source or the detector efficiency must be sufficiently high for the singles, doubles, and triples count rates to represent statistically meaningful quantities. Often, the strength of the source and the allotted counting time are such that only the singles and doubles count rates are statistically meaningful. In this latter case, the ratio of (α,n) to the spontaneous fission must be estimated through some other means. With a simulated (α,n) rate, the two equations related to the singles and doubles count rates can be used to determine the sample mass. In order to determine the ratio of (α,n) to spontaneous fission rate of the source, the isotopic composition of the sample as well as the light element impurities inside the source must be known. Ideally, there would be a way to dynamically determine the (α,n) rate of the source from count rate information. In this paper, it is shown that the (α,n) rate of the source can be determined by using information about the ratio of the number of coincident neutrons at 180° to the number at 90°. By using this information, the three aforementioned unknowns can be dynamically determined through the sole use of singles and doubles count rates.

Verification and validation of the MCNPX-PoliMi code for simulations of neutron multiplicity counting systems

Neutron coincidence counting is widely used in nuclear safeguards. Simulations of these systems can be performed using Monte Carlo codes such as MCNPX to aid in calibration or measurement design. However, the MCNPX coincidence-counting routines treat particle histories individually, therefore the dead time of the acquisition electronics is not treated. The MCNPX-PoliMi code provides the ability to model detailed effects such as data-acquisition electronics and system dead times. A specialized post-processing code has been developed to interpret the collision-log file and determine the response of a 3He multiplicity counter. The MCNPX-PoliMi simulation provides the full neutron multiplicity distribution measured by the 3He tubes. This distribution is used to compute the singles, doubles, and triples rates which are the quantities used to determine 235U mass. MCNPX-PoliMi has previously been validated with passive multiplicity measurements. In this study, a detailed analysis of the measurement system operating in active mode is presented for uranium-oxide standards ranging from 0.5 to 4.0kg with a Canberra JCC-51 active well coincidence counter. MCNPX-PoliMi calculations are also compared with MCNPX. The two codes agree to within 1% for the cases with negligible dead times. The simulations are validated with measurements performed at the Y-12 National Security Complex.

Analysis of uncertainties affecting the Monte Carlo simulation of a neutron multiplicity counter

Abstract When applying numerical simulation to physical problems it is important to be able to clearly understand the quality that can be reached and the uncertainty associated with the modelling. In this paper we will deal with the Monte Carlo simulation of neutron coincidence and multiplicity counters. These instruments are intensively used in nuclear safeguards for the measurement and verification of the mass of fissile materials in nuclear installations. Their numerical simulation, either for detector design or for numerical calibration, is now used more and more frequently as support to inspections. Several benchmark exercises have been organised in order to assess the capabilities of Monte Carlo simulation to correctly predict realistic practical cases in neutron coincidence and multiplicity counting. Nevertheless most of them allowed reaching conclusions only on the overall uncertainties associated to the modelling through the comparison between calculated and measured integral data. In this paper we aim to analyse more deeply the sources of these uncertainties, separating the different possible causes and analysing parametrically the individual factors. We will include influencing factors deriving from the physical model of the detector, of the acquisition electronics, of the sample and from the nuclear data. The final total uncertainties are consistent with the results of the integral benchmarks.

Dead time corrections for neutron multiplicity counting

Dead time corrections for neutron multiplicity counting

Estimating alarm thresholds and the number of components in mixture distributions

Mixtures of probability distributions arise in many nuclear assay and forensic applications, including nuclear weapon detection, neutron multiplicity counting, and in solution monitoring (SM) for nuclear safeguards. SM data is increasingly used to enhance nuclear safeguards in aqueous reprocessing facilities having plutonium in solution form in many tanks. This paper provides background for mixture probability distributions and then focuses on mixtures arising in SM data. SM data can be analyzed by evaluating transfer-mode residuals defined as tank-to-tank transfer differences, and wait-mode residuals defined as changes during non-transfer modes. A previous paper investigated impacts on transfer-mode and wait-mode residuals of event marking errors which arise when the estimated start and/or stop times of tank events such as transfers are somewhat different from the true start and/or stop times. Event marking errors contribute to non-Gaussian behavior and larger variation than predicted on the basis of individual tank calibration studies. This paper illustrates evidence for mixture probability distributions arising from such event marking errors and from effects such as condensation or evaporation during non-transfer modes, and pump carryover during transfer modes. A quantitative assessment of the sample size required to adequately characterize a mixture probability distribution arising in any context is included.

A new alternative singles rate dead time correction algorithm for use in neutron multiplicity analysis

In this article we extend the neutron multiplicity counting dead time correction scheme devised by Dytlewski to include Singles. In this way a self consistent, item specific, Singles dead time correction is obtained under the same physical assumptions and approximations as are used for the Doubles and Triples corrections. This correction for the Singles count rate also affects the Doubles and Triples dead time correction because it is also the correction needed for the number of coincidence gates opened. We develop the new dead time formula, which we call the Croft–Dytlewski formula, as an action on the Accidentals histogram, and apply it to simulated pulse train data for which the correct results are known. The new formula was found to perform better than the standard ad-hoc Dytlewski Singles dead time correction formula, particularly for highly correlated high count rate items.

Computation of Neutron Multiplicity Statistics Using Deterministic Transport

Nuclear nonproliferation efforts are supported by measurements that are capable of rapidly characterizing special nuclear materials (SNM). Neutron multiplicity counting is frequently used to estimate properties of SNM, including neutron source strength, multiplication, and generation time. Different classes of models have been used to estimate these and other properties from the measured neutron counting distribution and its statistics. This paper describes a technique to compute statistics of the neutron counting distribution using deterministic neutron transport models. This approach can be applied to rapidly analyze neutron multiplicity counting measurements without relying on the point reactor kinetics model.

Experimental validation of a coupled neutron–photon inverse radiation transport solver

Abstract Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron–photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of α-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.

Dead-time correction for any multiplicity using list mode neutron multiplicity counters: A new approach – Low and medium count-rates

Abstract In the field of neutron multiplicity counting, modern list mode counters provide increased possibilities for neutron data analysis. Here a new method to correct dead-time using a multichannel list mode neutron counter is described. As it will become clear in this article, the data analysis can be done “on the fly” without further data storage. This will allow an instrument to be built having this method implemented, which will give directly dead-time corrected results. In practice, a classical theory of dead-time correction is applied to the final Totals, Reals (or Doubles) and possibly Triples, whereas for higher multiplicities no dead-time correction has been implemented so far. In contrast to that, this approach directly corrects the multiplicity distributions of the Reals plus Accidentals ( R + A ) and Accidentals ( A ) obtained by multiplicity counting. Hence this dead-time correction holds for any kind of multiplicity (Totals, Doubles, Triples, Quadruples, Quintuples, etc.) because the calculation of these values can then be derived directly from the corrected multiplicity distribution of R + A and A .

A combined neutron and gamma-ray multiplicity counter based on liquid scintillation detectors

Abstract Multiplicity counters for neutron assay have been extensively used in materials control and accountability for nonproliferation and nuclear safeguards. Typically, neutron coincidence counters are utilized in these fields. In this work, we present a measurement system that makes use not only of neutron (n) multiplicity counting but also of gamma-ray ( γ ) multiplicity counting and the combined higher-order multiples containing both neutrons and gamma rays. The benefit of this approach is in using both particle types available from the sample, leading to a reduction in measurement times compared with single-particle measurements. We present measurement results of n, γ , nn, n γ , γ γ , nnn, nn γ , n γ γ and γ γ γ multiples emitted by Mixed-Oxide (MOX) samples measured at Idaho National Laboratory (INL). The MOX measurement is compared to initial validation of the detection system done using a 252Cf source. The dual radiation measuring system proposed here uses extra measurables to improve the statistics when compared to a neutron-only system and allows for extended analysis and interpretation of sample parameters. New challenges such as the effect of very high intrinsic gamma-ray sources in the case of MOX samples are discussed. Successful measurements of multiple rates can be performed also when using high-Z shielding.

Initial evaluation for a combined neutron and gamma ray multiplicity counter

Multiplicity counters for neutron assay have been extensively used in materials control and accountability for nonproliferation and nuclear safeguards. Typically, neutron coincidence counters are utilized in these fields. In this paper we present a measurement system that makes use not only of neutron (n) multiplicity counting but also of gamma ray (γ) multiplicity counting and the combined higher-order multiplets containing both neutrons and gamma rays. The benefit of this approach is in using both particle types available from the sample, leading to a reduction in measurement time when using more measurables. We present measurement results of n, γ, nn, nγ,γγ, nnn, nnγ,nγγ and γγγ multiplets emitted by a 252Cf source and a 239Pu–Be source. A dual radiation measuring system proposed here could use extra measurables either to improve the statistics when compared to a neutron-only system, or alternatively be used for extended analysis and interpretation of sample parameters.The study presented here provides an answer to the questions regarding the ability to measure both neutron and gamma ray coincidences at once. The results show that the measurements system proposed in this study has a potential to be valuable in the area of nuclear nonproliferation and homeland security.

A note on the multiplicity expressions in nuclear safeguards

The purpose of this note is to point out notational inconsistencies and actual errors in some of the basic and often cited expressions of Ensslin et al. [Application Guide to Neutron Multiplicity Counting, Los Alamos Report LA-13422-M, 1998], even if the final formulae are correct. These expressions describe the measured multiplicity counting rates, i.e. singles, doubles, and triples rates S, D, and T, as parameters of a multiplying sample, through analytical expressions of the factorial moments of the neutrons generated in the sample by one initial source event. They serve as the basis of unfolding these parameters by the inversion of the formulae. The motivation for this brief communication is twofold. First, Ref. [N. Ensslin, W.C. Harker, M.S. Krick, D.G. Langner, M.M. Pickrell, J.E. Stewart, Application guide to neutron multiplicity counting, Los Alamos Report LA-13422-M, 1998] is the most widely available and by far the most cited source of the derivation, which is supposedly also used as a tutorial for newcomers. Hence it is important that the correct derivation is available. Indeed it can be seen that the erroneous expressions have already propagated widely in citations. Second, similar derivations have become recently interesting for the multiplicities of gamma photons. Hence we want to create both a citable reference as well as guidance for such on-going work in further applications of the theory.

Real-time simulation of neutron counters

In this paper we discuss the long-term perspective of the application of numerical modelling to neutron (coincidence or multiplicity) counting. The advantage of this technique for nuclear safeguards is discussed with particular emphasis to the numerical calibration of detectors. As a particular case of numerical modelling, real-time simulation would allow overcoming some problems related to pre-calibration of neutron counters, especially in all those cases when the sample characteristics deviate from the calibration conditions. Nevertheless it requires a significant increase in computing capabilities. The requirements for real-time simulations deriving from inspection needs are discussed and compared to current (and possible future) computing capabilities.

Variance results for the second and third reduced sample moments in neutron multiplicity counting for randomly triggered or signal-triggered counting gates

Neutron multiplicity counting is an established method to estimate the spontaneous fission rate, and therefore also the plutonium mass for example, in a sample that includes other neutron sources. The extent to which the sample and detector obey the “point model” assumptions impacts the estimate's total measurement error, but, in nearly all cases, for the random error contribution, it is useful to evaluate the variances of the second and third reduced sample moments of the neutron source strength. Therefore, this paper derives exact expressions for the variances and covariances of the second and third reduced sample moments for either randomly triggered or signal-triggered non-overlapping counting gates, and compares them to the corresponding variances in simulated data. Approximate expressions are also provided for the case of overlapping counting gates. These variances and covariances are useful in figure of merit calculations to predict assay performance prior to data collection. In addition, whenever real data are available, a bootstrap method is presented as an alternate but effective way to estimate these variances.

Comparison of list mode data acquisition and shift register measurements using the large epithermal neutron multiplicity counter (LEMC)

Neutron multiplicity analysis has been a valuable technique for safeguards measurements of plutonium oxide and mixed oxides. Historically, most of these measurements have been performed using shift register based electronics. The shift register data acquisition lacks certain flexibility because the basic coincidence parameters (e.g., pre-delay, gate width, and long delay) must be fixed prior to the start of the measurement and the values may potentially, therefore, be sub optimal. List mode or time stamped data acquisition records the arrival time of each pulse thereby preserving the history of the pulse stream and allowing analysis and reanalysis using software analogs to the shift register circuit with adjustable parameters. Until recently, the data rates encountered in the assay of modest amounts of plutonium using efficient multiplicity counters were beyond the capacity of readily available personal computers. The calibration of the large epithermal neutron multiplicity counter (LEMC) for assay of plutonium scrap materials is used as a vehicle to compare the performance of the multiplicity shift register and a commercially available list mode acquisition module.

Development of the ACP safeguards neutron counter for PWR spent fuel rods

An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21–24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

Sensitivity analysis of physical/operational parameters in neutron multiplicity counting

In this paper, we perform a sensitivity study on the influence of various physical and operational parameters on the results of neutron multiplicity counting. The purpose is to have a better understanding of the importance of each component and its contribution to the measurement uncertainty. Then we will be able to determine the optimal conditions for the operational parameters and for detector design and as well to point out weaknesses in the knowledge of critical fundamental nuclear data.

Development of a liquid scintillator neutron multiplicity counter (LSMC)

A new neutron multiplicity counter is being developed that utilizes the fast response of liquid scintillator detectors. The ability to detect fast (vs. moderated) fission neutrons makes possible a coincidence gate of the order of tens of nanoseconds (vs. tens of microseconds). A neutron counter with such a narrow gate will be virtually insensitive to accidental coincidences, making it possible to measure items with a high single neutron background to greater accuracy in less time. This includes impure Pu items with high (α, n) rates as well as items of low-mass HEU where a strong active interrogation source is needed. Liquid scintillator detectors also allow for energy discrimination between interrogation source neutrons and fission neutrons, allowing for even greater assay sensitivity. Designing and building a liquid scintillator multiplicity counter (LSMC) requires a symbiotic effort of simulation and experiment to optimize performance and mitigate hardware costs in the final product. We present preliminary Monte-Carlo studies using the GEANT toolkit.

Design and performances of the Scrap Neutron Multiplicity Counter

JRC has developed the Scrap Neutron Multiplicity Counter (SNMC): an advanced neutron multiplicity counter for the verification of inhomogeneous Pu samples, such as scrap material in MOX fuel fabrication plants. The innovative features of this counter with respect to existing ones rely on two aspects: (i) an optimised design based on Monte Carlo calculations in order to select the most appropriate materials, geometry and detector disposition for maximum efficiency and (ii) novel electronics based on digital signal processing (DSP) reducing the system dead time. The paper recalls the design process, the electronics, the construction and assembly of the counter. Then the results of the first experimental tests will be reported. We will show the characterization of the main physical parameters of the counter, the calibration and the verification of a wide variety of plutonium bearing samples available in the PERLA laboratory at JRC Ispra. This will include pure homogeneous samples (Pu dioxide powders, metal Pu, MOX powders and pellets) and some tests on heterogeneous samples representative of scrap material.

Development of a Neutron Mu1tiplicity Counter System for the Nuc1ear Material Accounting and Safeguards in KAERl

A well-type neutron multiplicity counter(WNMC) system was installed in a hot cell for the purpose of nuclear material control, accounting (NMC&A) and safeguards. The WNMC system determined the nuclear material amount on the basis of the Cm-monitoring method by measuring the neutron multiplicity and the number of coincidence events. The developed system consisted of 20 He-3 tubes with a 2.54 cm diameter and 50 cm length to cover a 25 cm long uniform neutron efficiency region along the tube length direction. As a neutron moderator in order to increase the neutron absorption rate at the He-3 tube, a well type high density polyethylene(HDPE) with a 50cm diameter was used. The efficiently optimized cylindrical Cd-plate was used in order to satisfy the uniformity along the tube direction. The absolute neutron detection efficiency of the HDPE system was measured about 14.0 % and the die-away time was 51.6µs. This system will be used not only for NMC8A but also the process for material quality control, for example, decladding performance of rod-cuts and the homogeneity of a mixing powder, waste and ingot.