Fissile Mass Research Articles

A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

Rapid nondestructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the Oak Ridge National Laboratory High Flux Isotope Reactor Neutron Activation Analysis facility has developed a generalized nondestructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and makes use of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a complete characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% recovery bias have been conducted on standards of 235U and 239Pu as low as 12ng in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 198ng of fissile mass with less than 7% recovery bias. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation facilities, and account for increasingly complex sample matrices.

Neutron Multiplicity Counting: Credible Regions for Reconstruction Parameters

From nuclear materials accountability to homeland security, the need for improved nuclear material detection, assay, and authentication has grown over the past decades. Starting in the 1940s, neutron multiplicity counting techniques have enabled quantitative evaluation of masses and multiplications of fissile materials. In this paper, we propose a new method to compute uncertainties on these parameters using a model-based sequential Bayesian processor, resulting in credible regions in the fissile material mass and multiplication space. These uncertainties will enable us to evaluate quantitatively proposed improvements to the theoretical fission chain model. In addition, because the processor can calculate uncertainties in real time, it is a useful tool in applications such as portal monitoring: monitoring can stop as soon as a preset confidence of nonthreat is reached.

Improving neutron multiplicity counting for the spatial dependence of multiplication: Results for spherical plutonium samples

The fissile mass deduced from a neutron multiplicity counting measurement of high mass dense items is underestimated if the spatial dependence of the multiplication is not taken into account. It is shown that an appropriate physics-based correction successfully removes the bias. It depends on four correction coefficients which can only be exactly determined if the sample geometry and composition are known. In some cases, for example in warhead authentication, available information on the sample will be very limited. MCNPX-PoliMi simulations have been performed to obtain the correction coefficients for a range of spherical plutonium metal geometries, with and without polyethylene reflection placed around the spheres. For hollow spheres, the analysis shows that the correction coefficients can be approximated with high accuracy as a function of the sphere׳s thickness depending only slightly on the radius. If the thickness remains unknown, less accurate estimates of the correction coefficients can be obtained from the neutron multiplication. The influence of isotopic composition is limited. The correction coefficients become somewhat smaller when reflection is present.

Fissile mass estimation by pulsed neutron source interrogation

Passive methods for detecting correlated neutrons from spontaneous fissions (e.g. multiplicity and SVM) are widely used for fissile mass estimations. These methods can be used for fissile materials that emit a significant amount of fission neutrons (like plutonium). Active interrogation, in which fissions are induced in the tested material by an external continuous source or by a pulsed neutron source, has the potential advantages of fast measurement, alongside independence of the spontaneous fissions of the tested fissile material, thus enabling uranium measurement.Until recently, using the multiplicity method, for uranium mass estimation, was possible only for active interrogation made with continues neutron source. Pulsed active neutron interrogation measurements were analyzed with techniques, e.g. differential die away analysis (DDA), which ignore or implicitly include the multiplicity effect (self-induced fission chains).Recently, both, the multiplicity and the SVM techniques, were theoretically extended for analyzing active fissile mass measurements, made by a pulsed neutron source. In this study the SVM technique for pulsed neutron source is experimentally examined, for the first time. The measurements were conducted at the PUNITA facility of the Joint Research Centre in Ispra, Italy. First promising results, of mass estimation by the SVM technique using a pulsed neutron source, are presented.

Lead Slowing-Down Spectrometry Analysis of Data from Measurements on Nuclear Fuel

Improved nondestructive assay of isotopic masses in used nuclear fuel would be valuable for nuclear safeguards operations associated with the transport, storage, and reprocessing of used nuclear fuel. Our collaboration is examining the feasibility of using lead slowing-down spectrometry techniques to assay the isotopic fissile masses in used nuclear fuel assemblies. We present the application of our analysis algorithms to measurements conducted with a lead spectrometer. The measurements involved a single fresh fuel pin and discrete 239Pu and 235U samples. We are able to describe the isotopic fissile masses with root-mean-square errors over seven different configurations to 6.3% for 239Pu and 2.7% for 235U. Significant effort is yet needed to demonstrate the applicability of these algorithms for used-fuel assemblies, but the results reported here are encouraging in demonstrating that we are making progress toward that goal.

A system for the measurement of delayed neutrons and gammas from special nuclear materials

The delayed neutron counting system at the Royal Military College of Canada has been upgraded to accommodate concurrent delayed neutron and gamma measurements. This delayed neutron and gamma counting system uses a SLOWPOKE-2 reactor to irradiate fissile materials before their transfer to a counting arrangement consisting of six 3He and one HPGe detector. The application of this system is demonstrated in an example where delayed neutron and gamma emissions are used in complement to examine 233U content and determine fissile mass with an average relative error and accuracy of −2.2 and 1.5 %, respectively.

Characterization of Old Nuclear Waste Packages Coupling Photon Activation Analysis and Complementary Non-Destructive Techniques

Radiological characterization of nuclear waste packages is an industrial issue in order to select the best mode of storage. The characterization becomes crucial particularly for waste packages produced at the beginning of the French nuclear industry. For the latter, available information is often incomplete and some key parameters are sometimes missing (content of the package, alpha-activity, fissile mass...) In this case, the use of non-destructive methods, both passive and active, is an appropriate solution to characterize nuclear waste packages and to obtain all the information of interest. In this article, we present the results of a complete characterization carried out on the TE 1060 block, which is a nuclear waste package produced during the 1960's in Saclay. This characterization is part of the DEMSAC (Dismantling of Saclay's facilities) project (ICPE part). It has been carried out in the SAPHIR facility, located in Saclay and housing a linear electron accelerator. This work enables to show the great interest of active methods (photon activation analysis and high-energy imaging) as soon as passive techniques encounter severe limitations.

Measurement Techniques for Warhead Authentication with Attributes: Advantages and Limitations

A survey of different measurement techniques applicable in the context of the attribute approach for warhead authentication is presented to demonstrate current capabilities and gaps. Therefore, their technical advantages and limitations will be assessed, in particular in terms of correctly characterizing an item whose configuration is not entirely known to an inspecting party. Passive and active neutron and gamma measurements assessing the presence of fissile material, isotopic composition and fissile mass of plutonium and uranium are considered.

Investigating the breeding capabilities of hybrid soliton reactors

Nuclear energy industry asks for an optimized exploitation of available natural resources and a safe operation of reactors. A closed fuel cycle requires the mass of fissile material depleted in a reactor to be equal to or less than the fissile mass produced in the same or in other reactors. In this work, a simple closed cycle scheme is investigated, grounded on the use of a conceptual thermal water-cooled and moderated subcritical hybrid soliton reactor (HSR). The concept is a specific Accelerator Driven System (ADS) operating at lower power than usual pressurized water reactors (PWRs). This type of reactor can be inherently safe, since shutdown is achieved by simply interrupting the accelerator's power supply. In this work a preliminary investigation is attempted concerning the existence of conditions under which the operation of a thermal HSR in breeding regime is possible. For this purpose, a conceptual encapsulated core has been defined by choosing the magnitude of a set of parameters which are important from the neutronic point of view, such as core geometry and fuel composition. Indications of breeding operation regime for thermal HSR systems are sought by performing preliminary simulations of this core. For this purpose, the Monte Carlo code ANET, which is being developed based on the high energy physics code GEANT is utilized, as being capable of simulating particles’ transport and interactions produced, including also simulation of low energy neutrons transport. A simple analytical model is also developed and presented in order to investigate the conditions under which breeding in HSR is possible, which supports the ANET simulation findings.

A novel method for active fissile mass estimation with a pulsed neutron source

Neutron interrogation facilities for mass evaluation of Special Nuclear Materials (SNM) samples are divided into two main categories: passive interrogation, where all neutron detections are due to spontaneous events, and active interrogation, where fissions are induced on the tested material by an external neutron source. While active methods are, in general, faster and more effective, their analysis is much harder to carry out.In the paper, we will introduce a new formalism for analyzing the detection signal generated by a pulsed source active interrogation facility. The analysis is aimed to distinct between fission neutrons from the main neutron source in the system, and the surrounding “neutron noise”.In particular, we derive analytic expressions for the first three central moments of the number of detections in a given time interval, in terms of the different neutron sources.While the method depends on exactly the same physical assumptions as known models, the simplicity of the suggested formalism allows us to take into account the variance of the external neutron source—an effect that was so far neglected.

Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71.

A method for the estimation of fissile mass by measuring the number of neutron signals within a specific time interval

This paper presents a new method for fissile mass estimation by passive neutron interrogation. The method uses the stochastic transport equation to express the first three central moments (mean, variance and skewness) for the number of detected neutrons within a given time interval in terms of the fissile mass, the (α,n) rate and the multiplicity factor. Solving these three equations allows estimation of the fissile mass. As the primary validation of the method, it was implemented for several fissile mass measurements, which were conducted by the Nuclear Security Unit of the Joint Research Center (JRC) in Italy and by the Institute of Isotopes in Hungary.

Characterization of a Mixed Multiplicity Counter Based on Liquid Organic Scintillators

A measurement system capable of multiplicity measurements for both neutrons and gamma rays has been developed. The benefit of such an approach is in the increased number of available observables. A pure neutron assay results in three observables for third-order multiples, while a combined neutron/gamma-ray assay results in nine observables for the same order of multiples. The idea is to use the additional observables to achieve greater accuracy when determining unknown parameters of the sample such as the fissile mass. The measurement system is based on liquid scintillation detectors (EJ-309s) which feed detected pulses to a digital data-acquisition system. The excellent pulse shape discrimination capabilities of the EJ-309s allow for accurate differentiation between gamma-ray pulses and neutron pulses. The PSD is vital to correctly identify the different multiples up to the third order: n, γ , nn, nγ , γγ , nnn, nnγ , nγγ and γγγ . Previous investigation of the measurement system showed that good counting statistics can be achieved within minutes for spontaneous-fission sources such as <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">252</sup> Cf. In this paper, we present new measurement results and corresponding Monte Carlo simulations aimed at characterizing the measurement system. Comparison of the measured and simulated multiples is discussed in detail and a relatively good agreement is found.

Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability to measure the fissile content in spent fuel assemblies, For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the {sup 244}Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective {sup 239}Pu mass was introduced by weighting the relative contribution to the signal of {sup 235}U and {sup 241}Pu compared to {sup 239}Pu and the calibration curves of DDA count rate vs. {sup 239}Pu{sub eff} were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10{sup 9} n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

Time-Spectral Analysis Methods for Spent Fuel Assay Using Lead Slowing-Down Spectroscopy

Nondestructive techniques for measuring the mass of fissile isotopes in spent nuclear fuel is a considerable challenge in the safeguarding of nuclear fuel cycles. A nondestructive assay technology that could provide direct measurement of fissile mass, particularly for the plutonium (Pu) isotopes, and improve upon the uncertainty of today's confirmatory methods is needed. Lead slowing-down spectroscopy (LSDS) has been studied for the spent fuel application previously, but the nonlinear effects of assembly self shielding (of the interrogating neutron population) have led to discouraging assay accuracy for realistic pressurized water reactor fuels. In this paper, we describe the development of time-spectral analysis algorithms for LSDS intended to overcome these self-shielding effects. The algorithm incorporates the tabulated energy-dependent cross sections from key fissile and absorbing isotopes, but leaves their mass as free variables. Multi-parameter regression analysis is then used to directly calculate not only the mass of fissile isotopes in the fuel assembly (e.g., Pu-239, U-235, and Pu-241), but also the mass of key absorbing isotopes such as Pu-240 and U-238. Modeling-based assay results using this self-shielding relationship indicate that LSDS has the potential to directly measure fissile isotopes with less than 5% average relative error for pressurized water reactor assemblies with burnup as high as 60 GWd/MTU. Shortcomings in the initial self-shielding model and potential improvements to the formulation are described.

Fissile mass determination by the active neutron correlation technique with the generalised Poisson parameters or its factorial moments

An analytical model is derived for a process with a fissile sample irradiated periodically by a pulsed neutron source in a neutron moderator assembly. The generated induced fission neutrons are detected in a specialised thermal neutron detector assembly, dimensioned for a short thermal neutron lifetime 1/ λ. This known lifetime is much shorter than the thermal neutron lifetime 1/ Λ of the moderator assembly. Neutrons arriving in the detector assembly are registered in periodically opened observation intervals of duration τ. This is repeated after each pulse of the neutron generator. The probability distribution of these signals is treated as a generalised inhomogeneous Poisson distribution. It is characterised by the parameter Γ j ( t) for the detection of j, j=1,2,3,…, neutrons in the thermal neutron detector assembly during the observation interval [ t− τ, t] after time t− τ≥ s 0 following each source pulse at time s=0. The normalised factorial moment m ( μ ) * ( t ) of Γ j ( t) with order μ is proportional to the number of events, where all μ signals are correlated i.e., from a common fission event. Γ j ( t) and m ( μ ) * ( t ) x are expressed as a function of the probability distribution b x ( t), for x=0,1,2,3,…, and as a function of its normalised factorial moments m ( μ )( t), for μ=0,1,2,3,…, respectively. Both Γ j ( t) and m ( μ ) * ( t ) depend each on a component decreasing exponentially with the argument Λt for all j and μ values, due to induced fission events. These also have a component independent of t, due to spontaneous fission and background events. This relationship is used to discriminate between induced fission and stationary events. It serves to determine the fissile mass G F , the detection probability ε F for slowed down fission neutrons, the mass G S of spontaneous fission neutron emitters, the detection probability ε S for slowed down spontaneous fission neutrons and neutrons from (α, n) reactions. A solution for two thermal neutron lifetimes 1/ Λ i , i=1,2 of the moderator assembly is given. It is valid for fissile material embedded in a waste matrix affecting the lifetime 1/ Λ of the neutron moderator assembly.

Fissile and fertile nuclear material measurements using a new differential die-away self-interrogation technique

This paper presents a new technique for the measurement of fissile and fertile nuclear materials in spent fuel and plutonium-laden materials such as mixed oxide (MOX) fuel. The technique, called differential die-away self-interrogation, is similar to traditional differential die-away analysis, but it does not require a pulsed neutron generator or pulsed beam accelerator, and it can measure the fertile mass in addition to the fissile mass. The new method uses the spontaneous fission neutrons from 244Cm in spent fuel and 240Pu effective neutrons in MOX as the “pulsed” neutron source, with an average of ∼2.7 neutrons per pulse. The time-correlated neutrons from the spontaneous fission and the subsequent induced fissions are analyzed as a function of time to determine the spontaneous fission rate, the induced fast-neutron fissions, and the induced thermal-neutron fissions. The fissile mass is determined from the induced thermal-neutron fissions that are produced by reflected thermal neutrons that originated from the spontaneous fission reaction. The sensitivity of the fissile mass measurement is enhanced by the use of two measurements, with and without a cadmium liner between the sample and a hydrogenous moderator that surrounds the sample. The fertile mass is determined from the multiplicity analysis of the neutrons detected soon after the initial triggering neutron is detected. The method obtains good sensitivity by the optimal design of two different neutron die-away regions: a short die-away for the neutron detector region and a longer die-away for the sample interrogation region.

Improvement of non-destructive fissile mass assays in α low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

Within the framework of radioactive waste control, non-destructive assay (NDA) methods may be employed. The active neutron interrogation (ANI) method is now well-known and effective in quantifying low α-activity fissile masses (mainly 235U, 239Pu, 241Pu) with low densities, i.e. less than about 0.4, in radioactive waste drums of volumes up to 200 l. The PROMpt Epithermal and THErmal interrogation Experiment (PROMETHEE [F. Jallu, A. Mariani, C. Passard, A.-C. Raoux, H. Toubon, Alpha low level waste control: improvement of the PROMETHEE 6 assay system performances. Nucl. Technol. 153 (January) (2006); C. Passard, A. Mariani, F. Jallu, J. Romeyer-Dherber, H. Recroix, M. Rodriguez, J. Loridon, C. Denis, PROMETHEE: an alpha low level waste assay system using passive and active neutron measurement methods. Nucl. Technol. 140 (December) (2002) 303–314]) based on ANI has been under development since 1996 to reach the incinerating α low level waste (LLW) criterion of about 50 Bq[α] per gram of crude waste (≈50 μg Pu) in 118 l drums on the date the drums are conditioned. Difficulties arise when dealing with matrices containing neutron energy moderators such as H and neutron absorbents such as Cl. These components may have a great influence on the fissile mass deduced from the neutron signal measured by ANI. For example, the calibration coefficient measured in a 118 l drum containing a cellulose matrix (density d = 0.144 g cm −3) may be 50 times higher than that obtained in a poly-vinyl-chloride matrix ( d = 0.253 g cm −3). Without any information on the matrix, the fissile mass is often overestimated due to safety procedures and by considering the most disadvantageous calibration coefficient corresponding to the most absorbing and moderating calibration matrix. The work discussed in this paper was performed at the CEA Nuclear Measurement Laboratory in France. It concerns the development of a matrix effect correction method, which consists in identifying and quantifying the matrix components by using prompt gamma-rays following neutron capture. The method aims to refine the value of the adequate calibration coefficient used for ANI analysis. This paper presents the final results obtained for 118 l waste drums with low α-activity and low density. This paper discusses the experimental and modelling studies and describes the development of correction abacuses based on gamma-ray spectrometry signals.

Investigation of Criticality Effects of Deuterium and Beryllium in Package Containing Fissile Material

The maximum mass of deuterium and beryllium in fissile exempted packages is strictly limited under subparagraph 672(a) of the IAEA Regulations for the Safe Transport of Radioactive Material (TS-R-1).1) This is because the restriction of fissile material mass in the subparagraph is based upon the assumption of hydrogenous moderation and reflection and that deuterium and beryllium are potentially more effective than hydrogen in terms of criticality. However, the basis of the deuterium and beryllium limits in the subparagraph is not clear and it may be inappropriate to impose a limit on their mass alone. In order to evaluate the reactivity effects of deuterium and beryllium in hydrogenous moderation and reflection, a series of calculations is performed using simple models with various parameters. The parameters include fissile nuclides, fissile concentrations, and concentrations of deuterium or beryllium in the moderator or the reflector. According to the calculation results, the concentrations of deuterium or beryllium as well their mass in the moderator or the reflector should be considered when establishing criticality based limits. This shows that, when beryllium or deuterium are present in very low concentrations within a package containing fissile material, the total mass of each may not be the only relevant factor to consider in terms of criticality.

Caustic Precipitation of Plutonium and Uranium with Gadolinium as a Neutron Poison

The caustic precipitation of plutonium and uranium from Pu- and U-containing waste solutions has been investigated to determine whether gadolinium could be used as a neutron poison for precipitation with greater than a fissile mass containing both Pu and enriched U. Precipitation experiments were performed using both process solution samples and simulant solutions with a range of 2.6 to 5.16 g/l U and 0 to 4.3:1 U:Pu. Analyses were performed on solutions at intermediate pH to determine the partitioning of elements for accident scenarios. When both Pu and U were present in the solution, precipitation began at pH 4.5 and by pH 7, 99% of Pu and U had precipitated. When complete neutralization was achieved at pH >14 with 1.2 M excess OH-, greater than 99% of Pu, U, and Gd had precipitated. At pH >14, the particle sizes were larger, and the distribution was a single mode. The ratio of hydrogen to fissile atoms in the precipitate was determined after both settling and centrifuging and indicates that sufficient water was associated with the precipitates to provide the needed neutron moderation for Gd to prevent a criticality in solutions containing up to 4.3:1 U:Pu and up to 5.16 g/l U.