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All sources NASA NTRS DOE OSTI

DOE OSTI2019

NUREG/CR-7239, “Review of Fissile Material Exemptions and General Licenses in 10 CFR Part 71,” and Proposed Revisions to Harmonize with IAEA SSR-6 Fissile Material Requirements

Title 10 of the United States (US) Code of Federal Regulations (CFR), Part 71, Packaging and Transportation of Radioactive Material, includes requirements for the transport of fissile material in packages. These regulations provide requirements for packaging that is certified by the US Nuclear Regulatory Commission (NRC) to transport fissile material. The regulations also include provisions that allow for shipment in packages that are not required to be certified by the NRC. These provisions include criteria for exemptions from classification as fissile material (§71.15) and permit general licenses for limited quantities of fissile material and plutonium-beryllium special form sources (§71.22 and §71.23, respectively). This paper will discuss recommendations that NRC has made on the use of the fissile material exemptions and general license provisions in a recently published NUREG/CR, and changes to these provisions proposed for an upcoming rulemaking action.NUREG/CR-7239, “Review of Fissile Material Exemptions and General Licenses in 10 CFR Part 71,” was recently developed by Oak Ridge National Laboratory (ORNL) for the NRC. This document provides technical recommendations to fissile material licensees on how to apply the fissile material exemption and general license regulations. This paper will outline this NUREG/CR, including: 1) the technical basis, intent, and anticipated usage of each of the fissile material exemption and general license provisions; 2) the recent regulatory history of the Part 71 fissile material exemptions and general licenses; and 3) example problems which illustrate how the provisions might be applied and justified.In 2012, IAEA made several significant changes to the fissile material requirements in SSR-6, including the fissile exception provisions and provisions similar to the U.S. fissile material general licenses. In 2016, the NRC began a rulemaking effort to consider these changes (among others), and to decide how to best harmonize the domestic U.S. regulations. This paper will review the 2012 changes made to SSR-6, discuss the NRC staff consideration of those changes, and detail the proposed revisions to 10 CFR Part 71 that NRC is considering for the fissile material exemptions and general licenses.

Barto, Andrew↗

DOE OSTI2014

Scope and verification of a Fissile Material (Cutoff) Treaty

A Fissile Material Cutoff Treaty (FMCT) would ban the production of fissile material - in practice highly-enriched uranium and separated plutonium - for weapons. It has been supported by strong majorities in the United Nations. After it comes into force, newly produced fissile materials could only be produced under international - most likely International Atomic Energy Agency - monitoring. Many non-weapon states argue that the treaty should also place under safeguards pre-existing stocks of fissile material in civilian use or declared excess for weapons so as to make nuclear-weapons reductions irreversible. This paper discusses the scope of the FMCT, the ability to detect clandestine production and verification challenges in the nuclear-weapons states.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2017

Characterization of Fissile Assemblies Using Low-Efficiency Detection Systems

Here, we have investigated the possibility that the amount, chemical form, multiplication, and shape of the fissile material in an assembly can be passively assayed using scintillator detection systems by only measuring the fast neutron pulse height distribution and distribution of time intervals Δt between fast neutrons. We have previously demonstrated that the alpha-ratio can be obtained from the observed pulse height distribution for fast neutrons. In this paper we report that we report that when the distribution of time intervals is plotted as a function of logΔt, the position of the correlated neutron peak is nearly independent of detector efficiency and determines the internal relaxation rate for fast neutrons. If this information is combined with knowledge of the alpha-ratio, then the position of the minimum between the correlated and uncorrelated peaks can be used to rapidly estimate the mass, multiplication, and shape of fissile material. This method does not require a priori knowledge of either the efficiency for neutron detection or the alpha-ratio. Although our method neglects 3-neutron correlations, we have used previously obtained experimental data for metallic and oxide forms of Pu to demonstrate that our method yields good estimates for multiplications as large as 2, and that the only constraint on detector efficiency/observation time is that a peak in the interval time distribution due to correlated neutrons is visible.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Review of Fissile Material Exemptions and General Licenses in 10 CFR Part 71 - 14578

Title 10 of the United States (US) Code of Federal Regulations (CFR), Part 71, Packaging and Transportation of Radioactive Material includes requirements for the transport of fissile material in packages. These regulations provide requirements for packaging that is certified by the US Nuclear Regulatory Commission (NRC) to transport fissile material, but the regulations also include provisions that allow for shipment of fissile material in packages that are not required to be certified by the NRC. These provisions include criteria for exemptions from classification as fissile material (10 CFR 71.15) and permit general licenses for limited quantities of fissile material and plutonium-beryllium special form sources (10 CFR 71.22 and 10 CFR 71.23, respectively). These regulations were revised by NRC as part of a broad rulemaking effort in 2004. These regulations are used frequently for a wide-variety of fissile material shipping operations and shippers must interpret the regulations for particular shipping situations. The quantity of fissile material that presents a criticality safety risk depends on the other materials in the system that might absorb neutrons emitted from fission or moderate (slow) those neutrons to lower energies where the probability of them creating an additional fission is highest. Besides the absorption and moderation of neutrons, other parameters that must be considered to ensure criticality safety are: the mass, concentration, or isotopic distribution of the fissile material; and the system geometry and neutron reflection of the package from surrounding materials. These parameters were carefully considered in developing the requirements for the exemptions from the classification of fissile material and the requirements for general licenses for fissile material. This paper will review the technical basis, intent, and anticipated usage of each of the fissile material exemption and general license provisions. Additionally, this paper will discuss the recent regulatory history of the Part 71 fissile material exemptions and general licenses, including the 2015 rulemaking effort. Finally, this paper will outline an upcoming NUREG/CR document regarding the fissile material exemptions and general licenses, and discuss recommendations that are being formulated for potential future revisions to the regulations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Which parameters might help to predict an impact on reactivity for mixed fissile units in a storage array?

Several studies have already been performed on mixed fissile units in storage arrays and concluded that the neutron spectrum of the fissile material, the shape of fissile units, the interstitial moderation and structural materials may have an impact on reactivity. For this paper, calculations are performed on a simplified infinite planar array with one level of fissile material units axially reflected by concrete. Fissile material is modelled by a cylinder shape centered in the array mesh selected (120 cm height and 60 cm pitch). Either air or interstitial moderation by water is modelled between the fissile units. Three parameters are studied: the fissile materials (with various moderation conditions), the cylinder dimension and the nature of the structural material. The main following conclusions can be drawn from these calculations. First, for several mixed-array cases, air-spaced mixed arrays present a high increase on reactivity whereas interstitial moderation between fissile units leads to no or little impact. Secondly, fissile unit shapes do not explain directly the mixing impact on reactivity. Thirdly, structural materials have an important impact on reactivity in mixed array (either an increase or a decrease) due to their absorbing and/or moderating properties either when air or interstitial moderation are modelled between fissile units. Fourthly, no simple parameter was found to predict the impact on reactivity of mixed array in all studied configurations. However, a comparison of different neutron spectrum parameters can help to predict the absence or potential presence of a 'mixing effect'. Indeed, if all parameters are similar, no 'mixing effect' is observed. But, if these parameters present differences, there might be a 'mixing effect'. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Minimum critical values for various fissile media with very low content of plutonium mixed in different kinds of matrix - 14340

For materials slightly contaminated by fissile nuclides (such as radioactive sludge), it could be useful to take into account the low fissile density in order to define the nuclear criticality safety limits (mass, volume,...) when this density is lower than the one of the optimally moderated fissile-water mixture (e.g. 30 g/dm{sup 3} for {sup 239}Pu). In such a case, if the material is modelled by the usual fissile-water mixture only, the fissile medium still contains a void fraction when the optimal moderation ratio is reached. Then, the question is to know whether modeling the other elements of the material can impact the nuclear criticality safety limits, in particular when they are good scattering elements. This paper aims at studying the impact of replacing the void fraction in plutonium fissile media, with various apparent densities, by different kinds of matrix containing scattering elements. Minimum critical values (mass, volume, diameter and thickness) have been determined for plutonium media, composed of 100 % {sup 239}Pu, with densities of 2, 5.8 and 7.3 g/dm{sup 3} in different kinds of matrix containing scattering elements (graphite, carbon, alumina, silica, magnesia, fluorite and calcium oxide). Different contents of the aforementioned scattering elements in the fissile media were studied and additional moderation was ensured by water. The results highlight that minimum critical values are lower when considering a fissile medium containing graphite, carbon, silica, alumina and magnesia with additional water moderation, compared to plutonium-water mixture only with the same low fissile apparent density. However, there is still a benefit of crediting density (critical values for a low fissile density are still higher than critical values without a density control). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Design and testing of a combined neutron and gamma assay system for measuring fissile material in fuel and waste items - 15340

A combined neutron and gamma ray assay system for the measurement of fissile material in both fuel and waste items has been designed, modelled, developed and tested. The purpose of the system is to measure and characterize a wide range of fissile materials including uranium, plutonium, americium and mixed oxide fuel arising from a variety of processes and facilities across the United Kingdom. Dounreay Site Restoration Limited at Dounreay currently stores these items form the United Kingdom Nuclear Decommissioning Authority's un-irradiated fuel inventory. This inventory comprises historic materials from ex-United Kingdom Atomic Energy Authority nuclear fuel projects and experiments together with fast reactor fuel materials and some commercial fuel items as delivered to Dounreay for recovery. The combined assay system has been designed and constructed to support the operation of a fuel processing facility, which will be operated at Dounreay to treat these historic legacy fuel materials, making them suitable for long term storage. Appropriate characterization of the fissile material is required for safety, criticality control and nuclear materials safeguards purposes prior to final disposition. The system consists of both gamma ray and neutron measurement components. The results of a neutron measurement are combined with the gamma ray isotopic ratio measurement results to generate the assay result for the fissile material being characterized. The gamma ray component of the system employs a safeguards quality high purity Germanium detector with electro-mechanical cooling and the measurements of fissile material containers are made in a shielded chamber with a tin and copper graded lining. Containers of fissile material are rotated during the gamma ray measurement and a variable automatic steel collimator aperture mechanism is employed to adjust the measured count rate to control detector dead time. Gamma ray spectra are analysed using the plutonium and uranium isotopic ratio code PC/FRAM. The gamma ray system can also be configured to perform a direct assay of small quantities of uranium bearing material. For this assay measurement a series of cadmium filters can be positioned in front of the detector to reduce the contribution from 59 keV gamma rays from Am-241. In addition to gamma ray isotopic measurements, both active and passive neutron assay is performed by the system. The neutron-measuring component consists of a thick walled cylindrical polyethylene chamber, which provides neutron moderation for 30 one-inch diameter He-3 detector tubes located in re-entrant channels in the polyethylene moderator. Two removable polyethylene-inserts incorporating lower plug assemblies are provided, one with and one without a cadmium lining. Also two top plug units are provided, again one with cadmium and one without a cadmium liner. The Neutron Subsystem can be operated both as a passive and as an active neutron coincidence counter. In both fast and thermal active neutron mode, the neutron component employs two americium-beryllium (Am-Be) alpha-n neutron sources, one positioned in the top and the other in the bottom plug unit. The neutron sources are removed when the counter is operated in passive mode. Passive neutron detection efficiency and active neutron performance has been modelled using the MCNP Monte Carlo computer simulation code as part of the design process. In the absence of fissile samples, operation of the gamma ray component of the system has been tested initially using calibrated Eu-152 and Ba-133 gamma ray sources and the use of archived spectra from fissile material measurements. Initial testing of the neutron system has employed a calibrated Cf-252 spontaneous fission neutron source. Neutron testing has been confirmed using further MCNP simulations. The results of test measurements with the system are presented and compared with the results of MCNP measurement simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Anomalies of the fissile mass control inside a controlled geometry leading to important safety issues - 14298

This article focuses on the very usual case of the fissile mass control inside a controlled geometry, whatever the fissile density in the fissile-moderator mixture is. Although this case is very usual (maximal loading of fissile material in a transport cask, maximal fissile content in a 'geometrically favorable' tank, etc.), some unexpected behaviors ('anomalies') may be met, leading to important nuclear criticality safety issues. This article will show the three following anomalies and the associated safety issues: - the 'non-smooth' variation of the minimum critical mass for a small variation of the geometry size (a cylinder case is considered throughout this article), - the very important differences in critical values, for intermediate-size geometries, between two fissile media having very close minimum critical masses and dimensions, - the change in the hierarchy of different fissile media when the size of the geometry changes. The next part of the article will demonstrate that these anomalies are, from a physics point of view, due to the nature of the fissile material and, from a methodology point of view, due to the fact that this 'Mass- Geometry Control' problem is generally solved without looking at the critical mass for the whole range of geometries. The latter happens because this whole resolution is not trivial and requires either a lot of calculations or human effort to estimate the mass limits as a function of the geometry. Another article will discuss this question of the 'whole' resolution of the problem. (authors)

42 ENGINEERING↗

DOE OSTI2025

Distinguishing fissile uranium isotopes using an active well neutron coincidence counter

Proposed thorium-based nuclear fuel cycles are likely to require quantification and verification of 233 U within nuclear material. Because of their similar fission cross sections, active neutron nondestructive assay (NDA) systems may respond similarly to 233 U and 235 U. Traditional safeguards equipment has been optimized for 235 U and 238 U quantification associated with conventional uranium/plutonium fuel cycles and may not be directly applicable to 233 U quantification when mixed with other actinides. This work used models of the large volume active well coincidence counter (LV-AWCC) at Oak Ridge National Laboratory to evaluate the performance of this neutron NDA system to differentiate fissile uranium isotopes. The models were developed to simulate NDA system performance in response to a number of triangular radiation signature training device sources within the central cavity or well. This work predicted that the LV-AWCC can effectively differentiate 233 U from 235 U in certain modes of operation. In active mode, the LV-AWCC with the cadmium liner results in different doubles count rates between the fissile isotopes for a given fissile uranium mass. Without the cadmium liner, the uranium isotopes provide a statistically indistinguishable doubles count rate response for the fissile masses considered in this work (up to approximately 150 g). The cadmium liner serves to harden the neutron interrogation spectrum, which better exploits the notable difference in the 233 U and 235 U fission cross sections at approximately 1 eV. In passive mode, the two fissile isotopes exhibit different doubles and singles count rates regardless of liner presence because the passive source strength of 233 U is approximately 2 orders of magnitude stronger than that of 235 U due to the shorter half-life and correspondingly higher (α, n) yield. We conclude that using neutron interrogation in the LV-AWCC, two measurements are needed to quantify 233 U content in mixed uranium items. The first measurement is used to determine the total fissile uranium mass using a mode that cannot distinguish fissile isotopes (i.e., where a similar response is observed for both fissile uranium isotopes such as active doubles without cadmium or using a thermal neutron interrogation source). In conclusion, the second measurement is used to determine the 233 U content by using a differentiating technique (e.g., passive doubles, passive doubles to singles ratio, active doubles with cadmium).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2013

Evaluation Of Glass Density To Support The Estimation Of Fissile Mass Loadings From Iron Concentrations In SB8 Glasses

The Department of Energy – Savannah River (DOE-SR) has provided direction to Savannah River Remediation (SRR) to maintain fissile concentration in glass below 897 g/m 3 . In support of that guidance, the Savannah River National Laboratory (SRNL) provided a technical basis and a supporting Microsoft® Excel® spreadsheet for the evaluation of fissile loading in Sludge Batch 5 (SB5), Sludge Batch 6 (SB6), Sludge Batch 7a (SB7a), and Sludge Batch 7b (SB7b) glass based on the iron (Fe) concentration in glass as determined by the measurements from the Slurry Mix Evaporator (SME) acceptability analysis. SRR has since requested that the necessary density information be provided to allow SRR to update the Excel® spreadsheet so that it may be used to maintain fissile concentration in glass below 897 g/m 3 during the processing of Sludge Batch 8 (SB8). One of the primary inputs into the fissile loading spreadsheet includes an upper bound for the density of SB8-based glasses. Thus, these bounding density values are to be used to assess the fissile concentration in this glass system. It should be noted that no changes are needed to the underlying structure of the Excel-based spreadsheet to support fissile assessments for SB8. However, SRR should update the other key inputs to the spreadsheet that are based on fissile and Fe concentrations reported from the SB8 Waste Acceptance Product Specification (WAPS) sample.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Transitions From Stochastic to Point Kinetics Models In Fissile Solutions

Fissile solutions form an integral part of the fuel cycle, being used to store and process both fresh fuel and waste. Fissile solutions are also found within Aqueous Homogeneous Reactors (AHRs) including research reactors such as CRAC and SILENE and also potential future systems which could be used for the production of medical isotopes. In the case of fissile solutions in the fuel cycle, avoiding criticality is a priority but accidents have occurred in which a significant number of fissions occurred. In the case of AHRs, a sustained neutron population and fission rate is desired. In both cases the ability to simulate transients where changes in reactivity cause the neutron population to increase are important. A popular method for the simulation of such transients is the point kinetic method which allows the capture of the major features of a transient at a fraction of the computational runtime and code complexity of more comprehensive modelling technologies. A limitation of the point kinetics method and other mainstream methods of predicting neutron behaviour is that they assume the neutron population is large enough that the mean behaviour will be observed. When the neutron population is low, however, the deviation around the mean behaviour is expected to be relatively large as there is an aleatoric uncertainty arising from the uncertain fate of individual neutrons. Such limitations are of particular importance when the reactivity of a system is rapidly increasing while the neutron population is low, which is often true in accident scenarios involving fissile solutions. This provides the motivation for the coupling of point kinetics and a probabilistic models of fissile solutions. This summary aims to explore some of the behaviours resulting from a low neutron population and rising reactivity using a coupled probabilistic and point kinetics model of a fissile solution. As the effects of a low neutron population are the primary concern of this summary, the point kinetics and thermal hydraulic models used will be relatively simple but these models should be sufficient to capture the most important behaviours of the system in the first minute or so of a transient. This work is an extension of the work found in [1]. In that work the methodology was applied to fast pulses produced by systems such as Godiva and its results were found to compare favourably to experimental and analytical results, such as the burst wait time, in [2]. This work focuses on the application of this methodology to a fissile solution but the method could be applied to a number of criticality safety or reactor operation scenarios through suitable modification of the point kinetics model and would produce a more accurate prediction of the peak power obtained when a system first becomes critical compared to the use of a standard point kinetics model. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

The additive volume density law implemented in the Latec workbench to deal with fissile media that contain moderator elements - 14342

The moderation ratio is one of the key parameters for safety assessment in criticality-safety studies. It can be expressed as a moderator to fissile atoms ratio and is related to the concentration of fissile isotopes. The relationship linking these parameters illustrates a physical behaviour of the mixture, which could be very complex depending on the different kinds of media and their associated physical states encountered in the nuclear fuel cycle. To guarantee the safety when modelling a given mixture, criticality-safety practitioners consider either a best estimate approach by fitting experimental data, or, a bounding approach with a fictitious fissile mixture. This theoretical model is generally more penalizing in terms of k{sub eff} and does not involve numeric algorithms for the resolution reducing the overall uncertainty. This is the case for the additive volume density law, which is based on the assumption that the volume of a mixture is the sum of the volumes of all its compounds. This naturally leads to differentiate all the compounds in the mixture and, in particular, the fissile compound and the moderator. A first algorithm has been derived from this approach and implemented in the LATEC criticality workbench developed at IRSN, but a second assumption is required to solve the problem encountered when moderator's elements have also to be used to define the fissile compound. A typical example of this major issue is an UZrH fissile compound moderated by water. The reference moderator element of the water is obviously hydrogen that belongs to the fissile compound. This paper is dedicated to the implementation in the LATEC criticality workbench of an algorithm based on the additive volume density law that is able to deal with such a configuration. The major advantage of this new algorithm is the versatility in terms of input data. The mixture compounds and the moderation can be expressed in multiple ways still using the same algorithm. A validation work has been performed on different UZrH compounds and for different expressions of moderation. The calculated isotopic concentrations have been compared to theoretical calculations using a more straightforward algorithm but that should be adapted for every change in terms of moderation expression. The results obtained are in good agreement with validation results. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Neutron Multiplicity Counting Moments for Fissile Mass Estimation in Scatter-Based Neutron Detection Systems

Neutron multiplicity counting (NMC) techniques are widely used for nuclear materials accountability and international safeguards applications to quantitatively evaluate characteristic properties pertaining to fissile material. Mathematical models for NMC moments have been previously derived for systems that use capture-based detectors; however, these models are not applicable when scatter-based detectors are used because of “neutron cross talk.” Neutron cross talk occurs when a single neutron scatters and deposits energy above threshold into multiple detectors causing spurious increase in multiplicity counts; this, in turn, has caused fissile mass to be overestimated when not treated. In this paper, we propose new mathematical models derived from point kinetics to correct for neutron cross-talk effects up to any arbitrary order N, where N denotes the maximum number of counts a single neutron can cause. The new models were used to estimate the fissile mass of plutonium metal and oxide samples with effective 240 Pu mass ranging from 2.5 to 250 g. The adequacy of the models was confirmed using simulations of a conceptual scatter-based neutron multiplicity counter (e.g., organic scintillators) using MCNPX v2.7e with the PoliMi fission event generating extension. The fissile mass estimates with no correction for neutron cross-talk events yielded an average relative deviation from the true 240 Pu eff mass of 55.94% and 84.56% for metal and oxide samples, respectively. When neutron cross-talk events of order N = 2 are included in the model, the fissile mass estimates yielded an average relative deviation of 11.89% for metal and 13.21% for oxide samples. Accounting for neutron cross-talk events of order N = 3 resulted in fissile mass estimates with an average relative deviation of 9.58% and 10.51% for metal and oxide samples, respectively. These mass estimates were compared to a reference case (i.e., no neutron crosstalk effects) that yielded an average relative deviation of 6.81% and 4.77% for metal and oxide samples, respectively. In conclusion, the discrepancy between the estimates from the proposed model and the reference case is attributed to the assumed value of N, which sets a finite upper bound on the order of cross-talk events the model treats (i.e., the model for N = 3 assumes that a neutron will never cause more than three counts).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Impact of lump characteristics on self-shielding corrections of active neutron interrogation on fissile material in waste - 15405

Non-destructive active neutron assay techniques such as active neutron coincidence counting, differential die away analysis or the californium shuffler method rely on the interrogation of fissile nuclear materials by an external neutron source. Whether it is a spontaneous fission or (α,n) neutron source, like {sup 252}Cf or AmLi; or a fusion neutron generator, all of these techniques rely on induced secondary fission events to provide prompt or delayed neutron signals that serve as the basis for the fissile material quantification. The accuracy of these fissile mass measurements is therefore dependent on the neutron interrogation efficiencies determined for calibration reference standards and whether the measured items are comparable to the calibration condition. In such favorable conditions, it is possible to establish a reliable empirical non-linear relationship between apparent (uncorrected) and real or true fissile mass. However, in waste measurements, this is unlikely to be the case; thus a calibration approach that takes account of the neutron self-shielding effect is required. By incorporating the Self-Shielding Factor (SSF) of the calibration reference standards, a linear mass calibration factor [s{sup -1}. {sup 239}Pu{sub eq} g{sup -1}] can be found; this calibration is accurate for waste in which the fissile material is 'infinitely dilute' i.e. the waste exhibits no self-shielding; this is assumed to be representative of dispersed material. The self-shielding correction and its systematic uncertainty are then left to be treated as part of the Total Measurement Uncertainty (TMU) evaluation associated with the assay result. The derivation of the TMU contribution associated with the self-shielding effect in nuclear waste is a complex problem for which a reliable evaluation has to be performed. This is usually based on assuming 'worst case scenarios' for the distribution of material that can arise in a particular waste stream. Such an approach is conservative, but often very penalizing, especially if one considers larger lumps (those containing grams of fissile material); for such lumps, the self-shielding correction becomes very significant, potentially altering the mass of {sup 239}Pu{sub eq} by over an order of magnitude. A worst case 'nuclear safety value' may not be the most appropriate choice for building a consignment best estimate or repository sentencing value for an item. In the present work the authors describe a numerical approach to study the variability of neutron self-shielding factors associated with distributions of a wide range of lump characteristics in terms of size, shape, chemical composition, density and isotopic composition. The described novel calculation tool is capable of accurately predicting the real impact of random distribution of lumps in a waste material, and allowed to show that given reasonable, defensible, assumptions on the composition of the lumps in waste stream, the systematic error component of the DDA TMU can be significantly reduced. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Nonuniform character of the population of spin projections K for a fissile nucleus at the scission point and anisotropies in the angular distributions of fragments originating from the induced fission of nuclei

It is shown that the emergence of anisotropies in the angular distributions of fragments originating from the spontaneous and induced fission of oriented actinide nuclei is possible only if nonuniformities in the population of the projectionsM (K) of the fissile-nucleus spin onto the z axis of the laboratory frame (fissile-nucleus symmetry axis) appear simultaneously in the vicinity of the scission point but not in the vicinity of the outer saddle point of the deformation potential. The possibilities for creating the orientation of fissile nuclei for spontaneous and induced fission and the effect of these orientations on the anisotropies under analysis are considered. The role of Coriolis interaction as a unique source of the mixing of different-K fissile-nucleus states at all stages of the fission process is studied with allowance for the dynamical enhancement of this interaction for excited thermalized states of the nucleus involved that is characterized by a high energy density. It is shown that the absence of thermalization of excited states of the fissile nucleus that appear because of the effect of nonadiabaticity of its collective deformation motion in the vicinity of the scission point is a condition of conservation of the influence that transition fission states formed at the inner and outer fission barriers exerts on the distribution of the spin projections K for lowenergy spontaneous nuclear fission. It is confirmed that anisotropies observed in the angular distributions of fragments originating from the fission of nuclei that is induced by fast light particles (multiply charged ions) are due to the appearance of strongly excited equilibrium(nonequilibrium) states of the fissile nucleus in the vicinity of its scission point that have a Gibbs (non-Gibbs) distribution of projections K.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Review of Exemptions and General Licenses for Fissile Material in 10 CFR 71

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2023

Increasing the Fissile Mass Loading of High-Level Waste Glass Canisters to Greater Than 2,500 g/m 3 at the Savannah River Site

To eliminate future fissile mass loading constraints for the Savannah River Site H-Canyon Facility and Liquid Waste system, the Savannah River National Laboratory recommends a repository evaluation of a uranium fissile mass loading in glass at 7,144 g/m 3 in addition to the existing International Atomic Energy Agency safeguards and security limit of 2,500 g/m 3 total plutonium. This recommended increase above the authorized 2,500 g/m 3 fissile mass loading limit is based on concentrations of uranium and plutonium shown to produce an acceptable glass waste form rather than projections of the maximum fissile mass loading in future sludge batches. An authorized fissile mass loading limit greater than 2,500 g/m 3 will increase facility flexibility, reduce the number of high-level waste canisters produced, and potentially avoid an increase in the Department of Energy Environmental Management mission life without compromising safety or glass product quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Accelerating fissile material detection with a neutron source

A method of discriminating fissile material from non-fissile material with a digital data acquisition system that collects data at high rate, and processes large volumes of data directly to count neutrons from the unknown source and detect excess grouped neutrons to identify fission. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and inducing neutron emission therefrom, and a DC power supply that exhibits electrical ripple of less than one part per million. A neutron count histogram and Poisson count distribution are overlaid to provide a visual indication of the difference in correlation of natural and induced emitted neutrons from the radiation source to characterize the neutron source as fissile material or non-fissile material.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗