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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 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 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 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↗

DOE OSTI2018

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

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

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 OSTI2020

Accelerating fissile material detection with a neutron source

A system discriminating fissile material from nonfissile material wherein a digital data acquisition unit 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↗

DOE OSTI2014

The Future of Nuclear Archaeology: Reducing Legacy Risks of Weapons Fissile Material

This report describes the value proposition for a "nuclear archeological" technical capability and applications program, targeted at resolving uncertainties regarding fissile materials production and use. At its heart, this proposition is that we can never be sure that all fissile material is adequately secure without a clear idea of what "all" means, and that uncertainty in this matter carries risk. We argue that this proposition is as valid today, under emerging state and possible non-state nuclear threats, as it was in an immediate post-Cold-War context, and describe how nuclear archeological methods can be used to verify fissile materials declarations, or estimate and characterize historical fissile materials production independently of declarations.

Fissile material, plutonium, Highly Enriched Urani↗

DOE OSTI2015

LANL's Role in the U.S. Fissile Material Disposition Program

The process of Fissile Material Disposition is in part a result of the Advanced Recovery and Integrated Extraction System (ARIES), which is an agreement between the U.S. and Russia to dispose of excess plutonium used to make weapons. LANL is one sight that aides in the process of dismantling, storage and repurposing of the plutonium gathered from dismantled weapons. Some uses for the repurposed plutonium is fuel for commercial nuclear reactors which will provide energy for citizens.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2018

Accelerating fissile material detection with a neutron source

A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly to count neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and a DC power supply that exhibits electrical ripple on the order of less than one part per million. Certain voltage multiplier circuits, such as Cockroft-Walton voltage multipliers, are used to enhance the effective of series resistor-inductor circuits components to reduce the ripple associated with traditional AC rectified, high voltage DC power supplies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2019

Accelerating fissile material detection with a neutron source

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Fissile material holdup measurement systems: an historical review of hardware and software

The measurement of fissile material holdup is accomplished by passively measuring the energy-dependent photon flux and/or passive neutron flux emitted from the fissile material deposited within an engineered process system. Both measurement modalities--photon and neutron--require the implementation of portable, battery-operated systems that are transported, by hand, from one measurement location to another. Because of this portability requirement, gamma-ray spectrometers are typically limited to inorganic scintillators, coupled to photomultiplier tubes, a small multi-channel analyzer, and a handheld computer for data logging. For neutron detection, polyethylene-moderated, cadmium-back-shielded He-3 thermal neutron detectors are used, coupled to nuclear electronics for supplying high voltage to the detector, and amplifying the signal chain to the scaler for counting. Holdup measurement methods, including the concept of Generalized Geometry Holdup (GGH), are well presented by T. Douglas Reilly in LA-UR-07-5149 and P. Russo in LA-14206, yet both publications leave much of the evolutionary hardware and software to the imagination of the reader. This paper presents an historical review of systems that have been developed and implemented since the mid-1980s for the nondestructive assay of fissile material, in situ. Specifications for the next-generation holdup measurements systems are conjectured.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2017

Fissile Material Stockpile Declarations and Cooperative Nuclear Archaeology

With the beginning of the Cold War, the United States and the Soviet Union quickly began to produce fissile material (plutonium and highly enriched uranium, HEU) for military purposes on an industrial scale. By the mid-1950s, both countries were already making ton-quantities of fissile material per year to supply their growing nuclear arsenals. They were soon joined by the United Kingdom (1951), France (1955), China (1964), and Israel (1965)—and later by India, Pakistan, and finally North Korea. By the time most of these production efforts ended in the early 1990s, the global stockpile of military fissile material had reached fantastic levels. While the combined global nuclear weapon inventory had peaked at about 64,000 in 1986,1 the military fissile-material stockpile ultimately reached almost 2200 tons (240 tons of plutonium and 1960 tons of highly enriched uranium), which is sufficient to make about 200,000 simple nuclear weapons.2 Since the end of the Cold War, almost 700 tons of highly enriched uranium have been down-blended and other material is now reserved for naval fuel. Overall, the military material available for weapon purposes today is sufficient for more than 150,000 weapons (Figure 1).

Glaser, Alexander↗

DOE OSTI2017

Reactor Neutronics: Impact of Fissile Material

Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235 U is replaced with 239 Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratio between the 235 U system and the 239 Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239 Pu as fissile material can be up to 45% larger than when using 235 U as fissile material, and the required fissile concentration of 239 Pu is up to 48% lower than that of 235 U to achieve criticality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

A Simplified Quantitative Model for Fissile Material Production from Fast Spallation Neutrons

The Accelerator Driven System (ADS) as a reactor to produce fissile material, or for partitioning & transmutation purposes, has potential as an inherently safe reactor since it can be shut down by simply turning off the accelerator. Multiple studies have focused on the intranuclear and internuclear physics involved in the spallation process for high-energy projectiles impacting high-Z targets relevant to ADS applications. To quantify spallation neutronics, and thereby fissile material production, generally requires numerical methods such as MCNP, and such calculations are specific to the geometry under consideration. This study uses published cross sections from the ENDF database and an assumed generic cylindrical geometry for the ADS target and fertile blanket to derive analytic expressions for the production of Pu-239 and U-233, without needing to run MCNP codes, yet matching published data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Policy and Technical Issues Facing a Fissile Material (Cutoff) Treaty. Chapter 14

We report the largest obstacle to creating nuclear weapons, starting with the ones that destroyed Hiroshima and Nagasaki, has been to make sufficient quantities of fissile materials – highly enriched uranium (HEU) and plutonium – to sustain an explosive fission chain reaction.1 Recognition of this fact has, for more than fifty years, underpinned both the support for and the opposition to adoption of an international treaty banning at a minimum the production of more fissile materials for nuclear weapons, commonly referred to as a fissile material cutoff treaty (FMCT).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

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 OSTI2019

Fissile material detection using neutron time-correlations from photofission

The detection of special nuclear materials (SNM) in commercial cargoes is a major objective in the field of nuclear security. In this work we investigate the use of two-neutron time-correlations from photo-fission using the Prompt Neutrons from Photofission (PNPF) detectors in Passport Systems Inc.’s (PSI) Shielded Nuclear Alarm Resolution (SNAR) platform for the purpose of detecting ~5 kg quantities of fissionable materials in seconds. The goal of this effort was to extend the secondary scan mode of this system to differentiate fissile materials, such as highly enriched uranium, from fissionable materials, such as low enriched and depleted uranium (LEU and DU). Experiments were performed using a variety of material samples, and data were analyzed using the variance-over-mean technique referred to as Y 2F or Feynman-α. Results were compared to computational models to improve our ability to predict system performance for distinguishing fissile materials. Simulations were then combined with empirical formulas to generate receiver operating characteristics (ROC) curves for a variety of shielding scenarios. We show that a 10 second screening with a 200 μA 9 MeV X-ray beam is sufficient to differentiate kilogram quantities of HEU from DU in various shielding scenarios in a standard cargo container.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗