Accelerated Design and Certification of a Small Plutonium Air Transport Package
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Bignell, John [Sandia National Laboratories (SNL-N↗
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Scanning Electron Microscopy and X-Ray Diffraction Analysis of Tank 18 Samples
The F-Area Tank Farm (FTF) Performance Assessment (PA) utilizes waste speciation in the waste release model used in the FTF fate and transport modeling. The waste release modeling associated with the residual plutonium in Tank 18 has been identified as a primary contributor to the Tank 18 dose uncertainty. In order to reduce the uncertainty related to plutonium in Tank 18, a better understanding of the plutonium speciation in the Tank 18 waste (including the oxidation state and stoichiometry) is desired. Savannah River National Laboratory (SRNL) utilized Scanning Electron Microscopy (SEM) and X-ray Diffraction (XRD) to analyze Tank 18 samples to provide information on the speciation of plutonium in the waste material. XRD analysis of the Tank 18 samples did not identify any plutonium mineral phases in the samples. These indicates the crystalline mineral phases of plutonium are below the detection limits of the XRD method or that the plutonium phase(s) lack long range order and are present as amorphous or microcrystalline solids. SEM analysis of the Tank 18 samples did locate particles containing plutonium. The plutonium was found as small particles, usually <1 μm but ranging up to several micrometers in diameter, associated with particles of an iron matrix and at low concentration in other elemental matrices. This suggests the plutonium has an affinity for the iron matrix. Qualitatively, the particles of plutonium found in the SEM analysis do not appear to account for all of the plutonium in the sample based on concentrations determined from the chemical analysis of the Tank 18 samples. This suggests that plutonium is also distributed throughout the solids in low concentrations.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Bi-Modal Model for Neutron Emissions from PuO{sub 2} and MOX Holdup
The measurement of uranium and plutonium holdup in plants during process activity and for decommissioning is important for nuclear safeguards and material control. The amount of plutonium and uranium holdup in glove-boxes, pipes, ducts, and other containers has been measured for several decades using both neutron and gamma-ray techniques. For the larger containers such as hot cells and glove-boxes that contain processing equipment, the gamma-ray techniques are limited by self-shielding in the sample as well as gamma absorption in the equipment and associated shielding. The neutron emission is more penetrating and has been used extensively to measure the holdup for the large facilities such as the MOX processing and fabrication facilities in Japan and Europe. In some case the totals neutron emission rates are used to determine the holdup mass and in other cases the coincidence rates are used such as at the PFPF MOX fabrication plant in Japan. The neutron emission from plutonium and MOX has 3 primary source terms: 1) Spontaneous fission (SF) from the plutonium isotopes, 2) The (α,n) reactions from the plutonium alpha particle emission reacting with the oxygen and other impurities, and 3) Neutron multiplication (M) in the plutonium and uranium as a result of neutrons created by the first two sources. The spontaneous fission yield per gram is independent of thickness, whereas, the above sources 2) and 3) are very dependent on the thickness of the deposit. As the effective thickness of the deposit becomes thin relative to the alpha particle range, the (α,n) reactions and neutrons from multiplication (M) approach zero. In any glove-box, there will always be two primary modes of holdup accumulation, namely direct powder contact and non-contact by air dispersal. These regimes correspond to surfaces in the glove-box that have come into direct contact with the process MOX powder versus surface areas that have not had direct contact with the powder. The air dispersal of PuO{sub 2} particles has been studied for several decades by health physicists, because the primary health hazard of plutonium is breathing the airborne particles. The air dispersal mechanism results from the smaller particles in the top layer of powder that are lifted into the air by the electrostatic charge buildup from the alpha decay process, and the air convection carries the particles to new more distant locations. If there is open plutonium powder in a glove-box, the surfaces at more distant locations will become contaminated over time. The range of an alpha particle in a solid or powder is a function of the particle energy, the material density, and the atomic number A of the material. The average energy of a plutonium alpha particle is ∼5.2 MeV and the range in air is ∼37 mm. The range in other materials can be estimated via the Bragg-Kleenman equation. For plutonium, A is 94, and the typical density for a single particle is ∼11.5 g/cm{sup 3}, but for a powder, the density would be less because of the air packing fraction. The significance of the small diameter is that the range of the alpha particle is ∼50 μm for powder density 2.5 and significantly less for a single particle with density 11.5, so the thin deposit of separate small particles will have a greatly reduced (α,n) yield. The average alpha transit length to the surface in the isolated MOX particle would be < 2.5 μm; whereas, the range of the alpha particle is much longer. Thus, most of the alpha particles would escape from the MOX particle and be absorbed by the walls and air. The air dispersal particles will have access to a large surface area that includes the walls, whereas, the powder contact surface area will be orders of magnitude smaller. Thus, the vast majority of the glove-box surface area does not produce the full (α,n) reaction neutron yield, even from the O{sub 2} in the PuO{sub 2} as well as any impurity contamination such as H{sub 2}O. To obtain a more quantitative estimate of the neutron (α,n) yields as a function of holdup deposit thickness, we have used MCNPX calculations to estimate the absorption of alpha particles in PuO{sub 2} holdup deposits. The powder thickness was varied from 0.1 μm to 5000 μm and the alpha particle escape probability was calculated. As would be expected, as the thickness approaches zero, the escape probability approaches 1.0, and as the thickness gets much greater than the alpha particle range (∼50 μm), the escape probability becomes small. Typically, the neutron holdup calibration measurement are performed using sealed containers of thick MOX that has all 3 sources of neutrons [SF, (α,n), and M], and no significant impurities. Thus, the calibration counting rates need to include corrections for M and (α,n) yields that are different for the holdup compared with the calibration samples. If totals neutron counting is used for the holdup measurements, the variability of the (α,n) term needs to be considered.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Deliverable for FαST project: Ln Resin based PLE
This memo describes the fabrication of a polymer ligand extractant based on Eichrom's LN-1 resin. This work has been in support of the Fast Alpha Spectrometry Tool (FαST) project. The first part of LANL's role in this project is to evaluate new extractants for use in polymer ligand extractants (PLEs). The first new extractant evaluated is Di(2-ethyl hexyl) phosphoric acid (HDEHP), which is an effective metal extractant. It has very efficient chelating properties for a wide variety of metal ions. HDEHP is an amphiphillic molecule with two long hydrocarbon chains and a polar end with a phosphoryl oxygen (P=O) and an acidic -OH group as shown in Figure 1. HDEHP has shown effectiveness in extracting lanthanides, selective actinides, and other trivalent elements. Several authors have reported that lanthanides and elements with +3 oxidation state have similar extraction behavior in nitric acid. The distribution ratio for lanthanides rapidly decreases at lower nitric concentration then start to increase at higher concentration as shown in. The trivalent americium, curium, and yttrium exhibit similar trend as trivalent lanthanides. This extraction trend can be also observed from hydrogen chloride solution. This work describes the use of this ligand in a PLE to extract plutonium from solution. Polymer ligand films were prepared by dissolving HDEHP ligands and polystyrene beads in THF. The solution was directly deposited onto a 40 mm diameter stainless steel substrate using an automated pipette. HDEHP based PLEs with direct stippling method are shown in Figure 2. The solution was air dried at room temperature overnight to ensure complete evaporation of THF. The plutonium tracer solution was prepared in 0.01, 0.1, 1, and 8M nitric solutions to study the effect of nitric concentration in plutonium extraction. 0.1667 Bq 239 Pu tracer solution was directly stippled on each PLE and was allowed to equilibrate for 3 hours before removing the solution. The plutonium activity of each sample was measured by direct alpha counting to quantify the plutonium recovery by HDEHP PLE. The alpha spectra from alpha spectroscopy are shown in Figure 3. 1:5, 1:10, and 1:20 PLEs had sharp peak with low tailing. 1:2 had an extremely long tail, which is a possible indication that a large amount of ligands caused the film to not form a smooth surface. Also, it can be noted that 1:2 ratio PLE surface was not as rigid as the other ratio PLEs and it was prone to scratching during sample handing. The resolution of alpha spectra was quantified by measuring Full Width at Half of the Maximum (FWHM) using Bortels equation. The tailing component of the peak was also measured along with FWHM. The peak resolutions and tailing measurements for 0.1M nitric solution samples are given in Table 1. The best resolution was achieved with 1:5 PLE and worst was given by 1:2 PLE. The plutonium recovery by HDEHP PLE was dependent on both nitric concentration and ligand to polymer ratio. 1:2 PLE consistently had the highest recovery followed by 1:5 as shown in Figure 4. It should be noted that 1:2 ratio PLEs consistently had long tailing and the ROI of the spectrum had to be increased to encompass total counts from the tracer. 1:10 and 1:20 PLEs had close to zero percent recovery in all nitric concentration except for 0.01M. The highest plutonium recovery was observed for 0.1M nitric acid. 1:5 PLE gave the best combination of alpha spectroscopy resolution and plutonium recovery. Radiography image of samples were generated to study the plutonium distribution on the PLE surface. Samples were placed on an imaging plate (Fujifilm BAS-TR 2025) for 24 hours and the plate was scanned using GE Typhoon FLA 7000 system. The radiography image in Figure 5 shows uneven distribution with hot spots along the edge and in the center of the samples. These hot spots may be the result of highly localized concentration of ligands or surface defects that were observed in SEM. This unevenness in distribution may cause inaccurate activity measurement by alpha spectroscopy due to a bias in the energy calibration. PLEs based on HDEHP have been evaluated. The concentration of ligand in polystyrene is critical in the performance of the PLE; both for extraction efficiency as well as for alpha spectrometry results. The optimal PLE using this ligand has a concentration of 1:5 ligand to polystyrene. Under laboratory conditions, this configuration can produce similar results to those obtained for DIPEX based PLEs from previous studies. The concentration is very important to the results obtained for this ligand. This PLE is a good candidate for use in later field studies. Our next report will detail the evaluation of PLEs utilizing crown ether ligands.
36 MATERIALS SCIENCE↗
Use of Separator with Regular Granular Filling in Extraction Processes - 12209
The separator with a regular granular filling has been developed to be used for the operations of the removal of carbon-black impurities from the aqueous flow; the intra-cycle regeneration of back-extractants; and the concentrating back-extraction of plutonium in the extraction SNF reprocessing technology. The process conditions for those operations have been experimentally tested using this separator. Thus, the separator for operations of concentrating reextraction of plutonium and regeneration recycling extractant allows to: - derive plutonium from the organic stream into an extremely concentrated form, reducing the amount of liquid radioactive waste (raffinate, decantate) plutonium branches to a minimum; - controlling the processes of concentration of uranium and plutonium, get uranium-plutonium product with required concentration and the amount of metal with required ratio; - eliminate contamination of the emulsion flows; - simplify the purification of uranium from the extract plutonium; - test results showed that when working on a solution 'hydrazine-DTPA' in the separator it is achieved the extraction of plutonium re-extract - 82% TC - at 56-65%. With the transition to the reextraction by solution 'U (IV)-hydrazine', of plutonium into re-extract increases to 95%, technetium - up to 83%; - regenerate the extractant recycling minimizing the amount of liquid radioactive waste generated as a technology used as well as alternative 'salt-free' systems; - it is found that the regeneration of the extractant solution of soda and EDA in the separator at a load of 3.2 m{sup 3} / (m{sup 2}.h) it can be achieved almost complete phase separation. Residual carryover of 0,005-0,006% emulsion is on the verge of analytical detection. - It is shown that in the depth of separation phase a separation of the extractant regeneration of quality superior is more than 5-7 times to the extraction. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Benchmark of the Chlorine Worth Study Experiments in Support of Chlorine Nuclear Data Validation for Nuclear Criticality Safety
The Chlorine Worth Study (CWS) was a critical experiment to address an urgent need for thermal chlorine nuclear data validation in plutonium systems. This urgent need is tied directly to plutonium recycle and recovery operations in the plutonium facility at Los Alamos National Laboratory, where exceptionally conservative criticality safety limits are used because no credit is taken for the neutron capture by chlorine. The experiment used weapons-grade plutonium metal plates clad in stainless steel, known as the PANN (plutonium aluminum no nickel) ZPPR (zero power physics reactor) plates. The plutonium was reflected and moderated by high-density polyethylene and included combinations of polyvinyl chloride (PVC) and chlorinated polyvinyl chloride (CPVC) as absorbers. The experiment and benchmark included three configurations mimicking 30 g 239 Pu/L plutonium, 300 g 239 Pu/L plutonium, and 600 g 239 Pu/L plutonium in an aqueous chloride solution. Uncertainties in the benchmark included five broad categories: (1) criticality measurement, (2) mass and density, (3) dimensions, (4) material compositions, and (5) positioning. The largest contribution to the overall uncertainties for all three cases came from the material compositions, in particular the PVC and CPVC absorber compositions. A detailed model was created to be a near match (that is within expectations of transport code users) and a simplified model was created to minimize offset dimensions and expedite modeling for code validation. Sample calculations were completed in MCNP6.3 with ENDF/B-VIII.0 and ENDF/B-VII.1 nuclear data. For the detailed and simplified models, the average difference between the computed and experimental k eff was 951 pcm. CWS will serve as the key validation experiment for nuclear criticality safety in support of aqueous chloride operations. The sensitivity to the chlorine capture cross section is orders of magnitude greater than other existing benchmarks. The current limits, as defined by nuclear criticality safety, are 520 g Pu per batch, i.e. the minimum critical mass of the Pu solution infinitely reflected by water [Criticality Handbook: Volume II, (1969)]. This extremely conservative critical mass limit does not credit any neutron capture by chlorine (in particular neutron capture by 35 Cl) and greatly impedes the throughput required for current and future operations.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Effects of Aging on PuO2∙xH2O Particle Size in Alkaline Solution
Between 1944 and 1989, 54.5 metric tons of the United States’ weapons-grade plutonium and an additional 12.9 metric tons of fuel-grade plutonium were produced and separated from irradiated fuel at the Hanford Site. Acidic high-activity wastes containing around 600 kg of plutonium were made alkaline and discharged to underground storage tanks from separations, isolation, and recycle processes to yield average plutonium concentration of about 0.003 grams per liter (or ~0.0002 wt%) in the ~200 million liter tank waste volume. The plutonium is largely associated with low-solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g., iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of plutonium hydrous oxide, PuO2∙xH2O, could undergo sufficient crystal growth through dissolution and reprecipitation in the alkaline tank waste to potentially become separable from neutron absorbing constituents by settling or sedimentation. Thermodynamic considerations and laboratory studies of systems chemically analogous to tank waste show that the plutonium formed in the alkaline tank waste by precipitation through neutralization from acid solution probably entered as 2–4-nm PuO2∙xH2O crystallite particles that, because of their low solubility and opposition from radiolytic processes, grow from that point at exceedingly slow rates, thus posing no risk of physical segregation.
aging↗
Project 57 Air Monitoring Report: January 1 through December 31, 2017
On April 24, 1957, the Atomic Energy Commission (AEC) (now the Department of Energy [DOE]) conducted the Project 57 safety experiment in western Emigrant Valley northeast of the Nevada National Security Site (NNSS, formerly the Nevada Test Site) on lands withdrawn by the Department of Defense (DOD) for the Nevada Test and Training Range (NTTR). The test was undertaken to develop (1) a means of estimating plutonium distribution resulting from a non-nuclear detonation; (2) biomedical evaluation techniques for use in plutonium-laden environments; (3) methods of surface decontamination; and (4) instruments and field procedures for prompt estimation of alpha contamination (Shreve, 1958). Although the test did not result in the fission of nuclear materials, it did disseminate plutonium across the land surface. Following the experiment, the AEC fenced the contaminated area and returned control of the surrounding land to the DOD. Various radiological surveys were performed in the area and in 2007, the DOE expanded the demarked Contamination Area (CA) by posting signs 200 to 400 ft (60 to 120 m) outside of the original fence. Plutonium in soil attaches preferentially to smaller particles (Tamura, 1985; Friesen, 1992; Murarik et al., 1992; and Misra et al., 1993). Therefore, redistribution of soil particles by wind (dust) and water are the mechanisms most likely to transport plutonium beyond the boundary of the Project 57 CA. Monitoring was implemented in 2011 by Desert Research Institute (DRI) to determine if radionuclide contamination was detectable in samples of airborne dust and to characterize meteorological and environmental parameters that influence dust transport. Initially, two monitoring stations consisting of radiological, meteorological, and dust sampling equipment were installed near the southeast and northeast corners of the CA. In January 2015, the original monitoring stations were dismantled and moved farther to the west along the CA boundary. This move was made to place the monitoring stations downwind of ground zero and the High Contamination Area (HCA) during the dominant northerly and southerly winds. Samples of particles suspended in the air are collected every two weeks and submitted for laboratory assessments of gross alpha and gross beta radioactivity, and for determinations of gamma-emitting radionuclides. In calendar year (CY) 2017, the mean gross alpha concentration at monitoring station P57-4 was approximately 4.25 times the value at monitoring station P57-3. In CY2015 and CY2016, the difference in mean gross alpha concentration at the two stations was approximately 1.5. The mean gross alpha concentrations at P57-3 and P57-4 were approximately 1.9 and 7.9 times, respectively, the mean gross alpha concentrations at the surrounding Community Environmental Monitoring Program (CEMP) stations. The minimum gross alpha concentrations reported for the Project 57 stations are in the range of the minimum values observed at the CEMP stations. However, maximum gross alpha concentrations are greater than the maximum concentrations observed at the surrounding CEMP stations. Gamma spectroscopy analyses resulted in the detection of seven radionuclides. Except for americium-241 (Am-241), all reported radionuclides occur naturally in the environment. The Am-241 detection by gamma spectroscopy was confirmed by alpha spectroscopy. During each quarter of CY2017, four suspended particulate matter samples were selected for alpha spectroscopy analysis. Two were selected each quarter from the regular biweekly samples collected at each of the two monitoring stations. Of the eight samples from P57-3, one was reported to have plutonium-238 (Pu-238) above the minimum detectable activity (MDA) and five were reported to have plutonium 239/240 (Pu-239/240) above the MDA. Of the eight samples from P57-4, three were reported to have Pu-238 above the MDA and six were reported to have Pu-239/240 above the MDA. Soil material is also transported by saltation, which is a wind driven phenomena that bounces sand-sized soil particles across the land surface because they are too heavy to be suspended in the air. Samples of particles transported by saltation were collected downwind and upwind of the CA at both monitoring stations in May 2017 after a 190-day deployment and again in September 2017 after a 114-day deployment. Mass of the saltation samples was determined after oven drying. The September sample had greater mass than the May sample. The same relationship was observed for the suspended particulate matter samples during periods of saltation sample collection. The difference in the collected mass appears to be because of the amount of soil moisture. Higher soil moisture exerts additional isotopic bonds between soil particles and water molecules, and between water molecules that envelop soil particles, making it more difficult for wind to raise the particles as dust. In general, samples collected from saltation traps facing south contained more material than samples collected from traps facing north. This suggests that although saltation material may have been moving back and forth in conjunction with the two dominant wind directions, there was a net trend for saltation material to be transported toward the north. Radiological analyses of the saltation samples show that the smaller particle size faction (<63 μm) generally produced the highest activity levels of Am-241, Pu-238, and Pu-239/240. In general, the isotope activity levels decreased as the particle size increased. These results are consistent with the expectation that the greatest potential for transport of radiological contamination is with the smaller particles. Thermoluminescent dosimeters (TLDs) indicated that the average annual external radioactivity dose at the monitoring stations is slightly higher than the dose determined at the surrounding CEMP stations, but approximately half of the estimated national average dose received by the general public from exposure to natural sources. The TLDs at the Project 57 monitoring stations are exposed to both natural sources (terrestrial and cosmic) and radioactive releases from the Project 57 CA. Winds in excess of approximately 15 mph (24 km/hr) begin to generate dust movement by saltation or suspension, as determined by the monitoring instruments. Saltated particles, PM10 (i.e., inhalable dust), and PM2.5 (i.e., fine particulate dust) exhibit an approximately exponential increase with increasing wind speed. The greatest concentrations of dust occurred for winds exceeding 20 mph (32 km/hr). When the wind/dust analysis was performed for winds separated into the dominant northerly and southerly wind directions, it was evident that the northerly winds generated the higher PM10 concentrations and had the dominate influence on the PM10 concentration determined for all wind speeds when all wind directions were evaluated together. During the year, winds in excess of 20 mph (32 km/hr) occurred approximately one percent of the time. Although winds sufficient to generate dust occur at the Project 57 site, they are infrequent and of short duration. A preliminary assessment of individual wind events suggests that dust generation in response to specific wind conditions is highly variable. Wind speeds that would be expected to generate noticeable dust may not do so in every instance. Additionally, saltation transport in response to wind conditions was less common than expected. These variations were likely because of the influence of meteorological and environmental parameters other than wind.
54 ENVIRONMENTAL SCIENCES↗
NMIS with Imaging and Gamma Ray Spectrometry for Pu, HEU, HE and Other Materials
The Nuclear Material Identification System (NMIS) has been under development at ORNL and the National Nuclear Security Administration (NNSA) Y-12 National Security Complex since 1984. In the mid-1990s, what is now the US Department of Energy (DOE) Office of Nuclear Verification (ONV) realized that it was a useful technology for future arms control treaty applications and supported further development of the system. In 2004, fast-neutron imaging was incorporated into the system. In 2007, the ONV decided to develop a fieldable version of the system, designated as FNMIS, for potential use in future treaties. The FNMIS is being developed to be compatible with the eventual incorporation of gamma-ray spectrometry and an information barrier. This report addresses how and what attributes could be determined by the FNMIS system with gamma-ray spectrometry. The NMIS is a time-dependent coincidence system that incorporates tomographic imaging (including mapping of the fission sites) and gamma-ray spectrometry. It utilizes a small, lightweight (30 lb), portable deuterium-tritium (DT) neutron (14.1 MeV) generator (4 x 10 7 neutrons/second) for active interrogation and can also perform passive interrogation. A high-purity germanium (HPGe) gamma-ray detector with multichannel analysis can be utilized in conjunction with the source for active interrogation or passively. The system uses proton recoil scintillators: 32 small 2.5 x 2.5 x 10.2-cm-thick plastic scintillators for imaging and at least two 2 x 2 arrays of 27 x 27 x 10-cm-thick plastic scintillators that detect induced fission radiation. The DT generator contains an alpha detector that time and directionally tags a fan beam of some of the neutrons emitted and subdivides it into pixels. A fast (1 GHz) time correlation processor measures the time-dependent coincidence among all detectors in the system. A computer-controlled scanner moves the small detectors and the source appropriately for scanning a target object for imaging. The system is based on detection of transmitted 14.1 MeV neutrons, fission neutrons, and gamma rays from spontaneous, inherent source fission of the target, fission neutrons and gamma rays induced by the external DT source, gamma rays from natural emissions of uranium and plutonium, and induced gamma-ray emission by the interaction of the 14.1 MeV neutrons from the DT source. The NMIS can and has been used with a time-tagged californium spontaneous fission source. It has also been used with pulsed interrogation sources such as LINACs, DT, and deuterium-deuterium (DD) sources. This system is uniquely suited for detection of shielded highly enriched uranium (HEU), plutonium, and other special nuclear materials and detection of high explosives (HE) and chemical agents. The NMIS will be adapted to utilize a trusted processor that incorporates information barrier and authentication techniques using open software and then be useful in some international applications for materials whose characteristics may be classified. The proposed information barrier version of the NMIS system would consist of detectors and cables, the red (classified side) computer system, which processes the data, and the black (unclassified side) computer, which handles the computer interface. The system could use the 'IB wrapper' concept proposed by Los Alamos National Laboratory and the software integrity (digital signatures) system proposed by Sandia. Since it is based entirely on commercially available components, the entire system, including NMIS data acquisition boards, can be built with commercial off-the-shelf components. This system is being developed into a fieldable system (FNMIS) for potential arms control treaties by the ONV. The system will be modularly constructed with the RF shielded modules connected to the processor by appropriate control and signal cable in metal conduit. The FNMIS is presently being designed for eventual incorporation of gamma-ray spectrometry and an information barrier to protect classified information. The system hardware and software can be configured to obtain the following: plutonium presence, plutonium mass, Pu-240/239 ratio, plutonium geometry, plutonium metal vs non-metallic (absence of metal), time (age) since processing for plutonium (or last purification), uranium presence, uranium mass, uranium enrichment, uranium geometry, uranium metal vs non-metallic compound (absence of metal), beryllium presence and mass, tritium and deuterium gas bottle presence, HE, and chemical weapons. A matrix of the quantities determined, the method of determination, whether active (external neutron source) or passive, and the measurement equipment involved is given in the Tables 1-4. Some of these attributes can be obtained by multiple data analysis methods. The gamma-ray spectrometry methods for HEU, plutonium, and HE have been developed by other laboratories, are well known, and will be incorporated.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Introduction to Pits and Weapons Systems (U) [Slides]
A Nuclear Explosive Package includes the Primary, Secondary, Radiation Case and related components. This is the part of the weapon that produces nuclear yield and it converts mechanical energy into nuclear energy. The pit is composed of materials that allow mechanical energy to be converted to electromagnetic energy. Fabrication processes used are typical of any metal fabrication facility: casting, forming, machining and welding. Some of the materials used in pits include: Plutonium, Uranium, Stainless Steel, Beryllium, Titanium, and Aluminum. Gloveboxes are used for three reasons: (1) Protect workers and public from easily transported, finely divided plutonium oxides - (a) Plutonium is very reactive and produces very fine particulate oxides, (b) While not the 'Most dangerous material in the world' of Manhattan Project lore, plutonium is hazardous to health of workers if not properly controlled; (2) Protect plutonium from reactive materials - (a) Plutonium is extremely reactive at ambient conditions with several components found in air: oxygen, water, hydrogen, (b) As with most reactive metals, reactions with these materials may be violent and difficult to control, (c) As with most fabricated metal products, corrosion may significantly affect the mechanical, chemical, and physical properties of the product; and (3) Provide shielding from radioactive decay products: α, γ, and η are commonly associated with plutonium decay, as well as highly radioactive materials such as 241 Am and 238 Pu.
36 MATERIALS SCIENCE↗
JOWOG 22/2 - Actinide Chemical Technology (July 9-13, 2012)
The Plutonium Science and Manufacturing Directorate provides world-class, safe, secure, and reliable special nuclear material research, process development, technology demonstration, and manufacturing capabilities that support the nation's defense, energy, and environmental needs. We safely and efficiently process plutonium, uranium, and other actinide materials to meet national program requirements, while expanding the scientific and engineering basis of nuclear weapons-based manufacturing, and while producing the next generation of nuclear engineers and scientists. Actinide Process Chemistry (NCO-2) safely and efficiently processes plutonium and other actinide compounds to meet the nation's nuclear defense program needs. All of our processing activities are done in a world class and highly regulated nuclear facility. NCO-2's plutonium processing activities consist of direct oxide reduction, metal chlorination, americium extraction, and electrorefining. In addition, NCO-2 uses hydrochloric and nitric acid dissolutions for both plutonium processing and reduction of hazardous components in the waste streams. Finally, NCO-2 is a key team member in the processing of plutonium oxide from disassembled pits and the subsequent stabilization of plutonium oxide for safe and stable long-term storage.
36 MATERIALS SCIENCE↗
Expanded Intercomparison of Nuclear Data Libraries Using Jupiter and Jupiter High-240 Experiments
There is a limited availability of plutonium experiments with sensitivity to lead in the ICSBEP (International Handbook of Evaluated Criticality Safety Benchmark Experiments) Handbook. The Jupiter and Jupiter High-240 experiments were performed at the National Criticality Experiments Research Center as a collaborative effort between Los Alamos National Laboratory and the Japan Atomic Energy Agency to assess lead void coefficients in a plutonium-lead system containing weapons- and reactor-grade plutonium, respectively. Concurrent with benchmark development, an intercomparison of calculations using different nuclear data libraries has been performed to assess the usability of the experimental data for nuclear data adjustment in a “softer-that-fast” neutron energy spectrum. Eigenvalue calculations using MCNP with the ENDF/B-VIII.0 and TENDL-2021 nuclear data libraries calculate closest to the benchmark values for Jupiter. Calculations using JENDL-5 and ENDF/B-VIII.1 match best with the Jupiter High-240 values. Lead void worth calculations using the various nuclear data libraries are all within 3σ of their respective measured values. Perturbation studies between ENDF/B-VIII.0 and ENDF/B-VIII.1 demonstrate an approximate increase in calculated eigenvalues for the Jupiter series experiments by ~240 pcm for plutonium (mostly 239 Pu) and ~120 pcm for lead accompanied by a decrease contributed by ~113 pcm for copper and ~13 pcm for stainless steel. Nuclear data sensitivities and uncertainties investigated using Whisper show slightly lower sensitivity to scatter than a lead-reflected plutonium sphere but greater sensitivity to neutron capture. The sensitivities between Jupiter and Jupiter High-240 for lead are very similar for both ENDF/B-VIII.0 and ENDF/B-VIII.1 nuclear data. These benchmarks are more sensitive to neutron capture in lead than other plutonium benchmark experiments and would be useful for both lead and 240 Pu validation. In conclusion, with the high degree of compensating effects between copper, lead, and plutonium cross sections, additional isolated Pb-Pu and Cu-Pu benchmarks would be beneficial in improving these nuclear data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor
This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2 -PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Colloid-Facilitated Actinide Transport in a Cementitious-Impacted SRS Groundwater
The objective of this report is to provide a literature review relevant to the question of whether colloid facilitated transport of plutonium is enhanced in a cementitious groundwater system on the Savannah River Site (SRS). Contaminant transport is commonly described as taking place in a system with a mobile aqueous phase and an immobile solid phase. There has been an increasing awareness of a third phase, a mobile solid phase, also referred to as a mobile colloidal phase. Mobile colloids consist of organic and/or inorganic submicron-particles that move with groundwater flow. When radionuclides are associated with the mobile colloids, the net effect is that radionuclides can move faster through the subsurface system than would be predicted by transport models that do not include mobile colloids. It is important to distinguish between subsurface colloids and subsurface mobile colloids. The subsurface environment includes an enormous reservoir of colloids, but only a tiny fraction, if any, are mobile. Mobile colloid formation is commonly described as involving a three-step process: genesis, stabilization, and transport. It was concluded that there is a strong likelihood that submicron colloids of plutonium exist near the source term, (i.e., the genesis step is likely completed). While such particles have not been directly detected, it is highly likely that plutonium could either attach to submicron particles in the sediment or that submicron plutonium fragments exist at the source. However, the tendency for these plutonium colloids to move in the SRS subsurface is extremely low, especially in engineered cementitious environments where the ionic strength of the solution and the elevated concentrations of divalent cations greatly curtails colloid suspension stability (i.e., the stabilization step is likely not completed). Together these data strongly indicate that plutonium-bearing colloids would likely exist in the near field of a tank closure facility, but the colloids would be unlikely to be mobile, thereby providing a vector for enhanced transport. Under cementitious leachate impacted groundwater conditions, it is reasonable to assume plutonium transport occurs primarily as a two-phase system, a mobile aqueous phase and an immobile solid phase.
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Impact of the deployment schedule of fast breeding reactors in the frame of French act for nuclear materials and radioactive waste management
In the frame of the French Act of June 28, 2006 on 'a sustainable management of nuclear materials and radioactive waste' EDF R and D assesses various research scenarios of transition between the actual French fleet and a Generation IV fleet with a closed fuel cycle where plutonium is multi-recycled. The basic scenarios simulate a deployment of 60 GWe of Sodium-cooled Fast Reactors (SFRs) in two steps: one third from 2040 to 2050 and the rest from 2080 to 2100 (scenarios 2040). These research scenarios assume that SFR technology will be ready for industrial deployment in 2040. One of the many sensitivity analyses that EDF, as a nuclear power plant operator, must evaluate is the impact of a delay of SFR technology in terms of uranium consumptions, plutonium needs and fuel cycle utilities gauging. The sensitivity scenarios use the same assumptions as scenarios 2040 but they simulate a different transition phase: SFRs are deployed in one step between 2080 and 2110 (scenarios 2080). As the French Act states to conduct research on minor actinides (MA) management, we studied different options for 2040 and 2080 scenarios: no MA transmutation, americium transmutation in heterogeneous mode based on americium Bearing Blankets (AmBB) in SFRs and all MA transmutation in heterogeneous mode based on MA Bearing Blankets (MABB). Moreover, we studied multiple parameters that could impact the deployment of these reactors (SFR load factor, increase of the use of MOX in Light Water Reactors, increase of the cooling time in spent nuclear fuel storage...). Each scenario has been computed with the EDF R and D fuel cycle simulation code TIRELIRE-STRATEGIE and optimized to meet various fuel cycle constraints such as using the reprocessing facility with long period of constant capacity, keeping the temporary stored mass of plutonium and MA under imposed limits, recycling older assemblies first... These research scenarios show that the transition from the current PWR fleet to an equivalent fleet of Generation IV SFR can follow different courses. The design of SFR-V2B that we used in our studies needs a high inventory of plutonium resulting in tension on this resource. Several options can be used in order to loosen this tension: our results lead to favour the use of axial breeding blanket in SFR. Load factor of upcoming reactors has to be regarded with attention as it has a high impact on plutonium resource for a given production of electricity. The deployment of SFRs beginning in 2080 instead of 2040 following the scenarios we described creates higher tensions on reprocessing capacity, separated plutonium storage and spent fuel storage. In the frame of the French Act, we studied minor actinides transmutation. The flux of MA in all fuel cycle plants is really high, which will lead to decay heat, a and neutron emission related problems. In terms of reduction of MA inventories, the deployment of MA transmutation cycle must not delay the installation of SFRs. The plutonium production in MABB and AmBB does not allow reducing the use of axial breeding blankets. The impact of MA or Am transmutation over the high level waste disposal is more important if the SFRs are deployed later. Transmutation option (americium or all MA) does not have a significant impact on the number of canister produced nor on its long-term thermal properties. (authors)
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Criticality safety analysis on fissile materials in Fukushima reactor cores
The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.
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Lung Cancer in the Mayak Workers Cohort: Risk Estimation and Uncertainty Analysis
The workers at the Mayak nuclear facility near Ozyorsk, Russia are a primary source of information about exposure to radiation at low-dose rates, since they were subject to protracted exposures to external gamma rays and to internal exposures from plutonium inhalation. Here we re-examine lung cancer mortality rates and assess the effects of external gamma and internal plutonium exposures using recently developed Monte Carlo dosimetry systems. Using individual lagged mean annual lung doses computed from the dose realizations, we fit excess relative risk (ERR) models to the lung cancer mortality data for the Mayak Workers Cohort using risk-modeling software. We then used the corrected information matrix (CIM) approach to widen the confidence intervals of ERR by taking into account the uncertainty in doses represented by multiple realizations from the Monte Carlo dosimetry systems. Findings of this work revealed that there were 930 lung cancer deaths during follow-up. Plutonium lung doses (but not gamma doses) were generally higher in the new dosimetry systems than those used in the previous analysis. This led to a reduction in the risk per unit dose compared to prior estimates. The estimated ERR/Gy for external gamma-ray exposure was 0.19 (95% CI: 0.07 to 0.31) for both sexes combined, while the ERR/Gy for internal exposures based on mean plutonium doses were 3.5 (95% CI: 2.3 to 4.6) and 8.9 (95% CI: 3.4 to 14) for males and females at attained age 60. Accounting for uncertainty in dose had little effect on the confidence intervals for the ERR associated with gamma-ray exposure, but had a marked impact on confidence intervals, particularly the upper bounds, for the effect of plutonium exposure [adjusted 95% CIs: 1.5 to 8.9 for males and 2.7 to 28 for females]. In conclusion, lung cancer rates increased significantly with both external gamma-ray and internal plutonium exposures. Accounting for the effects of dose uncertainty markedly increased the width of the confidence intervals for the plutonium dose response but had little impact on the external gamma dose effect estimate. Adjusting risk estimate confidence intervals using CIM provides a solution to the important problem of dose uncertainty. This work demonstrates, for the first time, that it is possible and practical to use our recently developed CIM method to make such adjustments in a large cohort study.
uncertainty, radiation risk estimation, Mayak Prod↗
Solidified Organics Testing
The Radiochemical Engineering Development Center currently has plutonium-containing liquid organic waste that must be solidified before final disposal as solid debris waste. The documented safety analysis for the Radiochemical Engineering Development Center limits the amount of plutonium that can be released from solid debris waste forms during a fire scenario, which is less than 10 -4 airborne release fraction. Experimental results for plutonium release fractions from burning various materials has been compiled by the Department of Energy [DOE, 1994]. Combustion of organic liquids gave an average airborne release fraction of 6 × 10 -3 and combustion of surface-contaminated solids (paper, plastic, etc.) gave an average airborne release fraction of 5 × 10 -4 . Burning of 30% tri- n -butyl phosphate (TBP) in kerosene gave cerium release fractions of 5.6–7.7 × 10 -3 [Mishima and Schwendiman, 1973]. Earlier testing had evaluated many different grout-based formulations to solidify plutonium-containing organic waste liquids, and the ones used in these tests were shown to produce monolithic, fairly homogeneous solids. Each of the waste forms tested would meet the waste acceptance criteria for disposal at the Waste Isolation Pilot Plant as transuranic waste, but there were questions about whether they would meet the plutonium release criteria during a fire. The purpose of the work described here was to quantify the airborne release fraction for cerium, as a surrogate for plutonium, in these proposed waste forms. Cerium is often used as a simulant for plutonium in bench-scale testing, and cerium nitrate hexahydrate [Ce (NO 3 ) 3 ·6H 2 O] was added to the organic solutions that were solidified to prepare the surrogate waste forms for these tests. The waste forms were subjected to the designated fire scenario of 600ºC for 1 hour, and any cerium in the off-gas was trapped using a quartz filter followed by a scrubber containing 0.1 M NaOH, which is the same as used by Mishima and Schwendiman, [1973].
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