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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Synchrotron Based Microstructural Characterization Method Development for Pu Oxides

Identification of new plutonium oxide signatures is a high priority focus area of the technical nuclear forensics and nonproliferation communities. Comprehensive materials science studies of plutonium oxide are needed to advance the understanding of the microcrystalline properties and their relationship to macro-scale morphologies and signatures. Plutonium oxide particle growth is depenedent on the microcrystalline strucutures, grain size, and orientation. Microcrystalline structures have been found to vary for plutonium oxide based on the oxidation state of plutonium, concentration of reactants, and the physiochemical conditions of precipitation. Empirical observation confirms the relationship between processing conditions and particle morphology. High sensitivity, non destructive analysis of individual microcrystalline grains requires the use of bright sources of X-rays produced by synchrotron radiation, such as those at the Advanced Photon Source. The light source can reveal higher resolution crystallographic mapping, more sensitive chemical analysis, and higher fidelity morphological information than any current methods at SRNL Progress of FY19 was the analysis of the synchrotron data. Areas of focus included XANES oxidation state and EXAFS to measure bond distances with modeling. Most measurements for XANES indicated that the plutonium was in the +4 oxidation state, however, some indicated that there could also be a +5 oxidation state. The EXAFS measurements showed typical Pu – O bonding distance of 2.33 and Pu – Pu bonding distances of 3.81 A; however, some samples also showed that there could be shorter Pu – O distances of approximately 1.81 A, indicative of plutonium in the +5 oxidation state.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Scientific Solutions to Nuclear Waste Environmental Challenges

The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount of time it takes for one-half of the material to undergo radioactive decay.) In general, the ideal material would need to be durable for approximately 10 half-lives to allow the activity to decay to negligible levels. However, the potential health effects of each radionuclide vary depending on what type of radiation is emitted, the energy of that emission, and the susceptibility for the human body to accumulate and concentrate that particular element. Consequently, actual standards tend to be based on limiting the dose (energy deposited per unit mass) that is introduced into the environment. The Environmental Protection Agency (EPA) has the responsibility to establish standards for nuclear waste disposal to protect the health and safety of the public. For example, the Energy Policy Act of 1992 directed the EPA to establish radiation protection standards for the Yucca Mountain geologic repository for nuclear wastes. The standards for Yucca Mountain were promulgated in 2008, and limit the dose to 15 millirem per year for the first 10,000 years, and 100 milirem per year between 10,000 years and 1 million years (40 CFR Part 197; http://www.epa.gov/radiation/yucca/2008factsheet.html). So, the challenge is two-fold: (1) develop a material (a waste form) that is capable of immobilizing the waste over geologic time scales, and (2) develop a process to convert the radioactive sludge in the tanks into this durable waste form material. Glass: Hard, durable, inert, and with infinite chemical versatility Molten glass is a powerful solvent liquid, which can be designed to dissolve almost anything. When solidified, it can be one of the most chemically inert substances known to man. Nature's most famous analogue to glass is obsidian, a vitreous product of volcanic activity; formations over 17 million years old have been found. Archaeologists have found man-made glass specimens that are five thousand years old.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Modulation of Tetravalent Actinide Oxalate Morphology through Organic Compounds

Spent nuclear fuel of pressurized water reactors mainly consists of uranium, plutonium, minor actinides and fission products. The long term radiotoxicity of wastes is mainly due to plutonium and minor actinides such as neptunium, americium and curium. So, in order both to reduce the radiotoxicity of the ultimate wastes and to save uranium resources, researches for current and new generations of nuclear plants focus on the recycling of plutonium and minor actinides, to fabricate new mixed oxides nuclear fuel pellets (MOX) or transmutation targets. In this aim, production of actinide oxides is necessary. Currently, according to the Purex process, the spent fuel is dissolved in nitric acid and then valuable actinides (uranium and plutonium) are separated and purified from fission products and minor actinides in the Areva plant 'La Hague'. The resulting plutonium nitric solution is converted into a solid phase by oxalic conversion: the oxalic precipitation is followed by the calcination of the obtained oxalate into plutonium oxide. Considering either the MOX fuel or the transmutation targets fabrication from actinide oxides, the control of the particle structure and microstructure is a key factor related to the ability of oxide powder to sinter and shape as pellets. In the oxalic conversion process, the morphology of the precipitated oxalate is strongly influenced by the precipitation conditions and is kept during the calcination from oxalate into oxide. In this context, the aim of this work is to control the morphology of the final actinide oxide by changing the oxalic precipitation conditions. A lot of possibilities can be considered to modulate the morphology of actinide oxalates. Numerous parameters such as temperature, pressure, the presence of a magnetic field, the operating process (seeding, emulsions)... can modify the precipitated particles shape. Additives are also able to play a major role at the precipitation stage by means of different mechanisms involving either complexation, surface covering or structure insertion. This study focuses on this way to modify the morphology of actinides oxalates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and {sup 238}U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with {sup 238}U alone should be reprocessed in the usual way.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Chemical Differences Between Sludge Solids at the F and H Area Tank Farms

The primary source of waste solids received into the F Area Tank Farm (FTF) was from PUREX processing performed to recover uranium and plutonium from irradiated depleted uranium targets. In contrast, two primary sources of waste solids were received into the H Area Tank Farm (HTF): a) waste from PUREX processing; and b) waste from H-modified (HM) processing performed to recover uranium and neptunium from burned enriched uranium fuel. Due to the differences between the irradiated depleted uranium targets and the burned enriched uranium fuel, the average compositions of the F and H Area wastes are markedly different from one another. Both F and H Area wastes contain significant amounts of iron and aluminum compounds. However, because the iron content of PUREX waste is higher than that of HM waste, and the aluminum content of PUREX waste is lower than that of HM waste, the iron to aluminum ratios of typical FTF waste solids are appreciably higher than those of typical HTF waste solids. Other constituents present at significantly higher concentrations in the typical FTF waste solids include uranium, nickel, ruthenium, zinc, silver, cobalt and copper. In contrast, constituents present at significantly higher concentrations in the typical HTF waste solids include mercury, thorium, oxalate, and radionuclides U-233, U-234, U-235, U-236, Pu-238, Pu-242, Cm-244, and Cm-245. Because of the higher concentrations of Pu-238 in HTF, the long-term concentrations of Th-230 and Ra-226 (from Pu-238 decay) will also be higher in HTF. The uranium and plutonium distributions of the average FTF waste were found to be consistent with depleted uranium and weapons grade plutonium, respectively (U-235 comprised 0.3 wt% of the FTF uranium, and Pu-240 comprised 6 wt% of the FTF plutonium). In contrast, at HTF, U-235 comprised 5 wt% of the uranium, and Pu-240 comprised 17 wt% of the plutonium, consistent with enriched uranium and high burn-up plutonium. X-ray diffraction analyses of various FTF and HTF samples indicated that the primary crystalline compounds of iron in sludge solids are Fe 2 O 3 , Fe 3 O 4 , and FeO(OH), and the primary crystalline compounds of aluminum are Al(OH) 3 and AlO(OH). Also identified were carbonate compounds of calcium, magnesium, and sodium; a nitrated sodium aluminosilicate; and various uranium compounds. Consistent with expectations, oxalate compounds were identified in solids associated with oxalic acid cleaning operations. The most likely oxidation states and chemical forms of technetium are assessed in the context of solubility, since technetium-99 is a key risk driver from an environmental fate and transport perspective. The primary oxidation state of technetium in SRS sludge solids is expected to be Tc(IV). In salt waste, the primary oxidation state is expected to be Tc(VII). The primary form of technetium in sludge is expected to be a hydrated technetium dioxide, TcO 2 ⋅ xH 2 O, which is relatively insoluble and likely co-precipitated with iron. In salt waste solutions, the primary form of technetium is expected to be the very soluble pertechnetate anion, TcO 4 - . The relative differences between the F and H Tank Farm waste provide a basis for anticipating differences that will occur as constituents of FTF and HTF waste residue enter the environment over the long-term future. If a constituent is significantly more dominant in one of the Tank Farms, its long-term environmental contribution will likely be commensurately higher, assuming the environmental transport conditions of the two Tank Farms share some commonality. It is in this vein that the information cited in this document is provided - for use during the generation, assessment, and validation of Performance Assessment modeling results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

MPACT Fast Neutron Multiplicity System Design Concepts

This report documents work performed by Idaho National Laboratory and the University of Michigan in fiscal year (FY) 2012 to examine design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes. This project seeks to develop a new type of neutron-measurement-based plutonium assay instrument suited for assaying advanced fuel cycle materials. Some current-concept advanced fuels contain high concentrations of plutonium; some of these concept fuels also contain other fissionable actinides besides plutonium. Because of these attributes the neutron emission rates of these new fuels may be much higher, and more difficult to interpret, than measurements made of plutonium-only materials. Fast neutron multiplicity analysis is one approach for assaying these advanced nuclear fuels. Studies have been performed to assess the conceptual performance capabilities of a fast-neutron multiplicity counter for assaying plutonium. Comparisons have been made to evaluate the potential improvements and benefits of fast-neutron multiplicity analyses versus traditional thermal-neutron counting systems. Fast-neutron instrumentation, using for example an array of liquid scintillators such as EJ-309, have the potential to either a) significantly reduce assay measurement times versus traditional approaches, for comparable measurement precision values, b) significantly improve assay precision values, for measurement durations comparable to current-generation technology, or c) moderating improve both measurement precision and measurement durations versus current-generation technology. Using the MCNPX-PoliMi Monte Carlo simulation code, studies have been performed to assess the doubles-detection efficiency for a variety of counter layouts of cylindrical liquid scintillator detector cells over one, two, and three rows. Ignoring other considerations, the best detector design is the one with the most detecting volume. However, operational limitations guide a) the maximum acceptable size of each detector cell (due to PSD performance and maximum-acceptable per-channel data throughput rates, limited by pulse pile-up and the processing rate of the electronics components of the system) and b) the affordability of a system due to the number of total channels of data to be collected and processed. As a first estimate, it appears that a system comprised of two rows of detectors 5" Ø $\times$ 3" would yield a working prototype system with excellent performance capabilities for assaying Pu-containing items and capable of handling high signal rates likely when measuring items with Pu and other actinides. However, it is still likely that gamma-ray shielding will be needed to reduce the total signal rate in the detectors. As a first step prior to working with these larger-sized detectors, it may be practical to perform scoping studies using small detectors, such as already-on-hand 3" Ø $\times$ 3" detectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144 Ce as a Surrogate for Pu Accountancy

This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144 Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium in the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144 Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for using cerium, which is rather easy to analyze using passive nondestructive analysis gamma-ray spectrometry, as a surrogate for plutonium in the final analysis of TMI-2 melted fuel debris. The generation of this report is motivated by the need to perform nuclear material accountancy measurements on the melted fuel debris that will be excavated from the damaged nuclear reactors at the Fukushima Daiichi nuclear power plant in Japan, which were destroyed by the Tohoku earthquake and tsunami on March 11, 2011. Lessons may be taken from prior U.S. work related to the study of the TMI-2 core debris to support the development of new assay methods for use at Fukushima Daiichi. While significant differences exist between the two reactor systems (pressurized water reactor (TMI-2) versus boiling water reactor (FD), fresh water post-accident cooing (TMI-2) versus salt water (FD), maintained containment (TMI-2) versus loss of containment (FD)) there remain sufficient similarities to motivate these comparisons.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

Final Demolition and Disposition of 209-E Critical Mass Laboratory - 12267

The 209-E Critical Mass Laboratory was constructed in 1960 to provide a heavy shielded reactor room where quantities of plutonium or uranium in solution could be brought to near-critical configurations under carefully controlled and monitored conditions. In the late 1980's, the responsible contractor, Pacific Northwest National Laboratory (PNNL), was directed by the Department of Energy (DOE) to prepare the facility for unoccupied status. The facility was demolished under a Removal Action Work Plan pursuant to the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The funding for this project was provided by the American Recovery and Reinvestment Act (ARRA). The primary rooms of concern with regards to contamination in 209-E facility, which is over 9,000 square feet, are the criticality assembly room (CAR), the mix room, and the change room. The CAR contained two reactor hoods (HO-140 and HO-170), which each had a high efficiency particulate air (HEPA) filter system. The CAR contained 13 tanks ranging from 38 L (10 gal) to 401 L (106 gal). Tanks TK-109 and TK-110 are below grade, and were removed as part of this demolition and disposition remedy. Nonradiological and radiological hazardous substances were removed, decontaminated, or fixed in place, prior to demolition. Except for the removal of below grade tanks TK-109 and TK-110, the facility was demolished to slab-on-grade. PNNL performed stabilization and deactivation activities that included removal of bulk fissile material and chemicals, flushing tanks, stabilizing contamination within gloveboxes and hoods, and packaging and removing waste. The removal of the contaminated plutonium equipment and materials from the 209E facility presented a number of challenges similar in nature to those associated with the inventory reduction and cleanup activities at the Plutonium Finishing Plant. Although there were no bulk fissile materials or chemicals within the facility, there were residual radiological materials (isotopes of plutonium and americium) in the tanks and hoods. The complexity of the remedy was present because of the various configurations of the tanks and hoods, combined with the residual contaminants. Because of the weight and dimensional configuration, size reduction of the slab tanks, as well as removal and disposal of the different material used for moderation and absorption, were two examples of challenges that were resolved to complete the remedy. One of the key methods developed and implemented at the facility was the design and construction of a shroud to allow the cutting of the Pu contaminated tanks. The shroud design, development and implementation at the 209E Project was an example of enhanced work planning and task hazards analysis with worker involvement. This paper will present the lessons learned from the 209E facility inventory reduction activities including the shroud and other methodologies used. The initial Lessons Learned discussion for this project was scheduled for late January 2012. This facility is the first open-air demolition of a highly contaminated plutonium-contaminated facility accomplished by CH2M Hill under the Plateau Remediation Contract. The demolition was completed without spread of contamination to the workers and the surrounding area. As with any project of this complexity, there are significant accomplishments, as well as experience that can be applied to future demolition of plutonium-contaminated facilities on the Hanford Site. These experiences will be documented at a later date. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Large Volume Airborne Contamination Monitoring To Support Nuclear Processes' Deactivation and Decommissioning

Current D and D operations at Hanford have demonstrated a flaw in the current state of the art capability of defining airborne contamination boundaries - Airborne particulate emissions of Pu-239 from CM2H operations on the Hanford Pu Finishing Plant were detected well beyond areas controlled for airborne Pu, putting numerous workers at risk for radiological assimilations - Current air samplers and CAM systems were surveying insufficient volumes of air to accurately predict where respiratory protection was required - Hanford is located in a high radon environment, making air monitoring for alpha emitting actinides challenging in high alpha radon induced backgrounds. - Inexpensive HEPA based home and industrial air purifiers filter significantly higher volumes of air than commercially available continuous-air monitoring (CAM) systems. - Inexpensive models capable of filtered air volumes exceeding 1000 times that of a CAM. - Detector system can be built to detect x-rays generated from actinide decay. - Branching ratios of x-rays are 4 orders of magnitude more intense from Plutonium decay than its gamma emissions. - Detector system can be a low resolution systems, as x-ray region of interest is not terribly congested. - Detection systems evaluated will be designed based on slabs of NaI or pixelated NaI or CsI panels. - Simple graphical user interface software will be developed to operate the detection system. - Concept is to deploy some of these air-purifying systems in nonradiological areas around SRNL to confirm radon can be rejected with confidence. - Generate some filters contaminated with plutonium via electroplated Pu or lab generated simulated particles. - then deploy units in known airborne radiological environments to establish systems ability to accurately measure actinide based hot particles Lab analyses will follow up the system analyses to ensure hot particles were correctly identified. Analysis by Scintillation: - Application of scintillation media to the surface of the filters is being explored. - Evaluating slabs of ZnS(Ag), application of powdered ZnS(Cu) and spray on Perkin Elmer Enhance. - Scintillation events would then be digitized with a digital camera and quantified. - Comparing against sensitivity of a PMT or SiPD readout. - Currently evaluating a Thorlabs 8 Megapixel Monochrome Scientific CCD Camera, hermetically sealed cooled package with a wide angle lens. - Wide angle lens allows complete view of HEPA filter from 7 inches away. - Images taken of glow in the dark paint, as well as plutonium-induced fluorescence. - Plutonium was flamed mounted on a 1 inch diameter stainless steel planchet, covered with a layer of mylar and a section of scintillating ZnS obtained from Eljon. - Currently working on reducing signal-to-noise levels to boost sensitivity. - Dark box to hold contaminated or electroplated source covered filters under fabrication. Analysis by x-ray Spectroscopy: - Branching ratios for x-ray emissions from Plutonium isotopes are orders of magnitude more intense than gamma-ray emissions. - The nuclear databases are incomplete on the branching ratios of some of the isotopes. - Evaluating some custom built SrI2 x-ray spectrometers vs a conventional windowed NaI detector. - SrI2 detectors are carbon-composite-windowed 51 mm x 51 mm. - MCNP calculations to establish self absorbance of filter media on 17 keV x-ray. Radon Rejection: - Activated charcoal pre-filter acts as a Radon trap. - Adding a time lapse feature to camera to aid in radon rejection. SRS Plutonium Fuel Form D and D Operations: Currently deploying air sampler to D and D operation of Pu-238 facility at SRS to generate some field samples to analyze.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Water and Solute Travel Time Analysis for Soils Corrective Action Units 375, 411, 412, 413, 414, and 415, Revision 0

This document presents a water and solute (uranium and plutonium) travel analysis of the potential for residual radiological materials to impact groundwater at select Soils Activity sites. These sites comprise the following corrective action units (CAUs): CAU 375, Area 30 Buggy Unit Craters [Buggy and Test Cell A (TCA) sites], CAU 411, Double Tracks Plutonium Dispersion (Nellis), CAU 412, Clean Slate I Plutonium Dispersion (TTR), CAU 413, Clean Slate II Plutonium Dispersion (TTR), CAU 414, Clean Slate III Plutonium Dispersion (TTR), CAU 415, Project 57 No. 1 Plutonium Dispersion (NAFR). The scope of this document is limited to these CAUs because the radionuclide inventories from the experiments conducted at these sites are not accounted for in the source term for the current Underground Test Area (UGTA) groundwater CAUs. Thus, this analysis was necessary to determine the time frame of potential impacts to groundwater by radionuclide contamination from these CAUs.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Pit Technologies at a Glance

The National Nuclear Security Administration has tasked Los Alamos National Laboratory (LANL) with important national security work: to develop a plutonium pit manufacturing capability and deliver a minimum of 30 plutonium pits per year in the years ahead. A pit is a hollow sphere of plutonium that, when uniformly compressed by explosives inside a warhead or bomb, causes a nuclear explosion. New pits are needed to maintain the nation’s nuclear stockpile and are a key component of nuclear deterrence and world peace. LANL’s Pit Technologies (PT) division manufactures these pits safely and securely. The work takes place in the Plutonium Facility (PF-4), the nation’s most modern plutonium science and manufacturing facility. LANL is currently the only place in the country with the capability to manufacture plutonium pits.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Collated experience of the use of multi-group analysis (MGA) in non-destructive waste assay systems - 15257

The quantification of plutonium is an important aspect of non-destructive assay of Special Nuclear Material (SNM). The calculation of {sup 240}Pu{sub eff} can be done through prior knowledge of the composition of the waste but more often is done through the analysis of a measured plutonium gamma-ray spectrum performed using an isotopic analysis code. In this work, the {sup 240}Pu{sub eff} results from a variety of non-destructive waste assay systems in the UK are presented and compared using MGA v9.63C and v10. The body of the results include real measurement results of a variety of waste materials containing plutonium, the results of reanalysis of these initial waste measurements with changes to certain setup parameters and the results of reanalysis with a later version of MGA (v10). Additionally, measurements of test matrices with known source isotopics are measured and the results analysed and presented here. The use of the 'waste' flag parameter introduced to allow measurement of unusual spectra in v10 is also evaluated. When using MGA as part of the waste assay characterisation process in automatic or batch mode, results have proven to be very dependent on the quality of the spectra measured and on the setup parameters that the code is reliant upon. With a modern version of the code, the likelihood of under-reporting the total plutonium content of a 200 litre/55 gallon drum using PNCC and MGA isotopics in combination is low. Additionally, the latest 'hardened' code will also reduce the number of cases where MGA fails to run and the use of default isotopics results in an over-pessimistic evaluation of the total plutonium mass. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Solid cancer incidence other than lung, liver and bone in Mayak workers: 1948–2004

Background: Cancer incidence in the Mayak Production Association (PA) cohort was analysed to investigate for the first time whether external gamma-ray and internal plutonium exposure are associated with raised incidence of solid cancers other than lung, liver and bone (other solid cancers). Methods: The cohort includes 22 366 workers of both sexes who were first employed between 1948 and 1982. A total of 1447 cases of other solid cancers were registered in the follow-up period until 2004. The Poisson regression was used to estimate the excess relative risk (ERR) per unit of cumulative exposure to plutonium and external gamma-ray. Results: A weak association was found between cumulative exposure to external gamma-ray and the incidence of other solid cancers (ERR/Gy = 0.07; 95% confidence intervals (CIs): 0.01–0.15), but this association lost its significance after adjusting for internal plutonium exposure. There was no indication of any association with plutonium exposure for other solid cancers. Among 16 individual cancer sites, there was a statistically significant association with external exposure for lip cancer (ERR/Gy = 1.74; 95% CI: 0.37; 6.71) and with plutonium exposure for pancreatic cancer (ERR/Gy = 1.58; 95% CI; 0.17; 4.77). Conclusion: This study of Mayak workers does not provide evidence of an increased risk of other solid cancers. The observed increase in the risk of cancer of the lip and pancreas should be treated with caution because of the limited amount of relevant data and because the observations may be simply due to chance.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

241 Am Ingrowth and Its Effect on Internal Dose

Generally, plutonium has been manufactured to support commercial and military applications involving heat sources, weapons and reactor fuel. This work focuses on three typical plutonium mixtures, while observing the potential of 241 Am ingrowth and its effect on internal dose. The term “ingrowth” is used to describe 241 Am production due solely from the decay of 241 Pu as part of a plutonium mixture, where it is initially absent or present in a smaller quantity. Dose calculation models do not account for 241 Am ingrowth unless the 241 Pu quantity is specified. This work suggested that 241 Am ingrowth be considered in bioassay analysis when there is a potential of a 10% increase to the individual’s committed effective dose. It was determined that plutonium fuel mixtures, initially absent of 241 Am, would likely exceed 10% for typical reactor grade fuel aged less than 30 years; however, heat source grade and aged weapons grade fuel would normally fall below this threshold. In conclusion, although this work addresses typical plutonium mixtures following separation, it may be extended to irradiated commercial uranium fuel and is expected to be a concern in the recycling of spent fuel.

241Am↗
DOE OSTI2017

A Rapid Method for Quantification OF 242Pu in Urine Using Extraction Chromatography and ICP-MS

Occupational exposure to plutonium is generally monitored through analysis of urine samples. Typically, plutonium is separated from the sample and other actinides, and the concentration is determined using alpha spectroscopy. Current methods for separations and analysis are lengthy and require long count times. A new method for monitoring occupational exposure levels of plutonium has been developed, which requires fewer steps and overall less time than the alpha spectroscopy method. In this method, the urine is acidified, and a 239 Pu internal standard is added. The urine is digested in a microwave oven, and plutonium is separated using an Eichrom TRU Resin column. The plutonium is eluted, and the eluant is injected directly into the Inductively Coupled Plasma–Mass Spectrometer (ICP-MS). Compared to a direct “dilute and shoot” method, a 30-fold improvement in sensitivity is achieved. This method was validated by analyzing several batches of spiked samples. Based on these analyses, a combined standard uncertainty plot, which relates uncertainty to concentration, was produced. As a result, the MDA 95 was calculated to be 7.0 × 10 –7 μg L –1 , and the Lc95 was calculated to be 3.5 × 10 –7 μg L –1 for this method.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Radiation Effects on Mortality from Solid Cancers Other than Lung, Liver, and Bone Cancer in the Mayak Worker Cohort: 1948–2008

Radiation effects on mortality from solid cancers other than lung, liver, and bone cancer in the Mayak worker cohort: 1948–2008. The cohort of Mayak Production Association (PA) workers in Russia offers a unique opportunity to study the effects of prolonged low dose rate external gamma exposures and exposure to plutonium in a working age population. We examined radiation effects on the risk of mortality from solid cancers excluding sites of primary plutonium deposition (lung, liver, and bone surface) among 25,757 workers who were first employed in 1948–1982. During the period 1948–2008, there were 1,825 deaths from cancers other than lung, liver and bone. Using colon dose as a representative external dose, a linear dose response model described the data well. The excess relative risk per Gray for external gamma exposure was 0.16 (95% CI: 0.07 – 0.26) when unadjusted for plutonium exposure and 0.12 (95% CI 0.03 – 0.21) when adjusted for plutonium dose and monitoring status. There was no significant effect modification by sex or attained age. Plutonium exposure was not significantly associated with the group of cancers analyzed after adjusting for monitoring status. Site-specific risks were uncertainly estimated but positive for 13 of the 15 sites evaluated with a statistically significant estimate only for esophageal cancer. Comparison with estimates based on the acute exposures in atomic bomb survivors suggests that the excess relative risk per Gray for prolonged external exposure in Mayak workers may be lower than that for acute exposure but, given the uncertainties, the possibility of equal effects cannot be dismissed.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2013

Nuclear power fleets and uranium resources recovered from phosphates

Current light water reactors (LWR) burn fissile uranium, whereas some future reactors, as Sodium fast reactors (SFR) will be capable of recycling their own plutonium and already-extracted depleted uranium. This makes them a feasible solution for the sustainable development of nuclear energy. Nonetheless, a sufficient quantity of plutonium is needed to start up an SFR, with the plutonium already being produced in light water reactors. The availability of natural uranium therefore has a direct impact on the capacity of the reactors (both LWR and SFR) that we can build. It is therefore important to have an accurate estimate of the available uranium resources in order to plan for the world's future nuclear reactor fleet. This paper discusses the correspondence between the resources (uranium and plutonium) and the nuclear power demand. Sodium fast reactors will be built in line with the availability of plutonium, including fast breeders when necessary. Different assumptions on the global uranium resources are taken into consideration. The largely quoted estimate of 22 Mt of uranium recovered for phosphate rocks can be seriously downscaled. Based on our current knowledge of phosphate resources, 4 Mt of recoverable uranium already seems to be an upper bound value. The impact of the downscaled estimate on the deployment of a nuclear fleet is assessed accordingly. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Model nitride irradiated nuclear fuel: production, reaction with water and dilution in nitric acid

Samples of the model nuclear fuel (MNF) were made from separately synthesized nitride powders uranium-plutonium, zirconium, lanthanum and metal additives of simulators (Mo, Pd, Rh, Ag) fission products. Synthesis of initial nitride components was carried out from individual oxides, using a carbo-thermal restoration method. From MNF samples baked at a temperature of 1750 C. degrees, were made ceramographic specimens which were investigated by a scanning electron microscope. The analysis showed that distribution of the MNF components and structure of the samples corresponds to distribution of these components in the irradiated nitride fuel. The samples of MNF of nitride fuel were used for carrying out researches on dissolution in water and nitric acid. Experiments on studying the interaction of MNF with water have been made at 20, 50 and 80 C. degrees. The speed of leaching has been determined by a way of measuring the activity of water (Bq/l) in time. It is shown that an increase of temperature leads to an increase of the speed of leaching of plutonium. The formation of a precipitation, allegedly polymeric forms of plutonium, has been observed. The estimated speed of leaching of plutonium from MNF in water at 80 C. degrees is -0,0064 μgPu/(mm{sup 2}*h). From elements of FP simulators, molybdenum appears to be the most significantly leached. The dissolution of MNF in nitric acid (7,8 and 9,4 mol/l) has been carried out at boiling temperature (106-109 C. degrees). During the process of dissolution, gases were emitted. The assessment of composition of the emitted gases has been carried out. During the filtering of the solutions a precipitate whose weight makes about 2% from the weight of initial fuel has been found. Precipitate represents small powder of metal with gray color. Precipitate was investigated by a scanning electron microscope. The analysis of ranges of absorption of solution showed that the Pu(VI) share to the general content of plutonium in solution can reach 40%. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗