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At least 19 records

A new TEVA-DGA chromatography procedure to separate Pu, Am, and Np from bulk U materials

Trace actinides present within uranium (U) material are diagnostic signatures of U processing history. For example, the emplacement of U material within the high neutron-flux environment of a nuclear reactor can cause neutron capture of U to form 241-plutonium ( 241 Pu). Plutonium-241 then decays to progeny isotopes 241-americium ( 241 Am) and 237-neptunium ( 237 Np); therefore, the presence and relative abundances of these trace actinides is indicative of the material’s production history and intended employment (Mayer et al., 2013). During a pre-detonation nuclear forensics investigation, measurement of trace actinide 241 Pu, 241 Am, and 237 Np concentrations by an analytical laboratory may be requested. However, chemically purifying these elements from bulk U is a challenging and involved procedure requiring several sequential chromatography columns across 4+ days. To expedite Pu-Am-Np separation from bulk U material, and to improve the yield recovery of these elements, this DHS Postdoc Fellowship has worked to create a new chromatography separation chemistry.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Fabrication and testing of antimony doped bismuth tri-iodide semiconductor gamma-ray detectors

We fabricated antimony (Sb) doped bismuth tri-iodide (BiI 3 ) radiation detectors from large single crystals that were grown using the modified vertical Bridgman technique. Detectors were prepared by subjecting the crystal surfaces to different mechanical and chemical treatments. Surface quality of the detectors was evaluated using optical microscopy. The influence of surface quality on detector performance was analyzed by measuring the leakage current for each of the detectors. The radiation response of the detectors was measured using an Americium ( 241 Am) gamma-ray source at room temperature. The first successful use of BiI 3 detectors for gamma-ray spectroscopy is reported here with energy resolution of 7.5% at 59.5 keV. The mobility-lifetime product for electrons was also estimated to be about 5.2 x 10 -4 cm 2 /V.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Estimating the AmLi Neutron Spectrum from Measured Ring Ratios

These are a set of slides on estimating the AmLi neutron spectrum from measured ring ratios. The IAEA uses an AmLi source in the Uranium Neutron Coincidence Collar (UNCL) to verify compliance with nonproliferation treaties. The UNCL requires calibration with known uranium samples. The AmLi spectrum isn’t known well enough to allow simulated calibrations. Alphas lose energy traveling through AmO 2 particle of unknown size. Energy reduction below Li threshold enhances O contribution. Unknown Li matrix material affects neutron production and thermalization. There is large variation in spectra from each element. Other topics covered include: applications, physics considerations, current spectra, measurement overview, measurement results - variation between sources, simulations, spectra fitting, other simulations, and conclusions.

07 ISOTOPE AND RADIATION SOURCES↗

Low-energy gamma ray inspection of brazed aluminum joints

Americium 241 serves as a suitable radioisotope /gamma ray source/ and exposure probe for radiographic inspection of brazed aluminum joints in areas of limited accessibility. The powdered isotope is contained in a sealed capsule mounted at the end of a spring-loaded pushrod in the probe assembly.

Brown, J. A.↗

Neutron irradiation of Am-241 effectively produces curium

Computer study was made on the production of multicurie amounts of highly alpha-active curium 242 from americium 241 irradiation. The information available includes curium 242 yields, curium composition, irradiation data, and production techniques and safeguards.

Anderson, R. W.↗

Lunar Surface Stirling Power Systems Using Am-241

For many years NASA has used the decay of Pu-238 (in the form of the General Purpose Heat Source (GPHS)) as a heat source for Radioisotope Thermoelectric Generators (RTG), which have provided electrical power for many NASA missions. While RTG's have an impressive reliability record for the missions in which they have been used, their relatively low thermal to electric conversion efficiency (-5% efficiency) and the scarcity of Plutoinium-238 (Pu-238) has led NASA to consider other power conversion technologies. NASA is considering returning both robotic and human missions to the lunar surface and, because of the long lunar nights (14 earth days) isotope power systems are an attractive candidate to generate electrical power. NASA is currently developing the Advanced Stirling Radioisotope Generator (ASRG) as a candidate higher efficiency power system that produces greater than 160 watts with 2 GPHS modules at the beginning of life (BOL) (-30% efficiency). The ASRG uses the same Pu-238 GPHS modules, which are used in RTG, but by coupling them to a Stirling convertor provides a 4-fold reduction in the number of GPHS modules. This study considers the use of Americium 241 (Am-241) as a substitute for the Pu-238 in Stirling convertor based Radioisotope Power Systems (RPS) for power levels from 1 O's of watts to 5 kWe. The Am-241 is used as a replacement for the Pu-238 in GPHS modules. Depending on power level, different Stirling heat input and removal systems are modeled. It was found that substituting Am-241 GPHS modules into the ASRG reduces power output by about 1/5 while maintaining approximately the same system mass. In order to obtain the nominal 160 watts electrical output of the Pu-238 ASRG requires 10 Am-241 GPHS modules. Higher power systems require changing from conductive coupling heat input and removal from the Stirling convertor to either pumped loops or heat pipes. Liquid metal pumped loops are considered as the primary heat transportation on the hot end and water pumped loop/heat pipe radiator is considered for the heat rejection side for power levels above 1 kWe.

Schmitz, Paul C.↗

Preliminary Analysis: Am-241 RHU/TEG Electric Power Source for Nanosatellites

The Februay 2013 Space Works Commercial report indicates a strong increase in nano/microsatellite (1-50 kg) launch demand globally in future years. Nanosatellites (NanoSats) are small spacecraft in the 1-10 kg range, which present a simple, low-cost option for developing quickly-deployable satellites. CubeSats, a special category of NanoSats, are even being considered for interplanetary missions. However, the small dimensions of CubeSats and the limited mass of the NanoSat class in general place limits of capability on their electrical power systems (especially where typical power sources such as solar panels are considered) and stored energy reserves; restricting the power budget and overall functionality. For example, leveraging NanoSat clusters for computationally intensive problems that are solved collectively becomes more challenging with power related restrictions on communication and data-processing. Further, interplanetary missions that would take NanoSats far from the sun, make the use of solar panels less effective as a power source as their required area would become quite large. To overcome these limitations, americium 241 (Am-241) has been suggested as a low power source option. The Idaho National Laboratory, Center for Space Nuclear Research reports that:  (Production) requires small quantities of isotope - 62.5 g of Pu-238; 250 g Am- 241 (for 5 We); Am-241 is available at around 1 kg/yr commercially; Am-241 produces 59 kev gammas which are stopped readily by tungsten so the radiation field is very low. Whereby, an Am-241 source could be placed in among the instruments and the waste heat used to heat the platform; and  amounts of isotope are so low that launch approval may be easier, especially with tungsten encapsulation. As further reported, Am-241 has a half-life that is approximately five times greater than that of Pu- 238 and it has been determined that the neutron yield of a 241-AmO(sub 2) source is approximately an order of magnitude lower than that of a 238-PuO(sub 2) source of equal mass and degree of (sup 16)O enrichment. Also it has been demonstrated that shielded heat sources fuelled by oxygen-enriched 238-PuO(sub 2) have masses that are up to 10 times greater than those fuelled by oxygenenriched 241-AmO(sub 2) with equivalent thermal power outputs and neutron dose rates at 1 m radii. For these reasons, Am-241 is well suited to missions that demand long duration electrical power output, such as deep spaceflight missions and similar missions that use radiation-hard electronics and instrumentation that are less susceptible to neutron radiation damage.

Robertson, Glen A.↗

Performance Assessment of a New Gamma Spectroscopy Toolkit: Peak Doctor and the Spectral Nondestructive Assay Platform (SNAP) - 17397

High resolution gamma spectroscopy (HRGS) is a widely used technique for non-destructive characterization of nuclear materials, and assay of radioactive waste. The ability to accurately evaluate raw spectral data is an important step in producing high-quality results. A spectral analysis application known as Peak Doctor has been developed under a Cooperative Research and Development Agreement (CRADA) between LANL and Pajarito Scientific Corporation (PSC). The software has undergone several years of testing and validation by the Radioactive Waste Characterization Team of the LANL Waste Management Services Group (WM-SVS). PeakDoctor is designed to work with the Spectral Nondestructive Assay Platform (SNAP{sup TM}) to provide a complete HRGS analysis engine. SNAP{sup TM} performs model specific corrections for attenuation and geometry. These applications are the major components of a new toolkit designed primarily for portable far-field non-destructive assay. Peak Doctor performs a very accurate peak fit analysis with the goal of improving upon the capabilities of existing commercially available routines. It is designed to be user-configurable and can be customized to any given type of detector and any given type of spectrum. Peak Doctor was developed to resolve many of the inherent complexities in gamma ray spectra including overlapping peaks, complex multiplet regions, large step peaks, backscatter regions and Compton edges. Unlike most spectral analysis platforms, Peak Doctor does not rely on the energy calibration and detector resolution parameters that are saved by the operator in the spectrum file. Instead, key parameters are pre-defined for individual detectors. The software fits a continuum using a series of splines between a series of junctions or 'knots'. Fitting techniques are used to join the splines together to create a near-continuous multi-element curve. Optimizing the location of the knots is one of Peak Doctor's most powerful features. The knots are initially configured using a pre-defined list and are removed or added automatically during the fitting process. Knots are also created when new features are identified such as Step Peaks and Compton Edges. The user can also manually specify knots in order to improve the continuum fit. The initial fit is improved on by removing any outlier regions (which are converted to peaks) and the continuum fitting process is repeated. At this stage, an experienced user can work interactively with the fit by manually converting peak regions to continuum or vice versa. The final step is to fit a series of peaks above the continuum based on detector parameters. Single and multiplet peaks are then processed by either the tailing (piecewise) or non-tailing (Gaussian) Levenberg Marquardt non-linear peak fitting routine depending on the user defined selection. Only a rough knowledge of the detector characteristics (energy calibration and resolution) is required for the fitting process to proceed with good accuracy. Any peaks that are missed can be manually added by the user and will also be fitted. Discontinuities in the peak / continuum fit can be rapidly identified by reviewing the 'residuals scale' plot. For waste streams that are well characterized, it is often the case that routinely encountered spectra will have similar peaks and continuum features. For these cases a handy 'One Step' method is available that proceeds in an automated fashion without user interaction. Peak Doctor has been configured with a set of default spectrum types including plutonium / transuranic waste (including Weapons Grade Pu, Heat Source Pu, Am-241, Am-243 and Np-237), depleted or enriched uranium, nuclear power plant / intermediate level waste (with fission and activation products such as Cs-137 and Co-60) as well as Naturally Occurring Radioactive Materials (NORM). The user can easily add new spectrum types for specific waste streams or other types of radiological / special nuclear materials. The output from PeakDoctor (including net peak area, centroid energy and a list of minimum detectable activity levels at pre-defined regions) is used by SNAP{sup TM} to complete the data analysis process. Isotope activities are calculated using configurable template-driven geometry and attenuation correction methods. The software accounts for matrix and container wall effects, collimator / angular corrections, detector efficiency (based on calibration curves) and stand-off distance. SNAP{sup TM} uses published algorithms incorporating a series of decision points during its operation to allow the analyst to provide logical input to guide the software during the analysis of each unique spectrum. The Peak Doctor / SNAP{sup TM} toolkit has been extensively tested and validated under a variety of conditions using 'benchmark' reference spectra. Test scenarios reported include HPGe-based measurements of well characterized uranium and plutonium standards and radioactive sources with a variety of different geometries and matrix materials including surrogate drums and boxes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of potential hazards associated with 241 Am loaded resins from nitrate media

LANL has been contacted to provide possible assistance in safe disposition of a number of 241 Am-bearing materials associated with local industrial operations. Among the materials are ion exchange resins which have been in contact with 241 Am and nitric acid, and which might have potential for exothermic reaction. The purpose of this paper is to analyze and define the resin forms and quantities to the extent possible from available data to allow better bounding of the potential reactivity hazard of the resin materials. An additional purpose is to recommend handling procedures to minimize the probability of an uncontrolled exothermic reaction.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Decontamination methods testing for the waste isolation pilot plant - 15691

On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. Several practical, easily deployable methods of decontaminating WIPP salt, using a surrogate contaminant and americium ({sup 241}Am), were developed and tested. The effectiveness of the methods is evaluated qualitatively, and to the extent practical, quantitatively. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and water washing is easy and rapid to deploy. The amount of wastewater produced (2 L/m{sup 2}) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from water washed coupons found no residual removable contamination. Thus, whatever contamination is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.'s Polymeric Barrier System proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

SU-F-P-24: Radiological Disperse Device

Purpose: We are now living in a society of constant fear of terrorism. This topic is pertaining to give a general knowledge of what is a radiological dispersion device or RDD and in case of its detonation, what are the options open to public for a safe action in terms of reducing the exposure and knowing the proper steps. These RDD are also called dirty bombs. Methods: Compared to nuclear weapons, dirty bombs are easy to make. In order for a terrorist organization to construct and detonate a dirty bomb, it must acquire radioactive material by stealing it or buying it through legal or illegal channels. Possible RDD material could come from the millions of radioactive sources used in the industry, for medical purposes and in academic applications mainly for researches. These are, americium-{sup 241}, californium-{sup 252}, caesium-{sup 137}, cobalt-{sup 60}, iridium-{sup 192}, plutonium-{sup 238}, polonium-{sup 210}, radium-{sup 226} and strontium-{sup 90}. Results: Prompt detection of the type of radioactive material used will greatly assist advising people on the protective measures, like sheltering in place, or quickly leaving the immediate area. The effects of radiation are determined by:°the amount of radiation absorbed by body°the type of radiation °the distance from the radiation to an individual°the means of exposure absorbed by the skin, inhaled, or ingested; and length of time exposed. Conclusion: In any facility it is now much more important to keep a log list of all radioactive materials in use. In case there is a dirty bomb explosion, the chaos and economic cost could be enormous. The economic cost for the evaluation of the contamination, survey of people and surroundings and the after treatment, decontamination cost and effort will be a big challenge in any country. So awareness and preparation is the start to face this new type of challenge.

61 RADIATION PROTECTION AND DOSIMETRY↗

Tonopah Test Range Air Monitoring: CY2018 Meteorological, Radiological, and Wind Transported Particulate Observations

In 1963, the U.S. Department of Energy (DOE) (formerly the Atomic Energy Commission [AEC]), implemented Operation Roller Coaster on the Tonopah Test Range (TTR) and an adjacent area of the Nevada Test and Training Range (NTTR) (formerly the Nellis Air Force Range). This operation resulted in radionuclide-contaminated soils at the Double Tracks site and Clean Slate I, II, and III sites. This report documents observations made during ongoing monitoring of radiological, meteorological, and dust conditions at stations installed adjacent to the Clean Slate sites and at the TTR Sandia National Laboratories (SNL) Range Operations Center (ROC). The primary objective of the monitoring effort is to support the DOE Environmental Management Nevada Program (EM NV) in the safe remediation of the environmental legacy of nuclear device development and testing by determining if wind blowing across the Clean Slate sites is transporting particles of radionuclide-contaminated soil beyond the physical and administrative boundaries of the sites, and to provide information for designing long-term monitoring of the sites. The monitoring program in 2018 included five stations. Station 400, located within TTR Area 3 and near the ROC, monitors conditions near the local workforce center. Stations 401 and 403 are located on the northern and southeastern perimeter fence lines, respectively, of the Clean Slate III site. Stations 404 and 405 are along the northern and eastern boundary fence, respectively, at Clean Slate II. The stations are generally downwind of the contaminated areas during either northwesterly or southerly winds, which are the predominant wind directions. The stations—which are similar in design to the Community Environmental Monitoring Program (CEMP) stations operating at locations surrounding the Nevada National Security Site and TTR—include meteorological instruments and continuous-flow, low-volume air samplers. Three pressurized ionization chambers for measuring gamma energy are in use at Stations 400, 401, and 403. Saltation sensors and saltation traps are deployed at the stations adjacent to the Clean Slate sites. Detailed meteorological data are recorded on data loggers, with periodic uploads via a satellite system to the Western Regional Climate Center (WRCC) at Desert Research Institute (DRI), to monitor instrument and site conditions. Air filter samples are collected every two weeks and the deposits in the saltation traps are collected when a sufficient sample mass for analysis has accumulated (generally a six- to eight-month interval). Monitoring at TTR in calendar year (CY) 2018 provided surveillance of both ambient conditions, and ground disturbance associated with remediation activities at Clean Slate II and III. The first half of CY2018 data from Stations 404 and 405 at Clean Slate II were collected during remediation activities at that site, and data collected during the final third of the year at Stations 401 and 403 at Clean Slate III coincided with remediation work at the site. Ambient conditions prevailed at Station 400 throughout the year. Dust transport by suspension and saltation is strongly dependent on wind speed. Concentrations of PM 10 (particulate matter of aerodynamic diameter ≤10 micrometers [μm], a size fraction of small particles that are suspended in the air and can be easily inhaled) remain low until winds exceed approximately 32 kilometers per hour (km/hr) (20 miles per hour [mph]). Saltation particle counts also increase above the same general threshold wind speed. Wind speeds in excess of 32 km/hr (20 mph) occur less than three percent of the time at the sites. Ground disturbing remediation activities were carried out at Clean Slate II from October 31, 2017, until July 26, 2018. Station 404 is located closest to the remediation area and is downwind of the area when winds are from the south. By late November, analysis of dust from the air sampler at Station 404 registered increases in plutonium-239+240 ( 239+240 Pu) concentrations. Peak plutonium concentrations, accompanied by elevated gross alpha concentrations and the detection of americium-241 ( 241 Am) by gamma spectroscopy, occurred between March and May 2018. During spring and early summer of 2018, elevated radionuclide concentrations were also measured in dust collected from the air sampler at Station 405, which is more than twice the distance from the remediation activity as Station 404. A minor but sustained increase in gross alpha concentration of suspended dust collected at Stations 400, 401, and 403 was also observed starting in May 2018 and continuing into August. Stations 401 and 403 are approximately 4.8 to 6.4 km (three to four miles) from Clean Slate II and Station 400 is approximately 13 km (eight miles) from that site. The radionuclide content of material that saltated into the passive Big Spring Number Eight (BSNE) traps during the Clean Slate II remediation period increased in both the traps facing toward and away from the fenced contamination area at Station 404, with the largest increases in concentration (almost five times the previous 239+240 Pu concentration) observed in the traps facing the fenced area. The radionuclide concentration in material in the trap facing away from the fence more than doubled from that measured in 2017. A similar increase was observed for the saltation trap facing away from the fenced area at Station 405, although no increase was observed for material captured in the trap facing the fenced area. Particles ranging in size from 50 to 500 µm generally move by saltation processes and are not expected to significantly contribute to dose for someone breathing in the standard respiratory zone for adult humans (1.5 to 1.7 m [4.9 to 5.6 ft] above ground level) because they are too large to be entrained in the air as suspended dust. Remediation activities at Clean Slate III began on September 5, 2018, and continued until a brief hiatus started on December 20, 2018. Although Station 401 is much closer to the remediation area than Station 403, the first increase in gross alpha and 239+240 Pu concentrations was noted in dust from filters collected at Station 403 in early November 2018. The highest radionuclide concentrations of 2018 at the Clean Slate III stations were recorded in air samples taken from Stations 401 and 403 in late December 2018 and these concentrations were comparable between the two station locations. Gamma rate data measured by the PICs was slightly higher than the annual average at Stations 401 and 403 after remediation activities began, but no single measurement was outside of the range observed in 2018 prior to remediation activities. No increase in radionuclide concentration was observed for the Clean Slate III remediation time period on air filters collected from the other stations. Radiological monitoring at Station 400—which is located at the ROC, and therefore the closest regularly staffed work location to the Clean Slate sites—recorded a higher mean gross alpha activity from dust collected on air sampler filters than in previous years. The individual biweekly air filters yielded measurements of gross alpha that were at or below the mean plus one standard deviation measured in CY2017, except for the period from April 25, 2018, through September 12, 2018. Nonetheless, gamma spectroscopy of all air filters from Station 400 identified only naturally occurring radionuclides, with no detection of americium-241 ( 241 Am) (which is used as an indicator of plutonium-241[ 241 Pu]). Alpha spectroscopy of selected air filters collected at Station 400 did not detect plutonium-238 ( 238 Pu) or 239+240 Pu above the analytical detection limit. Gross beta and gamma exposure rate data were comparable to data from regional CEMP stations. No saltation traps are deployed at Station 400. Considering the ambient monitoring periods during CY2018, the monitoring stations adjacent to the Clean Slate sites recorded above-background 238 Pu and 239+240 Pu concentrations (SNL, 1998; Turner et al., 2003) in samples collected by passive saltation traps and on selected air filters. These observations were consistent with previous years and indicated that radionuclide movement occurred under ambient conditions by processes of both saltation and suspension. The presence of radionuclides in all saltation traps, both upwind and downwind of the fenced contamination areas, was consistent with previous radiological survey results (EG&G, 1979) that show lower concentration radiologically contaminated material dispersed outside the fenced area. Although 239+240 Pu concentrations in the air filter samples were sometimes orders of magnitude higher than background, the concentrations were below risk-based action levels defined for an industrial area worker scenario (U.S. DOE, 2014). The interception of airborne dust with above-background radionuclide concentrations demonstrates that dispersal processes are continuing. Strong winds are the driver for both suspension and saltation of dust and soil. The recorded wind speed and PM10 during remediation activities at Clean Slate II and III identified discrete wind events lasting approximately five to eight hours each, which accounted for the greatest PM 10 concentrations at the monitoring stations. These events captured the effect of winds from the northwest and from the south. The observed differences in dust loading and radionuclide concentrations for the monitoring stations when they were either upwind or downwind of the contamination area identify the remediation area as a source of plutonium contaminated dust. The increase in concentration of radionuclides in airborne dust during the remediation activities, compared with measurements under previous ambient conditions (mean 239+240 Pu concentrations during remediation activities are approximately nine to 40 times higher compared to pre-remediation concentrations, depending on monitoring station), highlights the impact of land disturbance on wind-borne contaminant transport from the Clean Slate test areas. The meteorological and particulate matter monitoring indicated that conditions for wind-borne contaminant movement exist at the Clean Slate sites, and that transport of radionuclide-contaminated soil by both suspension and saltation occurred under ambient conditions and accelerated when the sites were disturbed. The 239+240 Pu alpha spectroscopy results indicated that airborne concentrations during remediation activities ranged from 2 × 10-16 to 380 × 10-16 µCi/ml. These results can be compared to the derived air concentration limit for 239 Pu of 6 × 10-11 µCi/ml for occupational workers contained in Code of Federal Regulations 10 CFR 835, Occupational Radiation Protection. Even though the air sampling results document an increase in concentration during remediation activities, the observed concentrations are less than 0.06 percent of the occupational limit. The Corrective Action Unit (CAU) closure strategy uses a risk-based approach, whereby acceptable contaminant concentrations are determined as a function of anticipated human exposure.

54 ENVIRONMENTAL SCIENCES↗

Recycle of Am-241 obtained from long term stored plutonium for use in radioisotope power systems - 15035

Pu-238 has been used as a power source for spacecraft since the early days of space exploration. It has proven to be an effective source of power where the use of solar generated power is impractical. Its application for use in spacecraft power systems was developed by the USA and Russia who have spent many $ 100s of millions on its development and production. However, the original facilities for the production of Pu-238 are no longer in use and existing worldwide stockpiles have reduced to levels sufficient for only one or two further missions. Production of Pu-238 requires considerable facilities including a nuclear reactor and reprocessing plants that are very expensive to build and operate. The European Space Agency (ESA) has assessed the options for post-launch power generation in future European space missions and has made the decision to pursue the use of Am-241 as an alternative isotope to power future European Radioisotope Power Systems (RPS). What makes Am-241 attractive for the European market is its 100% isotopically pure production from the decay of Pu-241 in separated civil plutonium stockpiles. Historically, Europe has relied on collaborations with the USA or Russia to access these nuclear power sources. During 2009, the European Space Agency (ESA) funded a project to examine the cost and practicality of establishing a European source of material suitable for Radioisotope Power Systems (RPS). The study concluded that a small number of radioisotopes could be used in RPSs. On further review two leading candidates, Am-241 and Pu-238, were selected for further study. Whilst Pu-238 was recognized as the isotope of choice for use in RPSs, the costs of setting up a European based production facility was prohibitively expensive. Although Am-241 has a number of limitations compared to Pu-238, the much reduced production costs make Am-241 an attractive option for European based production of RPSs. Despite its lower power density of ∼ 0.11 Wth/g, Am-241 is viewed as a potential alternative to Pu-238 because of its availability within the nuclear fuel cycle. During the reprocessing of spent nuclear fuel plutonium is separated as a chemically pure PuO{sub 2} product, containing around 3-11% w/w of the Pu-241 isotope, and placed into storage. During storage the Pu-241 isotope beta decays to Am-241 with a half-life of 14.4 years. {sub 94}{sup 241}Pu → (β 14.4 y) {sub 95}{sup 241}Am. This process has resulted in many 100 kgs of americium growing into the civil plutonium stockpile in the UK; material that is of value in generating heat sources and is able to be recycled. The key benefit of Am-241 as a potential European RPS source isotope is the availability of feed material along with the facilities to process it. It is expected that aged civil PuO{sub 2} will contain 25-50 g Am-241 per kg PuO{sub 2} depending on the source. A conceptual flowsheet using facilities for separating americium from plutonium is expected to produce on average 8.4 kg Am/year. ESA are now following their published road-map of studies to develop flight ready RTGs and RHUs. This has included the development and testing of the flowsheet to be used for production of Am-241. The project has focused on underpinning the conceptual flowsheet, using aged plutonium in NNL's PuMA laboratory. This has underpinned the design of a plant to produce a steady supply of Am-241. The scope of the overall project is 'store to store' - the starting point is aged plutonium currently housed in existing Sellafield stores and the end point is the storage of separated plutonium and americium powders on the Sellafield site. The full scale process is planned to be housed within the NNL's Central Laboratory in an existing facility designed for plutonium active operations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Design and Measurement of a Low-Noise 64-Channels Front-End Readout ASIC for CdZnTe Detectors

Cadmium zinc telluride (CdZnTe) detectors, as one of the principal detectors for the next-generation X-ray and γ-ray imagers, have high energy resolution and supporting electrode patterning in the radiation environment at room-temperature. In the present, a number of internationally renowned research institutions and universities are actively using these detector systems to carry out researches of energy spectrum analysis, medical imaging, materials characterization, high-energy physics, nuclear plant monitoring, and astrophysics. As the most important part of the readout system for the CdZnTe detector, the front-end readout application specific integrated circuit (ASIC) would have an important impact on the performances of the whole detector system. In order to ensure the small signal to noise ratio (SNR) and sufficient range of the output signal, it is necessary to design a front-end readout ASIC with very low noise and very high dynamic range. In addition, radiation hardness should be considered when the detectors are utilized in the space applications and high energy physics experiments. In this paper, we present measurements and performances of a novel multi-channel radiation-hardness low-noise front-end readout ASIC for CdZnTe detectors. The readout circuits in each channel consist of charge sensitive amplifier, leakage current compensation circuit (LCC), CR-RC shaper, S-K filter, inverse proportional amplifier, peak detect and hold circuit (PDH), discriminator and trigger logic, time sequence control circuit and driving buffer. All of 64 readout channels' outputs enter corresponding inputs of a 64 channel multiplexer. The output of the mux goes directly out of the chip via the output buffer. The 64-channel readout ASIC is implemented using the TSMC 0.35 μm mixed-signal CMOS technology. The die size of the prototype chip is 2.7 mm x 8 mm. At room temperature, the equivalent noise level of a typical channel reaches 66 e{sup -} (rms) at zero farad for a power consumption of 8 mW per channel. The linearity error is lower than 1% and the overall gain of the readout channel is 165 V/pC. The crosstalk between the channels is less than 3%. By connecting the readout ASIC to a CdZnTe detector, we obtained a γ-ray spectrum, the energy resolution is 5.1% at the 59.5-keV line of {sup 241}Am source. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Q-Spectroscopy

A new tool for rapid single-particle analysis in safeguards is described. This quantitative analysis tool has advantages over conventional methods for isotopic analysis of actinide-bearing samples. The method can be used to determine chemical age via 241 Am/ 241 Pu ''atomic clock''.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Study of some health physics parameters of bismuth-ground granulated blast furnace slag shielding concretes

The Bismuth-ground granulated blastfurnace slang (Bi-GGBFS) concrete samples were prepared. The weight percentage of different elements present in Bi-GGBFS Shielding concretewas evaluated by Energy Dispersive X-ray Microanalysis (EDX). The exposure rate and absorbed dose rate characteristics were calculated theoretically for radioactive sources namely {sup 241}Am and {sup 137}Cs. Our calculations reveal that the Bi-GGBFS concretes are effective in shielding material for gamma radiations.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Reactor pond walls deposit investigation during dismantling MR reactor - 15028

Specialists of the NRC 'Kurchatov Institute' now carry out the dismantling of the multiloop research MR reactor. During this works the water level was decreased in the MR reactor pond on 4 meter. Dependence of exposure dose rate above the pond from deposits on the pond walls was confirmed. Detailed research of deposits on MR reactor pond walls was made, using α-, β-, γ- spectrometric methods. Correlation coefficients between concentration Pu and U radionuclides and concentration {sup 137}Cs and {sup 241}Am was calculated. The surface activity of the main dose-forming radionuclide {sup 137}Cs on the walls of the pond was measured. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Radiochemical Methodology Applied To Determination of gamma-Emitting Radionuclides In Waste Samples From IEA-R1 Reactor - 17265

The Brazilian reactor IEA-R1 has been in operation since 1957. It is an open pool-type nuclear research reactor, operating at 2 to 5 MW that uses water as coolant, moderator and biological shield. Polypropylene filter cartridges are part of the water treatment system of the reactor and become radioactive as they remove suspended solid material from the cooling water during the operation of the reactor. Annually, the replacement of the filters from the reactor's water treatment system generates about 36 units. The development of a method for the characterization of the radionuclides present in the filters is required for routine operation in the Radioactive Waste Management Facility (RWMF) in order to determine the radioisotopic inventory of the waste packages. Wastes from nuclear reactors contain transuranic elements, activation and fission products, only few of them emitting gamma radiation measurable by simple gamma scanning methods. The method of the Scaling Factors (SF) uses empirically determined proportions between the concentrations of Difficult to Measure (DTM) radionuclides and that of easily measurable gamma emitters called Key Nuclide (KN), to calculate DTM concentrations. The objective of this study is to describe a method of the dissolution of the filter cartridges for determination of gamma emitting radionuclides. To determine the activity of a sample by means of of gamma spectrometry, it is necessary previous knowledge of the efficiency curve for a given geometry, which can be obtained by calibrating the efficiency and energy of the detector. The need to dissolve the filter cartridges has arisen given the difficulty in obtaining a standard source with the same filter geometry. Therefore, for this work, the calibration curve was constructed using a mixed gamma-ray standard with known activities, same geometry, density, physical state and chemical composition of the source to be analyzed. A sample with background count (BG) was prepared and analyzed under the same conditions as the samples and the result was subtracted from the results obtained for each filter sample. After many tests, it was possible to establish a method for the dissolution of filter cartridges for the determination of gamma-emitting radionuclides. A mixed gamma-ray standard was produced in liquid form, containing the following radionuclides: {sup 241}Am, {sup 137}Cs and {sup 60}Co, which have photo-peaks at the following energies: 59.54, 661.66 and 1332.49 keV, respectively. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗