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New reference materials for trace-levels of actinide elements in plutonium
In this report two plutonium oxides were prepared as unique reference materials for measurement of actinide elements present as trace constituents. Each reference material unit is approximately 200 mg of PuO 2 powder in a quartz glass bottle. Characterized attributes of the oxides included mass fractions of plutonium, americium, neptunium, and uranium. Isotope-amount ratios were also determined for plutonium and uranium, but neptunium and americium were observed to be monoisotopic 237 Np and 241 Am. Measurements for characterization and verification of the attributes show that plutonium and trace actinides are homogeneous with the exception of limited heterogeneity for uranium, primarily observed for the 238 U isotope. Model purification ages calculated from measured americium and uranium attribute values are consistent with material histories and indicate that these impurities are predominantly due to the decay of plutonium isotopes.
A multi-collector ICP-MS method for quantification of plutonium, uranium, and americium in hair and nails of occupationally or medically exposed individuals
The 239 Pu, 238 U, and 241 Am concentrations and 239 Pu/ 240 Pu, 235 U/ 238 U, and 236 U/ 238 U atom ratios were measured in the hair and nail samples from individuals who donated their bodies to the United States Transuranium and Uranium Registries. Further, the concentration of 239 Pu ranged from 0.22–15.8 ng/kg. The 240 Pu/ 239 Pu isotopic ratios ranged from 0.026 to 0.127 which is consistent with weapons-grade plutonium. Concentration of uranium fell between 1.84 µg/kg and 29.5 µg/kg and 235 U/ 238 U ratios ranged from 4.8×10 -3 to 7.6×10 -3 . Elevated 236 U/ 238 U atom ratios were measured in two cases and ranged from 5.0×10 -6 – 2.4×10 -5 indicating exposure to spent or reprocessed uranium material. The concentration of 241 Am was measured in four hair samples and ranged from 0.02–0.21 ng/kg.
Rapid isotopic analysis of uranium, plutonium, and americium in post-detonation debris simulants by RIMS
Uranium interference is a major problem in conventional 238 Pu analysis. In RIMS, a slight detuning of the resonance laser extinguishes the Pu signal so that the 238 U interference can be accurately quantified and subtracted.
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.
Analysis of a Bulk 237 Np Oxide Sample for Trace Actinides
A neptunium oxide sample was dissolved and trace actinide constituents were measured by isotope dilution ICP-MS (for 239–244 Pu, 233–238 U, 237 Np, 241–243 Am) and alpha spectrometry ( 238 Pu, 242 Cm, 244 Cm). This material contained significant quantities of transuranic elements, notably americium and curium. The isotopic composition of the material will continue to evolve over time due to radioactive decay; the measurements reported here include all isotopes decay corrected to a reference date of 1/1/2022. These factors may be considerations in criticality studies of the related 237 Np sphere.
Radiotoxicity Characterization of Multi-Recycled Thorium Fuel - 12394
As described in companion papers, Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the transuranic (TRU) contained in the used nuclear fuel. The potential of thorium as a TRU burner is described in another paper presented at this conference. This paper analyzes the long-term impact of thorium on the front-end and backend of the fuel cycle. This is accomplished by an assessment of the isotopic make-up of Th in a closed cycle and its impact on representative metrics, such as radiotoxicity, decay heat and gamma heat. The behavior in both thermal and fast neutron energy ranges has been investigated. Irradiation in a Th fuel PWR has been assumed as representative of the thermal range, while a Th fuel fast reactor (FR) has been employed to characterize the behavior in the high-energy range. A comparison with a U-fuel closed-cycle FR has been undertaken in an attempt of a more comprehensive evaluation of each cycle's long-term potential. As the Th fuel undergoes multiple cycles of irradiation, the isotopic composition of the recycled fuel changes. Minor Th isotopes are produced; U-232 and Pa-231 build up; the U vector gradually shifts towards increasing amounts of U-234, U-235 etc., eventually leading to the production of non negligible amounts of TRU isotopes, especially Pu-238. The impact of the recycled fuel isotopic makeup on the in-core behavior is mild, and for some aspects beneficial, i.e. the reactivity swing during irradiation is reduced as the fertile characteristics of the fuel increase. On the other hand, the front and the back-end of the fuel cycle are negatively affected due to the presence of Th-228 and U-232 and the build-up of higher actinides (Pu-238 etc.). The presence of U-232 can also be seen as advantageous as it represents an obstacle to potential proliferators. Notwithstanding the increase in the short-term radiotoxicity and decay heat in the multi-recycled fuel, the Th closed cycle has some potentially substantial advantages compared to the U cycle, such as the smaller actinide radiotoxicity and decay heat for up to 25,000 years after irradiation. In order for these benefits to materialize, the capability to reprocess and remotely manufacture industrial amounts of recycled fuel appears to be the key. Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the TRU contained in the current UNF. The general approach and the potential of thorium as TRU burner is described in other papers presented at this conference. The focus of this paper is to analyze the long-term potential of thorium, once the legacy TRU has been exhausted and the thorium reactor system will become self-sufficient. Therefore, a comparison of Th closed cycle, in fast and thermal neutron energy ranges, vs. U closed cycle, in the fast energy range, has been undertaken. The results presented focus on selected backend and front-end metrics: isotopic actinide composition and potential implications on ingested radiotoxicity, decay heat and gamma heat. The evaluation confirms potential substantial improvements in the backend of the fuel cycle by transitioning to a thorium closed cycle. These benefits are the result of a much lower TRU content, in particular Pu-241, Am-241 and Pu-240, characterizing the Th vs. U actinide inventories, and the ensuing process waste to be disposed. On the other hand, the larger gamma activity of Th recycled fuel, consisting predominantly of hard gammas from U-232's decay products, is a significant challenge for fuel handling, transportation and manufacturing but can be claimed as beneficial for the proliferation resistance of the fuel. It is worth remembering that in our perspective the Th closed cycle and the U closed cycle will follow a transmutation phase which will likely take place over several decades and dictate the technologies required. These will likely include remote fuel manufacturing, regardless of the specific system adopted for the transmutation, which could then be inherited for the ensuing closed cycles. Finally, specific data related to the fuel manufacturing and separation technologies and their performance in the prospected industrial scale deployment, are key for further quantification of the potential merits of the options explored. Further studies in this direction should be warranted before making definitive conclusion. (authors)
Sensitivity analysis of Monju using ERANOS with JENDL-4.0
This paper deals with sensitivity analysis using JENDL-4.0 nuclear data applied to the Monju reactor. In 2010 the Japan Atomic Energy Agency - JAEA - released a new set of nuclear data: JENDL-4.0. This new evaluation is expected to contain improved data on actinides and covariance matrices. Covariance matrices are a key point in quantification of uncertainties due to basic nuclear data. For sensitivity analysis, the well-established ERANOS [1] code was chosen because of its integrated modules that allow users to perform a sensitivity analysis of complex reactor geometries. A JENDL-4.0 cross-section library is not available for ERANOS. Therefore a cross-section library had to be made from the original nuclear data set, available as ENDF formatted files. This is achieved by using the following codes: NJOY, CALENDF, MERGE and GECCO in order to create a library for the ECCO cell code (part of ERANOS). In order to make sure of the accuracy of the new ECCO library, two benchmark experiments have been analyzed: the MZA and MZB cores of the MOZART program measured at the ZEBRA facility in the UK. These were chosen due to their similarity to the Monju core. Using the JENDL-4.0 ECCO library we have analyzed the criticality of Monju during the restart in 2010. We have obtained good agreement with the measured criticality. Perturbation calculations have been performed between JENDL-3.3 and JENDL-4.0 based models. The isotopes {sup 239}Pu, {sup 238}U, {sup 241}Am and {sup 241}Pu account for a major part of observed differences. (authors)
Verification of EPA's ''Preliminary Remediation Goals for radionuclides'' (PRG) electronic calculator
The U.S. Environmental Protection Agency (EPA) requested an external, independent verification study of their updated “Preliminary Remediation Goals for Radionuclides” (PRG) electronic calculator. The calculator provides PRGs for radionuclides that are used as a screening tool at Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and Resource Conservation and Recovery Act (RCRA) sites. These risk-based PRGs establish concentration limits under specific exposure scenarios. The purpose of this verification study is to determine that the calculator has no inherit numerical problems with obtaining solutions as well as to ensure that the equations are programmed correctly. There are 167 equations used in the calculator. To verify the calculator, all equations for each of seven receptor types (resident, construction worker, outdoor and indoor worker, recreator, farmer, and composite worker) were hand calculated using the default parameters. The same four radionuclides (Am-241, Co-60, H-3, and Pu-238) were used for each calculation for consistency throughout.
Estimation of multi-group cross section covariances for {sup 235,238}U, {sup 239}Pu, {sup 241}Am, {sup 56}Fe, {sup 23}Na and {sup 27}Al
This paper presents the methodology used to estimate multi-group covariances for some major isotopes used in reactor physics. The starting point of this evaluation is the modelling of the neutron induced reactions based on nuclear reaction models with parameters. These latest are the vectors of uncertainties as they are absorbing uncertainties and correlation arising from the confrontation of nuclear reaction model to microscopic experiment. These uncertainties are then propagated towards multi-group cross sections. As major breakthroughs were then asked by nuclear reactor physicists to assess proper uncertainties to be used in applications, a solution is proposed by the use of integral experiment information at two different stages in the covariance estimation. In this paper, we will explain briefly the treatment of all type of uncertainties, including experimental ones (statistical and systematic) as well as those coming from validation of nuclear data on dedicated integral experiment (nuclear data oriented). We will illustrate the use of this methodology with various isotopes such as {sup 235,238}U, {sup 239}Pu, {sup 241}Am, {sup 56}Fe, {sup 23}Na and {sup 27}Al. (authors)
Proliferation resistant fuel for pebble bed modular reactors
We show that it is possible to denature the Plutonium produced in Pebble Bed Modular Reactors (PBMR) by doping the nuclear fuel with either 3050 ppm of {sup 237}Np or 2100 ppm of Am vector. A correct choice of these isotopes concentration yields denatured Plutonium with isotopic ratio {sup 238}Pu/Pu {>=} 6%, for the entire fuel burnup cycle. The penalty for introducing these isotopes into the nuclear fuel is a subsequent shortening of the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full Power Days (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and 1965 MWd/ton reduction in fuel discharge burnup. (authors)
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)
A fuel for sub-critical fast reactor
Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and {sup 238}U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.
Modeling of Am-241 as an alternative fuel source in a radioisotope thermoelectric generator
Radioisotope thermoelectric generators (RTGs) have been utilized in the USA to power satellites and space exploration equipment for over half a century. In our preceding paper, Dustin and Borrelli (2021), we assessed the performance of 241 Am as an alternative to 238 Pu by modeling the specific decay-heat and gamma spectra of several potential radioisotopes in SCALE-ORIGEN, and comparing these with 238 Pu An analysis of heat output, heat stability, safety, availability and required mass showed that of the isotopes reviewed, 241 Am had the greatest potential for replacing 238 Pu, with additional design considerations. In our current work, we have assessed radiation transport for two configurations of an 241 Am powered Multi-Mission Radioisotope Thermoelectric Generator (MMRTG), due to the potential health effects for operators, fabricators, and others with the potential to be exposed to neutrons and lower energy gamma rays. We have modeled the use of americium as a fuel source replacement for plutonium in current MMRTG designs. The assessment has been performed using MCNP6, in terms of neutron and gamma emissions that might interfere with sensitive equipment and power system production. The radiation transport behavior in terms of dose rate of the 241 Am fueled models is compared with that of a 238 Pu fueled model. Finally, additional work was performed to validate modeling simplifications with regards to thermocouple material configurations.
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.
AMPP Division Rack Cards [Posters/Cards]
The Actinide Material Processing and Power (AMPP) Division supports plutonium 238 science and manufacturing for NASA exploration, Defense Programs, Advanced Recovery and Integrated Extraction System (ARIES) for nuclear non-proliferation, Materials Recovery & Recycle to enable the future of our operations, and americium production to support well logging for the United States.
Application of the JENDL-4.0 nuclear data set for uncertainty analysis of the prototype FBR Monju
This paper deals with uncertainty analysis of the Monju reactor using JENDL-4.0 and the ERANOS code 1. In 2010 the Japan Atomic Energy Agency - JAEA - released the JENDL-4.0 nuclear data set. This new evaluation contains improved values of cross-sections and emphasizes accurate covariance matrices. Also in 2010, JAEA restarted the sodium-cooled fast reactor prototype Monju after about 15 years of shutdown. The long shutdown time resulted in a build-up of {sup 241}Am by natural decay from the initially loaded Pu. As well as improved covariance matrices, JENDL-4.0 is announced to contain improved data for minor actinides 2. The choice of Monju reactor as an application of the new evaluation seems then even more relevant. The uncertainty analysis requires the determination of sensitivity coefficients. The well-established ERANOS code was chosen because of its integrated modules that allow users to perform sensitivity and uncertainty analysis. A JENDL-4.0 cross-sections library is not available for ERANOS. Therefor a cross-sections library had to be made from the original ENDF files for the ECCO cell code (part of ERANOS). For confirmation of the newly made library, calculations of a benchmark core were performed. These calculations used the MZA and MZB benchmarks and showed consistent results with other libraries. Calculations for the Monju reactor were performed using hexagonal 3D geometry and PN transport theory. However, the ERANOS sensitivity modules cannot use the resulting fluxes, as these modules require finite differences based fluxes, obtained from RZ SN-transport or 3D diffusion calculations. The corresponding geometrical models have been made and the results verified with Monju restart experimental data 4. Uncertainty analysis was performed using the RZ model. JENDL-4.0 uncertainty analysis showed a significant reduction of the uncertainty related to the fission cross-section of Pu along with an increase of the uncertainty related to the capture cross-section of {sup 238}U compared with the previous JENDL-3.3 version. Covariance data recently added in JENDL-4.0 for {sup 241}Am appears to have a non-negligible contribution. (authors)
GrayQb Single-Faced Version 2 Open Environment Test - 16344
The GrayQb{sup TM} Single Faced Version 2 (SF2) is a non-destructive examination device developed to generate radiation contour maps showing source locations and relative radiological levels present in an area under examination. This device allows for characterization of radioactively contaminated areas such as hot cells, gloveboxes, small and large rooms, hallways, and waste tanks. The goal of GrayQb{sup TM} (pronounced 'Gray Cube') is to provide a low cost easily deployable device to aid in the process of decontamination and decommissioning (D and D). The GrayQb{sup TM} SF2 device is approximately a six inch cube weighing seven (7) pounds composed of a tungsten shield with integrated collimator, radiation imaging medium (Phosphor Storage Plate), and digital camera. This paper details the design upgrades incorporated into the new version of the GrayQb{sup TM} SF2 device and the characterization testing of this upgraded device. Device results from controlled testing performed at the Savannah River Site (SRS) Health Physics Instrument Calibration Laboratory (HPICL) are presented, as well as results from the open environment field testing performed in the E-Area Low Level Waste Storage Area. Follow on work performed deploying the GrayQb SF2 device at the Hanford Plutonium Reclamation Facility (PRF) to assist in the radiological characterization of the canyon will also be presented. Testing of the redesigned GrayQb{sup TM} SF2 device demonstrated its ability to detect low energy isotopes, e.g. Am-241 and Pu-238, not only indoors in controlled environments with well-defined source locations, but also in open environment field areas with a significant high energy background. Additionally, it was demonstrated that the device could detect higher energy isotopes (Cs-137 and Co-60) in a controlled setting and with additional testing and/or modifications to device and software, would likely be able to detect these isotopes in the open environment field testing. Follow on 2015 work performed at Hanford proved GrayQb to be a useful tool in the radiological characterization of the PRF canyon. Distribution tendencies of the radiological hot spots were identified that provided insights which will assist in more efficient planning of the continued D and D effort. Hot spots were identified which can be targeted during the future D and D effort and used as markers during the NDA assessment to quantify the radiological environment. (authors)