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Fukushima Daiichi reactor source term attribution using cesium isotope ratios from contaminated environmental samples
Source term attribution of environmental contamination following the Fukushima Daiichi Nuclear Power Plant (FDNPP) disaster is complicated by a large number of possible similar emission source terms (e.g. FDNPP reactor cores 1–3 and spent fuel ponds 1–4). Cesium isotopic analyses can be utilized to discriminate between environmental contamination from different FDNPP source terms and, if samples are sufficiently temporally resolved, potentially provide insights into the extent of reactor core damage at a given time. In this work, rice, soil, mushroom, and soybean samples taken 100–250 km from the FDNPP site were dissolved using microwave digestion. Radiocesium was extracted and purified using two sequential ammonium molybdophosphate-polyacrylonitrile columns, following which 135 Cs/ 137 Cs isotope ratios were measured using thermal ionization mass spectrometry (TIMS). Results were compared with data reported previously from locations to the northwest of FDNPP and 30 km to the south of FDNPP. 135 Cs/ 137 Cs isotope ratios from samples 100–250 km to the southwest of the FDNPP site show a consistent value of 0.376 ± 0.008. 135 Cs/ 137 Cs versus 134 Cs/ 137 Cs correlation plots suggest that radiocesium to the southwest is derived from a mixture of FDNPP reactor cores 1, 2, and 3. Conclusions from the cesium isotopic data are in agreement with those derived independently based upon the event chronology combined with meteorological conditions at the time of the disaster. In conclusion, cesium isotopic analyses provide a powerful tool for source term discrimination of environmental radiocesium contamination at the FDNPP site. For higher precision source term attribution and forensic determination of the FDNPP core conditions based upon cesium, analyses of a larger number of samples from locations to the north and south of the FDNPP site (particularly time-resolved air filter samples) are needed. Published in 2016. This article is a U.S. Government work and is in the public domain in the USA.
Radioactive Cesium Isotope Ratios as a Tool for determining Dispersal and Re-dispersal Mechanisms Downwind from the Nevada Nuclear Security Site
Fractionation of the two longer lived radioactive cesium isotopes (135 and 137) produced by above ground nuclear tests have been measured and used to clarify the dispersal mechanisms of cesium deposited in the area between the Nevada Nuclear Security Site (NNSS) and Lake Mead in the southwestern United States. Fractionation of these isotopes is due to the 135 decay chain requiring several days to completely decay to Cs and the 137 decay chain less than one hour. Since the Cs precursors are gases, iodine and xenon, the 135Cs plume was deposited farther downwind than the 137Cs plume. Samples were obtained from the Las Vegas arm of Lake Mead, separated into segments, gamma counted to identify layers of activity and analyzed for 135Cs/137Cs ratios. The layers proved to have nearly identical highly fractionated isotope ratios. This information is consistent with a model where the cesium was initially deposited onto the land area draining into Lake Mead and the composite from all the above ground shots subsequently washed onto the Lake by high intensity rain and windstorms producing a layering of Cs activity where each layer is a portion of the composite.
Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel
Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurized water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240 Pu/ 239 Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240 Pu/ 239 Pu and 137 Cs/ 135 Cs ratios, respectively. Lastly, this study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.
Cesium Isotope Ratios as Indicators of Nuclear Power Plant Operations
There are multiple paths by which radioactive cesium can reach the effluent from reactor operations. The radioactive 135Cs/137Cs ratios are controlled by these paths. In an effort to better understand the origin of this radiation, these 135Cs/137Cs ratios in effluents from three power reactor sites have been measured in offsite samples. These ratios are different from global fallout by up to six fold and as such cannot have a significant component from this source. A cesium ratio for a sample collected outside of the plant boundary provides integration over the operating life of the reactor. A sample collected inside the plant at any given time can be much different from this lifetime ratio. The measured cesium ratios vary significantly for the three reactors and indicate that the multiple paths have widely varying levels of contributions. There are too many ways these isotopes can fractionate to be useful for quantitative evaluations of operating parameters in an offsite sample, although it may be possible to obtain limited qualitative information for an onsite sample.
A preliminary assessment on the use of biochar as a soil additive for reducing soil-to-plant uptake of cesium isotopes in radioactively contaminated environments
A series of K d tracer batch experiments were conducted in this paper to assess the absorptive-desorption properties of Biochar as a potential agent to selectively sequester labile soil Cs or otherwise help reduce the uptake of Cs isotopes into plants. A parallel experiment was conducted for strontium. Fine-grained fractionated Woodlands tree Biochar was found to have a relatively high affinity for Cs ions (K d > 100) relative to coral soil (K d < 10) collected from the Marshall Islands. The Biochar material also contains an abundance of K (and Mg). Finally, these findings support a hypothesis that the addition of Biochar as a soil amendment may provide a simple yet effective method for reducing soil-to-plant transfer of Cs isotopes in contaminated environments.
ASSESSMENT OF THE RADIONUCLIDE COMPOSITION OF "HOT PARTICLES" SAMPLED IN THE CHERNOBYL NUCLEAR POWER PLANT FOURTH REACTOR UNIT
Fuel-containing materials sampled from within the Chernobyl Nuclear Power Plant (ChNPP) 4th Reactor Unit Confinement Shelter were spectroscopically studied for gamma and alpha content. Isotopic ratios for cesium, europium, plutonium, americium, and curium were identified and the fuel burnup in these samples was determined. A systematic deviation in the burnup values based on the cesium isotopes, in comparison with other radionuclides, was observed. The conducted studies were the first ever performed to demonstrate the presence of significant quantities of {sup 242}Cm and {sup 243}Cm. It was determined that there was a systematic underestimation of activities of transuranic radionuclides in fuel samples from inside of the ChNPP Confinement Shelter, starting from {sup 241}Am (and going higher), in comparison with the theoretical calculations.
Biokinetic models for radiocesium and its progeny
Over the next few years the International Commission on Radiological Protection (ICRP) will publish a series of reports containing updated biokinetic and dosimetric models and dose coefficients for occupational intake of radionuclides. The biokinetic modeling scheme continues a trend in modern ICRP reports toward physiologically realistic descriptions of the time-dependent behavior of absorbed radionuclides and ingrowing chain members. This paper proposes systemic biokinetic models for cesium isotopes and their chain members for use in these ICRP reports and examines dosimetric implications of the proposed models. Comparisons of A = tissue dose per unit input to blood based on current ICRP models for workers (ICRP Publication 68, 1994) with B = corresponding values based on the proposed biokinetic models (but using dosimetry models of Publication 68) yields the following ranges of ratios B:A for tissues addressed in current ICRP documents: 0.5-25 for 130Cs (T1/2 = 29.2 min), 0.6-9.5 for 134mCs (2.9 h), 0.8-2.2 for 129Cs (32.1 h), 0.7-1.7 for 131Cs (9.69 d), 0.8-1.3 for 136Cs (13.2 d), 0.7-1.1 for 134Cs (2.06 y), 0.5-1.9 for 137Cs (30.2 y), and 0.2-3.7 for 135Cs (2.3x106 y). The large differences in estimated tissue dose for some tissues and cesium isotopes, particularly short-lived isotopes, result mainly from differences in model predictions of the time-dependent distributions of cesium in the body. For example, the proposed and current ICRP models for cesium predict peak kidney contents of ~22% and ~0.4%, respectively, following intravenous injection of stable cesium. Based on the proposed models for cesium and its progeny, the only dosimetrically significant chain members of cesium isotopes are 137mBa, which represents 32-85% of the estimated tissue doses from injected 137Cs, and 134Cs, which represents 4-53% of the estimated tissue doses from injected 134mCs.
Characterization Techniques for TRU Waste from the Analytical Laboratory Hot Cells at Idaho National Laboratory - 19610
Since the first reactor went hot in 1942, there has been a never ending desire to improve the nuclear reactor and the fuel that it burns. The Idaho National Laboratory has been a key player in reactor and fuel development. As more research projects begin, the spectrum of waste parameters grows. Over the years, the Analytical laboratory (AL) at the INL Materials and Fuels Complex (MFC) has supported a large range of various research projects. Each project brings its own challenge and analytical technique that, in turn, yields its own unique waste parameters. The AL Hot Cells at INL consist of 6 small cells that can support around 12 different fuel research projects at a time along with infrastructure and environmental samples. Each different project yields different levels of transuranic (TRU) material, RCRA constituents, dose, and contamination levels. The waste parameters from each project dictate a path forward for the waste as Low-Level Waste (LLW), Contact Handled TRU (CH-TRU), Mixed Contact Handled TRU (M-CH-TRU), Remote Handled TRU (RH-TRU), or Mixed Remote Handled TRU (M-RHTRU). With visual examination techniques and process knowledge as an aid, there have been waste handling methods developed that allow debris and sample material to be segregated. Unique characterization mechanisms have been applied for each debris and sample material that are generally more accurate than smear and surface contamination data. For debris found in the Hot Cells, analytical data is gathered and conservatively summarized for each fuel type and research method. This consists of data from Quadrupole Inductively Coupled Plasma Mass Spectrometry (Q-ICP-MS), Gamma Spectroscopy, and some data verification through Thermal Ionization Mass Spectrometry (TIMS) and Multi-Collector Inductively Coupled Plasma Mass Spectrometry (MC-ICP-MS). This data gives a range of TRU isotope concentrations. Fuel age is taken into account to determine the dominant TRU activity isotopes. Cesium concentrations are also gathered. Each parameter/fuel type then ties a measured Cesium number to TRU content based on historical analytical data. Debris in the hot cells is segregated by project and packaged in individual 2 gallon metal paint cans. Each can is counted by gamma spectroscopy in a calculated and measured geometry. Using process knowledge to determine the fuel or project type, newly counted cesium numbers can be applied to the historic analytical data and TRU content can be estimated. The TRU content allows waste debris to be segregated between CH-TRU and RH-TRU. Packaging techniques can allow for the minimization of RHTRU waste. Sample material received into the AL must be dissolved into a liquid matrix in order to be analyzed in most cases. When analysis is complete, these liquids go through a consolidation, complexation, neutralization, and solidification process where real time analytical data can be used to calculate TRU content and RCRA constitutes. These calculations allow for no further analytical data to be gathered. The waste matrix is considered stable and acceptable by all Acceptable Knowledge (AK) documents. The unique techniques used in characterizing and handling the Analytical Laboratories Hot Cell waste can be applied to various situations and various laboratories. They allow for more accurate data for waste constituents and also for the minimization of RH-TRU waste. (authors)
Evaluation of observed and modelled aerosol lifetimes using radioactive tracers of opportunity and an ensemble of 19 global models
Aerosols have important impacts on air quality and climate, but the processes affecting their removal from the atmosphere are not fully understood and are poorly constrained by observations. This makes modelled aerosol lifetimes uncertain. In this study, we make use of an observational constraint on aerosol lifetimes provided by radionuclide measurements and investigate the causes of differences within a set of global models. During the Fukushima Dai-Ichi nuclear power plant accident of March 2011, the radioactive isotopes cesium-137 ( 137 Cs) and xenon-133 ( 133 Xe) were released in large quantities. Cesium attached to particles in the ambient air, approximately according to their available aerosol surface area. 137 Cs size distribution measurements taken close to the power plant suggested that accumulation-mode (AM) sulfate aerosols were the main carriers of cesium. Hence, 137 Cs can be used as a proxy tracer for the AM sulfate aerosol's fate in the atmosphere. In contrast, the noble gas 133 Xe behaves almost like a passive transport tracer. Global surface measurements of the two radioactive isotopes taken over several months after the release allow the derivation of a lifetime of the carrier aerosol. We compare this to the lifetimes simulated by 19 different atmospheric transport models initialized with identical emissions of 137 Cs that were assigned to an aerosol tracer with each model's default properties of AM sulfate, and 133 Xe emissions that were assigned to a passive tracer. We investigate to what extent the modelled sulfate tracer can reproduce the measurements, especially with respect to the observed loss of aerosol mass with time. Modelled 137 Cs and 133 Xe concentrations sampled at the same location and times as station measurements allow a direct comparison between measured and modelled aerosol lifetime. The e-folding lifetime τ e , calculated from station measurement data taken between 2 and 9 weeks after the start of the emissions, is 14.3 days (95 % confidence interval 13.1–15.7 days). The equivalent modelled τ e lifetimes have a large spread, varying between 4.8 and 26.7 days with a model median of 9.4 ± 2.3 days, indicating too fast a removal in most models. Because sufficient measurement data were only available from about 2 weeks after the release, the estimated lifetimes apply to aerosols that have undergone long-range transport, i.e. not for freshly emitted aerosol. However, modelled instantaneous lifetimes show that the initial removal in the first 2 weeks was quicker (lifetimes between 1 and 5 days) due to the emissions occurring at low altitudes and co-location of the fresh plume with strong precipitation. Deviations between measured and modelled aerosol lifetimes are largest for the northernmost stations and at later time periods, suggesting that models do not transport enough of the aerosol towards the Arctic. The models underestimate passive tracer ( 133 Xe) concentrations in the Arctic as well but to a smaller extent than for the aerosol ( 137 Cs) tracer. As a result, this indicates that in addition to too fast an aerosol removal in the models, errors in simulated atmospheric transport towards the Arctic in most models also contribute to the underestimation of the Arctic aerosol concentrations.
Cs-135 Content of Cesium Capsules and Strontium/Cesium Heat Sources - 18509
The United States Department of Energy (DOE) is responsible for disposing of 1,335 cesium capsules and 34 strontium and cesium heat sources that are currently stored at the Hanford Site. The documented radionuclide inventories for these wastes report quantities of Sr-90 and Cs-137 but do not report quantities of other strontium and cesium isotopes, some of which need to be estimated for the purpose of conducting a postclosure performance assessment for the eventual disposal facility. The purpose of this paper is to present a method for estimating the quantity of Cs-135 present in the cesium capsules and in the strontium/cesium heat sources. (authors)
Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation
The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.
Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation - 18594
The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry. (authors)
Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry
Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.
Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry
Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2 ) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.
Qualification of Processing Batches for Tank Closure Cesium Removal Through In-Tank Batch Contact Testing - 20470
The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)
Post-Closure Environmental Sampling and Inspection at the Amchitka Island, Alaska, Site - 17251
In the 1960's and early 1970's, three underground nuclear tests were conducted on Amchitka Island, Alaska. In 1965, the US Department of Defense-in conjunction with the US Atomic Energy Commission (AEC)-conducted the first of these nuclear tests, Long Shot, to provide data that would improve US capabilities for detecting underground nuclear explosions. In 1969, AEC conducted the second test, Milrow, to explore the feasibility of detonating a much larger device. In 1971, AEC conducted the third test, Cannikin, which was the largest underground test ever conducted by the US. The Long-Term Surveillance Plan (LTSP) for the Amchitka, Alaska, Site details how the US Department of Energy Office of Legacy Management (LM) intends to fulfill its mission to protect human health and the environment at and around the Amchitka Island site. The LTSP calls for monitoring to be performed every 5 years, at least during the initial phase of the project. The purpose of the monitoring is to develop a baseline of activity concentrations for selected radionuclides in biota, water, and soil, both on Amchitka Island and at the reference location on Adak Island, approximately 322 kilometers (200 miles) northeast of Amchitka. In addition to the environmental monitoring activities, seven mud pit caps that contain diesel-contaminated soils are inspected. The plan for biological and water sampling (both seawater and freshwater) during the May 2016 sampling event was developed through close coordination with the primary stakeholders, including the Alaska Department of Environmental Conservation, the Aleutian Pribilof Islands Association, and the US Fish and Wildlife Service. The environmental sampling in 2016 is a continuation of LM's evaluation of Amchitka Island's environment. Previous results from similar monitoring programs in 2004 and 2011 indicate that the environment is not being impacted by radionuclide migration and uptake from the underground nuclear testing conducted beneath Amchitka Island; consequently, subsistence and commercial-catch seafood is safe for human consumption. The collection of biota samples (rockfish, Irish lords, greenling, and rockweed) for cesium-134 ({sup 134}Cs), cesium-137 ({sup 137}Cs), plutonium-239 ({sup 239}Pu), plutonium-240 ({sup 240}Pu), and freshwater and seawater samples for {sup 134}Cs, {sup 137}Cs, and tritium ({sup 3}H) was the primary objective of the 2016 sampling effort. Rockfish, Irish lords, and greenling are resident, nonmigratory fish species that spend their entire life cycle in the marine environment, and rockweed is an inter-tidal species that may concentrate cesium. The seawater samples were collected primarily to determine if leakage from the Long Shot detonation could be detected and, secondarily, to compare with the background seawater samples collected from Adak Island. Cesium isotopes are also being analyzed to better understand the potential impact that the Fukushima-Daiichi release might have on the Amchitka and Adak environments. The target detection limits used for the 2016 sampling are lower than those used in previous studies. The lower detection limits make it possible for LM to delineate background activity levels for the radionuclides of interest. Data from the 2016 sampling event go beyond assessing food safety and focus on detecting leakage from the underground nuclear explosions. Additionally, the baseline activity concentrations obtained for the reference area (Adak Island) will enable identification of quantitative statistical trends in the results of future sampling programs. Sampling programs conducted in 2021 and beyond will have an objective to assess the statistical trends of radioactivity levels in the Amchitka Island environment. (authors)
Emerging Technologies and Techniques for Wide Area Radiological Survey and Remediation
Technologies to survey and decontaminate wide-area contamination and process the subsequent radioactive waste have been developed and implemented following the Chernobyl nuclear power plant release and the breach of a radiological source resulting in contamination in Goiania, Brazil. These civilian examples of radioactive material releases provided some of the first examples of urban radiological remediation. Many emerging technologies have recently been developed and demonstrated in Japan following the release of radioactive cesium isotopes (Cs-134 and Cs-137) from the Fukushima Dai-ichi nuclear power plant in 2011. Information on technologies reported by several Japanese government agencies, such as the Japan Atomic Energy Agency (JAEA), the Ministry of the Environment (MOE) and the National Institute for Environmental Science (NIES), together with academic institutions and industry are summarized and compared to recently developed, deployed and available technologies in the United States. The technologies and techniques presented in this report may be deployed in response to a wide area contamination event in the United States. In some cases, additional research and testing is needed to adequately validate the technology effectiveness over wide areas. Survey techniques can be deployed on the ground or from the air, allowing a range of coverage rates and sensitivities. Survey technologies also include those useful in measuring decontamination progress and mapping contamination. Decontamination technologies and techniques range from non-destructive (e.g., high pressure washing) and minimally destructive (plowing), to fully destructive (surface removal or demolition). Waste minimization techniques can greatly impact the long-term environmental consequences and cost following remediation efforts. Recommendations on technical improvements to address technology gaps are presented together with observations on remediation in Japan.