Processing biological samples from simulated radiological terrorist events using Rapid DNA instruments

Two commercially available portable Rapid DNA instruments were evaluated for their ability to process 1 µL and 10 µL saliva samples deposited on metal and plastic surfaces and contaminated with surrogates of cesium (Cs)-137, strontium (Sr)-90 and cobalt (Co)-60; radioactive materials potentially released during a nuclear weapon accident or a radiological dispersal device detonation. A comparable success rate was noted for both Rapid DNA instruments when considering the number of complete and balanced DNA profiles, the number of profiles with a minimum of 10 autosomal STR loci (out of 23 [FlexPlex™ 27] or 21 [GlobalFiler™ Express]), and the possibility to search a national DNA database in Canada and the United States. Cobalt had an adverse impact on the quality of the megaplex short tandem repeat (STR) DNA profiles derived on each instrument for two of the three contamination levels tested in this study, i.e., 0.05 M and 0.1 M as reflected by a reduced number of detected alleles and decreased profile peak heights. Strontium exhibited some adverse effect on the Rapid DNA results when used at the highest contamination level (0.1 M) whereas cesium had none. No new artifacts were observed in the Rapid DNA profiles of samples spiked with the non-radiogenic surrogates. Importantly, in the context of a radiological/nuclear (RN) event, the ANDE™ 6C offers the possibility to dispose of all radioactive materials associated with contaminated samples quickly using a chip on which all steps of the Rapid DNA process are performed whereas the RapidHIT™ ID accumulates radioactive materials for many days before disposal. An individual handling 25 samples in a week (5 per day) on the RapidHIT™ ID at a 30.5 cm distance with a 5 min exposure to the radioactive source estimated at every run would exceed the 0.042 µSv/5 min limit with gamma dose rates for Cs at 0.13 mSv and for Co at 3.8 mSv. Beta dose rates calculated for the surrogate isotopes at the three concentrations tested were also above the recommended radiation exposure limit of 1 mSv/yr (0.042 µSv/5 min). Various potential mechanisms of action behind the interference noted for Sr and Co at high concentrations are presented. These elements may play a role in the steps prior to PCR (at the DNA molecule by binding to bases or to phosphate groups), during PCR (at the DNA polymerase as cofactors for catalytic sites), or even during amplified DNA fragment detection (as fluorescence quenchers).

Guidance Needed for Continued Employment of Long-term Contaminated Individuals

Operational practices in nuclear power utilities often restrict job assignments for workers with internal contamination due to the difficulty of monitoring for new intakes and strict radioactive material controls. However, restriction of job assignments for an extended period (i.e., months to years) may be too conservative. An industry consensus on guidance in these conditions would be helpful for Radiation Safety Officers (RSOs) to allow long term contaminated workers to have less work restrictions.

Measurement of single- and double-escape HPGe efficiency ratios for 60Co

As a by-product of another measurement, ratios of the single-escape (SE) and double-escape (DE) efficiencies relative to the full-energy-peak efficiency (FE) have been measured for two HPGe detectors for ⁶⁰Co. For a 2.5-cm-thick 95 cm³ crystal the results were SE/FE = 0.000 48 ± 0.000 20 and 0.003 25 ± 0.000 24 for 1173 and 1332 keV gamma-rays, respectively, and DE/FE = 0.000 90 ± 0.000 17 and 0.003 41 ± 0.000 11 for 1173 and 1332 keV, respectively. For a 3.0-cm-thick 84 cm³ crystal the results were SE/FE = 0.000 67 ± 0.000 32 and 0.003 79 ± 0.000 27 for 1173 and 1332 keV respectively, and DE/FE = 0.001 05 ± 0.000 28 and 0.004 29 ± 0.000 16 for 1173 and 1332 keV, respectively. These measurements may be of relevance in connection with Monte Carlo calculations of HPGe detector efficiencies, and may also suggest a path towards improved atomic cross-section measurements.

RE-EVALUATION OF THE AIR KERMA PRIMARY STANDARD OF THE SMU IN 60Co AND 137Cs GAMMA RAY BEAMS

The primary standard for air kerma of the Slovak Institute of Metrology (SMU) as a part of national standard of dosimetric quantities of gamma radiation NE028/02, based on cylindrical graphite walled cavity ionization chamber was established in 2002 for the 60Co gamma ray beams. Since then the main role of SMU was to continuously maintain, develop and refurbish the standard with related accessory to ensure international equivalence. Since 2002, a re-evaluation of the active volume of the ionizing chamber, correction factors, the key data, due to adoption of ICRU Report 90(1), and changes of the calibrations methods and corrections of the equipment were made to obtain improved parameters. The range was extended for 137Cs gamma ray beams and key comparisons with the Bureau International des Poids et Mesures (BIPM) were performed. ICRU Report 90(1) implementation led to a decrease of the Gray of -0.84% for 60Co and -0.80% for 137Cs.

DETERMINING THE DOSE RATE DEPENDENCE OF DIFFERENT ACTIVE PERSONAL DOSEMETERS IN STANDARDIZED PULSED AND CONTINUOUS RADIATION FIELDS

In 2016, the EURADOS working group 12 started a project to gather reliable data concerning the behavior of active personal dosemeters (APDs) used by European hospitals in standardized pulsed and hospital fields. This paper concentrates on the tests that deal with the dose rate dependence of APDs in laboratory tests. The dependence on the dose rate was determined for 10 dosemeter types with continuous radiation of a 60Co source and pulsed radiation of the reference radiation facility of Physikalisch-Technische Bundesanstalt. Furthermore, irradiations were done with continuous radiation of a 137Cs source of the Karlsruher Institute of Technology and with continuous X-rays of the two radiation qualites in radiation beams emerging from the X-ray source assembly, RQR5 and RQR8, of Nofer Institute of Occupational Medicine. For the first time, APDs have been systematically investigated under laboratory conditions in a pulsed reference fields according to ISO/TS 18090-1 and the experience gained by these test procedures will be included in the new IEC/TS 63050.

INTERCOMPARISON ON INTERNAL DOSE ASSESSMENT FOR NUCLEAR POWER PLANTS IN KOREA

The intercomparison test is a quality assurance activity performed for internal dose assessment. In Korea, the intercomparison test on internal dose assessment was carried out for nuclear facilities in May 2018. The test involved four nuclear facilities in Korea, and seven exposure scenarios were applied. These scenarios cover the intake of 131I, a uranium mixture, 60Co and tritium under various conditions. This paper only reviews the participant results of three scenarios pertinent to the operation of nuclear power plants and adopts the statistical evaluation method, used in international intercomparison tests, to determine the significance values of the results. Although no outliers were established in the test, improvements in the internal dose assessment procedure were derived. These included the selection of intake time, selection of lung absorption type according to the chemical form and consideration of the contribution of previous intake.

ASSESSMENT OF THE ABSORBED DOSES IN THE ORGANS IN CASE OF RADIATION EMERGENCY WITH THE SEALED GAMMA-SOURCES

The majority of the radiation accidents with early acute clinical effects were associated with the orphan sources used in industrial and medical facilities. These accidents involved members of general public, who were entirely unaware of the exposure to the radiation. In such situations, the exposure commonly occurs when the source is in contact with a body of a victim, primarily located in pockets of clothing or in hands. In this research, the average absorbed doses in internal organs, skin and tissues close to the source were assessed using the phantom modeling of contact human exposure by the sealed 192Ir, 137Cs and 60Co gamma sources. The results allow estimating the RBE-weighted absorbed dose values in organs and tissues to assess the possibility and severity of deterministic medical effects caused by the exposure and to compare them with the reference levels established by IAEA for performing the protective and medical actions.

Removal of Co, Sr and Cs ions from simulated radioactive wastewater by forward osmosis

The removal of Co, Sr and Cs ions form simulated radioactive wastewater using forward osmosis (FO) process was investigated. The effect of various factors on nuclide transport was examined, including membrane orientation, NaCl concentration, flow velocity, and the main factors were identified by correlation analysis. The mechanisms of nuclides transfer through membrane were explored. The results indicated that the active layer facing draw solution (AL-DS) had higher nuclide flux than AL-FS. At AL-FS mode, the highest flux of Co, Sr and Cs were only 1.54, 10.22 and 15.63 mg m^-2 h^-1 respectively by cellulose triacetate with embedded polyester screen support (CTA-ES) membrane. At AL-DS mode, the flux of Co and Cs increased when NaCl concentration and flow velocity increased. Convection, diffusion and electrostatic interactions were found to influence the nuclide transport all together. The Pearson correlation and partial correlation analysis identified that the diffusion coefficient of nuclides and reverse NaCl flux were the most important factors affecting nuclide flux through cellulose triacetate membrane. The water flux, NaCl concentration, flow velocity and partition coefficient were not the main affecting factors for nuclide flux.

A prototype tracing-technique to assess the mobility of dispersed earthworm casts on a vegetated hillslope using caesium-134 and cobalt-60

Soil transport on fully vegetated land surfaces is typically detachment limited. Rates of soil and nutrient transport, and ultimately long-term landscape evolution, are controlled by processes that supply soil material for entrainment and transport. Despite their on-going nature, many such processes operate at low rates and have not been subject to detailed investigation. We present preliminary findings from a prototype tracing approach to quantify one such process; namely to determine the relative mobility of sediment from earthworm casts on a fully vegetated hillslope surface. A 0.6 ∗ 0.5 m bounded area of pasture was prepared and fifteen intact earthworm casts representing 203 g of soil were labelled with an estimated 216 Bq of caesium-134 (¹³⁴Cs) activity and evenly distributed across the upslope half of the plot, 0.3-0.6 m from the downslope outlet. A further 15 intact casts representing 190.7 g of soil were labelled with 224 Bq of cobalt-60 (⁶⁰Co) activity and distributed between 0.3 and 0.0 m from the same outlet. All labelled casts were exposed to natural weather events over 76 days, during which time 186.3 mm of rainfall generated 16 runoff samples. A mass balance was used to partition labelled sediment from the unlabelled material. A total of 27.17 g of ⁶⁰Co-labelled casts, equivalent to 14.2% of the original mass deployed, was recovered from a distance of ≤0.3 m from their original locations. In contrast, 8.77 g of ¹³⁴Cs-labelled casts, equivalent to 4.3% of the original mass deployed, was recovered from a distance ≥0.3 m from their original locations. Some runoff-derived samples recorded an over-enrichment of radionuclide material, which suggests that intact casts may sorb more material than the original assumption predicts. Ways in which sorption can be more accurately quantified to improve the accuracy of the tracing approach are outlined.

Effects of ocean acidification on 109Cd, 57Co, and 134Cs bioconcentration by the European oyster (Ostrea edulis): Biokinetics and tissue-to-subcellular partitioning

The uptake and depuration kinetics of dissolved ¹⁰⁹Cd, ⁵⁷Co and ¹³⁴Cs were determined experimentally in the European flat oyster Ostrea edulis (Linnaeus, 1758) under different pH conditions (i.e., 8.1, 7.8 and 7.5) for 59 days. Uptake and depuration rates were variable within these elements; no effects were observed under different pH conditions for the uptake biokinetics of ¹⁰⁹Cd and ⁵⁷Co and depuration of ¹⁰⁹Cd and ¹³⁴Cs in oyster. The uptake and depuration rate constants of ¹³⁴Cs differed during the exposure phase between treatments, while the steady state concentration factors (CF_ss) were similar. The resulting Cs activity that was purged during short- and long-term depuration phases differed, while the remaining activities after thirty-nine days depuration phase (RA_39d) were similar. Co-57 depuration was affected by pCO₂ conditions: RA_39d were found to be significantly higher in oysters reared in normocapnia (pCO₂ = 350 μatm) compared to high pCO₂ conditions. Co-57 tissue distribution did not differ among the variable pCO₂ conditions, while ¹⁰⁹Cd and ¹³⁴Cs accumulated in soft tissue of oysters were found to be higher under the highest pCO₂. Additionally, Cd, Co and Cs were stored differently in various compartments of the oyster cells, i.e. cellular debris, metal-rich granules (MRG) and metallothionein-like proteins (MTLP), respectively. The subcellular sequestration of the elements at the end of the depuration phase did not differ among pH treatments. These results suggest that bioconcentration and tissue/subcellular distribution are element-specific in the oyster, and the effects of higher pCO₂ driven acidification and/or coastal acidification variably influence these processes.

Decontamination of select infrastructure materials after a radiological incident using a water-based formulation

This paper summarizes the results of the decontamination of the infrastructure materials concrete, limestone, brick and asphalt contaminated with ⁶⁰Co, ⁸⁵Sr, ¹³⁷Cs and ²⁴¹Am. The paper focuses on the effect of differences in substrate properties and of the pH of the radionuclide solution used for surface contamination on adsorption or ion exchange of the radionuclides and how these factors affect the decontamination effectiveness. A six-component chemical formulation was used and a process effectiveness of up to 76% was obtained depending on the substrate and radionuclide. Asphalt was the easiest material to decontaminate because of its more hydrophobic nature. Concrete and limestone (and to some extent brick) were less effectively decontaminated as their porous surfaces allowed penetration of radionuclides into water-filled pores in the substrate facilitating adsorption or ion exchange and making them difficult to remove. Brick was the most difficult material to decontaminate because the major component of brick is clay which retains most mono- and divalent ions. The removal of ⁶⁰Co, ⁸⁵Sr and ¹³⁷Cs from the surfaces of concrete, limestone and brick increased when the pH of the radionuclide solutions was moderately acidic to neutral compared to when they were highly acidic. The variability in the test results was similar to that observed in other studies using other decontamination methods, attributed to the inhomogeneity of the substrates used and considered representative of real infrastructure materials.

Comparison of experimental and calculated shielding factors for modular buildings in a radioactive fallout scenario

Experimentally and theoretically determined shielding factors for a common light construction dwelling type were obtained and compared. Sources of the gamma-emitting radionuclides ⁶⁰Co and ¹³⁷Cs were positioned around and on top of a modular building to represent homogeneous fallout. The modular building used was a standard prefabricated structure obtained from a commercial manufacturer. Four reference positions for the gamma radiation detectors were used inside the building. Theoretical dose rate calculations were performed using the Monte Carlo code MCNP6, and additional calculations were performed that compared the shielding factor for ¹³⁷Cs and ¹³⁴Cs. This work demonstrated the applicability of using MCNP6 for theoretical calculations of radioactive fallout scenarios. Furthermore, the work showed that the shielding effect for modular buildings is almost the same for ¹³⁴Cs as for ¹³⁷Cs.

Epimerization of Alanyl-Alanine Induced by γ-Rays Irradiation in Aqueous Solutions

Living organisms have homochiral L-amino acids in proteins and homochiral D-mononucleotides in nucleic acids. The chemical evolutionary process to protein homochirality has been discussed for many years. Although many scenarios have been proposed for homochirality in the monomeric compounds, homochirality in amino acids and mononucleotides does not always guarantee homochirality in polypeptides and polynucleotides. Integrated scenarios containing the pathways from monomer to polymer should be proposed because in the pathways oligomers and polymers as well as monomers racemize (or epimerize), degrade, and condense. This research addresses epimerization and degradation of dipeptides under γ-rays irradiation by a cobalt-60 (60Co) radiation source. The different rate constants of epimerization between diastereomeric dipeptides in the research suggest that the potential pathway toward homochirality could be much more complex.

Geological and physicochemical controls of the spatial distribution of partition coefficients for radionuclides (Sr-90, Cs-137, Co-60, Pu-239,240 and Am-241) at a site of nuclear reactors and radioactive waste disposal (St. Petersburg region, Russian Federation)

The paper presents a study of the sorption properties of sediments of different geological ages and lithological types, governing radionuclide retention in the subsurface (up to 160 m beneath the surface) within the area of potential influence of the Northwestern Center of Atomic Energy (NWCAE), St. Petersburg region, RF. The focus of this work is mostly on the sedimentary rocks of two types, i.e., weakly cemented sandstone and lithified clay formations of Cambrian and Vendian series. The first lithological unit is associated with a groundwater reservoir (Lomonosov aquifer), and the second one, with both a relative aquitard in the upper part of the Vendian formation (Kotlin clay) and a regional aquifer (Gdov aquifer) in the lower part of the formation. The main mechanisms responsible for the variability of the sorption distribution coefficient (Kd, defined as the ratio of the concentration of solute on solid phase to its concentration in solution at equilibrium) was identified for radionuclides such as Sr-90, Cs-137, Co-60, Pu-239,240, and Am-241. It was shown that the main factors contributing to the chemical heterogeneity of the Cambrian sandstone were related to the presence of secondary minerals (iron and magnesium oxides and hydroxides produced by the weathering process) in trace amounts, forming correlated layer structures. The statistical analysis of nonlinear isotherms confirmed this conclusion. For the Vendian formation, a determinate trend was established in the Kd change over depth as a result of temporal trends in the sedimentation process and pore-water chemistry. The geostatistical characteristics and the spatial correlation models for describing linear sorption of different radionuclides are not identical, and the exhibition of chemical heterogeneity of sedimentary rock of a particular lithological type depends on radionuclide chemistry. Moreover, variogram analysis for some Kd data sets (both in Cambrian and Vendian formations) demonstrates the total absence of auto-correlation (pure nugget effect). It can be supposed that sampling distances did not allow the evaluation of small-scale variability in sediment sorption properties.

2240 IRRADIATED PERSONAL DOSEMETERS FOR THE IC2014 EURADOS INTERCOMPARISON AT SEIBERSDORF

The Dosimetry Laboratory Seibersdorf was the irradiation laboratory for the IC2014 EURADOS intercomparison for individual monitoring. One hundred and twelve whole-body dosimetry systems with 30 dosemeters per dosimetry system participated in the intercomparison. In total, 2240 personal dosemeters were irradiated in photon reference radiation fields on the ISO water slab phantom in terms of personal dose equivalent Hp(10) and Hp(0.07). Two nuclide irradiation systems ((137)Cs and (60)Co gamma radiation) and a 320-kV X-ray radiation unit were used. The irradiation plan consisted of nine irradiation set-ups with five different ISO/IEC photon radiation qualities (S-Cs, S-Co, RQR 7, W-80 and W-150) and two different angles of radiation incidence (0° and 60°). Reference dose equivalent values were in the range of ∼1-500 mSv. Reference values were based on air kerma rate traceable to primary standard calibrations and appropriate conversion coefficients resulting in total expanded measurement uncertainties <6 %.

SECONDARY STANDARD CALIBRATION, MEASUREMENT AND IRRADIATION CAPABILITIES OF THE INDIVIDUAL MONITORING SERVICE AT THE HELMHOLTZ ZENTRUM MÜNCHEN: ASPECTS OF UNCERTAINTY AND AUTOMATION

The individual monitoring service at the Helmholtz Zentrum München has adopted the recommendations of the ISO 4037 and 6980 standards series as base of its dosimetric systems for X-ray, gamma and beta dosimetry. These standards define technical requirements for radiation spectra and measurement processes, but leave flexibility in the implementation of irradiations as well as in the resulting uncertainty in dose or dose rate. This article provides an example for their practical implementation in the Munich IAEA/WHO secondary standard dosimetry laboratory. It focusses on two aspects: automation issues and uncertainties in calibration.

Calculation of Coincidence Summing Correction Factors for an HPGe detector using GEANT4

The aim of this paper was to calculate the True Coincidence Summing Correction Factors (TSCFs) for an HPGe coaxial detector in order to correct the summing effect as a result of the presence of (88)Y and (60)Co in a multigamma source used to obtain a calibration efficiency curve. Results were obtained for three volumetric sources using the Monte Carlo toolkit, GEANT4. The first part of this paper deals with modeling the detector in order to obtain a simulated full energy peak efficiency curve. A quantitative comparison between the measured and simulated values was made across the entire energy range under study. The True Summing Correction Factors were calculated for (88)Y and (60)Co using the full peak efficiencies obtained with GEANT4. This methodology was subsequently applied to (134)Cs, and presented a complex decay scheme.

A new large-volume metal reference standard for radioactive waste management

A new large-volume metal reference standard has been developed. The intended use is for calibration of free-release radioactivity measurement systems and is made up of cast iron tubes placed inside a box of the size of a Euro-pallet (80 × 120 cm). The tubes contain certified activity concentrations of (60)Co (0.290 ± 0.006 Bq g(-1)) and (110m)Ag (3.05 ± 0.09 Bq g(-1)) (reference date: 30 September 2013). They were produced using centrifugal casting from a smelt into which (60)Co was first added and then one piece of neutron irradiated silver wire was progressively diluted. The iron castings were machined to the desirable dimensions. The final material consists of 12 iron tubes of 20 cm outer diameter, 17.6 cm inner diameter, 40 cm length/height and 245.9 kg total mass. This paper describes the reference standard and the process of determining the reference activity values.

Energy response of GR-200A thermoluminescence dosemeters to 60Co and to monoenergetic synchrotron radiation in the energy range 28-40 keV

The response of LiF:Mg,Cu,P thermoluminescence dosemeters (type GR-200A) to monoenergetic radiation of energy 28, 35, 38 and 40 keV was evaluated with respect to irradiation with a calibrated (60)Co gamma-ray source. High-precision measurements of the relative air kerma response performed at the SYRMEP beamline of the ELETTRA synchrotron radiation facility (Trieste, Italy) showed a significant deviation of the average response to low-energy X-rays from that to (60)Co, with an over-response from 6 % (at 28 keV) to 22 % (at 40 keV). These data are not consistent with literature data for these dosemeters, where model predictions gave deviation from unity of the relative air kerma response of about 10 %. The authors conclude for the need of additional determinations of the low-energy relative response of GR-200A dosemeters, covering a wider range of monoenergetic energies sampled at a fine energy step, as planned in future experiments by their group at the ELETTRA facility.

Comparison of the NMIJ and the ARPANSA standards for absorbed dose to water in high-energy photon beams

The authors report the results of an indirect comparison of the standards of absorbed dose to water in high-energy photon beams from a clinical linac and (60)Co radiation beam performed between the National Metrology Institute of Japan (NMIJ) and the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA). Three ionisation chambers were calibrated by the NMIJ in April and June 2013 and by the ARPANSA in May 2013. The average ratios of the calibration coefficients for the three ionisation chambers obtained by the NMIJ to those obtained by the ARPANSA were 0.9994, 1.0040 and 1.0045 for 6-, 10- and 15-MV (18 MV at the ARPANSA) high-energy photon beams, respectively. The relative standard uncertainty of the value was 7.2 × 10(-3). The ratio for (60)Co radiation was 0.9986(66), which is consistent with the results published in the key comparison of BIPM.RI(I)-K4.

Characterization of the Natural Organic Matter (NOM) in groundwater contaminated with (60)Co and (137)Cs using ultrafiltration, Solid Phase Extraction and fluorescence analysis

Spot samples of shallow groundwaters have been taken between the years 2004 and 2010 near a site formerly used for the dispersal of radioactive liquid wastes. Three sampling points, one clean (upstream), and two downstream of the contamination source, were processed by ultrafiltration (5000 Da cut-off) and Solid Phase Extraction (SPE) to determine the association of selected artificial radionuclides ((60)Co, (137)Cs) with Natural Organic Matter (NOM). The last two sampling episodes (2008 and 2010) also benefited from fluorescence analysis to determine the major character of the NOM. The fluorescence signals are reported as humic-like, fulvic-like and protein-like, which are used to characterize the different NOM types. The NOM from the clean site comprised mostly fine material, whereas the colloidal content (retained by ultrafiltration) was higher (e.g., 15-40% of the Total Organic Carbon - TOC). Most of the 137Cs was present in the colloidal fraction, whereas (60)Co was found in the filtered fraction. Fluorescence analysis, on the other hand, indicated a contrasting behavior between the clean and contaminated sites, with a dominance of protein-like material, a feature usually associated with human impacts. Finally, SPE removed almost quantitatively the protein-like material (>90%), whereas it removed a much smaller fraction of the (137)Cs (<28%). This finding indicates that the (137)Cs preferential binding occurs with a fraction other than the protein-like NOM, likely the fulvic-like or humic-like portion.

Direct megavoltage photon calibration service in Australia

The Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) maintains the Australian primary standard of absorbed dose. Until recently, the standard was used to calibrate ionisation chambers only in (60)Co gamma rays. These chambers are then used by radiotherapy clinics to determine linac output, using a correction factor (k Q) to take into account the different spectra of (60)Co and the linac. Over the period 2010-2013, ARPANSA adapted the primary standard to work in megavoltage linac beams, and has developed a calibration service at three photon beams (6, 10 and 18 MV) from an Elekta Synergy linac. We describe the details of the new calibration service, the method validation and the use of the new calibration factors with the International Atomic Energy Agency's TRS-398 dosimetry Code of Practice. The expected changes in absorbed dose measurements in the clinic when shifting from (60)Co to the direct calibration are determined. For a Farmer chamber (model 2571), the measured chamber calibration coefficient is expected to be reduced by 0.4, 1.0 and 1.1 % respectively for these three beams when compared to the factor derived from (60)Co. These results are in overall agreement with international absorbed dose standards and calculations by Muir and Rogers in 2010 of k Q factors using Monte Carlo techniques. The reasons for and against moving to the new service are discussed in the light of the requirements of clinical dosimetry.

Applicability test of methods for discrimination between external and internal contamination using an ORTEC whole body counter

Whole-body counters (WBCs) are generally used for monitoring internal radioactive contamination of radiation workers in nuclear power plants. However, it has been found that external contamination is occasionally counted as internal contamination. The previous study was conducted to provide guidance on measuring the exact internal radioactivity using a Canberra WBC. However, there is strong need to verify the application of the previous study to whole-body counting using a different type of WBC, ORTEC StandFAST II. Thus, in this study, several experiments were conducted focusing on the discrimination between external and internal contamination. Finally, it was found that counts from the front and back are still effective to distinguish external contamination from internal contamination for whole-body counting.

Monte Carlo tools to supplement experimental microdosimetric spectra

Tissue-equivalent proportional counters (TEPCs) are widely used in experimental microdosimetry for characterising the radiation quality in radiation protection and radiation therapy environments. Generally, TEPCs are filled with tissue-equivalent gas mixtures, at low gas pressure, to simulate tissue site sizes similar to the cell nucleus (1 or 2 µm). The TEPC response using Monte Carlo (MC) codes can be applied to supplement experimental measurements. Most of general-purpose MC codes currently available recourse to the condensed-history approach to model the electron transport and do not transport low-energy electrons (<1 keV), which can lead to systematic errors, especially in thin layers and in gas-condensed medium interfaces. In this work, a comparison between experimental microdosimetric spectra of (60)Co and (137)Cs radiation at different simulated sizes (from 1.0 to 3.0 μm) in pure propane versus simulated spectra obtained with two general-purpose codes FLUKA and PENELOPE, which include a detailed simulation of electron-photon transport in arbitrary materials, including gases, is presented.

Clinical data from one year follow-up of victims of the radiation accident with 60Co in Bulgaria

A severe radiation accident occurred on 14 June 2011 in an industrial irradiation facility for medical equipment sterilization in Bulgaria. Five people were exposed for 5-10 min to a 60Co source containing 137 TBq. The Emergency Department of the National Centre of Radiobiology and Radiation Protection (NCRRP), Bulgaria, put into practice the plans for providing medical care in radiation accidents and the procedures developed for assessment of injury severity, the decision-making algorithm regarding subsequent treatment, and the therapy for persons affected. The activities performed for initial assessment of the severity of injury of irradiated patients were published in 2012. Based on predictive assessments of the severity of radiation damage, it was decided that the victims required hospitalization at a specialized hematology clinic. Percy Hospital in Paris was chosen for this purpose. The aim of this report is to present the results of 1-y follow-up for three of the victims. Sadly, 1 mo after the accident, Patient 4 died from a heart attack. The medical opinion was that this was not a direct outcome of the irradiation. Patient 5 was only followed up for 4 mo (118 d) because medical follow-up is voluntary, and despite repeated calls, the patient did not respond. Medical examinations by a physician as well as hematology and biochemical tests were performed using standard laboratory methods. The obtained results were compared to the victims' personal reference limits obtained from annual health monitoring. After the accident, the recovery to normal content of peripheral blood cells was observed in all victims. Nevertheless, there were observed cases of thrombocytopenia, granulocytopenia, and leucocytopenia at various times after exposure. During the period of observation, morphological changes in red blood cells such as anisomicrocitosis, macrocytes, megalocytes, and polychromatic erythrocytes were demonstrated. During the 1-y observation period, all victims showed immediate hematopoietic recovery within 2 mo after irradiation. Slight hypocellularity of bone marrow was noted at the end of the year.

Monitoring of tritium, 60Co and 137Cs in the vicinity of the warm water outlet of the Paks Nuclear Power Plant, Hungary

Danube water, sediment and various aquatic organisms (snail, mussel, predatory and omnivorous fish) were collected upstream (at a background site) and downstream of the outlet of the warm water channel of Paks Nuclear Power Plant. Gamma emitters, tissue free-water tritium (TFWT) and total organically-bound tritium (T-OBT) measurements were performed. A slight contribution of the power plant to the natural tritium background concentration was measured in water samples from the Danube section downstream of the warm water channel. Sediment samples also contained elevated tritium concentrations, along with a detectable amount of (60)Co. In the case of biota samples, TFWT exhibited only a very slight difference compared to the tritium concentration of the Danube water, however, the OBT was higher than the tritium concentration in the Danube, independent of the origin of the samples. The elevated OBT concentration in the mollusc samples downstream of the warm water channel may be attributed to the excess emission from the nuclear power plant. The whole data set obtained was used for dose rate calculations and will be contributed to the development of the ERICA database.

On the application of an environmental radiological assessment system to an anthropomorphic surrogate

Recent developments have seen the expansion of the system of radiological protection for humans to one including protection of the environment against detrimental effects of radiation exposure, although a fully developed framework for integration of human and ecological risk assessment for radionuclides is only at an early stage. In the context of integration, significant differences exist between assessment methodologies for humans and the environment in terms of transfer, exposure, and dosimetry. The aim of this elaboration was to explore possible implications of the simplifications made within the system of environmental radiological protection in terms of the efficacy and robustness of dose-rate predictions. A comparison was conducted between human radiological assessment and environmental radiological assessment for an anthropomorphic surrogate, the results for which, produced by both the environmental and human-oriented risk assessment systems, were critically compared and contrasted. The adopted approach split the calculations into several parts, these being 1) physical transfer in an ecosystem, 2) transfer to humans, 3) internal doses to humans, and 4) external doses to humans. The calculations were carried out using both a human radiological assessment and ecological risk assessment system for the same surrogate. The results of this comparison provided indications as to where the 2 systems are amenable to possible integration and where such integration may prove difficult. Initial stage transport models seem to be an obvious component amenable for integration, although complete integration is arguably unattainable as the differences between endpoints mean that the relevant outputs from the models will not be the same. For the transfer and dosimetry components of 2 typical methodologies, it seems that the efficacy of the environmental system is radionuclide-dependent, the predictions given by the environmental system for (90) Sr and (60) Co being unsatisfactory and those for (239) Pu and (210) Po being evidently poor. Integration in this context might take the form of exploring the biokinetic models developed for humans with regard to selected animals and radionuclides. External dose assessment for environmental and human systems provide results for the surrogate that correspond quite closely providing an indication that integration in this regard is perhaps unnecessary.

Biokinetic analysis code development and applications to visualise the distribution of intake activity

Measurements for internal dose assessment are required to be conducted based on the distribution of radionuclides in the body, which may change depending on the lapsed time. In this study, a biokinetic analysis code, which can be used in practical radiation control is developed, and the results of (60)Co and (137)Cs biokinetics are visualised as examples by drawing the depositions for each organ and tissue in a figure of the body as a function of lapsed time. In addition, based on visualised biokinetics, precautions for in vivo measurements are also discussed. These discussions led to the conclusion that the information of visualised biokinetics is useful for actual measurements in practical radiation control.

Fate and transport of radiocesium, radiostrontium and radiocobalt on urban building materials

Kinetics of (137)Cs, (60)Co and (85)Sr sorption on powdered building materials in aqueous suspensions at 20 °C for interaction times of 1, 7, 14 and 28 days were studied. The (137)Cs distribution coefficient (Kd) values for all building materials except limestone practically did not change during 28 days of sorption. The Kd ((85)Sr) was several orders of magnitude lower than for (60)Co. The highest values were observed for asphalt and granite. An effective method to study the radionuclide distribution in depth of building materials using layer-by-layer sanding was developed. Using the developed method, the (137)Cs, (60)Co and (85)Sr distribution with depth of selected building materials at different air humidity, time and temperature was studied. Relative humidity (RH) was found to influence significantly the (85)Sr depth distribution in the case of granite (unlike (137)Cs and (60)Co). While (85)Sr penetrated to 0.5 mm in depth of granite at RH 30%, at RH 87% the depth of (85)Sr penetration to granite reached up to 7 mm.

Clearance monitoring using hand-held devices: operational implementation and challenges

Clearance monitoring includes all measures taken and measurements performed to ensure that all pertinent regulatory conditions are met for clearance of waste material. For bulk material, only a limited number of monitoring techniques are feasible for providing the necessary proof of compliance in the time frames and at costs commensurate with operator and regulator expectations for decommissioning projects. Operationally, measurements using hand-held devices, such as contamination monitors using scintillator or proportional counter technology, might provide sufficiently short response times at acceptable accuracies to segregate the waste streams according to their level of contamination and allowing for a decision on their further processing. Recognizing the various limitations hand-held devices might display in bulk material screening during clearance monitoring measurements, the detection limits with respect to material self-attenuation and non-homogenous activity distributions have been investigated.

Particle size characterization of aerosols generated during surface contaminated concrete demolition

The purpose of this study was to measure mass and activity distributions from the aerosols generated during the demolition of surface contaminated concrete. Air samples were collected using a cascade impactor during the mechanical hammering and dismantlement of radiologically contaminated high level waste process vaults from which the piping and components had been previously removed. The experimentally determined distribution parameters were compared with the 5.0-μm particle size referenced in the U.S. Department of Energy (U.S. DOE) regulations pertaining to internal deposition (10CFR835 and ICRP66). Mass distribution parameters were compared with their activity distribution counterparts. The Mass Median Aerodynamic Diameter (MMAD) was determined to be 4.2 μm with a Mass Geometric Standard Deviation (GSDM) of 2.3 μm, and the Co Activity Median Aerodynamic Diameter (AMAD) was determined to be 3.9 μm with an Activity Geometric Standard Deviation (GSDA) of 2.3 μm. These results are consistent with the ICRP66 5.0-μm reference particle size and the Derived Air Concentration (DAC) values referenced in 10CFR835 and utilized throughout the U.S. DOE complex.

Determining photon energy absorption parameters for different soil samples

The mass attenuation coefficients (μs) for five different soil samples were measured at 661.6, 1173.2 and 1332.5 keV photon energies. The soil samples were separately irradiated with (137)Cs and (60)Co (370 kBq) radioactive point gamma sources. The measurements were made by performing transmission experiments with a 2″ × 2″ NaI(Tl) scintillation detector, which had an energy resolution of 7% at 0.662 MeV for the gamma-rays from the decay of (137)Cs. The effective atomic numbers (Zeff) and the effective electron densities (Neff) were determined experimentally and theoretically using the obtained μs values for the soil samples. Furthermore, the Zeff and Neff values of the soil samples were computed for the total photon interaction cross-sections using theoretical data over a wide energy region ranging from 1 keV to 15 MeV. The experimental values of the soils were found to be in good agreement with the theoretical values. Sandy loam and sandy clay loam soils demonstrated poor photon energy absorption characteristics. However, clay loam and clay soils had good photon energy absorption characteristics.

Distribution of artificial gamma-ray emitting radionuclide activity concentration in the top soil in the vicinity of the Ignalina Nuclear Power Plant and other regions in Lithuania

The impact of the operating Ignalina Nuclear Power Plant (INPP) on the contamination of top soil layer with artificial radionuclides has been studied. Results of the investigation of artificial gamma-ray emitting radionuclide distribution in soil in the vicinity of the INPP and distant regions in Lithuania in 1996-2008 (INPP operational period) show that nowadays (137)Cs remains the most important artificial gamma-ray emitting radionuclide in the upper soil layer. Mean (137)Cs activity concentrations in the top soil layer in the vicinity of the INPP were found to be significantly lower compared to those in remote regions of Varėna and Plungė (~300 km from INPP). In 1996 and 1998 mean (137)Cs activity concentrations were in the range of 28-45 Bq/kg in the nearest vicinity to the INPP, 103 Bq/kg in Varėna and 340 Bq/kg in Plungė region. (137)Cs activity concentrations were 5-20 times lower in meadow soil (4-14 Bq/kg) compared to swamp and forest soil. (60)Co, the INPP origin radionuclide, was detected in samples only in 1996 and 2000, and the activity concentration of (60)Co was found to be in the range from 0.4 to 7.0 Bq/kg at the sampling ground nearest to the INPP. Average annual activity concentrations of the INPP origin (137)Cs and (60)Co in the air and depositions in the INPP region were modeled using Pasquill-Gifford equations. The modeling results of (137)Cs and (60)Co depositional load in the INPP vicinity agree with the experimentally obtained values. Our results provide the evidence that the operation of INPP did not cause any significant contamination in soil surface.

A standard for absorbed dose rate to water in a 60Co field using a graphite calorimeter at the National Metrology Institute of Japan

A primary standard for the absorbed dose rate to water in a ⁶⁰Co radiation field has been newly established at the National Metrology Institute of Japan. This primary standard combines the calorimetric measurements using a graphite calorimeter with the ionometric measurements using a thick-walled graphite cavity ionisation chamber. The calorimeter is operated in the constant temperature mode using AC Wheatstone bridges. The absorbed dose rate to water was determined to be 12 mGy s⁻¹ at a point of 1 m from the radiation source and at a water depth of 5 g cm⁻². The uncertainty on the calibration coefficient in terms of the absorbed dose to water of an ionisation chamber using this standard was estimated to be 0.39 % (k=1).

Biological microdosimetry based on radiation cytotoxicity data

Researchers in the field of radiation microdosimetry have attempted to explain the relative biological effectiveness (RBE) of different ionising photon radiation sources on the basis of the singly stochastic, microdose metric lineal energy y, which only addresses physical stochasticity related to energy (ε) deposition via single events in the critical targets (cell nuclei assumed here). Biological stochasticity related to variable nuclei geometries and cell orientations (relative to the incoming radiation) is usually not addressed. Here a doubly stochastic microdose metric, the single-event hit size q (=ε/T), is introduced which allows the track length T to be stochastic. The new metric is used in a plausible model of metabolic-activity-based in vitro cytotoxicity of low-dose ionising photon radiation. The cytotoxicity model has parameters E{q} (average single-event hit size with q assumed to be exponentially distributed) and E{α}, which is the average value of the cellular response parameter α. E{α} is referred to as the biological signature and it is independent of q. Only E{q} is needed for determination of RBE. The model is used to obtain biological-microdosimetry-based q spectra for 320-kV X-rays and (137)Cs gamma rays and the related RBE for cytotoxicity. The spectra are similar to published lineal energy y spectra for 200-kV X-rays and (60)Co gamma rays for 1-μm biological targets.

Measurements of (60)Co in massive steel samples exposed to the Hiroshima atomic bomb explosion

To study discrepancies in retrospective Hiroshima dosimetry, the specific activity of (60)Co in 16 steel samples from Hiroshima was measured using gamma-ray spectrometry in underground laboratories. There is general agreement between these new activity measurements and the specific activities derived from previously calculated dose values on the one hand and former measurements of samples gathered at distances less than 1,000 m from the center of the explosion (< 1,000 m slant range) on the other. It was found that activities at long range (> 1,300 m slant range) were mainly cosmogenically induced. Furthermore, at long range, these results are in disagreement with older measurements whose specific activity values were 10 to 100 times higher than predicted by computer model calculations in DS86 and DS02. As a consequence, the previously reported discrepancy is not confirmed.

Distribution of 60Co in steel samples from Hiroshima

This paper describes ultra low-level gamma-ray spectrometry measurements of the (60)Co activity distribution inside one 52 mm and one 41 mm thick steel sample. The samples had been exposed to the Hiroshima atomic bomb and were from the Aioi bridge and the Yokogawa bridge. Both samples were measured in a recent study aiming to back up model calculation of Hiroshima dosimetry. The (60)Co activity distributions found in this study support the assumptions made in the previous study.

Characterization of local soils and study the migration behavior of radionuclide from disposal site of LILW

Migration behavior of radionuclide is one of the most important factors to be considered for the long-term safety assessment of a radioactive waste disposal facility in a wet geological formation. In the present study, laboratory based column experiments have been carried out to assess the radionuclide migration behavior of ¹³⁷Cs and ⁶⁰Co and to evaluate the retardation factor through a clay soil layers using gamma spectrometry. Investigation was performed for a range of particle sizes and fixed column lengths to determine an appropriate value of migration rate of ¹³⁷Cs and ⁶⁰Co. The distribution pattern of particle size in soil samples were measured by sieved method. Two different particle sizes (≤ 90 μm and mixed size) were used in the column experiments. The migration rate in the clay type soil layer of particle size ≤ 90 μm was found by the order of ⁶⁰Co > ¹³⁷Cs. The maximum migration length of ⁶⁰Co in the soil layer was found to be 0-25 cm, however in the case of ¹³⁷Cs it was found at a maximum length of 0-10 cm. The distribution coefficient of ⁶⁰Co was found nearly same as that of ¹³⁷Cs. The retardation factor was found to be 1.79 and 1.94 for ⁶⁰Co and ¹³⁷Cs, respectively. The experimental breakthrough from this study indicates that the amount of radioactive cesium and cobalt released depends upon the composition of the soils.

Evaluating the performance of the ORTEC® Detective™ for emergency urine bioassay

The performance of the ORTEC(®) Detective™ as a field deployable tool for emergency urine bioassay of (137)Cs, (60)Co, (192)Ir, (169)Yb and (75)Se was evaluated against ANSI N13.30. The tested activity levels represent 10 % RL (reference level) and 1 % RL defined by [Li C., Vlahovich S., Dai X., Richardson R. B., Daka J. N. and Kramer G. H. Requirements for radiation emergency urine bioassay techniques for the public and first responders. Health Phys (in press, 99(5), 702-707 (2010)]. The tests were conducted for both single radionuclide and mixed radionuclides at two geometries, one conventional geometry (CG) and one improved geometry (IG) which improved the MDAs (minimum detectable amounts) by a factor of 1.6-2.7. The most challenging radionuclide was (169)Yb. The measurement of the mixture radionuclides for (169)Yb at the CG did not satisfy the ANSI N13.30 requirements even at 10 % RL. At 1 % RL, (169)Yb and (192)Ir were not detectable at either geometry, while the measurement of (60)Co in the mixed radionuclides satisfied the ANSI N13.30 requirements only at the IG.

Vertical migration of radionuclides in the vicinity of the chernobyl confinement shelter

Studies of vertical migration of Chernobyl-origin radionuclides in the 5-km zone of the Chernobyl Nuclear Power Plant (ChNPP) in the area of the Red Forest experimental site were completed. Measurements were made by gamma spectrometric methods using high purity germanium (HPGe) detectors with beryllium windows. Alpha-emitting isotopes of plutonium were determined by the measurement of the x-rays from their uranium progeny. The presence of 60Co, 134,137Cs, 154,155Eu, and 241Am in all soil layers down to a depth of 30 cm was observed. The presence of 137Cs and 241Am was noted in the area containing automorphous soils to a depth of 60 cm. In addition, the upper soil layers at the test site were found to contain 243Am and 243Ñm. Over the past 10 years, the 241Am/137Cs ratio in soil at the experimental site has increased by a factor of 3.4, nearly twice as much as would be predicted based solely on radioactive decay. This may be due to "fresh" fallout emanating from the ChNPP Confinement Shelter.

Independent regulatory examination of radiation situation in the areas of spent nuclear fuel and radioactive wastes storage in the Russian far east

This paper describes the findings of the radiation situation analysis on-site near Sysoeva and Razbojnik Bays. The results of radiation monitoring performed by radiological laboratory of DalRAO and studies performed by the experts from the Burnasyan Federal Medical Biophysical Centre have been used in the course of analysis. On the industrial sites, gamma dose rate reaches 60 µSv h(-1), and the specific activities of man-made radionuclides in soil reach 2.5 × 10(4) Bq kg(-1) for (137)Cs, 7.6 × 10(3) Bq kg(-1) for (90)Sr and 2.0 × 10(3) Bq kg(-1) for (60)Co. Beyond the industrial sites, there are three local parts of the area on the coast and in the off-shore water area, contaminated with man-made radionuclides. Gamma dose rate reaches 8 µSv h(-1). The radionuclide contents in soil at this area reach 3.6 × 10(3), 2.8 × 10(3) and 19 Bq kg(-1) for (137)Сs, (90)Sr and (60)Со, respectively. At the remaining part of the area nearby Sysoeva Bay, the radiation situation complies with natural background.

Radiation fields and dose assessments in Korean nuclear power plants

In the primary systems of nuclear power plants (NPPs), various radionuclides including fission products and corrosion products are generated due to the complex water chemistry conditions. In particular, (3)H, (14)C, (58)Co, (60)Co, (137)Cs, and (131)I are important or potential radionuclides with respect to dose assessment for workers and the management of radioactive effluents or dose assessment for the public. In this paper, the dominant contributors to the dose for workers and the public were reviewed and the process of dose assessment attributable to those contributors was investigated. Furthermore, an analysis was carried out on some examples of dose to workers during NPP operation.

The thermoluminescence efficiency of Li2B4O7:Cu and of CaSO4:Tm for photons

The intrinsic thermoluminescence (TL) efficiency of a TL detector relates the absorbed dose in the detector material to the light yield observed upon evaluation. Knowledge of the TL efficiency is of interest when performing numerical simulations of detector response, where only absorbed dose can be predicted. Here, the experimental determination of TL efficiency for calcium sulphate (CaSO(4):Tm) and lithium borate (Li(2)B(4)O(7):Cu) is reported. These materials are widely used in Panasonic dosemeter badges. The results of the study are in agreement with predictions from track structure theory and microdosimetry, relating an enhanced light yield at low X-ray energies to supralinear behaviour of the TL phosphor.

Personal monitor glass badge: theoretical dosemeter response calculated with the Monte Carlo transport code MCNPX

This paper describes the results of the simulation of a radiophotoluminescent (RPL) dosemeter with the Monte Carlo transport code MCNPX. The aim of this study is to calculate the response with MCNPX of the RPL dosemeter in terms of equivalent doses H(p) (0.07) and H(p)(10) using X-ray photon radiation qualities N series, together with S-Cs and S-Co nuclide radiation qualities, specified in ISO 4037-1. After comparison with reference values versus experimental results, the deviation of the theoretical responses of the RPL dosemeter proved to be lower than 5 % for reference values and lower than 10 % for experimental results. This good correlation validates the model over the energy range studied.

Characterisation of the PSI whole body counter by radiographic imaging

A joint project between the Paul Scherrer Institut (PSI) and the Institute of Radiation Physics was initiated to characterise the PSI whole body counter in detail through measurements and Monte Carlo simulation. Accurate knowledge of the detector geometry is essential for reliable simulations of human body phantoms filled with known activity concentrations. Unfortunately, the technical drawings provided by the manufacturer are often not detailed enough and sometimes the specifications do not agree with the actual set-up. Therefore, the exact detector geometry and the position of the detector crystal inside the housing were determined through radiographic images. X-rays were used to analyse the structure of the detector, and (60)Co radiography was employed to measure the core of the germanium crystal. Moreover, the precise axial alignment of the detector within its housing was determined through a series of radiographic images with different incident angles. The hence obtained information enables us to optimise the Monte Carlo geometry model and to perform much more accurate and reliable simulations.

Dose reconstruction using mobile phones

Electronic components inside mobile phones are regarded as useful tools for accident and retrospective dosimetry using optically stimulated luminescence (OSL) and thermoluminescence. Components inside the devices with suitable properties for luminescence dosimetry include, amongst others, ceramic substrates in resistors, capacitors, transistors and antenna switches. Checking the performance of such devices in dosimetric experiments is a crucial step towards developing a reliable dosimetry system for emergency situations using personal belongings. Here, the results of dose assessment experiments using irradiated mobile phones are reported. It will be shown that simple regenerative dose estimates, derived from various types of components removed from different mobile phone models, are consistent with the given dose, after applying an average fading correction factor.

Development of thermal neutron-sensitive glass dosemeter containing lithium

New radiophotoluminescence (RPL) phosphate glass containing (6)Li was successfully made from the powder of NaPO(3), Al(PO(3))(3), LiOH, HPO(3) and AgCl. The ternary diagram of NaPO(3)-Al(PO(3))(3)-LiPO(3) has clarified the region where satisfactory RPL characteristics of the glass are kept up. The synthesised phosphate glass indicated good RPL characteristics on the condition that the content of LiPO(3) was below 10 wt%. Gamma-ray irradiation experiments showed that the newly synthesised phosphate glass had satisfactory linearity and wide dynamic range in dose measurement and low variation in sensitivity. It was confirmed from thermal neutron irradiation experiments that a pair of the newly synthesised phosphate glass containing enriched (6)Li and (7)Li, or (n)Li and enriched (7)Li could be effectively used for the evaluation of thermal neutron dosimetry.

Using electron beam radiation to simulate the dose distribution for whole body solar particle event proton exposure

As a part of the near solar system exploration program, astronauts may receive significant total body proton radiation exposures during a solar particle event (SPE). In the Center for Acute Radiation Research (CARR), symptoms of the acute radiation sickness syndrome induced by conventional radiation are being compared to those induced by SPE-like proton radiation, to determine the relative biological effectiveness (RBE) of SPE protons. In an SPE, the astronaut's whole body will be exposed to radiation consisting mainly of protons with energies below 50 MeV. In addition to providing for a potentially higher RBE than conventional radiation, the energy distribution for an SPE will produce a relatively inhomogeneous total body dose distribution, with a significantly higher dose delivered to the skin and subcutaneous tissues than to the internal organs. These factors make it difficult to use a (60)Co standard for RBE comparisons in our experiments. Here, the novel concept of using megavoltage electron beam radiation to more accurately reproduce both the total dose and the dose distribution of SPE protons and make meaningful RBE comparisons between protons and conventional radiation is described. In these studies, Monte Carlo simulation was used to determine the dose distribution of electron beam radiation in small mammals such as mice and ferrets as well as large mammals such as pigs. These studies will help to better define the topography of the time-dose-fractionation versus biological response landscape for astronaut exposure to an SPE.

Mobility of radionuclides in soil/groundwater system: comparing the influence of EDTA and four of its degradation products

The adsorption behavior of 241Am, 60Co, 137Cs and 85Sr in the presence and absence of chelating ligands (ethylenediaminetetraacetic acid, ethylenediaminediacetic acid, hydroxyethyliminodiacetic acid, iminodiaceiticacid and methyliminodiacetic acid) was investigated. Sorption affinity in the absence of chelating ligands followed: Am(III)>Co(II)>Cs(I)>Sr(II). The presence of chelating ligands generally had little effect on sorption of 85Sr and 137Cs with Kd values 110 and 690 mL g(-1), respectively. But at 0.02 M of ethylenediaminetetraacetic or hydroxyethyliminodiacetic, the Kd decreased to 5 or 63 mL g(-1), respectively, where thermochemical modeling indicated almost all 85Sr is complexed with these ligands. The Kd values for 241Am and 60Co generally decreased with increasing chelating agent concentrations. In notable cases, the Kd values for Am increased at specific concentrations of 10(-3) M for IDA, MIDA and 10(-4) M for EDDA. This is proposed to be due to formation of a ternary surface complex.