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.