Final report of Hanford Thyroid Disease Study released.
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Dose-Response Relationship, Radiation↗
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Communicable disease and health protection quarterly review: April to June 2006: from the Health Protection Agency.
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Animals↗
Is acute lymphoblastic leukemia in children virus-related?
The sharp peak in incidence of acute lymphoblastic leukemia at ages 2 and 3 years strongly suggests the effect of an agent, whether viral or not, to which either exposure occurs only in the earliest months of life or to which immunity develops very rapidly. Suspected clusters of childhood leukemia in the neighborhoods of two British nuclear reprocessing facilities led Leo Kinlen to postulate that large-scale immigration into areas that had previously been remote and isolated offers opportunities for spread of viral infections to which most urban populations become immune at a very early age: leukemia may be a rare manifestation of infection by one or more of such viruses. He and his group have presented evidence in support of this hypothesis. Lack of increase in childhood leukemia in the contexts of the massive evacuation of mothers and children from British cities during the Second World War, and of the considerable immigration into previously isolated islands of Greece during the last several decades, indicates that some large movements of children have occurred without providing the circumstances postulated by Kinlen. The marked inverse association of leukemia risk with birth order, noted almost 30 years ago, remains unexplained and deserves to be recalled when considering the possibility of viral involvement in the etiology of acute lymphoblastic leukemia of children.
Adolescent↗
In vivo measurement of Pu dissolution parameters of MOX aerosols and related uncertainties in the values of the dose per unit intake.
The aim of this study was to compare dissolution parameter values for Pu from industrial MOX with different Pu contents. For this purpose, preliminary results obtained after inhalation exposure of rats to MOX containing 2.5% Pu are reported and compared to those obtained previously with MOX containing 5% Pu. Dissolution parameter values appear to increase when the amount of Pu decreases. Rapid fractions, f(r), of 4 x 10(-3) (s.d. = 2 x 10(-3)) and 1 x 10(-3) (s.d. = 6 x 10(-4)) and slow dissolution rates, s(s) of 2 x 10(-4) d(-1) (standard deviation, sigma = 5 x 10(-5)) and 5 x 10(-5) d(-1) (sigma = 1 x 10(-5)) were derived for MOX containing 2.5 and 5% of Pu, respectively. Simulations were performed to assess uncertainties on dose due to experimental errors. The relative standard deviations of the dose per unit intake (DPUI) due to f(r) (4-8%), are far less than those due to s(s) (about 20%), which is the main parameter altering the dose. Although quite different dissolution parameter values were derived, similar DPUIs were obtained for MOX aerosols containing 2.5 and 5% Pu which appear close to that for default Type S values.
Absorption↗
Assessment of exposure to depleted uranium.
In most circumstances, measurement of uranium excreted in urine at known times after exposure is potentially the most sensitive method for determining the amount of depleted uranium (DU) incorporated. The problems associated with this approach are that natural uranium is always present in urine because of the ingestion of natural uranium in food and drink, and that the uncertainties in the intakes as assessed from excretion measurements can be quite large, because many assumptions concerning the exposure characteristics (time pattern of exposure, route of intake, chemical form, solubility, biokinetics within the body) must be made. Applying currently available methods and instruments for the measurement of uranium in urine samples, DU incorporations of levels relevant with respect to potential health hazards can be detected reliably, even a long time after exposure.
Environmental Exposure↗
Characterisation and dissolution of depleted uranium aerosols produced during impacts of kinetic energy penetrators against a tank.
Aerosols produced during impacts of depleted uranium (DU) penetrators against the glacis (sloping armour) and the turret of a tank were sampled. The concentration and size distribution were determined. Activity median aerodynamic diameters were 1 microm (geometric standard deviation, sigma(g) = 3.7) and 2 microm (sigma(g) = 2.5), respectively, for glacis and turret. The mean air concentration was 120 Bq m(-3), i.e. 8.5 mg m(-3) of DU. Filters analysed by scanning electron microscopy (SEM) and X ray diffraction showed two types of particles (fine particles and large molten particles) composed mainly of a mixture of uranium and aluminium. The uranium oxides were mostly U3O8, UO2.25 and probably UO3.01 and a mixed compound of U and Al. The kinetics of dissolution in three media (HCO3-, HCl and Gamble's solution) were determined using in-vitro tests. The slow dissolution rates were respectively slow, and intermediate between slow and moderate, and the rapid dissolution fractions were mostly intermediate between moderate and fast. According to the in-vitro results for Gamble's solution, and based on a hypothetical single acute inhalation of 90 Bq, effective doses integrated up to 1 y after incorporation were 0.54 and 0.56 mSv, respectively, for aerosols from glacis and turret. In comparison, the ICRP limits are 20 mSv y(-1) for workers and 1 mSv y(-1) for members of the public. A kidney concentration of approximately 0.1 microg U g(-1) was predicted and should not, in this case, lead to kidney damage.
Aerosols↗
Measuring aerosols generated inside armoured vehicles perforated by depleted uranium ammunition.
In response to questions raised after the Gulf War about the health significance of exposure to depleted uranium (DU), the US Department of Defense initiated a study designed to provide an improved scientific basis for assessment of possible health effects on soldiers in vehicles struck by these munitions. As part of this study, a series of DU penetrators were fired at an Abrams tank and a Bradley fighting vehicle, and the aerosols generated by vehicle perforation were collected and characterised. A robust sampling system was designed to collect aerosols in this difficult environment and monitor continuously the sampler flow rates. The aerosol samplers selected for these tests included filter cassettes, cascade impactors, a five-stage cyclone and a moving filter. Sampler redundancy was an integral part of the sampling system to offset losses from fragment damage. Wipe surveys and deposition trays collected removable deposited particulate matter. Interior aerosols were analysed for uranium concentration and particle size distribution as a function of time. They were also analysed for uranium oxide phases, particle morphology and dissolution in vitro. These data, currently under independent peer review, will provide input for future prospective and retrospective dose and health risk assessments of inhaled or ingested DU aerosols. This paper briefly discusses the target vehicles, firing trajectories, aerosol samplers and instrumentation control systems, and the types of analyses conducted on the samples.
Aerosols↗
Excretion of depleted uranium by Gulf War veterans.
During the Persian Gulf War, in 1991, approximately 100 US military personnel had potential intakes of depleted uranium (DU), including shrapnel wounds. In 1993, the US government initiated a follow-up study of 33 Gulf War veterans who had been exposed to DU, many of whom contained embedded fragments of DU shrapnel in their bodies. The veterans underwent medical evaluation, whole-body counting, and urinalysis for uranium by kinetic phosphorescence analysis (KPA). Data are available from seven individuals who exceeded the detection limit for whole-body counting and also had elevated urinary uranium. Urinary excretion rates, in microg U g(-1) creatinine, were determined in 1997 and 1999. The body contents, in mg DU, were determined in 1997; it is assumed there were no significant decreases in total body content in the interim. For the 1997 data, the mean fractional excretion was (2.4 +/- 2.8) x 10(-5) g(-1) creatinine, and for the 1999 data, the mean was (1.1 +/- 0.6) x 10(-5) g(-1) creatinine. However, these means are not significantly different, nor is there any correlation of excretion rate with body content. Thus, human data available to date do not provide any basis for determining the effects of particle surface area, composition and solubility, and biological processes such as encapsulation, on the excretion rate.
Air Pollutants, Radioactive↗
Doses and risks from the ingestion of Dounreay fuel fragments.
The radiological implications of ingestion of nuclear fuel fragments present in the marine environment around Dounreay have been reassessed by using the Monte Carlo code MCNP to obtain improved estimates of the doses to target cells in the walls of the lower large intestine resulting from the passage of a fragment. The approach takes account of the reduction in dose due to attenuation within the intestinal wall and self-absorption of radiation in the fuel fragment itself. In addition, dose is calculated on the basis of a realistic estimate of the anatomical volume of the lumen, rather than being based on the average mass of the contents, as in the current ICRP model. Our best estimates of doses from the ingestion of the largest Dounreay particles are at least a factor of 30 lower than those predicted using the current ICRP model. The new ICRP model will address the issues raised here and provide improved estimates of dose.
Administration, Oral↗
Assessment of intakes of mixtures of radionuclides.
Australia has several uranium mines and a large number of mineral sand mines, with associated processing facilities. Exposures resulting from these mining and processing operations usually involve intakes of mixtures of radionuclides. This work describes the development of a suite of first order, linear compartment models, based on the ICRP Publication 66 respiratory tract model, and an analytical solution to the decay equations, for assessing the consequences of such intakes. The computer programs based on these models directly compute excretion, organ retention and organ and whole-body doses for intakes of either single radionuclides or any mixture of radionuclides belonging to the same radioactive decay chain. The intake can be via inhalation, ingestion or injection, and can be acute, chronic or of limited duration. The starting concentration and degree of secular (dis)equilibrium can be specified for each radionuclide. No assumptions need to be made about the relative magnitudes of the radioactive half-lives of the different nuclides.
Air Pollutants, Radioactive↗
A United States perspective on long-term management of areas contaminated with radioactive materials.
The US has far-reaching and extensive experience in the long-term management of areas contaminated with radioactive materials. This experience base includes the Department of Energy's continued follow-up with Hiroshima and Nagasaki from the 1940s at the Radiological Effects Research Foundation in Hiroshima, Japan, the long-term management of the Marshall Islands Programme, the clean-up of the US nuclear weapons complex and the ongoing management of accident sites such as in Palomares, Spain. This paper discusses the lessons learnt and best practices gained from this far-reaching and extensive experience in the long-term management of areas contaminated with radioactive materials.
Decision Support Systems, Management↗
The shielding design process--new plants to decommissioning.
BNFL have over 25 years experience of designing nuclear plant for the whole-fuel cycle. In the UK, a Nuclear Decommissioning Authority (NDA) is to be set up to ensure that Britain's nuclear legacy is cleaned up safely, securely and cost effectively. The resulting challenges and opportunities for shielding design will be substantial as the shielding design process was originally devised for the design of new plants. Although its underlying principles are equally applicable to decommissioning and remediation of old plants, there are many aspects of detailed application that need to adapt to this radically different operating environment. The paper describes both the common issues and the different challenges of shielding design at different operational phases. Sample applications will be presented of both new plant and decommissioning projects that illustrate not only the robust nature of the processes being used, but also how they lead to cost-effective solutions making a substantive and appropriate contribution to radiological protection goals.
Computer-Aided Design↗
Comparison of measured and calculated dose rates for the Castor HAW 20/28 CG.
In January 2003 neutron and gamma dose rate measurements at a CASTOR HAW 20/28 CG were performed by the Bundesamt für Strahlenschutz at Gorleben. First, commercial dose rate measurement devices were used, then spectral measurements with a Bonner sphere system were made to verify the results. Axial and circumferential dose rate profiles were measured near the cask surface and spectral measurements were performed for some locations. A shielding analysis of the cask was performed with the MCNP Monte Carlo Code with ENDF/B-VI cross section libraries. The cask was modelled 'as built', i.e. with its real inventory, dimensions and material densities and with the same configuration and position as in the storage facility. The average C/E-ratios are 1.3 for neutron dose rates and 1.4 for gamma dose rates. Both the measured and calculated dose rates show the same qualitative trends in the axial and circumferential direction. The spectral measurements show a variation in the spectra across the cask surface. This correlates with the variation found in the C/E-ratios. At cask midheight good agreement between the Bonner sphere system and the commercial device (LB 6411) is found with a 7% lower derived H*(10) dose rate from the Bonner sphere system.
Computer-Aided Design↗
Transmutation of actinides in power reactors.
Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.
Actinoid Series Elements↗
Modelling of radiation field around spent fuel container.
Operation of nuclear reactors leads to the production of spent nuclear fuel (SNF). There are two basic strategies of SNF management: ultimate disposal of SNF in geological formations and recycle or repeated utilisation of reprocessed SNF. In both options, there is an urgent necessity to study radiation properties of SNF. Information about SNF radiation properties is required at all stages of SNF management. In order to reach more effective utilisation of nuclear materials, new fuel cycles are under development based on uranium-plutonium, uranium-thorium and some other types of nuclear fuel. These promising types of nuclear fuel are characterised by quite different radiation properties at all the stages of nuclear fuel cycle (NFC) listed above. So, comparative analysis is required for radiation properties of different nuclear fuel types at different NFC stages. The results presented here were obtained from the numerical analysis of the radiation field around transport containers of different SNF types and in SNF storage. The calculations are carried out with the application of the computer code packages SCALE-4.3 and MCNP-4C. Comparison of the dose parameters obtained for different models of the transport container with experimental data allowed us to make certain conclusions about the errors of numerical results caused by the approximate geometrical description of the transport container.
Computer Simulation↗
ISFSI site boundary radiation dose rate analyses.
Across the globe nuclear utilities are in the process of designing and analysing Independent Spent Fuel Storage Installations (ISFSI) for the purpose of above ground spent-fuel storage primarily to mitigate the filling of spent-fuel pools. Using a conjoining of discrete ordinates transport theory (DORT) and Monte Carlo (MCNP) techniques, an ISFSI was analysed to determine neutron and photon dose rates for a generic overpack, and ISFSI pad configuration and design at distances ranging from 1 to -1700 m from the ISFSI array. The calculated dose rates are used to address the requirements of 10CFR72.104, which provides limits to be enforced for the protection of the public by the NRC in regard to ISFSI facilities. For this overpack, dose rates decrease by three orders of magnitude through the first 200 m moving away from the ISFSI. In addition, the contributions from different source terms changes over distance. It can be observed that although side photons provide the majority of dose rate in this calculation, scattered photons and side neutrons take on more importance as the distance from the ISFSI is increased.
Computer Simulation↗
Survey on lost and poorly controlled radioactive sources in non-medical applications in Iran.
Despite well-developed Regulatory Authority systems for radiation protection, there are some concerns about lost-controlled and poorly controlled radioactive sources owing to abundance and dispersion of the users. This survey covers the current situation of the categories 3 through 5 of radioactive sources (based on the categorisation of the IAEA) in non-medical applications in Iran. The investigation performed on a selected group of non-medical users of radioactive sources in different provinces of Iran. Findings show that unauthorised supply and legacy sources in addition to abandonment of sources under regulatory control are the main reasons for lost-controlled sources. Considering the possibility of increasing poorly controlled sources which can lead to increase in the number of lost-controlled sources, further developments are needed in national regulatory control as well as international cooperation and discipline in this regard.
Humans↗
Project salt vault: a demonstration disposal of high-level radioactive solids in Lyons, Kansas, salt mine.
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Hot Temperature↗