Table of contents

The full text of the Invited Editorial is given in the PDF file.

This review updates material prepared for the UK Government Committee Examining Radiation Risks from Internal Emitters (CERRIE) and also refers to the new recommendations of the International Commission on Radiological Protection (ICRP) and other recent developments. Two conclusions from CERRIE were that ICRP should clarify and elaborate its advice on the use of its dose quantities, equivalent and effective dose, and that more attention should be paid to uncertainties in dose and risk estimates and their implications. The new ICRP recommendations provide explanations of the calculation and intended purpose of the protection quantities, but further advice on their use would be helpful. The new recommendations refer to the importance of understanding uncertainties in estimates of dose and risk, although methods for doing this are not suggested. Dose coefficients (Sv per Bq intake) for the inhalation or ingestion of radionuclides are published as reference values without uncertainty.

The primary purpose of equivalent and effective dose is to enable the summation of doses from different radionuclides and from external sources for comparison with dose limits, constraints and reference levels that relate to stochastic risks of whole-body radiation exposure. Doses are calculated using defined biokinetic and dosimetric models, including reference anatomical data for the organs and tissues of the human body. Radiation weighting factors are used to adjust for the different effectiveness of different radiation types, per unit absorbed dose (Gy), in causing stochastic effects at low doses and dose rates. Tissue weighting factors are used to take account of the contribution of individual organs and tissues to overall detriment from cancer and hereditary effects, providing a simple set of rounded values chosen on the basis of age- and sex-averaged values of relative detriment. While the definition of absorbed dose has the scientific rigour required of a basic physical quantity, the same is not true of the ICRP protection quantities equivalent and effective dose (i.e. those measured in sieverts). The ICRP quantities are intended for practical application in radiological protection and the choice of radiation and tissue weighting factors used in their calculation involves simplifying assumptions regarded as acceptable for this purpose. Best estimates of doses and risks to individuals and specific population groups may be calculated using ICRP biokinetic and dosimetric approaches, but would require the use of best available information on RBE and age-, sex- and population-specific risk factors.

Consideration of uncertainties is important in applications such as the assessment of the probability of cancer causation for an individual and in estimating doses in epidemiological studies. While the ICRP system of protection does not take explicit account of uncertainties, an understanding of the various contributions to uncertainty can be seen to be of value when making judgments on the optimisation of protection.

This paper provides a review of the 2007 recommendations of the International Commission on Radiological Protection (ICRP). These new recommendations take account of the latest biological and physical information and consolidate the additional guidance provided by ICRP since 1990. The changes to the scientific data are not substantial. ICRP has retained its fundamental hypothesis for the induction of stochastic effects of linearity of dose and effect without threshold and a dose and dose-rate effectiveness factor (DDREF) of 2 to derive nominal risk coefficients for low doses and low dose rates. While the overall detriment from low radiation doses has remained unchanged, ICRP has made adjustments to the values of the radiation and tissue weighting factors. In particular, the tissue weighting factor for breast has increased while that for gonads has decreased. There are some presentational changes to the system of protection. While ICRP has maintained the three fundamental principles—justification, optimisation of protection, and dose limitation—it has attempted to develop a more holistic approach to radiological protection covering all exposure situations—planned, existing and emergency—and all radiation sources, whether of natural or artificial origin. This approach should ensure that attention is focused on those exposures that can reasonably be controlled. It has also strengthened the principle of optimisation of protection with a particular emphasis on the use of constraints for planned exposure situations and reference levels for existing and emergency exposure situations. Dose constraints and reference levels are categorised into three bands which should assist in rationalising the many values of dose restrictions given in earlier ICRP publications. There are no changes to the dose limits. ICRP also indicates its intentions with respect to the development of further guidance on the protection of the environment. The fact that these new recommendations are more a matter of consolidation of previous ICRP recommendations and guidance should provide confidence that the system of protection established by and large in its present form several decades ago has reached a certain level of maturity. As such, no major changes to radiological protection regulations based on the 1990 recommendations should be necessary.

In the first few hours after an accidental release of radioactivity to the atmosphere it is likely that limited measurements of radioactivity in the environment will be available on which to make decisions concerning protection measures and radiation monitoring activities, and that monitoring data will be supplemented by the predictions of models. There will be imprecision associated with these predictions, partly resulting from lack of knowledge (for example, about the nature of the release and the actual state of the weather), partly due to imprecision in the models themselves and partly due to intrinsic imprecision associated with the accuracy of the measurements. This study considers the relative importance of the key parameters influencing assessment imprecision and discusses the implications for emergency response systems.

For neutron energies ranging from 1 keV to 20 MeV, the kerma coefficients for elements H, C, N, O, light water, and ICRU tissue were deduced respectively from microscopic cross sections and Monte Carlo simulation (MCNP code). The results are consistent within admitted uncertainties with values evaluated by an international group (Chadwick et al 1999 Med. Phys. 26 974–91). The ambient dose equivalent generated in the ISO-recommended neutron field for an Am–Be neutron source (ISO 8529-1: 2001(E)) was obtained from the kerma coefficients and Monte Carlo calculation. In addition, it was calculated directly by multiplying the neutron fluence by the fluence-to-ambient dose conversion coefficients recommended by ICRP (ICRP 1996 ICRP Publication 74 (Oxford: Pergamon)). The two results agree well with each other. The main feature of this work is our Monte Carlo simulation design and the treatments differing from the work of others in the calculation of neutron energy transfer in non-elastic processes.

A portable system, using electrostatic collection, for the measurement of exhaled thoron activity in humans is described, together with details of the basic theory, equipment, calibration procedures, measurement and preliminary use. The portable system built on experience at the Argonne National Laboratory (ANL) to achieve a reduction in measurement time from 30 h to 200 min, and to increase the total efficiency of the system from 50% (at ANL) to 70%. The limit of detection of the system (2σ above zero) is 0.007 Bq of thoron activity outside the body. The total standard error of this system is 47% for a thorium lung burden of 0.22 Bq. The average background of this scintillation detector was 0.003 counts min⁻¹.

Brick, sand, marble and cement are mainly used for the construction of dwellings in Pakistan. Therefore, knowledge of the presence of natural radioactivity in these materials is of great importance in order to assess the radiological hazards associated with them. In this context, specific activities of ²²⁶Ra, ²³²Th and ⁴⁰K were measured in brick, sand, marble and cement samples collected from different localities of the North West Frontier province and federally administered tribal areas, Pakistan, using a P-type coaxial high-purity germanium spectrometer. In brick samples, the average measured activities for ²²⁶Ra, ²³²Th and ⁴⁰K were 30 ± 15, 41 ± 21 and 523 ± 182 Bq kg⁻¹, whereas in sand samples, these values were 19 ± 9, 30 ± 15 and 769 ± 461 Bq kg⁻¹, respectively. In marble samples, the average specific activities of ²²⁶Ra, ²³²Th and ⁴⁰K were found to be 18 ± 19, 18 ± 21 and 299 ± 328 Bq kg⁻¹, whilst in cement samples they were 24 ± 6, 18 ± 4 and 244 ± 29 Bq kg⁻¹, respectively. Radium equivalent activities were also calculated and were found to be 129 ± 54, 121 ± 57, 67 ± 60 and 68 ± 9 Bq kg⁻¹ for brick, sand, marble and cement samples, respectively. The annual average effective doses from these samples were 0.37 ± 0.15, 0.33 ± 0.15, 0.20 ± 0.17 and 0.20 ± 0.03 mSv, respectively. External and internal hazard indices were less than one for all the samples studied.

Activity concentrations of ²³⁸U, ²³²Th and ⁴⁰K in rocks and soil samples collected from Sannur cave, Beni Suef governorate, eastern desert of Egypt, were determined using the high-resolution gamma spectrometry technique. The results show that the concentrations of the naturally occurring radionuclides are the following: ²³⁸U ranged from 8.51 ± 1.23 to 20.66 ± 2.12 Bq kg⁻¹, ²³²Th ranged from 7.69 ± 1.02 to 22.73 ± 1.60 Bq kg⁻¹ and ⁴⁰K ranged from 185.74 ± 0.42 to 2084.70 ± 23.30 Bq kg⁻¹. The radium equivalent activity (Ra_eq), the absorbed dose rate (D), and the external hazard index (H_ex) were also calculated and compared to the international recommended values. The radon concentration and radon exhalation rate from the rock and soil samples were measured using the Can technique. The average value of annual effective dose for cave workers is 1.98 mSv y⁻¹, while for visitors it is 2.4 µSv per visit. The radon exhalation rate varies from 0.21 ± 0.03 to 1.28 ± 0.02 Bq m⁻² h⁻¹. A positive correlation has been observed between uranium content and radon exhalation rate.

Natural radioactivity due to the presence of ²²⁶Ra, ²³²Th and ⁴⁰K in selected building materials (cement, sand, bricks, gypsum and ceramic) used in Egypt was measured using a gamma-ray spectrometer with an HPGe detector. The average activity concentrations observed in different building materials ranged from 10.0 ± 1.3 to 109 ± 6, <2 to 55.8 ± 2.2 and 5.5 ± 1.7 to 684 ± 34 Bq kg⁻¹ for ²²⁶Ra, ²³²Th and ⁴⁰K, respectively. Based on these, together with previously reported results, the effective doses received by the residents of different types of house within all Egyptian governorates were assessed using the WinMat computer program. The results were below 1 mSv a⁻¹ in all cases. The collective effective dose indoors was assessed as 15 000 man Sv and the excess effective dose due to building materials was 0.07 mSv a⁻¹.

The Khewera Mines located in Pakistan contain the world's second largest reserves of rock salt. Rock salt is used in Pakistan in food recipes. It was decided to investigate the concentrations of naturally occurring radionuclides in rock salt from the Khewera Mines. Samples of rock salt were collected from 10 different locations and analysed by gamma spectrometry. The mean activity concentrations of ²²⁶Ra, ²³²Th and ⁴⁰K were 790 ± 262, 640 ± 162 and 23 000 ± 6000 mBq kg⁻¹, respectively. The mean annual effective dose due to the intake of natural radionuclides from rock salt for adults was estimated to be 0.0638 ± 0.015 mSv, which is lower than the average annual effective dose of 0.29 mSv received per caput worldwide due to the ingestion of natural radionuclides, as reported by the United Nations Scientific Committee on the Effects of Atomic Radiation in 2000.

The Health and Safety Executive (HSE) published its revised Safety Assessment Principles for Nuclear Facilities (SAPs) in December 2006. The SAPs are primarily intended for use by HSE's inspectors when judging the adequacy of safety cases for nuclear facilities. The revised SAPs relate to all aspects of safety in nuclear facilities including the technical discipline of criticality safety. The purpose of this paper is to set out for the benefit of a wider audience some of the thinking behind the final published words and to provide an insight into the development of UK regulatory guidance. The paper notes that it is HSE's intention that the Safety Assessment Principles should be viewed as a reflection of good practice in the context of interpreting primary legislation such as the requirements under site licence conditions for arrangements for producing an adequate safety case and for producing a suitable and sufficient risk assessment under the Ionising Radiations Regulations 1999 (SI1999/3232 www.opsi.gov.uk/si/si1999/uksi_19993232_en.pdf).

The full text of the letters in this issue is given in the PDF file.

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The full text of the meeting reports in this issue is given in the PDF file.

The full text of the book reviews in this issue is given in the PDF file.

Summaries of the articles in this issue are given in the PDF file.