Table of contents

Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.

This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945–1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965–1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985–2005) there was very little new nuclear development, Sizewell B (the first and so far only PWR power reactor in the UK) being colocated with an early Magnox station on the rural Suffolk coast. Renewed interest in nuclear new build from 2005 onward led to a number of sites being identified for new reactors before 2025, all having previously hosted nuclear stations and including the semi-urban locations of the 1960s and 1970s. Finally, some speculative comments are made as to what a 'fifth phase' starting in 2025 might look like.

The United States radiation medical countermeasures (MCM) programme for radiological and nuclear incidents has been focusing on developing mitigators for the acute radiation syndrome (ARS) and delayed effects of acute radiation exposure (DEARE), and biodosimetry technologies to provide radiation dose assessments for guiding treatment. Because a nuclear accident or terrorist incident could potentially expose a large number of people to low to moderate doses of ionising radiation, and thus increase their excess lifetime cancer risk, there is an interest in developing mitigators for this purpose. This article discusses the current status, issues, and challenges regarding development of mitigators against radiation-induced cancers. The challenges of developing mitigators for ARS include: the long latency between exposure and cancer manifestation, limitations of animal models, potential side effects of the mitigator itself, potential need for long-term use, the complexity of human trials to demonstrate effectiveness, and statistical power constraints for measuring health risks (and reduction of health risks after mitigation) following relatively low radiation doses (<0.75 Gy). Nevertheless, progress in the understanding of the molecular mechanisms resulting in radiation injury, along with parallel progress in dose assessment technologies, make this an opportune, if not critical, time to invest in research strategies that result in the development of agents to lower the risk of radiation-induced cancers for populations that survive a significant radiation exposure incident.

This paper's goal is to assess secondary neutron doses received by paediatric patients treated for intracranial tumours using a 178 MeV proton beam. The MCNPX Monte Carlo model of the proton therapy facility, previously validated through experimental measurements for both proton and neutron dosimetry, was used. First, absorbed dose was calculated for organs located outside the clinical target volume using a series of hybrid computational phantoms for different ages and considering a realistic treatment plan. In general, secondary neutron dose was found to decrease as the distance to the treatment field increases and as the patient age increases. In addition, secondary neutron doses were studied as a function of the beam incidence. Next, neutron equivalent dose was assessed using organ-specific energy-dependent radiation weighting factors determined from Monte Carlo simulations of neutron spectra at each organ. The equivalent dose was found to reach a maximum value of ∼155 mSv at the level of the breasts for a delivery of 49 proton Gy to an intracranial tumour of a one-year-old female patient. Finally, a thorough comparison of the calculation results with published data demonstrated the dependence of neutron dose on the treatment configuration and proved the need for facility-specific and treatment-dependent neutron dose calculations.

In surface and interstitial high-dose-rate brachytherapy with either ⁶⁰Co, ¹⁹²Ir, or ¹⁶⁹Yb sources, some radiosensitive organs near the surface may be exposed to high absorbed doses. This may be reduced by covering the implants with a lead shield on the body surface, which results in dosimetric perturbations. Monte Carlo simulations in Geant4 were performed for the three radionuclides placed at a single dwell position. Four different shield thicknesses (0, 3, 6, and 10 mm) and three different source depths (0, 5, and 10 mm) in water were considered, with the lead shield placed at the phantom surface. Backscatter dose enhancement and transmission data were obtained for the lead shields. Results were corrected to account for a realistic clinical case with multiple dwell positions. The range of the high backscatter dose enhancement in water is 3 mm for ⁶⁰Co and 1 mm for both ¹⁹²Ir and ¹⁶⁹Yb. Transmission data for ⁶⁰Co and ¹⁹²Ir are smaller than those reported by Papagiannis et al (2008 Med. Phys. 35 4898–4906) for brachytherapy facility shielding; for ¹⁶⁹Yb, the difference is negligible. In conclusion, the backscatter overdose produced by the lead shield can be avoided by just adding a few millimetres of bolus. Transmission data provided in this work as a function of lead thickness can be used to estimate healthy organ equivalent dose saving. Use of a lead shield is justified.

The main objective of automatic environmental radiation warning networks is to quantify a set of radiological parameters corresponding to the medium being monitored (water, air, etc) in the shortest possible time so as to be able to provide rapid and precise information on the medium's radiological status, and on any alterations that may occur and their severity. Specifically, in this paper we present the substantial improvements that have been carried out in an automatic near-real-time radiation monitoring of a water system belonging to Radiation Alert Network of Extremadura (RARE) in southwest Spain. These improvements are based on the incorporation of (i) a gamma spectrometry system with solid scintillation detectors and compact digital electronics, (ii) continuous measurement of the water flow that is being monitored, (iii) improvements in the maintenance tasks required to optimise the operation of this type of equipment and (iv) the controlled and automated collection of water samples so that, in the case of a possible radiological anomaly, it will be possible to perform ulterior specific complementary determinations in a low-background laboratory.

This study evaluated the secondary cancer risk to various organs due to radiation treatment for breast cancer. Organ doses to an anthropomorphic phantom were measured using a photoluminescent dosimeter (PLD) for breast cancer treatment with 3D conformal radiation therapy (3D-CRT), intensity modulated radiation therapy (IMRT), and volumetric modulated arc therapy (VMAT). Cancer risk based on the measured dose was calculated using the BEIR (Biological Effects of Ionizing Radiation) VII models. The secondary dose per treatment dose (50.4 Gy) to various organs ranged from 0.02 to 0.36 Gy for 3D-CRT, but from 0.07 to 8.48 Gy for IMRT and VMAT, indicating that the latter methods are associated with higher secondary radiation doses than 3D-CRT. The result of the homogeneity index in the breast target shows that the dose homogeneity of 3D-CRT was worse than those of IMRT and VMAT. The organ specific lifetime attributable risks (LARs) to the thyroid, contralateral breast and ipsilateral lung per 100 000 population were 0.02, 19.71, and 0.76 respectively for 3D-CRT, much lower than the 0.11, 463.56, and 10.59 respectively for IMRT and the 0.12, 290.32, and 12.28 respectively for VMAT. The overall estimation of LAR indicated that the radiation-induced cancer risk due to breast radiation therapy was lower with 3D-CRT than with IMRT or VMAT.

A previous audit revealed a high frequency of adult fingers visualised on neonatal intensive care unit (NICU) chest radiographs—representing an example of inappropriate occupational radiation exposure. Radiation safety education was provided to staff and we hypothesised that the education would reduce the frequency of adult fingers visualised on NICU chest radiographs. Two cross-sectional samples taken before and after the administration of the education were compared. We examined fingers visualised directly in the beam, fingers in the direct beam but eliminated by technologists editing the image, and fingers under the cones of the portable x-ray machine. There was a 46.2% reduction in fingers directly in the beam, 50.0% reduction in fingers directly in the beam but cropped out, and 68.4% reduction in fingers in the coned area. There was a 57.1% overall reduction in adult fingers visualised, which was statistically significant (Z value − 7.48, P < 0.0001). This study supports radiation safety education in minimising inappropriate occupational radiation exposure.

The Waste Management Department of Nuclear and Energy Research Institute (IPEN) is responsible for the safety management of the waste generated at all internal research centers and that of other waste producers such as industry, medical facilities, and universities in Brazil. These waste materials, after treatment, are placed in an interim storage facility. Among them are ²²⁶Ra needles used in radiotherapy, siliceous cake arising from conversion processes, and several other classes of waste from the nuclear fuel cycle, which contain Ra-226 producing ²²²Rn gas daughter.

In order to estimate the effective dose for workers due to radon inhalation, the radon concentration at the storage facility has been assessed within this study. Radon measurements have been carried out through the passive method with solid-state nuclear track detectors (CR-39) over a period of nine months, changing detectors every month in order to determine the long-term average levels of indoor radon concentrations. The radon concentration results, covering the period from June 2012 to March 2013, varied from 0.55 ± 0.05 to 5.19 ± 0.45 kBq m⁻³. The effective dose due to ²²²Rn inhalation was further assessed following ICRP Publication 65.

The characteristics of alpha radiation have for decades been demonstrated in UK schools using small sealed ²⁴¹Am sources. There is a small but steady number of schools who report a considerable reduction in the alpha count rate detected by an end-window GM detector compared with when the source was new. This cannot be explained by incorrect apparatus or set-up, foil surface contamination, or degradation of the GM detector. The University of Liverpool and CLEAPSS collaborated to research the cause of this performance degradation. The aim was to find what was causing the performance degradation and the ramifications for both the useful and safe service life of the sources. The research shows that these foil sources have greater energy straggling with a corresponding reduction in spectral peak energy. A likely cause for this increase in straggling is a significant diffusion of the metals over time. There was no evidence to suggest the foils have become unsafe, but precautionary checks should be made on old sources.

Volumetric computed tomography dose index (CTDI_vol) is an important dose descriptor to reconstruct organ doses for patients combined with the organ dose calculated from computational human phantoms coupled with Monte Carlo transport techniques. CTDI_vol can be derived from weighted CTDI (CTDI_w) normalised to the tube current–time product (mGy/100 mAs), using knowledge of tube current–time product (mAs), tube potential (kVp), type of CTDI phantoms (head or body), and pitch. The normalised CTDI_w is one of the characteristics of a CT scanner but not readily available from the literature. In the current study, we established a comprehensive database of normalised CTDI_w values based on multiple data sources: the ImPACT dose survey from the United Kingdom, the CT-Expo dose calculation program, and surveys performed by the US Food and Drug Administration (FDA) and the National Lung Screening Trial (NLST). From the sources, the CTDI_w values for a total of 68, 138, 30, and 13 scanner model groups were collected, respectively. The different scanner groups from the four data sources were sorted and merged into 162 scanner groups for eight manufacturers including General Electric (GE), Siemens, Philips, Toshiba, Elscint, Picker, Shimadzu, and Hitachi. To fill in missing CTDI values, a method based on exponential regression analysis was developed based on the existing data. Once the database was completed, two different analyses of data variability were performed. First, we averaged CTDI values for each scanner in the different data sources and analysed the variability of the average CTDI values across the different scanner models within a given manufacturer. Among the four major manufacturers, Toshiba and Philips showed the greatest coefficient of variation (COV) (=standard deviation/mean) for the head and body normalised CTDI_w values, 39% and 54%, respectively. Second, the variation across the different data sources was analysed for CT scanners where more than two data sources were involved. The CTDI values for the scanners from Siemens showed the greatest variation across the data sources, being about four times greater than the variation of Toshiba scanners. The established CTDI database will be used for the reconstruction of CTDI_vol and then the estimation of individualised organ doses for retrospective patient cohorts in epidemiologic studies.

This paper describes the latest developments at the Institute for Energy Technology (IFE) in Norway, in the field of real-time 3D (three-dimensional) radiation risk assessment for the support of work simulation in nuclear environments. 3D computer simulation can greatly facilitate efficient work planning, briefing, and training of workers. It can also support communication within and between work teams, and with advisors, regulators, the media and public, at all the stages of a nuclear installation's lifecycle. Furthermore, it is also a beneficial tool for reviewing current work practices in order to identify possible gaps in procedures, as well as to support the updating of international recommendations, dissemination of experience, and education of the current and future generation of workers.

IFE has been involved in research and development into the application of 3D computer simulation and virtual reality (VR) technology to support work in radiological environments in the nuclear sector since the mid 1990s. During this process, two significant software tools have been developed, the VRdose system and the Halden Planner, and a number of publications have been produced to contribute to improving the safety culture in the nuclear industry.

This paper describes the radiation risk assessment techniques applied in earlier versions of the VRdose system and the Halden Planner, for visualising radiation fields and calculating dose, and presents new developments towards implementing a flexible and up-to-date dosimetric package in these 3D software tools, based on new developments in the field of radiation protection. The latest versions of these 3D tools are capable of more accurate risk estimation, permit more flexibility via a range of user choices, and are applicable to a wider range of irradiation situations than their predecessors.

This paper reviews data related to the biokinetics of phosphorus in the human body and proposes a biokinetic model for systemic phosphorus for use in updated International Commission on Radiological Protection (ICRP) guidance on occupational intake of radionuclides. Compared with the ICRP's current occupational model for systemic phosphorus (Publication 68, 1994), the proposed model provides a more realistic description of the paths of movement of phosphorus in the body and greater consistency with experimental, medical, and environmental data regarding its time-dependent distribution. For acute uptake of ³²P to blood, the proposed model yields roughly a 50% decrease in dose estimates for bone surface and red marrow and a six-fold increase in estimates for liver and kidney compared with the model of Publication 68. For acute uptake of ³³P to blood, the proposed model yields roughly a 50% increase in dose estimates for bone surface and red marrow and a seven-fold increase in estimates for liver and kidney compared with the model of Publication 68.

Gamma emitting radioactive by-products generated during the cyclotron irradiation of ¹⁸O labelled water by protons to produce ¹⁸FDG (fluorodeoxyglucose) for positron emission tomography are well characterised. However, the production of tritium (³H) through the ¹⁸O(p,t)¹⁶O nuclear reaction has not been investigated in detail. The aim of this study was to measure tritium activity produced during a large number of ¹⁸FDG production runs in order to obtain a better perspective on its impact on radioactive waste management, particularly as regards storage and disposal. Tritium was assayed by liquid scintillation counting in recovered ¹⁸O water from 24 separate production runs. The mean (SD) values of activity and activity concentration were 170 (20) kBq and 81 (8) kBq ml⁻¹ respectively. Both quantities were positively correlated with the activity of ¹⁸F. Tritium was detected in much lower concentration in water used to rinse the target vessel. The activity of tritium is such that it is exempt from regulatory control and may be combined with bulk non-active waste for disposal as Very Low Level Waste. However, variations in the irradiation conditions or the procedures for the collection of recovered water might result in its classification as Low Level Waste, necessitating a more complex disposal regime.

Existing data used to calculate the barrier transmission of scattered radiation from computed tomography (CT) are based on primary beam CT energy spectra. This study uses the EGSnrc Monte Carlo system and Epp user code to determine the energy spectra of CT scatter from four different primary CT beams passing through an ICRP 110 male reference phantom. Each scatter spectrum was used as a broad-beam x-ray source in transmission simulations through seventeen thicknesses of lead (0.00–3.50 mm). A fit of transmission data to lead thickness was performed to obtain α, β and γ parameters for each spectrum. The mean energy of the scatter spectra were up to 12.3 keV lower than that of the primary spectrum. For 120 kVp scatter beams the transmission through lead was at least 50% less than predicted by existing data for thicknesses of 1.5 mm and greater; at least 30% less transmission was seen for 140 kVp scatter beams. This work has shown that the mean energy and half-value layer of CT scatter spectra are lower than those of the corresponding primary beam. The transmission of CT scatter radiation through lead is lower than that calculated with currently available data. Using the data from this work will result in less lead shielding being required for CT scanner installations.

Long term outdoor radon measurements were recorded in Ireland using CR-39 track etch detectors. A measurement protocol was designed for this study, which was optimized for the relatively low radon concentrations expected outdoors. This protocol included pre-etching the detectors before exposure to allow radon tracks to be more easily distinguished from background. The average outdoor radon concentration for the Republic of Ireland was found to be 5.6 ± 0.7 Bq m⁻³. A statistically significant difference between inland and coastal radon concentrations was evident but no difference between mean radon concentrations on the east coast and those on the west coast was observed.

We recently published effective doses per time-integrated activity (mSv MBq⁻¹ s⁻¹) for paediatric and adult family members exposed to an adult patient released from hospital following I-131 therapy. In the present study, we intend to provide medical physicists with a methodology to estimate family member effective dose in daily clinical practice because the duration of post-radiation precautions for the patient–family member exposure scenario has not been explicitly delineated based on the effective dose. Four different exposure scenarios are considered in this study including (1) a patient and a family member standing face to face, (2) a patient and a family member lying side by side, (3) an adult female patient holding a newborn child to her chest and (4) a one-year-old child standing on the lap of an adult female patient following her I-131 therapy. The results of this study suggest that an adult female hyperthyroidism (HT) patient who was administered with 740 MBq should keep a distance of 100 cm from a 15-year-old child for six days and the same distance from other adults for seven days. The HT female patient should avoid holding a newborn against her chest for at least 16 days following hospital discharge, and a female patient treated with 5550 MBq for differentiated thyroid cancer should not hold her newborn child for at least 15 days following hospital discharge. This study also gives dose coefficients allowing one to predict age-specific effective doses to family members given the measured dose rate (mSv h⁻¹) of the patient. In conclusion, effective dose-based patient release criteria with a modified NRC two-component model provide a site medical physicist with less restrictive and age-specific radiation precaution guidance as they fully consider a patient's iodine biokinetics and photon attenuation within both the patient and the exposed family members.

A recent work has shown that the current ICRP biokinetic model for the transfer of caesium radionuclides from food to human breast milk was able to describe with satisfactory accuracy ¹³⁷Cs activity concentrations in human breast samples collected a few weeks after the Chernobyl accident as well as in samples collected some years later. However, systematic discrepancies were observed for the predictions of the activity concentrations in urine samples. In the present work, modifications to the model were investigated with the aim of improving the agreement between model predictions and data. It turned out that the disagreement for the urine data was ascribable to the mathematical simplifications used by the ICRP to describe urinary excretion in the first few days after delivery. However, the predictive performances of the model remained unchanged even when differences in the bioavailability of caesium from the ingested food types were considered or metabolic interactions between caesium and potassium were introduced into the model formulation.

This article discusses technical issues related to compliance assessment of ICNIRP 2010 basic restrictions. Several difficulties are identified in this study when assessing the spatial average and 99th percentile value of the electric field. These issues are mainly attributed to the lack of clarity in the guideline specifications, which leads to inadequate or irreproducible results. Effects on compliance results due to such ambiguous procedures are hereby investigated, with particular focus on technical issues rather than biological ones. Examples spanning from simple canonical test cases to realistic applications have been selected to highlight the strong variability in dosimetry results. Based on our findings, revisiting the ICNIRP 2010 guidelines is strongly recommended, and proposed alternative solutions are outlined.

PHE has undertaken a simple dose assessment for members of the public living in the UK at the time of the accident at the Fukushima Daiichi nuclear power station in March 2011. PHE reported that there was no public health risk to the UK from the release of material from the accident in a statement made on 29 March 2013. This assessment confirms the initial estimate of the doses which were about the same as a person in the UK would receive in an hour from natural background.

The determination of radionuclide activity concentration requires a prior knowledge of the full-energy peak (FEP) efficiency at all photon energies for a given measuring geometry. This problem has been partially solved by using procedures based on Monte Carlo simulations, developed in order to complement the experimental calibration procedures used in gamma-ray measurements of environmental samples. The aim of this article is to apply GEANT4 simulation for calibration of two HPGe detectors, for measurement of liquid and soil-like samples in cylindrical geometry. The efficiencies obtained using a simulation were compared with experimental results, and applied to a realistic measurement. Measurement uncertainties for both simulation and experimental values were estimated in order to see whether the results of the realistic measurement fall within acceptable limits. The trueness of the result was checked using the known activity of the measured samples provided by IAEA.

Cone-beam computed tomography (CBCT) is a relatively new technique for imaging of extremities. It provides high-resolution images with lower effective dose compared to conventional CT. However following the ALARA principle, CBCT-imaging protocols and practices must also be optimised to minimize the dose absorbed by the patient as well as personnel. The aim of this study is to evaluate the effect of a novel scanner-attached radiation shield on the dose absorbed by the patient and on the amount of scattered radiation around the scanner.

An orthopedic CBCT scanner was applied for comparing the doses with and without the shield during an elbow and a knee scan. A homogeneous 8 cm PMMA phantom with either an anthropomorphic Alderson phantom or a 16 cm PMMA phantom simulated the tissues of a patient. Measurements were made for several scan parameters using calibrated dose meters.

The results show that the radiation shield significantly decreased the doses measured on the patient during CBCT scans of the elbow and the knee. The usage of the shield decreased the absorbed doses by up to 95.5%. Also scattered radiation around the gantry decreased notably. The use of the shield is highly recommended, especially for pediatric patients.

In the UK, as elsewhere, there is potential to improve how radiological challenges are addressed through improvement in, or development of, a strong radiation protection (RP) safety culture. In preliminary work in the UK, two areas have been identified as having a strong influence on UK society: the healthcare and nuclear industry sectors. Each has specific challenges, but with many overlapping common factors. Other sectors will benefit from further consideration.

In order to make meaningful comparisons between these two principal sectors, this paper is primarily concerned with cultural aspects of RP in the working environment and occupational exposures rather than patient doses.

The healthcare sector delivers a large collective dose to patients each year, particularly for diagnostic purposes, which continues to increase. Although patient dose is not the focus, it must be recognised that collective patient dose is inevitably linked to collective occupational exposure, especially in interventional procedures.

The nuclear industry faces major challenges as work moves from operations to decommissioning on many sites. This involves restarting work in the plants responsible for the much higher radiation doses of the 1960/70s, but also performing tasks that are considerably more difficult and hazardous than those original performed in these plants.

Factors which influence RP safety culture in the workplace are examined, and proposals are considered for a series of actions that may lead to an improvement in RP culture with an associated reduction in dose in many work areas. These actions include methods to improve knowledge and awareness of radiation safety, plus ways to influence management and colleagues in the workplace. The exchange of knowledge about safety culture between the nuclear industry and medical areas may act to develop RP culture in both sectors, and have a wider impact in other sectors where exposures to ionising radiations can occur.

The full text of the Obituary is given in the PDF file.

The full text of the Obituary is given in the PDF file.

The full text of this article is available in the PDF provided.

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