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Data Interpretation, Statistical↗
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Estimated neutron dose to embryo and foetus during commercial flight.
A study has been carried out to assess the radiation exposure from cosmic-ray neutrons to the embryo and foetus of pregnant aircrew and air travellers in consideration of the radiation exposure from cosmic-ray neutrons to the embryo and foetus. A Monte Carlo analysis was performed to determine the equivalent dose from neutrons to the brain and body of an embryo at 8 weeks and to the foetus at the 3, 6 and 9 month periods. Neutron fluence-to-absorbed dose conversion coefficients for the foetal brain and for the entire foetal body (isotropic irradiation geometry) have been determined at the four developmental stages. The equivalent dose rate to the foetus during commercial flights has been further evaluated considering the fluence-to-absorbed dose conversion coefficients, a neutron spectrum measured at an altitude of 11.3 km and an ICRP-92 radiation-weighting factor for neutrons. This study indicates that the foetus can exceed the annual dose limit of 1 mSv for the general public after, for example, 15 round trips on commercial trans-Atlantic flights.
Aircraft↗
Differential absorbed dose distributions in lineal energy for neutrons and gamma rays at the mono-energetic neutron calibration facility.
Absorbed dose distributions in lineal energy for neutrons and gamma rays of mono-energetic neutron sources from 140 keV to 15 MeV were measured in the Fast Neutron Laboratory at Tohoku University. By using both a tissue-equivalent plastic walled counter and a graphite-walled low-pressure proportional counter, absorbed dose distributions in lineal energy for neutrons were obtained separately from those for gamma rays. This method needs no knowledge of energy spectra and dose distributions for gamma rays. The gamma-ray contribution in this neutron calibration field >1 MeV neutron was <3%, while for <550 keV it was >40%. The measured neutron absolute absorbed doses per unit neutron fluence agreed with the LA150 evaluated kerma factors. By using this method, absorbed dose distributions in lineal energy for neutrons and gamma rays in an unknown neutron field can be obtained separately.
Absorption↗
Central dose data management and analysis in IT-driven radiation protection strategies.
The applications of information technology in health care are now widespread and continue to grow. Medical imaging is at the forefront of this revolution and the introduction of digital detection methods to replace film is now addressing the diagnostic X-ray market, the most routinely employed imaging modality. The introduction of picture archiving and communication systems, hospital and radiology information systems is well underway, and the integration of radiation protection initiatives into these developments is desirable. In north-west UK, a project aimed at developing and implementing IT-driven radiation protection strategies has been underway for the past 10 y. Such strategies are geared towards the support of European Commission patient dose directive 97/43 EURATOM, in particular the need to implement clinical audit, patient dose audit and to establish dose reference levels. This paper demonstrates the national and local requirements for establishing a central dose data management system for use in radiation protection strategies. In particular, such a system can help develop and support the role of a medical physics expert in optimisation. The scientific requirements for such an approach are presented in this paper, and a prototype system is described. Preliminary results obtained with the central data management facility are also presented and the implication for analysing multiple site dose data in optimisation strategies for digital radiographic technology is highlighted.
Calibration↗
Retrospective patient dose analysis of a digital radiography system in routine clinical use.
The log files generated in the flat-panel detector based digital radiography systems (Philips Digital Diagnost) at 11 different hospitals were used to acquire data regarding the dose-area product (DAP) and the entrance air kerma (EAK) for all X-ray examinations performed in routine clinical use. A retrospective statistical analysis of the data with respect to the dose levels was performed for several examination types. The resulting average dose levels were compared with the diagnostic reference levels (DRLs). While for the same type of examination the median dose level at different hospitals could differ by up to a factor of 5, almost all the dose levels were found to be much below the published DRL. Only in one hospital for one type of examination (Thorax PA) the DRL for the DAP was slightly exceeded, while for the same site and examination the EAK was still below the DRL. In conclusion, examination log files can be used for retrospective dose analysis and correlation with DRLs.
Dose-Response Relationship, Radiation↗
Gamma dose from activation of internal shields in IRIS reactor.
The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.
Algorithms↗
Developing instruments for measurement of plutonium activity in the environmental samples in the case of radiation accident.
According to the Russian norms, the permissible specific activity of 239Pu in the air and water are the lowest. Still the problem of low (compared to permissible) Pu activity measurement for environmental samples arises, for instance, in the case of radiation accident during transportation or storage of excess weapons-grade plutonium transferred to the civil sphere of utilisation. The base method of Pu activity measurement is alpha spectrometry by using silicon detector. A new model of alpha spectrometer, based on silicon surface-barrier detector with background level less than 1-3 pulses per day has been developed. A simple unfolding method for the determination of Pu activity of samples with various thickness based on alpha spectrum measurement has been also developed. L-X photon detection is also used for reducing the measurement threshold of Pu activity determination. A method based on detecting L-X photons of Pu, excited by X rays and beta particles, is applied for further reducing the threshold.
Alpha Particles↗
Characteristics of high-energy neutrons estimated using the radioactive spallation products of Au at the 500-MeV neutron irradiation facility of KENS.
We carried out a shielding experiment of high-energy neutrons, generated from a tungsten target bombarded with primary 500-MeV protons at KENS, which penetrated through a concrete shield in the zero-degree direction. We propose a new method to evaluate the spectra of high-energy neutrons ranging from 8 to 500 MeV. Au foils were set in a concrete shield, and the reaction rates for 13 radionuclides produced by the spallation reactions on the Au targets were measured by radiochemical techniques. The experimental results were compared with those obtained by the MARS14 Monte-Carlo code. A good agreement (between them) was found for energies beyond 100 MeV. The profile of the neutron spectrum, ranging from 8 to 500 MeV, does not depend on the thickness of the concrete shield.
Algorithms↗
A system for the calculation and visualisation of radiation field for maintenance support in nuclear power plants.
A system has been developed to improve the efficiency of maintenance work while decreasing the radiation exposure of maintenance personnel in nuclear power plants. The input data for dose rate calculation are automatically generated by using computer-aided design data. Changes for the input data corresponding to the progress of maintenance work, such as installation of a radiation shield and removal of a component, are easily input interactively on a graphical user interface (GUI). A new method was proposed which searches the sets of source and detector points between which gamma-ray attenuation is changed by the component movement. The calculation is performed only for the changed sets, so that the change of the three-dimensional dose rate distribution is calculated rapidly according to the work progress. The dose rate distribution and the radiation exposure of maintenance personnel are displayed three-dimensionally in colour with plant components and pipes on the GUI.
Body Burden↗
Integrated nuclear data utilisation system for innovative reactors.
A five-year research and development project on an integrated nuclear data utilisation system was initiated in 2002, for developing innovative nuclear energy systems such as accelerator-driven systems. The integrated nuclear data utilisation system will be constructed as a modular code system, which consists of two sub-systems: the nuclear data search and plotting sub-system, and the nuclear data processing and utilisation sub-system. The system will be operated with a graphical user interface in order to enable easy utilisation through the Internet by both nuclear design engineers and nuclear data evaluators. This paper presents an overview of the integrated nuclear data utilisation system, describes the development of a prototype system to examine the operability of the user interface and discusses specifications of the two sub-systems.
Computer Graphics↗
A new low-energy bremsstrahlung generator for GEANT4.
The 2BN bremsstrahlung cross section is a well-adapted distribution to describe the radiative processes at low-electron kinetic energy (E(k) < 500 keV). In this work a method to implement this distribution in a Monte Carlo generator is developed.
Computer Simulation↗
GENII-LIN: a new object-oriented interface for the GENII code.
GENII-LIN is a new object-oriented interface for GENII, a well-known analysis tool for the health impact evaluation of accidental or chronic release of radionuclides to the environment. GENII-LIN has been developed under the Linux OS, chosen because it is a stable, reliable and open source. The suite codes that constitute the original GENII simulation kernel have been re-implemented and built, using the G77, the Fortran 77 module of the GCC collection. The code modules and the flow of data among them are at present fully controlled by a management interface written in C++, which is the core of the package. GENII-LIN comes with a new graphical user interface (GUI), built using the QT libraries by Trolltech. Through the wizard-like GUI, the user has full control over the code and can easily handle both the input and the output files.
Body Burden↗
Pressure vessel calculations for VVER-440 reactors.
For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.
Algorithms↗
Measurements of the response functions of a large size NE213 organic liquid scintillator for neutrons up to 800 MeV.
The response functions of 25.4 cm (length) x 25.4 cm (diameter) NE213 organic liquid scintillator have been measured for neutrons in the energy range from 20 to 800 MeV at the Heavy-Ion Medical Accelerator in Chiba (HIMAC) and at the Research Center for Nuclear Physics (RCNP) of Osaka University. At HIMAC, white (continuous) energy spectrum neutrons were produced by the 400 MeV per nucleon carbon ion bombardment on a thick graphite target, whose energy spectrum has already been measured by Kurosawa et al., [Nucl. Sci. Eng. 132, 30 (1999)] and the response functions of the time-of-flight-gated monoenergetic neutrons in a wide energy range from 20 to 800 MeV were simultaneously measured. At RCNP, the quasi-monoenergetic neutrons were produced via 7Li(p,n)7Be reaction by 250 MeV proton beam bombardment on a thin 7Li target, and the TOF-gated 245 MeV peak neutrons were measured. The absolute peak neutron yield was obtained by the measurement of 478 keV gamma rays from the 7Be nuclei produced in a Li target. The measured results show that the response functions for monoenergetic neutrons < 250 MeV have a recoil proton plateau and an edge around the maximum light output, which increases with increasing incident neutron energy, on the other hand > 250 MeV, the plateau and the edge become unclear because the proton range becomes longer than the detector size and the escaping protons increase. It can be found that the efficiency of the 24.5 cm (diameter) x 25.4 cm (length) NE213 for the 250 MeV neutrons is -10 times larger than the 12.7 cm (length) x 12.7 cm (diameter) NE213, which is widely used as a neutron spectrometer.
Electrons↗
Neodymium neutron cross section measurements.
Neutron capture and transmission measurements were performed by the time-of-flight technique at the Rensselaer Polytechnic Institute LINAC using metallic neodymium samples. The capture measurements were made at the 25-m-long flight station with a 16-segment NaI(Tl) multiplicity detector, and the transmission measurements were performed at 15 and 25 m flight stations with a 6Li glass scintillation detector. After the data were collected and reduced, resonance parameters were determined by simultaneously fitting the transmission and capture data with the multilevel R-matrix Bayesian code SAMMY. The resonance parameters for all naturally occurring neodymium isotopes lie within the energy range of 1.0-500 eV. The resulting resonance parameters were used to calculate the capture resonance integral with this energy region and were compared to calculations obtained when using the resonance parameters from ENDF-B/VI. The RPI parameters gave a resonance integral value of 32 +/- 0.5 b that is approximately 7% lower than that obtained with the ENDF-B/VI parameters. The current measurements significantly reduce the statistical uncertainties on the resonance parameters when compared with previously published parameters.
Chromium Isotopes↗
Characterisation of radioactive waste products associated with plant decommissioning.
The inventory of radioactivity that must be considered in the decommissioning of a typical 1000 MWe Spanish pressurised water reactor (PWR) was investigated as part of a generic plant decommissioning study. Analyses based on DORT models (in both R-Z and R-theta geometries) were used with representative plant operating history and core power distribution data in defining the expected neutron environment in regions near the reactor core. The activation analyses were performed by multiplying the DORT scalar fluxes by energy-dependent reaction cross sections (based on ENDF/B-VI data) to generate reaction rates on a per atom basis. The results from the ORIGEN2 computer code were also used for determining the activities associated with certain nuclides where multi-group cross section data were not available. In addition to the bulk material activation of equipment and structures near the reactor, the activated corrosion-product (or 'crud') deposits on system and equipment surfaces were considered. The projected activities associated with these sources were primarily based on plant data and experience from operating PWR plants.
Computer Simulation↗
Activation analysis of concrete and graphite in the experimental reactor RUS.
The decommissioning and dismantling of nuclear installations after their service life involves the necessary disassembling, handling and disposing of a large amount of radioactive equipment and structures. In particular, the concrete that has been used as a biological reactor shield and graphite that has been used as a moderator-reflector represent the majority of waste, requiring geological disposal. To reduce this undesirable volume to the minimum and to successfully plan the dismantling and disposal of radioactive materials to storage facilities, the activations of the structures should be accurately evaluated. In the framework of the decommissioning and the dismantling of the experimental reactor of the University of Strasbourg, detailed activation estimates have been conducted to characterise the graphite and the structural materials present in the reactor environment. For this purpose, the chemical compositions of fresh graphite samples and different types of concrete have been determined by activation analysis in the research reactors OSIRIS and ORPHEE of CEA Saclay (France). Then, the activations of graphite, concrete and other materials have been calculated in the whole reactor, as a function of the three main nuclear data libraries, i.e. ENDF, JEF and JENDL. In parallel, the activations of representative graphite and concrete samples have been measured experimentally. The comparison of theoretical predictions with experimental values validates the approach and the methodology used in the present study and tests the consistency and the reliability of the nuclear data used for activation analysis. We believe that a similar approach could also be used for the decommissioning of industrial nuclear reactors.
Algorithms↗
Remanent dose rates around the collimators of the LHC beam cleaning insertions.
The LHC will require an extremely powerful and unprecedented collimation system. As approximately 30% of the LHC beam is lost in the cleaning insertions, these will become some of the most radioactive locations around the entire LHC ring. Thus, remanent dose rates to be expected during later repair or maintenance interventions must be considered in the design phase itself. As a consequence, the beam cleaning insertions form a unique test bed for a recently developed approach to calculate remanent dose rates. A set of simulations, different in complexity, is used in order to evaluate methods for the estimation of remanent dose rates. The scope, as well as the restrictions, of the omega-factor method are shown and compared with the explicit simulation approach. The latter is then used to calculate remanent dose rates in the beam cleaning insertions. Furthermore, a detailed example for maintenance dose planning is given.
Computer Simulation↗