Analytical shielding calculations for a proton therapy facility

The University of Pennsylvania is building a proton therapy facility in collaboration with Walter Reed Army Medical Center. The proposed facility has four gantry rooms, a fixed beam room and a research room, and will use a cyclotron as the source of protons. In this study, neutron shielding considerations for the proposed proton therapy facility were investigated using analytical techniques and Monte Carlo simulated neutron spectra. Neutron spectra calculations were done using the GEANT4 (v6.2) simulation code for various materials: water, carbon, iron, nickel and tantalum to estimate the neutron production at proton beam energies of 100, 175 and 250 MeV. Dose equivalent calculations were performed using analytical methods at various critical points within the facility, by folding the GEANT4 produced neutron spectra with dose equivalent rate data from the National Council on Radiation Protection and Measurements (NCRP) Report #144.     

Design of radiation shielding for the proton therapy facility at the National Cancer Center in Korea

The design of radiation shielding was evaluated for a proton therapy facility being established at the National Cancer Center in Korea. The proton beam energy from a 230 MeV cyclotron is varied for therapy using a graphite target. This energy variation process produces high radiation and thus thick shielding walls surround the region. The evaluation was first carried out using analytical expressions at selected locations. Further detailed evaluations have been performed using the Monte Carlo method. Dose equivalent values were calculated to be compared with analytical results. The analytical method generally yielded more conservative values. With consideration of adequate occupancy factors annual dose equivalent rates are kept <1 mSv y(-1) in all areas. Construction of the building is expected to be completed near the end of 2004 and the installation of therapy equipment will begin a few months later.     

Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.     

Thermal neutron equivalent dose assessment around the KFUPM neutron source storage area using NTDs. King Fahd University of Petroleum and Minerals

Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li2B4O7) layer for thermal neutron detection via 10B(n,alpha)7Li and 6Li(n,alpha)3H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks x cm(-2) x microSv(-1) using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSvx h(-1) up to 11 microSv x h(-1). The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv x h(-1), which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems.     

Monte Carlo simulations of the photon calibration fields at the underground laboratory of PTB

A unique photon calibration facility operated by Physikalisch-Technische Bundesanstalt (PTB) provides photon fields with area dose rates in the order of the natural environmental radiation and even below. This facility is located in an underground laboratory in the Asse salt mine at a depth of 490 m below ground, where the ambient dose equivalent rate is only 2 nSv h(-1). Radioactive sources of the nuclides (241)Am, (57)Co, (137)Cs, (60)Co and (226)Ra are used to generate photon fields with different characteristics. In the past, the basic properties of the photon field, especially the area dose rate at the reference point and the mean energy of the photon spectra, were calculated by using analytic methods. However, information about scattered photons is only accessible through an investigation of spectra by performing Monte Carlos simulations. Therefore, the photon spectra at the reference point of the calibration facility were calculated using the Monte Carlo transport code MCNP. The results obtained by using this method are of relevance for the traceability of the reference dose rate values to PTB's primary standards, as well as for the determination of the mean photon energy of the spectra. The latter was calculated with respect to the different quantities 'photon fluence', 'air kerma' and 'ambient dose equivalent'. The origin of the scattered component in the photon spectrum is investigated in detail by studying the photon field produced by the quasi-monoenergetic gamma emitter (137)Cs (E(γ) = 662 keV) under various geometrical conditions. Implications of the Monte Carlo simulations on the traceability of the dose rate reference values as well as on the assessment of uncertainties will be described.     

Personal radiation doses in PET/CT facility: measurements vs. calculations

The estimation of shielding requirement of a new positron emission tomography (PET) facility is essential. Because of penetrating annihilation photons, not only radiation safety in the vicinity of patients should be considered, but also rooms adjacent to uptake and imaging rooms should be taken into account. Before installing a PET/CT camera to nuclear medicine facilities of Helsinki University Central Hospital (HUCH), a typical PET imaging day was simulated using phantoms. Phantoms were filled with 300 +/- 36 MBq of (18)F isotope and dose rates were measured at 12 central locations in the laboratory. In addition to measurements, dose rates were also calculated using guidelines of AAPM Task Group 108. The relationship between the measured and calculated dose rates was found to be good and statistically significant, using Pearson's correlation test. The evaluated monthly doses were compared with personal dosemeter readings. AAPM's report gives practical tools for evaluation of radiation shielding. Calculations can be carried out successfully for existing hospital complexes too. However, calculations should be carried out carefully, because especially doors, windows and partitions can easily cause underestimation of shielding requirements as shown in this work.     

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.     

The refined shielding design for the cyclotron room of the Buddhist Tzu Chi General Hospital

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the inadequate maze design. The double differential neutron source from the 18O(p,n)18F reaction was adopted for the calculation. The weight window variance reduction technique, where the weight window was set by applying the adjoint flux, has been implemented in the MCNP run to facilitate the calculation of the dose rates outside the cyclotron room. Dose rates including neutron and gamma-ray components were calculated for some maze shielding modifications.     

Analysis and improvement of cyclotron thallium target room shield

Because of high neutron and gamma-ray intensities generated during bombardment of a thallium-203 target, a thallium target-room shield and different ways of improving it have been investigated. Leakage of neutron and gamma ray dose rates at various points behind the shield are calculated by simulating the transport of neutrons and photons using the Monte Carlo N Particle transport computer code. By considering target-room geometry, its associated shield and neutron and gamma ray source strengths and spectra, three designs for enhancing shield performance have been analysed: a shielding door at the maze entrance, covering maze walls with layers of some effective materials and adding a shadow-shield in the target room in front of the radiation source. Dose calculations were carried out separately for different materials and dimensions for all the shielding scenarios considered. The shadow-shield has been demonstrated to be one suitable for neutron and gamma dose equivalent reduction. A 7.5-cm thick polyethylene shadow-shield reduces both dose equivalent rate at maze entrance door and leakage from the shield by a factor of 3.     

Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent.     

Radioprotection calculations for MEGAPIE

The MEGAwatt PIlot Experiment (MEGAPIE) liquid lead-bismuth spallation neutron source will commence operation in 2006 at the SINQ facility of the Paul Scherrer Institut. Such an innovative system presents radioprotection concerns peculiar to a liquid spallation target. Several radioprotection issues have been addressed and studied by means of the Monte Carlo transport code, FLUKA. The dose rates in the room above the target, where personnel access may be needed at times, from the activated lead-bismuth and from the volatile species produced were calculated. Results indicate that the dose rate level is of the order of 40 mSv h(-1) 2 h after shutdown, but it can be reduced below the mSv h(-1) level with slight modifications to the shielding. Neutron spectra and dose rates from neutron transport, of interest for possible damage to radiation sensitive components, have also been calculated.     

Evaluation of the dose rate distribution for an air-type 60Co irradiation facility

The dose rate distributions in the 29,000-Ci 60Co irradiation facility in National Tsing Hua University were investigated by measurements and calculations. The dose rate measurements were performed using radiochromic dye films and an Exradin A2 ion chamber mounted on a PC-controlled motorised vertical translation stage. The calculations were made by using the three-dimensional point kernel code QAD-CGGP with detailed source composition and geometry modelling. The scattered gamma rays from the walls of the irradiation cell were also evaluated by using the Monte Carlo code MCNP.     

B对Pb-Mg-Al-B屏蔽功能材料的显微组织与屏蔽性能的影响

利用立式高频感应炉制备出铸造Pb-Mg-Al-B屏蔽材料,并对其显微组织和屏蔽性能进行了研究。结果表明,随着B的添加,材料组织中AlB2颗粒增多;室温抗拉强度可达105MPa,布氏硬度为160;厚度为20mm、B含量为2.0%(质量分数)的Pb-Mg-Al-B屏蔽材料对能量为250、118和65keV的X射线屏蔽率分别为90.29%、99.22%和97.9%,对γ射线的屏蔽率达到49.75%(137Cs源)和34.21%(60Co源),对中子的屏蔽率高达92.7%,说明屏蔽材料具有强度高、X(γ)射线与中子屏蔽性能优异,具备结构-功能(屏蔽)一体化的特点。     

A time-dependent dose model for glovebox processing of fissile material

A new modeling system for high-intensity neutral particle radiation fields is presented. The code PANDEMONIUM calculates external effective dose rates from neutrons and photons produced at specific locations within an industrial-size plutonium processing facility. The new version of PANDEMONIUM introduces time-dependent neutronics for source multiplication coupled with transient source and detector positions. The code is designed to provide quick and acceptably accurate total effective dose estimates for scenarios and facilities for which conventional methods prove to be too impractical or costly to model. The energy range of the code has also been extended to include the effects of prompt-fission photons.     

Neutron shielding calculations in a proton therapy facility based on Monte Carlo simulations and analytical models: criterion for selecting the method of choice

Proton therapy facilities are shielded to limit the amount of secondary radiation to which patients, occupational workers and members of the general public are exposed. The most commonly applied shielding design methods for proton therapy facilities comprise semi-empirical and analytical methods to estimate the neutron dose equivalent. This study compares the results of these methods with a detailed simulation of a proton therapy facility by using the Monte Carlo technique. A comparison of neutron dose equivalent values predicted by the various methods reveals the superior accuracy of the Monte Carlo predictions in locations where the calculations converge. However, the reliability of the overall shielding design increases if simulation results, for which solutions have not converged, e.g. owing to too few particle histories, can be excluded, and deterministic models are being used at these locations. Criteria to accept or reject Monte Carlo calculations in such complex structures are not well understood. An optimum rejection criterion would allow all converging solutions of Monte Carlo simulation to be taken into account, and reject all solutions with uncertainties larger than the design safety margins. In this study, the optimum rejection criterion of 10% was found. The mean ratio was 26, 62% of all receptor locations showed a ratio between 0.9 and 10, and 92% were between 1 and 100.     

Shielding design studies for a neutron irradiator system based on a 252Cf source.

This study aims to investigate a shielding design against neutrons and gamma rays from a source of 252Cf, using Monte Carlo simulation. The shielding materials studied were borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP4B was used to design shielding for 252Cf based neutron irradiator systems. By normalising the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independent of the intensity of the actual 252Cf source. The results show that the total dose equivalent rates were reduced significantly by the shielding system optimisation.     

Induced radioactivity in the forward shielding and semiconductor tracker of the ATLAS detector

The radioactivity induced in the forward shielding, copper collimator and semiconductor tracker modules of the ATLAS detector has been studied. The ATLAS detector is a long-term experiment which, during operation, will require to have service and access to all of its parts and components. The radioactivity induced in the forward shielding was calculated by Monte Carlo methods based on GEANT3 software tool. The results show that the equivalent dose rates on the outer surface of the forward shielding are very low (at most 0.038 microSv h(-1)). On the other hand, the equivalent dose rates are significantly higher on the inner surface of the forward shielding (up to 661 microSv h(-1)) and, especially, at the copper collimator close to the beampipe (up to 60 mSv h(-1)). The radioactivity induced in the semiconductor tracker modules was studied experimentally. The module was activated by neutrons in a training nuclear reactor and the delayed gamma ray spectra were measured. From these measurements, the equivalent dose rate on the surface of the semiconductor tracker module was estimated to be < 100 microSv h(-1) after 100 d of Large Hadron Collider (LHC) operation and 10 d of cooling.     

Shielding evaluation of a medical linear accelerator vault in preparation for installing a high-dose rate 252Cf remote afterloader

In support of the effort to begin high-dose rate 252Cf brachytherapy treatments at Tufts-New England Medical Center, the shielding capabilities of a clinical accelerator vault against the neutron and photon emissions from a 1.124 mg 252Cf source were examined. Outside the clinical accelerator vault, the fast neutron dose equivalent rate was below the lower limit of detection of a CR-39 etched track detector and below 0.14 +/- 0.02 muSv h(-1) with a proportional counter, which is consistent, within the uncertainties, with natural background. The photon dose equivalent rate was also measured to be below background levels (0.1 muSv h(-1)) using an ionisation chamber and an optically stimulated luminescence dosemeter. A Monte Carlo simulation of neutron transport through the accelerator vault was performed to validate measured values and determine the thermal-energy to low-energy neutron component. Monte Carlo results showed that the dose equivalent rate from fast neutrons was reduced by a factor of 100,000 after attenuation through the vault wall, and the thermal-energy neutron dose equivalent rate would be an additional factor of 1000 below that of the fast neutrons. Based on these findings, the shielding installed in this facility is sufficient for the use of at least 5.0 mg of 252Cf.     

Radiation dose in the high background radiation area in Kerala, India

A systematic radiological survey has been carried out in the region of high-background radiation area in Kollam district of Kerala to define the natural gamma-radiation levels. One hundred and forty seven soil samples from high-background radiation areas and five samples from normal background region were collected as per standard sampling procedures and were analysed for (238)U, (232)Th and (40)K by gamma-ray spectroscopy. External gamma dose rates at all sampling locations were also measured using a survey meter. The activities of (238)U, (232)Th and (40)K was found to vary from 17 to 3081 Bq kg(-1), 54 to 11976 Bq kg(-1) and BDL (67.4 Bq kg(-1)) to 216 Bq kg(-1), respectively, in the study area. Such heterogeneous distribution of radionuclides in the region may be attributed to the deposition phenomenon of beach sand soil in the region. Radium equivalent activities were found high in several locations. External gamma dose rates estimated from the levels of radionuclides in soil had a range from 49 to 9244 nGy h(-1). The result of gamma dose rate measured at the sampling sites using survey meter showed an excellent correlation with dose rates computed from the natural radionuclides estimated from the soil samples.