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.     

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.     

Two types of multi-moderator neutron spectrometers: Gamma-ray insensitive type and high-efficiency type

Two types of multi-moderator neutron spectrometers were developed; one is a gamma-ray insensitive type, and the other is a high-efficiency type. An indium activation detector is loaded in the former spectrometer, which can measure the photon-dominant pulsed neutron field such as in the primary photon beam of a high-energy medical electron accelerator. The latter, in which a ³He counter is loaded, is so sensitive that it can measure leakage neutrons from a well shielded facility or even the skyshine neutrons. The response functions of the spectrometers were measured by thermal and mono-energetic neutron standard fields, and were also calculated by the one-dimensional discrete ordinates transport code, ANISN. The measured and calculated responses showed generally good agreement. A benchmark measurement of ²⁵²Cf fission neutrons by using these two spectrometers agreed well with the calculated spectrum. The spectrometers were used in the measurements of neutrons produced by a medical electron accelerator and of skyshine neutrons from an intense 14 MeV neutron source facility.     

A comparison of skyshine computational methods

A variety of methods employing radiation transport and point-kernel codes have been used to model two skyshine problems. The first problem is a 1 MeV point source of photons on the surface of the earth inside a 2 m tall and 1 m radius silo having black walls. The skyshine radiation downfield from the point source was estimated with and without a 30-cm-thick concrete lid on the silo. The second benchmark problem is to estimate the skyshine radiation downfield from 12 cylindrical canisters emplaced in a low-level radioactive waste trench. The canisters are filled with ion-exchange resin with a representative radionuclide loading, largely 60Co, 134Cs and 137Cs. The solution methods include use of the MCNP code to solve the problem by directly employing variance reduction techniques, the single-scatter point kernel code GGG-GP, the QADMOD-GP point kernel code, the COHORT Monte Carlo code, the NAC International version of the SKYSHINE-III code, the KSU hybrid method and the associated KSU skyshine codes.     

Skyshine analysis using energy and angular dependent dose-contribution fluxes obtained from air-over-ground adjoint calculation

A generalised and convenient skyshine dose analysis method has been developed based on forward-adjoint folding technique. In the method, the air penetration data were prepared by performing an adjoint DOT3.5 calculation with cylindrical air-over-ground geometry having an adjoint point source (importance of unit flux to dose rate at detection point) in the centre. The accuracy of the present method was certified by comparing with DOT3.5 forward calculation. The adjoint flux data can be used as generalised radiation skyshine data for all sorts of nuclear facilities. Moreover, the present method supplies plenty of energy-angular dependent contribution flux data, which will be useful for detailed shielding design of facilities.     

Fail-safe ion chamber beam loss monitor system for personnel protection

A fail-safe beam loss monitor system has been designed for use in the Los Alamos Neutron Scattering Center (LANSCE) Radiation Security System (RSS). The latest version (Model III) has been in use since 1991. This system has been carefully designed to protect personnel from large radiation doses caused by errant beam conditions during beam transport. Due to limited shielding between the beam transport lines and occupied areas, personnel can safely occupy these areas only if the beam losses in the transport lines are sufficiently low. Although the usual solution is to increase the bulk shielding between the beam lines and the occupied areas, the physical dimensions of the site do not permit additional shielding to be added. In this paper we will discuss the design and implementation of the fail-safe beam loss monitor system.     

Recommended improvements to the DS02 dosimetry system's calculation of organ doses and their potential advantages for the Radiation Effects Research Foundation

The Radiation Effects Research Foundation (RERF) uses a dosimetry system to calculate radiation doses received by the Japanese atomic bomb survivors based on their reported location and shielding at the time of exposure. The current system, DS02, completed in 2003, calculates detailed doses to 15 particular organs of the body from neutrons and gamma rays, using new source terms and transport calculations as well as some other improvements in the calculation of terrain and structural shielding, but continues to use methods from an older system, DS86, to account for body self-shielding. Although recent developments in models of the human body from medical imaging, along with contemporary computer speed and software, allow for improvement of the calculated organ doses, before undertaking changes to the organ dose calculations, it is important to evaluate the improvements that can be made and their potential contribution to RERF's research. The analysis provided here suggests that the most important improvements can be made by providing calculations for more organs or tissues and by providing a larger series of age- and sex-specific models of the human body from birth to adulthood, as well as fetal models.     

MCNPX vs. DORT for SNS shielding design studies

Radiation transport occurs through the 18 m long access way adjacent to the Spallation Neutron Source accelerator tunnel and the 2.2 m thick massive shielding door which closes the access way. A variety of typical materials for accelerator shielding, such as concrete and steel, were used for construction of the door to study radiation penetration. A comparison was carried out using both Monte Carlo (code MCNPX) and discrete ordinates (code DORT) methods. The beam losses during the accelerator operation are the sources for the radiation calculations. Analyses show that the results from the two methods are in good agreement.     

Radiological assessment of the structural shielding adequacy of the radiotherapy facility at Korle-Bu Teaching Hospital, Accra, Ghana

A (60)Co radiotherapy source with an initial activity of 185 TBq has exhausted its useful half-life and has been replaced with a 222 TBq (60)Co source at the Korle-Bu Teaching Hospital; a radiological assessment was performed to ascertain the shielding integrity of the facility. Dose rate at selected critical positions were calculated for the old and new sources. Dose rate measurements were also performed at these critical locations for the new source. The dose rates at all the critical locations of the public and staff access areas were within the recommended dose rate limit of 0.5 and 7.5 μSv h(-1) for the public and controlled area, respectively. The concrete biological shielding at the facility at the moment is adequate enough to attenuate the gamma photons from the new 222 TBq (60)Co source. High dose rates were recorded at the entrance to the treatment room, it is therefore recommended that optimisation of procedures should be encouraged to restrict activities in this area.     

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.     

放射源运输容器力学设计和试验验证

目的 设计一种用于运输工业辐照用特殊形式放射源的运输容器,通过试验验证运输容器的安全性能。方法 根据《放射性物品安全运输规程》和《放射性物质安全运输条例》中B(U)型货包的规定进行设计,设计的运输容器主要包括屏蔽容器、吊篮、防护罩及运输托架4部分组成,并对运输容器进行正常运输工况和运输事故工况下的3次跌落试验。结果 试验对容器外防护罩和保温层造成部分损坏,但防火层完整,不影响包装的密封和屏蔽功能,试验结果表明运输容器满足标准要求。结论 运输容器在正常运输工况和运输事故工况下的跌落测试后,包装的屏蔽和密封仍然有效,即该运输容器在规定的正常运输工况和运输事故工况下能保证放射源的运输安全。     

An improved monte carlo technique for radiation shield design

Reduction of statistical error in Monte Carlo shielding calculations is achieved by using antithetic variates with geometrical surface splitting while solving the radiation transport problems. Numerical results are presented for sample calculations for slab shields of different dimensions. The present method shows great improvement in efficiency for thick slabs and therefore is useful for practical shielding calculations.     

Radiological considerations for bulk shielding calculations of national synchrotron light source-II

Brookhaven National Laboratory is designing a new electron synchrotron for scientific research using synchrotron radiation. This facility, called the “National Synchrotron Light Source II” (NSLS-II), will provide x-ray radiation of ultra-high brightness and exceptional spatial and energy resolution. It will also provide advanced insertion devices, optics, detectors and robotics, and a suite of scientific instruments designed to maximize the scientific output of the facility. The project scope includes the design, construction, installation, and commissioning of the following accelerators: a 200 MeV linac, a booster synchrotron operating from 200 MeV to 3.0 GeV, and the storage ring which stores a maximum of 500 mA current of electrons at an energy of 3.0 GeV. It is planned to operate the facility primarily in a top-off mode, thereby maintaining the maximum variation in stored beam current to <1%. Because of the very demanding requirements for beam emittance and synchrotron radiation brilliance, the beam life-time is expected to be quite low, on the order of 2 h. Analysis of the bulk shielding for operating this facility and the input parameters used for this analysis have been discussed in this paper. The characteristics of each of the accelerators and their operating modes have been summarized with the input assumptions for the bulk shielding analysis.     

Dose measurements around spallation neutron sources

Neutron dose measurements and calculations around spallation sources appear to be of great importance in shielding research. Two spallation sources were irradiated by high-energy proton beams delivered by the Nuclotron accelerator (JINR), Dubna. Neutrons produced by the spallation sources were measured by using solid-state nuclear track detectors. In addition, neutron dose was calculated after polyethylene and concrete, using a phenomenological model based on empirical relations applied in high-energy physics. The study provides an analytical and experimental neutron benchmark analysis using the transmission factor and a comparison between the experimental results and calculations.     

The TRADE experiment: shielding calculations for the building hosting the subcritical system

The TRADE project (TRiga Accelerator Driven Experiment), to be performed at the existing TRIGA reactor at ENEA Casaccia, has been proposed as a validation of the accelerator-driven system (ADS) concept. TRADE will be the first experiment in which the three main components of an ADS--the accelerator, spallation target and sub-critical blanket--are coupled at a power level sufficient to encounter reactivity feedback effects. As such, TRADE represents the necessary intermediate step in the development of hybrid transmutation systems, its expected outcomes being considered crucial--in terms of proof of stability of operation, dynamic behaviour and licensing issues--for the subsequent realisation of an ADS Transmutation Demonstrator. An essential role in the feasibility study of the experiment is played by radioprotection calculations. Such a system exhibits new characteristics with respect to a traditional reactor, owing to the presence of the proton accelerator. As beam losses always occur under normal operating conditions of an accelerator, shielding studies need to be performed not only around the reactor but also along the beam line from the accelerator to the spallation target. This paper illustrates a preliminary evaluation, using Monte Carlo methods, of the additional shielding to be located around the reactor structures, the beam transport line and the existing reactor building to allow access into the reactor hall and to restrict the doses outside to their legal limits.     

An evaluation of radiation damage to solid state components flown in low earth orbit satellites

The effects of total ionising radiation dose upon commercial off-the-shelf semiconductors fitted to satellites operating in low Earth orbit (LEO) conditions was evaluated. The evaluation was performed for the Korea Institute of Technology SATellite-1, (KITSAT-1) which was equipped with commercial solid state components. Two approximate calculation models for space radiation shielding were developed. Verification was performed by comparing the results with detailed three-dimensional calculations using the Monte-Carlo method and measured data from KITSAT-1. It was confirmed that the developed approximate models were reliable for satellite shielding calculations. It was also found that commercial semiconductor devices, which were not radiation hardened, could be damaged within their lifetime due to the total ionising dose they are subject to in the LEO environment. To conclude, an intensive shielding analysis should be considered when commercial devices are used.     

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.     

Shielding design calculations for beam dump facility of KOMAC

A project to construct the Korea Multi-purpose Accelerator Complex (KOMAC) is currently underway targeting a high-intensity proton beam with an average current of 4.8 mA. As for the first stage of construction, a 20 MeV linac is planned to be built by 2007 and additional DTL sections will be added to increase the proton energy to 100 MeV by 2012. In this paper, preliminary shielding estimates, such as the evaluation of the gamma ray and neutron dose rate around the beam dump, have been carried out with the three-dimensional (3-D) Monte Carlo transport code MCNPX in order to determine the shielding requirements. The tentative flux calculations using the 3-D deterministic code KATRIN, which can handle a coupled charged-neutral particle transport, were also performed and their results were compared with the MCNPX calculations.