Influence of neutron cross-section resonances on organ/tissue equivalent and effective dose coefficients for the ICRP voxel phantoms.

The materials which compose the ICRP Voxel phantoms used in the computation of conversion coefficients involve neutron interaction cross-sections that have resonances at specific energies. Depending on the energy bin structure used in the computations, these cross-section resonances may occur at energies that fall between energies at which dose coefficients are computed, thus their effects may not be completely accounted for in the reported coefficients. In the present study, a highly refined energy grid that closely follows the resonance structure in the phantom material cross-sections was identified and used to calculate dose coefficients. Both the equivalent organ/tissue doses for male and female voxel phantoms were computed as well as their summation to obtain the effective dose coefficients. The used refined energy grid tracks very closely the cross-sections in the vicinity of the resonances. The resulting refined energy grid coefficients are compared to coefficients for the coarser energy grid used in ICRP Publication 116. Additionally, reference spectra have been folded with both the fine and coarse sets of conversion coefficients. The resulting total effective doses for these reference spectra are used to assess the adequacy of the dose coefficients calculated on the original ICRP 116 energy grid. The dose coefficients were similarly computed for the local skin dose on the trunk of the body using the ICRU Report 95 phantom. The overall impact of the resonances on the organ/tissue equivalent dose, the effective dose, and the local skin dose are presented and discussed. In general, it was found that resonances can impact neutron dose coefficients, but in most cases the wide range of neutron energies encountered minimized this effect. The impact of resonances was further limited when computing effective dose due to organ/tissue summing and sex-averaging. For the neutron fields studied here, the impact was below 5%.

Vertically Well-Aligned ZnO Nanoscintillator Arrays with Improved Photoluminescence and Scintillation Properties.

ZnO nanoarrays were grown via a low-temperature hydrothermal method. Solutions, each with different additive combinations, were prepared and evaluated. The effects of the additives involved in the growth procedure, i.e., ammonium hydroxide and sodium citrate, were studied in terms of the morphological, optical and scintillation properties of the ZnO nanostructures. Measurement of the nanorod (NR) length, corresponding photoluminescence (PL) and scintillation spectra and their dependence on the additives present in the solution are discussed. ZnO NRs grown on a silica substrate, whose UV transmission was found to be better than glass, showed high-quality structural and optical properties. It was found that the addition of sodium citrate significantly reduced defects and correspondingly increased the intrinsic near-band-edge (NBE) UV emission intensity at ~380 nm. To obtain high-quality nanostructures, samples were annealed in a 10% H2 + 90% N2 atmosphere. The anneal in the forming gas atmosphere enhanced the emission of the UV peak by reducing defects in the nanostructure. NRs are highly tapered towards the end of the structure. The tapering process was monitored using time growth studies, and its effect on PL and reflectance spectra are discussed. A good alpha particle response was obtained for the grown ZnO NRs, confirming its potential to be used as an alpha particle scintillator. After optimizing the reaction parameters, it was concluded that when ammonium hydroxide and sodium citrate were used, vertically well-aligned and long ZnO nanoarrays with highly improved optical and scintillation properties were obtained.

Fetal atomic bomb survivor dosimetry using the J45 series of pregnant female phantoms with realistic survivor exposure scenarios: comparisons to dose estimates in the DS02 system.

A significant source of information on radiation-induced biological effects following in-utero irradiation stems from studies of atomic bomb survivors who were pregnant at the time of exposure in Hiroshima, and to a lesser extent, from survivors in Nagasaki. Dose estimates to the developing fetus for these survivors have been assigned in prior dosimetry systems of the Radiation Effects Research Foundation as the dose to the uterine wall within the non-pregnant adult stylized phantom, originally designed for the dosimetry system DS86 and then carried forward in DS02. In a prior study, a new J45 (Japanese 1945) series of high-resolution phantoms of the adult pregnant female at 8 weeks, 15 weeks, 25 weeks, and 38-weeks post-conception was presented. Fetal and maternal organ doses were estimated by computationally exposing the pregnant female phantom series to DS02 free-in-air cumulative photon and neutron fluences at three distances from the hypocenter at both Hiroshima and Nagasaki under idealized frontal (AP) and isotropic (ISO) particle incidence. In this present study, this work was extended using realistic angular fluences (480 directions) from the DS02 system for seven radiation source terms, nine different radiation dose components, and five shielding conditions. In addition, to explore the effects of fetal position within the womb, four new phantoms were created and the same irradiation scenarios were performed. General findings are that the current DS02 fetal dose surrogate overestimates values of fetal organ dose seen in the J45 phantoms towards the cranial end of the fetus, especially in the later stages of pregnancy. For example, for in-open exposures at 1000 m in Hiroshima, the ratio of J45 fetal brain dose to DS02 uterine wall dose is 0.90, 0.82, and 0.70 at 15 weeks, 25 weeks, and 38-weeks, respectively, for total gamma exposures, and are 0.64, 0.44, and 0.37 at these same gestational ages for total neutron exposures. For organs in the abdominal and pelvic regions of the fetus, dose gradients across gestational age flatten and later reverse, so that DS02 fetal dosimetry begins to underestimate values of fetal organ dose as seen in the J45 phantoms. For example, for the same exposure scenario, the ratios of J45 fetal kidney dose to DS02 uterine wall dose are about 1.09 from 15 to 38 weeks for total gamma dose, and are 1.30, 1.56, and 1.75 at 15 weeks, 25 weeks, and 38 weeks, respectively, for the total neutron dose. Results using the new fetal positioning phantoms show this trend reversing for a head-up, breach fetal position. This work supports previous findings that the J45 pregnant female phantom series offers significant opportunities for gestational age-dependent assessment of fetal organ dose without the need to invoke the uterine wall as a fetal organ surrogate.

Fetal and Maternal Atomic Bomb Survivor Dosimetry Using the J45 Pregnant Female Phantom Series: Considerations of the Kneeling and Lying Posture with Comparisons to the DS02 System.

ABSTRACT: Organ dosimetry data of the atomic bomb survivors and the resulting cancer risk models derived from these data are currently assessed within the DS02 dosimetry system developed through the Joint US-Japan Dosimetry Working Group. In DS02, the anatomical survivor models are limited to three hermaphroditic stylized phantoms-an adult (55 kg), a child (19.8 kg), and an infant (9.7 kg)-that were originally designed for the preceding DS86 dosimetry system. As such, organ doses needed for assessment of in-utero cancer risks to the fetus have continued to rely upon the use of the uterine wall in the adult non-pregnant stylized phantom as the dose surrogate for all fetal organs regardless of gestational age. To address these limitations, the Radiation Effects Research Foundation (RERF) Working Group on Organ Dose (WGOD) has established the J45 (Japan 1945) series of high-resolution voxel phantoms, which were derived from the UF/NCI series of hybrid phantoms and scaled to match mid-1940s Japanese body morphometries. The series includes male and female phantoms-newborn to adult-and four pregnant female phantoms at gestational ages of 8, 15, 25, and 38 wk post-conception. In previous studies, we have reported organ dose differences between those reported by the DS02 system and those computed by the WGOD using 3D Monte Carlo radiation transport simulations of atomic bomb gamma-ray and neutron fields for the J45 phantoms series in their traditional "standing" posture, with some variations in their facing direction relative to the bomb hypocenter. In this present study, we present the J45 pregnant female phantoms in both a "kneeling" and "lying" posture and assess the dosimetric impact of these more anatomically realistic survivor models in comparison to current organ doses given by the DS02 system. For the kneeling phantoms facing the bomb hypocenter, organ doses from bomb source photon spectra were shown to be overestimated by the DS02 system by up to a factor of 1.45 for certain fetal organs and up to a factor of 1.17 for maternal organs. For lying phantoms with their feet in the direction of the hypocenter, fetal organ doses from bomb source photon spectra were underestimated by the DS02 system by factors as low as 0.77, while maternal organ doses were overestimated by up to a factor of 1.38. Organs doses from neutron contributions to the radiation fields exhibited an increasing overestimation by the DS02 stylized phantoms as gestational age increased. These discrepancies are most evident in fetal organs that are more posterior within the mother's womb, such as the fetal brain. Further analysis revealed that comparison of these postures to the original standing posture indicate significant dose differences for both maternal and fetal organ doses depending on the type of irradiation. Results from this study highlight the degree to which the existing DS02 system can differ from organ dosimetry based upon 3D radiation transport simulations using more anatomically realistic models of those survivors exposed during pregnancy.

Comparison of out-of-field normal tissue dose estimates for pencil beam scanning proton therapy: MCNP6, PHITS, and TOPAS.

Monte Carlo (MC) methods are considered the gold-standard approach to dose estimation for normal tissues outside the treatment field (out-of-field) in proton therapy. However, the physics of secondary particle production from high-energy protons are uncertain, particularly for secondary neutrons, due to challenges in performing accurate measurements. Instead, various physics models have been developed over the years to reenact these high-energy interactions based on theory. It should thus be acknowledged that MC users must currently accept some unknown uncertainties in out-of-field dose estimates. In the present study, we compared three MC codes (MCNP6, PHITS, and TOPAS) and their available physics models to investigate the variation in out-of-field normal tissue dosimetry for pencil beam scanning proton therapy patients. Total yield and double-differential (energy and angle) production of two major secondary particles, neutrons and gammas, were determined through irradiation of a water phantom at six proton energies (80, 90, 100, 110, 150, and 200 MeV). Out-of-field normal tissue doses were estimated for intracranial irradiations of 1-, 5-, and 15-year-old patients using whole-body computational phantoms. Notably, the total dose estimates for each out-of-field organ varied by approximately 25% across the three codes, independent of its distance from the treatment volume. Dose discrepancies amongst the codes were linked to the utilized physics model, which impacts the characteristics of the secondary radiation field. Using developer-recommended physics, TOPAS produced both the highest neutron and gamma doses to all out-of-field organs from all examined conditions; this was linked to its highest yields of secondary particles and second hardest energy spectra. Subsequent results when using other physics models found reduced yields and energies, resulting in lower dose estimates. Neutron dose estimates were the most impacted by physics model choice, and thus the variation in out-of-field dose estimates may be even larger than 25% when considering biological effectiveness.

Compact portable sources of high-LET radiation: Validation and potential application for galactic cosmic radiation countermeasure discovery.

Implementation of a systematic program for galactic cosmic radiation (GCR) countermeasure discovery will require convenient access to ground-based space radiation analogs. The current gold standard approach for GCR simulation is to use a particle accelerator for sequential irradiation with ion beams representing different GCR components. This has limitations, particularly for studies of non-acute responses, strategies that require robotic instrumentation, or implementation of complex in vitro models that are emerging as alternatives to animal experimentation. Here we explore theoretical and practical issues relating to a different approach to provide a high-LET radiation field for space radiation countermeasure discovery, based on use of compact portable sources to generate neutron-induced charged particles. We present modeling studies showing that DD and DT neutron generators, as well as an AmBe radionuclide-based source, generate charged particles with a linear energy transfer (LET) distribution that, within a range of biological interest extending from about 10 to 200 keV/µm, resembles the LET distribution of reference GCR radiation fields experienced in a spacecraft or on the lunar surface. We also demonstrate the feasibility of using DD neutrons to induce 53BP1 DNA double-strand break repair foci in the HBEC3-KT line of human bronchial epithelial cells, which are widely used for studies of lung carcinogenesis. The neutron-induced foci are larger and more persistent than X ray-induced foci, consistent with the induction of complex, difficult-to-repair DNA damage characteristic of exposure to high-LET (>10 keV/µm) radiation. We discuss limitations of the neutron approach, including low fluence in the low LET range (<10 keV/µm) and the absence of certain long-range features of high charge and energy particle tracks. We present a concept for integration of a compact portable source with a multiplex microfluidic in vitro culture system, and we discuss a pathway for further validation of the use of compact portable sources for countermeasure discovery.

Specific Absorbed Fractions for Spontaneous Fission Neutron Emitters in the ICRP Reference Pediatric Voxel Phantom Series.

ABSTRACT: Specific absorbed fractions (SAFs) are key components in the workflow of internal exposure assessment following the intake of a radionuclide, allowing quick conversion of particle energy released in a source region to the expected absorbed dose in target regions throughout the body. For data completeness, SAFs for spontaneous fission neutron emitters are currently needed for the recently adopted ICRP reference pediatric voxel phantom series. With 77 source regions within each reference individual and 28 radionuclides decaying via spontaneous fission, full Monte Carlo simulation requires significant computation time. In order to reduce this burden, a novel method for neutron SAF estimation was undertaken. The Monte Carlo N-Particle version 6.1 (MCNP6) simulation package was chosen to simulate the 252 Cf Watt fission neutron spectrum originating from 15 source regions in each phantom; dose estimation within 41 target tissues allowed for assessment of the SAF value for each source-target pair. For the remaining source regions, chord length distributions were computed using MATLAB code to determine the separation between the source-target pairs within the pediatric phantom series. These distance distributions were used in conjunction with a 252 Cf neutron dose point kernel calculated in soft tissue, which was modified to account for the source region's depth from the surface of the body. Lastly, the 252 Cf SAF dataset was extended to the other 27 spontaneous fission neutron emitters based on differences in the Watt fission spectrum parameters of each radionuclide. This methodology has been shown to accurately estimate spontaneous fission neutron SAFs to within 20% of the Monte Carlo estimated value for most source-target pairs in the ICRP reference pediatric series.

Developing A Radiation Protection Hub.

The Where are the Radiation Professionals (WARP)? statement issued by the National Council on Radiation Protection and Measurements estimates that in 10 y, there will be a human capital crisis across the radiation safety community. The ability to respond to this shortage will be amplified by the fact that many radiation protection (health physics) academic programs will find it difficult to justify their continued existence since they are low-volume programs, both in terms of enrollment and research funding, compared to the research funding return and visibility of more highly subscribed and highly funded academic disciplines. In addition, across the national laboratory complex, radiation protection research groups have been disbanded or dramatically reduced in size. The loss of both of these national resources is being accelerated by low and uncertain government funding priorities. The most effective solution to this problem would be to form a consortium that would bring together the radiation protection research, academic, and training communities. The goal of such a consortium would be to engage in research, education, and training of the next generation of radiation protection professionals. Furthermore, the consortium could bring together the strengths of different universities, national laboratory programs, and other entities in a strategic manner to accomplish a multifaceted research, educational, and training agenda. This vision would forge a working and funded relationship between major research universities, national laboratories, 4-y degree institutions, technical colleges, and other partners.

Overview of the ICRP/ICRU adult reference computational phantoms and dose conversion coefficients for external idealised exposures.

This paper reviews the ICRP Publications 110 and 116 describing the reference computational phantoms and dose conversion coefficients for external exposures. The International Commission on Radiological Protection (ICRP) in its 2007 Recommendations made several revisions to the methods of calculation of the protection quantities. In order to implement these recommendations, the DOCAL task group of the ICRP developed computational phantoms representing the reference adult male and female and then calculated a set of dose conversion coefficients for various types of idealised external exposures. This paper focuses on the dose conversion coefficients for neutrons and investigates their relationship with the conversion coefficients of the protection and operational quantities of ICRP Publication 74. Contributing factors to the differences between these sets of conversion coefficients are discussed in terms of the changes in phantoms employed and the radiation and tissue weighting factors.

Calibration of indium response functions in an Au-In-BSE system up to 800 MeV.

Calibration of the response functions of a gold (Au)-indium (In) dual foil Bonner sphere extended (BSE) system was described. The response of the In and Au foil of the system was calculated using MCNPX code with different activation cross-sectional libraries: (ACTL and ENDF VI for gold and ACTL and 532DOS2 for In). To verify and correct the calculated response functions the Bonner sphere set (BSS) was irradiated using (252)Cf and (241)AmBe sources of known neutron strengths for neutrons ranging from thermal to 20 MeV, and was irradiated at the 800-MeV neutron beam of the Los Alamos Neutron Science Center. The neutron spectrum of the 800 MeV beam was determined using time-of-flight (TOF) technique. We observed that the uncertainty of activation cross section in the resonance region can result in great uncertainty in the MCNPX-calculated response functions of activation foil-based BSS. The MCNPX-calculated response functions must be corrected using neutron sources of known spectrum and strength.

Secondary neutron spectra from modern Varian, Siemens, and Elekta linacs with multileaf collimators.

Neutrons are a by-product of high-energy x-ray radiation therapy (threshold for [gamma,n] reactions in high-Z material -7 MeV). Neutron production varies depending on photon beam energy as well as on the manufacturer of the accelerator. Neutron production from modern linear accelerators (linacs) has not been extensively compared, particularly in terms of the differences in the strategies that various manufacturers have used to implement multileaf collimators (MLCs) into their linac designs. However, such information is necessary to determine neutron dose equivalents for different linacs and to calculate vault shielding requirements. The purpose of the current study, therefore, was to measure the neutron spectra from the most up-to-date linacs from three manufacturers: Varian 21EX operating at 15, 18, and 20 MV, Siemens ONCOR operating at 15 and 18 MV, and Elekta Precise operating at 15 and 18 MV. Neutron production was measured by means of gold foil activation in Bonner spheres. Based on the measurements, the authors determined neutron spectra and calculated the average energy, total neutron fluence, ambient dose equivalent, and neutron source strength. The shapes of the neutron spectra did not change significantly between accelerators or even as a function of treatment energy. However, the neutron fluence, and therefore the ambient dose equivalent, did vary, increasing with increasing treatment energy. For a given nominal treatment energy, these values were always highest for the Varian linac. The current study thus offers medical physicists extensive information about the neutron production of MLC-equipped linacs currently in operation and provides them information vital for accurate comparison and prediction of neutron dose equivalents and calculation of vault shielding requirements.

Effects of tertiary MLC configuration on secondary neutron spectra from 18 MV x-ray beams for the Varian 21EX linear accelerator.

The effect of the jaw configuration and the presence and configuration of the tertiary multileaf collimator (MLC) on the secondary neutron spectra for an 18 MV Varian 21EX linear accelerator (linac) is investigated in detail. The authors report the measured spectra for four collimator (jaw-and-MLC) configurations. These configurations represent the extreme settings of the jaws and MLC and should therefore describe the range of possible fluence and spectra that may be encountered during use of this linac. In addition to measurements, a Monte Carlo model was used to simulate the four collimator configurations and calculate the energy spectra and fluence at the same location as it was measured. The Monte Carlo model was also used to calculate the sources of neutron production in the linac head for each collimator configuration. They found that photoneutron production in the linac treatment head is dominated by the order in which the primary photon beam intercepts the high-Z material. The primary collimator, which has the highest position in the linac head (in a fixed location), is the largest source of secondary neutrons. Thereafter, the collimator configuration plays a role in where the neutrons originate. For instance, if the jaws are closed, they intercept the beam and contribute substantially to the secondary neutron production. Conversely, if the jaws are open, the MLC plays a larger role in neutron production (assuming, of course, that it intercepts the beam). They found that different collimator configurations make up to a factor of 2 difference in the ambient dose equivalent.

Calculation of effective dose from measurements of secondary neutron spectra and scattered photon dose from dynamic MLC IMRT for 6 MV, 15 MV, and 18 MV beam energies.

Effective doses were calculated from the delivery of 6 MV, 15 MV, and 18 MV conventional and intensity-modulated radiation therapy (IMRT) prostate treatment plans. ICRP-60 tissue weighting factors were used for the calculations. Photon doses were measured in phantom for all beam energies. Neutron spectra were measured for 15 MV and 18 MV and ICRP-74 quality conversion factors used to calculate ambient dose equivalents. The ambient dose equivalents were corrected for each tissue using neutron depth dose data from the literature. The depth corrected neutron doses were then used as a measure of the neutron component of the ICRP protection quantity, organ equivalent dose. IMRT resulted in an increased photon dose to many organs. However, the IMRT treatments resulted in an overall decrease in effective dose compared to conventional radiotherapy. This decrease correlates to the ability of an intensity-modulated field to minimize dose to critical normal structures in close proximity to the treatment volume. In a comparison of the three beam energies used for the IMRT treatments, 6 MV resulted in the lowest effective dose, while 18 MV resulted in the highest effective dose. This is attributed to the large neutron contribution for 18 MV compared to no neutron contribution for 6 MV.

Bremsstrahlung doses from natural uranium ingots.

In the past, some privately owned commercial facilities in the United States were involved in producing or processing radioactive materials used in the production of atomic weapons. Seven different geometrical objects, representative of the configurations of natural uranium metal potentially encountered by workers at these facilities, are modelled to determine gamma ray and bremsstrahlung dose rates. The dose rates are calculated using the MCNP5 code and also by using the MICROSHIELD point-kernel code. Both gamma ray and bremsstrahlung dose rates are calculated and combined to obtain a total dose rate. The two methods were found to be in good agreement despite differences in modelling assumptions and method differences. Computed total dose rates on the surface of these objects ranged from approximately 51-84 microSv h(-1) and 17-95 microSv h(-1) using the MCNP5 and the MICROSHIELD modeling, respectively. The partitioning of the computed dose rates between gamma rays and bremsstrahlung were the same order of magnitude for each object.

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.

Measurements of secondary neutron dose from 15 MV and 18 MV IMRT.

Secondary neutron dose-equivalents were determined for conventional and intensity modulated radiation therapy (IMRT) prostate treatments for 15 and 18 MV X-ray beams. Conventional and IMRT treatment plans were generated to deliver 45 Gy to the prostate, seminal vessicles and external and internal iliac lymph nodes. Neutron spectra were determined by unfolding measurements from a TLD-based Bonner sphere system. Treatments using 18 MV IMRT and conventional plans result in neutron ambient dose-equivalents of 687 and 112 mSv, respectively. Delivery of the 15 MV IMRT and conventional plans results in neutron ambient dose-equivalents of 327 and 52 mSv, respectively. The data illustrate that using lower photon energies for IMRT reduces the secondary neutron dose, while still achieving comparable treatment volume coverage and sparing critical normal tissue.

Determination of dosimetric characteristics of OptiSeed(TM) a plastic brachytherapy (103)Pd source.

A new (103)Pd plastic brachytherapy source, OptiSeed(TM) Model 1032P, is being introduced by International Brachytherapy sa (IBt). Measurements of the dose distributions about the source were performed using LiF thermoluminescent dosimeters (TLD-100) in Virtual Water(TM). MCNP5 calculations were performed to determine the dose distributions in Virtual Water(TM) and liquid water. The source dose rate constant, radial dose function, anisotropy function and anisotropy factor have been determined following the updated AAPM TG-43 recommendations. The measured dose rate constant in the Virtual Water(TM) phantom was determined to be 0.727+/-6.9% cGyh(-1)U(-1), and the computed value is 0.716+/-2.1% cGyh(-1)U(-1). The Monte-Carlo simulation yielded a dose rate constant of 0.665+/-2.1% cGyh(-1)U(-1) in water. The measured dose rate constant in water is 0.675+/-7.5% cGyh(-1)U(-1). It is determined by multiplying the dose rate constant measured in the Virtual Water(TM) phantom with the ratio of the value calculated in water to that in Virtual Water(TM). The average of the measured and calculated dose rate constant is 0.670+/-5.5% cGyh(-1)U(-1). The radial dose functions of the new source were measured for distances ranging from 1 to 7 cm in a Virtual Water(TM) phantom. The anisotropy functions in Virtual Water(TM) phantom were measured for distances of 2, 3, 5, and 7 cm. The Monte-Carlo computed radial dose functions, anisotropy functions, and anisotropy factors in both Virtual Water(TM) phantom and water are reported.

Investigation of secondary neutron dose for 18 MV dynamic MLC IMRT delivery.

Secondary neutron doses from the delivery of 18 MV conventional and intensity modulated radiation therapy (IMRT) treatment plans were compared. IMRT was delivered using dynamic multileaf collimation (MLC). Additional measurements were made with static MLC using a primary collimated field size of 10 x 10 cm2 and MLC field sizes of 0 x 0, 5 x 5, and 10 x 10 cm2. Neutron spectra were measured and effective doses calculated. The IMRT treatment resulted in a higher neutron fluence and higher dose equivalent. These increases were approximately the ratio of the monitor units. The static MLC measurements were compared to Monte Carlo calculations. The actual component dimensions and materials for the Varian Clinac 2100/2300C including the MLC were modeled with MCNPX to compute the neutron fluence due to neutron production in and around the treatment head. There is excellent agreement between the calculated and measured neutron fluence for the collimated field size of 10 x 10 cm2 with the 0 x 0 cm2 MLC field. Most of the neutrons at the detector location for this geometry are directly from the accelerator head with a small contribution from room scatter. Future studies are needed to investigate the effect of different beam energies used in IMRT incorporating the effects of scattered photon dose as well as secondary neutron dose.

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.