In-phantom dosimetry and spectrometry of photoneutrons from an 18 MV linear accelerator

A combination of three superheated drop detectors with different neutron energy responses was developed to evaluate dose-equivalent and energy distributions of photoneutrons in a phantom irradiated by radiotherapy high-energy x-ray beams. One of the three detectors measures the total neutron dose equivalent and the other two measure the contributions from fast neutrons above 1 and 5.5 MeV, respectively. In order to test the new method, the neutron field produced by the 10 cm X 10 cm x-ray beam of an 18 MV radiotherapy accelerator was studied. Measurements were performed inside a tissue-equivalent liquid phantom, at depths of 1, 5, 10 and 15 cm and at lateral distances of 0, 10, and 20 cm from the central axis. These data were used to calculate the average integral dose to the radiotherapy patient from direct neutrons as well as from neutrons transmitted through the accelerator head. The characteristics of the dosimeters were confirmed by results in excellent agreement with those of prior studies. Track etch detectors were also used and provided an independent verification of the validity of this new technique. Within the primary beam, we measured a neutron entrance dose equivalent of 4.5 mSv per Gy of photons. It was observed that fast neutrons above 1 MeV deliver most of the total neutron dose along the beam axis. Their relative contribution increases with depth, from about 60% at the entrance to over 90% at a depth of 10 cm. Thus, the average energy increases with depth in the phantom as neutron spectra harden.     

Application of fission track detectors to californium-252 neutron dosimetry in tissue near the radiation source

Fission track detectors were applied to a unique problem in neutron dosimetry. Measurements of neutron doses were required at locations within a tumor of 1 cm diameter implanted on the back of a mouse and surrounded by a square array of four 252Cf medical sources. Measurements made in a tissue-equivalent mouse phantom showed that the neutron dose rate to the center of the tumor was 2.18 rads micrograms-1 h-1 +/- 8.4%. The spatial variation of neutron dose to the tumor ranged from 1.88 to 2.55 rads micrograms-1 h-1. These measurements agree with calculated values of neutron dose to those locations in the phantom. Fission track detectors have been found to be a reliable tool for neutron dosimetry for geometries in which one wishes to know neutron dose values which may vary considerably over distances of 1 cm or less.     

In vivo neutron dosimetry during high-energy Bremsstrahlung radiotherapy

PURPOSE: A new technique is presented for in vivo measurements of the dose equivalent from photoneutrons produced by high-energy radiotherapy accelerators. METHODS AND MATERIALS: The dosimeters used for this purpose are vials of superheated halocarbon droplets suspended in a tissue-equivalent gel. Neutron interactions nucleate the formation of bubbles, which can be recorded through the volume of gel they displace from the detector vials into graduated pipettes. These detectors offer inherent photon discrimination, dose-equivalent response to neutrons, passive operation, and small sensitive size. An in vivo vaginal probe was fabricated containing one of these neutron detector vials and a photon-sensitive diode. Measurements were carried out in patients undergoing high-energy x-ray radiotherapy and were also repeated in-phantom, under similar irradiation geometries. RESULTS AND CONCLUSION: Neutron doses of 0.02 Sv were measured in correspondence to the cervix, 50 cm from the photon beam axis, following a complete treatment course of 46.5 Gy with an upper mantle field of 18-MV x-rays. This fraction of dose from neutrons is measured reliably within an intense photon background, making the technique a valid solution to challenging dosimetry problems such as the determination of fetal exposure in radiotherapy. These measurements can be easily carried out with tissue-equivalent phantoms, as our results indicate an excellent correlation between in vivo and in-phantom dosimetry.     

Calibration of a prototype in vivo total body composition analyser using 14 MeV neutron activation and the associated particle technique

A prototype in vivo total body composition analyser has been constructed for determining the total body contents of nitrogen (TBN), carbon (TBC) and oxygen (TBO) in young experimental animals such as sheep or pigs by 14 MeV neutron activation using a commercially available associated particle sealed tube neutron generator (APSTNG). The instrument was calibrated by scanning phantoms of different sizes in the mass range 10-36 kg, filled with a mixture of elements as found in a normal human body. Good agreement was found between the measured and expected values of N, O and C when two phantoms of similar dimensions but of different composition were scanned. With four 15 cm x 15 cm cross section and 45 cm long NaI(T1) gamma detectors and a radiation dose of approximately 20 microSv due to neutrons, the expected precisions for a 28 kg animal, CV% (based on counting statistics) are N: 9.3, C: 2.3 and O: 1.4.     

The albedo effect on neutron transmission probability

Shields, used for protection against radiation, are often pierced with vacuum channels for passing cables and other instruments for measurements. The neutron transmission through these shields is an unavoidable phenomenon. In this work we study and discuss the effect of channels on neutron transmission through shields. We consider an infinite homogeneous slab, with a fixed thickness (20 lambda, with lambda the mean free path of the neutron in the slab), which contains a vacuum channel. This slab is irradiated with an infinite source of neutrons on the left side and on the other side (right side) many detectors with windows equal to 2 lambda are placed in order to evaluate the neutron transmission probabilities (Khanouchi, A., Aboubekr, A., Ghassoun, J. and Jehouani, A. (1994) Rencontre Nationale des Jeunes Chercheurs en Physique. Casa Blanca Maroc; Khanouchi, A., Sabir, A., Ghassoun, J. and Jehouani, A. (1995) Premier Congré International des Intéractions Rayonnements Matière. Eljadida Maroc). The neutron history within the slab is simulated by the Monte Carlo method (Booth, T. E. and Hendricks, J. S. (1994) Nuclear Technology 5) and using the exponential biasing technique in order to improve the Monte Carlo calculation (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500-504; Jehouani, A., Ghassoun, J. and Aboubker, A. (1994) In Proceedings of the 6th International Symposium on Radiation Physics, Rabat, Morocco). Then different geometries of the vacuum channel have been studied. For each geometry we have determined the detector response and calculated the neutron transmission probability for different detector positions. This neutron transmission probability presents a peak for the detectors placed in front of the vacuum channel. This study allowed us to clearly identify the neutron channeling phenomenon. One application of our study is to detect vacuum defects in materials.     

Monte Carlo dosimetry study of a 6 MV stereotactic radiosurgery unit

Small-field and stereotactic radiosurgery (SRS) dosimetry with radiation detectors, used for clinical practice, have often been questioned due to the lack of lateral electron equilibrium and uncertainty in beam energy. A dosimetry study was performed for a dedicated 6 MV SRS unit, capable of generating circular radiation fields with diameters of 1.25-5 cm at isocentre using the BEAM/EGS4 Monte Carlo code. With this code the accelerator was modelled for radiation fields with a diameter as small as 0.5 cm. The radiation fields and dosimetric characteristics (photon spectra, depth doses, lateral dose profiles and cone factors) in a water phantom were evaluated. The cone factor (St) for a specific cone c at depth d is defined as St(d, c) = D(d, c)/D(d, c(ref)), where c(ref) is the reference cone. To verify the Monte Carlo calculations, measurements were performed with detectors commonly used in SRS such as small-volume ion chambers, a diamond detector, TLDs and films. Results show that beam energies vary with cone diameter. For a 6 MV beam, the mean energies in water at the point of maximum dose for a 0.5 cm cone and a 5 cm cone are 2.05 MeV and 1.65 MeV respectively. The values of St obtained by the simulations are in good agreement with the results of the measurements for most detectors. When the lateral resolution of the detectors is taken into account, the results agree within a few per cent for most fields and detectors. The calculations showed a variation of St with depth in the water. Based on calculated electron spectra in water, the validity of the assumption that measured dose ratios are equal to measured detector readings was verified.     

Characteristics of neutron beam generated by 500 MeV proton beam

A neutron irradiation facility was constructed at PARMS, University of Tsukuba to produce an ultrahigh energy neutron beam with a depth dose distribution superior to an x-ray beam generated by a modern linac. This neutron beam was produced from the reaction on a thick uranium target struck by a 500 MeV proton beam from the booster synchrotron of the High Energy Physics Laboratory. The percentage depth dose of this neutron beam was nearly equivalent to that of x-rays around 20 MV and the dose rate was 15 cGy per minute. The relative biological effectiveness (RBE) of this neutron beam has been estimated using the cell inactivation effect and the HMV-I cell line. The survival curve of cells after neutron irradiation has a shoulder with n and Dq of 8 and 2.3 Gy, respectively. The RBE value at the 10(-2) survival level for the present neutron beam as compared with 137Cs gamma rays was 1.24. The results suggest that the biological effects of ultrahigh energy neutrons are not large enough to be useful, although the depth dose distribution of neutrons can be superior to that of high energy linac x-rays.     

Monte Carlo calculation of dose enhancement by neutron capture of 10B in fast neutron therapy

Since 1978 the Essen Medical Cyclotron Facility has been used for fast neutron therapy. The treatment of deep-seated tumours by d(14) + Be neutron beam therapy (mean energy = 5.8 MeV) is still limited because of the steep decrease in depth-dose distribution. The interactions of fast neutrons in tissue leads to a thermal neutron distribution. These partially thermalized neutrons can be used to produce neutron capture reactions with 10B. Thus incorporation of 10B in tumours treated with fast neutrons will increase the relative local tumour dose due to the reaction 10B (n, alpha) 7Li. The magnitude of dose enhancement by 10B depends on the distribution of the thermal neutron fluence, 10B concentration, field size of the neutron beam, beam energy and the specific phantom geometry. The slowing down of the fast neutrons, resulting in a thermal neutron distribution in a phantom, has been computed using a Monte Carlo model. This model, which includes a deep-seated tumour, was experimentally verified by measurements of the thermal neutron fluence rate in a phantom using neutron activation of gold foil. When non-boronated water phantoms were irradiated with a total dose of 1 Gy at a depth of 6 cm, the thermal fluencies at this depth were found to be 2 x 10(10) cm-2. The absorbed dose in a tumour with 100 ppm 10B, at the same depth, was enhanced by 15%.     

Chromosome aberrations in human fibroblasts induced by monoenergetic neutrons. I. Relative biological effectiveness

The relative biological effectiveness (RBE) of neutrons for many biological end points varies with neutron energy. To test the hypothesis that the RBE of neutrons varies with respect to their energy for chromosome aberrations in a cell system that does not face interphase death, we studied the yield of chromosome aberrations induced by monoenergetic neutrons in normal human fibroblasts at the first mitosis postirradiation. Monoenergetic neutrons at 0.22, 0.34, 0.43, 1, 5.9 and 13.6 MeV were generated at the Accelerator Facility of the Center for Radiological Research, Columbia University, and were used to irradiate plateau-phase fibroblasts at low absorbed doses from 0.3 to 1.2 Gy at a low dose rate. The reference low-LET, low-dose-rate radiation was 137Cs-gamma rays (0.66 MeV). A linear dose response (Y = alphaD) for chromosome aberrations was obtained for all monoenergetic neutrons and for the gamma rays. The yield of chromosome aberrations per unit dose was high at low neutron energies (0.22, 0.34 and 0.43 MeV) with a gradual decline with the increase in neutron energy. Maximum RBE (RBEm) values varied for the different types of chromosome aberrations. The highest RBE (24.3) for 0.22 and 0.43 MeV neutrons was observed for intrachromosomal deletions, a category of chromosomal change common in solid tumors. Even for the 13.6 MeV neutrons the RBEm (11.1) exceeded 10. These results show that the RBE of neutrons varies with neutron energy and that RBEs are dissimilar between different types of asymmetric chromosome aberrations and suggest that the radiation weighting factors applicable to low-energy neutrons need firmer delineation. This latter may best be attained with neutrons of well-defined energies. This would enable integrations of appropriate quality factors with measured radiation fields, such as those in high-altitude Earth atmosphere. The introduction of commercial flights at high altitude could result in many more individuals being exposed to neutrons than occurs in terrestrial workers, emphasizing the necessity for better-defined estimates of risk.     

Synchrotron radiation in the study of the variation of dose response in thermoluminescence dosimeters with radiation energy

Thermoluminescence dosimetry (TLD) is a versatile technique with many applications for dosimetry of ionising radiation. However, in the range of kilovoltage x-rays which is widely used for diagnostic and therapeutic medical applications, problems arise from the differing dose response of most TL dosimeters with the radiation energy. The dose response of various TL detector types was investigated in mono-energetic x-ray beams of 26.8, 33.2, 40, 80.4 and 99.6keV from a synchrotron radiation source at the National Laboratory for High Energy Physics in Japan. This response was studied as a function of TL material (LiF:Mg,Ti, LiF:Mg,Cu,P and Al2O3), the detector geometry and size, and their thermal history. Due to the asymmetric diffraction from a Si crystal employed to produce monoenergetic photons there was more than 50% dose inhomogeneity in some of radiation fields used. Therefore, the different TL dosimeter types were rotated around and the results related to the reading of a set of "standard" LiF:Mg,Ti ribbons which were included in all experiments as reference detectors. No significant influence of the detector shape (physical size, thickness) on the dose response with energy could be found. However, the pre-irradiation thermal history influences the dose response with radiation energy: a fast cool down of LiF:Mg,Ti after a high temperature anneal will increase the sensitivity by more than a factor of two. The relatively new TLD material LiF:Mg,Cu,P (GR-200, obtained from Solid Dosimeter & Detector Laboratories, Beijing) was found to be approximately 100 times more sensitive than the standard LiF:Mg,Ti. In addition it proved to be more tissue equivalent for photon radiation between 27keV and 40keV. The performance of LiF:Mg,Cu,P makes it a very interesting TL material deserving further evaluation for applications in diagnostic and therapeutic x-rays.     

高剂量氦离子辐照对新型中子增殖铍钨合金表面结构的影响

为了确保未来核聚变反应堆的氘氚自持燃烧必需采用中子增殖材料来得到合适的氚增值比。金属铍被认为是最有前途的核聚变反应堆固态中子倍增材料,但其熔点低,高温抗辐照肿胀性能差,因此需要寻找和研发具有更高熔点和更耐辐照肿胀的新型中子增殖材料以满足更先进的聚变堆要求。本研究尝试提出并制备了一种更高熔点的铍钨合金（Be₁₂W）,通过X射线和扫描电子显微镜对它的相组成和表面结构进行分析。对新型铍钨合金进行高剂量的氦离子辐照,发现合金表面一次起泡的平均尺寸约为0.8 μm,面密度约为2.4×107 cm?2,而二次起泡的平均尺寸约为80 nm,面密度约为1.28×108 cm?2。分析氦辐照引起的表面起泡及其机制,并与纯铍和铍钛合金表面起泡的情况进行了对比。      

Neutron RBEs for cytopenia and repopulation of stroma and hematopoietic stem cells: mathematical models of marrow cell kinetics

The objectives of this study were to (a) extend previous bone-marrow cell kinetics models that have been published for ionizing photons to include neutron radiations, and (b) provide Relative Biological Effectiveness (RBE) values for time-specific cell killing (cytopenia) and compensatory cellular proliferation (repopulation in response to toxic injury) for neutron doses ranging from 0.01 to 4.5 Gy delivered uniformly over one minute, hour, day, week, and month. RBEs for cytopenia of a cell lineage were based on ratios of protocol-specific doses that determined the same cell population nadir, whereas the RBEs for repopulation of a lineage were based on the ratios of protocol-specific doses that corresponded to the same total number of cells killed over the radiation treatments, and which should be replaced for long-term survival of the animal. Time-dependent RBEs were computed for neutron exposures relative to the effect of 60Co gamma rays given as a prompt dose. By the use of these RBE factors, low or variable dose rates, dose fractionations given over long periods of time, and different protocols involving several radiation qualities can be converted realistically, and by standard convention, into an equivalent dose of a reference radiation comprised of x or gamma rays given either as a pulse or at any other reference dose rate for which risk information based on epidemiological or animal dose-response data are available. For stromal tissues irradiated by fission neutrons, time-dependent RBEs for cytopenia were computed to range from 4.24 to 0.70 and RBEs for repopulation varied from a high of 6.88 to a low of 2.24. For hematopoietic stem cells irradiated by fission neutrons, time-dependent RBEs for cytopenia were computed to range from 5.02 to 0.22 and RBEs for repopulation varied from a high of 5.02 to a low of 1.98. RBEs based on tissue-kerma-free-in-air would be about twofold lower for isotropic cloud or rotational exposure geometries because marrow dose from isotropic neutron fields suffer factor-of-two greater attenuation than the gamma doses from gamma photons. For certain doses and dose rates, the RBE values computed for compensatory cellular proliferation clearly demonstrate the behavior that is commonly referred to as an inverse dose-rate effect, i.e., protraction of exposure may-under certain conditions-increase the magnitude of the dose response. Furthermore, because of non-linear rates for repair and repopulation, the highest RBEs are not necessarily found for the lowest doses nor the lowest RBEs always found at the highest doses.     

Assessment of exposure rate and collective effective dose equivalent in the city of Baghdad due to natural gamma radiation

Exposure rate measurements in the city of Baghdad were initiated in June 1981. The average exposure rate, the average annual effective dose equivalent and the collective effective dose equivalent are assessed. The data are presented according to the eleven municipal divisions of the city. The average exposure rate was 6.9 microR h-1 and an average annual effective dose equivalent of 4.5 x 10(-4) Sv year-1. The collective effective dose equivalent is 1.81 x 10(3) man Sv year-1.     

Chemical control of breathing in identical twin athletes

The dosimetric characteristics of the electron beams from the Sagittaire (C.G.R. MeV) linear accelerator are presented. The variations in dose output with the distance from the source are indicated for the usual treatment distances. The measurements carried out demonstrating the deviations from the inverse square law for distances. The graphs for the variation in dose output in terms of the opening of the collimator are given for square and rectangular openings. The measurements demonstrate the preponderant influence of the short side 1 of the rectangular field of the relative variation in dose output. The variations in the graphs of the deep yield in terms of beam energy, distance from the source and field dimension were studied. For the different energies, the influence of the distance from the collimator and the dimension of the field on the dose at entry and on the depth of the maximum was shown. Measurements of the deep yields were carried out in the case of rectangular fields. In each case, the side of the equivalent square field corresponding to the same deep yield as a 1 X L rectangular field was determined; a comparison was carried out with the equivalent square field defined for colbat radiation.     

Neutron attenuation characteristics of polyethylene, polyvinyl chloride, and heavy aggregate concrete and mortars

Polyethylene and polyvinyl chloride pellets were introduced into concrete to improve its neutron attenuation characteristics while several types of heavy coarse aggregates were used to improve its gamma ray attenuation properties. Neutron and gamma ray attenuation were studied in concrete samples containing coarse aggregates of barite, pyrite, basalt, hematite, and marble as well as polyethylene and polyvinyl chloride pellets in narrow-beam geometry. The highest neutron attenuation was shown by polyethylene mortar, followed by polyvinyl chloride mortar; barite and pyrite concrete showed higher gamma ray attenuation than ordinary concrete. Broad-beam and continuous (infinite) medium geometries were used to study the neutron attenuation of samples containing polymers at different concentrations with and without heavy aggregates, the fitting equations were established, and from these the neutron removal coefficients were deduced. In a radiation field of neutrons and gamma rays, the appropriate concentration of polymer and heavy aggregate can be selected to give the optimum total dose attenuation depending on the relative intensities of each type of radiation. This would give much better design flexibility over ordinary concrete. The compressive strength tests performed on mortar and concrete samples showed that their value, in general, decreases as polymer concentration increases and that the polyvinyl chloride mortar showed higher values than the polyethylene mortar. For general construction purposes, the compression strength was considered acceptable in these samples.     

Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system

The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.     

Examining Duct Leakage on Prestressed Concrete Bridges via a Multiple Neutron Sources Method

Tendon corrosion and the leakage of water through the grouting voids are important contributors to the degradation of prestressed concrete (PC) bridges. Therefore, leakage inspections are beneficial in determining whether a tendon is corroding. This work addresses the inspection of duct leakages on PC bridges using a special multiple neutron source method. This approach is based on the principle of elastic collision between fast neutrons and hydrogen atoms during the emergence of thermalization. Multiple neutron sources should be combined with multiple detectors and an appropriately extended detection time. The probability of capturing thermal neutrons can thus be increased to inspect PC duct leakage. An equation was derived from a combination of theoretical studies and experimental outcomes as a reference for properly selecting the numbers of neutron sources and detectors as well as the detection time. The experimental results show that this approach increases the detection depth of leakage within concrete.     

Linear B-cell epitopes in the N-terminus of L2 of bovine papillomavirus type 4

In vivo dosimetry performed with semiconductor detectors is a reliable method for patient dose control. The purpose of this study is to evaluate the perturbations introduced in the patient's absorbed dose distribution by three types of commercially available diodes (Isorad, Sun Nuclear Corp.; model 114200, 114300 and 114400) from the same company and to present possible solutions for minimizing this side-effect.     

Combined use of FLUKA and MCNP-4A for the Monte Carlo simulation of the dosimetry of 10B neutron capture enhancement of fast neutron irradiations

Boron neutron capture enhancement (BNCE) of the fast neutron irradiations use thermal neutrons produced in depth of the tissues to generate neutron capture reactions on 10B within tumor cells. The dose enhancement is correlated to the 10B concentration and to thermal neutron flux measured in the depth of the tissues, and in this paper we demonstrate the feasibility of Monte Carlo simulation to study the dosimetry of BNCE. The charged particle FLUKA code has been used to calculate the primary neutron yield from the beryllium target, while MCNP-4A has been used for the transport of these neutrons in the geometry of the Biomedical Cyclotron of Nice. The fast neutron spectrum and dose deposition, the thermal flux and thermal neutron spectrum in depth of a Plexiglas phantom has been calculated. The thermal neutron flux has been compared with experimental results determined with calibrated thermoluminescent dosimeters (TLD-600 and TLD-700, respectively, doped with 6Li or 7Li). The theoretical results were in good agreement with the experimental results: the thermal neutron flux was calculated at 10.3 X 10(6) n/cm2 s1 and measured at 9.42 X 10(6) n/cm2 s1 at 4 cm depth of the phantom and with a 10 cm X 10 cm irradiation field. For fast neutron dose deposition the calculated and experimental curves have the same slope but different shape: only the experimental curve shows a maximum at 2.27 cm depth corresponding to the build-up. The difference is due to the Monte Carlo simulation which does not follow the secondary particles. Finally, a dose enhancement of, respectively, 4.6% and 10.4% are found for 10 cm X 10 cm or 20 cm X 20 cm fields, provided that 100 micrograms/g of 10B is loaded in the tissues. It is anticipated that this calculation method may be used to improve BNCE of fast neutron irradiations through collimation modifications.     

Conversion of ionization measurements to radiation absorbed dose in non-water density material

The radiation absorbed dose to non-water equivalent materials of interest in radiotherapy is the dose to lung and the dose to bone. The measurement and calculation of dose to the lung has been of great interest and much effort has gone into the development of accurate lung dose calculation methods. The radiation absorbed dose to the bone is usually not calculated and most absorbed dose calculations have been done without correcting for the presence of bone. For the lower megavoltage photon beams this may be appropriate, however, as the energy of the photon beam increases, the region of electronic disequilibrium becomes larger and pair production which depends on the atomic number of the material becomes significant. Therefore the bone will produce greater perturbations of the dose distribution. The dose to lung-equivalent material is uniquely obtained from ionization measurements. However, in bone-equivalent materials two different calculations of absorbed dose are possible: the absorbed dose to soft tissue plastic (polystyrene) within bone-equivalent material and the dose to the bone-equivalent material itself. Both can be calculated from ionization measurements in phantoms. These two calculations result in significantly different doses in a heterogeneous phantom composed of polystyrene and aluminium (a bone substitute). The dose to a thin slab of polystyrene in aluminium is much higher than the dose to the aluminium itself at the same depth in the aluminium. Monte Carlo calculations confirm that the calculation of dose to polystyrene in aluminium can be accurately carried out using existing dosimetry protocols. However, the conversion of ionization measurements to absorbed dose to high atomic number materials cannot be accurately carried out with existing protocols and appropriate conversion factors need to be determined.