Fast Neutron Dose Research Articles

Validation of dose planning calculations for boron neutron capture therapy using cylindrical and anthropomorphic phantoms

In this paper, the accuracy of dose planning calculations for boron neutron capture therapy (BNCT) of brain and head and neck cancer was studied at the FiR 1 epithermal neutron beam. A cylindrical water phantom and an anthropomorphic head phantom were applied with two beam aperture-to-surface distances (ASD). The calculations using the simulation environment for radiation application (SERA) treatment planning system were compared to neutron activation measurements with Au and Mn foils, photon dose measurements with an ionization chamber and the reference simulations with the MCNP5 code. Photon dose calculations using SERA differ from the ionization chamber measurements by 2–13% (disagreement increased along the depth in the phantom), but are in agreement with the MCNP5 calculations within 2%. The 55Mn(n,γ) and 197Au(n,γ) reaction rates calculated using SERA agree within 10% and 8%, respectively, with the measurements and within 5% with the MCNP5 calculations at depths >0.5 cm from the phantom surface. The 55Mn(n,γ) reaction rate represents the nitrogen and boron depth dose within 1%. Discrepancy in the SERA fast neutron dose calculation (of up to 37%) is corrected if the biased fast neutron dose calculation option is not applied. Reduced voxel cell size (⩽0.5 cm) improves the SERA calculation accuracy on the phantom surface. Despite the slight overestimation of the epithermal neutrons and underestimation of the thermal neutrons in the beam model, neutron calculation accuracy with the SERA system is sufficient for reliable BNCT treatment planning with the two studied treatment distances. The discrepancy between measured and calculated photon dose remains unsatisfactorily high for depths >6 cm from the phantom surface. Increasing discrepancy along the phantom depth is expected to be caused by the inaccurately determined effective point of the ionization chamber.

Fast-neutron dose evaluation in BNCT with Fricke gel layer detectors

Abstract Boron neutron capture therapy (BNCT) is a cancer radiotherapy that uses epithermal and thermal neutron beams. The determination of the absorbed dose in healthy tissue, separating the various dose contributions having different radiobiological effectiveness (RBE) is of great importance for therapy planning. However, a standard code of practice has not yet been established because suitable methods for dosimetry in BNCT are still in progress. A study about the characterization of the epithermal column of the LVR-15 research reactor in Řež (CZ) has been performed, in particular concerning the fast-neutron dose. This dose is not negligible and its determination is important owing to its high RBE. Fast-neutron and photon dose distributions in a water phantom have been measured by means of Fricke gel layer dosimeters. Even if gel layer dosimetry is not yet standardized, it is presently the only method for obtaining images of each dose contribution in BNCT neutron fields. The results were compared with values measured with thermoluminescence detectors, twin ionization chambers data taken from literature and Monte Carlo simulations.

Study of the improvement of TLD cards for personal neutron dosimetry

In this work, personal thermoluminescence dosimeter (TLD) cards type of GN-6770 (holder type 8806) from Harshaw were used for personal neutron dosimetry. The response of the dosimeters has been determined in terms of the personal absorbed dose and personal dose equivalent for different neutron energy components, based on the recommendations of ICRP-60 and ICRU-49. Neutron irradiation was performed using a 5 mCi Am–Be neutron source. The TLD reader, type Harshaw 6600, was installed and calibrated for accurate neutron doses equivalent to gamma-ray doses. It was found that fast neutron doses measured by TLD (badges or cards) are in agreement with those measured by neutron TE (tissue equivalent gas) ionization chambers and neutron monitors. Thermal neutron doses measured by TLD cards were overestimated when compared with those measured by neutron monitors. Additional Cd was used to reduce thermal neutron doses to be in agreement with actual thermal doses. Other configurations for TLD crystals are also suggested for accurate thermal neutron dose measurements.

Performance optimization of the CANDLE reactor for nuclear energy sustainability

Constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing (CANDLE) fast reactors show the following outstanding merits: (1) simple and safe, (2) high performance to nuclear nonproliferation and safeguard, (3) small volume of waste, (4) efficient fuel use. However, CANDLE reactor has two inherent problems: (1) destruction of cladding integrity caused by high fast neutron dose and high accumulated FP gas pressure and (2) longer core height caused by the shape of burning region and shift of the region along burning. In the present paper cladding integrity is tried to be kept by employing recladding and the core height is tried to be shorten by employing multi-channel once-through-then-out (MOTTO) cycle. MOTTO cycle adjusts the amount of discharged fuel for each pin to make the axial power peak align perpendicular to the core axis at the beginning of each cycle operation. We selected a LBE cooled fast reactor with nitride fuel for the present study. Thermal power rate is 1540 MW. The core-plus-blanket diameter is 4 m, and the height is minimized to be 1.6 m. It is similar to familiar pancake-type fast reactor designs and considered to be economically feasible together with other many cost effective characteristics.

Novel fluorescent nuclear track detector technology for mixed neutron-gamma fields

Abstract The latest achievements in fluorescent nuclear track detector (FNTD) technology are described. FNTDs are aluminum oxide crystals containing aggregate oxygen vacancy defects and doped with carbon and magnesium (Al 2 O 3 :C,Mg). Unlike most nuclear track detectors, Al 2 O 3 :C,Mg is sensitive to low linear energy transfer (LET) radiation including secondary electrons resulting from interactions of photons with the crystal. A new image processing method is investigated as a technique to discriminate and measure the doses of gamma and fast neutrons in mixed field conditions. Dose dependencies for both gamma and neutron irradiated FNTDs are shown. The new image processing method increased the dynamic range of detectable neutron doses from 4 orders of magnitude for track counting method to at least 6 orders of magnitude by combining track counting with the new image processing method. The new image processing method is combined with a detector configuration utilizing three converters: Teflon ® , polyethylene, and lithium fluoride.

Radiation protection measurements around a 12 MeV mobile dedicated IORT accelerator

The aim of this study is to investigate radioprotection issues that must be addressed when dedicated accelerators for intraoperative radiotherapy (IORT) are used in operating rooms. Recently, a new version of a mobile IORT accelerator (LIAC Sordina SpA, Italy) with 12 MeV electron beam has been implemented. This energy is necessary in some specific pathology treatments to allow a better coverage of thick lesions. At an electron energy of 10 MeV, leakage and scattered x-ray radiation (stray radiation) coming from the accelerator device and patient must be considered. If the energy is greater than 10 MeV, the x-ray component will increase; however, the most meaningful change should be the addition of neutron background. Therefore, radiation exposure of personnel during the IORT procedure needs to be carefully evaluated. In this study, stray x-ray radiation was measured and characterized in a series of spherical projections by means of an ion chamber survey meter. To simulate the patient during all measurements, a polymethylmethacrylate (PMMA) slab phantom with volume 30 x 30 x 15 cm3 and density 1.19 g / cm3 was used. The PMMA phantom was placed along the central axis of the beam in order to absorb the electron beams and the tenth value layer (TVL) and half value layer (HVL) of scattered radiation (at 0 degrees, 90 degrees, and 180 degrees scattering angles) were also measured at 1 m of distance from the phantom center. Neutron measurements were performed using passive bubble dosimeters and a neutron probe, specially designed to evaluate ambient dose equivalent H*(10). The x-ray equivalent dose measured at 1 m along the beam axis at 12 MeV was 260 microSv/Gy. The value measured at 1 m at 90 degrees scattering angle was 25 microSv/Gy. The HVL and TVL values were 1.1 and 3.5 cm of lead at 0 degrees, and 0.4 and 1 cm at 90 degrees, respectively. The highest equivalent dose of fast neutrons was found to be at the surface of the phantom on the central beam axis (2.9 +/- 0.6 microSv/Gy), while a lower value was observed below the phantom (1.6 +/- 0.3 microSv/Gy). The neutron dose equivalent at 90 degrees scattering angle and on the floor plane on the beam axis below the beam stopper was negligible. Our data confirm that neutron exposure levels around the new dedicated IORT accelerator are very low. Mobile shielding panels can be used to reduce x-ray levels to below regulatory levels without necessarily providing permanent shielding in the operating room.

Increase in Irradiation Beam Intensity by Using a Hybrid Target System in Cyclotron-Based Neutron Capture Therapy

A cyclotron-based epithermal neutron generator has been expected to achieve the early realization of hospital-based boron neutron capture therapy. The system using a hybrid target composed of a thin tungsten (W) layer and a thick beryllium (Be) one with 30MeV protons and a beam-shaping assembly was previously proposed as a better generator candidate than the system using only a thick W or Be layer. In this study, to clearly show that the hybrid target system is more clinically useful than the other two, all three target systems were compared using epithermal neutron intensity as a figure of merit under an equal condition of the harmful effect of fast neutron dose. We found that the hybrid system using a W layer of appropriate thickness increased the epithermal neutron intensity by 107 and 32% compared with systems using only W or Be, respectively, with the same harmful fast neutron component. It is concluded that the hybrid target system is more useful in terms of epithermal neutron intensity than systems with either W or Be as a generator using the 30MeV proton accelerator and beam-shaping assembly.

Special features of the mechanism of defect formation in CdS single crystals subjected to irradiation with high doses of fast reactor neutrons

Electrical and optical properties of CdS single crystals irradiated with a dose ≥1018 cm−2 of fast reactor neutrons have been studied. It is established that clusters of defects are formed in irradiated material, and cadmium vacancies are dominant in these clusters. In the case when the defect clusters are decomposed in the course of radiation-stimulated annealing or thermal annealing in the temperature range Φ ≈ (200–400)°C, the crystal lattice becomes enriched with Cd vacancies. It is assumed that subthreshold effects play an important role in formation of defect clusters, and that these effects are related to preferential excitation of the K shells in Cd atoms and their Coulomb ejection from the core of a cluster.

Mechanical properties and microstructural stability of 11Cr-ferritic/martensitic steel cladding under irradiation

The in-reactor creep rupture tests of 11Cr–0.5Mo–2W, V, Nb F/M steel were carried out in the temperature range from 823 to 943 K using materials open test assembly in the fast flux test facility and tensile and temperature-transient-to-burst specimens were irradiated in the experimental fast reactor JOYO at temperatures between 693 and 1013 K to fast neutron doses ranging from 11 to 102 dpa. The results of post-irradiation mechanical tests showed that there was no significant degradation in tensile and transient burst strengths even after neutron irradiation below 873 K, but that there was significant degradation in both strengths at neutron irradiation above 903 K. On the other hand, the in-reactor creep rupture times were equal or greater than those of out-reactor creep even after neutron irradiation at all temperatures. This creep rupture behavior was different from that of tensile and transient burst specimens.

Comparative study of dose calculations with SERA and JCDS treatment planning systems

Three treatment planning systems developed for clinical boron neutron capture therapy (BNCT) use are SERA developed by INL/Montana State University, NCTPlan developed by the Harvard–MIT and the CNEA group and JAEA computational dosimetry system (JCDS) developed by Japan Atomic Energy Agency (JAEA) in Japan. Previously, performance of the SERA and NCTPlan has been compared in various studies. In this preliminary study, the dose calculations performed with SERA and JCDS systems were compared in single brain cancer patient case with the FiR 1 epithermal neutron beam. A two-field brain cancer treatment plan was performed with the both codes. The dose components to normal brain, tumor and planning target volume (PTV) were calculated and compared in case of one radiation field and combined two fields. The depth dose distributions and the maximum doses in regions of interest were compared. Calculations with the treatment planning systems for the thermal neutron induced ( 10B and nitrogen) dose components and photon dose were in good agreement. Higher discrepancy in the fast neutron dose calculations was found. In case of combined two-field treatment plan, overall discrepancy of the maximum weighted dose was ∼3% for normal brain and PTV and ∼4% for tumor dose.

Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: Application to the treatment of experimental oral cancer

The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1×10 9 n cm −2 s −1 and the fast neutron flux was 2.5×10 6 n cm −2 s −1, indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in 6Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

A virtual model of the patient's head for BNCT

A virtual model of the patient's head for BNCT The aim of the present work was creating a virtual phantom of a human head for BNCT, as a part of the BNCT programme project. This model is an amplification of the simple model described in earlier publications. It takes into account the major head organs as well as the scalp and skull. The chemical composition of all tissues was modelled according to the recommendations of the ICRP. The organs were parameterized using mathematical formulas based on the human head magnetic resonance images. The model was used for calculating the thermal neutron flux and the injuring (fast neutron, nitrogen and gamma) dose components for the head irradiated using the therapeutic neutron beam, whose parameters were obtained as the result of the modelling of the filter/moderator system for the BNCT therapeutic beam from the MARIA reactor.

Nuclear reactions induced by deuterons and their applicability to skin tumor treatment through BNCT

In this work the D(d,n) 3He and 9Be(d,n) 10B reactions have been studied in a low-energy regime as neutron sources for skin tumor treatment in the frame of accelerator-based BNCT (AB-BNCT). The total neutron production and the energy and angular distributions for each reaction at different bombarding energies and for the thick targets considered (TiD 2, Be) have been determined using the available data in the literature. From this information, a feasibility study has been performed by means of MCNP simulations. The thermal, epithermal and fast neutron fluxes and doses at skin tumor positions (loaded with 40 ppm 10B) which are located on a whole-body human phantom have been simulated for different D 2O moderator depths. The best-case performance shows that a high tumor control probability (TCP) of 99% corresponding to a weighted dose in tumor of 40 Gy can be reached at the tumor position keeping the weighted dose in healthy tissue below 12.5 Gy, by means of the 9Be(d,n) 10B reaction at 1.1 MeV for a deuteron current of 20 mA and a 30 cm D 2O moderator in 52 min. The availability of low-energy neutrons in the 9Be(d,n) 10B reaction from the population of excited levels between 5.1 to 5.2 MeV in 10B and the convenience of a thin beryllium target are discussed. As a complement concerning alternatives to the Li(metal) + p reaction, the neutron yield of refractory lithium compounds (LiH, Li 3N and Li 2O) were calculated and compared with a Li metal target.

Development of a Young’s modulus model for Gilsocarbon graphites irradiated in inert environments

Dimensional and material property changes caused by fast neutron damage, radiolytic oxidation or a combination of both have been measured over many years on the Gilsocarbon graphites. The data have been gathered together in a comprehensive database and are being used to develop dose–damage relationships of irradiated graphite material properties for use in irradiated graphite component stress analyses up to a fast neutron dose of ∼200 × 10 20 n/cm 2 equivalent DIDO nickel dose (EDND) in the temperature range ∼300–650 °C. This paper covers development and validation of an empirical model for Young’s modulus of Gilsocarbon graphite when irradiated in an inert environment. The new model provides a good fit over the range of the primary modelling variables: dose, irradiation temperature, and graphite group. The proposed model is convenient to use in component stress analysis, as it features an analytic function of temperature and dose that eliminates the need in the existing model for interpolation on those variables.

Accumulation of thermal resistance in neutron irradiated graphite materials

A nuclear graphite, H451, and two high thermal conductivity graphite composites have been irradiated in the temperature range of 310–710 °C in the high flux isotope reactor and their thermal conductivities monitored in situ. Data were measured continuously up to a fast neutron dose of approximately 1 × 10 25 n/m 2 ( E > 0.1 MeV). Data are interpreted in terms of the added thermal resistance and materials compared on this basis. Following this analysis it is shown that for the three materials studied, which have significantly different initial thermal conductivity values, the accumulation of thermal resistance is greater for the materials with lower initial thermal conductivity. Given that vacancies dominate phonon scattering at these irradiation temperatures and dose levels, these data clearly indicate that materials of higher perfection have a slower rate of stable vacancy accumulation during irradiation.

Dimensional and material property changes to irradiated Gilsocarbon graphite irradiated between 650 and 750 °C

Samples of Gilsocarbon graphite have been recently irradiated in the High Flux Reactor at Petten. They were irradiated in an inert atmosphere to a fast neutron dose of between 3 and 10 dpa, at temperatures of between 650 and 750 °C, this irradiation dose equates to approximately dimensional change ‘turnaround’. The present understanding of the irradiation induced microstructural changes considered to lead to the irradiation induced dimensional and material property changes are discussed and correlations between the various property changes noted. However, at present there are few microstructural studies to confirm or refute these theories. A methodology proposed to convert graphite CTE measurement, for all types graphite, from one temperature range to another is shown to be applicable to both unirradiated and irradiated Gilsocarbon graphite. In addition an important relationship for the thermal conductivity temperature dependence of irradiated Gilsocarbon graphite is shown to be reasonable.

Development of a Brazilian gamma–neutron dosimeter

Aiming to improve the neutron dosimetry in Brazil a gamma–neutron mixed field dosimeter using the techniques of TLAD and SSNTD was developed. Commercially available materials were preferentially used. The dosimeter was projected to evaluate gamma, intermediate (Albedo) and fast neutron doses from 241AmBe sources. Dosimetric properties such as gamma and neutron doses and angular responses were evaluated.

Monte Carlo calculations of epithermal and fast neutron dose in a human head model for Boron Neutron Capture Therapy

Monte Carlo calculations of epithermal and fast neutron dose in a human head model for Boron Neutron Capture Therapy Boron Neutron Capture Therapy is a very promising form of cancer therapy, consisting in irradiating a stable isotope of boron (10B) concentrated in tumor cells with a low energy neutron beam. This technique makes it possible to destroy tumor cells, leaving healthy tissues practically unaffected. In order to carry out the therapy in the proper way, the proper range of the neutron beam energy has to be chosen. In this paper we continue the earlier started calculations of the optimum energy range for BNCT, taking into account the absorbed dose from fast neutrons.

Evaluation of D(d,n) 3 He reaction neutron source models for BNCT irradiation system design

A mathematical method was developed to calculate the yield, energy spectrum and angular distribution of neutrons from D(d,n) 3 He (D-D) reaction in a thick deuterium-titanium target for incident deuterons in energies lower than 1.0MeV. The data of energy spectrum and angular distribution were applied to set up the neutron source model for the beam-shaping-assembly (BSA) design of Boron-Neutron-Capture-Therapy (BNCT) using MCNP-4C code. Three cases of D-D neutron source corresponding to incident deuteron energy of 1000, 400 and 150 keV were investigated. The neutron beam characteristics were compared with the model of a 2.45 MeV mono-energetic and isotropic neutron source using an example BSA designed for BNCT irradiation. The results show significant differences in the neutron beam characteristics, particularly the fast neutron component and fast neutron dose in air, between the non-isotropic neutron source model and the 2.5 MeV mono-energetic and isotropic neutron source model.

A preliminary study on the use of 10Be in forensic radioecology of nuclear explosion sites

Cosmogenic 10Be, known for use in dating studies, unexpectedly is also produced in nuclear explosions with an atom yield almost comparable to (e.g.) 137Cs. There are major production routes via 13C( n, α) 10Be, from carbon dioxide in the air and the organic explosives, possibly from other bomb components and to a minor extent from the direct fission reaction. Although the detailed bomb components are speculative, carbon was certainly present in the explosives and an order of magnitude calculation is possible. The ( n, α) cross-section was determined by irradiating graphite in a nuclear reactor, and the resulting 10Be estimated by Accelerator Mass Spectrometry (AMS) giving a cross-section of 34.5 ± 0.7 mb (6–9.3 MeV), within error of previous work. 10Be should have applications in forensic radioecology. Historical environmental samples from Hiroshima, and Semipalatinsk (Kazakhstan) showed two to threefold 10Be excesses compared with the background cosmogenic levels. A sample from Lake Chagan (a Soviet nuclear cratering experiment) contained more 10Be than previously reported soils. 10Be may be useful for measuring the fast neutron dose near the Hiroshima bomb hypocenter at neutron energies double those previously available.