Neutron Kerma Factors Research Articles

Gadolinium neutron capture therapy: Calculation of 157Gd kerma factor and a dosimetry study using MCNP Monte Carlo code

In this paper, computational dosimetry studies for gadolinium-157 neutron capture therapy (157GdNCT) were carried out using MCNP Monte Carlo code, and the total equivalent dose rate was calculated in a Snyder head phantom with convenient and more accurate methods compared to the previous methods. The potential effectiveness of such therapy was assessed by calculating the therapeutic ratio based on the calculated dose distribution.The total neutron kerma factor for 157Gd was calculated according to the latest nuclear data. The 157Gd dose was calculated by kerma approximation and also directly by 157Gd capture reactions. The kerma of 157Gd photons was calculated in tumor and compared with the photon dose calculated from the transport of the produced electrons.The results reveal that, due to the severe neutron flux depression and high photon production in the tumor, the kerma approximation does not properly estimate the 157Gd dose and dose of 157Gd photons in a small-size tumor. Increasing the size of the tumor, the 157Gd kerma without considering neutron flux depression and photon kerma approach the dose of 157Gd and photons, respectively. If 157Gd is incorporated into the nucleus of the cancer cells, the therapeutic ratio of 157GdNCT will be greater than one and 157Gd might, therefore, be a good candidate for neutron capture therapy.

Study of neutron capture reaction-induced displacement damage cross section and KERMA factor

Radiative neutron capture reaction is generally the main reaction inducing atomic displacements and nuclear heating for low incident neutron energy, even though its contribution to total displacement damage is quite limited in general. The multiple photon emission without identifying the correlations between photons leads the calculated recoil energy to be uncertain. The present work recommends a formula for computing the average recoil energy by assuming the isotropic photon emissions, while the upper limit has been implemented in some codes. The recommended method is identical to the one used in NJOY if the photon distributions are in MF12-MF15, whereas the photon yield is not included in NJOY when photon yields are included in MF6. The presently recommended method for KERMA factor calculation is numerically validated against Tripoli-4 neutron-photon coupling Monte Carlo simulations. Due to the negligible contribution of neutron kick, the correction of the kinematic KERMA factor is quasi-identical to that of photon kick by the yield Y (or Y2 for upper limit). This conclusion is however not applicable for displacement damage calculation if the maximum photon kick is close to the threshold displacement energy. In such a case, detailed consideration of photon emission cascades rather than approximate analytical formulae should be included for accurate calculation of neutron capture reaction-induced damage cross section.

Remarks and improvements on neutron KERMA factors and radiation damage cross sections calculated by NECP-Atlas and NJOY21 using different evaluated nuclear data libraries

The nuclear data processing codes sometimes produce unphysical neutron KERMA factors and radiation damage cross sections, for example, a negative neutron KERMA factor. In order to improve neutron KERMA factors and radiation damage cross sections and provide more information for evaluators to improve nuclear data for these data calculation, it is necessary to investigate the reasons for the unphysical results. In this paper, the neutron KERMA factors and radiation damage cross sections are calculated based on CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3 and FENDL-3.1d. The results calculated using different evaluated nuclear data libraries are investigated and the reasons for the unphysical results are specified. Meanwhile, effective solutions and practical suggestions are presented to obtain more reasonable results.

ОШИБКИ В РАДИАЦИОННОМ НАГРЕВЕ УРАНА ПРИ РАСЧЕТАХ С ИСПОЛЬЗОВАНИЕМ БИБЛИОТЕКИ FENDL 3.1D

The work is devoted to the analysis of uranium radiation heating results under the action of neutron radiation. The revealed results discrepancies in the radiation heating modeling of the uranium 235 and 238 using FENDL 3.1d and ENDF nuclear data libraries motivate the study. At the same time, all initial data in the computational model, namely, the neutrons energy spectrum, mass of fissile material, geometry and surrounding structures composition were identical. In the course of the work, the results of energy release calculations in uranium cells were compared using various libraries. Analytical evaluation of the radiation heating results in test cells is given. The obtained data indicate erroneous values of neutron kerma factors in FENDL 3.1d nuclear data library. Fusion Evaluated Nuclear Data Library was created by the working group of the Nuclear Data Section of the IAEA in the mid-1990s. FENDL is focused on application in the design of fusion plants as well as plants for materials testing such as IFMIF. The expanded application required the addition of some materials to the library, the expansion of the energy range to higher energies and the addition of data for reactions caused by charged particles. Version 3.1d was released in January 2018 and is widely used in nuclear analysis of structural materials for the experimental fusion reactor ITER and the demonstration fusion reactor DEMO. The results of this work have been submitted to the ITER International Organization contact group for presentation to the developers of the FENDL library.

Studies on neutron and photon kerma parameters for human body organs

A study on neutron kerma factors and photon air-kerma for human organs is presented for neutron energy range 2.53?10-8 MeV to 29 MeV and photon energy range 1 keV to 20 MeV. The human organs water equivalence for photon and neutron, is also presented. The ratio of the mass-energy absorption coefficients of human organs to water was found constant and unity above 100 keV, whereas there was a large difference for energies below 100 keV. The neutron kerma factors of human organs and water are found of same order of magnitude whereas differs for air. The neutron kerma factors of human organs and tissue substitutes were found to be equal to water for neutron energies between 63 eV and 200 keV. The skeleton-cortical bone was found to be away from water equivalence for low-energy photons and high-energy neutrons.

Neutron kerma factors and water equivalence of some tissue substitutes

The Kerma factors and Kerma relative to the air and water of 24 compounds that are used as tissue substitutes were calculated for neutron energies ranging from 2.53×10−8 to 29MeV. The Kerma ratios of the tissue substitutes relative to air and water were calculated. The water equivalence of the selected tissue substitutes was observed above neutron energies of 100eV. The Kerma ratio relative to the air for poly-vinylidene fluoride and Teflon were nearest to unity at very low energy (up to 1eV) and above 63eV, respectively. It was found that the natural rubber was a water-equivalent tissue substitute compound. The results of the Kerma factors in our investigation show good agreement with those published in ICRU-44. We found that at higher neutron energies, the Kerma factors and Kerma ratios of the selected tissue substitute compounds were approximately the same, but though the differences were large for energies below 100eV.

SU-E-T-108: Development of a Novel Clinical Neutron Dose Monitor for Proton Therapy Based On Twin TLD500 Chips in a Small PE Moderator

Purpose: In proton therapy, it could be desirable to measure out-of-field fast neutron doses at critical locations near and outside the patient body. Methods: The working principle of a novel clinical neutron dose monitor is verified by MCNPX simulation. The device is based on a small PE moderator of just 5.5cm side length for easy handling covered with a thermal neutron suppression layer. In the simulation, a polystyrene phantom is bombarded with a standard proton beam. The secondary thermal neutron flux produced inside the moderator by the impinging fast neutrons from the treatment volume is estimated by pairs of α-Al2O3:C (TLD500) chips which are evaluated offline after the treatment either by TL or OSL methods. The first chip is wrapped with 0.5mm natural Gadolinium foil converting the thermal neutrons to gammas via (n,γ) reaction. The second chip is wrapped with a dummy material. The chip centers have a distance of 2cm from each other. Results: The simulation shows that the difference of gamma doses in the TLD500 chips is correlated to the mean fast neutron dose delivered to the moderator material. Different outer shielding materials have been studied. 0.5mm Cadmium shielding is preferred for cost reasons and convenience. Replacement of PE moderator material by other materials like lead or iron at any place is unfavorable. The spatial orientation of the moderator cube is uncritical. Using variance reduction techniques like splitting/Russian roulette, the TLD500 gamma dose simulation give positive differences up to distances of 0.5m from the treatment volume. Conclusion: Applicability and basic layout of a novel clinical neutron dose monitor are demonstrated. The monitor measures PE neutron doses at locations outside the patient body up to distances of 0.5m from the treatment volume. Tissue neutron doses may be calculated using neutron kerma factors.

Remarks on KERMA Factors in ACE files

Some neutron KERMA factors in ACE files are negative and extremely large if nuclear data libraries do not keep energy-balance. The status of neutron KERMA factors in the official ACE file of ENDF/B-VII.1 is examined. As a result, it is found out that neutron KERMA factors of nuclei more than 200 in ENDF/B-VII.1 have some problems. Effects of the inadequate KERMA factor are also investigated, which are large for neutron heat while those are small for total (neutron + gamma) heat. Users who use only neutron KERMA factors should check if the factors are adequate or not before they use the factors.

Extended use of alanine irradiated in experimental reactor for combined gamma- and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of 14C by LSC

Gamma- and neutron doses in an experimental reactor were measured using alanine/electron spin resonance (ESR) spectrometry. The absorbed dose in alanine was decomposed into contributions caused by gamma and neutron radiation using neutron kerma factors. To overcome a low sensitivity of the alanine/ESR response to thermal neutrons, a novel method has been proposed for the assessment of a thermal neutron flux using the 14N(n,p) 14C reaction on nitrogen present in alanine and subsequent measurement of 14C by liquid scintillation counting (LSC).

SU‐FF‐T‐436: Comparison of Neutron Doses to a Water Target Exposed to a 10 MeV Neutron Beam: Foil Activation Method Vs. Monte Carlo Simulations

We developed a mouse dosimetry model for neutron beam, and compared the calculated and simulated neutron doses for a water target (mouse) exposed to a 10 MeV neutron beam. We also estimated the longitudinal neutron dose profile for the water target. A water bottle (a simulated mouse, 6 cm × 1.3 cm radius) was exposed to a flux of 10 MeV neutrons produced via the d(d, n)He reaction using the Tandem accelerator at Triangle Universities Nuclear Laboratory at Duke University. An estimation of the neutron flux was obtained by neutron activations of two sets of aluminum and gold activation foils. One foil of each type was placed at the front and back sides of the water target. The flux was then calculated based on gamma spectroscopy of the irradiated foils using HPGe detectors. A Monte Carlo (MC) simulation based on FLUKA2008 code, was also used to estimate the 10 MeV neutron dose for the same geometry. The highest neutron flux measured using the activation technique was (8.538×106 +/− 6.89×104) n.cm−2.s−1 at the front end of the target. Using neutron kerma factors from ICRU data tables, the calculated dose rate to a water target at the location of the foil was (1.890+/− 0.015) Gy.hr−1. Using the data from the MC simulation, the calculated neutron dose to the front end of the water target was (1.870 +/− 0.001) Gy.hr−1. The dose calculated using the foil activation technique was in good agreement with the MC results. The slight discrepancy in the measured flux was likely the result of the closer position of the activation foils to the neutron source than the water bottle target, and also absorption within the activation foils. Assuming these results are in good agreement, we estimated the dose profile for the simulated mouse.

Monte Carlo evaluation of a photon pencil kernel algorithm applied to fast neutron therapy treatment planning

When dedicated software is lacking, treatment planning for fast neutron therapy is sometimes performed using dose calculation algorithms designed for photon beam therapy. In this work Monte Carlo derived neutron pencil kernels in water were parametrized using the photon dose algorithm implemented in the Nucletron TMS (treatment management system) treatment planning system. A rectangular fast-neutron fluence spectrum with energies 0–40 MeV (resembling a polyethylene filtered p(41)+ Be spectrum) was used. Central axis depth doses and lateral dose distributions were calculated and compared with the corresponding dose distributions from Monte Carlo calculations for homogeneous water and heterogeneous slab phantoms. All absorbed doses were normalized to the reference dose at 10 cm depth for a field of radius 5.6 cm in a 30 × 40 × 20 cm3 water test phantom. Agreement to within 7% was found in both the lateral and the depth dose distributions. The deviations could be explained as due to differences in size between the test phantom and that used in deriving the pencil kernel (radius 200 cm, thickness 50 cm). In the heterogeneous phantom, the TMS, with a directly applied neutron pencil kernel, and Monte Carlo calculated absorbed doses agree approximately for muscle but show large deviations for media such as adipose or bone. For the latter media, agreement was substantially improved by correcting the absorbed doses calculated in TMS with the neutron kerma factor ratio and the stopping power ratio between tissue and water. The multipurpose Monte Carlo code FLUKA was used both in calculating the pencil kernel and in direct calculations of absorbed dose in the phantom.

10.2172/10188235

This progress report concentrates on two major areas of dosimetry research: measurement of fast neutron kerma factors for several elements for monochromatic and white spectrum neutron fields and determination of the response of thermoluminescent phosphors to various ultra-soft X-ray energies and beta-rays. Dr. Zhixin Zhou from the Shanghai Institute of Radiation Medicine, People`s Republic of China brought with him special expertise in the fabrication and use of ultra-thin TLD materials. Such materials are not available in the USA. The rather unique properties of these materials were investigated during this grant period.

Benchmark Experiment on Vanadium-Alloy Assembly with D-T Neutrons - In-Situ Measurement -

A fusion neutronics benchmark experiment on vanadium-alloy was conducted. Various nuclear quantities related to both neutrons and gamma-rays were measured. It was found through benchmark analyses that all the cross section data of mainly vanadium in recent nuclear data files still involved serious problems. Improvement of these data is strongly required. According to discussion on neutron KERMA factors, the major finding was that neutron KERMA factors based on ENDF/B-VI above 10 MeV were too large due to the larger 51V(n,n'p)50Ti cross section.

Experimental measurement of nuclear heating in a graphite-cantered assembly in deuterium–tritium neutron environment for the validation of data and calculation

Abstract Within the framework of the ITER Task T-218 entitled ‘Shielding Blanket Neutronics Experiments’, nuclear heating measurements were conducted jointly by the USA and Japan using a micro calorimetric technique in a graphite-cantered assembly. An accelerator-based D-T neutron source at JAERI was used to provide a mixed neutron and photon field. The first measurements related to direct micro calorimetric measurements in individual graphite probes along the axis. In the second set, the first graphite probe was replaced, one by one, by single probes of beryllium, aluminum, silicon, silicon carbide, titanium, vanadium, chromium, iron, stainless steel 316, nickel, copper, zirconium, niobium, molybdenum, tungsten. Analysis of the measurements has been carried out using Monte Carlo code MCNP with FENDL-1, ENDF/B-VI and MCPLIB nuclear data libraries. A comparison of calculations (C) and experiments (E) shows a C/E ratio lying in a C/E band extending from 0.9 to 1.2 for beryllium, graphite, copper, chromium, iron, nickel, 316 stainless steel, titanium, vanadium, molybdenum, niobium and tungsten. However, larger deviations from unity are seen for C/E values for silicon, zirconium, and aluminum. Though FENDL-1 and ENDF/B-VI libraries provide very close nuclear heating rates for most of the probe materials, significant divergences are seen for silicon, silicon carbide, aluminum, titanium, zirconium, niobium, and molybdenum. The divergences are traceable to differences in neutron kerma factors as well as gamma production cross-sections of these materials.

The elemental composition of tumors: kerma data for neutrons.

The elemental composition of a variety of tumor samples, including squamous cell lung carcinoma, sarcoma, adenoid cystic carcinoma, melanoma, and rectal carcinoma have been measured by combustion analysis. Hydrogen, carbon, nitrogen, and oxygen content have been determined. Using the elemental neutron kerma data published in ICRU Report #46 the neutron kerma factors for the various tumor samples have been calculated in the energy range 11 eV to 29 MeV. The average neutron kerma values for all tumor samples are approximately 6%-7% lower than those for average soft tissue in the energy range of interest in fast neutron therapy (energies > 1 MeV).

Kerma factor evaluation and its application in nuclear heating experiment analysis

The accuracy of calculating ‘kerma factor’, which is the response function for nuclear heating is very important in many applications, particularly fusion systems. A computer code, called MAZE, has been developed primarily for evaluating neutron and gamma kerma factors based on an improved modeling and a capability to analayze basic data and evaluate consistency in preserving energy. Kinematics equations and energy conservation considerations are modeled in the code. The code has features that allow comparison and analysis among different methods of calculating the kerma factor and it provides users with data information needed to make user-oriented selections from two to three, often different, results. A kerma factor library, called MAZE-LIB, has been generated with MAZE for most of the materials of interest in fusion based on the most recent and accurate data available, FENDL library, which is the nuclear data library selected by ITER. The kerma factor library provided in this work is the result of an extensive analysis of nuclear data and careful selection from among often different results from the different calculational methods. It was shown that some data inconsistency problems in preserving the energy still exist in the FENDL library. Some problems with the widely used computer code NJOY were exposed during the process of the kerma factor evaluation, which is crucial at this stage of fusion research, since NJOY is the primary nuclear data processing code selected by ITER. Comparison with nuclear heating experimental data, derived from very recent integral experiments with 14 MeV neutrons and the analysis of the uncertainty in prediction have been carried out using the newly developed kerma factors. Comparative analysis shows that the new kerma factors generated in this work results in better agreement with the experiments than those obtained from previous work. The calculated-to-experimental values (C/Es) obtained in this work indicate an overestimation of the gamma production in the basic nuclear data library for certain elements. The integral and differential heating profiles along with knowledge of the contributing partial kerma factors, which are generated in this work, could in principle, provide input for sensitivity analysis and be used to improved the basic nuclear data and kerma factor evaluations.

Neutron kerma factors for tissue and particle detector materials from 15 to 150 MeV

Charged particle emission spectra in neutron-induced reactions for elements of tissue and detector materials at energies from 15 to 150 MeV have been calculated with the geometry dependent hybrid model. A comprehensive set of kerma factor values for H, C, O, N, Si, as well as for tissue and frequently used high-energy particle detectors is given.

Kerma factors for calculating the absorbed dose of 15–150 MeV neutrons

A hybrid geometrically dependent model has been used to calculate the yield of charged particles in inelastic neutron—nuclear reactions involving elements which form part of the human tissue and are used in particle detectors, for primary nucleon energies up to 150 MeV. The hybrid model is found to give considerably better agreement with experimental data than the cascade—evaporative and cascade—exciton models in the entire energy range considered. The error in calculating the kerma factors using the hybrid model is estimated as ∼6–10%, which is considerably more accurate that the previously published data. Calculated detailed energy dependences are given of the neutron kerma factors for C, O, N, H, Si, and also for soft biological tissue and certain particle detectors.

Heat Deposition Rate Measurements Using a Graphite Quasi-Adiabatic Calorimeter and Thermoluminescent Dosimeters in a Fusion Environment of the LOTUS Facility

Heat deposition rate measurements are made by an extremely sensitive quasi-adiabatic graphite calorimeter and thermoluminescent dosimeters (TLDs) in the fusion environment of the LOTUS facility. The response of a bare calorimeter and the response inside a large graphite cylindrical block are measured by irradiating with a mixed neutron and gamma field of the Haefely neutron generator. The reproducibility of these measurements is found to be better than 1% for a dose rate more than 60 cGy/min and better than 3.8% for dose rates in the range of 6 to 60 cGy/min. The heating rates are found to vary linearly with neutron source strength. The calculation to experiment (C/E) for the bare calorimeter is found to be 1.05, whereas inside the graphite block, C/E varies from 1.11 to 1.32. These measurements are analyzed by the MCNP Monte Carlo neutron and photon transport code using the BMCCS2, PHOTXS2, and EL2 cross-section libraries. The influence of wall-returned neutrons and gammas is found to be negligible. The origin of the discrepancies is found by measuring the gamma component of the heating at identical locations by conducting special geometry irradiation using several TLDs-700. The conditions that are employed considerably simplify the transformation of the TLD results to that of the graphite medium. A detailed data treatment is done with the TLD outputs to arrive at the gamma heating component at different locations in the graphite by employing the Burlin theory. The gamma production is found to be well represented in the calculations. On the other hand, measured and calculated net nuclear heating in the graphite differ considerably. A downward revision of the neutron kerma factor would be desirable.

Measurement of Neutron Kerma Factors in C and O: Neutron Energy Range of 20 MeV to 70 MeV

Measurement of Neutron Kerma Factors in C and O: Neutron Energy Range of 20 MeV to 70 MeV