Neutron Sensitivity Research Articles

Investigation of neutron converters for production of optically stimulated luminescence (OSL) neutron dosimeters using Al 2O 3:C

This paper presents the optically stimulated luminescence (OSL) properties of neutron dosimeters in powder and in the form of pellets prepared with a mixture of Al 2O 3:C and neutron converters. The neutron converters investigated were high density polyethylene (HDPE), lithium fluoride (LiF), lithium fluoride 95% enriched with 6Li ( 6LiF), lithium carbonate 95% enriched with 6Li ( 6Li 2CO 3), boric acid enriched with 99% of 10B ( H 3 10 BO 3 ) and gadolinium oxide (Gd 2O 3). The proportion of Al 2O 3:C and neutron converter in the mixture was varied to optimize the total OSL signal and neutron sensitivity. The neutron sensitivity and dose-response were determined for the OSL dosimeters using a bare 252Cf source and compared to the response of Harshaw TLD-600 and TLD-700 dosimeters ( 6LiF:Mg,Ti and 7LiF:Mg,Ti). The results demonstrate the possibility of developing an OSL dosimeter made of Al 2O 3:C powder and neutron converter with a neutron sensitivity (defined as the ratio between the 60Co equivalent gamma dose and the reference neutron absorbed dose) and neutron–gamma discrimination comparable to the TLD-600/TLD-700 combination. It was shown that the shape of the OSL decay curves varied with the type of the neutron converter, demonstrating the influence of the energy deposition mechanism and ionization density on the OSL process in Al 2O 3:C.

New cathode design boron lined proportional counters for neutron area monitoring application

A new cathode design boron lined proportional counter of 26 mm ID × 100 mm sensitive length SS304 cathode has been developed with boron-coated baffles separated by 3 mm spacers inserted in the sensitive volume perpendicular to the axis. The baffles and the spacers were coated with indigenously available 27.7% enriched 10B. The introduction of baffles enhanced the boron coated surface area by a factor of 2.8. Tests in 120 nv thermal neutron flux show that the counter has 0.84 cps/nv thermal neutron sensitivity, which shows enhancement in the sensitivity by a factor of 2.78 due to baffle structure. For comparison standard cylindrical cathode geometry counter coated with 92% enriched 10B on its inner wall with a coating thickness of 0.8 mg / cm 2 is developed with same outer dimensions for neutron area monitoring applications. The counter has 1 cps/nv thermal neutron sensitivity. Comparative tests carried out on counters with and without baffle structure show that the baffles enhance the neutron sensitivity and in 2 kR/h gamma background the effect of gamma pile up is similar on both the counters. The variation in cathode internal diameter due to baffle structure gives higher voltage plateau slope (2.8%/10 V) as compared to conventional cylindrical geometry counter (1.2%/10 V). The usability of boron lined counters for neutron area monitoring applications for the cylindrical geometry counter has been studied.

Organic Semiconductors for Detection of Ionizing Radiation

AbstractOrganic semiconductors are increasingly common in electronics and sensors, and are now under investigation for a novel type of radiation sensors at Sandia National Laboratories. This class of materials can offer wide band gaps, high resistivities, low dielectric constants, and high dielectric strengths, suggesting they may be suitable for solid-state particle counting detectors. A range of solution cast materials have been evaluated for this application, primarily in the family of poly(p-phenylene vinylene)s, or PPVs. The high ratio of hydrogen to carbon offers neutron sensitivity, while the low Z material provides some natural gamma discrimination. Compared to existing detectors, these materials could potentially offer large-scale radiation detection at a substantially reduced cost.While PPVs hold promise for radiation detection, the mechanical and electrical properties must be optimized and the processing effects understood. Polymers can offer significantly simplified processing compared to the more common crystals used in solid state detection, which can be size limited and fragile. However, organic semiconductors are very sensitive to processing conditions, and mobility can be affected by orders of magnitude by processing variables, without altering any chemistry. Additives can also have dramatic effects on both electrical and mechanical properties. We report on nanoparticle additives that cause an increase in photoresponse of approximately three orders of magnitude as compared to a polymer film without additives. We separately show an order of magnitude increase in photoresponse by exposing the polymer/fullerene composite to sub-bandgap light.Future work will analyze the feasibility of single particle detection and various geometries for optimization. Additional processing variables will also be investigated for further improvement of mobility and reduction of trap density.

Neutron dosimetry with TL albedo dosemeters at high energy accelerators

The GSF-Personal Monitoring Service uses the TLD albedo dosemeter as standard neutron personal dosemeter. Due to its low sensitivity for fast neutrons however, it is generally not recommended for workplaces at high-energy accelerators. Test measurements with the albedo dosemeter were performed at the accelerator laboratories of GSI in Darmstadt and DESY in Hamburg to reconsider this hypothesis. It revealed that the albedo dosemeter can also be used as personal dosemeter at these workplaces, because at all measurement locations a significant part of neutrons with lower energies could be found, which were produced by scattering at walls or the ground.

Photon and neutron dose discrimination using low pressure proportional counters with graphite and A150 walls

A graphite-walled proportional counter with low neutron sensitivity was used in combination with a tissue-equivalent proportional counter (TEPC) to separate the photon and neutron components in mixed radiation fields. Monte Carlo (MCNP4C) simulations of the photon and neutron responses of the two detectors were done to obtain correction factors for the sensitivity differences. In an alternative method the radiation components were determined using constant-yD-values for typical photon and neutron energy distributions. The results show no significant difference between the two methods and the measured neutron dose-equivalent agrees within +/-50% with Bonner sphere determined values. The experimental data were obtained in measurement campaigns organised within the EVIDOS-project.

In vivo and phantom measurements of the secondary photon and neutron doses for prostate patients undergoing IMRT

For intensity modulated radiation therapy (IMRT) treatments 6 MV photons are typically used, however, for deep seated tumors in the pelvic region, higher photon energies are increasingly being employed. IMRT treatments require more monitor units (MU) to deliver the same dose as conformal treatments, causing increased secondary radiation to tissues outside the treated area from leakage and scatter, as well as a possible increase in the neutron dose from photon interactions in the machine head. Here we provide in vivo patient and phantom measurements of the secondary out-of-field photon radiation and the neutron dose equivalent for 18 MV IMRT treatments. The patients were treated for prostate cancer with 18 MV IMRT at institutions using different therapy machines and treatment planning systems. Phantom exposures at the different facilities were used to compare the secondary photon and neutron dose equivalent between typical IMRT delivered treatment plans with a six field three-dimensional conformal radiotherapy (3DCRT) plan. For the in vivo measurements LiF thermoluminescent detectors (TLDs) and Al2O3 detectors using optically stimulated radiation were used to obtain the photon dose and CR-39 track etch detectors were used to obtain the neutron dose equivalent. For the phantom measurements a Bonner sphere (25.4 cm diameter) containing two types of TLDs (TLD-600 and TLD-700) having different thermal neutron sensitivities were used to obtain the out-of-field neutron dose equivalent. Our results showed that for patients treated with 18 MV IMRT the photon dose equivalent is greater than the neutron dose equivalent measured outside the treatment field and the neutron dose equivalent normalized to the prescription dose varied from 2 to 6 mSv/Gy among the therapy machines. The Bonner sphere results showed that the ratio of neutron equivalent doses for the 18 MV IMRT and 3DCRT prostate treatments scaled as the ratio of delivered MUs. We also observed differences in the measured neutron dose equivalent among the three therapy machines for both the in vivo and phantom exposures.

New measurement of neutron capture resonances inBi209

The neutron capture cross section of Bi-209 has been measured at the CERN n_TOF facility by employing the pulse-height-weighting technique. Improvements over previous measurements are mainly because of an optimized detection system, which led to a practically negligible neutron sensitivity. Additional experimental sources of systematic error, such as the electronic threshold in the detectors, summing of gamma-rays, internal electron conversion, and the isomeric state in bismuth, have been taken into account. gamma-Ray absorption effects inside the sample have been corrected by employing a nonpolynomial weighting function. Because Bi-209 is the last stable isotope in the reaction path of the stellar s-process, the Maxwellian averaged capture cross section is important for the recycling of the reaction flow by alpha decays. In the relevant stellar range of thermal energies between kT=5 and 8 keV our new capture rate is about 16 presently accepted value used for nucleosynthesis calculations. At this low temperature an important part of the heavy Pb-Bi isotopes are supposed to be synthesized by the s-process in the He shells of low mass, thermally pulsing asymptotic giant branch stars. With the improved set of cross sections we obtain an s-process fraction of 19 +/- 3 residual of 81 +/- 3 is also of relevance for the design of accelerator driven systems based on a liquid metal Pb/Bi spallation target.

Determination of the response to photons and thermal neutrons of new LiF based TL materials for radiation protection purposes

LiF based thermoluminescent materials are the most used passive detectors in the field of radiation protection. For personal and area dosimetry in mixed neutron and photon fields, /sup 6/LiF and /sup 7/LiF TLD pairs are employed for the simultaneous determination of neutron and photon doses. Therefore, an accurate knowledge of their sensitivity to photons and thermal neutrons is essential. Aimed at providing reliable sensitivity data for new LiF based TL materials, an experimental study was performed in the framework of collaboration between INFN Frascati National Laboratories and ENEA, Italy. The /sup 6/LiF(Mg, Cu, P) material from TLD Poland, was specially manufactured in six versions with different sensitive thickness (t=40,60,80,100,150 and 230 mg/spl middot/cm/sup -2/), which implies different neutron and photon sensitivity. Together with the /sup 7/LiF(Mg, Cu, P) material, namely MCP-7, all seven type of TLDs were irradiated in reference beams of photons and thermal neutrons, at the calibration facilities of INFN and ENEA. The study provides a set of sensitivity data and discusses the performance of the investigated materials for the dosimetry of mixed neutron and photon fields. All uncertainties or error bars reported in the paper refer to 95% confidence level (2 sigma).

TU‐B‐224A‐01: Shielding II:Linac Radiation Monitoring and Surveys: Instruments and Methods

Regulatory agencies typically require shielding integrity radiation surveys during commissioning of radiation therapy linear accelerators (linacs). While concrete barriers that provide adequate shielding for photons also provide adequate shielding for neutrons, facilities operating at energies above 10 MV shall be checked for neutrons at the door, maze entrance, and any other openings through the shielding. Laminated barriers shall be monitored for neutrons beyond the shielding.For the primary barrier measurements, the maximum field size is utilized without a phantom in the beam. Gantry angles of 0, 90, 180, 270 degrees as well oblique angles depending upon the shielding configuration are commonly used. Secondary barriers are surveyed with the maximum field size and a phantom in place.Photon surveys outside the barriers are performed typically with a calibrated ionization chamber which has both rate and integrate modes, at 30 cm from the barrier. Head leakage in the linac room can be established with the use of film wrapped around the linac head and integrating dosimeters.In this lecture neutron monitoring will be emphasized. Neutron measurements inside the treatment room are fraught with difficulties because of photon interference from the primary and leakage photon beam and the fact that neutron detection is spread over many decades of energy. Thus no single neutron detector can accurately measure neutron fluence or dose equivalent over the entire energy ranges. Additionally neutron detectors can have photon‐induced reactions when used in the primary photon beam. Further because therapy linacs are operated in a pulsed mode, the intense photon pulse overwhelms any active detector that detects particles electronically. Thus active detectors such as such as neutron rem‐meters, fluence meters and spectrometers neutron cannot be used inside the treatment rooms except at or near the maze entrance. They can be used outside the shielded treatment room.Passive monitors with high neutron sensitivity such as moderated activation foils (gold and indium) and threshold activation detectors (phosphorous) can be typically used inside the treatment room and inside the primary beam. Moderated activation foils can also be used inside the treatment room and outside the primary beam. Solid state neutron detectors (SSNTDs) such as CR‐39 ® and bubble detectors can be used inside the treatment room, but outside the primary beam. Bubble detectors can also be used for radiation surveys outside the shielded treatment room.Educational Objectives:1. Understand how to perform shielding integrity radiation surveys.2. Understand the various neutron monitoring methods and instruments.3. Understand under which conditions these monitors can be used.

Dosimetry at the Portuguese research reactor using thermoluminescence measurements and Monte Carlo calculations

This work presents an extensive study on Monte Carlo radiation transport simulation and thermoluminescent (TL) dosimetry for characterising mixed radiation fields (neutrons and photons) occurring in nuclear reactors. The feasibility of these methods is investigated for radiation fields at various locations of the Portuguese Research Reactor (RPI). The performance of the approaches developed in this work is compared with dosimetric techniques already existing at RPI. The Monte Carlo MCNP-4C code was used for a detailed modelling of the reactor core, the fast neutron beam and the thermal column of RPI. Simulations using these models allow to reproduce the energy and spatial distributions of the neutron field very well (agreement better than 80%). In the case of the photon field, the agreement improves with decreasing intensity of the component related to fission and activation products. (7)LiF:Mg,Ti, (7)LiF:Mg,Cu,P and Al(2)O(3):Mg,Y TL detectors (TLDs) with low neutron sensitivity are able to determine photon dose and dose profiles with high spatial resolution. On the other hand, (nat)LiF:Mg,Ti TLDs with increased neutron sensitivity show a remarkable loss of sensitivity and a high supralinearity in high-intensity fields hampering their application at nuclear reactors.

Study of materials for mixed field dosimetry by EPR spectroscopy

EPR dose reconstruction after accidental photon exposure based on materials irradiated in the vicinity of the victim (sucrose, medicine tablets, etc.) was used successfully in several cases referenced in the literature. However, accidental exposure may also occur with a neutron component such as in the Tokai-Mura criticality accident. The aim of this work is to investigate the potentiality of EPR dosimetry for mixed photon and neutron field exposure with different organic materials already used for photon exposure (sucrose) or with potential dosimetric properties (ascorbic acid, sorbitol, glucose, galactose, fructose, lactose and mannose). To assess the neutron sensitivity, the materials were exposed to a mixed radiation field of an experimental reactor with different neutron to photon ratios. The relative neutron sensitivity was found to range from 12 to 43% according to the materials. The potentiality of these materials for mixed field EPR dosimetry is discussed.

Extremely high dose neutron dosimetry using CR-39 and atomic force microscopy

Atomic force microscopy (AFM) has been applied to the analysis of CR-39 nuclear track detectors for high dose neutron dosimetry. As a feasible study to extract the neutron dose, we have employed a (239)Pu-Be neutron source with the traditional track density measurement of recoil proton etch pits from a high density polyethylene (CH(2)) radiator. After very short etching ( approximately 1 microm), etch pit densities were measured as a function of neutron fluence (neutron dose) up to 1.4 x 10(10) cm(-2) (6.6 Sv). Neutron sensitivity was also measured to be 6.6 x 10(-4). Maximum measurable neutron dose was estimated to be approximately 200 Sv by measuring the fraction of the total image area occupied by the etch pits.

Spent Fuel Monitoring with Silicon Carbide Semiconductor Neutron/Gamma Detectors

Abstract Silicon carbide semiconductor radiation detectors have been demonstrated for neutron and gamma-ray monitoring of spent nuclear fuel. Neutrons and gamma rays were monitored simultaneously over a 2050-h period, resulting in a gamma dose of over 6000 Gy to the SiC detector. No changes in the neutron and gamma-ray sensitivity were observed as a result of the gamma-ray exposure. After the spent-fuel measurements, the absolute neutron sensitivity was determined in a standard neutron field, and the degree of gamma-ray and neutron spectrum overlap were determined through exposures in intense gamma ray fields. No overlap was observed for gamma-ray dose rates up to 100 G/h.

Study of neutron sensitivity in CMS–RPC using MC simulation for two different setups

In order to investigate the neutron sensitivity we applied a realistic MC simulation based on GEANT3 code with MICAP and FLUKA interfaces for two different types of resistive plate chambers for CMS/LHC. In the present studies sensitivity calculations have been performed as a function of neutron energy in the range of 100 eV to 1 GeV. In the first setup using aluminum (GND+strips), the sensitivity for an isotropic neutron source was obtained to be S n < 9.04 × 10 - 3 at < 100 MeV, while for the second setup using copper (GND+strips) sensitivity for the same source configuration is S n < 9.91 × 10 - 3 at < 100 MeV . Both the results are relative to an RPC in double-gap configuration with an isotropic neutron source. Additionally in these setups similar characteristics have been observed for parallel neutron source configuration.

Response of RadFET dosimeters to high fluences of fast neutrons

We present irradiation experiments carried out on RadFETs in the high-intensity T2 neutron beam at the CRC-UCL in Belgium. The aim of the test was to characterize the neutron response of RadFETs in view of their use as an integrated part of a radiation-monitoring sensor for the CERN Large Hadron Collider (LHC) experiments. Two types of RadFETs were investigated up to a total neutron fluence of 3/spl times/10/sup 14/ cm/sup -2/ corresponding to a deposited dose of 744 Gy in silicon. The responses of bare devices to neutrons are compared to the commonly used reference measurements with gamma rays. It is found that the gamma ray calibration cannot directly be adopted to convert the RadFET signals into neutron dose. In a second experiment, the influence of a plastic packaging, simulated by polyethylene slabs of different thicknesses, was tested in the neutron beam and compared to GEANT4 Monte Carlo simulations. An increase of the RadFET neutron sensitivity by a factor of up to 7 due to the packaging is found. The influence of these findings on the conception of the radiation-monitoring sensor is discussed.

Measurements of the 90,91,92,94,96Zr( n, γ) cross-sections at n_TOF

Neutron capture cross sections of the 90,91,92,94,96 Zr have been measured over the energy range from 1 eV to 1 MeV at the spallation neutron facility n_TOF at CERN in 2003. The innovative features of the neutron beam, in particular the high instantaneous flux, the high energy resolution and low background, together with improvements of the neutron sensitivity of the capture detectors make this facility unique for neutron-induced reaction cross section measurements with much improved accuracy. The preliminary results of the Zr measurements show capture resonance strengths generally smaller than in previous measurements.

MO-D-T-617-02: Measurement of Photon and Neutron Doses Outside the Treatment Field for Prostate Patients Undergoing 18 MV IMRT

Purpose: For IMRT treatments 6 MV photons are typically used, however, for deep seated tumors in the pelvic region, higher photon energies are increasingly being employed. IMRT treatments require more monitor units (MUs) to deliver the same dose as conformal treatments, causing increased secondary radiation from leakage and scatter, as well as a possible increase in the neutron dose from photon interactions in the machine head. Here we provide in-vivo and phantom measurements of the photon secondary radiation and the neutron dose equivalent for 18 MV IMRT treatments. Method and Materials: We use different detectors with various photon and neutron sensitivities: TLD 600 and TLD 700 chips, a Neutrak detector containing Al〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 to detect photons using optically stimulated radiation and a CR-39 plastic sheet using track etching for neutrons from Landauer, and a separate CR-39 detector from Argonne National Labs. In -vivo patient measurements were obtained outside the field edge for 9 prostate patients undergoing 18 MV IMRT on two different commercial accelerators. We also compared the secondary photon dose for 6 prostate patients undergoing 6 MV IMRT with that for 18 MV IMRT. Additionally, a 25.4 cm diameter Bonner sphere containing TLD 700 and TLD 600 was used to compare the out of field secondary doses for typical 18 MV IMRT and 18 MV six-field conformal prostate treatments. Results: The patient measurements showed photon doses approximately 12 and 7 times greater than the neutron dose equivalent at 10 and 20 cm from the field edge, respectively. Initial Bonner sphere measurements showed neutron doses for IMRT greater than for conformal treatments. Conclusion: For prostate treatment the photon secondary dose for 18 MV and 6 MV IMRT are similar, and the neutron dose is higher than the dose from 18 MV conformal treatments by about the ratio of MUs.

Energy response of a PbWO4 scintillation detector to gamma and neutron radiation

A scintillation detector based on PbWO4 is used to measure gamma rays in a mixed neutron/gamma field from a high intensity pulsed reactor. In order to calibrate gamma sensitivities, a series of monoenergetic radioisotope sources are manufactured in a reactor. Two high-energy gamma sources were obtained from nuclear reactions 12C(p, γ)13N and 19F(p, αγ)16O by means of the 600 kV high voltage accelerator. The energy response of the detector to gamma rays in the energy range 0.15 MeV ⩽ En ⩽ 6.13 MeV was measured and the results are in good agreement with the MCNP simulations. In order to get the absolute sensitivities of the detector to neutrons, one experiment was carried out using 2.5 and 14.1 MeV neutrons from D(d, n)3He and T(d, n)4He reactions and another experiment was carried out using 0.565, 1.2, 1.8, 2.5 MeV neutrons from the T(p, n)3He reaction. The result shows the detector has a high gamma-to-neutron discrimination by comparing the gamma sensitivity at 1.25 MeV with the neutron sensitivity at 1.2 MeV. It is an appropriate gamma detector for pulsed gamma rays in a mixed neutron/gamma field.

Neutron sensitivity of thin gap chambers

Thin gap chambers (TGC) will be used for triggering forward muons in the ATLAS detector for the LHC at CERN. A large amount of neutron background is foreseen in the ATLAS experiment. This paper describes the measurements of the neutron sensitivities (detection efficiencies) of the TGCs. The sensitivities of both small and real size TGCs to 2.5 and 14 MeV mono-energetic neutrons were measured. For a small size TGC, sensitivities of 0.032% and 0.10% were measured to 2.5 and 14 MeV neutrons, respectively, whereas for a real size TGC, sensitivities of 0.048% and 0.13% were measured. These measured values were in reasonably good agreement with the simulations based on the Geant4.

Neutron imaging using an optical stimulated luminescent material: /spl alpha/-Al/sub 2/O/sub 3/:C+Gd/sub 2/O/sub 3//sup */

An optical stimulated luminescent (OSL) material /spl alpha/-Al/sub 2/O/sub 3/:C is sensitized for neutrons. Basic performances of neutron sensitivity, linearity, and fading characteristics are studied for three types of neutron sensitive OSL sheets using a prompt OSL reader. Neutron imaging is also studied using an imaging OSL reader. Preliminary results of neutron image will be shown in this paper.