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The paper presents the results of a study of the neutron field characteristics surrounding biological shielding of Russian research nuclear reactors to establish their expected behavior and enhance personnel neutron monitoring. The study covered five nuclear research reactors: IVV-2M, IRT-T, IRT-MEPhI, RBT-6, and SM-3. All reactors are pool-typed; some (RBT-6 and SM-3) have pressurized water, while others (IVV-2M, IRT-T, and IRT-MEPhI) have water under normal pressure. The neutron fields analyzed are located at the tank cover of reactors, in front of the reactor core covered with biological shielding, and in front of the horizontal experimental channel. The spectrum average neutron energy ranges from 0.01 MeV (RBT-6) to 1.0 MeV (SM-3). The fluence-to-ambient dose conversion coefficient varies from 15 pSv ∙ cm2 (RBT-6) to 260 pSv ∙ cm2 (SM-3). The spectra at the other measurement points exhibit the energy distribution characteristics of the neutron radiation flux density behind the biological shielding. • The paper presents the results of a study of the neutron field characteristics surrounding biological shielding of Russian research nuclear reactors to establish their expected behavior and enhance personnel neutron monitoring. • The study covered five nuclear research reactors: IVV-2M, IRT-T, IRT-MEPhI, RBT-6, and SM-3. • The neutron fields analyzed are located at the tank cover of reactors, in front of the reactor core covered with biological shielding, and in front of the horizontal experimental channel. • The spectrum average neutron energy ranges from 0.01 MeV (RBT-6) to 1.0 MeV (SM-3). • The fluence-to-ambient dose conversion coefficient varies from 15 pSv ∙ cm2 (RBT-6) to 260 pSv ∙ cm2 (SM-3). [ABSTRACT FROM AUTHOR]

This paper presents the evaluation of the fluence-to-ambient dose equivalent response function (referred to as F2D response function) of the Aloka TPS-451C neutron dose equivalent rate meter (hereafter referred to as a neutron meter) based on a singular value decomposition (SVD) approach. Measurements of neutron ambient dose equivalent rates, Ḣ ∗ (10) , using the neutron meter were conducted in various neutron standard fields of 241Am-Be source with different neutron fluence rate spectra. A series of Ḣ ∗ (10) values and corresponding neutron fluence rate spectra were used as the input data for unfolding the F2D response function of the neutron meter. To verify the SVD-based method, the fluence response functions of a well-known Bonner sphere spectrometer system were unfolded using the SVD method, and the results were compared with those provided in the IAEA technical report No. 403. The SVD method was then applied to determine the F2D response function of the Aloka TPS-451C neutron meter with the use of the ICRP-74 F2D response function and that predicted in a previous work as initial guesses. The consistency between initial guesses and the SVD unfolded F2D response function of the Aloka TPS-451C neutron meter implies the reliability of the SVD method for determining the response functions of neutron meters. • SVD-based computer code has been developed. • Fluence-to-ambient dose equivalent response function of a neutron meter was derived. • Standard uncertainties of response functions were discussed in the paper. [ABSTRACT FROM AUTHOR]

Dose evaluation and direct measurements are fundamental for radiation protection in non-conventional accelerator facilities, both before and after the primary and secondary shielding. In this paper, we will report about the experimental setup, data acquisition and analysis, together with FLUKA modeling, of the dose measurements test carried out in the Beam Test Facility (BTF) of the INFN - Frascati's National Laboratories (LNF), where an intense mixed field is produced and measured with thermoluminescent dosimeters. BTF is an extraction and transport line of DA Φ NE LINAC (Buonomo et al. 2021; Mazzitelli et al. 2003). It is optimized for electrons and positrons production in a wide range of intensity, energy (30 MeV–800 MeV), beam spot dimensions and divergence, using both primary and secondary beam of the DA Φ NE LINAC. Through the years, the BTF has gained an important role in particle detectors test and development with electron/positron beam. A small fraction of the BTF's shifts have been dedicated to radiation damage test using LINAC electron primary beam up to 5 × 1 0 10 e-/s. As radiation protection group of the LNF, we evaluated the dose when electrons impinging on a Pb target from: (i) photon Bremsstrahlung production; (ii) photoneutron production. Three dedicated tests with 503 MeV electrons impinging on a ∼ 16 cm thick Pb target have been carried out in February, June 2022 and in January 2023, using TLD700 and TLD600, measuring doses at several charge intervals. The aim of this study focuses on evaluating dosimetric quantities produced by the mixed field, air kerma for the photon component, and ambient dose equivalent for the neutron one, using thermoluminescence dosimeters calibrated with low-energy standards: Cs-137 and Am-Be. The approach adopted involves the use of Monte Carlo simulations of the experiment, both to benchmark against experimental measurements and to validate the results obtained for energies higher than those of calibration. The results of this comparison show excellent agreement between measured and simulated quantities in the forward direction, allowing us to conclude and confirm the validity of the calibrations themselves. • Fundamental dose evaluation and measurements in non-conventional accelerator facilities. • Experimental setup, data acquisition, and analysis at LNF-INFN's Beam Test Facility. • Dose measurements up to 500 MeV using TLD700 and neutron detector, calibrated sources. • Evaluation of air kerma for photons and ambient dose equivalent for neutrons. • Validation via FLUKA simulations for energies beyond calibration. • Measured and simulated quantities confirm calibration validity at higher energies. [ABSTRACT FROM AUTHOR]

Within the EURADOS (European Radiation Dosimetry Group) self-sustained programme of regular intercomparisons (ICs) for individual monitoring, three campaigns had been performed in 2012, 2017 and 2022 to assess the performance of neutron personal dosemeters routinely used to measure personal dose equivalent, H p (10). The irradiation schemes were established to provide participants with useful information regarding linearity, reproducibility and response of their dosimetry systems at different energies and angles of incidence. Small alterations were made in irradiation plans for subsequent exercises. Combinations of standard calibration fields and simulated workplace fields were employed, using bare and moderated neutron sources as well as monoenergetic and thermal neutron fields. Neutron energies used to evaluate dosemeter performance ranged from thermal to several MeV. Dose values varied between 0.3 mSv and 15 mSv. The number of participating laboratories in the three campaigns was stable: 31 in IC2012n, 32 in IC2017n and 29 in IC2022n. The dosimetry systems could have been categorized according to their detection principle into albedo and track dosemeters, or a combination of both. In 2012, a category 'other' had to be introduced to consider active personal dosemeters and fission track detectors. In 2012, participants were asked to provided results in a two-step procedure: (i) with no information on the radiation fields, and (ii) after receiving additional simplified a priori information on the energy distribution of the neutron fields to enable correction for dosemeter response. In 2017 and 2022, based on a one-step procedure, however, participants might have requested a priori information upon registration, which was reported in the certificate of participation. Intercomparison results were analysed against the acceptance criteria of the ISO 14146:2018 standard, although there were no standardized performance requirements for neutron personal dosemeters at the time of IC2012n. The paper compares the results of the three exercises, identifies major problems for individual monitoring services and shows trends in dosimetric performance. In all three intercomparisons, a few systems had shown problems related to calibration. Data do not allow to draw a universally applicable conclusion with regard to potentially superior performance of one particular dosimetry system. • Results from three EURADOS inter-comparisons on neutron dosimetry for individual monitoring. • Performance of neutron personal dosemeters tested in various radiation fields. • Statistical overview of personal dosemeter responses. [ABSTRACT FROM AUTHOR]

Organic scintillators detect neutrons by scattering reactions with hydrogen nuclei (protons) in organic compounds. The recoil protons then produce the light output from the scintillator. The energy spectrum of neutrons impinging on the scintillator has to be determined from this light output using a so-called unfolding algorithm. However, this problem is ill-posed and therefore difficult to solve. In this study, a new method for unfolding neutron spectra is presented. This method uses B-splines approximation of neutron energy spectrum. Results of this method are shown for test data and experimental data and compared to the results of the previously developed unfolding method. • A new method for unfolding neutron spectra using B-splines is presented in this paper. • This unfolding method was tested on test and experimental data. • A good agreement was found between the results of the proposed method and Expectation Maximization algorithm. [ABSTRACT FROM AUTHOR]

Abstract: DEXA (dual-energy X-ray absorption) is widely used in airport scanners, industrial scanners and bone densitometers. DEXA determines the properties of materials by measuring the absorption differences of X-rays from a bremsstrahlung tube source with and without filtering. Filtering creates a beam with a higher mean energy, which causes lower material absorption. The absorption difference between measurements (those with a filter subtracted from those without a filter) is a positive number that increases with the effective atomic number of the material. In this paper, the concept of using a filter to create a dual beam and an absorption difference in materials is applied to radiation from a 252Cf source, called NEUDEG (neutron and dual-energy gamma absorption). NEUDEG includes absorptions for fast neutrons as well as the dual photon beams and thus an incentive for developing the method is that, unlike DEXA, it is inherently sensitive to the hydrogen content of materials. In this paper, a model for the absorption difference and absorption sum in NEUDEG is presented using the combined gamma ray and fast neutron mass attenuation coefficients. Absorption differences can be either positive or negative in NEUDEG, increasing with increases in the effective atomic number and decreasing with increases in the hydrogen content. Sample sets of absorption difference curves are calculated for materials with typical gamma-ray and fast neutron mass attenuation coefficients. The model, which uses tabulated mass attenuated coefficients, agrees with experimental data for porcelain tiles and polyethylene sheets. The effects of “beam hardening” are also investigated. [Copyright &y& Elsevier]

Abstract: When applying numerical simulation to physical problems it is important to be able to clearly understand the quality that can be reached and the uncertainty associated with the modelling. In this paper we will deal with the Monte Carlo simulation of neutron coincidence and multiplicity counters. These instruments are intensively used in nuclear safeguards for the measurement and verification of the mass of fissile materials in nuclear installations. Their numerical simulation, either for detector design or for numerical calibration, is now used more and more frequently as support to inspections. Several benchmark exercises have been organised in order to assess the capabilities of Monte Carlo simulation to correctly predict realistic practical cases in neutron coincidence and multiplicity counting. Nevertheless most of them allowed reaching conclusions only on the overall uncertainties associated to the modelling through the comparison between calculated and measured integral data. In this paper we aim to analyse more deeply the sources of these uncertainties, separating the different possible causes and analysing parametrically the individual factors. We will include influencing factors deriving from the physical model of the detector, of the acquisition electronics, of the sample and from the nuclear data. The final total uncertainties are consistent with the results of the integral benchmarks. [Copyright &y& Elsevier]

Proton cross-section libraries and interaction models available in the Monte Carlo code MCNP6.2 were tested and compared for energies relevant to proton radiation therapy (100, 150, 226 MeV), and for three different phantom materials. Proton, neutron, and photon characteristics were studied. Primary proton parameters included proton total deposited energy and Bragg peak position. The results varied within ±0.5%. Secondary neutron characteristics included the number of originated secondary neutrons, their average initial energy, energy spectrum, depth-dose curve, radiation weighting factor, neutron equivalent dose inside phantoms, and neutron spectral fluence and ambient dose equivalent outside phantoms. Neutron radiation weighting factor inside a phantom varied within ±8% and neutron equivalent dose typically within ±30%. Large differences were observed in neutron fluence energy spectra both inside and outside a phantom. Secondary photon characteristics consisted of photon equivalent dose inside phantoms. The study demonstrates that the selection of input proton data may have a large impact on the simulated quantity, especially if the intention is to model secondary particles, particularly neutrons. • 5 proton cross-section tables and 5 interaction physics models were compared for protons from 100 to 226 MeV, in MCNP6.2. • Results related to primary protons are consistent. • Results related to secondary neutrons differ significantly. • The paper provides an estimate how large the differences can be, for various quantities. [ABSTRACT FROM AUTHOR]

Abstract: The European Commission has funded within its 6th Framework Programme a three-year project (2005–2007) called CONRAD, COordinated Network for RAdiation Dosimetry. The organizational framework for this project was provided by the European radiation Dosimetry Group EURADOS. Work Package 6 of CONRAD dealt with “complex mixed radiation fields at workplaces” and in this context it organised a benchmark exercise, which included both measurements and calculations, in a stray radiation field at a high-energy particle accelerator at GSI, Germany. The aim was to intercompare the response of several types of active detectors and passive dosemeters in a well-characterised workplace field. The Monte Carlo simulations of the radiation field and the experimental determination of the neutron spectra with various Bonner Sphere Spectrometers are discussed in Rollet et al. (2008) and in Wiegel et al. (2008). This paper focuses on the intercomparison of the response of the dosemeters in terms of ambient dose equivalent. The paper describes in detail the detectors employed in the experiment, followed by a discussion of the results. A comparison is also made with the H* (10) values predicted by the Monte Carlo simulations and those measured by the BSS systems. [Copyright &y& Elsevier]

Abstract: Strong fields of thermal and epithermal neutrons as used for Boron Neutron Capture Therapy (BNCT) show a relatively small gamma dose. However, for accurate dose estimation, both the neutron as well as the gamma component has to be considered. Different approaches have been studied with various thermoluminescence detectors (TLDs). As a result, CaF2:Tm detectors (TLD-300) are attributed to be most insensitive to thermal neutrons but not tissue-equivalent. In this paper we focus on the measurement of the gamma dose with TLD-300 in mixed irradiation fields. To overcome the disadvantage of being not tissue-equivalent, a study on energy compensation filters was performed. The paper comprises also calibrations of TLD-300 using energy compensation filters at the PTB-reference field for thermal neutrons, and first test experiments at the TRIGA Mainz including simulations. [Copyright &y& Elsevier]

Abstract This paper reports on applicability of a polyallyl-diglycol-carbonate (PADC) detectors of the CR-39 TASTRAK type as integrated neutron fluence sensors for monitoring of neutrons emitted from 14 MeV neutron generator. Neutron generators such as SODERN Model: GENIE 16 D are powerful emitters of fusion neutrons from DT reactions with yields up to 2 × 108 n·s−1. The presented generator is now available worldwide as a source of neutrons with a rather wide range of applications in medicine, industry and in the detection of hazardous materials. Before using in those measurements, the CR-39 TASTRAK type detector was precisely calibrated. Rectangular track detector samples prepared for neutron flux measurements are in the form of "sandwiches" consisting of a detector plate covered by a thick polythene (PE) radiator and in some part by a Cu-filter. In order to estimate how many neutrons could be converted into recoil-protons we have performed computer calculations by means of MC methods. Highlights • To use CR-39 detectors carefully calibration measurements were carried out. • The computer simulations were conducted in order to estimate conversion factor. • The activation technique were also used for verification of obtained results. [ABSTRACT FROM AUTHOR]

Abstract The main drawback of the double exposure holographic interferometry (DEHI) systems is that their initial setting up is a difficult and time-consuming task. This hinders their use in radiation dosimetry that requires observing necessary precautions to protect the users from radiation. In this paper, a detailed theoretical model capable of predicting the DEHI system response before its use for radiation dosimetry is developed which can partly overcome this limitation. In this model, radiation dosimetry by the DEHI systems is described in steady-state and transient modes. Physical background of hologram recordation and heat transfer within the absorbing material are fully described and it is shown how the whole process can be numerically modelled. Specifically, radiation dosimetry with two different homogeneous and inhomogeneous geometries is modelled in this work. For 3 MeV electrons in an inhomogeneous geometry, it is shown that the model results are in good agreement with the previous experimental results. It is also described how the model can be used to calculate the heat transfer correction factors in dosimetry by the DEHI systems. Highlights • A detailed model is developed for predicting the response of double exposure holographic interferometery (DEHI) systems. • In the model, radiation dosimetry by the DEHI systems is described in steady-state and transient modes. • The model is capable to account for the heat transfer effects in radiation dosimetry measurements by the DEHI systems. • For 3 MeV electrons, it is shown that the model results are in good agreement with the previous experimental results. • An approach is suggested for correcting for the impact of heat transfer in dosimetry experiments with the DEHI systems. [ABSTRACT FROM AUTHOR]

Abstract This paper discusses an instrument intercomparison performed in the high-energy field at the CERF facility at CERN between 2015 and 2017 (October 2015, May 2016, November 2016 and June 2017). Measurements were performed in several reference exposure locations with the CERN extended-range Bonner Sphere Spectrometer (BSS), a Berthold LB4611, the LINUS rem counter from CERN, the LUPIN rem counter from ELSE Nuclear, the FHT 762 Wide Energy Neutron Detection Instrument (WENDI-II) from Thermo Scientific, the LUDLUM MODEL 42–41 PRESCILA NEUTRON PROBE and two models of Tissue-equivalent proportional counters (TEPCs) from Far West Technologies. All results are compared with the latest FLUKA reference values from simulations performed in 2017. Highlights • The BSS , WENDI II, LINUS and LUPIN agree amongst them and with the FLUKA simulations within one sigma of uncertainty. • The LB6411 underestimates up to 35% as compared to FLUKA and the extended rem counters. • The PRESCILA overestimates the high-energy component and underestimate the evaporation part. Therefore, the PRESCILA overestimates the H*(10) on the CT and underestimates it on the IT. • Concerning the two TEPC , they agree within their range of uncertainty for all positions. The agreement with FLUKA is satisfactory on the CT/CS but the TEPCs show underestimation on the IT. [ABSTRACT FROM AUTHOR]

Recently, two independent electronic dosimeters for individual monitoring have been developed at the Helmholtz Center Munich, Germany, one for photons and one for neutrons. Calibration of these dosimeters was performed using mono-energetic neutrons (energy range: 144 keV–14.8 MeV) and photons (12 keV–1.3 MeV), and the resulting calibration curves were compared to Monte Carlo simulations using the GEANT4 tool kit. In the present paper, an integrated dosimeter prototype is described which combines both dosimeters and therefore allows measurement of total dose in mixed neutron/photon radiation fields. It is demonstrated that the neutron channel of the combined dosimeter is insensitive to photons, while the photon channel is insensitive to neutrons. The prototype was tested in the field of an Am-Be neutron source which also includes a photon component. When the Am-Be neutron source was shielded by 9 mm lead, the combined dosimeter indicated a dose rate of 122.6 ± 1.0 μSv/h from neutrons and 7.3 ± 0.1 μSv/h from photons. In contrast, when the Am-Be neutron source was not shielded by lead, the combined dosimeter indicated a dose rate of 125.8 ± 1.0 μSv/h from neutrons and 65.9 ± 0.1 μSv/h from photons indicating a significant contribution from photons to total dose. In a second experiment when the Am-Be neutron source was closer to the combined dosimeter and a 137 Cs source was added, a neutron dose rate of 421 ± 20 μSv/h and a photon dose rate of 849 ± 6 μSv/h was obtained. It is concluded that this instrument can be used in mixed photon-neutron fields with doses and dose rates typical for radiation protection purposes. [ABSTRACT FROM AUTHOR]

Identification of neutron sources is of interest in several fields, such as homeland security or basic nuclear physics. Organic scintillators offer the possibility to unfold fast-neutron energy spectra. Since organic scintillators are sensitive to neutrons and gamma-rays, a method to discriminate these two types of particles is required. This paper is focused on investigation of the discrimination properties of the filter method by using a new procedure. We compare the results of an experimental filter method to a digital pulse-shape discrimination (PSD) method based on charge integration. In addition, these methods are compared to the simulation results obtained with the MCNPX-PoliMi code. The experimental and numerical investigations were performed with a 252 Cf spontaneous-fission neutron source. The integration of the number of counts gave the relative differences between the experimental filter and digital PSD neutron pulse-height distributions (PHDs) and simulated PHDs less than approximately 5% in the range between 60 keVee and 1.715 MeVee of light output. Above 1.715 MeVee, the PSD method has advantages over the filter method, due to the filter method having significantly worse counting statistics. The results show that the filter method has potential for robust neutron measurements when the PSD method cannot be applied, such as for ‘old’ organic plastic scintillators without PSD capability. [ABSTRACT FROM AUTHOR]

Some drawbacks of monitoring radiation worker doses with personal dosimeters could be avoided by the introduction of computational dosimetry based on 3D cameras for tracking of workers, modelling the workplace and its radiation field, and dose calculation using computer simulations. The PODIUM (Personal Online DosImetry Using computational Methods) project was setup to demonstrate the feasibility of this approach. The goal of this paper is to demonstrate the feasibility of the computational dosimetry approach developed within the PODIUM project at a realistic neutron workplace field. A transport container with spent MOX fuel needles in a controlled area at SCK CEN was selected as a realistic neutron workplace field for this feasibility study. An MCNP6.2 model of this workplace was developed and successfully validated by comparing simulations with extensive measurements. After validation, the model was used to calculate an effective dose rate map of the workplace field. A single 3D camera was setup at the workplace to track the workers. Using a Python tool, the effective dose rate map, and the tracking file from the camera, the effective dose of the tracked worker could then easily be calculated. • Setup of an MCNP6.2 model of a neutron workplace field around a container with spent MOX fuel. • Validation of the MCNP6.2 model by means of an extensive measurement campaign. • Demonstration of computational staff dosimetry without personal dosimeters at a realistic neutron workplace field. [ABSTRACT FROM AUTHOR]

Rem-Counters are widely used in radiation protection practice for environmental neutron detection. They are usually based on both an active and a passive thermal neutron detector surrounded by a hydrogenate moderator designed to obtain a response function as close as possible to the H *(10) curve. Inclusion of high Z material is also used to extend the response function up to 1 GeV of neutron energy. Recently a dual detector Rem-Counter, both active and passive, was proposed by Agosteo et al. (2010) . The present paper focuses on the passive version that is based on a CR-39 track detector coupled to an enriched 10 B converter. A technique was developed for the track analysis based on the measurement of the mean LET of the charged particles produced by the neutron reaction on 10 B that has proven to be very effective in the background reduction. Moreover the simulated response function of the Rem-Counter was compared with the experimental results obtained with irradiations carried out with monoenergetic neutron beams, showing a quite good agreement. [ABSTRACT FROM AUTHOR]

Under standard conditions, several studies assess uncertainty values for individual dosimetry for photons, but seldom for neutrons. The Thermoluminescent Dosimetry Laboratory (LDT) of the Instituto de Radioproteção e Dosimetria (IRD), Brazil, has been running a neutron individual monitoring service using two different albedo monitors. This paper presents a study of the contribution of relevant sources of uncertainty for neutron dose evaluation, for both systems (called System 1 and System 2), using a reference 241 Am–Be field, at normal incidence. The combined and expanded uncertainties were calculated using GUM methodology and follows RP160 from the European Commission. This methodology was applied to calculate the uncertainties associated with the LDT assessment of neutron doses in the First Brazilian National Comparison on Measurements for Neutron Individual Monitors . The LDT participated in this comparison with its two systems, both presenting satisfactory performance. For System 1, at low neutron doses, the reproducibility of the apparent neutron dose is the more relevant source of uncertainty. However, for higher doses, the neutron calibration factor, NCF , becomes more important. For System 2, NCF is the main source of uncertainty for low and high doses. For occupational doses, the uncertainty will be much higher due to the need of additional correction factors, which depend on stray neutron field. [ABSTRACT FROM AUTHOR]

To accurately determine dose, albedo TLD dosimeters require application of a neutron energy correction factor (NECF). In this paper, the results from three different methods were used to determine NECFs. In the first method, NECF values were calculated as a ratio of the known "conventional true" dose and the neutron dose reported by albedo dosimeter(s) irradiated on a phantom. The results from this direct method were utilized for verification of two other methods. The second method used a specially designed device called the Neutron Area Monitor (NAM) model 5. The third method of NECF determination was based on the response of a BF 3 tube under different levels of moderation. The results of this work showed that the Navy-developed Neutron Area Monitor (NAM) model 5 measures NECF values correctly when irradiated with 252Cf (bare and D 2 O-moderated). This was an important result, because it validated the use of the NAM-5 in nuclear power and marine propulsion reactors, as their neutron energy spectrum is approximated by the field of a Cadmium-coated 30 cm diameter steel sphere of D 2 O moderated 252Cf. In the case of a broad neutron energy spectrum the 6Li response to the small amount of slow neutrons outweighs its response to the fast neutrons overwhelmingly responsible for the neutron dose rate. Based on the obtained results we can conclude that NAM-5 provides correct NECF determination if the fraction of slow neutrons in the neutron dose rate is more than 20%. A satisfactory correlation between the NECF determined by NAM-5 and the multi-moderated BF 3 tube method was observed, but the multi-moderated BF 3 tube values were ∼30% lower than those obtained by the NAM-5 at the same locations for broader neutron energy spectrum locations. • A neutron energy correction factor (NECF) for albedo dosimeters is introduced and defined. • Three different methods of NECF determination are suggested and discussed. • Pros and contras of the used approaches and causes of the identified discrepancies are discussed. [ABSTRACT FROM AUTHOR]

Abstract: The ANDANTE project began in January 2012 as a spin-off of results of an earlier Euratom project, ALLEGRO, designed to address the problems of medium- and long-term risks following radiotherapy. ANDANTE will investigate the relative risk of induction of cancer from exposure to neutrons compared to photons. The project will focus on three specific cancers that may be detected as second malignant neoplasms following paediatric radiotherapy: salivary gland, thyroid gland, and breast tissue. Stem cells from each of the types of tissue will be exposed to well characterised beams of both neutrons and photons. Biological markers of possible tumorigenesis will be used to develop RBE models for neutrons. The experimental beams will be measured in terms of fluence and energy spectra in order to provide date for a track structure model, which will be developed to simulate the exact experimental conditions and to explore the relationships between exposure parameters and response. The RBE model will be utilized in the assessment of follow-up data from paediatric photon radiotherapy patients. The out-of-beam mixed photon-neutron fields generated during proton therapy will be measured in phantoms, and an analytic algorithm will be developed for reconstructing the fields distant from the treatment site using data available from clinical records. One of the aims of the project is a critical analysis of the potential power of a multi-centre cohort of paediatric patients to give confidence in the neutron risk estimates in radiotherapy patients. This paper focuses on the challenges in the ANDANTE project posed by the need to know detailed neutron, photon, and charged particle fluences and energy spectra in order to determine the neutron RBE as a function of dose and energy. [Copyright &y& Elsevier]

Abstract: This paper presents in-phantom photoneutron equivalent doses induced by external radiotherapy, these are necessary to assess organ-averaged equivalent doses to derive the risk of secondary cancer development, associated with non-target organ exposures. The measurements were performed by Working Group 9 “Radiation Protection Dosimetry in Medicine” of the European Radiation Dosimetry Group (EURADOS) for simulations of clinical radiotherapy treatments of prostate cancer. The photoneutron dose measurements were carried out in a BOMAB-like water-filled phantom, by means of superheated emulsions (superheated drop detectors SDD and bubble damage BTI® detectors) and Poly-Allyl-Diglicol-Carbonate (PADC) solid state nuclear track detectors. Dose data were acquired in a three-dimensional matrix of reproducible measurement points, which are spaced according to detectors size. Four clinical protocols for the treatment of the prostate tumor were considered and compared, by measuring the doses delivered to the planning target volume (PTV) and to peripheral radiosensitive regions (i.e. colon-rectum and bladder). The clinical irradiations were performed in two clinical radiotherapy facilities based on Varian Clinac 2300 CD accelerator: Santa Chiara University Hospital (Pisa, Italy) and Centre of Oncology M. Skłodowska-Curie Memorial Institute (Krakow, Poland). An additional tomotherapy prostate cancer treatment was also simulated in Campo di Marte Hospital (Lucca, Italy). Radiation qualities of 6, 12, 15, 18 and 20 MV were used; all of these are capable of producing photoneutrons. Data from this work span most of the X-ray beam energies and prostate treatment modalities used in the current clinical practice. These data permit the assessment of doses absorbed by a radiotherapy patient either at the treatment volume or at out-of-field organs. Comparison of different dosimeters, under the same irradiation conditions, showed that dosimeters generally agreed within their 20% 1 SD uncertainty. Comparison of different treatment modalities in the two contributing clinical centers (Pisa and Krakow) were also possible, as well as a comparison of dose profiles resulting from the different treatment techniques, delivered at the same primary photon energy. It was in particular found that photon acceleration energies as low as 6 MV are able to produce a non-negligible photoneutron component, which causes an undue dose to the patient of the order of tens microsievert per unit photon dose delivered at the target volume. [Copyright &y& Elsevier]

Abstract: This paper describes the criteria behind the selection of neutron detection techniques for photoneutron dosimetry as well as the methods adopted to obtain dosimetric readouts. The work was conducted within the framework of Working Group 9 (WG9 Radiation Protection Dosimetry in Medicine), coordinated by the European Radiation Dosimetry Group (EURADOS). WG9 research aims at estimating the risk of second cancer induction due to radiation therapy. Therefore, a comprehensive experimental programme was devised to measure doses received by non-target organs-at-risk (OAR) during radiation therapy. The techniques described in this work were selected and used for the neutron dosimetric assessment during in-phantom simulations of clinical prostate radiotherapy treatments, carried out in three European facilities. Non-conformal standard fields were used as a common reference between different facilities. Performing neutron measurements near linacs is a complex task, because of the intense pulsed photon primary field. Therefore, photon insensitive dosimeters such as superheated emulsions (SE) and solid state nuclear track detectors (SSNTD) were chosen. Their readout procedures were carefully assessed. Methods were developed to count the large number of tracks and bubbles in SE. These are described in detail in the present work, along with a brief introduction to the detector physics. [Copyright &y& Elsevier]

Abstract: The first passive radon monitor was termed radon film-badge, in analogy with the neutron film-badge. This monitor, instead of a radon film-badge, should be considered the first diffusion chamber, the principle of which has been exploited in all existing passive radon-monitors. In the present paper, a true radon film-badge is proposed, which is formed by a thin-film radiator with suitable radon-sorption characteristics, facing a track detector. At the nanometer scale, the sorption in these radiators occurs by trapping the nanometers particles, formed by the radon atoms, into the nanoholes of technological advanced materials. When compared with existing passive radon-monitors, these radon film-badges present several advantages, such as a wide linearity of response, any desired response sensitivity, a unique simplicity in measuring radon in soil and in water. Radon-decay-products attach to airborne particles and thus acquire a wide range of sizes both in nanometer and micrometer scales. By employing a strategy similar to that used for neutrons, a very compact badge has been recently proposed, which shows promise for the solution of the complex problem of the radon-decay product dosimetry. [Copyright &y& Elsevier]

Abstract: Phototransfered thermoluminescence (PTTL) found its use in radiation dosimetry for dose reassessment primarily in photon radiation fields. After the dose has first been evaluated by a “normal” thermoluminescence (TL) measurement, the same detector is subjected to the PTTL dose reassessment procedure providing a new estimation by an independent measurement procedure. PTTL also gives a new possibility for a dose reassessment and for separate dose determination in the mixed fields. It was shown in our previous papers that it is possible to determine separately gamma and neutron dose components with a single detector due to different sensitivities of TL and subsequent PTTL to neutrons in comparison to gamma rays. The main disadvantage, the low sensitivity of both, TL and PTTL, to neutrons restricts the use only for the case of radiation accident. In this work we have investigated the possibilities of improvement the sensitivities of PTTL by heating the samples during UV irradiation and by using the sensitization technique involving a “pre-dose” followed by annealing. Two types of LiF:Mg,Ti detectors, TLD-100 and TLD-700, were examined by irradiation in 137Cs gamma ray field and with neutrons of 14.5 MeV. After UV irradiation at different temperatures, PTTL increases reaching the values 2.1 and 2.5 times higher at 90 °C than that at 25 °C for TLD-100 and TLD-700 respectively. Increasing level is approximately the same for gamma rays and neutrons. After sensitization, TL signal is one half of that before sensitization. PTTL signal has increased 11 times for gamma rays and 5 times for neutrons at 25 °C with significant increase with increasing temperature during UV irradiation. [Copyright &y& Elsevier]

Abstract: PADC detectors are commonly used both as radon detectors and as personal dosemeters for neutron radiation. In both cases the measurement in workplace fields are characterized by long term irradiations, lasting up to several months. During this period the detectors undergo to uncontrolled and unmonitored environmental conditions. In a recent work the same authors demonstrated that the environmental conditions, mainly temperature, can seriously affect the detector sensitivity. The temperature effect is to reduce the V ratio by decreasing the track etching velocity Vt . This causes a decrease in limit angle and efficiency. This paper describes the same ageing/fading test made with CR39 from a different supplier. In this case it seems that the temperature has a negligible effect, so no compensation algorithm is needed. [Copyright &y& Elsevier]

Abstract: A two-component thermoluminescent albedo neutron monitoring system was developed at Instituto de Radioproteção e Dosimetria, Brazil. As there is no Brazilian regulation for neutron individual monitoring service, the system was tested according to the ISO 21909 standard. This standard provides performance and test requirements for determining the acceptability of personal neutron dosemeters to be used for the measurement of personal dose equivalent, Hp(10), in neutron fields with energies ranging from thermal to 20 MeV. Up to 40 dosemeters were used in order to accomplish satisfactorily the requirements of some tests. Despite operational difficulties, this albedo system passed all ISO 21909 performance requirements. The results and problems throughout this characterization are discussed in this paper. [Copyright &y& Elsevier]

Abstract: The DIS-1 personal dosimeter is approved for assessment of personal dose H p(10) and H p(0.07) from γ and β radiation. Dose measurement in the DIS-1 is based on miniaturized ionization chambers. In mixed radiation fields with components other than photons and electrons, the dosimeter will not indicate personal dose, but the ionization chambers will produce a spurious signal that will add to the signal from γ and β radiation. The purpose of this paper is the evaluation of the magnitude of the spurious signal from other radiation types. For this purpose, irradiations of DIS-1 dosimeters were performed in a neutron calibration field and two mixed radiation fields at a particle accelerator. The results show, that the DIS-1 dosimeter indicates absorbed dose in arbitrary radiation fields. This property of the DIS-1 dosimeter can give critical information after accidental exposures or when investigating suspected overexposures. [ABSTRACT FROM AUTHOR]

Abstract: A novel radiation detector based on an artificial single crystal diamond was used to characterize in detail the energy distribution of neutron reference fields at the Physikalisch-Technische Bundesanstalt (PTB) and their contamination with charged particles. The monoenergetic reference fields at PTB in the neutron energy range from 1.5MeV up to 19MeV are generated by proton and deuteron beams impinging on solid and gas targets of tritium and deuterium. The energy of the incoming particles and the variation of the angle under which the measurement is performed produce monoenergetic reference fields with different mean energies and line shapes. Well established simulation codes allow these parameters to be calculated in detail, provided the properties of the targets are known. In this paper we present high resolution neutron spectrometry measurements of different monoenergetic reference fields. The results are compared with calculated spectra taking into account the actual target parameters. The influence of deviations from the ideal case, e.g. a non homogeneous tritium distribution in a solid Ti/T-target, was investigated. Line structures in the order of 80keV for a neutron energy of 9MeV were resolved. The shift of the mean energy and the increasing of the width of the neutron peak with increasing pressure in the gas target in the order of 30keV were measured. Another result is the determination of the contamination of the neutron field at 14MeV with high energy charged particles (protons) from side reactions inside the T-target. This effect is due to the thin backing of the targets in use at PTB. It depends on the age of the target and it has to be taken into consideration for irradiations at small distances for some detectors, especially when very old targets are used. The experiments have shown that this detector is an easy to operate compact neutron spectrometer with extremely good energy resolution and that detailed structures in the line shapes of monoenergetic neutron fields can be resolved without using time-of-flight techniques. [Copyright &y& Elsevier]

Abstract: In Japan, several monoenergetic and quasi-monoenergetic neutron reference fields are currently operating at several research institutes over a wide energy range 8 keV–390 MeV. These neutron fields are utilized for characterizing and developing neutron detectors as well as for neutron-induced cross section studies, shielding benchmark experiments and irradiation testing of micro-electronic devices. This paper reports the current status of these monoenergetic and quasi-monoenergetic neutron reference fields. [ABSTRACT FROM AUTHOR]

Abstract: Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward–backward (FB) and directionally coupled forward–backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light ion transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient. [Copyright &y& Elsevier]

Abstract: The development of automated track readers dates back to the first routine dosimetry applications of solid state nuclear track detectors. Many research-laboratory track readers and a few commercial systems were introduced in the past 25 years. The traditional application area for high-throughput track readers is radon testing in buildings and fast neutron dosimetry. The latter application, in particular, has considerably increased the need for efficient polyallyldiglycol carbonate (PADC) track reader systems with high throughput. The paper provides an overview of recent trends in the development of high-throughput PADC track readers and it gives a glimpse into the practice of designing this category of devices. A general overview is provided of the technical aspects relating to the optics, the hardware and the track recognition approaches. Ultimate theoretical limits in the performance as well as promising new approaches are addressed. Requirements and practical solutions are illustrated with reference to a commercially available system. [Copyright &y& Elsevier]

Abstract: The European Commission has funded within its 6th Framework Programme a three-year project (2005–2007) called CONRAD, COordinated Network for RAdiation Dosimetry. A major task of the CONRAD Work Package “complex mixed radiation fields at workplaces” was to organise a benchmark exercise in a workplace field at a high-energy particle accelerator where neutrons are the dominant radiation component. The CONRAD benchmark exercise took place at the Gesellschaft für Schwerionenforschung mbH (GSI) in Darmstadt, Germany in July 2006. In this paper, the results of the spectrometry using four extended -range Bonner sphere spectrometers of four different institutes are reported. Outside Cave A the neutron spectra were measured with three spectrometers at six selected positions and ambient dose equivalent values were derived for use in the intercomparison with other area monitors and dosemeters. At a common position all three spectrometers were used to allow a direct comparison of their results which acts as an internal quality assurance. The comparison of the neutron spectra measured by the different groups shows very good agreement. A detailed analysis presents some differences between the shapes of the spectra and possible sources of these differences are discussed. However, the ability of Bonner sphere spectrometers to provide reliable integral quantities like fluence and ambient dose equivalent is well demonstrated in this exercise. The fluence and dose results derived by the three groups agree very well within the given uncertainties, not only with respect to the total energy region present in this environment but also for selected energy regions which contribute in certain strength to the total values. In addition to the positions outside Cave A one spectrometer was used to measure the neutron spectrum at one position in the entry maze of Cave A. In this case a comparison was possible to earlier measurements. [Copyright &y& Elsevier]

Abstract: Extended-Range Bonner sphere spectrometers (ERBSSs) are well suited for measurements in high-energy stray neutron fields, where neutron energies extend from thermal to a few hundred MeV. These fields typically have a thermal peak, a fairly flat intermediate region, a medium energy peak at ∼1MeV and a high-energy peak at ∼ a few 100MeV. The data analysis is not straightforward: it requires unfolding with response functions that have a substantial amount of overlap, and the responses of the modified spheres (which provide information about the fluence at high energies) increase dramatically above ∼100MeV. In this paper, I try to determine which of the main features of the spectrum are well determined by the data given optimal methods of analysis, and which features are subject to ambiguity. To do this, I analyse ERBSS measurements using two methods of analysis that are particularly useful for this purpose, Bayesian parameter estimation and maximum entropy unfolding. [Copyright &y& Elsevier]

Abstract: An intercomparison of radiation protection devices in a high-energy stray neutron field was held in the framework of the COordinated Network for Radiation Dosimetry [CONRAD, http://www.eurados.org/conrad/conrad_overview.htm], financed by the European Commission. An irradiation hall of the GSI (Gesellschaft für Schwerionenforschung, Darmstadt, Germany) was the facility selected for this benchmark experiment. The measurements were carried out outside the shielding barriers of this facility, while bombarding a thick carbon target with a beam of 400MeV/u carbon ions. This paper describes the Monte Carlo simulations which were performed in order to characterize the radiation field at the measurement positions. Two codes were used for this purpose, namely FLUKA and MCNPX. Different approaches were adopted for the simulations and the results are compared and discussed. The results of the different simulations were in a satisfactory agreement. The comparisons with the measurements are discussed in Wiegel et al. [Intercomparison of radiation protection devices in a high-energy stray neutron field. Part II: Bonner sphere spectrometry. Radiat. Meas., 2008] and Silari et al. [Intercomparison of radiation protection devices in a high-energy stray neutron field. Part III: instrument response. Radiat. Meas., 2008]. [Copyright &y& Elsevier]

Abstract: In recent years voxel models have been used to investigate the validity of radiation protection quantities evaluated, till now, employing stylized reference models. An ICRP task group (DOCAL) is developing a male and female reference voxel models to estimate a new set of radiation protection quantities. It should be taken into account that, due the intrinsic subject-dependent anatomical differences, the various developed voxel models can imply a certain variability in the corresponding sets of estimated conversion coefficients. Therefore a qualitative evaluation of the range of such discrepancies should be of interest in terms of the sensitivity of organ absorbed dose to the specific model used for its computation. The NORMAN-05 model, updated at ENEA-Radiation Protection Institute on the basis of the original NORMAN model, developed by P. Dimbylow at NRPB-UK (now HPA) in the 1990s, is a good representation of the standard Caucasian-type man and could be a valuable tool for comparison with other voxel models in order to state the inherent uncertainties associated to the estimated sets of conversion coefficients. The present paper summarizes the results obtained with NORMAN-05 and the Monte Carlo code MCNPX for neutron conversion coefficients calculation, in the energy domain from thermal to 20MeV, in the kerma approximation that, for such energy range, as stated in the ICRU/ICRP recommendations, is a satisfactory approach. [Copyright &y& Elsevier]

Abstract: In the framework of the 2006 experimental benchmark organized at the GSI (Darmstadt, Germany) by the EC CONRAD network, a neutron dosimetry intercomparison was performed in a workplace field around a carbon target hit by 400MeV/u ions. The radiation protection group of the INFN-LNF participated to the intercomparison with a Bonner sphere spectrometer equipped with an active scintillator and a set of passive detectors, namely MCP-6s /MCP-7 TLD pairs from TLD Poland. Both active and passive spectrometers, independently tested and calibrated, were used to determine the field and dosimetric quantities in the measurement point. The FRUIT unfolding code, developed at the INFN-LNF radiation protection group, was used to unfold the raw BSS data. This paper compares the results of the active or passive spectrometers, obtaining a satisfactory agreement in terms of both spectrum shape and value of the integral quantities, as the neutron fluence or the ambient dose equivalent. These results allow qualifying the BSS based on TLD pairs as a reliable passive method to be used around high energy particle accelerators even in low dose rate areas. This is particularly useful in those workplaces where the active instruments could be disturbed by the presence of pulsed fields, large photon fluence or electromagnetic noise. [Copyright &y& Elsevier]

Abstract: This paper presents a measurement campaign carried out in six medical electron linear accelerator facilities to evaluate occupational exposure associated with photons and secondary neutrons for several encountered configurations. No strong correlation between dose and configuration has been observed. Assuming realistic hypotheses, the annual effective dose to staff due to external exposure has been estimated to 0.9mSv, 54% associated with the decay of activation products and 7% due to exposure to leakage neutrons. [Copyright &y& Elsevier]

Abstract: This paper demonstrates the possibility of using neutron converters incorporated into dosimeters to produce new neutron-sensitive OSL dosimeters suitable for personal monitoring. A comparison between different neutron converters indicate that lithium-based compounds (6LiF and result in the best neutron sensitivity, probably because of the relatively large range of the 2.75MeV triton produced by the neutron capture reaction. Based on these results, a new neutron-sensitive OSL composite material, identical to the one used in the Luxel™ and InLight™ dosimetry systems (Landauer Inc.) except for the smaller gamma sensitivity and enhanced neutron sensitivity, was produced and characterized. The gamma sensitivity of the new material is the sensitivity of the regular material (Luxel™), due to reduced amount of in its composition. The neutron sensitivity achieved in this new material is of the neutron sensitivity of 6LiF:Mg,Ti (TLD-600). [Copyright &y& Elsevier]

Abstract: We investigated the ESR response of new materials, alanine and ammonium tartrate to which gadolinium was added. The addition of gadolinium enhances sensitivity for -photons because of its high atomic number and an enhancement of sensitivity for thermal neutrons because of its high thermal neutron cross section and high linear energy transfer (LET) secondary particles produced after the reaction with neutrons. In particular, in this paper we analyzed the microwave power saturation properties of dosimeters of alanine and ammonium tartrate with or without gadolinium exposed to different LET beams. The power saturation trends of dosimeters exposed to photons and to thermal neutrons are different. For ammonium tartrate dosimeters the gadolinium addition intensifies significantly these differences and therefore, it improves the ammonium tartrate''s ability to discriminate the radiation LET . [Copyright &y& Elsevier]

Abstract: Measurement of the correct neutron fluence in various energy intervals in and around the neutron sources is important for the purpose of personnel and environmental neutron dosimetry. In the present paper we present a method for the measurement of neutron fluence using a CR-39 plastic track detector. The samples exposed to fission neutrons from a reactor with – fluence and to fast neutrons from a –Be source with a fluence of were etched for 14 time intervals starting from 10 to 570min in 6N NaOH at . The etched samples were exposed to the monochromatic light signal from an iodine–tungsten lamp and the transmittance was measured after each etching time interval using a UV Spectrophotometer. An exponential decay of the transmittance has been observed with increasing etching time. The behaviour of the fission neutrons having four different values of fluence has been observed to be distinctly different from that of the fast neutrons. It has also been observed that there is a linear relationship between the transmittance decay constant and fission neutron fluence. [Copyright &y& Elsevier]

Abstract: There are a number of etched-track neutron dosimetry systems in routine use for personal monitoring. In this paper, the operational and dosimetric characteristics of these systems are summarized. Brief details are given of the dosemeter design, the material used, its quality control procedures, background, processing and read methods, neutron energy range, energy and angle dependence of response, decision threshold, linearity, signal storage stability, calibration methods including normalization, effect of influence quantities, and the advantages and disadvantages of the systems in routine application. [Copyright &y& Elsevier]

Abstract: The BRADOS 1–3 projects were organised by the Russian Space Agency (RZA) between 2001 and 2003. The aim was to study the contribution of the primary galactic cosmic rays and of the secondary particles to the dose received by the crew of the International Space Station (ISS). Several laboratories participated in these experiments. Two different stacks (constructed by the team of the Atomic Energy Research Institute, AERI, Budapest, Hungary) composed of solid-state nuclear track detectors (SSNTD) were exposed inside the Service Module at different locations. The calibrations were made at the CERN high-energy neutron reference field named CERF (Geneva, Swiss). Applying a multiple track etching technique (2–20h etching time) and a sophisticated image analyser, the secondary neutron dose was deduced. The composition of stacks, the evaluation methods and the results will be presented here or referenced to previous papers. [Copyright &y& Elsevier]

Abstract: Polycarbonate (PC) detector is one of the common detectors for neutron and radon gas detection. Using this detector it is possible to measure the dose in mSv, by counting on an etched surface. In this paper, a special procedure has been suggested to determine the dose based on current drain during the etching process. In these experiments the effects of voltage, frequency, effective etched area, PC detector''s thickness, etched area (one side or two sides), etching solution temperature and dose absorbed by the PC foil have been studied. The results obtained show the current drain variation for a voltage of 200–1600V, a frequency of 2–10kHz, effective area with a diameter of 2–12cm, PC thickness of 125–250–375– and a temperature of etching solution of 25–. Lexan PC foils were exposed to doses of of neutrons. The unexposed foils were considered as the background (BG) foils. Most of the experiments were performed at a voltage of 800V, a frequency of 2 and 8kHz, foil thickness of , diameter of effective etched area of foils of 2, 6 and 12cm, temperatures of 25 and and the etching process from 0 up to overload stage. Overload stage occurs when the foil becomes so thin due to growth of the tracks that it leads to sparking between phase and null that makes a hole in the foil. Current drain curves versus the function of the etching time are absolutely different for various doses from zero (BG) to 10rad (BG up to ). This is true especially for the time interval from 3h of etching up to overload stage. In this way, it is possible to obtain a calibration of PC detector net current drain based on its absorbed dose. In this experiment, the number and diameter of tracks and their relation with drain current and PC foil residual thickness at overload stage have been studied. The same experiment has been performed for various concentrations of radon gas (Bq/m3) as well. [Copyright &y& Elsevier]

Abstract: This paper describes the fundamentals of a technique that enables the assessment of the uranium and thorium dose-rate component from fission tracks induced by neutrons. Apart from analysing elemental contents as low as g/g, the microscopic distribution of the alpha and beta dose-rates within petrologic sections can be spatially resolved down to . [Copyright &y& Elsevier]

Thermoluminescent detectors (TLDs) are widely used for the dosimetry of photons and electrons. They are less used for the radiation with higher linear energy transfer (LET). One of the reasons for that is that their TL relative efficiency η decreases for the most of them with increasing LET.The paper presents first a review of author''s experimental results in which η was established for charged particles having LET of the order from 1 to 100 keV/μm in tissue. Among TLDs studied were known materials like LiF:Mn; Ti; Al–P glass; CaSO4:Dy; Al2O3:Na; and Al2O3:C. It was found that the dependence of their η on LET is not the same for all TLDs studied.The response of the same materials to neutrons was also studied. It was found that both η as the relative response (RR) defined in terms of absorbed dose in tissue are different, they depend critically also on the composition. When a TLD contains nuclei like 6Li and 10B, their RR would be rather high. As far as η is concerned, the same tendencies were observed as for charged particles, i.e. when average LET of secondary particles formed in a TLD increases, their η generally decreases. [Copyright &y& Elsevier]

In the course of optical dating of a sequence of Dutch coversands, the distribution of equivalent doses was examined by applying the single-aliquot regenerative-dose protocol to small aliquots containing ∼0.5–1 mg of quartz grains. Broad Gaussian De distributions with relative standard deviations of 11–12% were observed. This was not expected, given the windblown nature of the deposits.This paper reports on a study on the causes of the broad dose distribution, using the quartz grains extracted from one of the samples from this coversand sequence. Investigations were made as to whether the spread might be related to the measurement procedures or to partial bleaching, but no evidence for these was found. The possibility of variations in microdosimetry was investigated by performing neutron activation analysis on small amounts (∼0.5–1 g) of sediment. A large scatter in U and Th concentrations throughout the profile was found. NAA of etched quartz extracts showed furthermore a non-negligible internal radioactivity. It is concluded that small-scale differences in the annual dose rate are the main cause for the observed distributions, although at this stage of our investigations it could not be established whether external or internal variations in microdosimetry are the dominant factor. The implications of our findings for dating these deposits are discussed in short. [Copyright &y& Elsevier]

The in-vessel radial shield configurations for a Prototype Fast Breeder Reactor (PFBR) consist of layers of steel and borated graphite/boron carbide with sodium as coolant. This in-vessel shield is provided to reduce the neutron flux at the intermediate heat exchanger so that secondary sodium activation in the heat exchanger can be kept to an acceptable level. It is necessary to optimise this in-vessel shield because an over design will increase the cost of the reactor and under design will increase secondary sodium activation. Mock up experiments to study neutron transport through various combinations of shield models consisting of layers of steel, sodium, borated graphite and boron carbide were carried out. In these shield model combinations, along with activation detectors, SSNT detectors were also kept to measure the reaction rate by the (n,f) technique. For this purpose, Lexan® polycarbonate track detector were used in combination with various fissionable nuclides such as natural uranium, thorium, neptunium, depleted uranium, etc., either in deposition form or foil form. The fission reaction rates per atom in various neutron energy regions were evaluated by counting the induced fission track density. These reaction rates are being interpreted using an appropriate computer code for unfolding the neutron spectrum. The paper describes the results obtained from the fission track data in the various sets of the experiments. [Copyright &y& Elsevier]

Neutron sources like 241Am–Be, 239Pu–Be, 252Cf and 14.6 MeV neutron generator are being used in oil exploration industries as well as in research institutions. While handling these neutron sources, personnel may be exposed to neutrons. Also personnel working in reactors, accelerators may receive dose from neutrons. These exposed individuals need to be monitored regularly for measurement of neutron doses. The individual neutron doses can be estimated by using Kodak NTA films and CR-39 Solid State Nuclear Track Detector with a polyethylene radiator to increase sensitivity in front in holder. Nearly 1450 personnel are being monitored regularly throughout the country on a quarterly basis. In India, the monitoring system adopted for individual neutron dose estimation having energy from 100 keV and above is described in this paper. Background counts of 0.20 mSv could be measured with CR-39 SSNTD foil system and it has been successfully introduced for Fast Neutron Personnel Monitoring for the country. [Copyright &y& Elsevier]

Extensive data have been gathered since the early 1990s on the response of different detectors based on the registration of neutron-induced fission in bismuth, gold, tantalum by the spark replica counter and the thin film breakdown counter. These detectors make it possible to exploit the excellent characteristics of the fission reactions in bismuth, gold and tantalum for the measurements of high-energy neutrons.Most of the investigations have been carried out at the quasi-monoenergetic neutron beam facility at The Svedberg Laboratory-TSL of the Uppsala University in cooperation with the Khlopin Radium Institute (KRI).The responses of different fission detectors in the intermediate range of neutron energy (35–180 MeV) have been evaluated: a region where the predictive power of available nuclear reaction models and codes is not reliable yet. For neutron energy greater than 200 MeV, the fission-detector responses have been derived from the data of the proton fission cross-sections.Finally, by using the ratio of the responses of these detectors, a simple and accurate way to evaluate the spectrum hardness can be obtained, thus providing a tool to obtain spectral information needed for neutron dosimetry without the need to know the entire spectrum.The experimentally measured spectra obtained to-date have different shapes and they are also different from those calculated.In this paper, a new approach will be reported to analyse the existing spectra by using response ratios of different detectors. Preliminary data have been already obtained for the high-energy neutron spectrum from the CERN concrete facility. [Copyright &y& Elsevier]

This paper reports the first attempt to use L-band spectroscopy for estimating the sensitivity of whole teeth to fast neutrons and gamma-rays. Three teeth were successively irradiated first with fast neutrons with a wide energy spectrum (mean energy around 30 MeV) up to ∼160 Gy and then with gamma-rays up to ∼14 Gy. After each irradiation, L-band (∼1 GHz) EPR spectra of each whole tooth surrounded by the surface–coil resonator were recorded, yielding a single composite line principally due to CO2− and native radicals. The sensitivities are estimated by the slopes of the linear dose response curves of the dosimetric CO2− radicals. The ratios of the gamma/neutron sensitivities were found to be in the range 8–9 (±2) for the three teeth. [Copyright &y& Elsevier]