Neutron Dose Research Articles

Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300μSv/h and 175-200μSv/h, respectively, when the reactor is operating at 10kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178keV. Apart from energy reduction, 10kW normalised neutron flux of south beam pit is also attenuated by the shield by 25000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

An Optimal Radiation Shield Design of a Small Modular Nuclear Power Reactor by Genetic Algorithm

One of the critical concerns in developing small modular nuclear reactors (SMRs) is employing efficient radiation shielding consistent with the design and implementation requirements. Besides observing the radiation shielding requirements, the design process entails developing compact, lightweight shielding as well as ensuring the safety of the staff and radiation-sensitive equipment around the reactor under different operating conditions. Conventional methods of radiation shielding design are based on experiments conducted. Therefore, the resultant design involves trial and error, rarely achieving an optimal, effective, and efficient design. In this study, a multi-objective method based on the genetic algorithm coupled with the MNCP calculation code has been used to enhance the radiation shielding system of a SMR. Using this method, the thickness of different shielding layers has been optimized to minimize the total dose (neutrons and gamma) at the output, the weight, and the overall volume of the shielding. To assess and compare the method used to an implemented conventional method, a sample design of the MRX nuclear power reactor is considered. Using the optimization technique, the radiation shielding of the reactor has been redesigned and compared with the original design. This paper discusses the existing deficiencies in the initial design of the primary shield of the reactor as a design implemented based on conventional methods. Similar to the original design, the main materials considered for radiation shielding in consecutive layers were water and steel, and the final layer contained lead. The thickness of the different layers and their arrangements have been optimized to assess the upgraded method. The calculations indicate that the overall thickness of the proposed shielding is suggested as 93.8 cm compared to 140 cm to obtain the total dose of neutrons and gammas, which is suggested as less than 10 µSv/h. The results imply a 38.56% reduction in volume and a 17.24% reduction in weight of the radiation shielding compared to the reference reactor design.

Cytogenetic validation of DS02R1-estimated dose for atomic bomb survivors in Hiroshima and Nagasaki with FISH

Introduction For Hiroshima and Nagasaki survivors, it has not been possible to calculate individual doses from the cytogenetic data and compare them with the physically estimated doses. This is because the cytogenetic studies used solid Giemsa staining which only provides the percent of cells bearing at least one stable-type aberration (most of the unstable-type aberrations had already disappeared), and a gamma-ray dose plus a 10-times neutron dose was used to integrate the data for both cities. Objectives To compare the FISH-derived gamma-ray dose with the DS02R1-derived gamma-ray dose after correcting for a contribution of the neutron dose. It was also an attempt to determine if the frequency of stable-type aberrations had remained unchanged after the exposure. Methods Stable exchange-type aberration data was obtained using the 2-color FISH method from 1,868 atomic bomb survivors in Hiroshima and Nagasaki. The collected frequency was first extended to a genome-equivalent frequency. Then, by using known induction rates of exchange-type aberrations in vitro caused by neutrons and gamma-rays, respectively, and the mean relationship between the neutron and gamma-ray doses in the DS02R1 estimates for the survivors, the gamma-ray effect was estimated from the total yield of translocations. Results It was found that over 95% of individual cytogenetic gamma-ray doses fell within the expected range of plus/minus about 1 Gy from the DS02R1 dose and the mean slope for the linear regression was 0.98, which reassures us of the validity of the DS02R1 study. Conclusions The present results demonstrate the validity of the individual DS02R1 doses, and that the frequency of stable-type aberrations in blood lymphocytes did not decay over the years, and thus is useful for retrospective dose evaluations of exposures which took place in the distant past.

Molecular mechanisms of neutron radiation dose effects on M1 generation peas

The dose effect of radiation has long been a topic of concern, but the molecular mechanism behind it is still unclear. In this study, dried pea seeds were irradiated with 252Cf fission neutron source. Through analyzing the transcriptome and proteome of M1 generation pea (Pisum sativum L.) leaves, we studied the molecular rule and mechanism of neutron dose effect. Our results showed three important rules of global gene expression in the studied dose range. The rule closely related to the neutron absorbed dose at the transcription and translation levels is: the greater the difference in neutron absorbed dose between two radiation treatment groups, the greater the difference in differential expression between the two groups and the control group. We also obtained important sensitive metabolic pathways of neutron radiation, as well as related key genes. Furthermore, the overall molecular regulation mechanism of dose effect was revealed based on the main functional items obtained. Our research results can be applied to appropriate radiation dose estimation and agricultural production practice.

Photon and Neutron Dose Estimation Using Monte Carlo Simulation in TrueBeam's Room.

The distribution of neutron ambient dose equivalent within the TrueBeam 10 MV photon chamber was investigated. The research used particle and heavy ion transport code system (PHITS) code and JENDL-5.0 to simulate the neutron ambient dose equivalent on and around TrueBeam's head. The simulated results were compared with the measured results using CR-39 detectors when TrueBeam radiated 5000 monitor units of 10 MV photons with field sizes 20 cm × 20 cm and 0.5 cm × 0.5 cm. Out of field size, the neutron ambient dose equivalents of the 0.5 cm × 0.5 cm field size are higher than those values of the 20 cm × 20 cm field size from 4% to 30%. The differences between the simulated value and the measured value of the neutron ambient dose equivalents at all points out of field size are smaller than 20%. The neutron ambient dose equivalents, simulated with PHITS and JENDL-5.0, are satisfied with the measured neutron ambient dose equivalent.

Using ANN for thermal neutron shield designing for BNCT treatment room

Occupational radiation protection should be applied to the design of treatment rooms for various radiation therapy techniques, including BNCT, where escaping particles from the beam port of the beam shaping assembly (BSA) may reach the walls or penetrate through the entrance door. The focus of the present study is to design an alternative shielding material, other than the conventional material of lead, that can be considered as the material used in the door and be able to effectively absorb the BSA neutrons which have slowed down to the thermal energy range of <1\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$< 1$$\\end{document} eV after passing through the walls and the maze of the room. To this aim, a thermal neutron shield, composed of polymer composite and polyethylene, has been simulated using the Geant4 Monte Carlo code. The neutron flux and dose values were predicted using an artificial neural network (ANN), eliminating the need for time-consuming Monte Carlo simulations in all possible suggestions. Additionally, this technique enables simultaneous optimization of the parameters involved, which is more effective than the traditional sequential and separate optimization process. The results indicated that the optimized shielding material, chosen through ANN calculations that determined the appropriate thickness and weight percent of its compositions, can decrease the dose behind the door to lower than the allowable limit for occupational exposure. The stability of ANN was tested by considering uncertainties with the Gaussian distributions of random numbers to the testing data. The results are promising as they indicate that ANNs could be used as a reliable tool for accurately predicting the dosimetric results, providing a drastically powerful alternative approach to the time-consuming Monte Carlo simulations.

Secondary neutrons in proton and light ion beam therapy: a review of current status, needs and potential solutions

The advantages of proton and light ion beam therapy compared to conventional photon radiation therapy are well-known, mainly thanks to the characteristic depth dose distribution of ions and their radio-biological effectiveness. Nevertheless, the use of ions implies different nuclear reactions that generate secondary particles, with neutrons among them. These secondary neutrons can travel far away from the treatment volume, their measurement is a challenging complex task, and their biological effects are particularly high for neutrons with energies in the MeV range. In this review, a comprehensive description of secondary neutron dosimetry in proton and light ion beam therapy is given. Many studies have been conducted on the quantification of the secondary neutron dose, most of them have been performed for proton beams, whereas for other ions like carbon, the available information is scarce. The main measurement campaigns are summarised, focusing on the type of detectors used. In line with the detectors’ advantages and limitations, measurements performed inside and outside anthropomorphic phantoms are considered. The role of Monte Carlo radiation transport simulations is discussed since many experimental detection techniques need additional simulations to provide dose estimates. A focus on the current challenges for the measurements of neutrons with energies above 20 MeV is given, as this is one of the main components of secondary neutrons produced by therapeutic ion beams. Finally, the potential clinical relevance of the available and needed secondary neutron dose data is discussed, in terms of its impact on the treatment of patients. For this, the relative biological effectiveness of neutrons and the potential risk of cancer induction re-incidence or secondary cancer due to secondary neutron doses play a key role.

Photon and neutron dose evaluation at the Beam Test Facility of the INFN - National Laboratory of Frascati

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 DAΦNE 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 DAΦNE 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 5×1010 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.

Dose equivalent consideration from neutron contamination in modified radiotherapy vault: a Monte Carlo study

Laminated barriers incorporating metal sheets provide effective protection for space-restricted radiotherapy centers. This study aimed to assess photoneutron contamination in smaller vaults protected by different compositions of multilayer barriers during simulated pelvic radiotherapy with 18 MV photon beams. Monte Carlo Simulations of 18 MV LINAC (Varian 2100 C/D) and Medical Internal Radiation Dose (MIRD) phantom were used to assess photoneutron contamination within reconstructed vaults incorporating different combinations of metal sheet and borated polyethylene (BPE) during pelvic radiotherapy. The findings highlight a 3.27 and 2.91 times increase in ambient neutron dose Hn* (10) along the maze of reconstructed vaults that use lead and steel sheets, respectively, compared to concrete. The Hn* (10) outside the treatment room increased after incorporating a metal sheet, but it remained within the permissible limit of 20 μSv/week for uncontrolled areas adjacent to the LINAC bunker, even with a workload of 1000Gy/week. Neutron equivalent doses in the patient’s organs ranged from 0.22 to 0.96 mSv Gy−1. There is no notable distinction in the organ’s neutron equivalent dose, fatal cancer risk, secondary radiation-induced cancer risk, and cancer mortality for various laminated barrier compositions. Furthermore, the use of metal sheets for vault wall reconstruction keeps the variation in cancer risk induced by photoneutrons below 6%, while risks of fatal cancer and cancer mortality vary less than 11%. While the metal portion of the laminated barrier raises the neutron dose, the addition of a BPE plate reduces concerns of increased effective dose and secondary malignancy risk.

Compact fusion blanket using plasma facing liquid Li-LiH walls and Pb pebbles

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 1019 n/cm2, and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH2), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements.

Practical Approach to Nuclear Shelter Wall Thickness Estimation

Abstract In the complex situation in which we live, preventive measures are increasingly being taken against the threat of nuclear attack and ionizing radiation, nuclear shelters are being built for private and public purposes. The purpose of the authors is to provide a practical approach to quickly calculate the thickness of the concrete shelter wall for such cases. The shelter wall thickness is assessed on the basis of published data on radiation dose calculations, corroborated by field data from nuclear tests, and information or calculations of the dose attenuation in the shelter wall. Doses from nuclear tests or simulations are summarized in tables in relation to bomb yield and distance from the explosion point. A dose criterion is chosen as the acceptable dose behind the shield. Neutron dose transmission factors, defined as ratio of the dose criterion to the dose without shielding are calculated. The concrete wall thickness corresponding to the neutron dose transmission factor was taken from reference sources. For gamma rays, the shield thickness is calculated based on an analytical relationship between dose without shield, the dose criterion, and the shield thickness. Data tables are provided with the estimated concrete wall thickness in relation to bomb yield and distance.

Effect of fibers and boron carbide on the radiation shielding properties of limestone and magnetite aggregate concrete

Research on new composite materials in the field of radiation shielding materials requires adequate means and meticulousness in their manufacture and characterization. In this study, concretes that use elements that theoretically improve their shielding properties are developed, such as fibers and boron-rich additions. Both magnetite and limestone concrete have been manufactured, incorporating steel fibers to improve the linear attenuation coefficient and polyvinyl alcohol to increase neutron scattering cross section. The incorporation of boron carbide is intended to increase the neutron absorption cross section. 5 mixes have been manufactured that have been tested with several gamma radiation sources (570–1280 keV) and with an Am–Be neutron source (111 GBq). The experimental results have been used to validate simulations carried out with MAVRIC® and have led to the modelling of a simplified spent nuclear fuel cask. Boron carbide produces a decrease of more than 50% on the neutron dose rate in contact with the cask. Incorporating PVA fibers supposes gain in neutron attenuation capacity of up to 13%, while fibers have a negligible effect on the photon dose rate.

Multiple alanine pellet dosimeter: Design and first test results

Existing nuclear accident dosimeters are based on the activity measurements of the neutron activation products. Those dose measurements require the use of expensive counting equipment. The other significant problem of existing accident dosimeters is that they need to be analyzed in a very short period, due to the short half-life of the activated products. The crystalline L-α-alanine is a well-known material for Electron Paramagnetic Resonance (EPR) dosimetry. It is widely used in radiation laboratories for both reference and transfer dosimetry. Alanine pellets are made from L-α-alanine powder mixed with 10–20% of a binder. The L-α-alanine pellets have ideal dosimetric properties. Alanine is sensitive to alpha, photon, and neutron irradiation. Until recently there was no way to distinguish separate dose contributions from multiple types of radiation. To overcome this shortfall, the multiple alanine pellet holder has been proposed, designed, and tested. It contains cadmium, lead, and 6LiF filters. The proposed approach allows for the separation of photon and neutron dose contributions. Use of the developed holder can also reduce negative effects of humidity and direct sun light on the alanine dosimetric properties. The developed holder design can be produced by a 3D printer at a low cost. Here we present and discuss the design and test results of the multiple alanine pellet dosimeter.

Neutron dose evaluation inside high-energy accelerator irradiation vault using LR-115

Recently, a new radiation therapy using a high-energy accelerator irradiation vault has attracted significant attention. This therapy is very effective in destroying cancer cells because it uses much higher energy than conventional radiation therapy. Nevertheless, it also has disadvantages due to its high energy and dose, resulting in the generation of secondary radiation such as X-rays and neutrons. In particular, neutrons have a higher radiation weighting factor than photons; therefore, they are more harmful to normal tissues. However, the popular neutron dosimeter CR-39 cannot evaluate high-dose neutrons.This paper proposes a novel method for measuring high-dose neutrons generated during cancer treatment. LR-115, which has low detection efficiency but is expected to be useful for evaluating high-dose neutrons, was used to evaluate the effective dose of neutrons in a high-energy accelerator irradiation room. The results were verified using MCNP 6.2, a Monte Carlo-based particle transport code. This study provides the lower limit for evaluating neutron effective dose using LR-115. The experimental results showed that the neutron effective dose increased linearly with beam current. These findings provide basic data for evaluating neutron effective dose for radiation protection.

Design and testing of an irradiation room with low room-scattering for neutron calibration

IntroductionThis laboratory plans to establish neutron reference radiation fields with three neutron sources to calibrate neutron-measuring devices. To perform calibration at multiple dose rates, neutron ambient dose equivalent rate H˙*(10) needs to range 1 μSv/h to 10 mSv/h. The lower limit requires that the maximum available calibration distance should be at least 4.5 m. MethodsTo reduce room-scattered neutrons and extend the available calibration distance, MC simulations were conducted to determine the material and size of the irradiation room. A 3″ Bonner sphere and a LB6411 environmental neutron dosimeter were used to characterize the irradiation room. ResultsA 14.32 × 14.32 × 12.00 m3 irradiation room was built based on simulation results. Floor, roof, and walls are made of 75 cm concrete covered by a coating layer of 2 cm BPE and 3 cm PE. Experimental maximum available calibration distance reaches 4.65 m. The range of H˙*(10) for calibration covers 1 μSv/h to 10 mSv/h. Neutron and photon H˙*(10) outside the room are within 0.19 μSv/h and 0.22 μSv/h, respectively.

Strength of bulk aluminum-boron alloys containing helium produced by 10B(n,α)7Li reaction in nuclear reactor

The study of metals and alloys containing helium has garnered significant attention within the nuclear energy community. However, there is limited research on the mechanical behavior of bulk alloys implanted with helium. This study investigates the mechanical properties of several Al-Boron alloys implanted with helium using controlled manipulation of helium doses via boron content under a consistent neutron dose. Results show that HemVn may contribute to strength by approximately 8.4–15 MPa and 16.8–23 MPa for helium doses 3.08 × 1019/cm3 and 6.84 × 1019/cm3, respectively, while lattice damages due to neutron-aluminum reaction contribute to strength by 24∼27 MPa. Subsequent annealing leads to the formation of helium bubbles, resulting in a slightly higher strengthening effect compared to HemVn. Additionally, the work hardening behavior of the alloys can be explained by the Voce model, drawing inspiration from the resemblance between helium bubbles and nanoprecipitates in 7xxx alloys. These findings provide insights to the nuclear energy community.

Neutron irradiation-induced shutdown dose rate estimation in EU DEMO divertor

The EU DEMO fusion reactor, which is part of the EUROfusion PPPT program and the successor to ITER, aims to facilitate the transition of fusion from the science laboratory to industrial use. Its key objective is net energy production. Its critical component, the divertor, manages power exhaust, ash, and helium removal. The key elements of the divertor include Plasma Facing Components (PFCs), shielding liners, reflector plates, divertor cassettes, and a vacuum vessel locking system. It is imperative that the shutdown dose rates are assessed to ensure workers' radiation protection during maintenance, plan the disposal of radioactive material, and ensure the overall safety of the plant. The shutdown dose rate tool employed in the present study is R2Smesh, which couples MCNP for radiation transport and FISPACT for neutron activation to calculate neutron-induced shutdown dose rates, utilizing an advanced, semi-heterogeneous EU DEMO divertor model, presenting a challenging task in neutron transport calculations.

Extended analysis of solid cancer incidence among nuclear industry workers in the UK 1955-2011: comparison of workers first hired in earlier and later periods.

To address points arising from the recent study of nuclear workers in the USA and the International Nuclear Workers Study (INWORKS), concerning the difference in solid cancer risk estimates between those first hired in earlier and later calendar years, subsidiary analyses were conducted on a cohort of 172 452 workers in the National Registry for Radiation Workers (NRRW) from the UK. A total of 18 310 incident first primary solid cancer cases were registered in the period from 1955 until 2011 in the NRRW cohort and workers accrued 5.25 million person-years of follow-up. Incidences rates of all solid cancers combined, lung cancer and solid cancer excluding lung cancer were examined in terms of external radiation doses in the full cohort and in a sub-cohort of workers who had no record of internal exposure monitoring and were defined by the periods of first hire before and after the beginning of the years 1960, 1965 and 1970. All analyses were carried out using Poisson Regression. These analyses demonstrated that only for lung cancer between the pre-1965 and post-1964 periods is there strong evidence for a difference in the risks using the NRRW full cohort. In the other calendar period breakdowns and for the other cancer groups, there is no clear evidence of differences in the risks. The NRRW estimation of risks between recent and early workers is not generally consistent with the US workers cohort or the INWORKS evaluations that later hired workers are at much higher solid cancer risk than earlier hired workers, although INWORKS contains a significant part of the latest updated NRRW cohort as well as the US data. The conclusion that the INWORKS and US study data demonstrate a real difference in excess solid cancer risk from external radiation exposure between earlier and later workers is premature. The results presented here should also be treated with caution because of the limited corroborating evidence from other published studies. Information on internal doses, neutron doses as well as non-radiation factors such as smoking and asbestos exposure would be needed to make definitive inferences.

Radiation shielding properties of composites of TiZrNbHfTa refractory high entropy alloy reinforced with TiZrNbHfTaOx high-entropy oxide

Radiation shielding materials play a critical role in advancing the next generation of nuclear reactors. Among these, high-entropy materials with refractory properties emerge as a new category for radiation shielding applications. This study introduces TiZrNbHfTa-TiZrNbHfTaOx composites tailored for such purposes. Initially, a TiZrNbHfTa refractory high-entropy alloy (RHEA) is synthesized via mechanical alloying, followed by the oxidation of the alloy to obtain its refractory high-entropy oxide (RHEO), TiZrNbHfTaOx. Composites are then prepared by blending 5%, 10%, and 20% weight fractions of RHEO with RHEA using mechanical alloying. Comparative analysis of shielding properties reveals a decrease in mass attenuation coefficient and effective atomic number with increasing RHEO content. While RHEA exhibits superior shielding performance against gamma photons and neutrons, the composite containing 20% RHEO emerges as the optimal shield against alpha and proton irradiations. The equivalent neutron dose rate and removal cross-section values are obtained 49.8% and 0.142 for the composite with 20% RHEO, respectively. These findings, supported by computational simulations in this study, contribute valuable insights for the advancement of novel shielding materials crucial for future nuclear reactor technologies.

Measurements of the neutron and gamma dose via TL/OSL in LiF:Mg,Ti and activation methods analysis of aluminum foils

The measurement of beta/gamma and neutron dose components in mixed radiation fields is described involving combined OSL and TL measurements. The accuracy of the technique is investigated by irradiation of LiF:Mg,Ti samples with 14.1 MeV neutrons and a comparison of the OSL/TL results with the neutron fluence determined by Al activation. The precision of the measurements is 6.1 to 6.7 % (1 SD) for the OSL/TL technique compared to 7.3 % for the Al foils. The advantage of the technique is the simultaneous determination of the two dose components on the same samples of 7LiF:Mg,Ti in a single dedicated TL/OSL reader.