Neutron Dosimetry Research Articles

Measurement of the 117mSn, 125mTe, 135mBa and 195mPt half-lives for reactor dosimetry application

Nuclear isomers are investigated to perform precise epithermal neutron dosimetry. One key physical property for reactor dosimetry is the precise knowledge of the isomer half-life. In the list of potential candidates, the half-life values available in the literature or in the recommended database for reactor dosimetry for 117mSn, 125mTe, 135mBa and 195mPt show some discrepancy. New half-life measurements are presented in this work. Those isomers were produced by neutron activation at the JSI TRIGA Reactor from tin, tellurium, barium (all three of them enriched in their parent of interest) and natural platinum dosimeters. Precise half-lives measurements were performed at the MADERE platform using gamma and X-ray spectrometry. Special care has been taken on the uncertainty determinations.

Biomarker dosimetry of acute low level of thermal neutrons and radiation adaptive response effect on rats

In this paper, we demonstrated the biological effects of acute low-dose neutrons on the whole body of rats and investigated the impact of that level of neutron dose to induce an in vivo radio-adaptive response. To understand the radio-adaptive response, the examined animals were exposed to acute neutron radiation doses of 5 and 10 mSv, followed by a 50 mSv challenge dose after 14 days. After irradiation, all groups receiving single and double doses were kept in cages for one day before sampling. The electron paramagnetic resonance (EPR) method was used to estimate the radiation-induced radicals in the blood, and some hematological parameters and lipid peroxidation (MDA) were determined. A comet assay was performed beside some of the antioxidant enzymes [catalase enzyme (CAT), superoxide dismutase (SOD), and glutathione (GSH)]. Seven groups of adult male rats were classified according to their dose of neutron exposure. Measurements of all studied markers are taken one week after harvesting, except for hematological markers, within 2 h. The results indicated lower production of antioxidant enzymes (CAT by 1.18–5.83%, SOD by 1.47–17.8%, and GSH by 11.3–82.1%). Additionally, there was an increase in red cell distribution width (RDW) (from 4.61 to 25.19%) and in comet assay parameters such as Tail Length, (from 6.16 to 10.81 µm), Tail Moment, (from 1.17 to 2.46 µm), and percentage of DNA in tail length (DNA%) (from 9.58 to 17.32%) in all groups exposed to acute doses of radiation ranging from 5 to 50 mSv, respectively. This emphasizes the ascending harmful effect with the increased acute thermal neutron doses. The values of the introduced factor of radio adaptive response for all markers under study reveal that the lower priming dose promotes a higher adaptation response and vice versa. Ultimately, the results indicate significant variations in DNA%, SOD enzyme levels, EPR intensity, total Hb concentration, and RDWs, suggesting their potential use as biomarkers for acute thermal neutron dosimetry. Further research is necessary to validate these measurements as biodosimetry for radiation exposure, including investigations involving the response impact of RAR with varied challenge doses and post-irradiation behavior.

Neutron dosimetry at HK-1 beam line of LVR-15 reactor for biomedical sample irradiations

Neutron dosimetry at HK-1 beam line of LVR-15 reactor for biomedical sample irradiations

Evaluation of neutron dosimetry capabilities with the MC-15 portable multiplicity counter

This work proposes a preliminary neutron dose rate estimation method for a neutron multiplicity detector through measurement- and simulation-based analyses. Uncharacterized neutron-emitting sources may be encountered in situations such as nuclear emergency response, safeguards, and treaty verification. These circumstances may present irradiation risk to personnel conducting field assay, search, and characterization measurements. It is therefore of interest to provide a field neutron dosimetry capability with the existing neutron multiplicity counting (NMC) capabilities. To date, no commercially-available neutron detection systems are capable of both accurate NMC and real-time neutron dosimetry. This work will focus on estimating dose rate using input from a single fielded NMC called the MC-15. The energy-dependent neutron detection efficiency response of the MC-15 was quantified in monoenergetic neutron simulations and evaluated in response to two neutron-emitting sources and to a polyethylene-moderated source. The results were compared to existing neutron dosimeters and established the proof of concept for further investigation of the MC-15 for dose estimation. Measurement results were also replicated in simulations; additional simulations were then conducted to expand upon the limited empirical data. The initial empirical results provided a conversion factor appropriate for use when measuring 252Cf neutrons that is independent of polyethylene shielding presence and thickness. The simulated data sets were then used to evaluate a fit equation allowing estimation of the neutron dose rate for less restricted geometric configurations and dependent only on the distance between the source and the detector. Additionally, the energy dependence of the efficiency response indicates that further empirical evaluations could provide energy-dependent conversion factors for broader neutron dosimetry capabilities with a wider range of neutron-emitting sources.

Generation of neutron spectra and emission yields from proton-induced reactions on natC and natLi targets using CR-39 detectors

Fast neutron spectrometry and dosimetry at a specified location in a compact radiation environment is difficult due to large size of the conventional neutron spectrometers. In such situations, a small-size bare CR-39 detector can be used as a neutron detector to quantify the fast neutron fraction. In the present study, CR-39 detectors were placed very close to natural lithium and carbon targets during the proton irradiations to acquire the emission neutron spectra. The neutron spectra and the ambient dose equivalents in the forward and lateral directions with respect to the incident proton beam were estimated for both targets at incident proton energies between 8 and 20 MeV. The measured neutron spectra using CR-39 detectors could identify the quasi mono-energetic neutron features from the Li(p,n) and 13C(p,n) reactions effectively. An important observation of the present study is the identification of the fast neutron signature from the 13C(p,n) system from the discrete state de-excitations of excited 14N composite nuclei. The theoretical evaluation of the neutron spectral features and relative neutron energy distributions were performed using the FLUKA: FLUktuierende KAskade, a Monte Carlo simulation package, and the estimates agreed with the experimental results for both systems. The neutron ambient dose equivalents were also estimated from the measured spectra. These neutron fluence and dose estimates at close vicinity to the target can serve as an essential basis for shielding calculations and planning the pertinent radiation protection strategies.

Performance of NB-CTMFD Detector Versus Ludlum 42-49B, and Fuji NSN3 Detectors for Hard (Am-Be) and Soft (Cf-252 Fission) Energy Spectra Neutron Sources Within Lead/Concrete Shielded Configurations

Abstract This paper presents the results of neutron detection efficiency and dosimetry between a nonborated centrifugally tensioned metastable fluid detector (NB-CTMFD) configured to detect fast (>0.1 MeV) versus two widely used combined thermal and fast neutron detection devices of similar form factors: moderated He-3 based Ludlum-42-49BTM neutron detector, and, the Fuji NSN3TM pressurized nitrogen-methane filled neutron detector. The one-on-one performance comparisons were conducted using hard (Am-Be) and soft (Cf-252 fission) neutron spectra isotope neutron sources positioned behind various levels of lead and concrete shielding ranging in thickness from 0 cm (unshielded) to up to 30 cm. The comparisons were conducted together with Monte Carlo N-Particle (MCNP) code simulations to account for 3-D effects and to relate the detection sensitivity with the dose rate. The MCNP model results for the neutron energy spectrum were validated versus experimental measurements using the H*TMFD. While the Ludlum and NSN3 detectors operate at a fixed sensitivity setting, the NB-CTMFD detector could be operated at various sensitivity levels by varying the tensioned metastable negative pressure (Pneg) from 0.3 MPa to 0.7 MPa. The NB-CTMFD (configured for fast: >0.1 MeV neutron detection at Pneg = 0.7 MPa) offered relative sensitivity enhancements of up to 15x greater than the Ludlum and ∼5x greater than the NSN3 detector for low shielding thicknesses. For larger thicknesses, the advantage factor for NB-CTMFD unit reduces with increased fast neutron removal via down-scattering, which benefits the Ludlum and NSN3 detectors. Our companion paper compares performance with a Cf-252 spectrum using a borated CTMFD (covering dual energy bins: thermal-epithermal and fast energy ranges) wherein, it is demonstrated that the advantage factor for the CTMFD (if borated) remains elevated at the 5–10x higher level for low and large shielding thicknesses and soft and hard neutron spectra.

Performance of Borated Centrifugally Tensioned Metastable Fluid Detector Detector Versus Ludlum 42-49B, Fuji NSN3 Detectors for Fission Energy Spectrum Neutron Detection With the Source Within Lead/Concrete Shielded Configurations

Abstract This paper presents the results of neutron detection efficiency and dosimetry between a borated centrifugally tensioned metastable fluid detector (B-CTMFD) configured to detect thermal to fast neutrons versus two widely used neutron detection devices of similar form factors: moderated He-3 based Ludlum-42-49B neutron detector, and, the Fuji Electric's NSN3TM (NNSN3) pressurized nitrogen-methane filled neutron detector. The one-on-one performance comparisons were conducted using soft (Cf-252 fission) neutron spectrum isotope neutron sources positioned behind various levels of lead and concrete shielding ranging in thickness from 0 cm (unshielded) to up to 30 cm. The comparisons were conducted with Monte Carlo N-Particle transport code (MCNP) code simulations to account for three-dimensional effects and to relate the absolute detection rate with the dose rate—to derive the sensitivity factor (cpm/μSv/h). While the Ludlum 42-49B and NSN3 detectors operate at a fixed sensitivity setting, the centrifugally tensioned metastable fluid detector (CTMFD) can be (and was) operated to detect and separate the contributions from epithermal (0.02 eV–0.2 MeV) and from fast (>0.2 MeV) neutrons at various sensitivity levels by varying the tensioned metastable negative pressure (Pneg) from 0.3 MPa to 0.7 MPa. The B-CTMFD (configured for detecting the full 0.02 eV to 12 MeV neutrons >0.1 MeV neutron detection at Pneg = 0.7 MPa) offered relative sensitivity enhancements of up to ∼22× greater than Ludlum and over 5× greater than NSN3, over the 0-15 cm range of Pb shielding, and the 0–30 cm concrete shield thickness. The contribution from detecting down scattered neutrons increases with increased thickness, especially for concrete shielding. The B-CTMFD design overcomes the detection penalty (up to 60% depending on shielding type and thickness) inherent in the nonborated centrifugally tensioned metastable fluid detector (NB-CTMFD), designed only to detect fast-energy neutrons—as described in the companion (Part-1) paper. However, unlike the NB-CTMFD, which used 100% decafluorapentane (DFP) (C5H2F10), the B-CTMFD requires the use of an azeotropic mixture of DFP, methanol, and tri-methyl borate (TMB—C3H9BO3, using natural boron) in 80:4:16 proportion. The B-CTMFD was about 6 times more sensitive than NB-CTMFD under the most heavily shielded condition and, taken together, also offered 2-energy bin neutron spectroscopic enablement, together with 22-5× higher absolute efficiency and relative sensitivity compared with the nonspectroscopic Ludlum (He-3) and NSN3 (methane-nitrogen) based detectors. From an intrinsic efficiency standpoint, the B-CTMFD operating at Pneg = 0.7 MPa state demonstrated ∼103× higher intrinsic efficiency over Ludlum 42-49B and NSN3 detectors, respectively.

New concept for neutron dosimetry in ion beam radiotherapy: Feasibility study

In ion beam radiotherapy (IBRT), the secondary neutrons contribute to whole-body exposure, potentially increasing the risk of radiation-induced cancer. The energy range of neutrons generated in the IBRT treatment room spreads from thermal to a few hundred MeV, with those between 0.1 and a few hundred MeV playing a crucial role in the ambient dose equivalent. This study introduces a novel approach for measuring the ambient dose equivalent that focuses on the neutron energy spectrum in the IBRT treatment room. Distinguished by the use of a hydrogen-filled proportional counter and a single moderator, this new method overcomes the weakness of the traditional Bonner Sphere Spectrometer by reducing the complexity and time required for measurements. By deriving the neutron energy spectrum, the neutron ambient dose equivalent can be accurately calculated. Our validation demonstrated that the neutron ambient dose equivalents derived using our method closely agree with the actual values, with discrepancies within 4%, underscoring the feasibility of this innovative approach to neutron dosimetry in the IBRT treatment room.

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.

Development of trench-shaped microstructure thermal neutron detector using α-Al2O3:C based on optically stimulated luminescence

Optical stimulated luminescence technique is increasingly used in neutron personal dosimetry, due to its capability of fast readout and potential for re-estimation of absorbed dose. The design configuration and performance of the microstructure optically stimulated luminescence neutron detector (MOSLND) using α-Al2O3:C was reported. To improve thermal neutron response, the MOSLND is structured with a matrix of microstructure trenches into the α-Al2O3:C substrate and backfilled with 6LiF. To obtain an optimum trench geometric parameter, several Monte Carlo simulations were performed using Geant4 code. The simulation results show that relatively high energy deposition can be realized for small cell dimensions with deep trenches, and is significantly increased compared with that of simple planar designs. Based on the results of calculation and practical limitation, the device was produced and tested using D2O-modrated 252Cf. The optimized minimal detectable dose (MDD) corresponding to signal integration time of 2 s is 7.2 μSv. The experimental results indicate that MOSLND can be a potential solution for personal neutron dosimetry.

Breakthrough electroneutron multi-response miniature dosimetry/spectrometry in medical accelerator

Breakthrough multi-response miniature dosimetry/spectrometry of electroneutrons (EN) was made on surface and in-depths of whole-body polyethylene phantom under 10 cm × 10 cm electron beam of 20 MV Varian Clinac 2100C electron medical accelerator commonly applied for prostate treatment. While dosimetry/spectrometry of photoneutrons (PN) has been well characterized for decades, those of ENs lagged behind due to very low EN reaction cross section and lack of sensitive neutron dosimeters/spectrometers meeting neutron dosimetry requirements. Recently, Sohrabi “miniature neutron dosimeter/spectrometer” and “Stripe polycarbonate dosimeter” have broken this barrier and determined seven EN ambient dose equivalent (ENDE) (µSv.Gy–1) responses from electron beam and from albedo ENs including beam thermal (21 ± 2.63), albedo thermal (43 ± 3.70), total thermal (64 ± 6.33), total epithermal (32 ± 3.90), total fast (112.00), total thermal + epithermal (l96 ± 10), and total thermal + epithermal + fast (208 ± 10.23) ENs. Having seven ENDE responses of this study and seven PNDE responses of previous study with the same accelerator obtained at identical conditions by the same principle author provided the opportunity to compare the two sets of responses. The PNDE (µSv.Gy–1) responses have comparatively higher values and 22.60 times at isocenter which provide for the first time breakthrough ENDE responses not yet reported in any studies before worldwide.

A computational- experimental model of reactor kinetic for investigating the linearity response of in-core neutron detectors of a low power research reactor

In this study, the linear response of fission chamber (FC) neutron detectors in Isfahan miniature neutron source reactor (MNSR) is investigated through experiments and calculations at different powers of the reactor. The linearity of the response of the main neutron detector in the MNSR is important due to its safety functions in reactor power control. Additionally, determining the linear response region of this detector presents a linear neutron dosimetry field even outside of the reactor core, i.e. at the outlet of external beam tubes. Moreover, it is very difficult to remove and calibrate this neutron detector that has been installed in the reactor core for a long time. For this purpose, an alternative method for measuring the linearity of the response of a neutron detector is extended by investigating the kinetic behavior of reactor. The aim of this method is to calculate the delayed neutrons population in the reactor core and compare with the FC detector response. In this method, the numerical solution method is used to solve the point kinetic equation with 15 delayed neutron groups (6 delayed neutron groups and 9 photo-neutron groups) after inserting a large negative reactivity into the MNSR core by considering temperature and xenon reactivity feedbacks during a coupling procedure. To obtain the xenon reactivity feedback, xenon concentration is calculated by coupling the 135Xe and 135I production equations and the εpPFnlPTnl parameter is obtained by an experimental method. Furthermore, the value of fission cross section (Σf) is calculated using a validated model in the Dragon code. The large reactivities are inserted at two initial power of 150 W and 30 kW to investigate the response of the main FC neutron detector in a cold xenon-free condition and by considering temperature and xenon reactivity feedbacks, respectively. The results demonstrate that the response of the MNSR FC detector is linear in the neutron flux range between 5 × 107 to 1 × 1012 n.cm−2.s−1, which is equivalent to reactor power of 1.5 W to 30 kW. This innovative method can be applied to calibrate the neutron detectors in other research and power reactors.

Secondary neutron dosimetry for conformal FLASH proton therapy.

Cyclotron-based proton therapy systems utilize the highest proton energies to achieve an ultra-high dose rate (UHDR) for FLASH radiotherapy. The deep-penetrating range associated with this high energy can be modulated by inserting a uniform plate of proton-stopping material, known as a range shifter, in the beam path at the nozzle to bring the Bragg peak within the target while ensuring high proton transport efficiency for UHDR. Aluminum has been recently proposed as a range shifter material mainly due to its high compactness and its mechanical properties. A possible drawback lies in the fact that aluminum has a larger cross-sectionof producing secondary neutrons compared to conventional plastic range shifters. Accordingly, an increase in secondary neutron contamination was expected during the delivery of range-modulated FLASH proton therapy, potentially heightening neutron-induced carcinogenic risks to thepatient. We conducted neutron dosimetry using simulations and measurements to evaluate excess dose due to neutron exposure during UHDR proton irradiation with aluminum range shifters compared to plastic rangeshifters. Monte Carlo simulations in TOPAS were performed to investigate the secondary neutron production characteristics with aluminum range shifter during 225MeV single-spot proton irradiation. The computational results were validated against measurements with a pair of ionization chambers in an out-of-field region ( 30cm) and with a Proton Recoil Scintillator-Los Alamos rem meter in a far-out-of-field region (0.5-2.5m). The assessments were repeated with solid water slabs as a surrogate for the conventional range shifter material to evaluate the impact of aluminum on neutron yield. The results were compared with the International Electrotechnical Commission (IEC) standards to evaluate the clinical acceptance of the secondary neutronyield. For a range modulation up to 26cm in water, the maximum simulated and measured values of out-of-field secondary neutron dose equivalent per therapeutic dose with aluminum range shifter were found to be and , respectively, overall higher than the solid water cases (simulation: ; measurement: ). The maximum far out-of-field secondary neutron dose equivalent was found to be ( ) and ( ) for the simulations and rem meter measurements, respectively, also higher than the solid water counterparts (simulation: ( ) ; measurement: ( ) ). We conducted simulations and measurements of secondary neutron production under proton irradiation at FLASH energy with range shifters. We found that the secondary neutron yield increased when using aluminum range shifters compared to conventional materials while remaining well below the non-primary radiation limit constrained by the IECregulations.

Characterization of tracks from alpha-particles in PADC detectors for passive environmental monitoring of fast neutron radiation field

Fast neutron dosimetry is particularly crucial in the realm of radioprotection, as it significantly contributes to evaluating the radiation protection measures. This assessment aims to ensure and enhance the effectiveness of implemented safeguards and precautions, thereby minimizing the risks and potential harm associated with radiation exposure. Given the substantial biological damage that fast neutrons can inflict when interacting with living tissues, accurate dosimetry is indispensable for ensuring the safety of personnel in these environments and for optimizing radiation therapy treatments. Employing an indirect detector technique, this research focuses on characterizing the geometrical and optical properties of tracks produced by alpha particles resulting from boron-neutron reactions and distinguish them from the background made up of tracks generated from radon decay and impurities on the surface of the detector. This methodology holds potential utility particularly in situations where dosimeters are not adequately stored: while the idea of employing plastic or alternative materials to shield dosimeters from radon may appear straightforward, there are numerous factors complicating its effectiveness and universal applicability, for example when the radon proof encapsulations are not sealed perfectly. Through the development of a robust protocol for fast neutron dosimetry, we can not only differentiate between tracks produced by alpha particles of varying energy levels but also quantify the dose resulting from exposure to the neutron field. A solid-state nuclear track detector system (Politrack, Mi.am, Italy) was used to address the critical need for measuring exposure to fast neutrons, thermalized by polyethylene spheres. This advancement facilitates the implementation of more effective radioprotection strategies and contributes to the overall safety of radiation therapy procedures.

Evaluation of TL and OSL signals of MgB4O7:Tm,Li prepared by the solution combustion method

Magnesium tetraborate (MgB4O7) doped with rare-earth elements is recognized for its suitability in thermoluminescence (TL) dosimetry due to its low effective atomic number (Zeff = 8.4) and the potential applications in neutron dosimetry. The study investigates the luminescent (TL and OSL) properties of MgB4O7 doped with thulium (Tm3+) and lithium (Li+). Samples were synthesized using the solution combustion method and TL and OSL responses were evaluated using a spectrometer and an automated OSL/TL reader. The emission lines of Tm3+ are consistent with previous literature, with Li+ enhancing TL and OSL responses. TL emission spectra unveil distinct bands corresponding to [Tm3+]* transitions. The electron paramagnetic resonance (EPR) spectroscopy analysis carried out in unirradiated, irradiated and annealed samples.The results show that the signals corresponding to BO32− and O2− are seen upon irradiation and disappear after annealing at 200 °C, indicating the participation of these centers in the TL process. The results show that the TL response of MgB4O7:Tm,Li is much more sensitive to gamma radiation than MgB4O7:Tm, with a minimum detectable value of 0.084 mGy and a linear response with dose for the range studied, from 0 to 500 mGy, with R2 of 0.998.

Converter thickness optimisation using Monte Carlo simulations of Fluorescent Nuclear Track Detectors for neutron dosimetry

Secondary neutrons generated during ion beam radiotherapy present a concern due to the potential dose deposition beyond the treatment volume, thereby elevating the risk of inducing secondary tumours. These neutrons can possess energies comparable to those of the primary ions, reaching magnitudes of several hundred MeV, posing a challenge for neutron detectors. Fluorescent Nuclear Track Detectors (FNTDs) are promising detectors for high-energy neutron dosimetry given their capability to detect particles with a low linear energy transfer. In this work, the sensitivity of FNTDs to neutron energies reaching 20MeV was analysed by experiments and Monte Carlo (MC) simulations, quantifying the recoil proton yield of FNTDs combined with polyethylene (PE) converters of different thicknesses. The FNTDs were read out using a dedicated FNTD reader, demonstrating a reasonable uncertainty by analysing a detector area of 0.1mm2. Investigations of different converter thicknesses reveal optimal detector sensitivity between 0.5mm to 1.0mm for a 241AmBe source, yielding a maximum sensitivity of (22.7±3.4) tracksmSv-1mm-2. Similar converter-FNTD configurations were assessed through MC simulations using FLUKA, yielding a correlation between detector sensitivity and converter thickness. Furthermore, an enhanced detector sensitivity for neutron energies up to 20MeV was found for the PE converter thickness of 4.0mm. The MC simulations can be used to optimise FNTD detector configurations for measuring higher neutron energies by maximising the recoil proton yield.

Development of Dy3+ doped lithium magnesium borate glass system for thermoluminescence based neutron dosimetry applications

Abstract The manuscript reports the synthesis of Dy3+ incorporated lithium magnesium borate glass by melt quenching technique. FTIR study revealed the presence of both BO3 as well as tetrahedral BO4 units through their characteristic frequencies. Photoluminescence (PL) study of unirradiated samples confirmed the presence of Dy dopant in the ‘+3’ oxidation states from the characteristic emissions at 482, 578, 666 and 716 nm corresponding to 4F9/2 → 6H15/2, 4F9/2 → 6H13/2 and 4F9/2 → 6H11/2, 4F9/2 → 6H9/2 transitions, respectively. Thermal neutron and gamma irradiated PL emission and lifetime characteristics were discussed in details based on the different defect centers. Thermal neutron irradiated TL study showed that the material has a broad and single dosimetry glow peak at about 450 K which showed high fading due to low temperature peak. TL based neutron sensitivity of LMB: Dy3+ was found to be about 37.4 times less than that of standard TLD-100 (LiF: Mg, Ti) powder. The net TL response from about 3 to 83 mSv of neutron dose was found to be linear (Adj. R2 = 0.9994) which is one of the most desirable properties for dosimetry applications. In addition, the TL trap parameters were evaluated using both deconvolution of TL glow curve and peak shape method as suggested by Chen which were found to be matching with each other.

Thermoluminescence properties of K2GdF5:Tb material irradiated in neutron-gamma fields of 241Am-Be source

This paper presents the thermoluminescence properties of K2GdF5:Tb material irradiated in a reference neutron-gamma field and its possibility to be used in neutron dosimetry. Double fluoride of K2GdF5 doped with 10 at% Tb3+ ions was synthesized under solid-state reaction conditions and irradiated in the reference neutron-gamma field of 241Am-Be source. The thermoluminescence glow curves were measured by the Rexon UL-320 TLD reader connected with the WinRex-320 processing program for obtaining the integrated thermoluminescence signals (counts). The results show that the temperature peaks of the irradiated K2GdF5:Tb material concerning the gamma and neutron fields at the distance of 100 cm from the source are at 230?C and 300?C, respectively. The sensitivity ratios of fast-to-thermal neutron and gamma-to-thermal neutron are 6.3 % and 18.2 %, respectively. The batch homogeneity and reproducibility are less than 30 % and 7.5 %, respectively. The linearity is less than 10 % in the dose range of less than 24 mSv. The detection threshold is less than 0.1 mSv, and the fading is 10.7 % after 90 days of irradiation. These thermoluminescence properties of K2GdF5:Tb material considerably satisfy the requirements to be used in neutron dosimetry.

A quantified approach to assessing the effectiveness of nuclear reactor pressure vessel neutron fluence monitoring programs

This paper presents a quantitative approach to characterize the effectiveness of neutron fluence monitoring programs using adjoint neutron transport calculations. The results of adjoint calculations are useful for this purpose because the adjoint solution is the expected contribution from one region of phase space to a response of interest, i.e., the importance of that region of phase space to the response. Thus, by identifying a response of interest (for this work, the fast (E > 1.0 MeV) neutron fluence rate at the peak fluence rate location in the reactor pressure vessel (RPV)), one can identify the fuel regions most important to that response. Once identified, the ability of various surveillance programs to adequately monitor the most-important fuel region(s) can be assessed. The approach described is applied to a three-loop Westinghouse pressurized water reactor to compare the effectiveness of surveillance-capsule neutron dosimetry versus ex-vessel neutron dosimetry to monitor the fuel assemblies that contribute to the peak fast (E > 1.0 MeV) neutron fluence rate in the RPV.

Reanalysis of graphite nuclear data limits for AGR dosimetry with updated carbon scattering data and Magnox plant measurements

The Advanced Gas-cooled Reactors (AGR) operating in the UK are, like the Magnox reactors that preceded them, graphite moderated and cooled by pressurised CO2 gas. As such, the nuclear data for carbon are of primary importance in neutron dosimetry and nuclear heating calculations. There is evidence from benchmark data and plant measurements of underprediction of neutron fluxes at the lower end of the fast neutron range for thicknesses of graphite greater than ~50 cm. Furthermore, the range of validation is limited to ~70 cm of graphite, with components beyond this being allocated bounding damage rates. Following the previous ISRD, opportunities were taken to: Look again at the carbon scattering cross-sections by incorporating the ENDF/B-VIII.0 Beta 4 carbon data to assess the impact on the analysis of benchmark data. Reassess validation data from the earlier Magnox reactors to put them on a more appropriate equivalent graphite thickness scale to the AGRs. The investigation into graphite nuclear data libraries concluded that the incorporation of ENDF/B-VIII Beta 4 carbon data into the JEF2.2 and ENDF/B-VII.1 data libraries with BINGO collision processing had only a limited effect on the calculated responses. The UKNDL data, with DICE collision processing, currently used in dosimetry calculations for EDF Energy’s AGRs produces the closest agreement with measurement for all calculated responses for graphite thicknesses greater than 10 cm. For reactor plant validation, applying the sensitivity of the 93mNb(n,n’) response to graphite density resulted in a 15 cm increase to the validation database when compared to a simple “straight-line” distance. However, caution is advised in applying the benchmark equivalent distance derived from locations within a graphite column, to ex-core plant measurements.