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Neutron spectrum unfolding is a crucial process in radiation protection and dosimetry. Unfolding codes using iterative algorithms require a criterion to stop the iterations. One approach often relies on the Root Mean Square Error (RMSE) criterion to assess the convergence of iterative algorithms. The aim of this work is to present a new criteria: Average Ratio Scaled (AVGS) and Relative Change in AVGS (dAVGS) to address specific challenges associated with RMSE. Extensive validation tests were conducted, covering a range of scenarios with results showing high level of agreement between the unfolded spectra and the reference.

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Bruce Power operates a first-of-its-kind isotope production system (IPS) that enables continuous production of 177Lu within Canada Deuterium Uranium (CANDU) commercial power reactors. Located on the reactivity mechanisms deck of Unit 7, just outside of reactor containment but in close proximity to the primary heat transport (PHT) pumps, this facility offers unique advantages for 177Lu production. However, employees working in this area encounter a radiation hazard which consists primarily of photoneutrons. These originate from the base of the PHT pumps and are only present when the reactor is operating. This study evaluates neutron exposure at Bruce Power's IPS by using a nested neutron spectrometer (NNS) to determine the neutron energy spectra and absolute dosimetric quantities such as the ambient dose equivalent, H*(10). The results from the NNS are then compared to surveys performed by a portable neutron rem meter (Model NP-2 by Nuclear Research Corporation), routinely used by Bruce Power staff for workplace monitoring. While the Model NP-2 generally showed consistent results across locations, a 50% dose correction factor was identified when operators were harvesting 177Lu from the IPS. This finding highlights an opportunity to reduce the neutron dose that is assigned to operators when producing 177Lu.

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Bonner Sphere Spectrometers (BSS) are widely used for neutron spectrometry. Spectra are obtained by unfolding detector readings. In this work, a Python Graphical User Interface Application (GUI/App) for spectrum unfolding is presented; SpecUnPy. In this App, the user can choose three unfolding algorithms: SPUNIT/MLEM/GRAVEL. There is no limit for energy bins or detectors and after unfolding, a ".xlsx" file and a graphical comparison can be downloaded. This paper presents SpecunPy and some tests performed to validate it.

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The response functions (RFs) of a Bonner Sphere Spectrometer (BSS) with a 6LiI thermal neutron detector were calculated using the Monte Carlo codes PHITS (version 3.26) and MCNPX (version 2.7.0), with their own default nuclear data libraries, and physics models. RFs were compared with other published data, obtained for the same spectrometer using the MCNP6.1 code with its own physics models. A discussion on the influence of using different nuclear data libraries and physics models using these codes/versions is analyzed.

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Boron neutron capture therapy (BNCT) is a promising cancer treatment that uses energetic ions released from 10B(n, α)7Li reactions. Accurate assessment of neutron energy spectra is important for simulation-based evaluation of neutron doses during BNCT. In this study, a proof-of-concept study was conducted for a neutron spectrometry technique that involves the use of a water phantom, which is commonly used for quality assurance in BNCT, as a moderator. The technique involves applying unfolding to the count rate distribution of the thermal neutron counter measured within the phantom to derive the energy spectrum. We performed experiments using a spherical 3He proportional counter in neutron fields generated by 252Cf and 241Am-Be sources. The results demonstrated that the spectrometer reasonably reproduced neutron spectra and showed the potential of using a water phantom as a moderator for such a technique.

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During production of radiopharmaceuticals, the radiation situation in cyclotron pit is an important parameter, which is being monitored to ensure fulfilment of the limits and conditions of safe operation. The neutron flux in the structural components of the accelerator is also an important parameter, because the secondary neutrons are responsible for activation of cyclotron structural components and may even affect structural changes in it. This paper aims to characterize the neutron field in inner positions of medical accelerator IBA 18/9 by activation detectors and by means of scintillation spectrometry. The backward angle measurement was realized also in special liquid water target (H218O) at U120M cyclotron to confirm the data obtained in IBA 18/9 cyclotron.

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For neutron spectrometry of the D2O-moderated 252Cf source with a Bonner sphere spectrometer (BSS), it is difficult to use the large and heavy shadow cone to correct the neutron scattering effect. To overcome this problem, Monte Carlo (MC) simulation method was applied to calculate the neutron scattering ratio and to establish the BSS response functions. The simulated response functions were verified by experimental measurements in reference mono-energetic neutron fields. MC simulation based scattering-correction was validated by measurement of 252Cf neutron field. The measured and simulated values of the neutron scattering ratio were very close with relative errors within ±6%. Finally, the neutron spectrum and the spectrum averaged conversion coefficients of the D2O-moderated 252Cf were measured using BSS after scattering-correction by MC simulation, and the results agreed with the values recommended by ISO 8529-1:2021. It shows that the MC simulation can be a useful substitute to shadow cones method for neutron scattering-correction.

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The aim of this work is to present a reproducible methodology for the evaluation of total equivalent doses in organs during proton therapy facilities. The methodology is based on measuring the dose equivalent in representative locations inside an anthropomorphic phantom where photon and neutron dosimeters were inserted. The Monte Carlo simulation was needed for obtaining neutron energy distribution inside the phantom. The methodology was implemented for a head irradiation case in the passive proton beam of iThemba Labs (South Africa). Thermoluminescent dosimeter (TLD)-600 and TLD-700 pairs were used as dosimeters inside the phantom and GEANT code for simulations. In addition, Bonner sphere spectrometry was performed inside the treatment room to obtain the neutron spectra, some relevant neutron dosimetric quantities per treatment Gy, and a percentual distribution of neutron fluence and ambient dose equivalent in four energy groups, at two locations. The neutron spectrum at one of those locations was also simulated so that a reasonable agreement between simulation and measurement allowed a validation of the simulation. Results showed that the total out-of-field dose equivalent inside the phantom ranged from 1.4 to 0.28 mSv/Gy, mainly due to the neutron contribution and with a small contribution from photons, 10% on average. The order of magnitude of the equivalent dose in organs was similar, displaying a slow reduction in values as the organ is farther from the target volume. These values were in agreement with those found by other authors in other passive beam facilities under similar irradiation and measurement conditions.

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A multi-shell neutron spectrometer with indium foil detector (In-MuNS) was developed to evaluate intense neutron fields that are generated in medical accelerators. The response matrix of this new spectrometer was calculated from 1 meV to 100 MeV using MCNP5 v.1.6 with ENDF/B-VIII.0 nuclear data. An experiment with a252Cf source with known emission rate was performed to validate the computational model of the spectrometer. This included detailed modelling of the irradiation room to evaluate the room-scattered field. The contribution of scattered neutrons to the induced activity in the foil reached 30% for the smallest sphere configuration (diameter 5.0 cm). The quotient between the experimental and simulated foil activity remained satisfactorily constant (1.03±0.04) as the sphere diameter varied, demonstrating the validity of the simulation model. In-MuNS proved to be a portable and compact alternative to conventional Bonner spheres.

RESUMEN

PURPOSE: To simulate secondary neutron radiation fields that had been measured at different relative positions during phantom irradiation inside a scanning proton therapy gantry treatment room. Further, to identify origin, energy distribution, and angular emission of the secondary neutrons as a function of proton beam energy. METHODS: The FLUKA Monte Carlo code was used to model the relevant parts of the treatment room in a scanned pencil beam proton therapy gantry including shielding walls, floor, major metallic gantry-components, patient table, and a homogeneous PMMA target. The proton beams were modeled based on experimental beam ranges in water and spot shapes in air. Neutron energy spectra were simulated at 0°, 45°, 90° and 135° relative to the beam axis at 2m distance from isocenter for monoenergetic 11×11cm2 fields from 200MeV, 140MeV, 75MeV initial proton beams, as well as for 118MeV protons with a 5cm thick PMMA range shifter. The total neutron spectra were scored for these four positions and proton energies. FLUKA neutron spectra simulations were crosschecked with Geant4 simulations using initial proton beam properties from FLUKA-generated phase spaces. Additionally, the room-components generating secondary neutrons in the room and their contributions to the total spectrum were identified and quantified. RESULTS: FLUKA and Geant4 simulated neutron spectra showed good general agreement with published measurements in the whole simulated neutron energy range of 10-10 to 103MeV. As in previous studies, high-energy (E≥19.6MeV) neutrons from the phantom are most prevalent along 0°, while thermalized (1meV≤E<0.4eV) and fast (100keV≤E<19.4MeV) neutrons dominate the spectra in the lateral and backscatter direction. The iron of the large bending magnet and its counterweight mounted on the gantry were identified as the most determinant sources of secondary fast-neutrons, which have been lacking in simplified room simulations. CONCLUSIONS: The results helped disentangle the origin of secondary neutrons and their dominant contributions and were strengthened by the fact that a cross comparison was made using two independent Monte Carlo codes. The complexity of such room model can in future be limited using the result. They may further be generalized in that they can be used for an assessment of neutron fields, possibly even at facilities where detailed neutron measurements and simulations cannot be performed. They may also help to design future proton therapy facilities and to reduce unwanted radiation doses from secondary neutrons to patients.

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The correct description of neutron transport in lead is an essential task for correct description of tritium production in the DEMO (DEMOnstration Power Station) breeding blanket because some concepts deal with lead as a major component: namely the WCLL (water cooled lithium lead blanket), HCLL (helium cooled lithium lead blanket), and DCLL (dual cooled lithium lead blanket). Concerning the improvement of the knowledge about the transport of fast neutrons in lead, a set of experiments and calculations was carried out to study this problem with a well-defined neutron beam. The neutron flux behind various lead arrangements positioned along the beam axis was measured using a stilbene scintillation crystal (10 mm × 10 mm) with neutron and gamma pulse shape discrimination. The measurement was performed along the beam axis and in the case of the thick target also above the axis, to estimate the neutron angular scatter in lead. The calculations were realized using MCNP6 with various nuclear data libraries. Discrepancies in the angular distribution description in the energy region of about 1 MeV were discovered by these experiments.

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The neutron flux distribution behind a reactor pressure vessel (RPV) is an important parameter that is monitored to determine neutron fluence in the RPV. Together with mechanical testing of surveillance specimens, these are the most important parts of in-service inspection programs that are essential for a realistic and reliable assessment of the RPV residual lifetime. The fast neutron fluence values are determined by a calculation. These calculation results are accompanied by measurements of induced activities of the activation foils placed in the capsules behind the RPV at selected locations, namely in azimuthal profile. In case of discrepancies between the measured and calculated activities of the activation foils placed behind the pressure vessel, it is difficult to determine the source of the deviation. During such analysis, there arises a question on the influence of power peaking near core boundary on neutron profile behind the RPV. This paper compares the calculated and measured increase of the neutron flux density distribution behind the reactor pressure vessel in the azimuthal profile that has arisen from the replacement of 164 fuel pins located close to reactor internals by pins with the higher enrichment. This work can be understood as the first step in the characterization of the effect of incorrectly calculated pin power or burn-up in the fuel assembly at the core boundary relative to the neutron flux distribution behind reactor pressure vessel. Based on a good agreement between the calculated and experimental values, it can be concluded that the mathematical model used to evaluate the power increase is correct.

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CSYP (CYlindrical SPectrometer) is a directional neutron spectrometer based on a single moderator embedding multiple thermal neutron detectors. Similarly to Bonner Spheres, CYSP responds from thermal up to GeV neutrons and the spectrum is obtained via few-channel unfolding methods. CYSP has the shape of a polyethylene cylinder with diameter 50 cm and height 65 cm. Owing on a thick collimator and on a specifically designed shielding structure, the internal detectors only respond to neutrons coming from a known direction. Internal thermal neutron detectors are one-cm2 6LiF-covered silicon diodes. Un upgraded version of CYPS was developed to work in low intensity applications, such as cosmic field measurements. It is called CYSP-HS (High-Sensitivity) and is equipped with large area 6LiF-covered silicon diodes (LATND, Large Area Thermal Neutron Detectors). Compared with the former CYSP, the sensitivity increased approximately by an order of magnitude. This paper presents CYSP-HS focusing on the new internal detectors, the response matrix and its verification in a reference field of Am-Be available at the Politecnico di Milano.

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Neutron field measurements around individual DPT type spent fuel transport and storage casks were performed at the storage installation of Trillo Nuclear Power Plant. Neutron spectra were determined at five different positions using a Bonner sphere system. These measurements were used to validate the dosimetry measurements of two neutron monitors. Axial and angular measurements were used to determine the ambient dose equivalent. The results are consistent with the cask design and no field correction is needed when using these monitors.

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In boron neutron capture therapy, it is important to evaluate the dose administered to a patient's body outside the tumour area. The exposure dose is evaluated by calculation; however, the calculated value must be validated using a measured value. The dose evaluations based on the measured neutron spectrum are investigated. Multi-foil activation, combined with a LiCaAlF6 scintillation detector and an imaging plate, is proposed as a measurement method. The proposed method can measure the neutron spectrum at various points quickly.

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Neutron fields inside and outside the independent spent fuel storage installation of Trillo Nuclear Power Plant are characterized exhaustively in terms of neutron spectra and ambient dose equivalent, measured by Bonner sphere system and LB6411 monitor. Measurements are consistent with storage casks and building shield characteristics, and also with casks distribution inside the building. Outer values at least five times lower than dose limit for free access area are found. Measurements with LB6411 and spectrometer are consistent with each other.

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A well-defined neutron spectrum is an essential tool for calibration and tests of spectrometry and dosimetry detectors, and evaluation methods for spectra processing. Many of the nowadays used neutron standards are calibrated against a fission spectrum which has a rather smooth energy dependence. In recent time, at the LVR-15 research reactor in Rez, an alternative approach was tested for the needs of fast neutron spectrometry detector calibration. This process comprises detector tests in a neutron beam, filtered by one meter of single-crystalline silicon, which contains several significant peaks in the fast neutron energy range. Tests in such neutron field can possibly reveal specific problems in the deconvolution matrix of the detection system, which may stay hidden in fields with a smooth structure and can provide a tool for a proper energy calibration. Test with several stilbene scintillator crystals in two different beam configurations supplemented by Monte-Carlo transport calculations have been carried out. The results have shown a high level of agreement between the experimental data and simulation, proving thus the accuracy of used deconvolution matrix. The chosen approach can, thus, provide a well-defined neutron reference field with a peaked structure for further tests of spectra evaluation methods and scintillation detector energy calibration.

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The neutron spectral fluence of an accelerator-based neutron source facility for boron neutron capture therapy (BNCT) based on a proton linac and a beryllium target was evaluated by the unfolding method using a Bonner sphere spectrometer (BSS). A 3He-proportional-counter-based BSS was used with weak beam during the development of the facility. The measured epithermal neutron spectra were consistent with calculations. The epithermal neutron intensity at the beam port was estimated and the results gave a numerical target for the enhancement of the proton beam intensity and will be used as reference data for measurements performed after the completion of the facility.

RESUMEN

BACKGROUND AND PURPOSE: Systematic investigation of the energy and angular dependence of secondary neutron fluence energy distributions and ambient dose equivalents values (H*(10)) inside a pencil beam scanning proton therapy treatment room using a gantry. MATERIALS AND METHODS: Neutron fluence energy distributions were measured with an extended-range Bonner sphere spectrometer featuring ³He proportional counters, at four positions at 0°, 45°, 90°, and 135° with respect to beam direction and at a distance of 2 m from the isocenter. The energy distribution of secondary neutrons was investigated for initial proton beam energies of 75 MeV, 140 MeV, and 200 MeV, respectively, using a 2D scanned irradiation field of 11 × 11 cm² delivered to a 30 × 30 × 30 cm³ PMMA phantom. Additional measurements were performed at a proton energy of 118 MeV including a 5 cm range-shifter (PMMA), yielding a Bragg peak position similar to that of 75 MeV protons. RESULTS: Ambient dose equivalent values from 0.3 µSv/Gy (75 MeV; 90°) to 24 µSv/Gy (200 MeV; 0°) were measured inside the treatment room at a distance of 2 m from the isocenter. H*(10) values were lower (by factors of up to 7.2 (at 45°)) at 75 MeV compared to those at 118 MeV with the 5 cm range-shifter. At 0° and 45°, an evaporation peak was found in the measured neutron fluence energy distributions, at neutron energies around MeV, which contributes about 50% to total H*(10) values, for all investigated proton beam energies. CONCLUSIONS: This study showed a pronounced increase of secondary neutron H*(10) values inside the proton treatment room with increasing proton energy without beam modifiers. For example, in beam direction this increase was about a factor of 50 when protons of 75 MeV and 200 MeV were compared. The existence of a peak of secondary neutrons in the MeV region was demonstrated in beam direction (0°). This peak is due to evaporation neutrons produced in the existing surrounding materials such as those used for the gantry. Therefore, any simulation of the secondary neutrons within a proton treatment room must take these materials into account. In addition, the results obtained here show that the use of a range-shifter increases the production of secondary neutrons inside the treatment room. Using a range-shifter, the higher neutron doses observed mainly result from the higher incident proton energy (118 MeV instead of 75 MeV when no range-shifter was used), due to higher neutron production cross-sections.

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This paper describes the measurement campaign held around the neutron radiography (NR) facility of the Algerian 1MW NUR research reactor. The main objective of this work is to characterize accurately the neutron beam provided at one of the radial channels of the NUR research reactor taking benefit of the acquired CRNA Bonner spheres spectrometer (BSS). The specific objective was to improve the image quality of the NR facility. The spectrometric system in use is based on a central spherical (3)He thermal neutron proportional counter combined with high density polyethylene spheres of different diameters ranging from 3 to 12in. This counting system has good gamma ray discrimination and is able to cover an energy range from thermal to 20MeV. The measurements were performed at the sample distance of 0.6m from the beam port and at a height of 1.2m from the facility floor. During the BSS measurements, the reactor was operating at low power (100W) to avoid large dead times, pulse pileup and high level radiation exposures, in particular, during spheres handling. Thereafter, the neutron spectrum at the sample position was unfolded by means of GRAVEL and MAXED computer codes. The thermal, epithermal and fast neutron fluxes, the total neutron flux, the mean energy and the Cadmium ratio (RCd) were provided. A sensitivity analysis was performed taking into account various defaults spectra and ultimately a different response functions in the unfolding procedure. Overall, from the obtained results it reveals, unexpectedly, that the measured neutron spectrum at the sample position of the neutron radiography of the NUR reactor is being harder with a predominance of fast neutrons (>100keV) by about 60%. Finally, those results were compared to previous and more recent measurements obtained by activation foils detectors. The agreement was fairly good highlighting thereby the consistency of our findings.