Neutron Dose Research Articles

Stray neutron radiation exposures from proton therapy: physics-based analytical models of neutron spectral fluence, kerma and absorbed dose

Objective. Patients who receive proton beam therapy are exposed to unwanted stray neutrons. Stray radiations increase the risk of late effects in normal tissues, such as second cancers and cataracts, and may cause implanted devices such as pacemakers to malfunction. Compared to therapeutic beams, little attention has been paid to modeling stray neutron exposures. In the past decade, substantial progress was made to develop semiempirical models of stray neutron dose equivalent, but models to routinely calculate neutron absorbed dose and kerma are still lacking. The objective of this work was to develop a new physics based analytical model to calculate neutron spectral fluence, kerma, and absorbed dose in a water phantom. Approach. We developed the model using dosimetric data from Monte Carlo simulations and neutron kerma coefficients from the literature. The model explicitly considers the production, divergence, scattering, and attenuation of neutrons. Neutron production was modeled for 120–250 MeV proton beams impinging on a variety of materials. Fluence, kerma and dose calculations were performed in a 30 × 180 × 44 cm3 phantom at points up to 43 cm in depth and 80 cm laterally. Main Results. Predictions of the analytical model agreed reasonably with corresponding values from Monte Carlo simulations, with a mean difference in average energy deposited of 20%, average kerma coefficient of 21%, and absorbed dose to water of 49%. Significance. The analytical model is simple to implement and use, requires less configuration data that previously reported models, and is computationally fast. This model appears potentially suitable for integration in treatment planning system, which would enable risk calculations in prospective and retrospective cases, providing a powerful tool for epidemiological studies and clinical trials.

Validation of a Monte Carlo Framework for Out-of-Field Dose Calculations in Proton Therapy.

Proton therapy enables to deliver highly conformed dose distributions owing to the characteristic Bragg peak and the finite range of protons. However, during proton therapy, secondary neutrons are created, which can travel long distances and deposit dose in out-of-field volumes. This out-of-field absorbed dose needs to be considered for radiation-induced secondary cancers, which are particularly relevant in the case of pediatric treatments. Unfortunately, no method exists in clinics for the computation of the out-of-field dose distributions in proton therapy. To help overcome this limitation, a computational tool has been developed based on the Monte Carlo code TOPAS. The purpose of this work is to evaluate the accuracy of this tool in comparison to experimental data obtained from an anthropomorphic phantom irradiation. An anthropomorphic phantom of a 5-year-old child (ATOM, CIRS) was irradiated for a brain tumor treatment in an IBA Proteus Plus facility using a pencil beam dedicated nozzle. The treatment consisted of three pencil beam scanning fields employing a lucite range shifter. Proton energies ranged from 100 to 165 MeV. A median dose of 50.4 Gy(RBE) with 1.8 Gy(RBE) per fraction was prescribed to the initial planning target volume (PTV), which was located in the cerebellum. Thermoluminescent detectors (TLDs), namely, Li-7-enriched LiF : Mg, Ti (MTS-7) type, were used to detect gamma radiation, which is produced by nuclear reactions, and secondary as well as recoil protons created out-of-field by secondary neutrons. Li-6-enriched LiF : Mg,Cu,P (MCP-6) was combined with Li-7-enriched MCP-7 to measure thermal neutrons. TLDs were calibrated in Co-60 and reported on absorbed dose in water per target dose (μGy/Gy) as well as thermal neutron dose equivalent per target dose (μSv/Gy). Additionally, bubble detectors for personal neutron dosimetry (BD-PND) were used for measuring neutrons (>50 keV), which were calibrated in a Cf-252 neutron beam to report on neutron dose equivalent dose data. The Monte Carlo code TOPAS (version 3.6) was run using a phase-space file containing 1010 histories reaching an average standard statistical uncertainty of less than 0.2% (coverage factor k = 1) on all voxels scoring more than 50% of the maximum dose. The primary beam was modeled following a Fermi–Eyges description of the spot envelope fitted to measurements. For the Monte Carlo simulation, the chemical composition of the tissues represented in ATOM was employed. The dose was tallied as dose-to-water, and data were normalized to the target dose (physical dose) to report on absorbed doses per target dose (mSv/Gy) or neutron dose equivalent per target dose (μSv/Gy), while also an estimate of the total organ dose was provided for a target dose of 50.4 Gy(RBE). Out-of-field doses showed absorbed doses that were 5 to 6 orders of magnitude lower than the target dose. The discrepancy between TLD data and the corresponding scored values in the Monte Carlo calculations involving proton and gamma contributions was on average 18%. The comparison between the neutron equivalent doses between the Monte Carlo simulation and the measured neutron doses was on average 8%. Organ dose calculations revealed the highest dose for the thyroid, which was 120 mSv, while other organ doses ranged from 18 mSv in the lungs to 0.6 mSv in the testes. The proposed computational method for routine calculation of the out-of-the-field dose in proton therapy produces results that are compatible with the experimental data and allow to calculate out-of-field organ doses during proton therapy.

Calculation of the Biological Efficiency of the Proton Component from 14.8 MeV Neutron Irradiation in Computational Biology with Help of Video Cards.

Fast neutron therapy, which previously has demonstrated effective results, but along with a large number of complications, can again be considered a promising treatment method in the treatment of cancer. One of the ways of analyzing the relative biological efficiency and accurate biological dose of fast neutrons in body tissues is to improve the algorithms of computational biology and mathematical modeling. A high-performance computing code was written which allows to estimate in real-time mode the biological dose of the proton component from the action of neutron radiation with an energy of 14.8 MeV. A comparative analysis of the computing performance on various video cards was also performed.

Development of personal neutron dosimeter with energy resolution ability

The prototype neutron personal dosimeter consists of three passive implanted planar silicon detectors. Polyethylene, B4C-containing polyethylene, and 6LiF act as neutron moderators, absorbers, and converters in front of each detector. This personal dosimeter has two modes of operation: real-time mode and unfolding mode. In real-time mode, it shows three neutron personal dose equivalent values, which represent the personal neutron equivalent of high energy (1–15 MeV), intermediate energy (1 × 10−3 MeV –1MeV), and low-energy (2.53 × 10−8 MeV −1 × 10−3 MeV), respectively. Unfolding method based on Genetic Algorithm was used to deduce neutron spectra from pulse height spectra of recoil proton. The dosimeter was calibrated in monoenergetic neutron fields and was tested in an 241Am–Be neutron field.

Effect of Neutron Irradiation on the Critical Event Size of Cleavage Fracture of the Chinese A508-3 Steel

In order to analyze the influence of neutron irradiation on the critical event of cleavage fracture for the Chinese A508-3 (RPV) steel, the fracture morphology, microstructure and grain size of the irradiated specimens are observed and analyzed by optical microscope (OM) and scanning electron microscope (SEM). The results show that with the increase of neutron doses, the fracture mode of Chinese RPV steel material changes gradually from ductile fracture to ductile-brittle mixed fracture transformation. The brittle part for fracture mode of ductile-brittle fracture is cleavage fracture, and the non-metallic inclusions, microstructure and grain size grades are not significantly different from those of the specimens without cleavage fracture. In the same material, with increasing of neutron doses, the critical event size of the cleavage fracture will be reduced, and the percentage of the grain that satisfies the grain size of the cleavage critical event will increase correspondingly, resulting in an increase of the probability of cleavage fracture of the material. The fine grain with uniform distribution can improve partly the fracture toughness of the material and withstand higher neutron doses under the same conditions.

Development of back-illuminated thin silicon diode applied to fast neutron sensor in active personal dosimeter

A back-illuminated thin Si diode was developed for fast neutron sensor in active personal neutron dosimeter for detecting fast neutrons below 1 MeV. Neutron response functions for the fast neutron sensor with the back-illuminated thin Si diode were evaluated based on measurements of monoenergetic neutron beams at the National Metrology Institute of Japan and Monte Carlo simulations using the Particle Heavy Ion Transport Code System in the neutron energy over 500 keV. The lower limit of neutron detection energy was decreased to 500 keV and neutron detection efficiencies were increased several times over other fast neutron sensors. It is possible to measure neutron doses in the neutron energy range of 500 keV–1 MeV using this novel fast neutron sensor, but the neutron doses were undetectable using other fast neutron sensors. The neutron detection efficiencies were closer to the personal dose equivalent conversion curve than those for other fast neutron sensors in this neutron energy range. Neutron doses of individuals employed at a nuclear facility were evaluated using the active personal neutron dosimeter, which consists of both the novel fast and slow neutron sensors.

Lifetime attributable risk of radiation induced second primary cancer from scattering and scanning proton therapy – A model for out-of-field organs of paediatric patients with cranial cancer

Background and PurposeProton therapy (PT) can reduce side effects for paediatric cranial malignancies. Despite the high number of paediatric patients treated with PT, radiation induced risk factors for second primary cancer (SPC) in out-of-field organs are unknown. This study estimated lifetime attributable risk (LAR) of SPC as a function of age and sex for out-of-field organs following passive scattering and scanning beam PT in paediatric brain tumours. Materials and MethodsMeasured neutron dose equivalent spectra for scattered and scanning PT were sourced from literature. The physical distance of 12 measured organs from paediatric CT dataset-based phantoms (5, 9 and 13 years-of-age) were applied to Schneider et al.’s analytical model using MATLAB (R2020B) to calculate the organ-specific LAR of SPC. ResultsScanning beam PT demonstrated smaller LAR (per 10,000 person years) of SPC compared to scattering. This was prominent for more radiosensitive organs, including the lung (320 vs 50), breast (1000 vs 150) and thyroid (350 vs 75), but not for all (i.e., rectum and reproductive organs were <10). For most organs, LAR was highest for 5-year-old females (i.e., breast LAR was 1,000 higher than for 13-year-olds), however, outliers existed for distal organs (i.e., stomach and lung). ConclusionThere was large variation in LAR estimates of out-of-field organs based on measured neutron dose equivalents. Younger female cranial paediatric patients were found at higher risk compared to males, especially for passive scattering PT. Not all organs had improved LAR using scanning beam PT for younger age groups.

Thick target neutron spectral yield and dose measurements with [formula omitted]C (p, n) system at intermediate proton energies between 8–20 MeV

The study of neutron parameters from a thick natural graphite target is of immense importance for obtaining the background neutron energy and yield corrections in case of proton irradiation of low Z targets, where graphite is used predominantly as the supporting target material. There are few measurements with thick graphite target reported in literature at proton energies up to 12 MeV and some at higher energies beyond 20 MeV. The present study focuses on the thick target neutron yield, dose and spectra measurements from thick natC(p,n) system at intermediate proton energies between 8–20 MeV. The experimental measurements were performed at 0°and 90°with respect to the incident beam direction. The generated data will complement the discontinuity in the neutron yield and spectral measurements, in the proton energy range of 12–20 MeV. These dosimetric estimates in the proton energy range of 8–20 MeV will be significant towards the personal radiation protection of occupational workers and neutron shielding evaluations.

Development of a multipurpose rig for material irradiation tests in BR2

A generic rig with a large range of possible irradiation conditions has been successfully developed for the BR2 material test reactor at SCK CEN. The rig contains five capsules in which active temperature control up to ∼300 °C is possible, and irradiation temperature and neutron dose can be chosen independently in each capsule. Each capsule can be lifted out of the core region of BR2 even during the operation of BR2 when the dose for the capsule reached the requested value. This results in material irradiation experimental data for which the irradiation dose is the only single parameter.

A Multi-Scale Simulation Study of Irradiation Swelling of Silicon Carbide.

Silicon carbide (SiC) is a promising structural and cladding material for accident tolerant fuel cladding of nuclear reactor due to its excellent properties. However, when exposed to severe environments (e.g., during neutron irradiation), lattice defects are created in amounts significantly greater than normal concentrations. Then, a series of radiation damage behaviors (e.g., radiation swelling) appear. Accurate understanding of radiation damage of nuclear materials is the key to the design of new fuel cladding materials. Multi-scale computational simulations are often required to understand the physical mechanism of radiation damage. In this work, the effect of neutron irradiation on the volume swelling of cubic-SiC film with 0.3 mm was studied by using the combination of molecular dynamics (MD) and rate theory (RT). It was found that for C-vacancy (CV), C-interstitial (CI), Si-vacancy (SiV), Si-interstitial (SiI), and Si-antisite (SiC), the volume of supercell increases linearly with the increase of concentration of these defects, while the volume of supercell decreases linearly with the increase of defect concentration for C-antisite (CSi). Furthermore, according to the neutron spectrum of a certain reactor, one RT model was constructed to simulate the evolution of point defect under neutron irradiation. Then, the relationship between the volume swelling and the dose of neutrons can be obtained through the results of MD and RT. It was found that swelling typically increases logarithmically with radiation dose and saturates at relatively low doses, and that the critical dose for abrupt transition of volume is consistent with the available experimental data, which indicates that the rate theory model can effectively describe the radiation damage evolution process of SiC. This work not only presents a systematic study on the relationship between various point defect and excess volume, but also gives a good example of multi-scale modelling through coupling the results of binary collision, MD and RT methods, etc., regardless of the multi-scale modelling only focus on the evolution of primary point defects.

The design and optimization of a radiation shield for the Tehran Research Reactor's neutron diffraction facility

In this paper, a novel optimal radiation shield for the Tehran research reactor's neutron powder diffraction facility was designed to reduce undesirable neutron and gamma dose rates. A layer-by-layer methodology was considered based on the design approach to achieve a dose rate of less than 30 μSv/hr. MCNPX was used for this purpose as a Monte Carlo radiation transport code. A combination of neutron and gamma absorbers and moderators was employed to create this shield. The results indicated that the shield's six distinct layers met the design objective, totaling 65 cm in thickness. This newly designed shield comprises 15 cm paraffin, 15 cm Portland iron concrete, another 15 cm paraffin, 5 cm solid boric acid, another 10 cm Portland iron concrete, and 10 cm lead. The neutron and photon dose rates are 9.35 and 12.81 μSv/hr, respectively, which is one-sixteenth and one-third of the current facility shield. Validation of the simulation was accomplished by comparing the results for the current shield to experimental data.

Silicon Carbide Neutron Detectors for Harsh Nuclear Environments: A Review of the State of the Art

Silicon carbide (SiC) semiconductor is an ideal material for solid-state nuclear radiation detectors to be used in high-temperature, high-radiation environments. Such harsh environments are typically encountered in nuclear reactor measurement locations as well as high-level radioactive waste and/or “hot” dismantling–decommissioning operations. In the present fleet of commercial nuclear reactors, temperatures in excess of 300 °C are often encountered, and temperatures up to 800 °C are anticipated in advanced reactor designs. The wide bandgap for SiC (3.27 eV) compared to more widely used semiconductors such as silicon (1.12 eV at room temperature) has allowed low-noise measurements to be carried out at temperatures up to 700 °C. The concentration of thermally induced charge carriers in SiC at 700 °C is about four orders of magnitude less than that of silicon at room temperature. Furthermore, SiC radiation detectors have been demonstrated to be much more resistant to the effects of radiation-induced damage than more conventional semiconductors, such as silicon, germanium, or cadmium zinc telluride (CZT), and have been demonstrated to be operational after extremely high gamma-ray, neutron, and charged-particle doses. The purpose of the present review is to provide an updated state of the art for SiC neutron detectors and to explore their applications in harsh high-temperature, high-radiation nuclear reactor applications. Conclusions related to the current state-of-the-art of SiC neutron detectors will be presented, and specific ideal applications will be discussed.

Joint EURADOS WG9-WG11 rem-counter intercomparison in a Mevion S250i proton therapy facility with Hyperscan pulsed synchrocyclotron

Objective. Proton therapy is gaining popularity because of the improved dose delivery over conventional radiation therapy. The secondary dose to healthy tissues is dominated by secondary neutrons. Commercial rem-counters are valuable instruments for the on-line assessment of neutron ambient dose equivalent (H*(10)). In general, however, a priori knowledge of the type of facility and of the radiation field is required for the proper choice of any survey meter. The novel Mevion S250i Hyperscan synchrocyclotron mounts the accelerator directly on the gantry. It provides a scanned 227 MeV proton beam, delivered in pulses with a pulse width of 10 μs at 750 Hz frequency, which is afterwards degraded in energy by a range shifter modulator system. This environment is particularly challenging for commercial rem-counters; therefore, we tested the reliability of some of the most widespread rem-counters to understand their limits in the Mevion S250i stray neutron field. Approach. This work, promoted by the European Radiation Dosimetry Group (EURADOS), describes a rem-counter intercomparison at the Maastro Proton Therapy centre in the Netherlands, which houses the novel Mevion S250i Hyperscan system. Several rem-counters were employed in the intercomparison (LUPIN, LINUS, WENDI-II, LB6411, NM2B-458, NM2B-495Pb), which included simulation of a patient treatment protocol employing a water tank phantom. The outcomes of the experiment were compared with models and data from the literature. Main results. We found that only the LUPIN allowed for a correct assessment of H*(10) within a 20% uncertainty. All other rem-counters underestimated the reference H*(10) by factors from 2 to more than 10, depending on the detector model and on the neutron dose per pulse. In pulsed fields, the neutron dose per pulse is a fundamental parameter, while the average neutron dose rate is a secondary quantity. An average 150–200 μSv/GyRBE neutron H*(10) at various positions around the phantom and at distances between 186 cm and 300 cm from it was measured per unit therapeutic dose delivered to the target. Significance. Our results are partially in line with results obtained at similar Mevion facilities employing passive energy modulation. Comparisons with facilities employing active energy modulation confirmed that the neutron H*(10) can increase up to more than a factor of 10 when passive energy modulation is employed. The challenging environment of the Mevion stray neutron field requires the use of specific rem-counters sensitive to high-energy neutrons (up to a few hundred MeV) and specifically designed to withstand pulsed neutron fields.

Ultrafast probing of plasma ion temperature in proton–boron fusion by nuclear resonance fluorescence emission spectroscopy

Aneutronic fusion reactions such as proton–boron fusion could efficiently produce clean energy with quite low neutron doses. However, as a consequence, conventional neutron spectral methods for diagnosing plasma ion temperature would no longer work. Therefore, finding a way to probe the ion temperature in aneutronic fusion plasmas is a crucial task. Here, we present a method to realize ultrafast in situ probing of 11B ion temperature for proton–boron fusion by Doppler broadening of the nuclear resonance fluorescence (NRF) emission spectrum. The NRF emission is excited by a collimated, intense γ-ray beam generated from submicrometer wires irradiated by a recently available petawatt (PW) laser pulse, where the γ-ray beam generation is calculated by three-dimensional particle-in-cell simulation. When the laser power is higher than 1 PW, five NRF signatures of a 11B plasma can be clearly identified with high-resolution γ-ray detectors, as shown by our Geant4 simulations. The correlation between the NRF peak width and 11B ion temperature is discussed, and it is found that NRF emission spectroscopy should be sensitive to 11B ion temperatures Ti &amp;gt; 2.4 keV. This probing method can also be extended to other neutron-free-fusion isotopes, such as 6Li and 15N.

Measurement of neutron equivalent dose in the thyroid, chiasma, and lens for patients undergoing pelvic radiotherapy: A phantom study

Radiotherapy is one of the most common methods for treating malignant diseases, whose ultimate goal is to deliver lethal doses to tumor cells. One of the unwanted consequences of radiotherapy is secondary radiation outside the treatment field, which imposes additional doses to healthy tissues and organs, specifically neutron doses, which we aim to evaluate. Therefore, this study aims to measure the fast neutron equivalent dose and the risk of secondary cancer in the thyroid, chiasm, and lenses in the treatment of the pelvic area. In this study, CR-39 detectors were used to measure the equivalent fast neutron dose in different sections of the RANDO Phantom (thyroid, chiasm, and lenses) and were irradiated by the energy of 18 MV on Varian Clinac 2100 C-D linear accelerator. CR-39 detectors were calibrated with predetermined doses by an Am–Be neutron source. Then, after etching and reading processes, the equivalent dose of fast neutrons was determined. According to the results, the fast neutron doses in the thyroid, right and left eye lenses, and chiasm were 0.613 ± 0.024, 0.835 ± 0.040, 0.866 ± 0.016, and 0.685 ± 0.045 mSv/Gy, respectively. Moreover, the secondary cancer risks in the unshielded organs are 0.004, 0.029, 0.030, and 0.025 for the thyroid, right and left eye lenses, and chiasm, respectively. In conclusion, the contribution of neutrons to the secondary doses in the out-of-field organs is significant and should not be ignored.

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

Abstract In this paper, shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW (Holtec International Transfer Cask Variable Weight) loaded with spent fuel assemblies (SFA) from nuclear power plant (NPP) Krsko spent fuel dry storage (SFDS) campaign one. The HI-TRAC VW is a multilayered cylindrical vessel designed to accept a multipurpose canister (MPC) during loading, unloading, and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using origen-s module of the scale code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron-induced gammas (n-γ reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, forward weighted consistent adjoint driven importance sampling (FW-CADIS) variance reduction scheme, as implemented in advantg code, was adopted for accelerating final mcnp6.2 calculations. The dose rates around HI-TRAC VW cask were calculated using mcnp6.2 code for all 16 cask loadings belonging to campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization - Holtec International (HI).

The response of various neutron dose meters considering the application at a high energy particle accelerator

Abstract The applicability of several neutron detectors for dose measurements at a neutron field typical for high energy particle accelerators is investigated. The response of four commercially available active neutron dose meters and two passive detectors to neutrons from a 241Am-Be(α,n) source and to neutrons at the CERN EU high energy reference field was determined experimentally and simulated using the Monte-Carlo code FLUKA. Fluence response functions and dose responses for the different detectors were calculated in the energy range between 1 keV and 10 GeV. The results show that the dose response to the high energy neutron field at CERN of the conventional rem-counters is lower by a factor of 2 to 2.5 if compared to the dose response to a 241Am-Be(α,n) neutron source. The rem-counters exhibiting an additional layer of lead inside the moderating structure showed dose readings which differ only up to 25 %. A thermoluminescent based neutron detector was tested for comparison. These passive detectors revealed a neutron response similar to a rem-counter and may be preferable for situations with highly pulsed beams.

Assessment of the occupational dose rate during fuel handling and storage of various fuel cycles used in the super-critical water reactor and the impact on deep geological repository performance

The Canadian concept of the Pressure-Tube Super-Critical Water-Reactor (PT-SCWR) uses Pu/Th in a once-through fuel cycle. The implication of using Pu/Th in a PT-SCWR on the safety of the workers in the Deep Geological Repository (DGR) facilities as well as on the public after the closure of the DGR was investigated.A comparison between the Pu/Th and other fuel cycles, namely, U/Th (two U/Th fuels with different enrichments and different UO2/ThO2 ratios), and the regular UO2 was also made. The enrichment of U-235 in U/Th and UO2 was adjusted to give the same exit burnup of the proposed Pu/Th fuel, namely 60 GWd per ton of heavy metals (THM).The dose rate as a function of distance from the surface of a canister holding different types of spent fuel, and the thickness of the shielding required to reduce this dose rate to 10 μSv/h were estimated.The dose rate after 50 years of decay for Pu/Th fuel channel is higher than that of uranium by about 8%. This will increase the dose to the workers in DGR facilities unless more shielding is used. Pu/Th will need 13% more thickness of concrete compared to UO2. Using U/Th fuel will also increase the dose received by workers, compared with uranium-only fuel, but less than using Pu/Th. The neutron dose rate of the Pu/Th fuel is the highest dose rate from all fuels studied.Pu/Th spent fuel will cause the highest dose to the public in case of an accident of the DGR after closure as it will produce the largest amount of I-129 per THM. Although the exact dose received by the critical group is not expected to exceed the dose limit for DGR, it is higher than alternative fuels.

Mortality from various diseases of the circulatory system in the Russian Mayak nuclear worker cohort: 1948–2018

This paper reports on the findings from the study of mortality from diseases of the circulatory system (DCS) in Russian nuclear workers of the Mayak Production Association (22 377 individuals, 25.4% female) who were hired at the facility between 1948 and 1982 and followed up until the end of 2018. Using the AMFIT module of the EPICURE software, relative risks (RRs) and excess RRs per unit absorbed dose (ERR/Gy) for the entire Mayak cohort, the subcohort of workers who were residents of the dormitory town of Ozyorsk and the subcohort of migrants from Ozyorsk were calculated based on maximum likelihood. The mean cumulative liver absorbed gamma-ray dose from external exposure was 0.45 (0.65) Gy (mean (standard deviation)) for men and 0.37 (0.56) Gy for women. The mean cumulative liver absorbed alpha dose from internal exposure to incorporated plutonium was 0.18 (0.65) Gy for men and 0.40 (1.92) Gy for women. By the end of the follow-up, 6019 deaths with DCS as the main cause of death were registered among Mayak Production Association workers (including 3828 deaths in the subcohort of residents and 2191 deaths in the subcohort of migrants) over 890 132 (622 199/267 933) person-years of follow-up. The linear model that took into account non-radiation factors (sex, attained age, calendar period, smoking status and alcohol drinking status) and alpha radiation dose (via adjusting) did not demonstrate significant associations of mortality from DCS, ischaemic heart disease (IHD) and cerebrovascular disease with gamma-ray exposure dose in the entire cohort, the resident subcohort or the migrant subcohort (either in men or women). For the subcohort of residents, a significant association with gamma dose was observed for mortality from ischaemic stroke in men with ERR/Gy = 0.43 (95% CI 0.08; 0.99); there were no significant associations with liver absorbed gamma dose for any other considered outcomes. As for internal exposure, for men no significant associations of mortality from any DCS with liver absorbed alpha dose were observed, but for women positive associations were found for mortality from DCS (the entire cohort and the resident subcohort) and IHD (the entire cohort). No significant associations of mortality from various types of DCS with neutron dose were observed either in men or women, although neutron absorbed doses were recorded in only 18% of the workers.

Propagation of radiation source uncertainties in spent fuel cask shielding calculations

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.