River ecological status is shaped by agricultural land use intensity across Europe.

Agriculture impacts the ecological status of freshwaters through multiple pressures such as diffuse pollution, water abstraction, and hydromorphological alteration, strongly impairing riverine biodiversity. The agricultural effects, however, likely differ between agricultural types and practices. In Europe, agricultural types show distinct spatial patterns related to intensity, biophysical conditions, and socioeconomic history, which have been operationalised by various landscape typologies. Our study aimed at analysing whether incorporating agricultural intensity enhances the correlation between agricultural land use and the ecological status. For this, we aggregated the continent's agricultural activities into 20 Areas of Farming-induced Freshwater Pressures (AFFP), specifying individual pressure profiles regarding nutrient enrichment, pesticides, water abstraction, and agricultural land use in the riparian zone to establish an agricultural intensity index and related this intensity index to the river ecological status. Using the agricultural intensity index, nearly doubled the correlative strength between agriculture and the ecological status of rivers as compared to the share of agriculture in the sub-catchment (based on the analysis of more than 50,000 sub-catchment units). Strongest agricultural pressures were found for high intensity cropland in the Mediterranean and Temperate regions, while extensive grassland, fallow farmland and livestock farming in the Northern and Highland regions, as well as low intensity mosaic farming, featured lowest pressures. The results provide advice for pan-European management of freshwater ecosystems and highlight the urgent need for more sustainable agriculture. Consequently, they can also be used as a basis for European Union-wide and global policies to halt biodiversity decline, such as the post-2027 renewal of the Common Agricultural Policy.

The colonization of an irradiated environment: the case of microbial biofilm in a nuclear reactor.

The investigation of the microbial community change in the biofilm, growing on the walls of a containment tank of TRIGA nuclear reactor revealed a thriving community in an oligotrophic and heavy-metal-laden environment, periodically exposed to high pulses of ionizing radiation (IR). We observed a vertical IR resistance/tolerance stratification of microbial genera, with higher resistance and less diversity closer to the reactor core. One of the isolated Bacillus strains survived 15 kGy of combined gamma and proton radiation, which was surprising. It appears that there is a succession of genera that colonizes or re-colonizes new or IR-sterilized surfaces, led by Bacilli and/or Actinobacteria, upon which a photoautotrophic and diazotrophic community is established within a fortnight. The temporal progression of the biofilm community was evaluated also as a proxy for microbial response to radiological contamination events. This indicated there is a need for better dose-response models that could describe microbial response to contamination events. Overall, TRIGA nuclear reactor offers a unique insight into IR microbiology and provides useful means to study relevant microbial dose-thresholds during and after radiological contamination.

Nuclear Cogeneration of Methanol and Acetaldehyde from Ethylene Glycol Using Ionizing Radiation.

Despite offering low-carbon and reliable energy, the utilization of nuclear energy is declining globally due to high upfront capital costs and longer returns on investments. Nuclear cogeneration of valuable chemicals from waste biomass-derived feedstocks could have beneficial impacts while harnessing the underutilized resource of ionizing energy. Here, we demonstrate selective methanol or acetaldehyde production from ethylene glycol, a feedstock derived from glycerol, a byproduct of biodiesel, using irradiations from a nuclear fission reactor. The influence of radiation quality, dose rate, and the absorbed dose of irradiations on radiochemical yields (G-value) has been studied. Under low-dose-rate, γ-only radiolysis during reactor shutdown rate (<0.018 kGy min-1), acetaldehyde is produced at a maximum G-value of 8.28 ± 1.05 µmol J-1 and a mass productivity of 0.73 ± 0.06% from the 20 kGy irradiation of neat ethylene glycol. When exposed to a high-dose-rate (6.5 kGy min-1), 100 kGy mixed-field of neutron + γ-ray radiations, the radiolytic selectivity is adjusted from acetaldehyde to generate methanol at a G-value of 2.91 ± 0.78 µmol J-1 and a mass productivity of 0.93 ± 0.23%. Notably, utilizing 422 theoretical systems could contribute to 4.96% of worldwide acetaldehyde production using a spent fuel pool γ-ray scheme. This research reports G-values and production capacities for acetaldehyde for high-dose scenarios and shows the potential selectivity of a nuclear cogeneration process to synthesize chemicals based on their irradiation conditions from the same reagent.

Radiation Response of Large-Area 4H-SiC Schottky Barrier Diodes.

We report on the effects of large-area 4H-SiC Schottky barrier diodes on the radiation response to ionizing particles. Two different diode areas were compared: 1 mm × 1 mm and 5 mm × 5 mm. 6LiF and 10B4C films, which were placed on top of the diodes, were used as thermal neutron converters. We achieved a thermal neutron efficiency of 5.02% with a 6LiF thermal neutron converter, which is one of the highest efficiencies reported to date. In addition, a temperature-dependent radiation response to alpha particles was presented. Neutron irradiations were performed in a JSI TRIGA dry chamber and an Am-241 wide-area alpha source was used for testing the alpha response of the 4H-SiC Schottky barrier diodes.

The dose accumulation and the impact of deformable image registration on dose reporting parameters in a moving patient undergoing proton radiotherapy.

INTRODUCTION: Potential changes in patient anatomy during proton radiotherapy may lead to a deviation of the delivered dose. A dose estimate can be computed through a deformable image registration (DIR) driven dose accumulation. The present study evaluates the accumulated dose uncertainties in a patient subject to an inadvertent breathing associated motion. MATERIALS AND METHODS: A virtual lung tumour was inserted into a pair of single participant landmark annotated computed tomography images depicting opposite breathing phases, with the deep inspiration breath-hold the planning reference and the exhale the off-reference geometry. A novel Monte Carlo N-Particle, Version 6 (MCNP6) dose engine was developed, validated and used in treatment plan optimization. Three DIR methods were compared and used to transfer the exhale simulated dose to the reference geometry. Dose conformity and homogeneity measures from International Committee on Radioactivity Units and Measurements (ICRU) reports 78 and 83 were evaluated on simulated dose distributions registered with different DIR algorithms. RESULTS: The MCNP6 dose engine handled patient-like geometries in reasonable dose calculation times. All registration methods were able to align image associated landmarks to distances, comparable to voxel sizes. A moderate deterioration of ICRU measures was encountered in comparing doses in on and off-reference anatomy. There were statistically significant DIR driven differences in ICRU measures, particularly a 10% difference in the relative D98% for planning tumour volume and in the 3 mm/3% gamma passing rate. CONCLUSIONS: T he dose accumulation over two anatomies resulted in a DIR driven uncertainty, important in reporting the associated ICRU measures for quality assurance.

4H-SiC Schottky Barrier Diodes for Efficient Thermal Neutron Detection.

In this work, we present the improved efficiency of 4H-SiC Schottky barrier diodes-based detectors equipped with the thermal neutron converters. This is achieved by optimizing the thermal neutron converter thicknesses. Simulations of the optimal thickness of thermal neutron converters have been performed using two Monte Carlo codes (Monte Carlo N-Particle Transport Code and Stopping and Range of Ions in Matter). We have used 6LiF and 10B4C for the thermal neutron converter material. We have achieved the thermal neutron efficiency of 4.67% and 2.24% with 6LiF and 10B4C thermal neutron converters, respectively.

Use of Gaussian process regression for radiation mapping of a nuclear reactor with a mobile robot.

Collection and interpolation of radiation observations is of vital importance to support routine operations in the nuclear sector globally, as well as for completing surveys during crisis response. To reduce exposure to ionizing radiation that human workers can be subjected to during such surveys, there is a strong desire to utilise robotic systems. Previous approaches to interpolate measurements taken from nuclear facilities to reconstruct radiological maps of an environment cannot be applied accurately to data collected from a robotic survey as they are unable to cope well with irregularly spaced, noisy, low count data. In this work, a novel approach to interpolating radiation measurements collected from a robot is proposed that overcomes the problems associated with sparse and noisy measurements. The proposed method integrates an appropriate kernel, benchmarked against the radiation transport code MCNP6, into the Gaussian Process Regression technique. The suitability of the proposed technique is demonstrated through its application to data collected from a bespoke robotic system used to conduct a survey of the Joz̆ef Stefan Institute TRIGA Mark II nuclear reactor during steady state operation, where it is shown to successfully reconstruct gamma dosimetry estimates in the reactor hall and aid in identifying sources of ionizing radiation.

Nuclear-driven production of renewable fuel additives from waste organics.

Non-intermittent, low-carbon energy from nuclear or biofuels is integral to many strategies to achieve Carbon Budget Reduction targets. However, nuclear plants have high, upfront costs and biodiesel manufacture produces waste glycerol with few secondary uses. Combining these technologies, to precipitate valuable feedstocks from waste glycerol using ionizing radiation, could diversify nuclear energy use whilst valorizing biodiesel waste. Here, we demonstrate solketal (2,2-dimethyl-1,3-dioxolane-4-yl) and acetol (1-hydroxypropan-2-one) production is enhanced in selected aqueous glycerol-acetone mixtures with γ radiation with yields of 1.5 ± 0.2 µmol J-1 and 1.8 ± 0.2 µmol J-1, respectively. This is consistent with the generation of either the stabilized, protonated glycerol cation (CH2OH-CHOH-CH2OH2+ ) from the direct action of glycerol, or the hydronium species, H3O+, via water radiolysis, and their role in the subsequent acid-catalyzed mechanisms for acetol and solketal production. Scaled to a hypothetically compatible range of nuclear facilities in Europe (i.e., contemporary Pressurised Water Reactor designs or spent nuclear fuel stores), we estimate annual solketal production at approximately (1.0 ± 0.1) × 104 t year-1. Given a forecast increase of 5% to 20% v/v% in the renewable proportion of commercial petroleum blends by 2030, nuclear-driven, biomass-derived solketal could contribute towards net-zero emissions targets, combining low-carbon co-generation and co-production.

Sterilization of polypropylene membranes of facepiece respirators by ionizing radiation.

Ionizing radiation has been identified as an option for sterilization of disposable filtering facepiece respirators in situations where the production of the respirators cannot keep up with demand. Gamma radiation and high energy electrons penetrate deeply into the material and can be used to sterilize large batches of masks within a short time period. In relation to reports that sterilization by ionizing radiation reduces filtration efficiency of polypropylene membrane filters on account of static charge loss, we have demonstrated that both gamma and electron beam irradiation can be used for sterilization, provided that the respirators are recharged afterwards.

Dose rate calculations at beam tube no. 5 of the JSI TRIGA mark II research reactor using Monte Carlo method.

Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate neutron and prompt gamma particles emitted from TRIGA research reactor during operation. Firstly, the method was validated by measuring dose rates around open beam port number 5 was unplugged. Neutron and gamma dose rates inside the reactor hall in the vicinity of the beam port were calculated and compared to the measurements. Due to the satisfactory agreement, the method was later used to design external shielding for the same beam port when it was upgraded - special mechanism was installed that allows irradiation of larger samples. Computational analysis of the proposed shielding configuration provided acceptable dose rate levels inside the reactor hall. When the shield was constructed, calculated dose rates were confirmed by the actual measurements. No modifications were needed.

Validation of the Serpent 2 code on TRIGA Mark II benchmark experiments.

The main aim of this paper is the development and validation of a 3D computational model of TRIGA research reactor using Serpent 2 code. The calculated parameters were compared to the experimental results and to calculations performed with the MCNP code. The results show that the calculated normalized reaction rates and flux distribution within the core are in good agreement with MCNP and experiment, while in the reflector the flux distribution differ up to 3% from the measurements.

Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements.

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system.

Capture cross section measurement analysis in the Californium-252 spectrum with the Monte Carlo method.

Absolute average capture cross sections of gold, thorium, tantalum, molybdenum, copper and strontium in (252)Cf spontaneous fission neutron spectrum were simulated for two types of experiment setups preformed by Z. Dezso and J. Csikai and by L. Green. The experiments were simulated with MCNP5 using cross section data from the ENDF/B-VII.0 library. The determination of neutron backscattering was calculated with the use of neutron flagging. Correction factors to experimentally measured values were determined to obtain average cross sections in a pure (252)Cf spontaneous fission spectrum. Influence of concrete wall thickness, air moisture and room size on the average cross section was analyzed. Correction factors amounted to about 30%. Corrected values corresponding to average cross sections in a pure (252)Cf spectrum were calculated for (197)Au, (232)Th, (181)Ta, (98)Mo, (65)Cu and (84)Sr. Average cross sections were also calculated with the RR_UNC software using IRDFF-v.1.05 and ENDF/B-VII.0 libraries. The revised average radiative capture cross sections are 75.5±0.1 mb for (197)Au, 87.0±1.6 mb for (232)Th , 98.0±4.5 mb for (181)Ta, 21.2±0.5 mb for (98)Mo, 10.3±0.3 mb for (63)Cu, and 34.9±6.5 mb for (84)Sr.

Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring.

Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor.

Computational analysis of irradiation facilities at the JSI TRIGA reactor.

Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities.