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Given the need for criteria to control the radiation doses due to radionuclide inhalation, in 1994 the International Commission on Radiological Protection presented a classification for radioactive compounds based on their pulmonary absorption rates. The Commission classified the compounds into fast, moderate, and slow categories and assigned to each material a default absorption class. Nevertheless, the proposed categories do not always resemble the actual behavior of the classified materials in the pulmonary environment. Therefore, the Commission itself suggested the assessment of the inhalation risk of a particulate substance referring to an in vivo study using the same material. Since it is not possible to trace in literature in vivo studies analyzing the physiological behavior of the totality of inhalable radioactive materials, the Commission suggested adopting in vitro systems simulating the pulmonary mechanism. For this reason, in the last 50 y, many simulating setups have been implemented, but none of these seemed to reproduce the lung dissolution dynamics effectively as the results were not comparable with the ones obtained using in vivo techniques. This paper aims to describe an innovative experimental apparatus implemented as an attempt to add another step toward the realization of a gold standard. In particular, the system was validated with BaSO 4 particulate, and the resulting error with respect to the physiological expected value figured as less than 4%. The system was then employed for the lung dissolution tests of radioactive graphite extracted from Politecnico di Milano's nuclear reactor to assess the radiobiological risk due to a slow lung absorption process that workers might run into during the reactor decommissioning.

This deliverable is the Proceedings book of the PRISMAP workshop 1 “PRISMAP workshop on emerging infrastructures and technical developments”. The Workshop on Emerging Infrastructures and Technical Developments was organised at Legnaro National Laboratories on 21-24 November 2022. The Workshop gave a general update on international facilities producing radionuclides, identified the most promising radionuclides on which production R&D shall focus in the near future, and presented the latest technical developments on targets, ion sources and mass separation techniques. This is one of the two events organised within the WP8: Involvement of Emerging Infrastructures. The workshop was organised in a hybrid format, registering 76 participants, out of which 39 in person and 37 remotely. The first day of the workshop included 15 talks dedicated to infrastructures and radionuclides programmes. The second day was dedicated to more technical topics, including 9 technical talks covering the major activities performed within WP10: on targets, ion sources and isotopes purification. The on-site participants could join a visit to SPES facilities, enjoying a unique opportunity to enter all sections of this ISOL facility which is getting ready to start the beam commissioning. Furthermore, a poster session has been organised, where 9 posters have been presented. Proceedings contributions from some of the posters are included in this Book of Proceedings.

The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects.

Rad je podijeljen na tri dijela. U prvom dijelu opisan je razvoj nuklearne energetike i tendencija budućeg razvoja kao i neki od osnovnih pojmova iz područja nuklearne energetike. U drugom dijelu izvršena je klasifikacija nuklearnih elektrana prema tipu reaktora. Obrađeno je devet tipova reaktora u manjoj ili većoj mjeri. Ovdje je naglasak najviše stavljen na tlakovodni i kipući reaktor kao dva tipa reaktora koji su danas u najvećem broju zastupljeni u nuklearnim elektranama, natrijem hlađeni brzi reaktor također je nešto detaljnije opisan kao najviše razvijen predstavnik reaktora IV generacije. U trećem dijelu postavljeni su matematički modeli za proračun termodinamičkih efikasnosti parnog kružnog ciklusa tlakovodnog i natrijem hlađenog reaktora, naposljetku su izračunate i uspoređene termodinamičke efikasnosti. The work is divided into three parts. The first part describes the development of nuclear energy and the tendency of future development, as well as some of the basic terms in the field of nuclear energy. In the second part, nuclear power plants were classified according to reactor type. Nine reactor types were described to a greater or lesser extent. Here, the emphasis is mostly placed on pressurized water reactor and boiling water reactor as the two types of reactors that are represented in the largest number in nuclear power plants today, the sodium cooled fast reactor is also described in more detail as the most developed representative of generation IV reactors. In the third part, mathematical models for calculating the thermodynamic efficiencies of the steam cycle of pressurized water and sodium cooled reactors were set, and finally, the thermodynamic efficiencies were calculated and compared.

This article analyses the trends and problems associated with the peculiarities of the development of nuclear energy in France. This country has always paid particular attention to the fuel and energy sector. Its significant potential allows it to occupy one of the top places in the world in the use of nuclear technologies in the national economy. At the same time, the relatively hostile policy of the European Union concerning nuclear energy in connection with the implementation of its course on the “green” economy creates many problems in the relations of Paris with the leadership of this integration association. Given that Germany in 2022 completely closes its nuclear power plants (NPP), and France becomes the leading country in the EU for the use of nuclear energy, the leadership of this country under the influence of political factors often has to make mutually exclusive decisions in this area. France’s position as the most significant nuclear power in the EU is complicated because it faces increasing competition in the global nuclear market. The leading positions on it are traditionally occupied by the United States, Russia, and China. Time will tell whether France will be able to cope with these threats. One thing is certain: France seeks to maintain and strengthen its position as the leading nuclear power in the EU and worldwide. France’s long-term experience in designing and building nuclear power plants and ensuring their safe operation can also be useful for Russia, as it is generally recognized that our country has been and is one of the world leaders in nuclear energy.

Refractory high-entropy alloys (RHEAs) have recently attracted widespread attention due to their outstanding mechanical properties at elevated temperatures, making them appealing for concentrating solar power and nuclear energy applications. Here, the corrosion behavior of equimolar HfTaTiVZr and TaTiVWZr RHEAs was investigated in molten FLiNaK eutectic salt (LiF-NaF-KF: 46.5−11.5−42 mol.%) at 650 °C. Potentiodynamic polarization, electrochemical impedance spectroscopy (EIS), and immersion test measurements were carried out for these two RHEAs and compared with Inconel 718 (IN718) superalloy and SS316 stainless steel under identical test conditions. Both TaTiVWZr and HfTaTiVZr refractory high-entropy alloys exhibited an order of magnitude lower corrosion rate than SS316. IN718 and TaTiVWZr showed similar corrosion rates. Corrosion products enriched with noble alloying elements formed in the case of TaTiVWZr and IN718 were stable and protective on the substrate. SS316 showed the lowest corrosion resistance and void formation along the exposed surface due to the active dissolution of Cr and Fe, which provided diffusion paths for the corroded species. The surface analysis results showed that IN718 underwent pitting corrosion, while TaTiVWZr experienced selective dissolution in the inter-dendritic area. In contrast, HfTaTiVZr and SS316 experienced corrosion at the grain boundaries.

Since nuclear wastes are the most important wastes in terms of health and the environment, they are evaluated differently within nuclear reactors as well as in terms of their use in medical and industrial applications. In some cases, emergency intervention is necessary due to the amount of radioactivity or the physical and/or chemical conditions. The purpose of this study is to investigate the adsorption properties of nano Ca3Y2B4O12 (CYBO) material synthesized by the sol-gel combustion method for the adsorption of Thorium (IV) from an aqueous medium. We tested how pH (3–8), the concentration of Th (IV) (25–125 mg/L), amount of adsorbent value (0.005–0.08 g) and temperature (20–60 °C), affect adsorption efficiency. The best possible combinations of these parameters were examined by Response Surface Methodology (RSM) and Artificial Neural Network (ANN). R2 values for RSM and ANN were 0.9964 and 0.9666, respectively. According to the models, the adsorption capacity under the optimum conditions determined for the RSM and ANN model was found to be 134.62 mg/g and 125.12 mg/g, respectively. © 2022 Elsevier Ltd, Türkiye Atom Enerjisi Kurumu, TAEK; Türkiye Bilimsel ve Teknolojik Araştırma Kurumu, TÜBİTAK: 1001-120M235, The authors acknowledge grants from the Scientific and Technological Research Council of Turkey (TUBITAK, project number: 1001-120M235 ). All radionuclide adsorption experiments performed in the study were carried out in Ege University Institute of Nuclear Sciences, which is also a partner in the project, accredited by the Turkish Atomic Energy Authority (TAEK)., The authors acknowledge grants from the Scientific and Technological Research Council of Turkey (TUBITAK, project number: 1001-120M235). All radionuclide adsorption experiments performed in the study were carried out in Ege University Institute of Nuclear Sciences, which is also a partner in the project, accredited by the Turkish Atomic Energy Authority (TAEK).

Nuclear technologies have strong potential and a unique role to play in delivering reliable low carbon energy to enable a net-zero society for future generations. However, to assure the sustainability required for its long-term success, nuclear will need to deliver innovative solutions as proposed in iMAGINE. One of the most attractive features, but also a key challenge for the envisaged highly integrated nuclear energy system iMAGINE, is the need for a demand driven salt clean-up system based on the principles of reverse reprocessing. The work described provides an insight into the dynamic interplay between a potential salt clean-up system and reactor operation in a plutonium-started core in a dynamic approach. The results presented will help to optimise the parameters for the salt clean-up process as well as to understand the differences which appear between a core started with enriched uranium and plutonium as the fissile material. The integrated model is used to investigate the effects of the initial fissile material on core size, achievable burnup, and long-term operation. Different approaches are tested to achieve a higher burnup in the significantly smaller Pu-driven core. The effects of different clean-up system throughputs on the concentration of fission products in the reactor salt and its consequences are discussed for general molten salt reactor design. Finally, an investigation into how a plutonium loaded core could be used to provide fuel for future reactors through fuel salt splitting is presented, with the outcome that one Pu-started reactor of the same size as a uranium-started core could deliver fuel for 1.5 new cores due to enhanced breeding. The results provide an essential understanding for the progress of iMAGINE as well as the basis for inter-disciplinary work required for optimising iMAGINE.

High-reliability organizations (HROs) including commercial airlines, the NASA Space Shuttle engineering team, US Naval aircraft carrier and nuclear submarine crews, and US nuclear power plants are relatively safe. However, these organizations experience system breakdowns often with catastrophic outcomes. This study focuses on risk information management strategies employed by a nuclear energy power plant located within 10 miles of a population center. The evacuation planning zone (EPZ) includes a hospital, several schools, and a public university. The nuclear plant provides written preparedness and evacuation information for all residents within the EPZ in the event of a radiological emergency. Focusing on the campus community within the EPZ, this study investigates individual awareness regarding the potential of a radiological event, the emergency information booklet, and the information provided within the booklet. We use descriptive statistics, frequency distribution, and cross tabulations (contingency tables) to establish awareness levels. Our study determines those participants who read the emergency instructions booklet are prepared to make an informed decision in the event of a radiological incident. We find college affiliation, educational level, university role, and age significantly related to emergency instruction booklet utilization. We also find gender is significantly linked to overall risk perception regarding a radiological event. Findings support previous research regarding women’s higher levels of risk aversion and pessimism involving dangerous new risky technologies and activities like nuclear energy plants. These findings support modifying policy to ensure nuclear facilities assess the efficacy of their warning systems in alerting the public. Furthermore, our findings provide guidance regarding the evaluation of the effectiveness of emergency instruction booklet distribution.

Molten salt reactors (MSRs), as one of the six main technologies of Gen IV, can meet the broad area of sustainability, economics, safety and reliability, proliferation resistance and physical protection goals. One of the main and first challenges in designing molten salt fast reactors (MSFRs) is the selection of an appropriate molten salt fuel system based on the envisaged applications and objectives. In this study’s series, a full-scope evaluation has been conducted about employing either chloride or fluoride salt fuels as the main competitors’ candidates for fuel salt in MSFR designs. Two distinguished projects, EVOL (CNRS, Grenoble-France), based on fluoride salt, and iMAGINE (The University of Liverpool, UK), based on chloride salts, were considered in order to achieve this goal as case studies. The first part of this series (part 1—this article) deals with the investigation of the thermophysical properties of the salt fuel system, criticality search and neutron-flux energy spectrum. An identical typical design of the MSFR core has been considered for a neutronic simulation by using MCNPX V2.7 based on the chemical composition of the fuel salt mentioned in both projects. The thermophysical evaluation has been conducted through literature research and theoretical methods based on the experimental values for the salt component properties. The results of the study are presented for thermophysical properties, including the melting point, vapour pressure/boiling point, specific heat capacity, thermal conductivity and density, in addition to neutronic simulation for the core critical dimension and neutron-flux spectrum of both the chloride- and fluoride-based salt fuel systems. In the discussion of the results, it is concluded that both the chloride and fluoride salt fuel systems have benefits that are seen on different comparative parameters. The delivered data will provide future decision makers with evidence for the salt choice for balancing their design objectives with the opportunities and expectations. Thus, a final selection of the most appropriate salt fuel system to be used in MSFRs will be postponed for more investigation in the final part of this article series, combining the data with different potential user profiles.

Traditionally, there has been a gap between reactor operation and the consideration of nuclear waste in the final disposal. Fuel is produced, and fuel must be disposed. In the view of the reactor operator, fuel has to be cleaned in the reprocessing, and new solid fuel has to be produced in the view of the chemist. iMAGINE is designed to overcome this separation through a breakthrough development applying an optimized, integrative approach from cradle to grave of nuclear energy production as a first step to come as close as possible to the vision of zero waste nuclear power. It is described here for the first time in three steps: reactor, fuel cycle, and waste, providing the rationality behind each of the choices made to come to the overall solution to open the discussion and thinking process on what could be achieved by a very innovative approach to integrated nuclear energy production. The opportunities regarding the handling of the remaining waste are discussed with a view on the expectation of the final disposal community, the study “Nuclear waste from small modular reactors”, and the IAEA report “waste from innovative types of reactors and fuel cycles—a preliminary study”. The aim of this work is not to find answers to each of the raised points, but to identify potential approaches and promising ways to go, as well as to stimulate a discussion among experts. In the best case, this could lead to a change of track for nuclear power to become even more sustainable and an important, trusted technology to help solve the net-zero challenge.

Under some unexpected conditions, drive rods and control-rod assemblies may not be disconnected. If this situation is not detected, the control rod will be lifted out of the reactor core together with the upper reactor internals. This situation will seriously affect the follow-up work and reduce the economy and safety protection of the nuclear power plant. To ensure safety, the tripping status must be checked after tripping. Follow-up work can be carried out after checking and confirming that all drive rods are in the tripping status. There are many problems for traditional inspection methods, such as misjudgment, low accuracy, and labor consumption. This paper proposes a visual inspection system for the uncoupling state of the control-rod drive rod of the nuclear reactor. The proposed method is based on the fitting model of the ellipse parameter of the drive-rod head and the height of the drive rod. The ellipse of the drive-rod head is firstly accurately detected. Then, a mathematical model between the ellipse parameter and the height of the drive rod is established. The measurement error caused by the swing of the head of the drive rod is eliminated. The accurate measurement of the height difference before and after the tripping of the drive rod is computed. Finally, the status of the uncoupling of the drive rod is judged according to the difference. Many experiments are carried out with our developed system. The experimental results show that the proposed system realizes remote operation, ensures the quality of trip-status inspection, improves work efficiency, and reduces the workload of staff.

Following the recent surge in harnessing clean energy sources to fast-track carbon neutrality, renewable and nuclear energies have been the best-rated sources of clean energy. Even though renewable energy presents an almost insignificant risk to public health and the environment, they are insufficient to support the growing demand for the high energy required for industrialization. Despite the competitive potential of nuclear energy to meet these demands, public concerns about its safety have significantly hindered its mass deployment in developing countries. Therefore, one of the primary considerations in commissioning a nuclear power plant is the establishment of emergency planning zones based on the reactor type and other siting criteria. Based on Ghana’s reactor type assessment (RTA), four reactor designs were considered in this study which are APR1400, HPR1000, VVER1200, and Nuscale Power Module. Using the NRC’s SNAP/RADTRAD and RASCAL codes, this research sought to investigate radionuclide doses released at the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), Control room (CR), and the 16 km recommended public safe zone during Fuel handling Accidents (FHA), Rod Ejection Accident (REA), and Long-Term Station Blackout (LTSBO). The results revealed that reactors’ power contributed to the source term activities and offsite consequences during REA and LTSBO, while FHA was predominantly affected by the number of fuel assemblies and a fraction of damaged fuel. Additionally, the accidents considered in this study followed a similar trend of impact in decreasing order of reactor power and the number of fuel assemblies; APR1400 < VVER1200 < HPR1000 < Nuscale. Nevertheless, all the doses were within regulatory limits.

Nuclear technologies have the potential to play a unique role in delivering low carbon energy for a future net-zero society. However, for long-term success, nuclear technologies will need to deliver innovative solutions as proposed in iMAGINE. One of the key challenges for the envisaged highly integrated nuclear energy system is the need for a demand-driven salt clean-up system. The work described provides an insight into the interplay between a potential salt clean-up system and the reactor operation in a dynamic approach. The results provided will help to optimise the parameters for the salt clean-up process by delivering a dynamically calculated priority list, identifying the elements with great influence on reactor operation. The integrated model is used to investigate the ideal time for the initiation of the clean-up as well as the effect of different throughputs through the clean-up system on criticality as well as on the concentration of the elements in the reactor salt. Finally, a staggered approach is proposed with the idea to phase in the chemical clean-up processes step by step to keep the reactor critical. The results provide an essential step for the progress of iMAGINE as well as a basis for the interdisciplinary work required to bring iMAGINE into real operation.

347 p. El objetivo de esta Tesis es el estudio de la distribución de la tasa de dosis dentro y fuera de un submarino de propulsión nuclear, considerando diferentes parámetros de influencia como son el término fuente considerado, los materiales de blindaje seleccionados o la configuración geométrica del modelo, todo ello considerado en operación normal.Se analizarán tanto el blindaje primario, el que envuelve a la vasija, como el blindaje secundario, que se sitúa en las zonas de los mamparos y el casco, formando parte del recinto de contención.El estudio de los parámetros de influencia se abordará de una forma analítica en primera instancia, profundizando posteriormente en el desarrollo de una herramienta (script de Python) sistemática que permitirá obtener como resultados los valores optimizados en dimensiones y materiales para un modelo de transporte con un término fuente dado.

The synthesis and extensive characterization of newly developed boron nitride nanosheet (BNNSs)-reinforced WNiCoFeCr high-entropy alloys (HEAs) are presented. The influence of B4C on the structural, physical, mechanical, and nuclear shielding characteristics of synthesized HEAs has been widely examined in terms of its monotonic effects on the behavior changes. The internal morphology and structural characteristics of the fabricated composites are first investigated using X-ray diffraction, scanning electron microscopy, and energy-dispersive spectroscopy. Wear testing is used to determine the coefficient of friction as a function of sliding distance. Experimental gamma ray and neutron setups are created to determine their shielding characteristics against nuclear radiation. Finally, the shielding characteristics of nuclear radiation for gamma ray and fast neutrons are compared extensively to those of many existing and new-generation shielding materials. Among the examined samples, the S2 sample with B4C and BNNSs reinforcement had the greatest mechanical characteristics. Our findings imply that increasing B4C directly contributes to the shielding qualities of nuclear radiation. The B4C created in the structure of BNNSs contributes to the overall properties of HEAs, which are crucial for nuclear applications, since HEAs are being examined as a component of future nuclear reactors. Additionally, B4C is a very versatile material that may be used in circumstances where mechanical and nuclear shielding properties need to be enhanced for a variety of radiation energies. © 2022, The Author(s), under exclusive licence to Springer-Verlag GmbH, DE part of Springer Nature. 2-s2.0-85134366577

Providing green energy with less emission is an important goal in the European union to achieve sustainable energy production by 2050. Intermediate stages such as the use of nuclear power facilitate the goal of transition to sustainable energy. Nuclear power is a reliable and carbon free technology that can be used to allow the energy transition. The scope of this work is to facilitate the energy transition to sustainability through hybrid systems. Numerous hybrid systems and their concepts were studied and adapted to make a new conceptual hybrid power plant. The case of hybrid considered in this work only constitutes renewable and nuclear hybrids. The very high temperature reactors (VHTR) and solar thermal power works and generates high temperatures. The best utilization of the heat from these sources would be to save them in salt storage and later used to match the demand requirements. Moreover, the VHTR is the most promising design to deal with high temperature. Thus from a VHTR design the cycle is modified to increase flexibility. A modified Brayton cycle is introduced in the power generation process with salt reheating and increasing the production by increasing the areaunder the curve in the TS diagram. Data used in the work involves the Spanish energy production and distribution. This data is used in the model calculation for future hybrid production. Finally, the hybrid system is tested for operating conditions and parameters, such as solar field area and tank size. Optimal operation conditions and parameters are expected to make the energy transition to become sustainable

"Retinoid" is the general term for vitamin A derivatives and chemical compounds that act like vitamin A. Vitamin A are composed of four isoprene units and are named according to their terminal functional group, such as retinol (OH, 1), retinal (CHO, 2), and retinoic acid (CO2H, 3). Vitamin A usually refers to retinol. In the past few decades, major advances in research on vitamin A have improved our understanding of its fundamental roles and physiological significance in living cells. In this review, three types of chemical biology studies using vitamin A analogs are described: (1) conformational studies of the chromophore in retinal proteins (rhodopsin, phoborhodopsin, and retinochrome), especially the conformation around the cyclohexene ring; (2) structure-activity relationship studies of retinoic acid analogs to create new signaling molecules for activating nuclear receptors; and (3) development of a new channelrhodopsin with an absorption maximum at longer wavelength to overcome the various demerits of channelrhodopsins used in optogenetics, as well as the stereoselective synthesis of retinoid isomers and their analogs using a diene-tricarbonyliron complex or a palladium-catalyzed cross-coupling reaction between vinyl triflates and stannyl olefins.

Background The etiology of brain cancer is poorly understood. The only confirmed environmental risk factor is exposure to ionizing radiation. Because nuclear reactors emit ionizing radiation, we examined brain cancer incidence rates in the USA in relation to the presence of nuclear reactors per state. Methods Data on brain cancer incidence rates per state for Whites by sex for three age groups (all ages, 50 and older, and under 50) were obtained from cancer registries. The location, number, and type of nuclear reactor, i.e., power or research reactor, was obtained from public sources. We examined the association between these variables using multivariate linear regression and ANOVA. Results Brain cancer incidence rates were not associated with the number of nuclear power reactors. Conversely, incidence rates per state increased with the number of nuclear research reactors. This was significant for both sexes combined and for males in the ‘all ages’ category (β = 0.08, p = 0.0319 and β = 0.12, p = 0.0277, respectively), and for both sexes combined in the’50 and older’ category (β = 0.18, p = 0.0163). Brain cancer incidence rates for counties with research reactors were significantly higher than the corresponding rates for their states overall (p = 0.0140). These findings were not explicable by known confounders. Conclusions Brain cancer incidence rates are positively associated with the number of nuclear research reactors per state. These findings merit further exploration and suggest new opportunities for research in brain cancer epidemiology.

Nuclear technologies have the potential to play a major role in the transition to a global net-zero society. Their primary advantage is the capability to deliver controllable 24/7 energy on demand. However, as a prerequisite for successful worldwide application, significant innovation will be required to create the nuclear systems of the 21st century, the need of the hour. The pros (low harmful emissions, high reliability, low operational expenses, and high energy density) and cons (environmental damage, fuel waste disposal concerns, limited uranium reserves, and long construction time-frame) of nuclear are discussed and analysed at different levels—the societal and public recognition and concerns (accidents, weapons, mining, and waste) as well as the scientific/engineering and economic level—to assure a demand-driven development. Based on the analysis of the different challenges, a vision for the nuclear system of the 21st century is synthesised consisting of three pillars—unlimited nuclear energy, zero waste nuclear, and accident-free nuclear. These three combined visions are then transformed into dedicated and verifiable missions that are discussed, in detail, regarding challenges and opportunities. In the following, a stepwise approach to the development of such a highly innovative nuclear system is described. Essential steps to assure active risk reduction and the delivery of quick progress are derived as answers to the critique on the currently observed extensive construction time and cost overruns on new nuclear plants. The 4-step process consisting of basic studies, experimental zero power reactor, small-scale demonstrator, and industrial demonstrator is described. The four steps, including sub-steps, deliver the pathway to a successful implementation of such a ground-breaking new nuclear system. The potential sub-steps are discussed with the view not only of the scientific development challenges but also as an approach to reduce the regulatory challenges of a novel nuclear technology.

Animal experiment findings suggest that high doses of ionizing radiation exposure (>1.0 Gy) may cause genetic and epigenetic effects in offspring. However, epidemiological studies of offspring of radiation-exposed parents did not find increased risks of any health effects. Findings of cellular/experimental investigations and studies of human health effects are contradicting, and further investigations are needed to help resolve ambiguities using updated and/or improved data. This paper provides a detailed description of a database of families of workers of the first Russian nuclear facility, Mayak Production Association, located in the Southern Urals in the Chelyabinsk region close to Ozyorsk city, which started its operation in 1948 and today consists of reactors, radiochemical and plutonium production plants, and auxiliary facilities. The Mayak worker cohort includes 22,377 individuals (25% females) who were hired at one of the main Mayak PA facilities between 1948 and 1982 and were externally or internally exposed to ionizing radiation over prolonged periods. Advantages of the cohort include its large size, extensive follow-up period (70 y), individually measured doses from external and internal exposure and the wide range of these doses, heterogeneity by gender/age/ethnicity/initial health status, complete data on vital status and causes of death, available medical information on morbidity and reproduction, available data on non-radiation factors, and stored biological specimens donated by more than one-third of the cohort members. Based on medical and dosimetry database "Clinics" containing raw data on workers of the study cohort, the Mayak workers' family and offspring database was created. To date, it comprises 12,195 family couples (a husband and a wife) and 16,585 offspring. Biological specimens are available for more than 1,000 family triads (a husband, a wife, and their child). Stages of assembling the database and its descriptive characteristics are presented in this paper. Examples of potential applications of the database for investigations of non-targeted and transgenerational radiation effects in offspring of exposed parents are discussed.

Supersonic shear layer development affects mixing between the primary and secondary fluids in a supersonic ejector substantially. The splitter plate plays a key role in shear layer development in the near-trailing-edge region. In this paper, supersonic shear layers with various splitter plate thicknesses are calculated and analyzed. Two types of instability modes are found when various splitter plate thicknesses are used. The shock wave strength and absolute unstable region determine the instability modes. Shock waves can limit instability development but an absolute unstable region promotes instability development. Velocity thickness growth rates are found to be the result of the merging process between two or more adjacent vortices in the transition region. The number of vortices that merge depends on the ratio of the vortex shedding and nature frequencies. The velocity thickness growth rates are larger under the local instability mode than under the global instability mode since three vortices merge under the former mode but two vortices merge under the latter. This study may provide a helpful reference for the design of supersonic ejectors.

Nuclear power has a prominent role to play in the energy landscape of the 21st century. To- gether with large reactor designs that make the most out of economies of scale, small, modular reactors (SMR) may also play a role thanks to their potential modular, factory-based and fast deployment, which have the potential to minimize the capital cost of nuclear, which is one of its main drawbacks. Among such designs, lead-cooled fast reactors (LFR) present some unique strengths, among others their lack of pressurization of the primary system and their potential to passively remove decay heat by means of natural convection. The SUNRISE project, in Sweden, is an effort by, among others, KTH to prove the feasability of this technology. In that context, the analytical design tool ADELE, which draws from the BELLA code produced at KTH, is developed. Its goal is to provide a fast-to-iterate algorithm that uses basic physics constraints to obtain the core design without relying on more computationally costly methods, like Monte Carlo simulations, which are regularly used for that purpose. Specifically, a system that is capable of removing all the residual heat via convective cooling is obtained from a short list of basic input values. This algorithm is implemented into a mobile application, with the goal of being available to Nuclear Engineering students. The resulting code is capable of replicating reference core designs with a reasonable degree of accuracy, given the simplicity of the assumptions used in the process.

Achieving nuclear fusion as an energy source on Earth is a practical goal that relies on continuing scientific and engineering innovation. Functional fusion reactors around the world today allow scientists and engineers to plan improvements that will eventually allow for greater energy output than the input required to operate the machine (including heating the plasma and operating the superconducting electromagnets that confine the plasma, among other energy inputs). The fusion reaction between nuclei of hydrogen isotopes is a carbon-free source of massive amounts of energy that could be paramount in a global turn towards greener energy. The fusion fuel needed to provide one person’s energy use for 100 years (assuming 20 kWh per day) is contained within roughly one and a half bathtubs of water and 3 laptop batteries. Given the enormous payoff of fusion, continued research and development are of great interest so that current challenges of heating and confining plasma, mitigating plasma disruptions, improving efficiency of magnets, and extending the lifetime of materials subjected to the harsh conditions surrounding the plasma may be overcome. Fusion reactor materials research carried out here at the BSC contributes to this ambitious goal. The idealistic goal for fusion materials research is to provide predictions about material behavior with the accuracy of quantum mechanical calculations at the scale of a full fusion reactor. Using strategic approximations and working at a small scale, computational fusion materials researchers can accurately reproduce and explain experimentally observed physical phenomena, such as the formation of microstructural defects in metals under neutron-irradiation, and offer the best predictions available for behavior of materials in future fusion reactor environments, where data about what will happen simply do not exist yet. In the study presented here, we examined the thermal conductivity, or how quickly a material allows heat to flow, of tungsten (W). W has been selected for plasma-facing components in ITER, which is currently under construction. We used LAMMPS atomic modelling of materials software and found that the thermal conductivity of W is significantly decreased in the presence of defects.

On February 24, 2022, Russia began a military invasion of Ukraine. Missile and air strikes were reported throughout the country, shortly followed by a large ground invasion from multiple directions. Four major theaters developed: the Kyiv offensive, the Northeastern Ukraine offensive, the Eastern Ukraine offensive, and the Southern Ukraine offensive, with continued missile and air strikes far into Western Ukraine. Advancing Russian military units launched an attack and captured the Chernobyl nuclear station. Russian troops dug trenches into the area commonly known as the “Red Forest,” violating the established radiation safety measures and threatening security within the Chernobyl Exclusion Zone. The placement of military units in such close proximity to the station also sparked concerns of possible damage occurring to the containment vessel constructed around the station’s wrecked fourth reactor. There are 15 operating nuclear reactors in Ukraine. Each is vulnerable to an attack or sabotage that could precipitate a malfunction and possible release of radioactive isotopes. In this short commentary, we will discuss radiobiologic data obtained after the analysis of historical nuclear power plant (NPP) accidents and emphasize new challenges for nuclear security when NPPs are found and are possible targets within a conflict zone.

Nuclear fission technologies have the potential to play a significant role in the energy mix of a net-zero and sustainable society. However, to achieve the sustainability goal two significant challenges remain: efficient and sustainable fuel usage and the minimization of long-term nuclear waste. Civil nuclear molten salt systems and technologies offer the opportunity to address both, delivering future reactors at scale for efficient and effective power production and nuclear waste burnup. Potentially, both objectives could be fulfilled in one reactor system, which could significantly improve sustainability indices. The key to this innovation is demand driven development of a significantly reduced fuel cycle with enhanced proliferation resistance which offers further potential for improvement. To achieve these goals, a transformative approach for salt clean-up during molten salt reactor operation is proposed, by concentrating on the detection and removal of key neutron poisoning elements which prevent the reactor from long-term operation. To enable this highly innovative development work, a novel analysis of the evolving elementary fuel composition, their concentrations, and their criticality influence is now provided in this work. This, combined with consideration of the oxidation states of each of these elements then provides the basis for the selection of these key poisons and the development of advanced separation processes and process monitoring. This work also discusses the importance of the effective integration of physics and chemistry when systems modelling in achieving these system development goals.

[ES] Ante el desafío que implica la reducción de los efectos del cambio climático, la industria nuclear se ha postulado como una buena alternativa para sustituir la producción de energía eléctrica a partir de combustibles fósiles. No obstante, debe constatar la seguridad de las centrales, para lo que resulta indispensable poder predecir su comportamiento ante escenarios operacionales y accidentales. A tal efecto, y dada la imposibilidad de disponer de datos de planta para analizar estos transitorios, se generan bases de datos en instalaciones a escala reducida a partir de experimentos, siendo necesarios métodos y estrategias de escalado que permitan extrapolar los comportamientos termohidráulicos. Pese a la relevante contribución que suponen los experimentos al campo de la seguridad nuclear, en ocasiones se cuestiona la validez de sus resultados para reproducir el comportamiento de las centrales. Este hecho motiva la ejecución de test counterpart entre distintas instalaciones, que contribuyen a abordar la problemática del escalado, así como a demostrar la adecuación de los códigos termohidráulicos para predecir una respuesta realista de los sistemas. La presente tesis doctoral explora la posibilidad de aumentar el número de experimentos counterpart a partir de la definición de nuevos escenarios y su simulación con el código termohidráulico TRACE5. Con este fin, se han desarrollado modelos de las instalaciones ATLAS y LSTF, y se han estudiado y simulado experimentos counterpart ya existentes entre dichas instalaciones. La identificación de los fenómenos termohidráulicos más significativos, y el análisis de su escalado y distorsión, configuran la base de conocimientos para abordar el diseño de los nuevos test. En la tesis, en particular, se plantea un escenario tipo station blackout para LSTF partiendo de las condiciones iniciales y de contorno de un test previo en ATLAS. La simulación del experimento confirma la idoneidad de ATLAS y LSTF para realizar experimentos counterpart, en los que la fenomenología relevante es similar, y pone de manifiesto algunas limitaciones de estas instalaciones en cuanto a la extrapolabilidad de ciertos fenómenos, debido a las distorsiones originadas por la diferencia de escala y tecnología., [CA] Davant del desafiament que implica la reducció dels efectes del canvi climàtic, la indústria nuclear s'ha postulat com una bona alternativa per a substituir la producció d'energia elèctrica a partir de combustibles fòssils. No obstant això, ha de constatar la seguretat de les centrals, per al que resulta indispensable poder predir el seu comportament davant d'escenaris operacionals i accidentals. A aquest efecte, i donada la impossibilitat de disposar de dades de planta per a analitzar aquests transitoris, es generen bases de dades en instal·lacions a escala reduïda a partir d'experiments, sent necessaris mètodes i estratègies d'escalat que permeten extrapolar els comportaments termohidràulics. Malgrat la rellevant contribució que suposen els experiments al camp de la seguretat nuclear, de vegades es qüestiona la validesa dels seus resultats per a reproduir el comportament de les centrals. Aquest fet motiva l'execució de test counterpart entre distintes instal·lacions, que contribuïxen a abordar la problemàtica de l'escalat, així com a demostrar l'adequació dels codis termohidràulics per a predir una resposta realista dels sistemes. La present tesi doctoral explora la possibilitat d'augmentar el nombre d'experiments counterpart a partir de la definició de nous escenaris i la seua simulació amb el codi termohidràulic TRACE5. Amb aquest fi, s'han desenvolupat models de les instal·lacions ATLAS i LSTF, i s'han estudiat i simulat experiments counterpart ja existents entre les dites instal·lacions. La identificació dels fenòmens termohidràulics més significatius, i l'anàlisi del seu escalat i distorsió, configuren la base de coneixements per a abordar el disseny dels nous test. En la tesi, en particular, es planteja un escenari tipus station blackout per a LSTF partint de les condicions inicials i de contorn d'un test previ en ATLAS. La simulació de l'experiment confirma la idoneïtat d'ATLAS i LSTF per a realitzar experiments counterpart, en els que la fenomenologia rellevant és semblant, i posa de manifest algunes limitacions d'aquestes instal·lacions quant a l'extrapolabilitat de certs fenòmens, a causa de les distorsions originades per la diferència d'escala i tecnologia., [EN] Faced with the challenge of reducing the effects of climate change, the nuclear industry has been postulated as a good alternative to replace the production of electricity from fossil fuels. However, it must verify the safety of the plants, for which it is essential to be able to predict their behavior in operational and accidental scenarios. To this end, and given the impossibility of having plant data to analyze these transients, databases are generated in reduced-scale facilities from experiments, being necessary scaling methods and strategies that allow the extrapolation of thermohydraulic behaviors. Despite the relevant contribution that experiments make to the field of nuclear safety, the validity of their results to reproduce the behavior of plants is sometimes questioned. This fact motivates the execution of counterpart tests between different facilities, which contribute to addressing scaling issues, as well as to demonstrate the adequacy of the thermal-hydraulic codes to predict a realistic response of the systems. This Ph.D. Thesis explores the possibility of increasing the number of counterpart experiments based on the definition of new scenarios and their simulation with the TRACE5 thermal-hydraulic code. In order to achieve this goal, models of the ATLAS and LSTF facilities have been developed, and counterpart experiments already existing between these facilities have been studied and simulated. The identification of the most significant thermal-hydraulic phenomena and the analysis of their scaling and distortion, configure the knowledge basis to approach the design of the new tests. In the Thesis, in particular, a station blackout scenario for LSTF based on the initial and boundary conditions of a previous test in ATLAS is proposed. The simulation of the experiment confirms the suitability of ATLAS and LSTF to perform counterpart experiments, in which the relevant phenomenology is similar. Moreover, it reveals some limitations of these facilities in terms of the extrapolability of certain phenomena, due to the distortions caused by the difference in scale and technology.

The knowledge of (n,2n) cross sections is required in shielding and breeding calculations. It's also important, in nuclear reactor applications, for neutron dosimetry research. In nuclear reaction mechanism, (n,2n) reaction channel is the dominant reaction for medium and heavy mass nuclei at 14 MeV energy range. The empirical and semi-empirical (n,2n) reaction cross sections have been investigated by many authors, but theoretical calculations have not been adequate because of the character of the nuclear structure is not exactly known. In this work, a new empirical formula has been proposed to calculate the (n,2n) reaction cross sections at 14-15 MeV neutron incident energy. In the calculations different fitting method have been used to obtain cross-section formula for target nuclei in the range 14 ≤ A ≤ 241 mass number. The (n,2n) experimental cross sections have been taken from EXFOR nuclear data library. In this new formula, the Flerov and Talyzin expression for the inelastic cross section σ

Recent incidents at nuclear facilities in Ukraine related to the attacks from Russian forces highlight the fragility of nuclear power plants and other nuclear facilities in war and the very real potential for another environmental nuclear disaster and associated health risks in Europe. Nuclear catastrophes from war can occur from radioactive materials released from war threatened nuclear power plants and other nuclear facilities in war zones, in addition to the direct threat from the deployment of nuclear weaponry and can result in immediate and long-term health impacts. Despite historical nuclear catastrophic events, including the Chernobyl nuclear power plant accident and atomic bombings of Hiroshima and Nagasaki, and that for more than a century epidemiologists have studied the consequences of radiation exposures, there are still major unanswered questions regarding radiation risks and human health. Epidemiologists will need to continue to quantify the health effects from exposure to environmental radiation, including background radiation, and are able to contribute to conversations about reliance on nuclear energy and alternative energy futures. As a society we are compelled to rethink our ties to nuclear energy, especially with the potential of increasing reliance on nuclear power amid oil and gas crisis and considering climate change, nuclear warfare, including nuclear weapon testing, and the fragility of humanity and health to even low doses of radiation from these and other natural and unnatural sources.

A fire in a nuclear reactor at Windscale Works (Sellafield, England) in October 1957 led to an uncontrolled aerial release of radionuclides. At the time of the accident air was sampled at various locations in Europe to monitor atmospheric pollution, and the opportunity was taken to measure the sampling filters for activity concentrations of iodine-131, caesium-137 and polonium-210 at the Harwell research establishment (United Kingdom); when it was not possible to perform measurements at Harwell, original measurement data were supplied. This programme of activity measurements was performed in the context of work by the Advisory Committee on Nuclear Radiation of the International Geophysical Year (IGY; July 1957-December 1958). The International Geophysical Year was an international programme of research into a comprehensive range of geophysical phenomena. The results of this measurement programme were originally reported in Harwell Memorandum AERE-M857 (1961) and this Harwell report is reproduced in this paper because of its historical interest and because it is no longer readily accessible to researchers.

The excess relative risk (ERR) of mortality for circulatory disease among nuclear workers was reanalyzed by taking into consideration the annual dose as the dose rate using publicly available epidemiological data of the Hanford site dedicated to the cohort study of nuclear workers in the US, the UK, and Canada. Values of the dose rate (cut-points) were chosen at 2 mSv y intervals from 2 to 40 mSv y, and risk estimates were made for 32,988 workers, considering the doses accumulated below and above each cut-point to have different effects. Similarly to that in the previous study for cancer by Sasaki et al., examinations of sensitivity analysis were also carried out for different risk models, lag periods, and impacts of adjusting the monitoring period to find the effect of the dose rate. As a result, emergence of a statistically significant difference between βL1, which is the ERR for the doses accumulated below the specified cut-point, and βH1, which is the ERR for that above the specified cut-point, was observed for cut-point of the dose rate of 4, 6, 34, 36, and 38 mSv y. While statistically negative values were estimated for βL1 (4 and 6 mSv y) and for βH1 (34, 36, and 38 mSv y), the overall relationship between the ERR and the cut-point of the dose rate was found to be similar to that obtained by the analysis of the mortality for all cancers excluding leukemia.

The Comprehensive Nuclear-Test-Ban Treaty (CTBT) specifies that an overall network of at least 40 International Monitoring System (IMS) stations should monitor the presence of radioxenon in the atmosphere upon its entry into force. The measurement of radioxenon concentrations in the air is one of the major techniques to detect underground nuclear explosions. It is, together with radionuclide particulate monitoring, the only component of the network able to confirm whether an event originates from a nuclear test, leaving the final proof to on-site inspection. Correct and accurate interpretation of radioxenon detections by State Signatories is a key parameter of the verification regime of the Preparatory Commission for the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO). In this context, the discrimination between the highly variable radioxenon background generated by normal operations of nuclear facilities and CTBT-relevant events is a challenging, but critical, task. To this end, the radioxenon background that can be expected at IMS noble gas systems must be sufficiently characterized and understood. All activities conducted to study the global radioxenon background are focused on the calibration and performance of the verification system as described in the Treaty. The unique CTBTO noble gas system network is designed to optimally covering the globe. By the end of 2019, 31 systems were put in operation, 25 of which being already certified. It took two decades from the first experimental setup of noble gas system in the field to reach this stage of maturity. In the meantime, it was an urgent need to gain empirical evidence of atmospheric radioxenon concentrations with the full spectrum of characteristics that IMS noble gas systems may be observing. This experience was significantly advanced through temporary measurement campaigns. Their objective was to gain the additional necessary knowledge for a correct understanding and categorization of radioxenon detections. The site selection for these campaigns put emphasis on regions with low coverage by the initially few experimental noble gas systems at IMS locations or where potential interferences with normal background might be observed. Short-term measurements were first initiated in 2008. Sites of potential interest were identified, and campaigns up to few weeks were performed. Based on the findings of these short campaigns, transportable systems were procured by the CTBTO. Longer temporary measurement campaigns were started afterwards and operated by local hosts in different regions of the globe. Site selections were based on purely scientific criteria. Objectives of the measurement campaigns were continually reassessed, and projects were designed to meet the scientific needs for radioxenon background understanding as required for nuclear explosion monitoring. As of today, several thousands of samples have been collected and measured. Spectra of temporary measurement campaigns were (and are still) analysed in the International Data Centre (IDC). As they are not part of the CTBT monitoring system, no IDC product is generated. Analysis results are stored in a non-operational database of the CTBTO and made available, together with raw data, to authorized users of States Signatories through a Secure Web Portal (SWP) and to scientific institutions for approved research projects through a virtual Data Exploitation Centre (vDEC) after signing a cost-free confidentiality agreement (https://www.ctbto.org/specials/vdec). This paper aims at providing an overview of the temporary measurement campaigns conducted by the CTBTO since the very first field measurements. It lays out scientific results in a systematic approach. This overview demonstrates the asset of radioxenon background measurement data that have been collected with a wide variety of characteristics that may be observed at IMS stations. It bears a tremendous opportunity for development, enhancement and validation of methodologies for CTBT monitoring. In 2018, a campaign started in Japan with transportable noble gas systems in the vicinity of the IMS station RN38 in Takasaki. It will be described separately once the measurements are completed.

In the Republic of Kazakhstan, within the framework of international initiatives on the non-proliferation of nuclear weapons, the IVG.1M research reactor is being converted to low enriched fuel. Due to the change in the fuel composition, it becomes necessary to assess the power density distribution in the fuel assemblies and in the core altogether. Close attention to this parameter is explained by the fact that the non-uniformity of the radial and axial power density fields determines the operating parameters of the reactor, its power, the core resource and the efficiency of fuel utilization. The purpose of this work is approbation of the activation gamma-spectrometry method for determining the power density distribution in the fuel of the IVG.1M research reactor after conversion. Additionally, for a preliminary assessment of the experimental outcomes, the paper presents the MCNP6 calculation results for the power distribution in fuel assemblies and the peaking factors.

In recent years Small Modular Reactors (SMR) have become very popular within the nuclear industry. These designs allow to reduce costs as well as to enhance the safety due to passive nuclear safety features. Within these systems, the integral Pressurized Water Reactors (iPWR) are very extended because they take advantage of the previous technology developed for Gen II and III PWRs. In this sense, previous Best Estimate system codes like RELAP5 or CATHARE seem to be reliable for Deterministic Safety Assessment (DSA) but need to be assessed for new passive systems in which natural circulation takes a key role. In the present paper, Energy Software Ltd., in collaboration with the UPC, has developed an iPWR input model for both NRC RELAP5 and ISS RELAPSCDAPSIM codes. These models, based on CAREM-25 publicly available data, simulate an SBO Fukushima likewise scenario. Results under Design Basis Accident (DBA) conditions are benchmarked to assess the reliability of the codes to reproduce the plant availability reported in the collected data. Passive systems like Safety Injections and Residual Heat Removal Exchangers have also been included to analyze the code capabilities to reproduce natural circulation under iPWR conditions. Finally, core damage progression is simulated with SCDAP components to analyze the severe accident related phenomena. Results of both simulations seem to confirm the 36 hours grace period for SBO scenario of the CAREM-25 design plus the extended 36 hours grace period associated to the availability of Emergency Injection System (EIS) in Loss of Coolant conditions reported by designer.

We open an avenue for discussing how we can pave the way for compliance with existing regulations is a far-reaching factor for settling nuclear fusion technology. Based on a model of the Korean Fusion Demonstration Reactor (K-DEMO) with a target fusion power of 2.2 GW, we assess the intrinsic safety determinants of internal energy sources, the expected radioactive waste, and the tritium management. Regarding these safety factors, we scrutinize the compatibility of the current legislative environment in Korea with K-DEMO and envisage foreseeable obstacles, such as licensing of the nuclear facilities and acceptability of the radioactive waste. Based on precedent licenses for the Korean Superconducting Tokamak Advanced Research (KSTAR) and lessons learned from the International Thermonuclear Experimental Reactor (ITER), we examine hazardous factors that would threaten regulatory compliance of K-DEMO. This approach can help shape a fusion-compatible framework for consolidating the necessary technical provisions and regulatory baselines reflecting social acceptance with a sense of safety. Fusion-compatible aspects in the regulatory environment are discussed, from fusion philosophy to subordinate administrative and technical guidelines, facility classification, and detailed methods guaranteeing integrity and safety. This paper will contribute to the timely settlement of fusion demonstration facilities and subsequent commercial plants.

[EN] This work outlines an approach for localizing anomalies in nuclear reactor cores during their steady state operation, employing deep, one-dimensional, convolutional neural networks. Anomalies are characterized by the application of perturbation diagnostic techniques, based on the analysis of the so-called ¿neutron-noise¿ signals: that is, fluctuations of the neutron flux around the mean value observed in a steady-state power level. The proposed methodology is comprised of three steps: initially, certain reactor core perturbations scenarios are simulated in software, creating the respective perturbation datasets, which are specific to a given reactor geometry; then, the said datasets are used to train deep learning models that learn to identify and locate the given perturbations within the nuclear reactor core; lastly, the models are tested on actual plant measurements. The overall methodology is validated on hexagonal, pre-Konvoi, pressurized water, and VVER-1000 type nuclear reactors. The simulated data are generated by the FEMFFUSION code, which is extended in order to deal with the hexagonal geometry in the time and frequency domains. The examined perturbations are absorbers of variable strength, and the trained models are tested on actual plant data acquired by the in-core detectors of the Temelín VVER-1000 Power Plant in the Czech Republic. The whole approach is realized in the framework of Euratom¿s CORTEX project., The research conducted was made possible through funding from the Euratom research and training programme 2014-2018 under grant agreement No. 754316 for the "CORe Monitoring Techniques And EXperimental Validation And Demonstration (CORTEX)" Horizon 2020 project, 2017-2021.

The European Pressurized Water Reactor (EPR) and Hualong One Pressurized Water Reactor (HPR) are two of the reactors under consideration by the Ghana Nuclear Power Programme. Radiotoxicity analysis of Spent Nuclear Fuel (SNF) assembly was carried out with these commercial Pressurized Water Reactor (PWR) nuclear power technology as case study. This will help determine which one is less radiotoxic on the environment between the two reactor technologies, in the long run. Burnup depletion calculation for the Uranium Oxide (UOX) fuel of these reactor technologies was simulated, using Monte Carlo Neutron Particle Extended (MCNPX), a code used in nuclear fuel management analysis, being a well validated code and also due to its versatile nuclei reactions cross section library. Determination of radiotoxicity for EPR and HPR SNF is the main objective of this study. The radiotoxicity was achieved taking into consideration the radioactive decay rate of the radionuclides and the Dose Factor of each radionuclide present in the SNF using the International Commission on Radiological Protection (ICRP) compendium of Dose Factors due to ingestion. The radiotoxicity for the two reactor's SNF were compared. The initial radiotoxicity for HPR SNF was higher in the duration below one hundred years but at about a hundred years and above, the radiotoxicity was higher for EPR SNF. The radiotoxicity was tremendously reduced for the reprocessed spent UOX fuel (with the Pu and U extracted) to be used as mixed oxide (MOX) fuel. The main finding is that Pu isotopes are the major contributors to the radiotoxicity of the SNF for the two reactors systems due to their very high radioactivity, long half-lifes and high dose factors as compared to other actinides and fission products present in the SNF.

In this project, a numerical model is developed to assess the thermal-hydraulic system code RELAP5-3D correlations’ used for two-phase flow in MYRRHA secondary circuit conditions. The well-known drift-flux model is applicable to a wide range of thermal-hydraulic problems in nuclear reactors. The selected correlations for the model, based on the 4 equations drift-flux model, show better performance at lower pressures due to its approach to interfacial interaction and it is suitable for a wide range of thermal-hydraulic problems in nuclear reactors. All field and constitutive equations were properly selected and derived to get as acceptable with respect to the simplified approach. For the solution methodology, an iterative scheme was chosen with convergence criteria on exit pressure. The results show great compliance with experimental results and for some conditions are even better than RELAP5-3D results.

The nuclear industry must address the issues faced by the current new-built plants in the west- ern world if it is to remain cost-competitive with other forms of electricity generation. Serial, factory-based production of small modular reactors may address all of these issues. In this context enters SEALER, a lead-cooled fast reactor designed in the most compact configuration possible for commercial power production in off-grid areas. Lead Cold and KTH Royal Institute of Technology are developing a multi-point dynamics code (BELLA) intended for use in the safety-informed conceptual design of lead-cooled fast reactors. Its development is motivated by relatively restricted options to apply modifications to currently available system codes. This master’s thesis aims to improve the accuracy of BELLA by implementing a module for simulation of nuclear fuel thermo-mechanical performance, including the effects of thermal expansion of the fuel and cladding, fission gas release, and swelling of the fuel. Moreover, it constitutes a comprehensive review of the reference code. The thermal expansion of the fuel pellet and cladding are estimated from temperature vari- ations and their respective thermal expansion coefficients. An analytical expression for the thermal conductivity of irradiated UO2 fuel is formulated considering the effects of poros- ity, solid fission products (precipitated and dissolved), and radiation damage. Fission gas release into the fuel-cladding gap is determined by assuming a total release above a burnup- dependent temperature threshold (Vitanza threshold) and a release-to-birth ratio of 1% oth- erwise. Gaseous swelling of the fuel is calculated from the density of the oxide pellet, which is described by a linear function of burnup. Solid-to-solid heat conduction between the fuel and the cladding is implemented, thus providing the possibility to simulate gap closure. The performance of SEALER under unprotected transient overpower conditions is examined using BELLA with and without the fuel thermo-mechanics extension. The results indicate that the fuel centreline temperature is underestimated by the reference code. Nevertheless, the safety margin to fuel melting is still significant. Conversely, the fuel outer temperature is overestimated. There are two reasons for this, depending on the level of burnup. At the beginning of life, the thickness of the gap decreases due to the thermal expansion of the fuel, thereby enhancing heat transfer from the fuel to the cladding. On the other hand, as burnup increases beyond 3% gap closure occurs, and improved heat conduction results from direct solid-to-solid contact.

In the mid-1970s, American civil defense authorities became increasingly concerned with the potential threat of fission reactor accidents. Research from the Defense Civil Preparedness Agency began to be aimed not just at the ever-present threat of nuclear warfare but also peacetime emergencies as part of a "dual use" philosophy. The Brookhaven National Laboratory received funding to create a prototype radioiodine air sampling system, with multiple publications detailing the creation of the air sampler itself and a special CD V-700 survey meter that accompanied it. In late March 1979, the system found its first operational employment at the reactor accident at Three Mile Island in Pennsylvania. Despite successful use and further study, the BNL Air Sampler was not widely fielded by the federal government. However, lessons learned from its employment and development may be applicable to contemporary radiological emergency response.

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.

Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of a zero-power reactor for the process of developing a new, innovative rector concept, such as that required for the molten salt fast reactor based on iMAGINE technology, which operates directly on spent nuclear fuel, is described here. It is based on historical developments as well as the current demand for experimental results and key factors that are relevant to the success of the next step in the development process of all innovative reactor types. In the systematic modelling and simulation of a zero-power molten salt reactor, the radius and the feedback effects are studied for a eutectic based system, while a heavy metal rich chloride-based system are studied depending on the uranium enrichment accompanied with the effects on neutron flux spectrum and spatial distribution. These results are used to support the relevant decision for the narrowing down of the configurations supported by considerations on cost and proliferation for the follow up 3-D analysis. The results provide for the first time a systematic modelling and simulation approach for a new reactor physics experiment for an advanced technology. The expected core volumes for these configurations have been studied using multi-group and continuous energy Monte-Carlo simulations identifying the 35% enriched systems as the most attractive. This finally leads to the choice of heavy metal rich compositions with 35% enrichment as the reference system for future studies of the next steps in the zero power reactor investigation. An alternative could be the eutectic system in the case the increased core diameter is manageable. The inter-comparison of the different applied codes and approaches available in the SCALE package has delivered a very good agreement between the results, creating trust into the developed and used models and methods.

The roof slab of the nuclear reactor supports all the components and sub-systems. Roof slab needs to resist the seismic loads in accordance with load-carrying criteria. The static stress analysis of the reactor roof slab reveals that high-stress concentration was present in the pump penetration shell (PPS) which supports the primary sodium pump. This paper presents the assessment of collapse load and optimization of pump penetration shell, through the reliability approach, accounting for material nonlinearity, geometrical nonlinearity and randomness in loading. In addition to that, the load-carrying capacity of PPS was determined considering two different materials, viz., IS2062 and A48P2. The design of experiments (DoE) was formulated considering the flange angle and flange thickness as parameters. An empirical model for load function was formulated from the results of the collapse load obtained for various combinations of design parameters. The above function was used to perform the reliability-based geometry optimization of PPS of the roof slab.

Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive net-zero energy production. However, a zero-power reactor experiment will be an essential first step into the process of delivering this technology. The topic of the control and shut down for a zero power reactor is for the first time introduced through a literature review and the reduction of the control approaches to a limited number of basic functions with different variations. In the following, the requirements for the control and shutdown system for a reactor experiment are formulated, and based on these assessments, an approach for the shutdown – splitting the lower part of the core with reflector, and an approach for the control – a vertically movable radial reflector are proposed. Both systems will be usable for a zero-power system with a liquid as well as with as a solid core and even more importantly, both systems somehow work on the integral system level without disturbing the central part of the core which will be the essential area for the experimental measurements. Both approaches have been investigated as a singular system as well as their interactions with one another and the sensitivity of the control system. The study has demonstrated that both proposed systems are able to deliver the required characteristics with sufficient shutdown margin and a sufficiently wide control span. The interaction of the system has been shown to be manageable and the sensitivity is on a very good level. The multi-group Monte-Carlo approach has been cross evaluated by a continuous energy test leading to good results, but they also demonstrate that there is room for improvement.

Deuterium plasma operations using a large fusion test device have been carried out since 2017 at the National Institute for Fusion Science. A small amount of tritium was produced by the fusion reaction, d(d, p)t. Then, a part of the tritium was released into the environment. Thus, monitoring the level of tritium in the environment around the fusion test facility is important. This is done before starting the deuterium plasma experiment. The environmental tritium concentrations indicated that they are at background levels in Japan. After starting the deuterium plasma experiment, the environmental tritium around the fusion test facility was within the range of environmental variation. This suggests that there was no impact of tritium on the environment during the first deuterium plasma experimental campaign.

Stand-off, in-situ, laser induced breakdown spectroscopy (LIBS) offers a rapid, safe, and cost-effective method for discrimination of radioactive waste materials arising during the operation of nuclear plants and from decommissioning activities. Characterisation of waste materials is a critical activity in understanding the nature of the waste, ensuring hazardous material is managed safely and that waste can be segregated for reuse, recycle or sentenced for appropriate disposal. Characterisation of materials, often in hostile environments, requires the ability to remotely differentiate between materials in terms of their chemical composition and radioactivity. This proposition was tested using a case study on nuclear grade graphite. Graphite has been used extensively as a moderator material in many nuclear reactors. Internationally, over 250,000 tons of various nuclear-grade graphite, and graphite-bearing, materials exist. These are a major issue for nuclear decommissioning and radioactive waste management, due to the long half-lives of the associated 14C and 36Cl isotopes. LIBS offers a method for discrimination of nuclear grade graphites and other carbon and non-carbon-bearing wastes. This paper describes the development of a workflow method, including LIBS measurement analysis, for the discrimination of pre-irradiated nuclear ‘Pile Grade A’ (PGA) graphite moderator rod and domestic lumpwood charcoal, which act as surrogates for nuclear grade graphite and other carbon-bearing wastes. A new analysis workflow comprising the examination of spectral characteristics, multivariate analysis and molecular isotopic spectroscopy is proposed to enable rapid segregation of graphite from a heterogeneous waste stream. Enhanced characterisation techniques have the potential to significantly reduce the cost of decommissioning large parts of legacy nuclear generation plants.

Low-probability, high-impact events are difficult to manage. Firms may underinvest in risk assessments for low-probability, high-impact events because it is not easy to link the direct and indirect benefits of doing so. Scholarly research on the effectiveness of programs aimed at reducing such events faces the same challenge. In this article, we draw on comprehensive industry-wide data from the U.S. nuclear power industry to explore the impact of conducting probabilistic risk assessment (PRA) on preventing safety-related disruptions. We examine this using data from over 25,000 monthly event reports across 101 U.S. nuclear reactors from 1985 to 1998. Using Poisson fixed effects models with time trends, we find that the number of safety-related disruptions reduced between 8% and 27% per month in periods after operators submitted their PRA in response to the Nuclear Regulatory Commission's Generic Letter 88-20, which required all operators to conduct a PRA. One possible mechanism for this is that the adoption of PRA may have increased learning rates, lowering the rate of recurring events by 42%. We find that operators that completed their PRA before Generic Letter 88-20 continued to experience safety improvements during 1990-1995. This suggests that revisiting PRA or conducting it again can be beneficial. Our results suggest that even in a highly safety-conscious industry as nuclear utilities, a more formal approach to quantifying risk has its benefits.

The radiological characterization of the topsoil of the L-54M reactor surroundings carried out in this work aims at obtaining the reference blank point for the forthcoming decommissioning operations and ascertain if unexpected radionuclide release occurred during the operational life of the plant. Standardised methods have been employed in order to collect representative samples and reliable results. Suitable sample pre-treatment procedures were applied. Gamma and beta spectrometric analyses were carried out to measure the activity concentrations of 60Co 137Cs, 152Eu, 241Am and 90Sr. These have been considered as representative radionuclides that could have been originated from reactor operations and that could still be present at four decades post reactor shutdown.