Mixed radiation field effects from a nuclear reactor on poly(aryl ether ether ketone): A melt viscosity study

Evaluation and modelling of the lithium fluoride based thermoluminescent detector response at the CERN-EU high-energy reference field (CERF)

Indonesia has three nuclear research reactors, nemely: RSG-GAS in Serpong, TRIGA2000 reactor in Bandung, and Kartini reactor in Yogyakarta. Besides being useful, nuclear reactor has inharent risk due to its radiation hazard. The research reactor nuclear contingency plan was prepared for nuclear preparedness. The study aimed to analyze the implementation of the research reactors’ nuclear contingency plans. The study used an analytic descriptive qualitative methods. The results showed that the nuclear contingency plans of the three research reactors have different implementation range due to their different hazard risks. However, the nuclear emergency responses were the same, they were: sheltering, evacuation, provision of KI tablets, and prohibition of consumption of local food and contaminated river/ ground water. The research reactors’ nuclear contingency plans have not discussed criteria used for taking nuclear emergency responses. The results also showed that the nuclear contingency plan of RSG-GAS in Serpong and the Kartini reactor in Yogyakarta have been implemented partially. While for the TRIGA2000 reactor nuclear contingency plan in Bandung has not been implemented yet. A legal umbrella is needed to implement the research reactor nuclear contingency plan. Thus, coordination and active role from all stakeholders are needed to be enhanced in order to be able to resolve the existing obstacles and challenges.

Weight loss and burst testing investigations of sintered silicon carbide under oxidizing environments for next generation accident tolerant fuels for SMR applications

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the Triga IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors.

The work presented in this paper characterize the fiber Bragg grating (FBG) for temperature measurement for the safe operation in material testing reactors (MTR). Fibre Bragg Grating is the key technology used to covers many applications in nuclear reactors in order to sense different physical parameters such as temperature, strain, etc. This paper investigates the modelling of Bragg grating based on MATLAB, through solved coupled mod equations in order to achieve maximum reflectivity. In addition, modelling of fiber Bragg grating as a temperature sensor for nuclear applications using opti-grating software is presented. Furthermore, Apodization techniques for signal processing also used to provide accurate measurements which required in nuclear applications.

Present work discusses the issues of creating the instrumentation for testing the semiconductor magnetic field sensors during their irradiation with neutrons in nuclear reactors up to fluences similar to neutron fluences in steady-state sensor locations in ITER. The novelty of the work consists in Hall sensor parameters being investigated: first, directly during the irradiation (in real time), and, second, at high irradiation levels (fast neutron fluence > 1018n/cm2). Developed instrumentation has been successfully tested and applied in the research experiments on radiation stability of magnetic sensors in IBR-2 (JINR, Dubna) and VVR-M (PNPI, Saint-Petersburg) reactors.

Power stabilization is a critical issue in nuclear reactors. A conventional PD controller is currently used in Egypt's second testing research reactor (ETRR-2). This controller is designed neglecting the nonlinear nature of the plant. In this paper, two fuzzy controllers are used instead of the conventional PD. The first fuzzy controller has a static rule set, while the second one has an adaptive rule set. Simulation results show that the adaptive fuzzy controller gives the best integral square error (ISE) index, while the static fuzzy controller still provides a better ISE than the conventional PD controller.

In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing conditions, such as a control-rod movement, induced by an operator, or an accident scenario. The current methods for transient analysis apply deterministic solvers to calculate the neutronic response, but not only do these deterministic solvers always need to discretise the problem, they also often apply more fundamental approximations such as homogenization or diffusion. Therefore a stochastic method is needed, which only has a statistical uncertainty. A challenge for applying the Monte Carlo method on transient analysis, is the different types of particles which must be simulated in a single simulation. It is common for a neutronics calculation to simulate the prompt neutrons, however the simulation of precursors is new. The main challenge when simulating precursors lies in their long lifetime, which is seconds, whereas the prompt neutrons have a lifetime of microseconds. This has been solved by dividing the transient problem into time intervals and forcing the precursors to produce a delayed neutron is all time intervals. This will ensure prompt-neutron chains in all intervals. The prompt neutrons will form prompt-neutron chains, which poses another challenge for the Monte Carlo method. The chain length of these prompt neutrons varies a lot and the can also split into many branches, which averages out in a real power reactor, but is difficult to simulate on a computer. To improve the chain-length statistics, a new variance reduction technique is developed, which dictates that a single neutron will emerge from each interaction. This emerging neutron can be the result of a scattering event or a fission event. Finally, a dynamic simulation scheme, which can run in parallel, has been devised, simulating all time intervals consecutively. Also, a method for sampling the initial conditions is created and this scheme is implemented in a purpose-built Monte Carlo code and in a general-purpose Monte Carlo code. The two codes have been tested in several cases and they behave as expected, agreeing with deterministic method where expected, but deviating when the deterministic methods are no longer valid. In order to simulate realistic transients in a power reactor, feedback has to be taken into account. To achieve this, the dynamic Monte Carlo method has been coupled to a thermal-hydraulics code, using an explicit scheme. The results of this coupled simulation are compared to a state-of-the-art coupled diffusion calculation in a NURISP-benchmark calculation and the results agree well, except for a small deviation towards the end. The exact source of this deviation should be further investigated, but the difference is small, when realising that there is already a deviation in the steady-state results. The results demonstrate the feasibility of performing a fully dynamic transient analysis, using only Monte Carlo for the neutronics part of the calculation. Transients can now be calculated, with detailed modelling of any complex geometry and with continuous energy, which is especially useful for newly developed reactor types and one-of-a-kind research reactors.

Proton recoil proportional counters have been successfully used in many laboratories worldwide for more than 30 years in order to measure high-resolution neutron energy spectra. The method is well elaborated and understood, nevertheless high expertise is required for its proper application. Table 4.1 summarises typical basic data for proton recoil proportional counters and the requirements for their application. It is noteworthy that any limiting parameter can vary to a certain degree depending on the quality of the detectors used (design, gas purity, response functions, etc) and their response matrices, the data analysis and the unfolding procedures applied with a correct evaluation of the uncertainties involved. It is also important to lend a critical eye to details during measurements (e.g. environmental and electronics problems) as well as in subsequent analysis and unfolding (e.g. oscillations due to unfolding artefacts or inadequacies in detector response matrices). It is recommended that any spectrometry system (procedures for measurement and data evaluation) should be tested and validated in well-known neutron fields e.g. 252Cf standard fission or 241Am-Be. One should, however, expect that, due to different room scatter conditions, deviations from the ISO spectra may occur, especially for low neutron energies. In order to demonstrate the capability of the recoil proton counter technique, two examples of typical neutron spectra are shown in Figures 4.20 and 4.21, both measured in mixed neutron-gamma fields at nuclear research reactors.

Indian nuclear research reactors provide facilities to carry out irradiation experiments on the core structural materials to assess the radiation damage in these materials. The radiation damage in the material strongly depends on the temperature of irradiation. The temperature-controlled instrumented irradiation capsule (TCIIC) is a device which facilitates irradiation of specimens at a constant higher temperature in a reactor. This paper discusses design of PID controller to establish a constant temperature in the TCIIC system. The control parameters are estimated through placement of dominant pole-zero in the “S-plane” and a stable region of the PID controller parameters was defined. The time response result from the proposed method has been compared with other well-known PID controller design methods and also verified experimentally. The proposed method shows that the controller parameters can provide a stable and reliable performance of temperature control needed in the irradiation experiment.

For many years we have been investigating the chemical effects in gases of mixed pile radiation and fission fragment radiation produced in a nuclear reactor (1). Mixed pile radiation consists in y-radiation, fast neutrons, some-radiation, and, in gases containing nitrogen, proton recoil radiation produced by the N14 (n, p)C14 reaction. Fission fragment radiation is obtained by placing U235 glass fibers in the vessel or flow system containing the gas and subjecting the U235 to thermal neutrons. When 0.1 mg of U235 or more is present per cubic centimeter of gas, the fission fragment radiation produced is the predominant radiation acting on the gas. Our investigations have shown that the chemical effects produced by mixed pile radiation are similar to those caused by fission fragment radiation. For example, when nitrous oxide is irradiated with either fission fragments or pile radiation, the same relative amounts of N2, 02, and NO2 are formed. Also, when gases are irradiated with either a, 3, or 7 ionizing radiation, the M/N values obtained are the same (within the limit of error) and the primary processes are probably similar or even identical. On the other hand, if fission fragments are used (which initially have charges up to +20), then many of the primary formed ions will have charges exceeding unity. Under our experimental conditions these doubly (for example, helium3), triply, or even higher ionized particles will instantaneously react with surrounding uncharged particles, forming ions of lower charge. Therefore, it is possible for the distribution of ions, excited ions, and excited neutral species to be

After early interest in the possible uses of uranium in 1937, Argentina's scientists and politicians showed an inclination to support nuclear development that has kept quite steady compared with other areas. The Argentinean government prohibited the export of uranium in 1945, because of the emerging possibility of producing nuclear energy. The creation of the Atomic Energy Commission soon followed, and the first experimental reactor was set critical in 1958. Since then, nuclear development has allowed the successful operation of two nuclear power reactors, a quite integrated nuclear fuel cycle, and sustained activity in the development, production and use of radioisotopes. Nowadays an Argentinean company competes with success in the experimental nuclear reactor market. After a period in which the nuclear sector has been largely ignored in the official interest, Argentina's authorities have launched a comprehensive plan intended to rehabilitate all aspects of nuclear activity.

TRIGA Bandung is a research nuclear reactor owned by Indonesia, located in Bandung with a power of 2 MWth. Nuclear research reactor TRIGA Bandung is used as a center for applied research and development in the field of application of the nuclear technologies. TRIGA Bandung is currently still using a cylindrical fuel element, this raises a new problem - the limited number of existing fuel element. The purpose of this research is the development of the preliminary core design of a nuclear research reactor TRIGA Bandung using fuel element plate MTR. The research method is modeling and simulation the preliminary design core of nuclear research reactor TRIGA Bandung using comparative method of porous media and non-porous media with CFD code. This research shows the velocity flow and temperature distribution and the influence of pressure from the comparison method of k-ε standard model and porous media model at the preliminary design in the core area of TRIGA Bandung research reactor with fuel element plate MTR.

Next generation high energy physics experiments will be more granular than those currently in operation, this means more demanding electronics to power the detectors and to process all collected data. Space constraints, cabling, cooling and, last but not least, efficiency are all parameters that need to be optimized during experiment design to have the best performance for data taking.\nThe CAEN R&D to develop a new generation of power supplies for hostile environment is presented and some results from test campaigns are discussed. This R&D named EASY6000, after its predecessors EASY3000 and EASY4000, was launched in 2020 designing and testing electronics capable to survive magnetic fields and mixed radiation fields (composed by gamma, neutral and charge hadrons) of the High Luminosity LHC (HL-LHC) experimental caverns at CERN.\nTo cover all needs, CAEN started an irradiation campaign in various steps and at various irradiation facilities, with the aim to investigate COTS (Commercial Off-The-Shelf) electronics behavior using one radiation at a time: neutron, gamma, protons, and then validate the final board design with a mixed field. Plus, we performed efficiency tests in magnetic fields up to 1 T using various orientations to exploit different symmetries in the boards design.\nDuring this talk we will focus on test campaigns performed in the last year. Some undertaken within the RADNEXT EU project and in collaboration with INFN and CERN, that include tests with proton, neutron, and gamma sources, of various components: ADCs, DACs, RAMs, FPGAs, 𝜇Controllers, Power Transistors and FETs, temperature and humidity sensors, etc. All the necessary pieces to design and build circuits and boards capable to survive in the experiments; these components alone cannot ensure the reliability to run an experiment in such conditions, thus also circuits and control loops must be tested. The results of the test campaigns will be discussed together with some mitigation techniques used to achieve the wanted reliability.\nFirst developments of power supply circuits and devices based on these blocks will be also presented, as well as the performances achieved so far in terms of reliability, power density, energy efficiency, noise figure, etc.