Graphite Reflector Research Articles

Core neutronic design of small modular molten salt reactor for submarine propulsion

Molten salt reactor (MSR) shows great promise as a Generation IV nuclear reactor concept with high thermal-electric conversion efficiency, inherent safety features, and online reprocessing capability. During operation, the molten salt reactor MSR does not require a long refueling shutdown owing to online fuel reprocessing, making it suitable for watercraft propulsion. This paper presents the neutronic design of the 250 MWt iso-breeder MSR concept for submarine propulsion. The reactor is fueled with LiF-BeF2-ThF4-233UF4 salt with a virtual one-and-a-half fluid configuration. The MSR core is moderated with beryllium oxide and equipped with a high-density graphite reflector to improve neutronic. All calculations on the reactor core design were performed using MCNP6.2 code with ENDF/B-VII.0 neutron cross-section library. As a part of the neutronic analysis, the calculated parameters were the effective multiplication factor (), temperature coefficient of reactivity (TCR), and breeding ratio (BR). The value of ± 1.005 was obtained using Th : U ratio of 98.42% : 1.58%. The TCR value was obtained at -2.56 pcm/K, while the BR value was 1.057. These initial values indicate that the marine propulsion MSR can achieve iso-breeding with inherent safety characteristics. By using MSR design, it is hoped that the submarines will be able to reach significantly longer operational range than its Diesel-powered counterpart..

Broadband metal-free graphite absorber for the shielding of far-infrared radiations

An ultrathin metal-free broadband absorber covering the far-infrared region is designed and implemented. A graphite resonator and reflector are used to provide the dual-resonance with merged spectrum. The graphite resonator is perturbed to improve the impedance bandwidth of the absorber at both the resonances. This helps in obtaining the wide and merged resonance to provide the wide absorption spectrum by maintaining the level of absorption more than 90%. The electromagnetic resonances taking place in the absorber structure are the reason behind the operation of the proposed absorber hence the circuit model analysis is carried out to validate its operation. Moreover, the shielding effectiveness of the absorber can vary from 40 to more than 150 dB proving its capability to be utilized in EM shielding, packaging and thermal storage applications in lower THz frequency band with ultrathin structure.

Neutronic characteristic analysis of the GAMA microreactor

The GAMA Microreactor is a low-power nuclear power plant dedicated for remote area with a power output of 300 kWe. The primary characteristics of the GAMA Microreactor are its system compactness, design simplicity, and safe operation without moving parts. The fuel is uranium hydride powder, which works simultaneously as a moderator, contained within a stainless-steel vessel and surrounded by a graphite reflector. Liquid sodium-filled heat pipes are embedded in the fuel powder to extract fission-generated heat via passive coolant flow due to capillary action. As a part of the design process, neutronic parameters of the GAMA Microreactor have been determined using SCALE6.2 and OpenMC codes. Among the analyzed parameters are the excess reactivity as a function of burn-up, shutdown margin, control rod worth, temperature coefficient of reactivity (TCR), fuel density coefficient of reactivity (DCR), and reactivity coefficient due to hydrogen dissociation. From the calculation results, the GAMA Microreactor has a maximum excess reactivity of 0.049 at the operational temperature (700 °C) and 0.0788 at ambient temperature. The TCR value is negative (−2.267 pcm/°C) at the beginning of the cycle (BOL) and −4.9618 pcm/°C at the end of the cycle (EOC). Hydrogen dissociation imposes negative reactivity on the reactor both at the BOL (−0.2631 Δk/k-%dissociation) and EOC (−0.3524 Δk/k-%dissociation). Therefore, the GAMA Microreactor maintains its inherent safety characteristics over a considerably long operation time.

Preliminary analysis of the core design for Indonesia Micro Reactor (IMR-13)

The development of microreactor technology can meet the energy needs of specific regions of Indonesia’s underdeveloped, foremost, and outermost areas. One of the latest microreactor designs is the Indonesia Micro Reactor (IMR-13), which has seed and blanket fuel and 5MWt power generation. The IMR-13 reactor has characteristic design features for operation for as many as ten years without refuelling. The reactor core comprises integral fuel elements consisting of a heat pipe, fuel matrix, fuel cladding, and graphite moderator. There are two fuel element types: seed fuel elements with 19.75 % low-enriched uranium oxide and blanket fuel elements with thorium oxide – graphite reflectors on the bottom, sides, and top surround the reactor core. Side reflectors have moving parts that function to regulate reactor power. There are no active components in this reactor. Heat transfer occurs by natural circulation in the heat pipe to the molten salt pool and then from the molten salt pool to the steam generator. This research focuses on the neutronic aspect of the reactor, particularly in calculating the seed and blanket ratio to allow the reactor to operate for ten years with minimal power peaking factor. The selected variant model has 979 seeds and 294 blankets. The variation model has a value of excess reactivity at the beginning of the life of 937 pcm. Each value of the shutdown margin for system first and system second sequentially is 3020 pcm and 2994 pcm. The temperature reactivity coefficient is negative, implying inherent safety. At the end of life, the burnup value was achieved at 4.97 GWd/MTU with an excess reactivity of 271 pcm. A parameter analysis of the burnup characteristics will follow in the selected variant model. The parameters examined were the core power distribution, delayed neutrons, the fast fission factor, conversion ratio, and decay heat.

The impact of fast spectrum graphite moderated systems on thermal spectrum graphite moderated reactors through the resonance physics of n+12C

Understanding the nuclear data used in predictive modeling and simulation of advanced reactors is vital for their development, licensing, and deployment. In this paper, we use the Kairos gFHR and the X-Energy Xe-100 as demonstration examples to test the scientific hypothesis that the resonance physics of n+12C allows for integral experiments with fast-neutron-spectra to be useful in helping to reduce nuclear data uncertainty on k-effective of new reactor systems with thermal-neutron spectra. This work hypothesizes that legacy criticality experiments hold value for these new reactors. HMF071 (highly enriched, metal-fuel, fast-spectrum critical experiment) (Parkey et al., 2014; Nuclear Science Committee of The Nuclear Energy Agency, 2016) is a series of legacy criticality benchmarks with graphite reflectors which were selected to test this scientific hypothesis by leveraging the large correlation visually identified in the covariance matrix of n+12C. This hypothesis will be tested through calculating the similarity and uncertainty-reduction provided from evaluating the effect of a fast-neutron-spectrum system, HFM071, on a thermal-neutron spectrum system. The sensitivity with respect to elastic scattering of carbon had greater than 70% similarity for one of the cases of HMF071 compared against each of the advanced reactor designs considered. This result quantifies the relevance that the fast spectrum experiments can have for thermal spectrum applications due to the resonance physics of 12C. With this connection quantified, the uncertainty reduction from the fast-spectrum experiments on these thermal-spectrum reactors was calculated using the SCALE module TSURFER (Rearden and Jessee, 2018; Wieselquist and Lefebvre, 2023). The HMF071 series of critical experiments were able to lower the nuclear data uncertainty associated with carbon on the Xe-100 and gFHR only by about 6 pcm using SCALE 6.2.4 and the ENDF/B-VII.1 nuclear data library. In addition, the HMF071 series of critical experiments were able to lower the nuclear data uncertainty associated with carbon on the Xe-100 and gFHR by about 11 pcm and 16 pcm respectively using SCALE 6.3.1 and the ENDF/B-VIII.0 nuclear data library. When all-materials were used in the TSURFER evaluation, the uncertainty reduction on the gFHR was about 39 pcm using SCALE 6.3.1 and the ENDF/B-VIII.0 nuclear data library. This is very small; however, due to the fact that carbon is such a well-studied element, any further reduction of uncertainty for thermal-spectrum systems is difficult (but useful) to obtain. These null results disproved our hypothesis and do not allow us to demonstrate this specific use case for evaluating legacy fast-neutron-spectrum experiments, which are otherwise often neglected, for modeling modern thermal-neutron spectrum graphite moderated/reflected reactors.

Thermally switchable metal-free THz electromagnetic shield using phase change material

A thermally switchable ultrathin terahertz (THz) metal-free electromagnetic shield can be developed with the help of phase change material, vanadium oxide (VO2). The thermal alterations in the phase change material can modulate its conductivity hence the switchable absorptivity can be achieved. An ultrathin, (height ≈λ/7, λ is free space wavelength) ultra-broadband, metal-free absorber structure containing a VO2 resonator, dielectric spacer and graphite reflector operating within 3.88−9.82 THz band is designed and numerically studied. Multiple reflections and destructive interference occurring in the dielectric delivers the absorber operation which is validated through the theory of multiple reflections. Absorber provides the stable response over the incident angle of more than 45°. Moreover, the shielding effectiveness of the developed thin periodic film remains more than 30 dB in the operating frequency range.

The computer model of a neutron fluxes forming system on a linear electron accelerator

A computer model of the system for generating neutron fluxes at the output of a linear electron accelerator has been developed in the Geant4 and PhysList QGSP BIC HP programming environment. With the help of the model, a number of virtual experiments on 107 primary neutrons were carried out. The dependence of the ratio of the number of neutrons, incident on the detector with a reflector to the number of neutrons without a reflector on the radius of curvature of the hemispherical dome of the reflector is determined, which makes it possible to establish the optimal dimensions of the working area. The energy spectra of neutrons incident on the detector are determined. It has been established that when using a graphite reflector 30 × 30 × 30 cm in size with a radius of curvature of the hemispherical dome of 5 cm, the number of neutrons at the location of the detector increases by 16.9 %, and the neutron background in the accelerator bunker decreases by 2.5 times, which is in good agreement with the real experiment made. Such a decrease in the radiation background, according to sanitary standards, will make it possible to increase the accelerator current by 2.5 times and increase the neutron flux.

Special Experimental Environment for Generation IV Reactors With Graphite Reflector

Abstract Nowadays, there is an increasing demand for new small modular reactors (SMR) reactors with a wide range of applications, often classified as a new Generation-IV reactors. Unfortunately, there is no commercially operating nuclear reactor meeting the characteristics of Generation-IV reactors in its technical design and features. However, Generation-IV nuclear reactors are intensively developed worldwide, including the Czech Republic. To support research activities linked with the development of these reactors, an appropriate experimental environment and resources simulating conditions expected in Generation-IV reactors with graphite are needed This work builds on previous research and presents a special arrangement of the LR-0 reactor core with a double-row graphite reflector, on which the influence of graphite on the neutron field shaping and power distribution was experimentally and computationally verified. The main reason for this article is to investigate the effect of the graphite reflector, which will serve as a support for the development of neutron instrumentation for Generation-IV reactors and other support for Generation-IV research activities. The results confirm the conclusions of previous research and also provide a better knowledge of the double-row graphite reflector neutron field. The presented and tested LR-0 core arrangement is included in a set of experiments supporting the research of reactor cores with graphite reflectors.

DEVELOPMENT AND TESTING OF A COMPUTER MODEL OF A SYSTEM FOR FORMING NEUTRON FLUXES ON A LINEAR ELECTRON ACCELERATOR

A computer model of the system for forming neutron fluxes at the output of a linear electron accelerator has been developed in the Geant4 programming environment. The system for forming neutron fluxes consists of a monoenergetic neutron source, a graphite reflector, a polyethylene moderator, lead protection and detector. With the help of the model, a number of virtual experiments on 107 primary neutrons were carried out for the energy range of the neutron source from 0.1 to 1.1 MeV. The dependence of the ratio of the number of neutrons falling on the detector with a reflector to the number of neutrons without a reflector is determined, on the energy of the neutron source. Energy spectra of neutrons falling on the detector are determined. It was established that the graphite reflector not only increases the number of particles falling on the detector, but also moderated the energy spectrum of neutrons. The neutron background in the accelerator bunker when using a reflector decreases from 2 to 7 times, depending on the energy of the neutron source.

The Computer Model of a Thermal Delayed Neutron Fluxes Forming System for Nuclear Medicine

In the work the computer model of a cell of a system for generating fluxes of therapeutic beams of delayed neutrons, based on the use of delayed fission neutrons, was developed in the Geant 4 environment. The principle of such a neutron source is that when a powerful electron beam interacts with a combined tungsten target and a target containing fissile material, a fission reaction occurs; as a result of which neutrons are emitted. If we move a target activated in this way several tens of meters into a neutron flux generation system consisting of a heater, protection, collimator and reflector, we will obtain a compact neutron source for nuclear medicine. A significant advantage of such a neutron source is the absence of gamma background from the electron accelerator and the combined target, and a bulky protection system is not required. In the Geant 4 environment, the geometry of this cell was developed and a series of experiments were carried out with 107 neutrons. The QGSP BIC HP physical sheet was used. A study of neutron energy spectra showed that more than half of the neutrons whose fluxes are formed using such a cell of the formation system have an energy <100 keV, which is suitable for use for therapeutic purposes. Analysis of the data obtained in a computer experiment made it possible to develop a modified cell of the system for generating streams of therapeutic beams of delayed neutrons, which differs from the basic one by the presence of a solid polyethylene moderator with holes for activated targets and a graphite reflector. Analysis of the data obtained showed that in this case the number of thermal neutrons hitting the detector increases 10 times compared to the base cell, and the energy of 80% of the particles does not exceed 5 keV, which is much better suited for therapeutic purposes.

A Whole-Core Transient Thermal Hydraulic Model for Fluoride Salt-Cooled Reactors

Abstract A thermal hydraulic model is developed for a solid pin-fueled fluoride-salt-cooled small modular advanced high temperature reactor (SmAHTR). This preconceptual SmAHTR was developed by the Oak Ridge National Laboratory (ORNL). For the fuel assembly configuration investigated in this study, the fuel and non-fuel pins are arranged in a hexagonal layout. The molten FLiBe salt coolant flows parallel to the bank of pins. A finite volume model is developed and used to compute temperatures in the solid regions (fuel and non-fuel pins, and the graphite reflectors) in the core. The temperature, flow, and pressure profiles for the coolant flowing through the pin bundles in the core are calculated using the conventional subchannel methodology. Pertinent closure relations are used to compute the hydraulic losses, momentum, and energy exchange between adjacent subchannels, and heat transfer between the solid and fluid regions. The resulting model can perform both steady-state and transient computations across the entire core. This fully implicit model also includes an adaptive time-stepping algorithm for automatic time-step adjustment. A preliminary code-to-code comparison demonstrates good agreement between the present subchannel-based model and a computational fluid dynamics (CFD)-based model for a transient case in which the core inlet flowrate varies with time. Following the code-to-code comparison, the thermal hydraulic model is used to analyze the protected loss of heat sink (P-LOHS) accident scenario. Key results such as the transient evolution of peak fuel, non-fuel, and coolant temperatures, and 3-D core temperature distribution are presented and discussed.

Recommendations for seismic analysis of a molten salt reactor

A Fluoride-salt cooled High-temperature Reactor (FHR) that uses circulating solid TRISO pebbles, as a fuel and positively buoyant graphite reflector blocks as a moderator is at an advanced stage of development. Seismic base isolation supports a pathway for cost competitive and rapid deployment of this molten salt reactor (MSR) in regions of different seismic hazard. This paper presents recommendations for analysis and modelling of different components of a base-isolated MSR, based on results of earthquake-simulator tests on a scale model of the reactor. Recommendations are provided for evaluating the sloshing behavior of the coolant in an annular region inside the reactor and for estimating hydrodynamic loads for use in analysis of structural and mechanical components. Hydrodynamic loading on the reflector blocks is discussed and a practical approach to estimate forces in the connections between them is presented. Approaches for modelling and analysis of two 2D horizontal base-isolation systems, utilizing spherical sliding bearings, are validated using results from experiments. The recommendations apply to reactor vessels with similar fluid-structure systems.

Neutronic Assessments towards a Novel First Wall Design for a Stellarator Fusion Reactor with Dual Coolant Lithium Lead Breeding Blanket

The Stellarator Power Plant Studies Prospective R&D Work Package in the Eurofusion Programme was settled to bring the stellarator engineering to maturity, so that stellarators and particularly the HELIAS (HELical-axis Advanced Stellarator) configuration could be a possible alternative to tokamaks. However, its complex geometry makes designing a Breeding Blanket (BB) that fully satisfies the requirements for such a HELIAS configuration, which is a difficult task. Taking advantage of the acquired experience in BB design for DEMO tokamak, CIEMAT is leading the development of a Dual Coolant Lithium Lead (DCLL) BB for a HELIAS configuration. To answer the specific HELIAS challenges, new and advanced solutions have been proposed, such as the use of fully detached First Wall (FW) based on liquid metal Capillary Porous Systems (CPS). The proposed solutions have been studied in a simplified 1D model that can help to estimate the relative variations in Tritium Breeding Ratio (TBR) and displacement per atom (dpa) to verify their effectiveness in simplifying the BB integration and improving the machine availability while keeping the main BB nuclear functions (i.e., tritium breeding, heat extraction and shielding). This preliminary study demonstrates that the use of FW CPS would drastically reduce the radiation damage received by the blanket by 29% in some of the selected configurations along with a small decrease of 4.9% in TBR. This could even be improved to just a 3.8% TBR reduction by using a graphite reflector. Such an impact on the TBR is considered affordable, and the results presented, although preliminary in essence, have shown the existence of margins for further development of the FW CPS concept for HELIAS, as they have been not found, at least to date, to be significant showstoppers for the use of this technological solution.

Design of a prototypical natural circulation water-based reactor cavity cooling system (RCCS) for a pebble-bed generic FHR

A prototypical natural circulation water-based reactor cavity cooling system (RCCS) for the UC Berkeley Mark-1 pebble-bed gFHR is designed based on one-dimensional thermal hydraulics modeling and optimization implemented in an in-house MATLAB code. The model employs a lumped core consisting of fuel pebble and graphite pebble regions, while the graphite reflector, vessel, and RCCS are represented as separate but energy-coupled regions. The model is derived based on steady state energy balance equations accounting for conductive heat transfer from the lumped core to the vessel and subsequent radiative heat transfer to the RCCS and convective heat transfer in the water. Mass flow rate is calculated based on momentum and thermal energy balance in the RCCS. Starting with ANL’s water NSTF as baseline, effects of RCCS design parameters such as size of the plate between pipes, pipe diameter, source-to-sink distance, surface emissivity, and pitch from the core are examined in an integral effects framework which accounts for subsequent variation of the mass flow rate, number of pipes, convection coefficient, mean region and surface temperatures, radiative view factors, and water outlet temperature. The importance of including a prototypical reactor design in RCCS design calculations is emphasized as the calculations show that temperature drops from the core to the RCCS fluid to transfer the heat tend to be the limitation not the heat removal capacity of the fluid. A closed-loop design is obtained with capability to safely remove up to 0.72% of the nominal reactor power at maximum estimated peak conditions and 0.40% at shutdown. The design is based on physics calculations and does not account for economic optimization.

Characterizing the seismic response of a molten salt nuclear reactor

AbstractOne Generation IV nuclear reactor, which uses a fluoride salt as a coolant, graphite reflector blocks as a moderator, and circulating buoyant TRISO pebbles as fuel is at an advanced stage of development. To characterize the seismic behavior of components of this reactor, validate numerical models for analysis, and develop recommendations for design, a set of earthquake‐simulator experiments on a scaled model of the reactor vessel and its internals was executed on a six‐degree‐of‐freedom earthquake simulator. The model was seismically isolated at its base using two types of spherical sliding bearings. The scaled model involved representations of the prototype reactor vessel, core barrel, reflector blocks, coolant, and spherical fuel pebbles. The material and geometric properties of different test components were selected based on a dynamic similitude scaling analysis and an approximate length scale of 0.4. Four sets of three‐component earthquake motions were used as inputs for testing. Instrumentation on the test specimen recorded the dynamic responses of the outer vessel, core barrel, and reflector‐block assembly, the hydrodynamic responses (sloshing and hydrodynamic pressure) of the liquid coolant, pebble consolidation under earthquake shaking, and the behavior of the isolation systems. This paper describes the design of the experiments and presents key results from the tests. The dynamic responses of the outer vessel, core barrel, and the reflector blocks revealed that the components responded as a unit for the intense shaking used in the experiments. The sloshing response of the fluid in a thin annulus near the perimeter of the vessel was heavily damped. The change in the packing fraction of the pebble bed under repeated, intense 3D earthquake shaking was less than 3%. Seismically isolating the vessel substantially reduced demands on its internal components.

Preliminary Neutronics Design and Analysis of the Fast Modular Reactor

General Atomics is developing a new 100-MW(thermal) fast modular reactor (FMR) that provides safe, carbon-free electricity and is capable of incremental capacity additions. The modular design allows it to be factory built and assembled onsite to keep the capital cost low, while the use of dry cooling facilitates siting to complement renewables in nearly any location. The FMR uses high-assay low-enriched uranium-dioxide fuel encapsulated by recognized irradiation-resistant silicon carbide composite (SiGA®) cladding that is derisked in the current accident-tolerant fuel program. The FMR fuel assembly is a hexagonal fuel bundle of 120 fuel rods. The total length of the fuel assembly is less than 4 m, with an active fuel length of 1.8 m. The fuel assemblies are configured in an annular core that is located and supported by the reactor internals. The coolant material is helium at a normal operating pressure of 7 MPa. The core is surrounded by zirconium silicide (Zr3Si2) and graphite reflector blocks. The fuel, coolant, internals, and reflectors are contained within a reactor pressure vessel. The preliminary nuclear design and analysis established the arrangement of the active core and reflector blocks. The nuclear design analyses of the FMR defined the design parameters, such as fuel enrichments, excess reactivity, fueling scheme, fuel cycle, power distribution, and control rod worth. The preliminary conceptual design determined the three-batch fueling scheme with the allowable total power peaking factor of 1.5. The average discharge burnup is 100 GW days per ton of uranium.

Nano Yttrium-90 and Rhenium-188 production through medium medical cyclotron and research reactor for therapeutic usages: A Simulation study

The main goal of the coordinated project development of therapeutic radiopharmaceuticals of Y-90 and Re-188 is to exploit advancements in radionuclide production technology. Here, direct and indirect production methods with medium reactor and cyclotron are compared to evaluate derived neutron flux and production yield. First, nano-sized 186W and 89Y specimens are suspended in water in a quartz vial by FLUKA simulation. Then, the solution is irradiated for 4 days under 9E+14 n/cm2/s neutron flux of reactor. Also, a neutron activator including three layers—lead moderator, graphite reflector, and polyethylene absorbent— is simulated and tungsten target is irradiated by 60 MeV protons of cyclotron to generate induced neutrons for 188W and 90Sr production via neutron capture. As the neutron energy reduced, the flux gradually increased towards epithermal range to satisfy (n/2n,γ) reactions. The obtained specific activities at saturation were higher than the reported experimental values because the accumulated epithermal flux and nano-sized specimens influence the outcomes. The beta emitters, which are widely utilized in brachytherapy, appeal an alternative route to locally achieve a rational yield. Therefore, the proposed method via neutron activator may ascertain these broad requirements.

Зарубежный опыт и подходы к извлечению графита из уран-графитовых реакторов

The article explores the best practices for graphite removal from the stacks of uranium-graphite reactors (GLEEP, WAGR, BGRR, HTGR) and reactors with graphite reflectors (DIORIT, DR 1) operated abroad. Based on several case studies, it considers different approaches, in particular: graphite crushing; underwater stack demolition by reactor shaft flooding and upper biological shield dismantling; dismantlement of the upper biological shield and layer-bylayer graphite block removal. It discusses current trends, existing challenges and long-term plans for the practical implementation of advanced approaches under international projects.

Preliminary study on HTR-10 operating at higher outlet temperature

High Temperature Gas-cooled Reactor (HTGR), which has well-known safety features and high temperature heat supply capability, is expected to be widely used for heat supply and technology heat utilization including the hydrogen production, and so contributing to the reduction of carbon dioxide emissions in various sectors. The 10 MW High Temperature gas-cooled test Reactor (HTR-10) had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. The first criticality of HTR-10 at air condition was realized on December 1, 2000, and the full power operation for 72 h on January 29, 2003.Supported by Chinese National S&T Major Project, HTGR for hydrogen production are now being studied. The physical and thermal hydraulic design to raise the outlet helium temperature of the HTR-10 reactor core from 700 °C to 850–1000 °C is carried out.In this paper, the preliminary thermal hydraulic design of the HTR-10 with the outlet helium temperature of 950 °C (HTR-10H) is introduced. The power density distribution, the fuel temperature distribution and the reactor pressure vessel (RPV) temperature are studied to identify what need to be focused on next. Besides, the DLOFC accident has been studied to evaluate the safety feature of the HTR-10 operating under higher core temperature and outlet temperature.Due to the lower power and power density, as well as the large heat capacity of the graphite reflector and the carbon brick, the maximal fuel temperature during the DLOFC would not exceed the limitation of 1620 °C (250th meeting of the reactor safety commission, 1990). Nevertheless, the higher fuel temperature during the normal operation should be of concern.

Optimization study of chlorine detection sensitivity in concrete based on prompt gamma analysis using 252Cf neutron source

Chloride attack is a serious problem that decreases the durability of concrete structures of bridges and highways. A compact neutron salt meter with a252Cf neutron source and germanium (Ge) gamma-ray detector based on prompt gamma-ray neutron activation analysis (PGNAA) has been proposed to determine the chlorine concentration in concrete structures. The Optimization of the dimensions of its components, such as polyethylene (PE) moderator, graphite reflector, and lead shield, as well as the positions of the 252Cf source and the Ge detector has been performed to make it highly sensitive for the detection of gamma-rays of chlorine, in addition to lightweight and small volume for in situ use. The results demonstrated that gain factors of 2.5 and 2.2 were obtained for gamma-ray intensity of chlorine and chlorine-to-hydrogen ratio (CHR), respectively, whereas the weight and volume became 19.1% and 23.4%, respectively, compared with the reference setup. The effectiveness of optimization was confirmed by preliminary experiments.