Next Generation Nuclear Research Articles

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

Pure metals for creation the nuclear structural materials

The paper presents the main results of research on the development of structural materials for equipment elements of existing and prospective nuclear power plants. The further development of nuclear energy largely depends on the development of advanced structural materials for new generation reactors and the improvement of materials for operating nuclear power plants. The high requirements for structural materials regarding the content of impurity elements can be met only when high-purity metals are used as initial components and new technologies for their production are used. The prospects of using new methods and approaches for evaluating and improving the physico-mechanical and physico-chemical properties (microhardness, corrosion resistance, radiation resistance, etc.) of structural materials for nuclear technologies were considered.

Development of a liquid metal subchannel code applied to ocean conditions and study on the effect of core geometry parameters and ocean motions on flow and heat transfer characteristics

The Generation IV International Forum (GIF) proposes a type of liquid metal reactor, due to its thermal properties and efficient heat transfer in a relatively small system. This special heat transfer characteristic allows the liquid metal reactor to be modular in design and flexible in installation on a nuclear-powered ship. Accurate prediction of the coolant temperature and flow distribution between subchannels is important to ensure nuclear safety in the complex ocean environment. In this study, an ocean version of subchannel code is developed by adding the motion terms in the momentum equations for the liquid metal reactor core based on the previously land-based subchannel code. Since the research on the characteristics of flow resistance and heat transfer mechanism under ocean conditions is not mature, the original empirical model is still used in the ocean version code. The constitutive models of the liquid metal are analyzed and incorporated into the ocean code. The code has been verified and validated by comparing with Computational Fluid Dynamics (CFD) simulation results, and experimental data. The effect of geometric parameters such as rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) is studied in the liquid metal lead 7 rod bundles. In addition, the effect of heaving start-up and shutdown motions between liquid metal lead and water is discussed in the 5 × 5 bare rod bundles. In general, the modified subchannel analysis code can well complete the calculation of liquid metal reactor under ocean conditions, as well as provide support for the design and safety analysis of a liquid metal cooled fast reactor.

Data and modeling sensitivity analysis for molten salt fast reactor benchmark – Static calculations

The Molten Salt Reactor (MSR) idea is increasingly being recognized in the nuclear field due to its potential safety, sustainability, and economic efficiency advantages. The Molten Salt Fast Reactor (MSFR) benchmark, introduced in 2019, highlighted variations in results tied to different neutron cross-section libraries. This study investigates the impact of utilizing the ENDF/B-VIII.0 and JEFF-3.3 cross-section libraries for MSFR benchmark assessment compared to the ENDF/B-VII.1 database. Monte Carlo based open source code OpenMC is used for the analyses. Rigorous sensitivity analyses assess the influence of individual components, including the cross-section database, resonance elastic scattering, and Thermal Scattering Law (TSL). Beyond the criticality assessments, parameters such as delayed neutron fraction, temperature coefficient of reactivity, and neutron spectrum are compared for different cross-section libraries. Our analyses reveal that incorporating new evaluations for 233U (n,γ) and fission cross-sections in ENDF/B-VIII.0 significantly alters criticality results, i.e., more than 1700 pcm difference is seen between libraries. Similarly, critical concentration using ENDF/B-VII.1 and JEFF-3.3 is over-predicted by approximately 3%. The variations in Thermal Scattering Law (TSL) files do not yield substantial differences in outcomes due to the fast spectrum of the reactor. In some cases, the treatment of resonance elastic scattering leads to reactivity differences greater than 50 pcm. The benchmark compares 233U-started and Minor Actinide (MA)-started core. From the reactor physics point of view, the MA-started core leads to a 29% higher (n, γ) reaction rate than the 233U-started core. A 3–4% smaller value of thermal reactivity coefficient is obtained using the ENDF/B-VIII.0 library compared to the ENDF/B-VII.1 value. Using the ENDF/B-VIII.0 for the MSFR benchmark signifies using newer and better data for the GEN-IV reactors neutron physics calculations.

Thermo-neutronic integrated coupling effects on nuclear reactor core calculations

Small Modular Reactors (SMRs) have gained significant attention as the next generation of nuclear reactors due to their unique features, such as improved economic viability and safety. This study focuses on the SMART reactor, the first certified design of an SMR, and investigates various Thermo-Neutronics aspects using an integrated coupling scheme of nuclear codes. The Neutronics cell and core calculations are performed using the DRAGON/DONJON codes, while the COBRA code is employed for thermal-hydraulic calculations. The analysis considers parameters including axial and radial power peaking factors (PPFs), critical heat flux (CHF), minimum departure from nucleate boiling ratio (MDNBR), axial average coolant temperature in hot channel and core, and maximum fuel temperature. These parameters are evaluated using the developed coupling system. Code-to-code verification is conducted to ensure the accuracy of the results, demonstrating good agreement. The coupling approach reveals significant effects compared to stand-alone code modeling, emphasizing the importance of considering the coupling between Neutronics and thermal-hydraulics, especially during the cycle length. This study contributes to the understanding of SMRs, particularly the SMART reactor, by providing insights into key parameters and their interdependencies. The integrated coupling scheme enhances the accuracy of the analysis and highlights the significance of considering coupling effects in the modeling process during the steady state and the cycle length.

Structural Changes in High-Entropy Alloys CoCrFeNi and CoCrFeMnNi, Irradiated by He Ions at a Temperature of 700 °C.

High-entropy alloys (HEA) are promising structural materials that will successfully resist high-temperature irradiation with helium ions and radiation-induced swelling in new generations of nuclear reactors. In this paper, changes in the elemental and phase composition, surface morphology, and structure of CoCrFeNi and CoCrFeMnNi HEAs irradiated with He2+ ions at a temperature of 700 °C were studied. Structural studies were mainly conducted using the X-ray diffraction method. The formation of a porous surface structure with many microchannels (open blisters) was observed. The average diameter of the blisters in CoCrFeMnNi is around 1.3 times smaller than in CoCrFeNi. It was shown that HEAs' elemental and phase compositions are stable under high-temperature irradiation. It was revealed that, in the region of the peak of implanted helium, high-temperature irradiation leads to the growth of tensile macrostresses in CoCrFeNi by 3.6 times and the formation of compressive macrostresses (-143 MPa) in CoCrFeMnNi; microstresses in the HEAs increase by 2.4 times; and the dislocation density value increases by 4.3 and 7.5 times for CoCrFeNi and CoCrFeMnNi, respectively. The formation of compressive macrostresses and a higher value of dislocation density indicate that the CoCrFeMnNi HEA tends to have greater radiation resistance compared to CoCrFeNi.

Study of radiation-induced structural changes in the near-surface layer of ZrO2 ceramics caused by He2+ irradiation

Abstract The aim of this study is to determine changes in the near-surface layer of ZrO2 ceramics caused by irradiation with low-energy He2+ ions, as well as the associated formation of defects and structural deformations. During the experimental work conducted, it was established that the accumulation of deformation-structural distortions is of two stage nature, having a direct dependence on irradiation fluence and, therefore, on atomic displacement value. It was determined that at small atomic displacement values (less than 10 dpa), the main mechanisms of structural distortions are caused by tensile residual stresses, the value of which is less than 0.1 GPa. At the same time, the deformation distortion of chemical bonds has a pronounced anisotropy associated with a more pronounced distortion of the Zr–OI chemical bonds, the distortion of which results in the formation of vacancy defects in the form of VO. During determination of alterations in optical characteristics depending on the atomic displacement value, it was found that the dominant role at small values of dpa (less than 10 dpa) is played by point defects, which influence the formation of obstacles in the form of absorbing centers. In this case, an increase in the irradiation fluence above 1017 ion/cm2 results in a growth in the linear refractive index, the change of which has a direct correlation with the value of residual stresses in the damaged layer. Certain dependencies of changes in structural features and their relationship with deformation distortions, as well as the accumulation of vacancy defects, can subsequently be used to predict the potential of using these ceramics as materials for new generation nuclear reactors.

Properties of W–Ta Materials of the Neutron-Producing Target of the Subcritical Assembly at the National Scientific Centre ‘Kharkiv Institute of Physics and Technology’ of the National Academy of Sciences of Ukraine

The works in the field of radiation materials science of target materials of neutron sources based on the subcritical assemblies controlled by linear accelerators of electrons or protons, so-called accelerator driven systems (ADS), are reviewed. Now, electronuclear ADS systems are the prototype of safe nuclear reactors of the 5th generation. In connection with the physical start-up of the neutron source facility of the National Scientific Centre ‘Kharkiv Institute of Physics and Technology’ of the N.A.S. of Ukraine (NSC ‘KhIPT’ NASU), the target of which is fabricated from powdered tungsten covered with tantalum, the issues of preparation technology, construction, and physical and mechanical properties of W–Ta materials of targets in non-irradiated and irradiated states are considered. The nuclear-physical processes of radiation damage to the target during co-irradiation of it with both neutrons of a subcritical assembly and high-energy e- and γ-beams are analyzed. The target resources of ADS systems are estimated. As noted, the difficulty of predicting the ‘survivability’ of the W–Ta target also relates to the fact that, except for the works carried out at the NSC ‘KhIPT’ NASU in 70–90th of the last century, there are no experimental works in the world on the radiation damage of reactor materials by high-energy electrons with an energy of 100 MeV and above.

Influence of injected ions on α’ formation under ion irradiation in Oxide Dispersion Strengthened Steels

Oxide Dispersion Strengthened (ODS) steels hold great promise for applications in next generation reactors. Under irradiation, a phase separation α/ α’ can occur within the Fe-Cr matrix of ODS steels that can alter their mechanical properties. This work presents, for the first time, the characteristics of α’ precipitates enhanced by ion irradiation at 400 °C and examines the influence of the implanted ions. Far from the implanted region, α’ is reported in significant density while at the implanted peak, the α’ density is considerably reduced. This suggests that ion implantation either reduces the fraction of α’ phase formed after irradiation or delays considerably its formation. Through atom probe tomography analysis and comparison with existing literature, the low impact of the damage rate and fluence on the α’ formation in ODS steels is highlighted. Interestingly, the efficiency of ballistic mixing of α’ appears to be less pronounced in ODS steels than in Fe-Cr systems.

Comparison of Monte Carlo Tools for Displacement Calculations for Protons and Heavy Ions Irradiation on Iron and EUROFER97

Abstract Ion irradiation of structural materials for the current and future generation nuclear reactors is a widespread technique used to simulate neutron irradiation in the context of material science research. During the designing phase of irradiation experiments Monte Carlo transport codes are often employed to quantify the radiation exposure, by estimating the displacements per atom (dpa) quantity. From the available displacement calculation codes, four popular choices were selected: cern-fluka, mcnp, phits, and srim. These codes will be used to simulate the irradiation effects of proton and Fe ions on pure iron (G379) and EUROFER97 alloy (a common material in fusion applications) targets. Results coming from the different software are presented in the form of damage profiles and of number of displacements as a function of the primary beam energy. This study identifies discrepancies between the codes' outputs and provides optimized simulation setup parameters.

Evidence that Femto-H2 Exists Leads to the Correct Mechanism of Cold Fusion (Cold Fusion is Caused by Femto-D2)

Femto-H2 is generated by compression of H- H bond of H2 confined at expandable T site, and authors’ mechanism of Cold fusion is that femto-D2 created at expandable T site cause fusion of D+D due to the shielding of coulomb repulsive force because femto- D2 has so dense electron between d-d that coulomb repulsive force is shielded to cause Fusion. One of the controversial heat generation reactors is QHe (Quantum Hydrogen Energy) developed by clean-planet, who explains that QHe is a heat-generating technology with hydrogen quantum diffusion. The diffusion is induced by heating a small amount of hydrogen saturated in nano-sized nickel-based composite material. Although they did not describe it cold fusion, it is commonly understood to be cold fusion. Therefore, this kind of Cold fusion debate is so confusing that no common standard theory has been established. Therefore, I would like to explain the mechanism of QHe based on author’s Mechanism of Cold Fusion. with H2 loading into multi-stacked Ni/Cu which has large number of reaction site of grain-boundary, where the hydrogen segregates to enhance the reaction of femto- H2 generation. And femto-H2 can fuse to Cu or Ni to be mainly Ze to generate 15~17MeV, which is comparable of energy by D+D=4He+24MeV.Because femto-H2 is so small to fuse to the metal nucleus at room temperature, and at high temperature, the nucleus vibrates at high speed and femto-H2 can fuse with the metal nucleus and once the heat generation start transmutation continues. Author also discovered phenomena that seems to be related to femto-H2, which are transmutation reactor by Dr. Ohmasa and car on H2O (Water Fuel CELL) by Stanly A Meyer. Ohmasa’s transmutation reactor uses Pd coating plate and loaded hydrogen by electrolysis generate femto-H2 which enter into H2O to transmute H2O to be helium-3 and oxygen-18 clusters. Water Fuel Cell can generate huge heat by the transmutation of metal with femto-H2. Therefore, author would like to request clean-planet to study the transmutation with femto-H2 in QHe reactor to prove author’s mechanism hypo, and Cold fusion community to start the discussion on the standard theory of Cold Fusion because author’s mechanism is inconsistent with nucleus model.

Long term reactivity operation for a high temperature testing reactor (HTTR) loaded with uranium nitride fuel

A High Temperature Testing Reactor (HTTR) is one of the fourth generation nuclear reactors, the reactor uses UO2 as typical fuel with different axial and radial enrichments. In the present research, uranium nitride (UN) is proposed as a fuel for the reactor for the same core enrichment distribution. The core multiplication factor, cycle length, discharge burnup are calculated and the results of UO2 fuel are compared with UN fuel results. Neutron spectrum, fissile isotopes behavior and Breeding capability are compared for the two fuel types. Kinetic parameters, Control rods and Burnable poisons worth are also evaluated. The results indicated that uranium nitride (UN) has a longer fuel cycle, higher breeding capability and harder thermal flux compared to UO2 fuel.

Site effect on seismic response of buried nuclear power plant structure

The development of multifunctional, multipurpose small modular next generation nuclear reactors has become a trend in designing small-scale nuclear power stations. In this study, taking a new next-generation buried nuclear power structure as the target structure, a 3D finite element model of the site-nuclear power plant structure system was established, and its response at different sites was studied. Using the internal substructure method, this study initially explored the impact of earthquake input and artificial boundary locations on the structural response. The results indicate that the internal substructure method enhances computational efficiency without compromising accuracy. For this model, it is recommended to input earthquake motion close to the structure, set the lateral boundary location to three times the width of the structure, and set the bottom boundary location to two times the depth of the structure. Subsequently, an investigation into the response of the structure at varying site wave velocities is conducted. The results show that as the wave velocity increases, the acceleration response intensifies, while the displacement response diminishes. However, the relationship is nonlinear; the response exhibits less variation with increasing site wave velocity. The seismic response of a structure is profoundly influenced by site conditions, emphasizing the imperative need for careful consideration of specific site characteristics in seismic analyses. In addition, the response pattern of the overall structure is significantly different from that of conventional large surface-sited nuclear power plants, and further research should be conducted on the seismic response characteristics of buried nuclear power plants.

Economics and finance of lead fast reactors: A systematic literature review

Generation IV (Gen-IV) nuclear reactor designs are receiving increasing attention because of their potential to achieve key goals in terms of sustainability, safety, reliability, and economics. One of the six technologies selected in the Gen-IV program is the lead-cooled fast reactor (LFR). This work focuses on LFRs, examining both pure lead and lead-bismuth eutectic designs. Through two systematic literature reviews, we consolidate the state-of-the-art in economics and finance for LFRs. The first review considers scientific literature and retrieves 12 articles. The second focuses instead on industrial literature, resulting in 12 additional documents. Economics literature is very scarce and sometimes of low quality. Economic estimations for the specific capital cost [$/kWe] and the cost of electricity [$/MWh] can vary by an order of magnitude (1500–25000 $/kWe and 30–350 $/MWh, respectively), while design organizations typically do not publicly share financial details. Finance literature is almost nonexistent. We report, in the final part of the work, notable knowledge gaps and further possible research areas.

Radiation shielding properties of composites of TiZrNbHfTa refractory high entropy alloy reinforced with TiZrNbHfTaOx high-entropy oxide

Radiation shielding materials play a critical role in advancing the next generation of nuclear reactors. Among these, high-entropy materials with refractory properties emerge as a new category for radiation shielding applications. This study introduces TiZrNbHfTa-TiZrNbHfTaOx composites tailored for such purposes. Initially, a TiZrNbHfTa refractory high-entropy alloy (RHEA) is synthesized via mechanical alloying, followed by the oxidation of the alloy to obtain its refractory high-entropy oxide (RHEO), TiZrNbHfTaOx. Composites are then prepared by blending 5%, 10%, and 20% weight fractions of RHEO with RHEA using mechanical alloying. Comparative analysis of shielding properties reveals a decrease in mass attenuation coefficient and effective atomic number with increasing RHEO content. While RHEA exhibits superior shielding performance against gamma photons and neutrons, the composite containing 20% RHEO emerges as the optimal shield against alpha and proton irradiations. The equivalent neutron dose rate and removal cross-section values are obtained 49.8% and 0.142 for the composite with 20% RHEO, respectively. These findings, supported by computational simulations in this study, contribute valuable insights for the advancement of novel shielding materials crucial for future nuclear reactor technologies.

Precipitation of oxide dispersion strength steels after long-term annealing at temperature of 475 °C

Oxide dispersion strengthened steels are key alloys used in nuclear installations. Steels with chromium content of up to 10 wt % are suitable due to their advantageous properties and these alloys are used for the construction of technological devices in the primary circuit of nuclear power plants. From other studies, it can be concluded that chromium has anti-corrosive properties due to the formation of a passivation layer, which results in reduced activation of the material by thermal neutrons. Macroscopic properties are determined by their microstructure, and therefore, the description of the microstructure is important. Transmission Mössbauer spectroscopy and atom probe tomography were used to characterise the material. This provides information about the physical and/or chemical environment of the resonant atoms can be obtained. Obtained spectral parameters reach saturation values from which the solubility limit of chromium in iron can be determined. In Cr-rich phase, the solubility limit can be estimated from the value of spectral parameters of the single-line in the spectrum annealed for the longest time. The suggested procedures are subsequently applied to the case studies of stainless steels suitable for the construction of various components of the III+/IVth generation of nuclear reactors (including fast and fusion reactors).

Comparison between DNS and RANS approaches for liquid metal flows around a square rod bundle

The thermal-hydraulic characteristics of liquid metal flows around rod bundles are of great interest for the research and design of fourth generation nuclear reactors. Currently, a large research effort is aimed at the development of accurate numerical models for low Prandtl number fluid flows, since the data available in the literature are quite scarce. Direct Numerical Simulation (DNS) is undoubtedly the most accurate approach, but its large requirements of computational resources and time make it less practical than other simplified methods such as the Reynolds-Average Navier Stokes (RANS) approach. The present paper provides a comparison between numerical results of a flow of liquid Lead-Bismuth Eutectic (LBE) at Pr=0.031 around four vertical cylindrical rods arranged in a square lattice, obtained by DNS and RANS. Several turbulence models are considered, including the standard k−ε, k−ω SST, and two Reynolds stress models, namely the ones by Launder, Reece and Rodi (LRR), and Speziale, Sarkar and Gatski (SSG). The accuracy of these models is assessed by comparing the mean Nusselt number, the pressure drop, and local field distributions with those obtained by DNS.

Neutronic simulation of Traveling Wave Reactor (TWR) core in multi-cycles using Monte Carlo method

Abstract An ingenious design, Traveling Wave Reactor (TWR) has been suggested for full exploitation of uranium resources in next generation of nuclear reactors. This design involves the slow propagation of a nuclear fission wave through a long initially subcritical core and the transmutation of nuclear fuel. It has been widely proven that the initiation and propagation of fission waves are feasible prompting the Terra Power Company to conduct preliminary commercialization studies on this project. In equilibrium state, the shapes of the neutron flux, nuclide densities and power density distribution remain constant but the burning region moves in axial (or radial) directions. In this case the in situ fissile material production and consumption drives the operation. Only natural or depleted uranium is required for fresh fuel region. However, in equilibrium state, the burning region contains a spectrum of fission products as well as higher actinides whose are not easily available for initial TWR core construction. One solution is based on the use of enriched uranium in the first cycle to ignite the fission wave and then employment of the composition in subsequent cycles up to the equilibrium state. The main objective of this work is the feasibility study of forming a fission wave as well as the neutronic analysis of a Gas-Cooled Traveling Wave Reactor in various cycles up to the equilibrium one. The MCNP Monte Carlo code was utilized for the analysis of criticality and burn-up calculation. In each cycle, the axial fresh fuel loading and spent fuel discharge, were subjected to analysis. Results showed that the equilibrium state can be obtained by using special arrays of absorbers, from the third cycle where the shape of neutron flux and the power density distribution in axial direction remain unchanged.

Off-design operation of supercritical CO2 Brayton cycle arranged with single and multiple turbomachinery shafts for lead-cooled fast reactor

The lead-cooled fast reactor (LFR) engenders novel requisites for the energy conversion system when applied to maritime ships and distributed energy supply systems. The supercritical carbon dioxide (sCO2) Brayton cycle, with the advantages of high efficiency, space-saving, simplicity, and flexibility, is a promising candidate for LFR. Here, we present the first study on the off-design operation of the sCO2 cycle for LFR, with the turbine and compressor arranged on a single shaft (coaxial-shaft) or two shafts (split-shaft). The off-design performances of the system are investigated using four rotational speed (RS) control-based hybrid control strategies. We showed that the split-shaft system controlled by separate turbine speed and compressor speed has a wider operating space of power load than the coaxial-shaft system. Among the four investigated control strategies, the RS-inventory hybrid control can achieve the best thermal efficiency during load variation by operating at the turbine speed and mass flow rate corresponding to the optimal turbine isentropic efficiency. The maximum operating space of power load (0%–100 %) can be achieved by the RS-turbine inlet pressure hybrid control and the RS-bypass hybrid control. Our study can provide guidance for the flexible and safe part-load operation for small-scale modular Gen-IV nuclear reactors.

Rotation-invariant rapid TRISO-fueled pebble identification based on feature matching and point cloud registration

The Pebble-bed reactor (PBR) utilizing TRISO-fueled pebbles is a promising Gen-IV reactor design due to its inherent safety and thermal efficiency. Identification of pebbles exiting the core may provide IAEA with additional information enhancing PBR safeguards help validate software predicting pebble flow behavior and burnup, and optimize fuel usage strategies should granular fuel management be needed. We propose a rapid, accurate pebble identification method applicable to commercial PBRs. Our approach relies on the unique spatial distribution of TRISO particles associated with each pebble, which can be extracted through X-ray CT. We developed an algorithm to identify a single pebble from 100,000 pebbles and achieved 100% identification accuracy in 90,000 trials, with arbitrary rotations and high noise levels. Remarkably, the outermost TRISO particle positions within a 0.6-mm shell suffice for unique identification. With an average identification time of under 50s per pebble, our method is compatible with PBR operational demands.