Nuclear Systems Research Articles

Current Progress on the Development of the Integrated Waste Management Consent-Based Siting Process in the United States

Consent-based siting is an approach to siting facilities that prioritizes the participation and needs of people and communities and seeks their willing and informed consent to accept a project in their community. The U.S. Department of Energy (DOE) is firmly committed to a consent-based approach to siting one or more spent nuclear fuel management facilities that enables meaningful and inclusive public participation, addresses community well-being and community needs, and centers equity and environmental justice as core values. At present, nuclear energy accounts for 19% of U.S. electricity production and half of the nation’s carbon-free energy. Nuclear energy is key to achieving the Nation’s goal of a 50% reduction in carbon emissions by the end of the decade, 100% clean electricity by 2035, and a net-zero emissions economy by 2050. Nuclear power will be key to achieving the Nation’s goals, as the United States strives to reduce carbon emissions; ensure energy independence; and maintain leadership in nuclear technology, safety, and security. However, the use of nuclear energy also requires an effective, integrated system for safely managing, storing, and permanently disposing of spent nuclear fuel. Since the 1950s, approximately 90 000 metric tons of heavy metal (MTHM) of spent nuclear fuel has been generated from commercial nuclear power generation in the United States. This inventory grows by approximately 2000 MTHM every year. This commercial spent nuclear fuel is currently safely and securely stored at over 70 sites in more than 30 states where it is enclosed in either steel-lined concrete pools of water or steel and concrete containers known as dry storage casks. The United States also has an inventory of noncommercial spent nuclear fuel and high-level radioactive waste, which makes up less than 5% of the total inventory. Management of the Nation’s spent nuclear fuel and high-level radioactive waste is the responsibility of the DOE, under the Nuclear Waste Policy Act of 1982, as amended. This responsibility includes finding sites to store and eventually dispose of this material. Per Congressional direction, DOE’s current activities are focused on establishing a federal interim storage capability for commercial spent nuclear fuel following a consent-based approach. However, a permanent disposal solution for spent nuclear fuel and high-level radioactive waste will still be needed. DOE will apply lessons learned from consent-based siting for one or more interim storage facilities to future siting efforts for other elements of an integrated nuclear waste management system. An integrated waste management system will include consolidated interim storage capacity, a permanent disposal pathway, and the transportation infrastructure needed to move spent nuclear fuel and high-level radioactive waste from nuclear power plants to storage and disposal facilities. This paper describes DOE’s current progress to identify one or more sites to develop a federal consolidated interim storage capability for commercial spent nuclear fuel using a consent-based siting process.

Study on start-up characteristics of a heat pipe cooled reactor coupled with a supercritical CO2 Brayton cycle

The new megawatt nuclear power system coupling a heat pipe cooled reactor with a supercritical CO2 Brayton cycle is considered one of the most promising energy conversion systems for large Unmanned Undersea Vehicles (UUVs). This system combines the characteristics of a heat pipe cooled reactor and a S-CO2 Brayton cycle, such as heat pipe start-up limitations and significant thermal inertia. It is unclear whether existing start-up control strategies for either heat pipe cooled reactors or S-CO2 Brayton cycle are suitable for this new system. Therefore, the start-up characteristics and the control strategies of the new megawatt nuclear power system are investigated in this study. A simulation code is developed for the study of the start-up characteristics of this system, which includes models for the reactor, heat pipes, and the S-CO2 Brayton cycle. The validation of the code is conducted based on the experiments as well as the design values. Good agreement between the calculation results and the experimental values or the design values are obtained. And then, the control scheme and start-up strategy of the new system are proposed. The start-up process is divided into four sub-stages, that is, core start-up from cold to critical state, heat pipes start-up, compressor start-up and turbine preheating, and branch switching. Different control strategies are applied to each sub-stage. The results show that the shaft raising rate of 2000 rpm/min is effective in overcoming expansion resistance without causing excessive overshoot in mass flowrate. This control strategy can achieve stable start-up of the new system without exceeding a temperature change rate of 10 K/min. This research may serve as a reference for the conceptual design of this new megawatt nuclear power system.

Development of a hybridized small modular reactor and solar-based energy system for useful commodities required for sustainable cities

The current study develops a hybridized small modular nuclear reactor and solar-based system designed specifically for sustainable communities in metropolitan areas to meet their power, heat, clean fuel, fresh water and food requirements. Both floating-type and bifacial-type photovoltaic arrays are integrated with a high-temperature gas-cooled small modular reactor. In the integrated system, a Rankine cycle subsystem for power generation, a solid oxide electrolyzer for hydrogen production, a pressure swing adsorption unit with ammonia reactor for nitrogen and then ammonia production, a multi-effect desalination unit for freshwater generation, and heat recovery units in order to provide heating to residential area, food drying facility, greenhouse, and fish farm facilities. For both floating and bifacial PV options, the 120 MWp PV modules are considered and integrated with two small modular reactor units at 250 MWth capacities each, 500 MWth in total. A thermodynamic analysis, which is based on energy and exergy approaches, is carried out. The time-dependent analyses are carried out for floating-type PV and bifacial-type PV options for the cities of Istanbul in Turkey; Barcelona in Spain; Los Angeles (LA) in the United States; Tokyo in Japan; and Toronto in Canada by using their hourly meteorological data, source and load capacities in a typical meteorological year. Among ten cases, between 112835 and 146180 tonnes of ammonia are generated in a typical year while meeting the heating and power requirements of sustainable communities in metropolitan areas. Among ten cases, the average overall energy and exergy efficiencies are found as 41.04% and 46.88%, respectively. The hybridization of such nuclear and solar systems achieves unique advantages by preventing intermittencies compared to the renewable-alone systems, reducing carbon emissions compared to fossil-based systems, and lowering the levelized costs of energy compared to the nuclear energy-alone systems.

Research Advances in the Application of the Supercritical CO2 Brayton Cycle to Reactor Systems: A Review

Amid the global emphasis on efficient power conversion systems under the “dual carbon” policy framework, the supercritical CO2 (SCO2) Brayton cycle is a noteworthy subject, owing to its pronounced efficiency, compact design, economic viability, and remarkable potential to increase the thermal cycle efficiency of nuclear reactors. However, its application across various nuclear reactor loops presents divergent challenges, complicating system design and analytical processes. This paper offers a thorough insight into the latest research on the SCO2 Brayton cycle, particularly emphasising its integration within directly and indirectly cooled nuclear reactors. The evolution of the Brayton cycle in nuclear reactor systems has been meticulously explored, focusing on its structural dynamics, key components, and inherent pros and cons associated with distinct reactor loops. Based on the theoretical frameworks and empirical findings related to turbomachinery and heat exchangers within the cycle, we chart a course for future enquiries into its critical components, underscoring the indispensable role of experimental investigations. This paper conclusively assesses the feasibility of deploying the SCO2 Brayton cycle in direct and indirect cooling contexts, offering a forward-looking perspective on its developmental trajectory. The SCO2 Brayton cycle may become a focal point for research, potentially creating avenues for nuclear energy endeavours.

Cross-Evaluation of the PU-MET-FAST criticality benchmark experiments in ICSBEP handbook (by the end of 2022)

Criticality experiments play an important role in the fields of criticality safety, neutron transport program validation and nuclear data validation. ICSBEP handbook collects about 5000 criticality benchmark experiments. Most of the criticality benchmark experiments in ICSBEP handbook are retrospective evaluation experiments, they are evaluated long after the experiments and their evaluations may be based on inaccurate or incomplete data. When these experiments are used in nuclear data validations and cross section adjustments, contradictions between some criticality benchmark experiments have been found. Therefore, it is necessary to analyze whether these retrospective evaluation criticality benchmark experiments are applicable in the fields of nuclear data validations (including cross section adjustments) and validations of nuclear system design. To solve this problem, we have developed cross-evaluation method of criticality benchmark experiments, which could evaluate the biases of criticality benchmark experiments from the point of statistics. In this article, we used this method to cross-evaluate the PU-MET-FAST criticality benchmark experiments in ICSBEP handbook. The results indicate that there are many contradictions between the retrospective evaluation experiments in the PU-MET-FAST category. Because of the limited number of experiments in this category, the number of experiments that are similar to each other after cluster analysis is small, which reduces the effectiveness of the cross-evaluation. There are only 1 or 2 experiments in some clusters. Even if there are contradictions in these clusters, it is impossible to judge which experiment has a higher quality. Based on these results, it is recommended to do new PU-MET-FAST criticality benchmark experiments to resolve contradictions between these retrospective evaluation experiments and to help validate the nuclear data more effectively.

An Investigation on the Church’s Impact on Elderly care in Ghana in the Context of Increasing Poverty and Political Polarisation

Social problems in society, such as the care of the aged, hinder people’s full potential and often arise from societal neglect. In Ghana, a significant percentage of the elderly population requires assistance with daily activities, but poverty and political polarization can hamper their care. The church has played a crucial role in addressing social problems, including the neglect, poverty, and poor health faced by the elderly. However, the church’s ability to intervene is limited, and traditional beliefs often stigmatize the elderly as witches or wizards. The healthcare sector also faces challenges in providing care for the elderly, with high costs and low coverage of health insurance. The study thus emphasized the importance of understanding the difficulties experienced by the elderly through theories such as Disengagement Theory and Convoy Theory. The study employed interviews to gather data from a sample of 12 individuals who were directly involved in caring for the elderly or had knowledge of their issues. The data analysis focused on the lived experiences of the interviewees. The weak social system, including pension systems and health insurance schemes, were found to be contributors to the struggles faced by the elderly. The elderly face challenges such as frailty, illness, poverty, and neglect. The study suggests that policymakers should review existing health insurance policies to better support the elderly. The church’s role in promoting a more inclusive and supportive society for the elderly is crucial, but it faces challenges in the changing cultural landscape and the dominance of the nuclear family system. Keywords: Old Age, Poverty, Political Polarization, Political System, Democracy and Ghana.

Decay Characteristics of Neutron Excess Argon Nuclei

Bevelacqua Resources, 7531 Flint Crossing Circle SE, Owens Cross Roads, AL 35763 USA bevelresou@aol.com In neutron star mergers, neutron excess nuclei and the r-process are important factors governing the production of heavier nuclear systems. A single-particle model evaluation of argon nuclei suggests that the heaviest Z = 18 nucleus will have mass 69 with filling of the 3s1/2 neutron shell. A = 51 – 69 argon isotopes have limited experimental half-life data, but the model predicts beta decay half-lives in the range of 0.473 – 20.3 ms. Based on previous calculations for Z = 9 - 17, 20, 26, and 30 systems and comparisons to the 51Ar - 56Ar, 58Ar, and 60Ar calculations, summarized in the Japanese Nuclear Data Compilation, the single-particle model results likely overestimate the half-lives of A = 51 – 69 neutron excess argon nuclei.

Colloidal processing, sintering and properties of aluminum borate Al18B4O33 porous ceramics

Aluminum borate (Al18B4O33) is one of the mixed oxides with extensive applications as due to its good thermomechanical properties. Its application requires of accessible processing techniques to provide controlled microstructures and shapes. This work presents an aqueous colloidal route to manufacture porous aluminum borate ceramics from micrometric Al18B4O33 powder (∼2.0 μm) obtained by high temperature reaction of alumina and boric acid. Zeta potential measurements determine that isoelectric point of the Al18B4O33 powders suspended in water happens at pH = 8.5 achieving a maximum positive value of 60 mV for pH < 4. Rheological characterization of concentrated slurries showed a pseudoplastic flowing behavior with a thixotropic cycle, being the yield point and the thixotropy value strongly dependents on the solid content. Shear thinning curves were fitted to a Windhab model that confirmed a face-edge interactions of the particles in the suspension.Porous ceramic with a green density of 50 % and a narrow micronic (1–2 μm) porosity are obtained by slip-casting of the optimal concentrated suspension (33 vol%). Sintering above the formation temperature showed a decrease in the density of the compacts as a consequence of the acicular grain growth. Four sintering conditions were tested to follow the structural and compositional evolution. It has been observed that thermal treatments up to 1300 °C do not affect the composition but the microstructure. At 1400 °C presence of alumina is detected indicating the initial decomposition of Al18B4O33 and evaporation of boron. At 1500 °C decomposition and evaporation phenomena are massive. The porous compacts sintered below the decomposition temperature are characterized by scanning electron microscopy, mercury intrusion porosimetry and mechanical strength, indicating that at 1300 °C Al18B4O33 achieves the more uniform microstructure and the better mechanical response.It should be noted that for the first time the Zeta potential curve for aluminum borate powder is reported. The processing methodology employed in this study, has potential application to the manufacture of porous materials with insulating applications to replace traditional ceramic materials working in harsh conditions, especially in nuclear energy systems with complex geometry by casting techniques and even by 3D printing.

New Covariant Density Functionals of Nuclear Matter for Compact Star Simulations

We generate three families of extended covariant density functionals of nuclear matter that have varying slope of symmetry energy and skewness at nuclear saturation density, but otherwise share the same basic parameters (symmetry energy, compressibility, saturation parameters, etc.) with the standard DDME2, DD2, and MPE functionals. Tables of the parameters of these new density functionals are given, which can be straightforwardly used in DDME2, DD2, and MPE parameterization-based codes. Furthermore, we provide tables of a large number of equations of state (81 for each family) that can be used in astrophysical simulations to assess the impact of variations of the not-well-known slope of symmetry energy and skewness of nuclear systems on the astrophysics of compact objects. We also provide tables of computed integral parameters (mass, radius, and tidal deformability) that can be used, e.g., for modeling gravitational waveforms. Finally, for the extended DDME2-based parameterization, we implement a first-order phase transition to quark matter to obtain a family of equations of state that accommodates a phase transition to quark matter. Analogous tables of the equations of state and integral parameters are provided for this case as well.

Virtual lattice method for efficient Monte Carlo transport simulation of dispersion nuclear fuels

The randomly dispersed Tristructural-isotropic (TRISO) particle fuel is an attractive nuclear fuel type widely used in advanced reactor designs. However, the massive randomly-distributed fuel elements and the prohibitive computational costs pose challenges for high-fidelity modeling and simulation of dispersion fuels. To address this, we present a virtual lattice optimization method for accelerating the Monte Carlo particle transport simulation of nuclear systems with dispersion fuels. The new method involves using a regular overlaid mesh on the dispersed fuel region to improve geometry processing in Monte Carlo simulations. We describe the optimization methodology in our paper and construct a theoretical performance model to determine the optimal lattice pitch size that minimizes calculation cost. We then develop and implement the algorithm of the virtual method in the open-source Monte Carlo code OpenMC. To demonstrate the efficacy of the new algorithm, high-fidelity HTR-PM full core models based on the Shidao-Bay nuclear power plant are constructed, including an explicit representation of up to 420,000 fuel pebbles. Criticality and depletion simulations are performed on large-scale models. Results show that the k-effective derived by the optimized code agrees well with both the experiment and the original code. Further performance comparison shows that the virtual lattice method exhibits lower time consumption, higher parallel efficiency, and lower storage requirement in comparison to the conventional physical lattice method. The study establishes the effectiveness of the proposed methodology in facilitating efficient and accurate high-fidelity simulations of large-scale dispersion fuel models.

Updates on grading mesothelioma.

Mesothelioma is a rare disease with an historically poor prognosis. Over the past decade, a grading system has been developed that is a powerful prognostic tool in epithelioid mesothelioma. Grading of epithelioid mesothelioma is now required or strongly recommended by expert consensus, the College of American Pathologists, the World Health Organization, and the International Mesothelioma Interest Group. The original nuclear grading system for epithelioid mesothelioma, developed in the United States, split epithelioid mesotheliomas into three prognostic groups with marked differences in survival. Now, this three-tiered nuclear grading system has been combined with the presence or absence of necrosis to form the currently recommended two-tiered grading system of low- and high-grade epithelioid mesothelioma. This review will focus on the development of this grading system in mesothelioma, the grading system's shortcomings, and the application of the grading system to cytology specimens and other extra-pleural sites. Lastly, this review will briefly discuss alternative grading systems and future considerations.

The squeezed dark nuclear spin state in lead halide perovskites

Coherent many-body states are highly promising for robust quantum information processing. While far-reaching theoretical predictions have been made for various implementations, direct experimental evidence of their appealing properties can be challenging. Here, we demonstrate optical manipulation of the nuclear spin ensemble in the lead halide perovskite semiconductor FAPbBr3 (FA = formamidinium), targeting a long-postulated collective dark state that is insensitive to optical pumping after its build-up. Via optical orientation of localized hole spins we drive the nuclear many-body system into this entangled state, requiring a weak magnetic field of only a few milli-Tesla strength at cryogenic temperatures. During its fast establishment, the nuclear polarization along the optical axis remains small, while the transverse nuclear spin fluctuations are strongly reduced, corresponding to spin squeezing as evidenced by a strong violation of the generalized nuclear squeezing-inequality with ξs < 0.5. The dark state corresponds to an ~35-body entanglement between the nuclei. Dark nuclear spin states can be exploited to store quantum information benefiting from their long-lived many-body coherence and to perform quantum measurements with a precision beyond the standard limit.

The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0

Numerical reactor technology is the development direction of advanced nuclear energy system simulation, which has the ability of high-fidelity multi-physical field coupling calculation. The thermal–hydraulic and neutron physics fields in reactor core exhibit strong feedback relationships. Establishing the neutronic thermal–hydraulic coupling calculation model of the whole core “pin-by-pin” is the key technology for the development of numerical reactors. In this study, the neutron diffusion analysis model and cross-section database have been developed based on the open source CFD platform OpenFOAM. The three-dimensional neutron diffusion analysis model based on the finite volume method has been validated by benchmark problems. By integrating the neutron diffusion analysis model and cross-section database into nuclear reactor core three-dimensional thermal–hydraulic characteristics analysis code CorTAF, the strongly coupled neutronic thermal–hydraulic coupling analysis code CorTAF-2.0 has been developed. The neutronic thermal–hydraulic coupling phenomena in the full core of AP1000 have been simulated using CorTAF-2.0, yielding the sub-channel level parameters distribution of the whole core. This work offers valuable references for reactor full-core neutronic thermal–hydraulic coupling analysis.

Assessment of alternative nuclear fuel cycles for the Brazilian nuclear energy system

Alternative nuclear fuel cycles could contribute to the future of nuclear fission technology reducing the use of natural resources, waste management and lessening greenhouse gas emissions. To increase understanding of future technologies, analysis of power generation systems are essential and supports long-term decision-making. On this basis, this research presents an evaluation and comparison considering the use of alternative nuclear fuels in Pressurized Water Reactors (PWR) based on the Brazilian nuclear energy system. Assessment results show consumption of resources, inventories of spent fuel, nuclear waste, greenhouse gas emissions and costs of nuclear fuels. Among the main results, Mixed Oxide (MOX) fuel appears as the most promising by multi-criteria analysis considering two of the three evaluated areas. On the other hand, if costs have an importance greater than 55%, the conventional fuel UOX appears as the most advantageous. Thorium-uranium dioxide (TRU-U) and Transuranic-thorium dioxide (TRU-Th) fuels are in intermediate positions, along with Uranium oxide (UOX), and may still be options depending on the intentions of decision-makers.

Combined Cycle Nuclear System Architecture for Crewed Mars Spacecraft Propulsion and Power

Nuclear thermal propulsion can potentially reduce the time of flight and spacecraft system mass needed for human spaceflight beyond cislunar space. This nuclear propulsion system has comparable thrust capability to chemically impulsive systems, which at about twice the specific impulse, can double the delta-velocity (ΔV) for the same propellant mass. However, the canonical problem for nuclear propulsion has always been that its benefits are shadowed by low technology readiness of a complex system. This paper describes a combined cycle nuclear thermal rocket (CCNTR) system architecture for propulsion and electrical power that comprises a 42-MWt-capable nuclear reactor core to provide 9.4 kN thrust on demand at a specific impulse of 940 s. The liquid hydrogen propellant flow through the rocket chamber cools the reactor during burns, thereby producing thrust while concurrently rejecting waste heat to space. The reactor also produces up to 100 kWe power for the spacecraft, eliminating the need for solar power generation and averting challenges associated with restarting a cold reactor for propulsive burns. Radiators reject the waste heat from electrical power production. Earth-to-Mars orbital transfers less than 100 days appear feasible assuming 680,000 kg of liquid hydrogen propellant and a vehicle dry mass of 83,000 kg that includes the 13,000 kg CCNTR system. Together, these results suggest that a CCNTR could be most promising to enable crewed missions to Mars.

Design and implementation of a virtual nuclear pulse signal generator

This paper presents a virtual nuclear pulse signal generator. The pulse generator can be used in specific scenarios such as calibration and maintenance of nuclear detection systems, validation of nuclear signal processing algorithms, and nuclear physics teaching experiments, as a substitute for real radioactive sources and detectors to generate nuclear pulse signals. This helps to avoid exposing experimental personnel to radiation hazards. The pulse generator is capable of generating specific nuclear pulse signals according to experimental requirements, and these signals has the same physical characteristics as real nuclear pulse signals, including amplitude distribution, time and amplitude randomness. In the pulse generator, we conducted theoretical analysis of the radioactive source and detector, and established a model library for both, to obtain crucial parameters of the nuclear pulse signals. The pulse generator is built on a hardware framework with field-programmable gate array (FPGA) as the core, utilizing these key parameters to generate signals with realistic physical characteristics of nuclear pulse signals, including amplitude and time characteristics. The counting rate of the output nuclear pulse signals can be adjusted by configuring the hardware parameters of the pulse generator.

Artificial Intelligence Detection System of Radioactive Nanocomposites in Liquid-Filled Containers for Nuclear Security

Nuclear terrorism resulting from the illicit trafficking of nuclear and radioactive materials consists of a serious threat against the security of countries. Hence, the transportation of hidden nuclear materials in large-scale cargos has emerged as an important public issue that requires immediate attention. The security architecture should merge the nuclear security systems and execute a procedure for the detection of nuclear and radioactive materials. In this regard, the detection of special nuclear materials (SNMs) in liquid-filled cargo containers is an essential matter of homeland security because of the difficulties imposed by liquids in performing efficient manual inspection. This paper presents a new artificial intelligence (AI) system implemented with fuzzy logic tools for detecting composite materials, including Pb-Th, Pb-U, and Pb-Co, in containers filled with water by utilizing optical signature information obtained with COMSOL. The developed AI system and its underlying method are validated for a scenario of detecting Pb-Th, Pb-U, and Pb-Co in water-filled containers.

Advancement Toward the Experimental Measurement of the Lead–Bismuth Eutectic Thermal Properties Using Differential Scanning Calorimetry Technique

Abstract The differential scanning calorimetry (DSC) method has recently emerged as a sophisticated and precise technique for promising contributions to the thermal analysis of various materials, including heavy liquid metal (HLM) coolants. However, there is a lack of experimental studies on the thermal properties of lead-based fluids, such as lead–bismuth eutectic (LBE) and lead–lithium eutectic, which are potential candidates for use as coolants, breeders, and neutron multipliers in advanced nuclear systems like the fourth-generation lead-cooled fast reactor. The available experimental data on the thermal properties of LBE and other lead-based fluids is limited, and the measurements have significant uncertainty. In addition, the composition of components used in the previous studies is inconsistent, and the environmental conditions were often unknown. Therefore, to fill these gaps and advance the thermal properties measurement technique for heavy liquid metal coolants, ENEA Brasimone, in collaboration with DICI-UNIPI, has installed a DSC instrument setup. The experiments performed at the installed DSC setup are focused on measuring some essential thermal properties of LBE using DSC. The experience gained from this work will facilitate the measurement of other fluids based on lead alloy, especially lead–lithium eutectic, a potential candidate for breeder, coolant, and neutron multiplier in demonstration power plant fusion reactors. This study represents the first effort to advance the DSC approach for accurately measuring the thermal characteristics of heavy liquid metals that are highly reactive, such as lead–lithium, which has significant potential in advanced nuclear systems.

Research on the steam power system operation condition assessment method based on PCA-AE

The traditional steam power system operation condition assessment method relies on expert knowledge and is difficult to achieve complete objectivity. Most data-driven methods require training models with system anomaly data. In response to the above issues, this article proposes the A Novel PCA-AE Steam Power System Operation Condition Assessment Method. This method utilizes the PCA model to obtain assessment indicator parameters, and uses the Autoencoder to automatically assign weights and reconstruct parameters. Finally, the vector similarity principle is used for quantitative assessment. Based on the simulation data of marine nuclear power system steam turbine and the experimental data of supercharged boiler, the feasibility and generalization ability of this method are verified. The new method can meet the requirements of different steam power systems, including the nuclear power system secondary circuits, and provide accurate quantitative assessment results of operating conditions to guide maintenance work. Meanwhile, this method does not rely on expert knowledge and abnormal data training sets, and has fewer limitations compared to traditional methods.

Multi-body entanglement and information rearrangement in nuclear many-body systems: a study of the Lipkin–Meshkov–Glick model

Multi-body entanglement and information rearrangement in nuclear many-body systems: a study of the Lipkin–Meshkov–Glick model