Reactor Neutron Research Articles

Preliminary neutronics and thermal analysis of a heat pipe cooled traveling wave reactor

Heat pipe cooled traveling wave reactor (HPTWR) is a novel concept of heat pipe reactor, which adopts neutron breeding wave to pursue Uranium-Plutonium breeding. In this paper, the cladding material, power output, and axial breeding fuel region length of the HPTWR are optimized in order to achieve the safe operation of reactor with reactor neutronics and HP thermal analysis. Monte Carlo neutron transport code RMC is used to obtain the neutronics results (reactivity, consumption and breeding of nuclear fuel, and power distribution). A heat pipe (HP) code based on thermal resistance–capacitance network model (TRCNM) is developed to analyze the heat transfer characteristics of HP using the power distribution of fuel element obtained by RMC code in the optimized HPTWR. Results show the optimized HPTWR can achieve 70 MWth power output and 46 years continuous operation. The highest temperature peak of HP wall in the highest power fuel element of reactor appears at beginning of cycle (BOC) and is about 1800 K. Due to the flattening of axial power distribution with propagation of traveling wave, the temperature peak of HP wall gradually decreases to about 1784.6 K at end of cycle (EOC), which can reduce the thermal stress of HP wall and improve the safety of reactor. The corresponding heat transfer margin of HP in the highest power fuel element is always above 19.0% during the whole reactor operation, which helps to ensure the safe operation of the optimized HPTWR.

Радиационная и пожарная опасность натриевого теплоносителя

Introduction. Nuclear power plants in the Russian Federation produce about 20 % of the total electricity. On the basis of fast neutron reactors using sodium as a coolant, Rosatom State Corporation is implementing the “Breakthrough” project aimed at the implementation of a nuclear fuel cycle using the energy potential of natural uranium. Study and generalization of information about fire and radiation hazards of sodium coolant is an urgent task to ensure safety in the operation of this type of reactors. Goals and objectives. The purpose of the article is an analytical study of information about sodium coolant radiation and fire hazard, published in domestic and foreign scientific literature. To achieve it, an analysis of fast neutron reactors operating in the world was carried out, the coolants used in fast reactors were considered. System analysis of radionuclides present in sodium coolant has been carried out, comparative diagrams according to radionuclide radiation properties are presented, the most dangerous radionuclides for humans present in the sodium coolant have been identified. The fire hazard of sodium metal and peculiarities of extinguishing fires associated with the leakage of sodium coolant in the primary and secondary circuits of the reactor plant are analyzed. Results and its discussion. At the present time, there is 1 experimental-industrial fast neutron reactor in operation in the world, located in Russia, and 1 experimental reactor in India and China. Sodium is used as a coolant in these reactors. As a result of the literature analysis from open sources, it was found that the main sources of impurities in the metal coolant are protective gas, structural and technological materials of the installation, and products of nuclear reactions. A systematic analysis of the properties of radionuclides present in the metal coolant made it possible to determine the most dangerous of them for human life and health. The fire hazard of sodium coolant is mainly due to its chemical activity. The main ways of extinguishing spilled sodium is isolation from oxygen by covering the puddle of metallic sodium with powdered fire-extinguishing MHS, powdered aluminum oxide or reduction of oxygen concentration in the air below 4 % by volume gas extinguishing with nitrogen, carbon dioxide or inert gases. Conclusions. As a result of the analysis of radionuclides present in the sodium coolant of a fast neutron reactor, it was found that the most dangerous for people are 24Na, 137Cs, 125Sb, 22Na, 239Pu, 54Mn, 110mAg, 131I. Based on the results of the analysis of the fire hazard of radioactive sodium, the most common fire extinguishing agents are established and the features of extinguishing fires that occur when a sodium coolant ignites are described.

Research on numerical calculation of reactor neutron noise in frequency domain and time domain

The theoretical calculation of neutron noise plays an important role in interpreting experiments and providing abnormal directivity information in the reactor. Therefore, it is necessary to study the numerical calculation methods of reactor neutron noise in frequency domain and time domain. Based on the theoretical framework of the Galerkin finite element method, this work derives the specific equations of reactor neutron noise in frequency domain and time domain, and then the neutron noise of a small core configuration under helium bubble perturbation was calculated. With the change of the perturbation flow of the helium bubble source, that is, the change of the characteristics of the noise source, the deviation of the calculated results of the neutron noise in frequency domain and time domain also changes sequentially and presents a nonlinear relationship with the noise source.

Development and evaluation of dose calculation algorithm with a combination of Monte Carlo and point-kernel methods for boron neutron capture therapy

We developed a ‘hybrid algorithm’ that combines the Monte Carlo (MC) and point-kernel methods for fast dose calculation in boron neutron capture therapy. The objectives of this study were to experimentally verify the hybrid algorithm and to verify the calculation accuracy and time of a ‘complementary approach’ adopting both the hybrid algorithm and the full-energy MC method. In the latter verification, the results were compared with those obtained using the full-energy MC method alone. In the hybrid algorithm, the moderation process of neutrons is simulated using only the MC method, and the thermalization process is modeled as a kernel. The thermal neutron fluxes calculated using only this algorithm were compared with those measured in a cubic phantom. In addition, a complementary approach was used for dose calculation in a geometry simulating the head region, and its computation time and accuracy were verified. The experimental verification indicated that the thermal neutron fluxes calculated using only the hybrid algorithm reproduced the measured values at depths exceeding a few centimeters, whereas they overestimated those at shallower depths. Compared with the calculation using only the full-energy MC method, the complementary approach reduced the computation time by approximately half, maintaining nearly same accuracy. When focusing on the calculation only using the hybrid algorithm only for the boron dose attributed to the reaction of thermal neutrons, the computation time was expected to reduce by 95% compared with the calculation using only the full-energy MC method. In conclusion, modeling the thermalization process as a kernel was effective for reducing the computation time.

Demonstration of Gd-GEM detector design for neutron macromolecular crystallography applications

The European Spallation Source (ESS) in Lund, Sweden will become the world's most powerful thermal neutron source. The Macromolecular Diffractometer (NMX) at the ESS requires three 51.2 × 51.2 cm2 detectors with reasonable detection efficiency, sub-mm spatial resolution, a narrow point-spread function (PSF), and good time resolution. This work presents measurements with the improved version of the NMX detector prototype consisting of a Triple-GEM (Gas Electron Multiplier) detector with a natural Gd converter and a low material budget readout. The detector was successfully tested at the neutron reactor of the Budapest Neutron Centre (BNC) and the D16 instrument at the Institut Laue-Langevin (ILL) in Grenoble. The measurements with Cadmium and Gadolinium masks in Budapest demonstrate that the point-spread function of the detector lacks long tails that could impede the measurement of diffraction spot intensities. On the D16 instrument at ILL, diffraction spots from Triose phosphate isomerase w/ 2-phosphoglycolate (PGA) inhibitor were measured both in the MILAND Helium-3 detector and the Gd-GEM. The comparison between the two detectors shows a similar point-spread function in both detectors, and the expected efficiency ratio compared to the Helium-3 detector. Both measurements together thus give good indications that the Gd-GEM detector fits the requirements for the NMX instrument at ESS.

Proliferation resistance of light water reactor versus pebble bed reactor

Intrinsic proliferation resistance (PR) against the weaponization of U and Pu in used nuclear fuel discharged from a 500 MWt pebble bed reactor (PBR) and an integral pressurized water reactor (iPWR) was assessed. Four inherent characteristics (spontaneous fission neutron emission, heat load, radiation exposure, and Rossi-α) of U and Pu in the used fuel were the basis for the PR evaluation. These four characteristics for a significant quantity of U (75 kg of 235U) and Pu (8 kg) metallic sphere made from used fuel were quantified by performing reactor neutronics simulations. The intrinsic PR values for PBR used fuel were found to be lower compared to the iPWR for all the cases analyzed due to the higher quantity and better quality of nuclear material. Online refueling with an effective safeguards approach and a better choice of PBR fuel composition are needed to increase the PR of its used fuel.

Development of an interatomic potential for mixed uranium-americium oxides and application to the determination of the structural and thermodynamic properties of (U,Am)O2 with americium contents below 50%

Uranium-americium mixed oxides are envisaged as targets in fast neutron reactors to allow the transmutation of americium in view of the reduction of its quantity in nuclear waste. The knowledge of structural and thermodynamic properties of (U,Am)O2 fuels is of crucial importance for fuel manufacturing, as well as for safe reactor operation. Atomic scale modelling has been proved an important complement to experimental characterizations to obtain the necessary data and bring insight into the behaviour of fuel materials over a wide range of compositions and conditions. An important feature of (U,Am)O2 compounds observed experimentally by several authors and confirmed by electronic structure calculations for all compositions, is the presence, not only of cations in the (+IV) oxidation state, but also of the Am3+ and U5+ species, inducing a complex behaviour as a function of composition. To allow the study of these mixed oxides using classical molecular dynamics, we optimized interaction parameters for the Am3+ and U5+ cations for the Cooper-Rushton-Grimes interatomic potential, which has yielded excellent results on several pure and mixed actinide oxides. These potential terms were parametrized on a small set of the experimental data available, namely the room temperature lattice parameters, the thermal expansion and the enthalpy increment for several americium contents. The potential obtained was validated against experimental data not included in the parametrization process and then applied to determine various structural and thermodynamic properties of U1-yAmyO2 compounds with 0≤y≤0.5 as a function of temperature and composition. The results yielded by this parametrization are in excellent agreement with the available experimental data and provide significant insight into the properties of uranium-americium mixed oxides, especially at high temperature.

Space-Dependent Calculation of the Multiplicity Moments for Shells With the Inclusion of Scattering

In recent work, we extended the methodology of multiplicity counting in nuclear safeguards by elaborating the one-speed stochastic transport theory of the calculation of the so-called multiplicity moments, i.e., the factorial moments of the number of neutrons emitted from a fissile item, following a source event from an internal neutron source [spontaneous fission and () reactions]. Calculations were made for solid spheres and cylinders, with the source being homogeneously distributed within the item. Recent measurements of the Rocky Flats Shells during the Measurement of Uranium Subcritical and Critical (MUSIC) campaign conducted by Los Alamos National Laboratory and assisted by the University of Michigan inspired us to extend the model to spherical shell geometry with a point source in the middle of the central cavity. Comparison of the calculated results with the experimental ones indicated that accounting for fission as the only neutron reaction (the standard procedure in the point model, adapted also in our work so far) was not sufficient for reaching good agreement with measurements. The model was therefore extended to include elastic scattering into the one-speed formalism, whereas the effect of inelastic scattering was accounted for in an empirical way. After these extensions, good agreement was found between the calculated and the measured values. The paper describes the extension of the theory and provides concrete quantitative results.

Development and verification of a Monte Carlo two-step method for lead-based fast reactor neutronics analysis

With the rise of economic and safety standards for nuclear reactors, new concepts of Gen-IV reactors and modular reactors showed more complex designs that challenge current tools for reactor physics analysis. A Monte Carlo (MC) two-step method was proposed in this work. This calculation scheme uses the continuous-energy MC method to generate multi-group cross-sections from heterogeneous models. The multi-group MC method, which can adapt locally-heterogeneous models, is used in the core calculation step. This calculation scheme is verified using a Gen-IV modular lead-based fast reactor (LFR) benchmark case. The influence of homogenized patterns, scatter approximations, flux separable approximation, and local heterogeneity in core calculation on simulation results are investigated. Results showed that the cross-sections generated using the 3D assembly model with a locally heterogeneous representation of control rods lead to an accurate estimation with less than 270 pcm bias in core reactivity, 0.5% bias in control rod worth, and 1.5% bias on power distribution. The study verified the applicability of multi-group cross-sections generated with the MC method for LFR analysis. The study also proved the feasibility of multi-group MC in core calculation with local heterogeneity, which saves 85% time compared to the continuous-energy MC.

Effect of reprocessing on neutrons of a molten chloride salt fast reactor

Due to their unique features, such as the inherent safety, simplified fuel cycle, and continuous on-line reprocessing, molten salt reactors (MSRs) are regarded as one of the six reference reactors in the Generation IV International Forum (GEN-IV). Molten chloride salt fast reactors (MCFRs) are a type of MSR. Compared to molten fluoride salt reactors (MFSRs), MCFRs have a higher solubility of heavy metal atoms, a harder neutron spectrum, lower accumulation of fission products (FPs), and better breeding and transmutation performance. Thus, MCFRs have been recognized as a type of MSR with great prospects for future development. However, as the most important feature for MSRs, the effect of different reprocessing modes on MCFRs must be researched in depth. As such, this study investigated the effect of different isotopes, especially FPs, on the neutronic performance of an MCFR, such as its breeding performance. Furthermore, the characteristics of the different reprocessing modes and MCFR rates were analyzed in terms of safety, radioactivity level, neutron economy, and breeding capacity. In the end, a reprocessing method suitable for MCFRs was determined through calculation and analysis, which provides a reference for the further research of MCFRs.

Neutronic simulation of medical radioisotope 99Mo and 177Lu production in IPR 14 MeV neutron generator facility

An accelerator based 14 MeV neutron generator is commissioned at Institute for Plasma Research India. The generator is based on the linear accelerator concept where the deuterium ion beam impinged to the tritium target to produce neutrons. The generator is designed to produce 5 × 1012 n/s (max.). The 14 MeV neutron source facilities are an emerging tool for the lab scale experiments and research. In order to utilize the generator for the welfare of humanity, the assessment is made for the production of medical radioisotopes using the neutron facility.The usage of radioisotopes in the treatment and diagnosis of a disease is an important factor in the healthcare sector. A series of calculations are conducted to generate radioisotopes, especially 99Mo and 177Lu those are having huge applications in the medical and pharmaceutical industries. 99Mo can be also generated through neutron reactions 98Mo(n, g)99Mo and 100Mo(n, 2n)99Mo apart from fission reaction. The cross section of 98Mo(n, g)99Mo is high in the thermal energy range whereas 100Mo(n,2n)99Mo occurs at a high energy range. 177Lu can be produced using the reactions 176Lu (n, g)177Lu and 176Yb (n, g)177Yb. The cross section of both 177Lu production routes is higher at thermal energy range. The neutron flux level near the target is around 1010 cm−2s−1. In order to enhance production capabilities, the neutron energy spectrum moderators are used to thermalize the neutrons. The materials used as a moderator are beryllium, HDPE, graphite, etc. Moderators enhance the capabilities of medical isotope production in neutron generators.

Finite element analysis of the main reactor vessel in the China Initiative Accelerator Driven System

The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The reactor in CiADS is a subcritical fast neutron reactor cooled by a liquid lead–bismuth eutectic. The reactor operates at high temperature and bears high thermal stress. In addition to the heavy weight of the whole reactor, the vessel will bear large effective stress. If the effective stress exceeds the limit of the material, defects and cracks may occur on the main reactor vessel, which will affect the safety performances of the reactor. Therefore, it is very important to analyze the effective stress field of the reactor vessel. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS. We can preliminarily prove that the maximum effective stress that the vessel will bear during the postulated Unprotected Loss of Flow (ULOF) and Unprotected Transient over Power (UTOP) accidents is less than the yield strength of 316L stainless steel. Therefore, we can preliminarily conclude that the current material selection and structural design of the CiADS vessel could survive the postulated transient accidents considering the effective stress effect.

The impact of TP53 status of tumor cells including the type and the concentration of administered 10B delivery agents on compound biological effectiveness in boron neutron capture therapy.

Human head and neck squamous cell carcinoma cells transfected with mutant TP53 (SAS/mp53) or neo vector (SAS/neo) were inoculated subcutaneously into left hind legs of nude mice. After the subcutaneous administration of a 10B-carrier, boronophenylalanine-10B (BPA) or sodium mercaptododecaborate-10B (BSH), at two separate concentrations, the 10B concentrations in tumors were measured using γ-ray spectrometry. The tumor-bearing mice received 5-bromo-2'-deoxyuridine (BrdU) continuously to label all intratumor proliferating (P) tumor cells, then were administered with BPA or BSH. Subsequently, the tumors were irradiated with reactor neutron beams during the time of which 10B concentrations were kept at levels similar to each other. Following irradiation, cells from some tumors were isolated and incubated with a cytokinesis blocker. The responses of BrdU-unlabeled quiescent (Q) and total (= P + Q) tumor cells were assessed based on the frequencies of micronucleation using immunofluorescence staining for BrdU. In both SAS/neo and SAS/mp53 tumors, the compound biological effectiveness (CBE) values were higher in Q cells and in the use of BPA than total cells and BSH, respectively. The higher the administered concentrations were, the smaller the CBE values became, with a clearer tendency in SAS/neo tumors and the use of BPA than in SAS/mp53 tumors and BSH, respectively. The values for BPA that delivers into solid tumors more dependently on uptake capacity of tumor cells than BSH became more alterable. Tumor micro-environmental heterogeneity might partially influence on the CBE value. The CBE value can be regarded as one of the indices showing the level of intratumor heterogeneity.

Prospects for the development of small nuclear power plants in Russia

This article reviews the main properties of the modular design of small nuclear power plants (SNPPs), shows the possibility of reducing the cost and construction time of this class of plants through factory production, the effect of serialization, and the reduction of redundant safety systems. The prospects for significant expansion of the field of application of nuclear technologies due to their modularity and the possibility to ensure high safety performance are presented. Possible areas of SNPP use for power supply of remote territories including the Arctic area, production of high-potential heat and hydrogen for industrial consumers, and other applications are analyzed. It shows the need to develop and implement a new nuclear energy technology platform based on SNPPs to solve the problems of global decarbonization of the world economy by significantly expanding the scope of nuclear energy technologies in addition to the currently developed technological platform of a closed nuclear fuel cycle with fast neutron reactors and the technological platform of controlled thermonuclear fusion. The authors propose to create a pilot site for testing technologies for captive production of hydrogen (heat) for an industrial consumer, as well as other technologies for the utilitarian use of SNPPs based on a pilot demonstration of a nuclear power plant with a high-temperature (about 1100C) fast neutron reactor with a gas coolant.

Development of oxide dispersion strengthened ferritic steel with enhanced strength–ductility combination through dispersion strengthening and back stress hardening

Oxide dispersion strengthened (ODS) ferritic steel is a potential structural material for fast neutron reactor and fusion reactor application. In this study, a new Fe–13Cr–2W-0.4Ti-0.6Zr/0.35Y2O3 ODS ferritic steel was designed and fabricated through hot isostatic pressing. Results show that the ODS ferritic steel consists of highly dispersive nano-oxide particles, identified as Y2(Ti0.4Zr0.6)2O7 and Y4Zr3O12, with average particle size of 5.34 nm and particle distribution density of 4.2 × 1023 m−3. In addition, heterogeneous coarse/ultrafine grains (CGs/UFGs) structure is also achieved in the matrix. The Fe–13Cr–2W-0.4Ti-0.6Zr/0.35Y2O3 ODS ferritic steel exhibits superior mechanical properties at both room temperature (σb = 1208 MPa, δ = 14%) and high temperature (σb, 700°C = 382 MPa, δ700°C = 32%). The high strength is mainly due to the synergistic effect of particle dispersion strengthening and back stress strengthening, while the good ductility can be attributed to the excellent deformation ability of the coarse grains and the back stress hardening effect induced by the heterogeneous CGs/UFGs structure.

Modeling of Dynamic Operation Modes of IVG.1M Reactor

This paper presents the results of a calculation code approach providing a solution to the point kinetics problem for the IVG.1M research reactor of the National Nuclear Center of the Republic of Kazakhstan and allowing the simulation of dynamic processes going on during reactor start-ups, including changes in the thermal state of all its elements, reactor regulator displacement, accumulation of absorbers in the fuel, and the beryllium reflector. A mathematical description of the IVG.1M point kinetics model is presented, which provides a calculation of the reactor neutron parameters, taking into account the dependence of reactivity effects on the temperature, changes in the isotopic composition of materials, and thermal expansion of core structural elements. An array of data values was formed of reactivity added by separate elements of the core when changing their thermal state and other reactor parameters, as well as an array of data with the parameters of heat exchange of coolant-based reactor structural elements. These are used in the process of solving the point kinetics problem to directly replace formal parameters, eliminating the need to calculate the values of these parameters at each calculation step. Preliminary calculations to form an array of values of reactivity effects was applied to the reactor by separate structural elements when their temperature changes were performed using the IVG.1M precision reactor calculation model. The model was validated by the reactor parameters in the critical state. Preliminary calculations to form an array of data with the parameters of heat exchange of coolant-based reactor structural elements were performed in ANSYS Fluent software using the calculation model that describes the IVG.1M reactor fuel element in detail. Validation of the developed calculation code based on the results of two start-ups of the IVG.1M reactor was performed and its applicability for the analysis of transient and emergency modes of reactor operation and evaluation of its safe operation limits was confirmed.

Application of the improved quasi-static coupled probability theory method to transient safety analysis of an experimental facility

The accelerator-driven sub-critical system is driven by external neutron sources, which are generated by the spallation reaction and maintain the stable operation of the sub-critical core, while the external neutron source increases the complexity of reactor neutron kinetic processes. This article focuses on simulation analysis of neutron space–time kinetics with a dynamic analysis code, which is developed based on the improved quasi-static approximation and Monte Carlo neutron transport method. The amplitude function and shape functions are solved to achieve the dynamics simulation process. Then, the transient responses of beam interruptions and reactivity insertions are calculated and analyzed for an experimental facility. To further verify the correctness and reliability of dynamic behavior, the simulation results are compared with experimental values, and the results show that the normalized neutron fluxes varying with time are in good agreement with the corresponding values. It can be concluded that the improved quasi-static coupled probability theory method can be used to solve the neutron space–time kinetic problem of the experimental facility, and the results are reliable.

Study on the calculational framework development of the advanced numerical reactor neutronics code SHARK

The SHARK program (Simulation-based High-fidelity Advanced Reactor physics Kit) is a high-fidelity heterogeneous neutronics code for the numerical reactor system being developed at the Nuclear Power Institute of China (NPIC). The program uses a Constructive Solid Geometry (CSG) framework to model various complex geometries. To enhance the flexibility and robustness during continuous development process, SHARK program attempts to support a rich set of methods, tools and library options within a unified general framework as a “toolkit”. For the multi-core clustered supercomputer architectures that are commonly used today, the SHARK program adopts a hybrid parallel strategy of MPI and OpenMP to achieve complementary advantages between them. In addition, the framework of SHARK program is designed with a true object-oriented manner. Through reasonable abstraction, inheritance and encapsulation, the maintainability and extensibility of the code are improved, and long-term team development is facilitated. Up to now, key modules for cross-section generation, heterogeneous transport calculation and microscopic depletion have been developed under the general frameworks. The main features of SHARK’s “resonance-transport-depletion” coupling system are elaborated in this paper, and some verification and validation (V&V) results in the current phase are presented and discussed.

New Neutron Imaging Facility NIFFLER at Very Low Power Reactor VR-1.

The paper describes the construction of the neutron imaging facility at the very low-power research reactor VR-1. The training reactor VR-1 is operated by the Czech Technical University in Prague, Czech Republic. It is mainly used for the education of students in the field of nuclear engineering as well as for the training of professionals. Neutron imaging is the new field of VR-1 reactor utilisation currently under development. Extremely low reactor power at the level of 100 W brought many challenges that were necessary to overcome to build and commission a sustainable neutron radiography facility. The paper describes the reactor's neutron flux verification and the basic concept and design of the neutron imaging instrumentation. The first experimental results were mainly dedicated to testing the detection system for different radial beam port configurations, different L/D ratios, and different exposure times. Preliminary results of neutron radiography and tomography measurements at VR-1 clearly showed the potential of using neutron imaging in low-power reactors such as the VR-1 reactor.

High deuteron and neutron yields from the interaction of a petawatt laser with a cryogenic deuterium jet

A compact high-flux, short-pulse neutron source would have applications from nuclear astrophysics to cancer therapy. Laser-driven neutron sources can achieve fluxes much higher than spallation and reactor neutron sources by reducing the volume and time in which the neutron-producing reactions occur by orders of magnitude. We report progress towards an efficient laser-driven neutron source in experiments with a cryogenic deuterium jet on the Texas Petawatt laser. Neutrons were produced both by laser-accelerated multi-MeV deuterons colliding with Be and mixed metallic catchers and by d (d,n)3He fusion reactions within the jet. We observed deuteron yields of 1013/shot in quasi-Maxwellian distributions carrying ∼8−10% of the input laser energy. We obtained neutron yields greater than 1010/shot and found indications of a deuteron-deuteron fusion neutron source with high peak flux (>1022 cm−2 s−1). The estimated fusion neutron yield in our experiment is one order of magnitude higher than any previous laser-induced dd fusion reaction. Though many technical challenges will have to be overcome to convert this proof-of-principle experiment into a consistent ultra-high flux neutron source, the neutron fluxes achieved here suggest laser-driven neutron sources can support laboratory study of the rapid neutron-capture process, which is otherwise thought to occur only in astrophysical sites such as core-collapse supernova, and binary neutron star mergers.