Neutron Cross Sections Research Articles

Transmutation of long-lived radwaste in fusion driven subcritical spent fuel burner reactor

In this study, Fusion Driven Subcritical Spent Fuel Burner Reactor (FDS-SFB), designed by the Frontier Development of Science Team (FDS Team) in the first decade of this century, has been neutronically modeled to compare the effect of using the thorium fuel cycle, instead of the uranium fuel cycle, on the transmutation of long-lived hazardous radwaste (including 241Am, 243Am, 244Cm, 237Np, 238Pu, 241Pu, 242Pu). Furthermore, the neutronic spectrum of the different parts of the FDS-SFB, and the variation of the effective multiplication factor (keff) of the system have been also investigated to study effects of changing the fuel cycle. For these purposes, SuperMC code has been applied for the simulations and performing the Monte Carlo and Monte Carlo-deterministic coupling calculations. ENDF/B-VII neutron cross-section library was used for applying needed cross-sections. The spent fuel used in the modeling had been discharged from a standard, 33,000 MWD/MTU, Pressurized-Water Reactor (PWR). After ten years of the FDS-SFB operation it was found that, for all of the loaded long-lived radwaste, the thorium fuel cycle had a much better efficiency in the transmutation process. The Monte Carlo-deterministic calculations indicated that the final mass differences between the transmuted radwaste i.e. 241Am, 243Am, 244Cm, 237Np, 238Pu, 241Pu, and 242Pu, in both uranium and thorium fuel cycles were 582.3 kg, 51.2 kg, 48.2 kg, 611.6 kg, 427.2 kg, 1320.3 kg, and 357.4 kg, respectively. Considering the total amount of the radwaste loaded in the FDS-SFB, the thorium fuel cycle, with a 3398 kg difference, had considerably better efficiency than the uranium fuel cycle in the incineration of the loaded long-lived radwaste.

Development of a High-Efficiency Device for Thermal Neutron Detection Using a Sandwich of Two High-Purity 10B Enriched Layers.

The shortage of 3He, a crucial element widely used as a neutron converter in neutron detection applications, has sparked significant research efforts aimed at finding alternative materials, developing appropriate deposition methods, and exploring new detector architectures. This issue has required the exploration of novel approaches to address the challenges faced in neutron detection. Among the available conversion materials, 10B has emerged as one of the most promising choices due to its high neutron-capture cross-section and relatively high Q value. In our previous papers, we delved into the possibility of depositing neutron conversion layers based on 10B using Pulsed Laser Deposition (PLD). We investigated and evaluated the performance of these layers based on various factors, including deposition conditions, substrate properties, and film thickness. Moreover, we successfully developed and tested a device that employed a single conversion layer coupled with a silicon particle detector. In this current study, we present the development of a new device that showcases improved performance in terms of efficiency, sensitivity, and discrimination against γ background signals. The background signals can arise from the environment or be associated with the neutron field. To achieve these advancements, we considered a new detection geometry that incorporates the simultaneous use of two 10B conversion layers, each with a thickness of 1.5 μm, along with two solid-state silicon detectors. The primary objective of this design was to enhance the overall detection efficiency when compared to the single-layer geometry. By employing this novel setup, our results demonstrate a significant enhancement in the device's performance when exposed to a neutron flux from an Am-Be neutron source, emitting a flux of approximately 2.2 × 106 neutrons per second. Furthermore, we established a noteworthy agreement between the experimental data obtained and the simulation results.

The influence of Monte Carlo radiation transport codes and hadron physics models on stray neutron dose estimations in proton therapy: An extended intercomparison

PurposeSelecting a Monte Carlo (MC) radiation transport code and nuclear physics model affects the simulation of secondary neutron fluence energy spectra and the estimation of the neutron dose equivalent in proton therapy. This study compares the MC codes FLUKA, MCNPX and GEANT4 from previous work with two additional MC codes, TOPAS (with either the Binary Intra-nuclear Cascade (BIC) or the Bertini intra-nuclear cascade model (Bert)) and PHITS (with either the Intra-Nuclear Cascade of Liege (INCL) or the Bert model), to facilitate a more comprehensive comparison. Material and methodsSecondary neutron fluence energy spectra were scored at different locations inside a water phantom, exposed to a 10 × 10 cm2 parallel mono-energetic proton beam with energies of 110 MeV, 150 MeV, 180 MeV and 210 MeV. The neutron dose equivalent was estimated by applying fluence-to-dose equivalent conversion coefficients. ResultsDiscrepancies between MC codes and physics models reach 90 % for the total neutron fluence. Differences are most pronounced in the forward direction of the beam due to a notable presence of high-energy neutrons and the higher uncertainty on the neutron cross-section data for these neutrons. Nuclear model selection within PHITS (INCL or Bert) and TOPAS (BIC or Bert) shows a worse agreement for the Bert model than for the other MC codes and nuclear models. The discrepancies in the neutron spectrum simulations propagate into differences in the estimated neutron dose equivalents. They are more prominent for distal positions (up to a factor of 3) than for lateral positions (up to a factor of 2) and decrease at higher proton energies. ConclusionsThis study shows considerable differences in the out-of-field neutron fluence energy spectra calculated by different MC codes and nuclear physics models. Differences in the neutron dose equivalent reach a factor of 3. The discrepancies are most apparent in distal positions and at lower proton energies.

Methodologies for 176Lu-176Hf Analysis of Zircon Grains from the Moon and Beyond.

Zircons are found in extraterrestrial rocks from the Moon, Mars, and some differentiated meteorite parent-bodies. These zircons are rare, often of small size, and have been affected by neutron capture induced by cosmic ray exposure. The application of the 176Lu-176Hf decay system to zircons from planetary bodies such as the Moon can help establish the chronology of large-scale differentiation processes such as the crystallization of the lunar magma ocean. Here, we present methods to measure the isotopic composition of Hf of extraterrestrial zircons dated using ID-TIMS U-Pb after chemical abrasion. We introduce a 2-stage elution scheme to separate Hf from Zr while preserving the unused Zr fraction for future isotopic analysis. The effect of neutron capture is also re-examined using the latest thermal neutron capture cross sections and epithermal resonance integrals. Our tests show that the precision of Hf isotopic analyses is close to what is theoretically attainable. We have tested this method to a limited set of zircon grains from lunar rocks returned by the Apollo missions (lunar soil 14163, fragmental polymict breccia 72275, and clast-rich breccia 14321). The model ages align with previously reported values, but further work is needed to assess the chronology of lunar magma ocean crystallization as only a handful of small zircons (5 zircons from 3 samples) were analyzed, and the precision of the analyses can be improved by measuring more and larger lunar zircon grains.

Recent Research on SiC/SiC Composites for PWR

Silicon carbide (SiC) continuous fiber reinforced SiC matrix composites (SiC/SiC) are considered as promising candidates for pressurized water reactor (PWR) due to their small neutron capture cross-section, excellent high temperature strength and irradiation resistance, etc. SPICRI have been working on the R&D of SiC/SiC composites cladding tubes for the last decade. This report will show the recent research in material development, properties and irradiation response of SiC/SiC composites. Several critical issues to be solved for PWR fuel cladding applications, such as mechanical properties, hermeticity, thermal properties and irradiation resistance are discussed here.

Development and Baseline Comparison of a New Pulsed-Neutron Spectroscopy Tool for Carbon- Oxygen Analysis and Three-Phase Saturation Monitoring

A new slim multidetector pulsed-neutron wireline-logging tool has been developed for openhole or casedhole formation evaluation saturation analysis and time-lapse monitoring. With a greater neutron source output and high-spectral resolution gamma ray detectors, the tool can be operated with reduced uncertainty or faster logging speeds. New, fully programable digital electronics provide a range of acquisition modes optimized for specific formation evaluation objectives. Neutrons are generated from a high-output pulsed-neutron generator, which propagates radially outward, passing through the borehole, completion material, and into the formation. Energy is lost through scattering, and neutrons are absorbed by the surrounding material. Gamma rays are emitted from scattering and absorption interactions at discrete energies, which can be measured by one of three spectral gamma scintillation detectors. The energy distribution of these incident gamma rays by inelastic and capture interactions is affected by the elemental composition of the material. A salinity-independent oil-water saturation assessment begins with the deconvolution of neutron-induced gamma ray inelastic spectra into constituent elemental components. Carbon-oxygen ratios (C/O) of elemental yields are then compared to a reference model to determine the relative saturation and porosity-filled volume of oil and water. In PNC (pulsed-neutron capture) acquisition mode, a salinity-dependent oil-water saturation assessment is determined from neutron capture cross section (sigma) measurements, which are compared to formation and porosity fluids using a mass balance approach. Gas saturation assessment is determined from gas-sensitive inelastic (RIN13) and capture (RATO13) ratios. Gas ratios are compared to a reference model to assess the relative saturation of fluids, typically distinguishing gas from water or gas from oil. Gas saturation is not limited to hydrocarbon components and can also be used for saturation analysis of H2, He, CO2, N2, and other non-hydrocarbon-bearing components. A new Omni-mode acquisition combining simultaneous PNC and C/O measurements was also developed. This acquisition mode provides advantages in reducing multipass logging typically required from separate PNC and C/O acquisitions. By including PNC neutron capture sigma, gas-sensitive neutron capture and inelastic measurements, and C/O inelastic measurements, the simultaneous Omni-mode (PNC+C/O) acquisition is specifically optimized for three-phase saturation analysis applications, including baseline CO2 sequestration evaluation for carbon capture, utilization, and storage (CCUS) CO2 and steamflood time-lapse monitoring. Results from a field example are presented to demonstrate the new technology with spectral C/O saturation analysis compared to traditional windows C/O analysis and to compare the performance of the next-generation tool to the legacy tool. Multiphase saturations from C/O and RIN13 measurements and compatibility with legacy interpretations for time-lapse saturation monitoring are also presented.

Measurement of fast neutron induced (n,γ) reaction cross-section of 152Sm, 154Sm and 150Nd in the energy range of 0.8 to 2 MeV

The (n,γ) reaction cross-section of rare earth stable fission products, such as 152Sm, 154Sm and 150Nd, in the neutron energy range of 0.8 to 2 MeV, obtained from 7Li(p,n)7Be reaction, was measured by neutron activation followed by off-line gamma-ray spectrometry. Natural strontium (natSr) has been established as a fast neutron flux monitor in this work, by measuring the natSr(n,γ+n,n′)87Srm reaction cross-section using 127I(n,γ)128I as monitor reaction. The cross-section for 86Sr(n,γ)87Srm has been determined for the first time to the best of our knowledge. The excellent decay properties of 87Srm and opposite energy dependence of 86Sr(n,γ)87Srm and 87Sr(n,n′)87Srm reactions makes natSr an excellent candidate for neutron flux monitor for wide neutron energy range. The measured (n,γ) cross-section data from the present study has been compared with the literature data, which is very limited along with large scatter in the energy range of 1–2 MeV for these nuclei. The results from the present study, clearly show a need for fresh cross-section evaluations for 152Sm(n,γ)153Sm. For, 154Sm(n,γ)155Sm and 150Nd(n,γ)151Nd reactions, the present data align with one of the data groups among the two which can be broadly identified having difference of about more than ∼50%. The present experimental data has been compared with TALYS 1.96 theoretical model calculations based on Fermi gas level density prescriptions which shows that the cross section values are close to the experimental data at some of the neutron energies. A comparison of the data on fast neutron capture cross-section for N = 90 isotones from the present study and from the literature showed a systematic decrease in the cross-section value with increasing neutron richness of the target nucleus. The cross-section data could be scaled with neutron-proton asymmetry parameter to show a nearly linear behaviour on a logarithmic scale. The data for 154Sm (N = 92) was also observed to follow the same trend.

Cosmic implications of the reaction: a nuclear astrophysical perspective

We conducted a comprehensive study of the neutron capture cross section of within the relevant astrophysical energy range. Through rigorous R-matrix analysis, we determined the Maxwellian-averaged cross section to be 11.98 0.25 μb at kT = 30 keV. This result is approximately four times higher than the thermal neutron capture cross section estimated in earlier studies assuming the law. The implications of our findings extend to the region of inhomogeneous Big-Bang models in nuclear astrophysics, where understanding the behaviour of neutron capture cross sections plays a crucial role in elucidating the intricate processes that shaped the early universe.

Fabrication and metrology of SiC cladding and seals for containment and heat transfer applications

Background SiC is a candidate material for high-temperature nuclear applications due to its high corrosion resistance, thermomechanical strength, and low neutron cross-section. Traditional manufacturing methods include sintering to a near-net shape with post-sinter finish machining. Methods Analysis of the metrology from fabricated SiC casings was conducted. SiC was fabricated using spark plasma sintering (SPS) under constant temperature and pressure (~2,000K; ~10MPa). Samples were then sandblasted to remove excess graphite and underwent metrological evaluation. Results Metrology indicates that dimensions may not be sufficiently predictable for net-shape fabrications. The process of mass dosing, human input, and sinter-forging lead to standard deviations of up to ~7%, suggesting that alternative tooling approaches are required to decrease the amount, and subsequent cost, of post-sinter machining. Dimensional compatibility trials using surrogate pellets show that machining is required. Conclusions The consequences of poor tolerances are discussed with respect to theoretical applications. This is an expected outcome for a ceramic fabrication process.

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

The accumulation of spent nuclear fuel (SNF) is presently considered as a hindrance of the massive deployment of nuclear power plant, especially regarding the transuranic (TRU) elements. Eliminating TRU through transmutation is one of the most feasible alternative as a technical solution to solve the issue. This study explores the possibility of TRU transmutation using molten salt reactor (MSR) in a heterogeneous configuration, where a solid TRU is coated inside the fuel channel filled with liquid salt fuel. Such configuration is proposed to allow higher TRU loading into fluoride salt mixture without compromising the safety of the reactor. TRU coating was applied in consecutively outward radial fuel channel layers with coating thicknesses of 2.5 mm and 5 mm. Calculation was performed using MCNP6.2 radiation transport code and ENDF/B-VII.0 neutron cross section library. From the results, TRU coating with smaller thickness and positioned closer to the centre of the core exhibit higher transmutation efficiency due to exposure to higher neutron flux. Highest transmutation efficiency was achieved at 67.93% after 160 days of burnup. This shows a potential of achieving highly efficient TRU using heterogeneous configuration in MSR core.

Measurement of cross section for the 232Th(n, f) reaction using a time projection chamber

Accurate cross sections of neutron induced fission reactions are needed in the development of nuclear science and technology. The time projection chamber (TPC) is viewed as the potential detector to improve the measurement accuracy of the fission cross sections based on its track reconstruction and particle identification capabilities. The TPC developed by the back-streaming white neutron source (Back-n) team of China Spallation Neutron Source (CSNS) was introduced to verify its ability in measuring the cross section of fission reaction. Using our TPC, and based on the mono-energetic neutron source at Peking University and 232Th/238U samples, we have measured the cross section of the 232Th(n, f) reaction at 5.0 MeV. The fission events have been well distinguished from the background alpha events. The present cross section is consistent with the evaluation data from different libraries, which indicates that the fission cross section measurement using our TPC is feasible. By reducing the distance between the two fission samples, the large uncertainty of the present result will be reduced in future experiments.

The influence of graphite moderator density on the neutronic performance and fuel economy of a Molten Salt Reactor

Typical thermal molten salt reactor (MSR) employs graphite as its neutron moderator, due to its chemical compatibility and low neutron absorption. It is not without issues, however, since graphite provides positive temperature feedback when the MSR is fueled with U-233. Reducing the graphite density can possibly ameliorate this issue, but at the cost of other neutronic parameters, such as fertile conversion into fissile nuclide, and fuel economy, such as fissile makeup cost. This study evaluates the influence of graphite moderator density on the neutronic performance and fuel economy of an MSR. Passive Compact Molten Salt Reactor (PCMSR) design was used as the MSR model. To evaluate the neutronic parameters, MCNP6.2 radiation transport code with ENDF/B-VII.0 neutron cross section library were used. The evaluated neutronic parameters were temperature coefficient of reactivity (TCR), kinetic parameters, and conversion ratio (CR). Meanwhile, the fuel economy evaluation comprised the initial fuel inventory cost and fissile makeup cost. From the calculation results, lowering graphite density remarkably improves TCR, but also decreased CR in a notable amount. While effective delayed neutron fraction (βeff) was not particularly affected, mean neutron generation time (Λ) was reduced in lower graphite density. Considering the fuel economy, initial fuel cost is not significantly affected, but the fissile makeup cost can be sensitive of graphite density.

The multigroup library processing method for coupled neutron and photon heating calculation of fast reactor

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

Systematics of Maxwellian-averaged neutron capture cross-sections (MACS) at 30keV

A formula for Maxwellian-Averaged Cross-Sections (MACS) at 30[Formula: see text]keV is derived. Based on the presented formula, a linear relationship was found between the log [Formula: see text] and [Formula: see text] of the target nucleus. From this relationship, a good agreement between the calculated MACS and the experimental data can be observed. It seems therefore possible to use this systematic law in order to analyze the experimental data. It would be useful in estimating the MACS of neighboring isotopes for which no experimental data are available.

Obtaining Boron Carbide and Nitride Matrix Nanocomposites for Neutron-Shielding and Therapy Applications

The very high capture cross-section of (epi)thermal neutrons by the boron isotope 10B makes elemental boron and its compounds and composites prospective for serving as materials intensively interacting with neutron irradiation. In their nanostructured form, boron-rich materials reveal properties that improve their radiation-performance characteristics. In this regard, new technologies have been proposed for the synthesis of nanocomposites with matrices of boron carbide B4C and hexagonal boron nitride h-BN. For the first time, boron carbide-tungsten and hexagonal boron nitride–(iron,magnetite) composites were obtained, respectively, in the form of layered/sandwich structures of components B4C and W and h-BN nanopowders coated/intercalated with magnetic nanoclusters of iron Fe or magnetite Fe3O4. Studying of their chemical/phase composition, structure/morphology, and some other properties leads to the conclusion that the developed B4C–W and h-BN–(Fe,Fe3O4) composites would be useful for solving important problems of boron-based neutron shielding and BNCT (Boron Neutron Capture Therapy), such as attenuating the gamma-radiation accompanying the absorption of neutrons by 10B nuclei and targeted delivery of 10B nuclei, as BNCT therapeutic agents, to tumor tissues using control by an external magnetic field, respectively.

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.

Evaluation of Neutron Cross-Section Data Using Gaussian Mixture Model and Digital Filter

For the accurate estimation of neutron cross-section data, evaluation of nuclear data is mandatory to fulfill the need of nuclear science and technology. In this work, the evaluation of 232Th(n,2n)231Th, 241Am(n,2n)240Am, 100Mo(n,2n)99Mo, and 96Mo(n,p)96Nb reaction cross-section data is carried out using a novel method. This novel method of evaluation employs a cluster-based piecewise evaluation followed by a digital filter for merging the evaluated curves. The clusters in the experimental data and model data are identified using the probabilistic method of the Gaussian Mixture Model. The clustered experimental data are then regressed using the polynomial regression technique. The model data are generated using the Talys 1.9 code, and the model deficiency due to the complex random nature of nuclear reactions is also accounted here using chi-squared analysis. Evaluation in each cluster is then carried out independently using the popular Kalman filter technique. For obtaining a single smooth evaluated curve for the whole energy range, the popular smoothing digital filter, the Savitzky-Golay Filter, is employed for the first time in nuclear data evaluation. The proposed evaluated curves and existing evaluated curves of 232Th(n,2n)231Th, 241Am(n,2n)240Am, 100Mo(n,2n)99Mo, and 96Mo(n,p)96Nb reactions from nuclear data libraries such as ENDF/BVIII.0, JEFF-3.3, JENDL-4.0, CENDL-3.1, and TENDL 2021 are compared and found to be in good agreement. It is also found that generally, evaluation methods are data dependent, and so, a single evaluation method may not be applicable for all reactions of all nuclides. Since piecewise evaluation is cluster dependent, selecting the appropriate cluster makes this method robust for almost all reactions of all nuclides. Also, it is proven that this novel method of evaluation is a promising method demonstrating the potential of this approach for evaluation based on the chi-squared goodness-of-fit test with respect to standard evaluated library ENDF/BVIII.0 and experimental data.

Estimation of the buildup of He-3 and Li-6 poisoning, He-4 and tritium activity in beryllium oxide reflector block in APSARA-U reactor

The Apsara-U research reactor, situated at the Bhabha Atomic Research Center, began its operation in 2018. This facility employs a fuel core configuration employing low-enriched uranium and reflector assemblies made of beryllium oxide. In neutron-rich environment of a nuclear reactor, these beryllium oxide reflector blocks participate in nuclear reactions identified as 9Be(n,α) and 9Be(n,2n). The 9Be(n,α) reaction initiates the formation of isotopes such as 4He, 6Li, 3He, and 3H in the process. The accumulation of gasses like 3He, 4He, and 3H causes the beryllium oxide reflector to swell, ultimately restricting how long it can effectively serve as a reflector or moderator material within the reactor. Extended periods of reactor shutdown result in a significant increase in the concentrations of 3He, mainly due to the radioactive decay of 3H. 6Li and 3He nuclei have significant neutron capture cross-sections, leading to a reduction in the reactor's reactivity, a phenomenon termed "poisoning". This effect varies depending on factors like neutron flux, neutron flux spectrum, and the reactor's operational history. Thorough investigations have been carried out to gain insights into the accumulation of 4He, 3He, 6Li, and 3H within beryllium oxide and their effects on reactivity considering the diverse levels of neutron exposure it encounters.

Development of a new Monte-Carlo neutron transport code for radiation shielding

A continuous energy Monte-Carlo method-based neutron transport code for fixed source in 3D solid geometry has been developed. The main objective is to integrate this code with recently developed Monte-Carlo gamma shielding code MCBLD, so as to build an in-house code for coupled neutron-gamma shielding study. Also it is intended to use of the new ENDF data library (ENDF/B-VIII) in this code. This paper gives an overview of the code in terms of processing neutron interaction data and numerical scheme to track scattered neutrons. Present capability of the code is demonstrated by presenting fluence due to mono-energetic neutron source as well as widely used spontaneous neutron source from 252Cf isotope, placed at the centre of different size iron spheres. Results are also generated from general purpose open-source Monte-Carlo code FLUKA. Overall good agreement between the two validates the numerical implementation of angle and energy dependent neutron cross-section data in this code.

Neutronic Chain Reactions for Polonium-210 Production

The production of the industrially significant radionuclide polonium-210 from the neutron irradiation of bismuth metal and the subsequent beta decay of bismuth-210 is highly inefficient due to the small neutron capture cross section of bismuth-209. In this paper, we report a previously undescribed self-sustaining nuclear chain reaction involving self-propagating neutron multiplication in bismuth salts that allow for rapid and cost-effective production of polonium-210. The reaction proceeds in a cycle of three alternating elementary steps – the capture of neutrons by bismuth-209 and the subsequent formation of polonium-210, the emission of high-energy alpha particles by polonium-210, and the production of more neutrons from (α, n) and (n,2n) reactions on light element and bismuth-209 nuclei respectively. Furthermore, the high hydrogen density of the compound also confers it intrinsic neutron moderation properties, increasing the neutron capture cross section of bismuth-209 at thermal neutron energies. The chain reaction was proven to have successfully occurred by irradiating a sample of the bismuth salt with a 80 μCi neutron source and monitoring the activity levels of the reaction. It was found that the activity of the reaction increased exponentially after an initial stable period following a derived formula for polonium production trends for the reaction, thus validating the occurrence of the reaction. Furthermore, alpha spectroscopy confirmed that polonium-210 had been produced by characterising the 5.30 MeV alpha emission peak of the reaction in addition to using beta spectroscopy to identify the parent nuclide bismuth-210, further proving that the reaction was successful. Hence, this paper reports the successful initiation and characterisation of a novel nuclear chain reaction, and its potential applications offered by a method of rapidly producing large quantities of polonium-210.