Source Reactor Research Articles

Development of a simple calculation tool of dose distributions in a phantom for boron neutron capture therapy

A simple dose calculation tool, SiDE, was developed for dose evaluation in a water phantom for boron neutron capture therapy, which makes the calculation time much shorter compared with the conventional particle transportation Monte Carlo codes and is applicable to any type of incident neutron spectra to the phantom. As the SiDE can not only calculate quantitatively the dose distribution in the phantom but also output dose indexes such as advantage depth and peak tumor dose, a comparison between different boron neutron capture therapy neutron sources can be easily performed. Consistency with a Monte Carlo transportation code was verified through comparison with the conventional dose calculation with the Particle and Heavy Ion Transport Code System, and the calculation time was nearly 1/90 in the SiDE. The dose distributions for a reactor and accelerator-based neutron sources were compared, and the differences were found to be small although large differences existed between the incident spectra.

SERPENT CODE POINT KINETIC PARAMETERS ANALYSIS: A NEXUS FOR NIGERIA RESEARCH REACTOR – 1 CORE CONVERSION USING ENRICHED URANIUM DIOXIDE FUEL

The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium

Search for sterile neutrino oscillations using RENO and NEOS data

We present a reactor model independent search for sterile neutrino oscillation using 2\,509\,days of RENO near detector data and 180 days of NEOS data. The reactor related systematic uncertainties are significantly suppressed as both detectors are located at the same reactor complex of Hanbit Nuclear Power Plant. The search is performed by electron antineutrino\,($\overline{\nu}_e$) disappearance between six reactors and two detectors with baselines of 294\,m\,(RENO) and 24\,m\,(NEOS). A spectral comparison of the NEOS prompt-energy spectrum with a no-oscillation prediction from the RENO measurement can explore reactor $\overline{\nu}_e$ oscillations to sterile neutrino. Based on the comparison, we obtain a 95\% C.L. excluded region of $0.1<|\Delta m_{41}^2|<7$\,eV$^2$. We also obtain a 68\% C.L. allowed region with the best fit of $|\Delta m_{41}^2|=2.41\,\pm\,0.03\,$\,eV$^2$ and $\sin^2 2\theta_{14}$=0.08$\,\pm\,$0.03 with a p-value of 8.2\%. Comparisons of obtained reactor antineutrino spectra at reactor sources are made among RENO, NEOS, and Daya Bay to find a possible spectral variation.

Bounds on new physics with data of the Dresden-II reactor experiment and COHERENT

Coherent elastic neutrino-nucleus scattering was first experimentally established five years ago by the COHERENT experiment using neutrinos from the spallation neutron source at Oak Ridge National Laboratory. The first evidence of observation of coherent elastic neutrino-nucleus scattering with reactor antineutrinos has now been reported by the Dresden-II reactor experiment, using a germanium detector. In this paper, we present constraints on a variety of beyond the Standard Model scenarios using the new Dresden-II data. In particular, we explore the constraints imposed on neutrino non-standard interactions, neutrino magnetic moments, and several models with light scalar or light vector mediators. We also quantify the impact of their combination with COHERENT (CsI and Ar) data. In doing so, we highlight the synergies between spallation neutron source and nuclear reactor experiments regarding beyond the Standard Model searches, as well as the advantages of combining data obtained with different nuclear targets. We also study the possible signal from beyond the Standard Model scenarios due to elastic scattering off electrons (which would pass selection cuts of the COHERENT CsI and the Dresden-II experiments) and find more stringent constraints in certain parts of the parameter space than those obtained considering coherent elastic neutrino-nucleus scattering.

Transmutation of long-lived fission products in an advanced nuclear energy system

Disposal of long-lived fission products (LLFPs) produced in reactors has been paid a lot attention for sustainable and clean nuclear energy. Although a few transmutation means have been proposed to address this issue, there are still scientific and/or engineering challenges to achieve efficient transmutation of LLFPs. In this study, we propose a novel concept of advanced nuclear energy system (ANES) for transmuting LLFPs efficiently without isotopic separation. The ANES comprises intense photoneutron source (PNS) and subcritical reactor, which consist of lead–bismuth (Pb-Bi) layer, beryllium (Be) layer, and fuel, LLFPs and shield assemblies. The PNS is produced by bombarding radioactive cesium and iodine target with a laser-Compton scattering (LCS) γ-ray beam. We investigate the effect of the ANES system layout on transmutation efficiency by Monte Carlo simulations. It is found that a proper combination of the Pb-Bi layer and the Be layer can increase the utilization efficiency of the PNS by a factor of ~ 10, which helps to decrease by almost the same factor the LCS γ-beam intensity required for driving the ANES. Supposing that the ANES operates over 20 years at a normal thermal power of 500 MWt, five LLFPs including 99Tc, 129I, 107Pd, 137Cs and 79Se could be transmuted by more than 30%. Their effective half-lives thus decrease drastically from ~ 106 to less than 102 years. It is suggested that this successful implementation of the ANES paves the avenue towards practical transmutation of LLFPs without isotopic separation.

Testing sterile neutrino mixing with present and future solar neutrino data

We investigate the sensitivity of solar neutrino data to mixing of sterile neutrinos with masses gtrsim eV. For current data, we perform a Feldman–Cousins analysis to derive a robust limit on the sterile neutrino mixing. The solar neutrino limit excludes significant regions of the parameter space relevant to hints from reactor and radioactive gallium source experiments. We then study the sensitivity of upcoming solar neutrino data, most notably elastic neutrino-electron scattering in the DARWIN and DUNE experiments as well as coherent neutrino-nucleus scattering in DARWIN. These high precision measurements will increase the sensitivity to sterile neutrino mixing by about a factor of 4.5 compared to present limits. As a by-product, we introduce a simplified solar neutrino analysis using only four data points: the low- and high-energy nu _e survival and transition probabilities. We show that this simplified analysis is in excellent agreement with a full solar neutrino analysis; it is very easy to handle numerically and can be applied to any new physics model in which the energy dependence of the nu _e transition probabilities is not significantly modified.

Statistical significance of the sterile-neutrino hypothesis in the context of reactor and gallium data

We evaluate the statistical significance of the 3+1 sterile-neutrino hypothesis using νe and overline{nu} e disappearance data from reactor, solar and gallium radioactive source experiments. Concerning the latter, we investigate the implications of the recent BEST results. For reactor data we focus on relative measurements independent of flux predictions. For the problem at hand, the usual χ2-approximation to hypothesis testing based on Wilks’ theorem has been shown in the literature to be inaccurate. We therefore present results based on Monte Carlo simulations, and find that this typically reduces the significance by roughly 1 σ with respect to the naïve expectation. We find no significant indication in favor of sterile-neutrino oscillations from reactor data. On the other hand, gallium data (dominated by the BEST result) show more than 5 σ of evidence supporting the sterile-neutrino hypothesis, favoring oscillation parameters in agreement with constraints from reactor data. This explanation is, however, in significant tension (∼ 3 σ) with solar neutrino experiments. In order to assess the robustness of the signal for gallium experiments we present a discussion of the impact of cross-section uncertainties on the results.

Performances of neutron cold multi-chopper instruments with different source natures among reactor, short and long pulse sources

Cold multi-chopper instrument is a popular instrument commonly used for variety of scientific fields, built in reactors and sharp pulse sources and will be built in a long pulse source. In this article I compare the performance of those instruments, IN5 at ILL, AMATERAS at J-PARC and CSPEC at ESS, and show a merit/demerit come from the different characters of the source itself, and finally give a suggestion for a further improvement.

Development of a polynomial model for power prediction of miniature neutron source reactors

Development of a polynomial model for power prediction of miniature neutron source reactors

Structural engineering of tin sulfides anchored on nitrogen/phosphorus dual-doped carbon nanofibres in sodium/potassium-ion batteries

Na/K-ion batteries (SIBs/PIBs) owing to their low cost, earth abundance, appropriate redox potential and comparable electrochemical performance, have gained ever-growing attention. Nevertheless, it remains a major challenge for high performance of SIBs/PIBs applications. Herein, with the assistance of waste chlorella as the reactor and phosphorous source, we designed few-layered tin sulfides immobilized on nitrogen and phosphorus dual-doped carbon nanofibres (SnSx-N/P-CNFs). The characterization and DFT calculation results demonstrate that N/P co-doping is expected to favour the distribution of electron density and strengthen ion reaction kinetics, which can enhance the Na+/K+ storage performance. As expected, as an anode material for SIBs, the SnSx-N/P–CNF electrode displays an impressive capacity (522 mAh g−1 after 50 cycles at 0.1 A g−1) and promising long-life cycling performance (214 mAh g−1 up to 32,000 cycles at 10 A g−1). Moreover, it also exhibits exceedingly impressive potassium-ion storage performance (468 mAh g−1 at 0.1 A g−1 after 100 cycles and 170 mAh g−1 exceed 10,000 cycles at 5 A g−1), which is the one of the optimal long-cycle properties reported for Sn-based anode for PIBs to date. Our work provides a reference for using biomass algae as the nitrogen and phosphorous source in constructing energy storage materials.

Feasibility study of Isfahan MNSR as a calibration thermal neutron source

In the present work, the thermal beam-tube of Isfahan Miniature Neutron Source Reactor (MNSR) with known neutron dose-equivalent rate and flux is investigated as a calibration source of thermal neutrons. It is shown that 90% of the neutrons emerging from this tube are thermal. Measuring the response of TLD-600 and TLD-700 dosimeters exposed to these neutrons reveals that the both dosimeters respond linearly to neutrons in the range of 0.1 to 38 mSv. Moreover, after subtracting the contribution of mixed gamma rays using the calibration curves obtained in a standard 60Co gamma rays field, the response of TLD-600 dosimeter still behaves linearly with dose-equivalent. Finally, it is concluded that the dose-equivalent rate and flux of thermal neutrons in Isfahan MNSR are stable with time and this reactor can reliably be used as a calibration source for thermal neutrons.

Diffusion of fission products in nuclear graphite: A review

• A literature review of the metallic FPs diffusion in nuclear graphite is presented. • The diffusion theories applied in the diffusivity determination are summarized. • Measurements of the diffusion coefficient of metallic FPs are discussed. • The variables affecting the diffusion coefficients of Cs, Sr, and Ag are discussed. Nuclear graphite in high-temperature gas-cooled reactors (HTGRs) acts as a moderator, a structural material, as well as an accommodation for fission products (FPs). Given the release of FPs under normal and accident conditions, the diffusion of FPs in nuclear graphite should be elucidated for the estimation of source term and reactor safety. However, the existing experimental data of diffusion are enriched in data of the nuclear graphite types (i.e., H451 and A3), that may not apply to next-generation nuclear graphite. The proposed diffusion theory and experimental systems are conducive to developinging a novel experimental method for the more accurate collection of the diffusion data and an insight into FPs diffusion in graphite. For this reason, this paper reviews the information regarding the diffusion of metallic FPs (e.g., Cs, Sr and Ag) in nuclear graphite. The diffusion theory adopted to calculate the diffusion coefficient is summarized. Emphasis is placed on the measurements for determination the diffusion coefficient and the discussion of detection limits, accuracy and experimental methods. Given the diffusion coefficient data, the variables of diffusion are addressed. The mentioned results lay a basis for gaining insights into the diffusion theory and can be exploited by employing with a valid measurement method to determine the diffusion coefficients of FPs in various nuclear graphite materials under a particular range of conditions.

Experimental Study on Biological Treatment of Copper in Acid Mine Drainage after Iron Removal

Acid mine drainage is characterized by low pH value, many types of heavy metals, high concentration of heavy metals and great environmental damage. Copper precipitation by sulfate reducing bacteria is a common method. In this experiment, the AMD after iron removal was treated by an upflow anaerobic bioreactor. The effects of HRT, carbon source type, carbon source dosage and reactor temperature on the operation of the reactor were studied. The optimum process parameters were determined: HRT was 12h, formic acid was used as carbon source and the dosage was 0.5g/L, and the temperature was 30 °C. Under the process conditions, the Cu concentration in the reactor effluent decreased to 0.043mg/L, and the recovery rate of Cu metal was 99.9%. The mechanism of copper deposition in the reactor was studied by characterization of the structure and morphology of the precipitated product and the analysis of the microbial community structure in the effluent of the reactor.

Investigative study of the radiation damage on fuel clad of miniature neutron source reactor using computational tools

Core conversion requires some evaluation of the reactor safety. Changes to the reactivity worth, shutdown margin, power density and material properties are crucial to the proper functioning of the reactor. The focus of this article is to study the neutron flux distribution in the reactor core and radiation damage on candidate clads. The Ghana Research Reactor-1 (GHARR-1) operates at maximum power of 30 kW in order to attain a flux of 1.0× 1012 n·cm–2·s for the high enriched uranium core. Using the GHARR-1 core geometry, considering 348 fuel pins, the multiplication factor (Keff) is calculated at enrichments of 10%, 12.5%, 16%, 20%, 30% and 90.2%. The spectrum of neutron flux generated in the 26 group is also calculated at the specified enrichments. The ion/particle interactions with the targets (clad) were studied in the Stopping and Range of Ion in Matter code to establish the best clad material based on recorded defects and vacancies generated. From the calculations and simulations, the best choice from the candidate clads based on the assessment is SiC. The calculation of the fuel campaign length gives 7.5 years. The defects sustained by the prospective clad showed low susceptibility to swelling and other forms of deformation.

Technical Note: Development of a generalized source model for flux estimation in nuclear reactors

Although very precise, the Monte Carlo method is difficult to exploit in an industrial context. It requires a relatively long calculation time to obtain results with a reasonable statistical accuracy. In the context of so-called fleet reactors, and especially, to facilitate the transposition from one reactor to another, it is proposed to develop a method allowing the modeling of the reactor source in a synthetic and adjustable way. One approach is to perform a complete Monte Carlo simulation and to generate during this simulation a phase-space file (PSF) by judiciously placing a storage surface. This PSF provides a precise distribution of the particle parameters and can be used as a source in a second simulation. This method is interesting but can lead to the introduction of a latent variance. An approach that overcomes this limitation is to create a so-called generalized source model based on histograms. By sampling the initial parameters of the particles, no particle can be generated twice with exactly the same parameters, which leads to a reduction of the latent variance present in PSF. In addition, histogram-based sampling is faster than PSF reading and also allows for correlations between particle parameters.

First results from a search for coherent elastic neutrino-nucleus scattering at a reactor site

The deployment of a low-noise 3 kg p-type point contact germanium detector at the Dresden-II power reactor, 8 meters from its 2.96 GW$_{th}$ core, is described. This location provides an unprecedented (anti)neutrino flux of 8.1$\times 10^{13} ~\bar{\nu_{e}}/$cm$^{2}$s. When combined with the 0.2 keV$_{ee}$ detector threshold achieved, a first measurement of CE$\nu$NS from a reactor source appears to be within reach. We report on the characterization and abatement of backgrounds during initial runs, deriving improved limits on extensions of the Standard Model involving a light vector mediator, from preliminary data.

Comparative analysis of core life-time for the NIRR-1 HEU and LEU cores

In this study, the end of lifetime for the Nigeria Research Reactor-1 high enriched uranium and the low enriched uranium fuelled core were analysed. The SCALE 6.2.3 computational sequences, CSAS6 and TRITON coupled with KENO-VI code were employed for Eigen value and depletion calculations respectively. The study was carried out to demonstrate the advantage of conversion of miniature neutron source reactor from high enriched uranium to the low enriched uranium fuel. The results obtained showed that the clean cold core excess reactivity values of the high enriched uranium and the low enriched uranium fuelled cores were comparable with the experimental data and indicate the suitability of the computational approach used in this study. Similarly, the integral worth of top beryllium shims was greater in the high enriched uranium core in comparison with the low enriched uranium fuelled core and could be attributed to the increased neutron leakage caused by hardening of the neutron spectrum due to higher inventory of U-238 in the low enriched uranium core. Furthermore, the lifetime of the low enriched uranium fuelled core was found to be higher when compared to the high enriched uranium core and could also be attributed to the higher loading of U-235 and U-238 in the low enriched uranium core. The longevity of the low enriched uranium fuelled core in comparison to the high enriched uranium indicate an advantage for its longer time of utilization in neutron activation analysis and underscore the importance of conversion from high enriched uranium to low enriched uranium fuel.

Integral experiments of technetium-99 using fast-neutron source reactor ‘YAYOI’

ABSTRACT The present study performed integral experiments of 99Tc by an activation method using a fast-neutron source reactor ‘YAYOI’ of the University of Tokyo to validate evaluated nuclear data libraries: JENDL-4.0, ENDF/B-VII and VIII. Technetium-99 samples were irradiated with reactor neutrons using a pneumatic system, and repeated 74 times to accumulate statistics. Reaction rates of 99Tc were obtained by measuring decay gamma rays emitted from 100Tc. The fast-neutron flux spectrum at the YAYOI was checked with flux monitors and MCNP calculations. The experimental reaction rate of 99Tc was compared with the reaction rates given using both the fast-neutron flux spectrum and evaluated nuclear data libraries. The present integral experiments supported the evaluated nuclear data library JENDL-4.0.

CALCULATION OF RADIOACTIVE SOURCE TERM RELEASE FROM FLEXBLUE NPP

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).

Progress in neutron techniques: towards improved polymer electrolyte membranes for energy devices

Design and implementation of advanced membrane formulations for selective transport of ions and molecular species are critical for creating the next generations of fuel cells and separation devices. It is necessary to understand the detailed transport mechanisms over time- and length-scales relevant to the device operation, both in laboratory models and in working systems under realistic operational conditions. Neutron scattering techniques including quasi-elastic neutron scattering, reflectivity and imaging are implemented at beamline stations at reactor and spallation source facilities worldwide. With the advent of new and improved instrument design, detector methodology, source characteristics and data analysis protocols, these neutron scattering techniques are emerging as a primary tool for research to design, evaluate and implement advanced membrane technologies for fuel cell and separation devices. Here we describe these techniques and their development and implementation at the ILL reactor source (Institut Laue-Langevin, Grenoble, France) and ISIS Neutron and Muon Spallation source (Harwell Science and Technology Campus, UK) as examples. We also mention similar developments under way at other facilities worldwide, and describe approaches such as combining optical with neutron Raman scattering and x-ray absorption with neutron imaging and tomography, and carrying out such experiments in specialised fuel cells designed to mimic as closely possible actual operando conditions. These experiments and research projects will play a key role in enabling and testing new membrane formulations for efficient and sustainable energy production/conversion and separations technologies.