Source Reactor Research Articles

Development and verification of open reactor simulator ADPRES

Abu Dhabi Polytechnic Reactor Simulator (ADPRES) is an open source reactor simulator recently developed at Abu Dhabi Polytechnic. It solves static and time-dependent three-dimensional, multi-group neutron diffusion equation in Cartesian geometries. Fourth order standard Nodal Expansion Method (NEM) is implemented in ADPRES for spatial discretization. The time-dependent diffusion equation is numerically solved by fully-implicit scheme. ADPRES input is straight-forward and designed with minimal keywords. ADPRES also has most features of today's modern reactor simulators. In this paper, ADPRES solutions for several reactor benchmarks were verified. The first part of PWR MOX/UO2 core benchmark was used to verify ADPRES on static reactor problem. For transient reactor problems, LMW three-dimensional and NEACRP 3D PWR transient benchmarks were used. ADPRES solutions were generally accurate and consistent with the solutions from other computer codes. The source codes can be found in the author’s GitHub repository.

Development of a 3D thermohydraulic-neutronic coupling model for accident analysis in research miniature neutron source reactor (MNSR)

Abstract To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

Spatial Organization of a cAMP/Ca2+‐Regulated Signaling Complex: A Structural Study of an AKAP79‐Scaffolded Complex Containing Type IIβ PKA and Calcineurin

cAMP‐dependent protein kinase (PKA) regulates many cell processes including development, differentiation, and memory. In its resting state, PKA is a tetrameric holoenzyme consisting of a regulatory (R) subunit dimer and two catalytic (C) subunits. Four isoforms of the R‐subunit exist (RIα, RIβ, RIIα, and RIIβ). Although each isoform varies in sequence and quaternary structure, all share a common domain organization. The strength, duration, and specificity of PKA activity within the cell are mostly dependent upon isoform‐specific subcellular localization. The D/D domain interface of the R‐ subunits serves as a docking site for A Kinase Anchoring Proteins (AKAPs) that target PKA to specific cellular locations. In cardiac myocytes and neurons, PKA is localized to Cav1.2 L‐type calcium channels by AKAP79 and up‐regulates calcium influx by phosphorylating the channel's cytosolic C‐terminal tail. AKAP79 also binds the Ca2+/calmodulin‐dependent protein phosphatase type 2B (PP2B, or calcineurin) at a site very close to that of PKA binding. Given the close proximity of the calcineurin and PKA binding sites on AKAP79, calcineurin has been suggested to play roles in the regulation of calcium channels, along with regulation of PKA activity itself. Using a variety of biochemical and biophysical techniques, we have isolated and characterized a stable complex of the C‐terminal tail of AKAP79, RIIβ holoenzyme, and calciuneurin (“ARC” for short). We show a stoichiometry of 1:1:1 (AKAP79ct: RIIβ2‐C2 :CNt) using native mass spectroscopy, and SAXS/SANS techniques confirm that this is a compact complex, suggesting a close interaction between PKA and calcineurin when bound to AKAP79. Altogether, we suggest beyond its function of localizing PKA and CN in close spatial proximity to the Cav1.2 L‐type calcium channel, the ARC signalosome fosters extended protein‐protein interactions and regulations that suggest tight spatiotemporal control. Support for this work was provided through NIH GM034921.Support or Funding InformationJames Hall is an IRACDA fellow and was supported by NIGMS/NIH award K12GM068524. This work was also supported by a grant from the National Institutes of Health GM034921 to S.S. Taylor. This Research used resources at the High Flux Isotope Reactor and Spallation Neutron Source, which are DOE Office of Science User Facilities operated by Oak Ridge National Laboratory. This work utilized SasView software, originally developed by the DANSE project under NSF award DMR‐0520547.This abstract is from the Experimental Biology 2019 Meeting. There is no full text article associated with this abstract published in The FASEB Journal.

Analysis of the reactivity coefficients and fuel burnup for MNSR research reactor using DRAGON5/DONJON5 codes

The aim of this report is to analyze and validate a calculation, using new versions of the lattice DRAGON5 code (2D), the DONJON5 code (3D), and four cross-sectional libraries are ENDF/B-VII.0, ENDF/B-VII.1, JENDL4.0, and IAEA library. In this paper, the reactivity coefficients of a Miniature Neutron Source Reactor (MNSR) during burn-up changes are calculated. Two-dimensional DRAGON5 assembly transport and depletion calculations are used to compute the group constants of fuel assemblies. After a number of cell calculations in DRAGON5, the resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model. The full-core diffusion calculations are performed with three-dimensional DONJON5 code, from which both effective multiplication factor, excess reactivity, fuel temperature coefficient (FTC), moderator temperature coefficient (MTC), moderator density coefficient (MDC), and the channel power peaking factor are estimated, for different nuclear data libraries. Moreover, the dependency of coefficients with neutronic, thermal-hydraulic, and geometrical aspects is considered. The results of the calculation show a good agreement of the core reactivity coefficients with the previous publications.

A neutron radiography beamline relying on the Isfahan Miniature Neutron Source Reactor

Abstract Concerning to the costs of the establishment of new research reactors and their waste issues, the world’s quest is to expand the capabilities of existing research reactors and to develop their application further. In this way, attempts have been made to extract an appropriate thermal neutron beam for neutron radiography purposes at the 30 kW Isfahan MNSR reactor. In this study, the result of calculations and experiments in support of the implementation of a thermal neutron radiography beamline at this reactor has been demonstrated. While many research reactors have beam tube equipment, the original design of the MNSR did not include beam tubes. Therefore, at first stage of our feasibility study, a vertical aluminum tube with a diameter of about 5 cm has been used as an external beam tube to validate calculated values of MCNP code with experimental measurements. The image quality has been determined using ASTM image quality indicators. It can conclude that the Isfahan MNSR is an appropriate reactor with suitable neutron flux for a neutron radiography system. Results show that it is possible to obtain an appropriate neutron radiography beam at this low power research reactor. Also, according to the physical restrictions of the MNSR and considering the least possible changes in the current reactor status, some beam tubes with larger beam diameters have been designed and the related beam parameters have been calculated and demonstrated.

Deflection predictions of involute-shaped fuel plates using a fully-coupled numerical approach

This paper describes the modeling and simulation of fluid structure interactions (FSI) of involute-shaped fuel plates used in nuclear research reactors. We believe this to be the first time that this type of application is described in the literature using a fully-coupled, and monolithic, finite element approach. The simulations are validated against plate deflection data for the conceptual design of the Advanced Neutron Source Reactor (ANSR), which was envisioned to be the world’s most powerful nuclear research reactor for neutron scattering and other applications, but was ultimately never completed. The high performance of the ANSR creates a bounding envelope for involute-shaped research reactors such as that used in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) which is undergoing research for the conversion from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. As such, the findings from the present FSI analyses carried out herein for the ANSR plates provide good guidelines and inform designers what should be expected for the next generation of plates in the HFIR. It is shown herein that the current approach can accurately capture the leading-edge deflections of the involute-shaped plates and simulations can predict the ‘S-shaped’ deflection of the first mode instilling confidence in the methodology.

Evaluation of the impact of cadmium poison and cadmium line channel on the dynamic behavior of miniature neutron source reactor (MNSR)

Nigerian Research Reactor-1 (NIRR-1) undergo different installation such as permanentCadmium line due to high demand for epithermal neutrons activation analysis (ENAA)irradiation by the clients and cadmium poison to regulate the reactivity of the reactor witha range value 4.0mk to 3.7mk. Safety and flux stability. Simulations were carried out viaMonte Carlo Transport Code MCNP5 for these installations but very few experimentalcalculation were done. The results of all these research work revealed that the effect ofcadmium-line on safety and flux stability is very small. But in the case of cadmium poison,the reactivity of the reactor after 10 years of operation was observe to be decreasing. Theresults obtained for excess core reactivity after (2.96 and 2.95) mk, and predicted power of14.65 kW for the two experiment with coolant temperature (12.10 and 12.08) °Crespectively. These show that the change in power of the reactor is very small. The outcomeof the research will help sample handling capabilities of NIRR-1 and provide useful data tothe MNSR.Key words: Epithermal neutron, ENAA Irradiation, NIRR-1.

Determination of neutron flux parameters (f, [formula omitted], φth and φe) of the irradiation sites of Isfahan MNSR reactor using empirical and MCNPX2.6 simulation approaches

Miniature Neutron Source reactors (MNSRs) are ideal for applications such as nuclear research, neutron activation analysis, etc., provided that their neutron flux parameters (φth, φe, f and α) are known. This paper present the results of the neutron flux parameters at inner and outer irradiation sites of Isfahan MNSR reactor determined through simulation with MCNPX2.6 code and experimentally using bare and cadmium-covered gold foils irradiation and bare triple (AuZr) monitor methods.In empirical approach, the obtained φth, φe, f [fbare] and α-values were 5.02 × 1011 ncm−2s−1 (±4.5%), 3.13 × 1010 ncm−2s−1 (±2.3%), 16.0 (±6.7%) [17.3 (±9.9%)] and −0.121 (±0.2%), for inner site; and 2.93 × 1011 ncm−2s−1 (±2.3%), 6.48 × 109 ncm−2s−1 (±3.1%), 45.2 (±5.4%) [42.5 (±7.1%)] and −0.011 (±1.8%), respectively, for outer site. In simulation approach, while, they were found to be 4.76 × 1011 ncm−2s−1 (±0.9%), 2.00 × 1010 ncm−2s−1 (±2.2%), 23.8 (±3.1%) and −0.078 (±0.13%), for the inner site; and 2.67 × 1011 ncm−2s−1 (±1.1%), 3.79 × 109 ncm−2s−1 (±5.4%), 70.4 (±6.5%) and −0.017 (±0.4%) for the outer site, respectively.Comparison of empirical and simulation results clearly revealed that: the inner site's φth and f values correspond with those measured during the first startup of the reactor; the values of φth are more reliable than φe,-values, as are the inner site's f and α results in comparison with outer site's values; the inner site's φth and φe are ∼1.7 and 5 times, respectively, larger than those of outer site; and the inner site 's α and f-values are more than 4.8 and 2.4 times larger and less than the outer site's values, respectively.

Feasibility study for production of <SUP align=right>99</SUP>Mo and <SUP align=right>99m</SUP>Tc by the neutron activation of <SUP align=right>98</SUP>Mo in the MNSR reactor

The calculated weekly specific activities of 99Mo and 99mTc produced from the irradiation of the MoO3 targets in the Miniature Neutron Source Reactor (MNSR) are presented in this paper. The productions of the 99Mo and 99mTc were modelled and calculated through a set of differential equations using the Mathcad program. The resonance self-shielding factor (Gres) was calculated using the MATSSF and MCNP4C codes. The effects of the physical parameters such as the neutron flux and irradiation time on the weekly specific activities of 99Mo and 99mTc have been analysed. It was found that the optimum irradiation scheme was achieved when the MNSR was operated for an extended period of 5 hours a day for 5 days a week at the neutron flux of 7.5×1011 n.cm‒2.s‒1. The weekly specific activities for the 99Mo and 99mTc which can be produced in the MNSR, filling each of the five inner irradiation sites with 20 g of the MoO3 targets and using the regular natural convection method to cool the reactor, were 7.20 and 5.05 mCi.g‒1, respectively.

Feasibility study for production of <SUP align=right>99</SUP>Mo and <SUP align=right>99m</SUP>Tc by the neutron activation of <SUP align=right>98</SUP>Mo in the MNSR reactor

The calculated weekly specific activities of 99Mo and 99mTc produced from the irradiation of the MoO3 targets in the Miniature Neutron Source Reactor (MNSR) are presented in this paper. The productions of the 99Mo and 99mTc were modelled and calculated through a set of differential equations using the Mathcad program. The resonance self-shielding factor (Gres) was calculated using the MATSSF and MCNP4C codes. The effects of the physical parameters such as the neutron flux and irradiation time on the weekly specific activities of 99Mo and 99mTc have been analysed. It was found that the optimum irradiation scheme was achieved when the MNSR was operated for an extended period of 5 hours a day for 5 days a week at the neutron flux of 7.5×1011 n.cm‒2.s‒1. The weekly specific activities for the 99Mo and 99mTc which can be produced in the MNSR, filling each of the five inner irradiation sites with 20 g of the MoO3 targets and using the regular natural convection method to cool the reactor, were 7.20 and 5.05 mCi.g‒1, respectively.

The Influence of Material Composition and Geometry of Heterogeneous Medium with Resonance Scattering on the Density of Fast Neutrons Remote from the Source

An option of high-density neutron flux generation locally, at a distance from the source, is considered in this paper. The energy dependence of the cross sections of neutron scattering by nuclei of some isotopes displays resonance behavior. Below the resonance level by energy, there is a deep drop of the scattering cross section dictated by the effect of resonance and potential scattering interference. When a slowing-down neutron gets the energy corresponding to a very small scattering cross section, it is capable of flying for several meters without interaction with nuclei of the medium. The possibilities of creating conditions for selection of high-energy neutrons emitted by a reactor or other source and moving in the required direction are investigated in the paper. The research described in the paper is focused on various factors (material composition, geometry) influencing the neutron radiation density and achievable neutron flux limits.

A Review of Ghana Research Reactor-1 (GHARR-1) Component Aging Degradation Problems and Ways of Mitigation

GHARR-1 facility is a Miniature Neutron Source Reactor with a rated power of 30 kW. GHARR-1 was commissioned on 8th March, 1995. The facility has been operating using the Micro-Computer Closed Loop System (MCCLS) and Control Console (CC) for 19 years. Age related degradation effects in safety related systems of nuclear reactors are managed to prevent safety margins from eroding below the acceptable limits provided in reactors design. This paper therefore provides an update on managing the safety aspects of ageing of structures, systems, and components (SSC) of GHARR-1. Managing the safety aspects of research reactors ageing requires a proactive, systematic, and integrated ageing management approach for the coordination of all activities relating to control, monitoring, and mitigation of ageing degradation of SSC through the lifecycle of the GHARR-1. This paper outlines the ageing management program and mitigation practices. Strategies for the ageing management include: periodic safety reviews, design features for components and unit replacement, and succession planning. Information sharing with other operating organizations is one of the means considered by GHARR-1 to attain excellence. This paper again concisely reviews and integrates information developed by other aging management studies and other available information related to understanding and managing age-related degradation effects.DOI: http://dx.doi.org/10.5755/j01.erem.74.3.19974

Prospects for exploring New Physics in Coherent Elastic Neutrino-Nucleus Scattering

Coherent Elastic Neutrino-Nucleus Scattering (CEvNS) is a Standard Model process that, although predicted for decades, has only been detected recently by the COHERENT collaboration. Now that CEvNS has been discovered, it provides a new probe for physics beyond the Standard Model. We study the potential to probe New Physics with CEvNS through the use of low temperature bolometers at a reactor source. We consider contributions to CEvNS due to a neutrino magnetic moment (NMM), Non-Standard Interactions (NSI) that may or may not change flavor, and simplified models containing a massive scalar or vector mediator. Targets consisting of Ge, Zn, Si, CaWO4, and Al2O3 are examined. We show that by reaching a percentage-level precision measurement on the CEvNS energy spectrum down to \U0001d4aa(10) eV, forthcoming experiments will improve by two orders of magnitude both the CEvNS-based NMM limit and the search for new massive mediators. Additionally, we demonstrate that such dedicated low-threshold CEvNS experiments will lead to unprecedented constraints on NSI parameters (particularly when multiple targets are combined) which will have major implications for the global neutrino physics program.

Rational design of a scalable bioprocess platform for bacterial cellulose production

Bacterial cellulose (BC) has been gaining importance over the past decades as a versatile material that finds applications in diverse industries. However, a secured supply is hindered by the slow production rate and batch-to-batch variability of the yield. Here, we report a rational approach for characterising the BC production process using Design of Experiment (DoE) methodology to study the impact of different parameters on desired process attributes. Notably, we found that the carbon source used for bacterial growth significantly impacts the interplay between the process variables and affects the desired outcomes. We therefore, propose that the highest priority process outcome in this study, the yield, is a function of the carbon source and optimal reactor design. Our systematic approach has achieved projected BC yields as high as ∼40 g/L for Gluconacetobacter hansenii 53582 grown on sucrose as the carbon source compared to the widely reported yields of ∼10 g/L.

The Neutron Macromolecular Crystallography Instruments at Oak Ridge National Laboratory: Advances, Challenges, and Opportunities

The IMAGINE and MaNDi instruments, located at Oak Ridge National Laboratory High Flux Isotope Reactor and Spallation Neutron Source, respectively, are powerful tools for determining the positions of hydrogen atoms in biological macromolecules and their ligands, orienting water molecules, and for differentiating chemical states in macromolecular structures. The possibility to model hydrogen and deuterium atoms in neutron structures arises from the strong interaction of neutrons with the nuclei of these isotopes. Positions can be unambiguously assigned from diffraction studies at the 1.5–2.5 Å resolutions, which are typical for protein crystals. Neutrons have the additional benefit for structural biology of not inducing radiation damage to protein crystals, which can be critical in the study of metalloproteins. Here we review the specifications of the IMAGINE and MaNDi beamlines and illustrate their complementarity. IMAGINE is suitable for crystals with unit cell edges up to 150 Å using a quasi-Laue technique, whereas MaNDi provides neutron crystallography resources for large unit cell samples with unit cell edges up to 300 Å using the time of flight (TOF) Laue technique. The microbial culture and crystal growth facilities which support the IMAGINE and MaNDi user programs are also described.

Evaluation of a New Reagent-Ion Source and Focusing Ion-Molecule Reactor for Use in Proton-Transfer-Reaction Mass Spectrometry.

We evaluate the performance of a new chemical ionization source called Vocus, consisting of a discharge reagent-ion source and focusing ion-molecule reactor (FIMR) for use in proton-transfer-reaction time-of-flight mass spectrometry (PTR-TOF) measurements of volatile organic compounds (VOCs) in air. The reagent ion source uses a low-pressure discharge. The FIMR consists of a glass tube with a resistive coating, mounted inside a radio frequency (RF) quadrupole. The axial electric field is used to enhance ion collision energies and limit cluster ion formation. The RF field focuses ions to the central axis of the reactor and improves the detection efficiency of product ions. Ion trajectory calculations demonstrate the mass-dependent focusing of ions and enhancement of the ion collision energy by the RF field, in particular for the lighter ions. Product ion signals are increased by a factor of 10 when the RF field is applied (5000-18 000 cps ppbv-1), improving measurement precision and detection limits while operating at very similar reaction conditions as traditional PTR instruments. Because of the high water mixing ratio in the FIMR, we observe no dependence of the sensitivity on ambient sample humidity. In this work, the Vocus is interfaced to a TOF mass analyzer with a mass resolving power up to 12 000, which allows clear separation of isobaric ions, observed at nearly every nominal mass when measuring ambient air. Measurement response times are determined for a range of ketones with saturation vapor concentrations down to 5 × 104 μg m-3 and compare favorably with previously published results for a PTR-MS instrument.

A suite-level review of the neutron powder diffraction instruments at Oak Ridge National Laboratory.

The suite of neutron powder diffractometers at Oak Ridge National Laboratory (ORNL) utilizes the distinct characteristics of the Spallation Neutron Source and High Flux Isotope Reactor to enable the measurements of powder samples over an unparalleled regime at a single laboratory. Full refinements over large Q ranges, total scattering methods, fast measurements under changing conditions, and a wide array of sample environments are available. This article provides a brief overview of each powder instrument at ORNL and details the complementarity across the suite. Future directions for the powder suite, including upgrades and new instruments, are also discussed.

Radiological Dose Assessment for the Ghana Research Reactor-1 at Shutdown Using Dispersion Model: Conversion from High-Enriched Uranium to Low-Enriched Uranium Fuel

Abstract Ghana Research Reactor-1 is a Miniature Neutron Source Reactor (MNSR) which is currently fueled with Highly-Enriched Uranium (HEU) aluminium alloy fuel. Efforts are underway to convert the research reactor fuel to a Low Enriched Uranium (LEU) oxide fuel. The project is a coordinated research work funded by the International Atomic Energy Agency through its Coordinated Research Project (CRP) on Core Conversion. The research project was started with thermal hydraulic and neutronic calculation on both fuel. A radiological dose assessment as part of safety assessment requirement needs to be carried out before the commencement of the core conversion project. As such, dose assessment was credibly estimated by employing a computer software (Health Physics Code HotSpot Version 3.0) developed by Lawrence Livermore National Laboratory. The code uses Gaussian plume model for atmospheric dispersion and deposition of radionuclides based on the meteorology and demography site information. The latest IAEA guidelines for radiological dose assessment was considered in the estimation of the released radionuclide doses. The anticipated ecological estimated radionuclides released provided a comprehensive theoretical and real bases for estimating the Committed Equivalent Dose (CED) covering the emergency and the low populated zone as expected in most severe accident scenarios. An isotope depletion analysis code ORIGEN-S coupled with MCNP5 code for neutron flux generation was used to study the possible available radionuclides present in the reactor core at shutdown. Some few released radionuclides were selected from the inventory generated from the HEU core. The selected radionuclides were used in the dispersion code for dose estimation. The total activity value of two selected radionuclides (Iodine and Cesium) from the inventory in Curies was 2.067E-03 and 6.20E-4 respectively. The values are based on the release fraction of the selected nuclides. The CED values estimated were found to be in agreement with the IAEA and US-NRC regulatory acceptable limit of 1mSv receive as public exposure and 50mSv for radiation worker exposure in a year. The study results can be recommended when establishing the required emergency planning zones around the Ghana Research Reactor-1 facility in future. DOI: http://dx.doi.org/10.5755/j01.erem.74.1.19948

Calculation of the activity inventory for the MNSR reactor using DRAGON4 code

A study on the fuel burnup and radioactive inventory of the Miniature Neutron Source Reactor (MNSR), after ten years of expected life of the reactor core, is conducted to validate the computer code DRAGON4 for the cluster geometry of the reactor. Excess reactivity of the reactor is calculated versus the fuel burnup time with different power levels 10, 15, 20, 25, and 30 kW. Furthermore, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity inventory of the MNSR core has been calculated using the DRAGON4 code and through comparison with the available results. A good agreement is observed between the present calculations of the DRAGON4 code, in addition to the previously published results of the GETERA and WIMSD4 codes.

Neutronic assessment of BDBA scenario at the end of Isfahan MNSR core life

Abstract The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (β eff ), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to 7.90 ± 0.01 mk which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).