Fast Flux Research Articles

FIPD: The SFR metallic fuels irradiation & physics database

The DOE Advanced Reactor Technology (ART) program has supported efforts to recover and preserve metallic fuel data generated throughout the US sodium-cooled fast reactor (SFR) program. Those efforts have been focused on establishing databases of the experimental data that were mainly generated during the Integral Fast Reactor (IFR) program including data generated at Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and Transient Reactor Test Facility (TREAT) reactors, as well as out of pile data. The data is essential for future licensing activities of metallic fuel based advanced fast reactors. This paper describes the development of the SFR Metallic Fuels Irradiation & Physics Database (FIPD) and covers the scientific knowledge available in the database. The architecture of the FIPD is described by showing the available reactor operation data, fabrication, and post-irradiation examination (PIE) data, and other documents. The applications of the fuel database are also discussed.

The low enriched uranium miniature neutron source reactor (LEU-MNSR) neutron spectrum characterization for k0-INAA

Ghana’s highly enriched uranium (90.2%) research reactor has been converted to a low enriched uranium (13.0%) type. Its neutron parameters have been characterized for the application of k0-instrumental neutron activation analysis. The magnitude of the LEU epithermal neutron shaping factor (α) at the irradiation site increased by about 25% as compared to the HEU, while the thermal neutrons decreased by about 5%. This indicates that the epithermal neutron spectrum is much dependent on the 1/E1+α. Similarly, the measured fast neutron flux increased by about 22%. Generally, the results show that the LEU core’s neutron spectrum is relatively harder, i.e., there is a marginal increase in high-energy neutrons. The reactor spectrum parameters have been used to validate the k0-IAEA software. Two certified reference materials (Carbonate rock GBW 07108 (GBW2) and NIST Standard Reference Material 1646a, Estuarine Sediment (EST)) and the synthetic multi-element standard (SMELS I, II & III) were analyzed based on a routine irradiation scheme. The process quality assurance was evaluated based on a u-score test. Most (80 %) of the elements determined were within |u-score| < 1.64, while a few were observed within 1.964 > |u-score| > 1.64. The results show a suitable characterization process and the general applicability of the k0-IAEA software.

Millicharged cosmic rays and low recoil detectors

We consider the production of a "fast flux" of hypothetical millicharged particles (mCPs) in the interstellar medium (ISM). We consider two possible sources induced by cosmic rays: (a) $pp\rightarrow$(meson)$\rightarrow$(mCP) which adds to atmospheric production of mCPs, and (b) cosmic-ray up-scattering on a millicharged component of dark matter. We notice that the galactic magnetic fields retain mCPs for a long time, leading to an enhancement of the fast flux by many orders of magnitude. In both scenarios, we calculate the expected signal for direct dark matter detection aimed at electron recoil. We observe that in Scenario (a) neutrino detectors (ArgoNeuT and Super-Kamiokande) still provide superior sensitivity compared to dark matter detectors (XENON1T). However, in scenarios with a boosted dark matter component, the dark matter detectors perform better, given the enhancement of the upscattered flux at low velocities. Given the uncertainties, both in the flux generation model and in the actual atomic physics leading to electron recoil, it is still possible that the XENON1T-reported excess may come from a fast mCP flux, which will be decisively tested with future experiments.

Assessment of tritium effluent from Prototype Generation IV Sodium-cooled Fast Reactor

A tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR were calculated by the Argonne Reactor Computation (ARC) code suite for fast reactor analysis, and due to the potential uncertainties in source terms and variations of cold trap performance and material permeability during reactor operation, the tritium distributions and tritium effluents were estimated for several bounding cases. The upper bounding gaseous and liquid tritium release rates from PGSFR are ~ 440.7 Ci/year and 14.7 Ci/year, respectively. The gaseous tritium effluent is about few orders of magnitude smaller than the regulatory constraint of the tritium concentration in the air. The peak tritium concentration in the steam-generator feedwater is ~ 1.1x10-3 μCi/g, which is required to be diluted prior to releasing as tritiated water. The accumulated tritium in the cold trap after five cycle operations is ~ 52,000 Ci. Because of significant conservatism in these upper bounding estimations, it is expected that the tritium effluent at the nominal case with plausible source terms would be about 2 orders of magnitude smaller than that of the upper bounding case.

Microstructure and hardening induced by neutron irradiation in single crystal, ITER specification and cold rolled tungsten

In this work, we have performed neutron irradiation and sub-sequent microstructural and micro-hardness measurements on a series of tungsten grades. The neutron irradiation was performed in BR2 reactor up to 0.2 dpa at 600, 800 and 1200 °C. The selected irradiation temperatures correspond to a large range of the operation of tungsten in ITER as armour for the divertor. The applied fast neutron flux is representative of ITER irradiation conditions in terms of damage rate. The primary purpose of the study was to establish the relation between the initial microstructure, the resulting irradiation-induced microstructure and corresponding irradiation hardening. The microstructure was studied using transmission electron microscopy (TEM) and hardening was assessed by micro-hardness measurements. The obtained TEM results were used to predict the hardening employing the dispersed barrier model and strength coefficients established for the single crystal from the open literature. Based on the performed analysis, it appeared that the main contribution to the hardening at high irradiation temperature originates from the voids, while the dislocation loops provide a comparable contribution (30–50%) only under irradiation at 600 °C. Excellent agreement between the model prediction and experimentally measured hardness increase was obtained for the single crystal and ITER specification grade without any additional adjustment, while the hardening induced in the cold rolled plate was essentially overestimated by the model. This result suggests that the cold rolled tungsten plate exhibits an alternative mechanisms of the plastic deformation besides the conventional dislocation glide and multiplication, and this alternative mechanism is impacted by the irradiation less severely compared to the prediction coming from the dispersed barrier model.

Propagating neutronic uncertainties for FFTF LOFWOS Test #13

The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.

Design and optimization of thermal neutron device based on deuterium-deuterium neutron generator

A thermal neutron device based on a portable D-D neutron generator has been designed and optimized by Monte Carlo method. The goal of optimal design is to maximize the thermal neutron flux at the output surface and improve the utilization efficiency of the neutron generator. The simulation of the reference device is experimentally verified. The contributions of different types and sizes of moderators, reflectors, and collimators to the thermal neutron intensity were simulated to obtain the optimum dimensions of the proposed design. Two parameters P and K were defined to describe the performance of device. It was determined that the final device uses high-density polyethylene (HDPE) as the moderator, nickel as the reflector, and HDPE as the collimator. The parameter K of the final device was 1.72. The P-value was 8 times that of the reference device. The thermal, epithermal + intermediate, fast, and total neutron flux on the output surface were 4.65, 2.52, 1.86, and 2.72 times of the reference device, respectively. The ratio of thermal to fast neutron flux ϕth/ϕfast increased from 0.52 to 1.30. The neutron irradiation device is supposed to be employed principally for neutron activation analysis, testing and calibration of neutron dosimeter and detector.

Verification of CENDL-3.2 Nuclear Data on VENUS-3 Shielding Benchmark by ARES Transport Code

The recently released CENDL-3.2 nuclear data library is deemed as an important achievement in the field of nuclear data research in China. To verify the applicability of the library to the shielding calculation of PWR and analyze the influence of multigroup cross-section parameters on the shielding calculation, ARES-MACXS module is used to process the MATXS format multigroup library based on CENDL-3.2 to generate multigroup working cross sections for PWR shielding calculation. VENUS-3 experimental facility has a clear and complete geometry. It is often used to test the ability of the advanced transport calculation method of calculating RPV fast neutron flux and to evaluate the accuracy of cross-section library. Different cross-section parameters are chosen for ARES to calculate VENUS-3 benchmark, and equivalent neutron flux of 58Ni(n,p)58Co, 115In(n,n′)115mIn and 27Al(n,α)24Na detectors is calculated according to the data provided by the benchmark report. The numerical results demonstrate that almost all the relative deviations between the calculated results and the experimental results are within 20%, which satisfies the requirement of shielding calculation. CENDL-3.2 is suitable for PWR shielding calculation. The comparison of various cross-section parameters results indicates that multigroup cross-section parameters have large effects on the transport calculation results.

State-of-the-art progress of gaseous radiochemical method for detecting of ionizing radiation

The article provides a review of the research results obtained during of more than 20 years concerning using the gaseous radiochemical method (GRCM) for detecting of ionizing radiation. This method based on threshold nuclear reactions with production of radioactive noble gas which does not interact with the materials of gaseous tract. The applications of GRCM in the diagnostics of neutrinos, neutrons, charged particles, thermonuclear plasma thermometry, and the study of the structure and dynamics of astrophysical objects, position-sensitive dosimetry of neutron targets with accelerator driving, spatial distribution of the fast neutron flux density in a nuclear reactor allowing the transformation of longitudinal coordinate of neutron flux distribution into a temporal distribution of the radiochemical gas decay counting rate (“barcode” semblance) and measurement of bombarding particles spectra are described. Experimental testing of the described technologies was made on the neutron target driven with the linear proton accelerator of Institute for Nuclear Research of Russian Academy of Sciences (INR RAS).

Modelling fission gas behaviour in fast reactor (U,Pu)O[formula omitted] fuel with BISON

The physics-based fission gas behaviour model available in the BISON fuel performance code provides satisfactory predictive capabilities for application to light-water reactor conditions. In this work, we present a model extension for application to fast reactor (U,Pu)O2 fuel. In particular, we detail the introduction of a lower bound to the number density of grain-face bubbles, representing a limit to the coalescence process once extensive bubble interconnection is achieved. This new feature is tested first against an experimental database for UO2-LWR, and secondly is validated against integral irradiation experiments for fast reactor (U,Pu)O2 fuel rods irradiated in the FFTF (Fast Flux Test Facility) and in the JOYO reactors. The comparisons of BISON results with the experimental data are satisfactory and demonstrate an improvement compared to the standard version of the code.

Fast flux control of 3D transmon qubits using a magnetic hose

Fast magnetic flux control is a crucial ingredient for circuit quantum electrodynamics (cQED) systems. So far, it has been a challenge to implement this technology with the high coherence 3D cQED architecture. In this paper, we control the magnetic field inside a superconducting waveguide cavity using a magnetic hose, which allows flux control of 3D transmon qubits on time scales less than 100 ns while maintaining a cavity quality factor larger than 106. The magnetic hose is designed as an effective microwave filter to not compromise the energy relaxation time of the qubit. The magnetic hose is a promising tool for fast magnetic flux control in various platforms intended for quantum information processing and quantum optics.

A new structure design to extend graphite assembly lifespan in small modular molten salt reactors

Dimensional change of the graphite assemblies due to irradiation is a key factor in limiting the service lifespan of the reactor core in small modular molten salt reactors. The graphite assembly structure, determining the fast neutron flux distribution and temperature distribution, has a notable impact on its dimensional change and hence its lifespan. In this article, we put forward an innovative solid hexagonal prism assembly (HPA) surrounded by fuel salt to prolong the lifespan of the graphite assemblies and compare it with the traditional hexagonal round channel assembly (RCA) with fuel salt flowing through its central round channel. Under the single-cell model, the calculation results prove that the neutron flux distribution and temperature distribution in the HPA are more uniform than those in the RCA. The HPA deforms more slowly than the RCA under the same operating conditions. The lifespan of the RCA, defined as when it expands back to its original size, is 7.5 years under the 100 MW/m3 power density of fuel salt, while the lifespan of the proposed HPA is 8.4 years. The HPAs also allow a further expansion in the radial direction due to its noncontact arrangement in the core. It will take 14.9 years for the HPA assemblies to contact each other, which displays a significant improvement in lifespan.

ENRICHMENT ZONING STUDY FOR THE VERSATILE TEST REACTOR

The primary mission of the Versatile Test Reactor (VTR) is to provide peak fast flux in excess of 4.0 x 1015n/cm2-s to support fuel and material testing. To achieve a high fast flux, it is beneficial to maximize the flux peaking in the center of the core. With a single enrichment zone, a highly peaked flux distribution produces a highly peaked power distribution. Coolant inlet orifices can be designed to handle the peaked power distribution but orifice design can be simplified if a more even radial power distribution can be achieved. An approach to reduce the power peaking factor is to use enrichment zoning, which would improve coolant flow homogeneity. Several alternative VTR core configurations are considered with two enrichment zones (15 wt% Pu and 20 wt% Pu). These alternative configurations require more assemblies to maintain reactivity than the reference VTR core, which leads to failure to achieve the design criterion for experimental fast flux with the target core power. Configurations using 20 wt% Pu with different fuel assembly designs having smaller and larger fuel volume fractions are also analyzed. The case having a larger fuel volume fraction reduces the number of fuel assemblies required for criticality, which keeps the experimental flux higher. Configurations with volume fraction zoning can slightly decrease the peaking factor while maintaining the desired fast flux, although some thermal hydraulic limits may not be satisfied. Volume fraction zoning configurations may offer benefits, but determining the feasibility of these configurations requires further thermal hydraulic design and analysis work beyond the scope of the present work.

Botnet detection used fast-flux technique, based on adaptive dynamic evolving spiking neural network algorithm

A botnet refers to a group of machines. These machines are controlled distantly by a specific attacker. It represents a threat facing the web and data security. Fast-flux service network (FFSN) has been engaged by bot herders for cover malicious botnet activities. It has been engaged by bot herders for increasing the lifetime of malicious servers through changing the IP addresses of the domain name quickly. In the present research, we aimed to propose a new system. This system is named fast flux botnet catcher system (FFBCS). This system can detect FF-domains in an online mode using an adaptive dynamic evolving spiking neural network algorithm. Comparing with two other related approaches the proposed system shows a high level of detection accuracy, low false positive and negative rates, respectively. It shows a high performance. The algorithm's proposed adaptation increased the accuracy of the detection. For instance, this accuracy reached (98.76%) approximately.

Botnet detection used fast-flux technique, based on adaptive dynamic evolving spiking neural network algorithm

A botnet refers to a group of machines. These machines are controlled distantly by a specific attacker. It represents a threat facing the web and data security. Fast-flux service network (FFSN) has been engaged by bot herders for cover malicious botnet activities. It has been engaged by bot herders for increasing the lifetime of malicious servers through changing the IP addresses of the domain name quickly. In the present research, we aimed to propose a new system. This system is named fast flux botnet catcher system (FFBCS). This system can detect FF-domains in an online mode using an adaptive dynamic evolving spiking neural network algorithm. Comparing with two other related approaches the proposed system shows a high level of detection accuracy, low false positive and negative rates, respectively. It shows a high performance. The algorithm's proposed adaptation increased the accuracy of the detection. For instance, this accuracy reached (98.76%) approximately.

CORE DESIGN ACTIVITIES OF THE VERSATILE TEST REACTOR – CONCEPTUAL PHASE

The Versatile Test Reactor (VTR) is a new fast spectrum test reactor being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. This includes neutron irradiation capabilities which would support alternate coolants including molten salt, lead/lead-bismuth eutectic mixture, gas, and sodium. The VTR aims at addressing most of the needs of the various stakeholders, which is primarily composed of advanced reactor technologists, developers and vendors, as well as a number of others interested parties.Design activities are underway targeting a first criticality date by 2026, with General Electric recently joining the project to contribute to the VTR plant design. Current efforts are focused on all aspects of the VTR design, with the core design being at the center of the initial steps. The VTR is currently proposed as a 300 MWth sodium-cooled fast reactor able to reach peak fast flux levels in excess of 4.0x1015 n/cm2-s (and total flux level of about 6.0x1015 n/cm2-s). In this configuration, it is using ternary metallic fuel with reactor-grade plutonium and 5% low-enriched uranium.

NEUTRONIC MODELLING OF THE FFTF CONTROL ROD WORTH MEASUREMENTS WITH DIFFUSION CODES

This paper presents an assessment of three deterministic core simulators with the focus on the neutronic performance in steady-state calculations of small Sodium cooled Fast Reactor cores. The selected codes are DYN3D, PARCS and the novel multi-physics solver GeN-Foam. By using these codes, the multi-group diffusion solutions are obtained for the selected twenty control rod worth measurements performed during the isothermal physics tests of the Fast Flux Test Facility (FFTF). The identical set of homogenized few-group cross sections applied in the calculations is generated with the Serpent Monte Carlo code. The numerical results are compared with each other as well as with the measured values. The obtained numerical results, such as the multiplication factors and control rod worth values, are in good agreement as compared to the experimental data. Furthermore, a comparison of the radial power distributions is presented between DYN3D, PARCS and GeN-Foam. Ultimately, the power distributions are compared to the full core Serpent solution, demonstrating an adequate performance of the selected deterministic tools. In overall, this study presents a verification and validation of the neutronic solvers applied by DYN3D, PARCS and GeN-Foam to steady-state calculations of SFR cores.

Iron-Mineralization-Induced Mesoporous Metal-Organic Frameworks Enable High-Efficiency Synergistic Catalysis of Natural/Nanomimic Enzymes.

Metal-organic frameworks (MOFs) have become a promising accommodation for enzyme immobilization and protection. However, the integration of multienzymes into MOFs may result in compromise of individual enzymatic activity. In this work, we report an iron mineralization strategy to facilely construct a mesoporous MOF, possessing excellent peroxidase-mimic bioactivity. Furthermore, the feasibility of in situ encapsulating natural enzymes within the developed mesoporous MOF nanozymes endows these natural/nanomimic enzyme hybrids with remarkably enhanced synergistic catalysis ability. Such activity enhancement is mainly due to (1) the fast flux rate of substances through the interconnected mesoporous channels and (2) the simultaneously increased loading amount of enzymes and iron within the MOFs caused by the iron mineralization process.

Investigation of irradiated metal of WWER-type reactor internals after 45 years of operation. Part 2. Calculated and experimental determination of the fast neutron fluence and damage dose

Fast neutron flux, fluence and damage dose have been determined when using experimentally measured specific activity of 54 Mn и 58 Co isotopes for metal of samples cut out from elements of pressure vessel internals of Novovoronezh NPP Unit No 3 (18Cr-10Ni-Ti steel, analog of AISI 321 steel). The results have been compared with the values calculated by KATRIN-2.5 computer code.

Verification and Validation of the Geant4 Monte Carlo Code Toolkit for DEMO TBR Evaluations

Verification and Validation analyses of the Monte Carlo code toolkit Geant4 were conducted with nuclear data from the JEFF-3.3 library to show the suitability for fusion applications. We present benchmarking against data of the engineering design oriented HCPB (Helium Cooled Pebble Bed) Tritium Breeder Module (TBM) mock-up experiment taken from the SINBAD compilation and nuclear analyses in comparison with MCNP of the DEMO reactor equipped with a HCPB breeding blanket. Missing functionalities for Geant4 were developed such as a track length detector with multiple energy bins and tally multiplication, reflective boundaries, and statistical checks in a post-processing step. The activation foil measurements in the TBM mock-up revealed Geant4 underestimates the uncollided and fast neutron flux increasingly with penetration depth, while being consistent with MCNP for the low energy neutron flux. Geant4 underestimates the experimental tritium activity in the breeder pellets by up to 2.7% more than MCNP, which is within the experimental uncertainty. The DEMO analyses yielded a 0.97% lower Tritium Breeding Ratio (TBR) of Geant4 than MCNP for a homogenized breeder blanket, and a 1.46% higher TBR of Geant4 than MCNP for a mostly heterogenic blanket. This good agreement with MCNP demonstrates Geant4’s suitability for DEMO TBR evaluations.