Neutron Emission Rate Research Articles

On the expected performance of the SORGENTINA-RF fusion neutron source

A study devoted to assessing the neutronic performance of the SORGENTINA-RF fusion neutron source is here presented. The calculation is performed considering the characteristics of the accelerated mixed deuterium/tritium ion beam from the ion source, the transport of ions into the titanium layer that constitutes the ion beam target, their implantation and the fusion reaction cross section. An investigation has been done by varying the ion composition of the beam in order to assess if and to what extent diatomic and triatomic ion species affect the neutron emission rate of the plant. This calculation is also instrumental to evaluate the cost-effectiveness of equipping the ion source with a filtering system to remove secondary ions from the beam.

Evaluation of functional loss to radiation detector in tomographic device for spent-fuel inspection by high-energy photons and neutrons: A preliminary study

The aim of this study was to investigate functional loss of radiation detectors used for spent fuel inspections due to radiations from spent fuel. Dose to a radiation detector was evaluated by Geant4 simulation. Subsequently, the radiation detector measured counts, according to the irradiation time, for the test source with presence of external 60Co and 252Cf sources in almost identical condition to those of the simulation, where gamma and neutron emission rates of the sources were similar to the actual spent fuels. The counts during the 60Co irradiation were changed with discrepancy of 1.17% while neutron beam decreased more than 25% of total counts after its irradiation. Additionally, abnormalities in the detector’s measured background counts were observed following the neutron irradiation. The results can be applied to efforts to define the tolerance dose for radiation detectors used in neutron-abundant environments such as nuclear power plants or spent-fuel storage facilities.

Measurements of the 252Cf half-life

Data from manganese bath measurements of the neutron emission rates from commercial 252Cf sources over a period of 40 years have been used to derive an accurate value for the 252Cf half-life. Particular care was taken to account for the presence of other californium and curium isotopes in the sources, and the resulting value has an uncertainty of 0.05%. The present value has been compared with earlier measurements in a review of this important half-life.

ON UNCERTAINTIES IN THE MANGANESE SULPHATE BATH TECHNIQUES.

The manganese sulphate bath is a globally recognized method for determining emission of neutrons from radionuclide since 1959. In combination with the measurement of neutron emission anisotropy, this method is standardly used for the primary standardization of dosimetry quantities of neutron radiation. To achieve results at the metrology level, it is essential to have all quantities which contribute to the overall uncertainty properly determined and contributing uncertainty minimized. This paper is focused on the quantification of the main uncertainty sources for in-house developed manganese bath. In previous analyses, the final uncertainty of neutron emission rate was <1%, which was considered as a target value. Nevertheless, the relative change from reference values was relatively high and out of this uncertainty. This relative change raised questions about the accuracy of contributed parameters, like vessel volume, concentration of MnSO4 or computational model. Therefore, this paper focuses on determination and re-evaluation of these parameters with corresponding uncertainties.

Monte Carlo Simulation of D-D Neutron Emission Rate in NBI-Heated Deuterium Plasma of LHD

Monte Carlo Simulation of D-D Neutron Emission Rate in NBI-Heated Deuterium Plasma of LHD

Design of ITER neutron flux monitor system back-end electronics

The primary aim of the International Thermonuclear Experimental Reactor (ITER) Neutron Flux Monitor (NFM) system is to diagnose the total neutron emission rate and fusion power of the plasma fusion reaction. The neutron flux is measured directly by the system and the fusion power can be calculated according to the neutron flux result. The NFM system is composed of detectors and back-end electronics. This paper focuses on the work of NFM#7 back-end electronics design and development. The back-end electronics consist mainly of four parts: Firstly, the pre-amplifiers are used to charge integrate and amplify the signal of fission chambers. Secondly, the output of pre-amplifiers will be transmitted to Signal Processing Unit (SPU) for analog signal conditioning, data processing, and measurement result uploading. Thirdly, pre-amplifiers and fission chambers have different electricity sources. Pre-amplifiers receive electricity generated from low voltage power whilst fission chambers receive electricity generated from high voltage power. Lastly, the Instrument and Control (I&C) is designed as the interface between NFM hardware and ITER central system and its functions are configuration & control, monitoring, calibration, data communication & archiving and clock synchronization. All the four parts are well integrated and tested in neutron field. The system design also follows the guidelines and standards of ITER.

Improved evaluation of safeguards parameters from spent fuel measurements with the Differential Die-Away (DDA) instrument

The Differential Die-Away (DDA) technique is a highly sensitive non-destructive assay method for characterizing and detecting the presence of fissile material within an item of interest. DDA utilizes a series of pulses from a neutron generator (NG) to actively interrogate an item of interest. The die-away time of the neutron population induced by this active interrogation and the integral of the total differential die-away signal can be used to characterize items such as nuclear waste drums and spent nuclear fuel assemblies. Los Alamos National Laboratory (LANL) conceptualized, designed, and fabricated a DDA instrument that was deployed for field test measurements at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Oskarshamn, Sweden. The instrument performed multiple static measurements at fixed locations and dynamic axial scans of 15 pressurized water reactor (PWR) and 10 boiling water reactor (BWR) spent fuel assemblies, collecting both passive and active measurement data. The static assays of the assemblies measured the differential die-away signal, die-away time, and total passive neutron emission rate to create calibration curves for the evaluation of assembly multiplication, burnup, initial enrichment, effective fissile mass, and total elemental plutonium mass. Each calibration curve was optimized by minimizing the relative root mean square error (RRMSE) of assembly assay results compared to declared assembly parameters. The same quantities were also measured with the axial scans, and the resulting data were applied in two ways: (1) in the creation of calibration curves to improve evaluation of the same safeguards parameters as static assays, and (2) for comparison to simulation. In most cases, across both PWR and BWR assemblies, axial scan data improved the estimation of the above parameters, quantified by decreasing the calibration curve RRMSE. These axial scan results demonstrate the ability of the DDA instrument and analysis method to characterize spent PWR and BWR fuel as well as, or better than, a static assay of the same assembly. Furthermore, the DDA instrument’s unique ability to obtain both active and passive data in a single, axial scan of an entire spent fuel assembly represents a more efficient and accurate way of assaying spent fuel for verification purposes. These results represent a significant advancement for characterizing spent nuclear fuel compared to current technologies.

Characterization of young Californium sources by means of [formula omitted]-ray spectroscopy

The spontaneous fission and neutron emission properties of Californium sources serve as a unique tool to analyze and validate the quality of detection of fission products. State-of-the-art methods for determining the age of a young Cf source were re-examined and modified. Further improvement included the determination of the single Cf isotopes within the Cf source and the neutron emission rate of the source.The analysis was performed by comparing the simulated γ-ray spectra of the Nucleonica suite with measured data of a young Cf source employing an HPGe detector. In addition, with the evaluation of the 133.5 keV peak (mainly due to the β-decay Ce-144/Pr-144) in relation to a peak at 497.1 keV (mainly due to the β-decay Ru-103/Rh-103) or 724.3 keV (multiplet, dominated by Ru-105/Rh-105 and Ce-145/Pr-145), the age-determination method of the investigated Cf source could be considerably improved to a confined age of 20 ± 3 months.By means of decay simulations, some deficiencies of current techniques in determining the isotope abundances were introduced. In particular, the quantities of Cf-252 and Cf-249 were analyzed extensively. Consequently, the source’s neutron emission rate matched well the emission rate from the source’s certification. The difficulties of defining the abundance of additional Cf isotopes within the source, albeit with lesser importance, are also illustrated.

Suppression of toroidal Alfvén eigenmodes by the electron cyclotron current drive in KSTAR plasmas

Advanced operation scenarios such as high poloidal beta (β P) or high q min are promising concepts to achieve the steady-state high-performance fusion plasmas. However, those scenarios are prone to substantial Alfvénic activity, causing fast-ion transport and losses. Recent experiments with the advanced operation scenario on KSTAR tokamak have shown that the electron cyclotron current drive (ECCD) is able to mitigate and suppress the beam-ion driven toroidal Alfvén eigenmodes (TAEs) for over several tens of global energy confinement time. Co-current directional intermediate off-axis ECCD lowers the central safety factor slightly and tilts the central q-profile shape so that the continuum damping in the core region increases. Besides, the rise of central plasma pressure and increased thermal-ion Landau damping contribute to TAE stabilization. While the TAEs are suppressed, neutron emission rate and total stored energy increase by approximately 45% and 25%, respectively. Fast-ion transport estimated by TRANSP calculations approaches the classical level during the TAE suppression period. Substantial reduction in fast-ion loss and neutron deficit is also observed. Enhancement of fast-ion confinement by suppressing the TAEs leads to an increase of non-inductive current fraction and will benefit the sustainment of the long-pulse high-performance discharges.

Study of ion cyclotron range of frequencies heating characteristics in deuterium plasma in the Large Helical Device

The characteristics of ion cyclotron range of frequencies (ICRF) minority ion heating with a hydrogen minority and deuterium majority plasma were studied by ICRF modulation injection experiments in the Large Helical Device (LHD). In recent experiments with deuterium plasma, no significant increase in the neutron emission rate due to ICRF second harmonic deuteron heating was observed. Therefore, in this study, the neutron emission rate was used to refer to the information regarding the thermal ion component. Like the results of the observations of the heating efficiencies at various minority proton ratios, the experimental results showed good agreement with the simple model simulation of ICRF wave absorption. During these experiments, the accelerated minority hydrogen ions were observed by neutral particle analyzers. The counting rates of the energetic particles were higher in the lines of sight passing through the helical ripple than across the magnetic axis, and the counting rate decreased as the minority hydrogen ion ratio increased. The dependence of the minority hydrogen ion ratio on the density of the energetic ions was consistent with the experimentally observed heating efficiencies and simulations. The heating efficiency of ICRF minority ion heating could be well explained by simple model simulation in the LHD deuterium experiment.

Development of radiation characteristics analysis code system for geological disposal and application to vitrified waste disposal with various LWR burnup conditions

The radiation characteristics analysis code system for geological disposal is developed to study the radiation influence of high-level waste on an engineered barrier system for the various conditions from a reactor burnup through geological disposal. The verification of the code system is performed by the comparisons with the data in the previous study which analyzed the radioactivity, the radiation emission rate, the radiation energy spectrum, and the radiation flux for the vitrified waste fabricated from spent UO2 fuel with 45 GWd/t discharge burnup. The calculation of neutron emission rate and energy spectrum shows the improvement by the introduction of updated data, and the calculation results of the code system show the good agreement with the reported data, considering the calculation conditions. The effect of the burnup condition of spent nuclear fuel in a light water reactor (LWR) on the radiation characteristics of vitrified waste is surveyed by the application of the code system. The numerical analyses are performed on the basis of the burnup conditions of two types of fuels (UO2 and MOX) and two burnup levels of current standard (45 GWd/t) and high burnup (70 GWd/t). The ratio of high-level liquid waste in vitrified waste is 20 wt% for all cases. The analysis results indicate that the utilization of MOX fuel increases the MA radioactivity drastically, thereby increasing the neutron emission rate of waste and the neutron flux on the inner surface of the overpack by a factor of over nine in comparison to the UO2 fuel waste disposal, whereas the utilization of high burnup fuel increases the emission rate and flux of neutron and γ-ray by 20 – 170 %. The radiation fluence is sufficiently small even by the change of burnup condition compared to the lifetime fluence of reactor pressure vessel employed in LWR.

Time-resolved secondary triton burnup 14 MeV neutron measurement by a new scintillating fiber detector in middle total neutron emission ranges in deuterium large helical device plasma experiments

A middle-sensitivity scintillating fiber detector (hereafter middle Sci-Fi detector) that works at a deuterium-tritium neutron flux of ~105-107 cm−2s−1 was utilized to measure secondary deuterium-tritium neutron emission rates with high temporal resolution at a total neutron emission rate of 1013 to 1015 n/s, where strong magnetohydrodynamic (MHD) instabilities were observed in the large helical device deuterium plasma experiments. The gain and angular characteristics of the middle Sci-Fi detector were evaluated in an accelerator-based deuterium-tritium neutron source in the intense 14 MeV neutron source facility at Osaka University. Observation of 1 MeV triton transport due to MHD instability was performed by a middle Sci-Fi detector whose deuterium-tritium neutron counting rate was approximately 20 times higher than that of the conventional Sci-Fi detector. Fusion-born triton transport due to energetic-particle-driven MHD instability was observed using the middle Sci-Fi detector due to its high detection efficiency and high discrimination ability of deuterium-tritium neutrons from the sea of deuterium-deuterium neutrons.

A study of beam ion and deuterium–deuterium fusion-born triton transports due to energetic particle-driven magnetohydrodynamic instability in the large helical device deuterium plasmas

Understanding energetic particle transport due to magnetohydrodynamic instabilities excited by energetic particles is essential to apprehend alpha particle confinement in a fusion burning plasma. In the large helical device (LHD), beam ion and deuterium–deuterium fusion-born triton transport due to resistive interchange mode destabilized by helically-trapped energetic ions (EIC) are studied employing comprehensive neutron diagnostics, such as the neutron flux monitor and a newly developed scintillating fiber detector characterized by high detection efficiency. Beam ion transport due to EIC is studied in deuterium plasmas with full deuterium or hydrogen/deuterium beam injections. The total neutron emission rate (S n) measurement indicates that EIC induces about a 6% loss of passing transit beam ions and a 60% loss of helically-trapped ions. The loss rate of helically-trapped ions, which drive EIC, is larger than the loss rate of passing transit beam ions. Furthermore, the drop of S n increasing linearly with the EIC amplitude shows that barely confined beam ions existing near the confinement-loss boundary are lost due to EIC. In full deuterium conditions, a study of deuterium–deuterium fusion-born triton transport due to EIC is performed by time-resolved measurement of total secondary deuterium–tritium neutron emission rate (S n_DT). Drop of S n_DT increases substantially with EIC amplitude to the third power and reaches up to 30%. The relation shows that not only tritons confined in confined-loss boundary, but also tritons confined in the inner region of a plasma, are substantially transported.

Analysis of radiation emission from MYRRHA spent fuel and implications for non-destructive safeguards verification

The radionuclide composition of, and emitted radiation in, spent nuclear fuel from the future MYRRHA facility have been studied using depletion simulations to understand potential consequences for safeguards verification using non-destructive assay. The simulations show that both the gamma-ray and neutron emission rates in spent MYRRHA assemblies are lower than in spent PWR UO2 and MOX assemblies. In addition, gamma-ray emission rates from 134Cs and 154Eu are considerably lower, and the total neutron emission rate in MYRRHA fuel is much less sensitive to fuel burnup and cooling time. The main reason is that the fast neutron spectrum in MYRRHA affects the radionuclide production in the fuel. One result is that 244Cm, the main contributor to the neutron emission in spent light water reactor fuel, has a limited production in MYRRHA. Consequently, neutron-detection techniques could be used to more directly assay the plutonium content of spent MYRRHA fuel.

Determination of a Pu-Be source neutron spectrum at Czech Metrology Institute

Contrary to the radionuclide neutron sources of 252Cf and 241Am-Be type, which are most widely used for the dosimetric instrumentation calibrations, the spectral source strength and spectrum averaged fluence to dose equivalent conversion coefficients of Pu-Be source are not recommended by an international standard. This work describes the determination of those parameters for the Pu-Be source used at Czech Metrology Institute (CMI) by means of a Bonner spectrometer in combination with neutron emission rate measurement in a manganese bath.

Neutron emission and fast ion simulation for high performance long pulses at EAST.

Neutron emission spectroscopy and neutron yield measurements are important for high neutral beam injection (NBI) power heating at the Experimental Advanced Superconducting Tokamak (EAST). The neutron yields mainly depend on the deposition from NBI to the deuterium plasmas in the EAST. We have recently used TRANSP with time dependent diagnostic results to simulate the transport process of 30s long pulse deuterium plasma discharges in the EAST, obtaining the time dependent fast ion distribution, neutron emission spectrum, and total neutron emission rate. Combined with the time trace of the result measured by a standard 235U fission chamber, the effects of different configurations of NBI heating in EAST fusion plasmas have been evaluated.

Sensitivity of the active neutron coincidence collar response during simulated and experimental fresh fuel assay

Verification of the fissile (and fertile) content in fresh nuclear fuel assemblies is conducted by the IAEA to enforce the Nuclear Non-Proliferation Treaty using the UNCL (Uranium Neutron Collar — Light Water Reactor Fuel). The UNCL uses an uncorrelated AmLi neutron source to interrogate the fuel, producing a signal of coincident fission neutrons (Doubles rate) used to assay 235U content of the fuel. The cost of producing calibration assemblies and limited availability of diverse commercial assemblies at any one time historically restricted the ability to explore the full parameter space experimentally. Monte Carlo simulations can overcome this, but introduce additional sources of uncertainty. Here the sensitivity of simulations and measurements to various parameters is assessed for a reference 16×16 PWR assembly of uniform 3.2% enrichment. Uncertainty contributions in this evaluation include: simulated AmLi neutron source spectrum, AmLi neutron emission rate, AmLi anisotropicity, high density polyethylene (HDPE) density, the precise position of the fuel assembly within the detector, and experimentally the statistical uncertainty. The overall total systematic uncertainty estimate for the simulation of the absolute/relative Doubles rates responses are estimated to be approximately 2.0%/1.5%, and for experimental measurements systematic uncertainty reduces to 1.1%. This analysis supports further work using the relative Doubles rates in place of measurements for updating and extending the UNCL analysis methodology as systematic uncertainty is reasonably small.

Characterization of the New Vertical Neutron Camera Designed for the Low Neutron Emission Rate Plasma in Large Helical Device

Characterization of the New Vertical Neutron Camera Designed for the Low Neutron Emission Rate Plasma in Large Helical Device

Development of a Wide Dynamic Range Neutron Flux Measurement Instrument Having Fast Time Response for Fusion Experiments

A wide-range neutron flux measurement instrument is developed herein for monitoring the total neutron emission rate and yield of the Large Helical Device (LHD) during deuterium experiments implemented from March 2017 in the National Institute for Fusion Science (NIFS), Japan. The instrument is designed for and installed on the Neutron Flux Monitoring (NFM) system, which measures the counting rate using a 235U Fission Chamber. By combining the pulse counting and Campbell methods, the Digital Signal Processing Unit (DSPU) realized a wide dynamic range of over six orders of magnitude from 1 × 103 counts/s (cps) to 5 × 109 cps. This study explains and discusses how the instrument is developed, including topics from the predevelopment activities to the verification test at the Kyoto University Critical Assembly (KUCA). Experimental results in the LHD using the finished products suggest that the NFM system works well during deuterium experiments.

Extending the total neutron emission rate of steady-state deuterium large helical plasma guided by a data-driven approach

A trial for extending the total neutron emission rate (Sn) was performed in the Large Helical Device (LHD) deuterium plasma experiment using a gas puff in order to show the way to realize a steady-state fusion reactor. In the 22nd experimental campaign of the LHD, a high Sn discharge was performed by the experimental scenario guided by a data-driven approach. A regression analysis of Sn was conducted beforehand using the externally controllable parameters based on the database of the previous three campaigns. Then, the high Sn record was successfully updated to be 3.7 × 1015 n/s in a steady-state discharge during the 22nd campaign in line with the regression expression. This study demonstrates that such a data-driven approach is effective for the gradual extension of Sn beyond the employed database.