Fission Neutrons Research Articles

Design and performance of an epithermal neutron detector based on PFN

In this paper, an epithermal neutron detector suitable for the Prompt Fission Neutron (PFN) uranium logging method is designed by Monte Carlo simulation. According to the simulation results, the epithermal neutron detector composed of a 1 mm cadmium (Cd) layer, a 5 mm high-density polyethylene (HDPE) layer, and a 3He tube is sensitive to epithermal neutrons. Through the simulation test of Monte Carlo in formations with different uranium content, the results show that the uranium content in the formation has an obvious linear relationship with E/T, and the detection limit can meet the minimum standard for uranium ore detection. It meets the design expectation and provides a basis for the optimal design of an epithermal neutron detector.

Matrix effect correction in active neutron interrogation for residual fissile mass assessment in radioactive waste drums

Metallic waste in the form of shells and nozzles remains following the reprocessing of spent fuel at La Hague plant (France). Before the final disposal, drums containing the waste are transported to the compaction facility where its volume is reduced by a factor of 5. With the objective of controlling the criticality/safety levels, an active neutron interrogation system at the entrance of the facility is used to assess the residual fissile mass remaining in the waste, which relies on prompt fission neutron detection (known as the differential die-away technique, DDA). The measured signal is proportionally linked to the fissile mass by a calibration coefficient. However, two effects can produce an inaccurate prediction. The number of induced fissions for the same fissile mass varies according to the waste matrix composition, particularly the neutron slow-down and absorption ratio. Secondly, the presence of fissile clusters can impact the neutron interrogation due to an increase in self-shielding effect. This work presents a matrix effect correction method based on the use of internal flux monitors (sensitive to the neutron absorption ratio) and the transmission signal (sensitive to the slow-down ratio). It relies on a numerical model of the whole measurement cell developed with the Monte-Carlo N-Particle transport (MCNP) code. The calibration coefficient of 72 different matrices, representative of the waste produced at La Hague, were simulated and a statistical model using a multilinear regression was then established. This simulated-based approach provides a more robust and comprehensive estimation of the residual fissile mass without impacting accuracy thanks to the flux monitors and the transmission signals.

Monitoring for Actinide Neutron Emissions From Spent Nuclear Fuel Under Extreme Gamma Fields Using Centrifugally Tensioned Metastable Fluid Detector Sensor Technology

Abstract This paper presents outcomes from research studies conducted independently at Purdue University as part of a collaborative project with the staff of Pacific Northwest National Laboratory (PNNL). Research work focused on assessments for meeting the challenge of monitoring actinide content in spent nuclear fuel (SNF) via characteristic neutron emissions [from spontaneous fission and (α,n) reactions] using the centrifugally tensioned metastable fluid detector (CTMFD) sensor technology using an inert organic decafluoropentane (DFP)(C5H2F10) sensing fluid. Traditional detectors readily saturate and/or cannot monitor neutron emissions under the expected 1012:1 gamma to neutron radiation environment. A challenge problem was posed to examine if a CTMFD could operate reliably over 1 h for conducting neutron spectroscopy at a 1 m standoff from a 30-y cooled SNF, in a ∼1012:1 gamma:neutron intensity environment resulting in a 150 Gy (15 kRad) accumulated dose for the CTMFD. The impacts on reliable operability were studied separately under expected gamma radiation energy and intensity for possible effects of: (i) radiolysis in the CTMFD sensing fluid from absorbed (<1.5 MeV) gamma radiation; (ii) photoneutron contamination signals from < 3 MeV high energy gamma photons interacting with the sensing fluid; and, (iii) malfunction of CTMFD component electronics from the absorbed gamma radiation. A Co-60 gamma irradiator was used for dose accumulation in the CTMFD electronic components and sensing fluids. A 14 MeV DT accelerator was used with a NaCl target to produce 3–4 MeV photons from activated 37S (via. neutron absorption in 37Cl) at SNF-commensurate intensities from SNF at 1-m standoff. Our examinations revealed the absence of any significant impact on CTMFD performance for meeting and exceeding the challenge problem metric. That is, we validated for no discernible impact of: 3–4 MeV gamma-produced photoneutrons when combined with a fission neutron source and radiolysis in the DFP sensing fluid through a 150 Gy absorbed dose. Past research results at Purdue University have validated for survivability of the key electronic components for absorbed gamma doses above the targeted 150 Gy level. This paper also provides extended evidence for survivability (from radiolysis) at higher gamma doses through 750 Gy with a borated DFP-sensing fluid formulation-based CTMFD.

Applicability to DEMO breeding blankets of neutron measurements techniques from fission reactors

Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement to the already planned diagnostic strategy.The breeding blankets are a key component of a fusion reactor such as DEMO. The expected performances and operational constraints are demanding, nevertheless some of those, as the neutron flux range and high temperature, are close to what is encountered in sodium-cooled fast neutron reactors.The strengths and drawbacks of neutron sensors are presented with respect to the performances and constraints. Self-powered neutron detectors and fission chambers benefit from a vast return of experiment in fission reactors and received recent developments in the framework of studies for sodium-cooled fast neutron reactors. The innovative optical fission chamber is discussed for its electromagnetic immunity and high dynamic range demonstrated up to seven decades. Dosimetry with activation foils is also considered for its ability to provide an absolute calibration and spectral unfolding.

Experimental study of the influence of borehole parameters on prompt fission neutron uranium logging and its corrections

In prompt fission neutron uranium logging, borehole environmental parameters affect the measured results and must be corrected. In order to explore the influence of borehole parameters on the interpretation of logging results, this paper builds a sandstone type uranium ore block model to simulate the field production drilling device based on the “Epithermal/Thermal neutron counting rate ratio” (E/T) theory. The effects of borehole diameter, thickness of iron tube and well fluid on the decay rate of epithermal and thermal neutrons and their uncertainty correction methods were investigated. The results show that the effect of borehole diameter on E/T is negligible. The iron tube thickness has a certain effect on the moderation and absorption of epithermal and thermal neutrons, and its E/T increases slightly with increasing thickness. The influence of iron tube thickness on E/T is corrected and the relative uncertainty is less than 5%. The well fluid thickness also affects the decay rate of epithermal and thermal neutrons, and its E/T follows the law of negative exponential attenuation. The influence of well fluid thickness on E/T is corrected and the relative uncertainty is less than 5%. This study provides technical guidance for field well survey of uranium deposit.

Neutronic Analysis for The Radial Direction Heterogeneous Core Configuration of GFR with Thorium Fuel

A neutronic analysis has been carried out for the radial direction heterogeneous core configuration in GFR with thorium nitride fuel to obtain the most optimal heterogeneous design has keff > 1 and excess reactivity below one percent. Neutronic analysis was performed using openMC version 0.13.2 based on open-source Monte Carlo simulation with ENDF/B-VII.1 data library. The first calculation is benchmarking openMC and SRAC where the largest error is 2.49% is caused by different data library. SRAC data library is JENDL 4.0. The second calculation is a homogeneous core configuration with 5% to 15% uranium-U233 fuel variation, the stability lies at 8% for the reference of heterogeneous core configuration. The heterogeneous core configuration was performed with three types of fuel percentage and five different geometry variations. Optimum design for geometry F1:F2:F3 = 1 ring:4 rings:1 ring using fuel type F1:F2:F3 = 6.5%:8%:9.5%. Optimum design for geometry F1:F2:F3 = 2 rings:2 rings:2 rings using fuel type F1:F2:F3 = 7.5%:8%:8.5%. Optimum design for geometry F1:F2:F3= 1 ring:2 rings:3 rings using fuel type F1:F2:F3=7%:8%:9%. Optimum design for geometry F1:F2:F3= 1 ring:3 rings:2 rings using fuel type F1:F2:F3 = 7.5%:8%:8.5%. Optimum design for geometry F1:F2:F3 = 1 ring:1 ring: 4 rings using fuel type F1:F2:F3 = 7.5%:8%:8.5%. The most optimum design among the other five designs is the F1:F2:F3 = 2 rings:2 rings:2 rings geometry design because it has a maximum keff of 1.00999 and excess reactivity of 0.99%. This design is carried out for an extended burn-up 15 years and still stable with the characteristics of the neutron flux and fission rate in this design decrease with burn-up time. Fission product in this design also decreased according to the purpose of the generation IV reactor is the prevention of nuclear weapons.

Microstructure evolution of γ-Al2O3/FeAl tritium permeation barrier coatings under 6.4 MeV Fe3+ ion irradiation

γ-Al2O3/FeAl coatings for tritium permeation barrier (TPB) application in fusion reactors were fabricated via electro-chemical deposition and selective oxidation. The TPB coatings consisted of a γ-Al2O3 scale (200 nm), an upper interlayer of FeAl (20 μm) and a lower interlayer of Fe3Al (10 μm). Their potential in-reactor performance was investigated based on 6.4 MeV Fe3+ ion irradiations at 400 °C, up to ion fluences of 1.0 × 1019 ions·m−2 and 1.0 × 1020 ions·m−2, at depth regions entailing the oxide scale and the top interlayer. Combined grazing incidence X-ray diffraction (GIXRD) and transmission electron microscopy (TEM) analysis confirmed the overall phase stability and grain structure stability in γ-Al2O3/FeAl coatings after irradiation. Cavities dominated the damage microstructure for both γ-Al2O3 and FeAl phases, but were not observed at the sub-layer interfaces. The underlying mechanisms for defect production were discussed. Finally, a brief comparison between heavy-ion and fission neutron damage effects in Al2O3-based TPB coatings is made, aiming to bring new insights on the coating integrity and performance reliability during fusion reactor operation.

A comparative analysis of gas-cooled fast reactor using heterogeneous core configurations with three and five fuel variations

GFR or Gas-cooled Fast Reactor is one type of fast generation-IV that uses a very high cooling temperature. Thus, it is necessary to have the right reactor core design so that the power distribution of neutrons produced reaches a safe and even limit point. The use of a uniform (homogeneous) reactor core can produce peaking power. This is very avoidable because it will cause a reactor accident. In this study, researchers tried to compare the results of the analysis for two heterogeneous reactor core designs including the configuration of 3 fuel variations and 5 fuel variations using UN-PuN fuel. This study aims to determine the keff value produced by both types of fuel variations during 5 years of burn-up and determine the characteristics of neutron flux, fission rate, and fission product during 15 years of burn-up. This study was started by calculating the homogeneous and heterogeneous core of 3 and 5 fuel variations with neutron transport simulation involving OpenMC. The calculation results show that the heterogeneous core configuration of 5 fuel variations for the keff value is more optimal than 3 fuel variations, because it has the smallest excess reactivity value. The neutron flux and fission rate characteristics for 5 fuel variations are more evenly distributed when compared to 3 fuel variations to maintain neutron lifetime and reactor life in operation. Burn-up residual plutonium material and minor actinide waste for 5 fuel variations have less mass than 3 fuel variations. The results of neutronic analysis of GFR reactors with heterogeneous reactor core designs for 5 fuel variations are better in terms of reactor criticality, neutron power distribution, and waste produced. Finally, optimization of the UN-PuN fuel volume fraction of 60 % provides the optimal keff value

Pencil beam kernel-based dose calculations on CT data for a mixed neutron-gamma fission field applying tissue correction factors.

Objective.For fast neutron therapy with mixed neutron and gamma radiation at the fission neutron therapy facility MEDAPP at the research reactor FRM II in Garching, no clinical dose calculation software was available in the past. Here, we present a customized solution for research purposes to overcome this lack of three-dimensional dose calculation.Approach.The applied dose calculation method is based on two sets of decomposed pencil beam kernels for neutron and gamma radiation. The decomposition was performed using measured output factors and simulated depth dose curves and beam profiles in water as reference medium. While measurements were performed by applying the two-chamber dosimetry method, simulated data was generated using the Monte Carlo code MCNP. For the calculation of neutron dose deposition on CT data, tissue-specific correction factors were generated for soft tissue, bone, and lung tissue for the MEDAPP neutron spectrum. The pencil beam calculations were evaluated with reference to Monte Carlo calculations regarding accuracy and time efficiency.Main results.In water, dose distributions calculated using the pencil beam approach reproduced the input from Monte Carlo simulations. For heterogeneous media, an assessment of the tissue-specific correction factors with reference to Monte Carlo simulations for different tissue configurations showed promising results. Especially for scenarios where no lung tissue is present, the dose calculation could be highly improved by the applied correction method.Significance.With the presented approach, time-efficient dose calculations on CT data and treatment plan evaluations for research purposes are now available for MEDAPP.

Research on fission and fusion neutron shielding performance of various materials

One of the most challenges in neutron shielding analyses is the selection of materials that protect the components of nuclear power plants and workers against radiation effects, including neutron and gamma fluxes. The purpose of this study is to examine the neutron reduction effect of various single and multi-element materials. To achieve that MCNP neutron transport and an analytical approach have been proposed to model experiments setup with complex geometry. The materials are inserted between neutron source and the water tank to reduce the neutron effect in the model. Firstly, fast neutron removal cross-section of each material has been calculated. Then, the surface flux tally has been performed to calculate neutron flux as a function of distance and energy for thermal and fast neutrons. The neutron fluxes are decreased depending on the distance and absorbing properties of each candidate. Regarding neutron attenuation performance, the materials have different neutron reduction abilities for each neutron range. Both MCNP and analytical results have been compared with experimental data. The obtained data for thermal and fast neutrons can be useful for the design and development of component and fission-fusion reactors.

A Simple Model of the Energy Threshold for Snowball Chambers

Cloud and bubble chambers have historically been used for particle detection, capitalizing on supersaturation and superheating, respectively. Here, we present new results from a prototype snowball chamber, in which an incoming particle triggers the crystallization of a purified, supercooled liquid. We demonstrate, for the first time, simulation agreement with our first results from 5 years ago: the higher temperature of the freezing of water and significantly shorter time spent supercooled compared to the control in the presence of a Cf-252 fission neutron source. This is accomplished by combining Geant4 modeling of neutron interactions with the Seitz nucleation model used in superheated bubble chambers, including those seeking dark matter. We explore the possible implications of using this new technology for GeV-scale WIMP searches, especially in terms of spin-dependent proton coupling, and report the first supercooling of WbLS (water-based liquid scintillator).

Prompt gamma rays of terbium induced by inelastic scattering of fission neutrons

AbstractPrompt gamma rays of terbium emitted after (n,nʹγ) inelastic scattering reactions induced by irradiation of a terbium(III) hexahydrate (TbCl3·6H2O) sample with a beam of fission neutrons were investigated with the instrument FaNGaS (Fast Neutron-induced Gamma-ray Spectrometry) at an angle of 90° between neutron beam and detector. At sample position, the fast-neutron flux was 1.13 × 108 cm−2 s−1and the neutron beam has an average energy of 2.30 MeV. We identified 124 prompt gamma lines from the159Tb(n,nʹγ)159Tb reaction. Presence of prompt gamma rays from oxygen and chlorine was used for a concise verification of recently published results. Relative gamma-ray intensities, effective cross sections and fast-neutron spectrum-averaged partial production cross sections of the gamma lines are given including comparisons with available literature data. We found a reasonable agreement and the multitude of unreported lines adds decisive value to nuclear spectroscopy. Additionally, we estimated the detection limit of terbium as 1 mg for a counting time of 12 h.

Determination of the absolute intensity of the 1205 keV [formula omitted]-ray emission from 91Y

A radiochemically pure solution of 91Y was produced by the thermal neutron fission of 235U followed by successive chemical separations to remove fission product impurities. The gamma emission rate of the 91Y 1205 keV gamma was measured using multiple high purity germanium gamma spectrometers previously calibrated for counting efficiency using a certificated mixed nuclide gamma standard. The activity concentration of the 91Y was subsequently standardised by liquid scintillation counting. From the combination the activity concentration and gamma emission intensity, the absolute intensity of the 1205 keV gamma emission was derived as 0.2297(39)%. This data agrees within the quoted uncertainties with the absolute intensity of 0.26(4)% published in nuclear data sheets A=91 (Baglin, 2013), but reduces the uncertainty by an order of magnitude.

Design and development of a neutron detection based novel criticality accident alarm system (CAAS) using liquid activation technique

Criticality Accident Alarm System (CAAS) is a mandatory system for prompt alarm and immediate evacuation of personnel during the unusual event of a criticality accident in all nuclear facilities using fissile materials. False criticality alarms are not desirable as they create a panic environment for radiation workers of the plant and the public. We have studied the use of different materials for such alarms and effective monitoring during such events. The present paper describes the design and development of a neutron detection based novel CAAS using the unique method of liquid activation technique of vanadyl sulphate solution, which can prevent false alarms. The neutronic properties of vanadium has been utilised for fast response. The performance evaluation of CAAS such as: testing, efficiency evaluation, threshold setting and various response studies using laboratory neutron sources including 252Cf fission neutron source, standard thermal neutron facility and reactor neutrons has been carried out by following the international recommendations and ensured the compliance. The results are also validated using Monte-Carlo FLUKA simulation. The concentration of vanadyl sulphate solution for liquid activation is optimized as 0.0082 g. cm−3 and the alarm threshold setting is evaluated as: 3881 CPS and 3.61 MBq of induced activity based on the current study. It is also worth mentioning that this liquid based system can be deployed in any reactor and fuel re-processing plants easily.

CGMF & FREYA Verification in Monte Carlo Code RMC

Monte Carlo simulation has become a crucial method internationally for simulating neutron multiplicity counting devices. This method requires sampling of fission neutron multiplicities, energies, and directions, making accurate simulation of fission events highly important. To meet this demand, the Fission Reaction Event Yield Algorithm (FREYA) and the Cascading Gammaray Multiplicity with Fission (CGMF) models have been integrated into RMC. These models can simulate individual fission events, preserving momentum, energy, and angular momentum, and then emit correlated particles. In addition to simulating fission neutrons, CGMF and FREYA models have been employed to compute fission photons, laying the foundation for subsequent photon multiplicity counting. Simultaneously, neutron multiplicity counting has been calculated using CGMF and FREYA, and point model equations have been used to invert Pu material, yielding favorable results.

Experimental setup of the 239Pu neutron capture and fission cross-section measurements at n_TOF, CERN

The experimental setup of the new measurement of 239Pu fission and capture cross-section in the n_TOF time-of-flight facility at CERN is presented. The measurement aims to address the needs and demands of nuclear data users. The experiment incorporates an innovative fast Fission Fragment Detector and the n_TOF Total Absorption Calorimeter, enabling the implementation of the fission tagging technique. Preliminary results exhibit the robust performance of the detector systems, along with the high quality of the new 239Pu samples. These samples were exclusively produced for this measurement by the European Commission’s Joint Research Centre in Geel.

Fuel Performance Comparison of Uranium Nitride and Uranium Carbide in VVER-1200 using OpenMC

Nuclear power is a reliable and large-scale source of GHG-free electricity. This study asses the viability of ATF fuel of uranium nitride (UN) and uranium carbide (UC) as fuel for the VVER-1200 reactor. A comprehensive overview of the VVER-1200 and Accident Tolerant fuels is conducted. A review of the development of ATFs identified UN and UC as viable fuels for the VVER reactor. The study utilizes OpenMC to model the VVER-1200 core and compares the behaviour of ATF with conventional fuel. Key findings include comparable k-eff values implying similar neutronic behaviour. UO2 and UC showed similar fission rates across the core while UN showed higher neutron flux and fission rate in the outer part of the core. The base Z44B2 showed increased flux and fission rate with UN as the fuel. ATF behaviour showed to be comparable to the UO2 and thus is a potential alternative to conventional fuels. ATFs provide an additional level of safety because of higher melting points and higher thermal conductivity. This study can be further improved to investigate the depletion of ATFs so that the behaviours of the core over large periods of time, fission products and operator safety can be assessed. Base case k-eff value of 1.24795 are comparable to k-eff values generated by UN and UC.

The effect of dominance ratio on the statistical convergence of sensitivity in Monte Carlo codes

Sensitivity computation in Monte Carlo-based codes is widely used and has been enhanced significantly. Although sensitivity converges in most reactor configurations, it encounters difficulties in large core designs. We suspect that this issue might be linked to the dominance ratio. To test this hypothesis, we develop a simple benchmark to validate whether statistical convergence of sensitivity depends on dominance ratio, and if it may be linked to other factors, such as neutron energy spectrum. The benchmark’s simplicity enables us to calculate the dominance ratio analytically and use eigenmodes decomposition to compute sensitivity to total fission neutron yields, substantially lowering computational costs. This method results in a good match when compared to the direct method which serves as a reference. We clearly see regular sensitivity convergence speed behavior linking it with the number of latent generations and the dominance ratio. Therefore, we establish a formula to recommend the number of latent generations required for sensitivity to converge, thus significantly saving computational resources.

Prompt-fission observable and fission yield calculations for actinides by TALYS

The nuclear reaction code TALYS adopts the Hauser-Feshbach statistical decay theory, to de-excite fission fragments. This involves for instance the evaporation of prompt fission neutrons and γ-rays. TALYS incorporates databases of primary fission fragment distribution which consists of primary fission fragment yield and data for excitation energy distribution of fission fragments. We conducted a sensitivity study on three parameters in TALYS and fitted them in order to reproduce experimental and evaluated data, in thermal neutron-induced fission of 235U. Moreover, we demonstrate a large-scale calculation of average prompt neutron and γ-ray multiplicities for 243 selected actinide isotopes.

Measurements of 235U(nth,f) Spectrum Averaged Cross Sections (SACS) in zero power reactors

This paper deals with the measurement of Spectrum Averaged Cross Sections in two different neutron fields formed in zero power reactors. The first was Benchmark Neutron Reference Field in the LR-0 reactor, and the second field was in the center of the vertical channel touching the fuel in the VR-1 reactor. The spectrum averaged cross section differs for both cases as the spectra differ, but after normalization to 235U PFNS using calculated correction, both results are in good agreement, thus confirming the spectra in both cases are similar in the 1 – 14 MeV region. A good agreement between lower threshold reactions averaged in actual reactor spectra and prompt fission neutron spectrum of 235U is reported as well.