Neutron Flux Gradients Research Articles

Gold content determination in small core-samples by instrumental neutron activation analysis

Application of instrumental neutron activation analysis (INAA) based on the relative method of concentration standardization was considered to determine Au content in small solid core-samples of the gold-bearing ores and rocks up to 10 g of the mass. The small core-samples about 10 mm in diameter and 22-23 mm in length were cut from the previously collected ore lumps or in-situ using a handheld drilling rig. The studied small core-samples and the polyethylene capsules of the same dimensions filled up with the corresponding certified reference material (CRM) were irradiated for 2 min by a lower density neutron flux with the help of the automated pneumatic transport system (PTS). Maximum mass and dimensions of the core-samples corresponded to the PTS design and to the conditions of its safe operation. A special device was made to fix the transport capsules in the stable counting geometry making possible to eliminate the influence of the neutron flux gradient during irradiation. Due to the substantial differences in CRM and solid sample densities, corrections for neutron self-shielding and for gamma-ray self-absorption by the core-samples were applied. The method was tried to analyze Au content in 170 small core-samples presenting different gold-bearing ores and country rocks from the Kazakhstan’s gold-barite-polymetallic deposit Maikain. This approach, rather simple methodically and requiring no unique equipment, can be used to assess gold resources together with the methods of geostatistics.

On the use of neutron flux gradient with ANNs for the detection of diverted spent nuclear fuel

One of the main tasks in nuclear safeguards is regular inspections of Spent Nuclear Fuel (SNF) assemblies to detect possible diversions of special nuclear material such as 235U and 239Pu. In these inspections, characteristic signatures of SNF such as emissions of neutrons and gamma rays from the radioactive decay, are measured and their consistency with the declared assemblies is verified to ensure that no fuel pins have been removed. Research in this field is focused on both the development of detection equipment and methods for the analysis of the acquired measurement data. In this paper, the use of the neutron flux gradient, which is not considered in regular SNF verification, is investigated in combination with the scalar neutron flux as input to artificial neural network models for the quantification of fuel pins in SNF assemblies. The training and testing of these ANN models rely on a synthetic dataset that is generated from Monte Carlo simulations of a typical intact pressurized water reactor assembly and with different patterns of fuel pins replaced by dummy pins. The dataset consists of unique scenarios so that the ANN can be assessed over “unknown” cases that are not part of the learning phase. Results show that the neutron flux gradient is advantageous for a more accurate reconstruction of diversions within SNF assemblies.

Radwaste analysis and package optimization with the F4E-radwaste tool and application to the ITER IVVS

A methodology to evaluate the neutron activation, categorize as radwaste, generate radiological checklists, and optimize the packaging for disposal of activated components according to French regulations is presented. The methodology obtains results like specific activity, dose rates and radwaste relevant parameters in a fine 3D mesh superimposed over the geometry of interest and can handle the typically complex models common in the nuclear fusion analyses like those of the ITER project. State-of-the-art radiation transport and activation codes are used along with a purposefully developed Python tool (F4E-radwaste), which performs the post-processing of the large amount of data generated and includes a graphical user interface. The 3D nature of the results allows the organization of packages of activated materials in a way that minimizes the amount of mass classified with higher radwaste levels and is subject to more stringent regulations and expensive disposals. An application case study of the methodology is presented for the In-Vessel Viewing System (IVVS) ITER system. The IVVS is especially suited as an example for this type of analysis as its elongated shape pointing towards the plasma results in a significant gradient of neutron flux and activation levels, and therefore radwaste categories. It is demonstrated how the methodology enables the reduction of the total higher radwaste level mass by optimizing the packaging of the IVVS radwaste.

Analysis of fluctuations in ex-core neutron detector signal in Krško NPP during an earthquake

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

Developments of the k0-NAA measurement model implemented in k0-INRIM software

The k0-INRIM software is a computer program that was recently developed to automatically evaluate combined uncertainty while performing mass fraction measurements adopting the k0-standardization method of Neutron Activation Analysis. In this paper, significant developments of the adopted measurement model, following a complete revision of the detector characterization procedure, are reported. In particular, the efficiency ratio between monitor and analyte γ-emissions accounts for conversions between counting positions, extended sample geometry and self-absorption. In addition, true-coincidence summing, neutron flux gradient, moisture and blank corrections are included in the model. The developed measurement model is implemented in the latest k0-INRIM release; this greatly improves flexibility of the software in most of the common situations encountered in routine analysis without losing its inherent characteristic focused on uncertainty evaluation through propagation of covariances via sensitivity coefficients. A performance test was carried out by measuring the Au mass fraction of a cylindrical sample prepared from electronic waste material. Results obtained by counting the sample at different positions with respect to the detector end-cap are presented.

Investigation and Validation of Unstructured Mesh Methodologies for Modeling Experimental Reactors

This paper summarizes a methodology developed at École Polytechnique Fédérale de Lausanne for the neutronic modeling of the CROCUS experimental reactor and proposes solutions to the challenges one may face while modeling a research reactor with a complex geometry. Indeed, the double-lattice configuration of CROCUS makes it difficult to use codes for neutron diffusion and transport relying on a structured mesh description. For this reason, and based on the available in-house competences, we decided to make use of the neutronic capabilities of the GeN-Foam multiphysics solver, which takes advantage of general finite volume methodologies on unstructured meshes to provide sufficient flexibility for the study of unconventional reactor designs. In this work, GeN-Foam is used to build a first SP3 model of CROCUS based on an unstructured mesh to have an explicit modeling of the double lattice and the water gap between the two lattices. Form functions are then used to reconstruct the intra-pin fission rates for validation against measured distributions. We also discuss the limitations of the SP3 approximation of neutron transport in regions with steep neutron flux gradients and the planned future developments.

Conceptual design and initial evaluation of a neutron flux gradient detector

Identification of the position of a localized neutron source, or that of local inhomogeneities in a multiplying or scattering medium (such as the presence of small, strong absorbers) is possible by measurement of the neutron flux in several spatial points, and applying an unfolding procedure. It was suggested earlier, and it was confirmed by both simulations and pilot measurements, that if, in addition to the usually measured scalar (angularly integrated) flux, the neutron current vector or its diffusion approximation (the flux gradient vector) is also considered, the efficiency and accuracy of the unfolding procedure is significantly enhanced. Therefore, in support of a recently started project, whose goal is to detect missing (replaced) fuel pins in a spent fuel assembly by non-intrusive methods, this idea is followed up. The development and use of a dedicated neutron detector for within-assembly measurements of the neutron scalar flux and its gradient are planned. The detector design is based on four small, fiber-mounted scintillation detector tips, arranged in a rectangular pattern. Such a detector is capable of measuring the two Cartesian components of the flux gradient vector in the horizontal plane. This paper presents an initial evaluation of the detector design, through Monte Carlo simulations in a hypothetical scenario.

RayActive: An all-in-one R2S neutron induced activation analysis tool based on new intrinsic resolution methods.

Neutron production in nuclear facilities can induce a SDDR (Shutdown Dose Rate) even after months. Different methods exist to simulate the SDDR of neutron activated materials. The most used method is the mesh-based R2S (Rigorous-Two-Steps). This method couples a Monte-Carlo code and an isotopic inventory code to compute the SDDR. The Monte-Carlo code computes the neutron flux on a user-defined mesh to take into account the neutron flux gradients in the cells. The aim of this article is to present RayActive, a prototype tool elaborated for computing the SDDR. RayActive is based on a modified R2S method with an almost exact identification of materials enclosed in each voxel of a superimposed mesh. The CAD based geometry definition for the model makes possible this identification. Moreover, RayActive is designed as an ”all-in-one” package without any other code coupling. RayActive is designed to be as easy to use as possible and each part of the computation has been optimized to minimize the computational time. Verification and validation of RayActive has been done on shutdown dose rate benchmarks made with FLUKA code and classical R2S process coupling MCNP and FISPACT codes. Good agreement was found for these benchmarks even if some differences occur for the validation benchmark. These discrepancies show the impact on the results of the modified R2S method employed in RayActive.

Pinwise Diffusion Solution of Partially MOX-Loaded PWRs with the GPS (GET PLUS SPH) Method

The generalized equivalence theory (GET) plus superhomogenization (SPH) [GET Plus SPH (GPS)] method, which is a new leakage correction method for the pin-by-pin reactor analysis of light water reactors, has been applied to benchmarks for partial loading of mixed oxide (MOX) fuel in pressurized water reactor (PWR) cores. In the GPS method, the pinwise, cross section–dependent SPH factors are parameterized as a function of normalized leakage, i.e., current-to-flux ratio. As partially MOX-loaded PWRs usually have a stiff gradient of neutron flux on nodal interfaces, the original GPS functions for UO2 cores are slightly modified to take into account the strong spectral interaction. To determine the coefficients of the GPS function, several colorset models are considered to obtain fitting data. In this work, the two-dimensional method of characteristics–based DeCART2D code is used for both colorsets and reference core calculations. The GPS method is implemented in an in-house, pin-by-pin diffusion solver with the pinwise coarse mesh finite difference method. To evaluate the performance of the GPS method on partially MOX-loaded PWRs, the Korea Advanced Institute of Science and Technology (KAIST) 1A benchmark is analyzed in this work. In addition, various small and large variants of the KAIST 1A benchmark are also analyzed using the same GPS functions to demonstrate the general applicability of the predetermined GPS functions. Based on the comprehensive results of this work, it is concluded that the GPS method can clearly improve the accuracy of the conventional GET-based, two-step, pin-by-pin core analyses.

Influence of mass sample thickness on quantitative analysis of neutron activation and its modification

With traditional neutron activation analysis, the increase of sample quality leads to some problems in both irradiation process and measurement process. These problems include the neutron flux gradient in the range of the sample, the decrease of the thermal neutron flux rate around the sample and the influence of self-shielding in the sample in the irradiation process, In the process of measurement, the self-attenuation of <i>γ</i>-ray in the sample and the geometric effect of the sample lead to the effect of the detector on the measurement of characteristic <i>γ</i>-ray emissivity due to the difference in the detection efficiency of each point of the sample. So the neutron activation analysis of mass sample needs to make some additional modifications. By using the neutron activation analysis technique, the content of <sup>24</sup>Mg and <sup>28</sup>Si in a large amount of flour can be detected, and the content of talc powder in the flour can be given, so that the quality of flour can be monitored. The flour mainly contains C, H, O, N, Ca, and F element, but the main chemical constituent of talc powder is Mg<sub>3</sub>[Si<sub>4</sub>0<sub>10</sub>](OH)<sub>2</sub>. Therefore, the measured content of Si and Mg element can be used to judge whether the flour contains talcum powder and to determine its exact content. When the content of <sup>24</sup>Mg and <sup>28</sup>Si in flour are measured by the neutron activation analysis, the variation of neutron flux and energy with thickness in the measured sample and the effect of <i>γ</i>-ray self-absorption will have great influence on the measurement results. The relationship between the neutron flux and energy and the thickness of the sample is simulated by MCNP5 (Monte Carlo N-particle transport code system 5), and the neutron fluxes at different thickness of the sample are measured by a <sup>3</sup>He proportional counter tube. The results show that the simulation results of MCNP5 are basically consistent with the experimental results. Using the simulation by MCNP5 and the measurements by a sodium iodide detector, the relationship between <i>γ</i>-ray self-absorption effect and sample thickness is studied, and the sample thickness is determined to be 6.6.cm that is adopted as an optimal experimental condition. Based on the simulated data, the function relationship between the counting of 1.779 MeV <i>γ</i>-ray and the thickness of the sample is obtained as follows: <i>A</i> = 1401 + 3815<i>x</i> – 720<i>x</i><sup>2</sup> + 64<i>x</i><sup>3</sup> – 2.8<i>x</i><sup>4</sup> + 0.05<i>x</i><sup>5</sup>. The curve trend of the experimental results is basically the same as that of the simulation results.

Spatial distribution of neutron flux in geological larger sample analysis at CDTN/CNEN, Brazil

Iron monitor wire was used to evaluate the neutron flux gradients inside 5 g-geological sample during irradiation. The spatial distribution for thermal flux was about 2 %/cm for radial and axial gradients and about 5 %/cm for radial and 2.5 %/cm for axial gradients for fast neutrons. The results were evaluated using reference material IAEA-SOIL-7 analysed as small and larger samples and tested with calculating E n numbers. This test pointed out that all results were consistent within a 95 % confidence interval. The contribution of the neutron flux gradients may be reduced for larger sample using the standard procedure of the k 0-instrumental neutron activation analysis.

Transmission Electron Microscopy of 304-type Stainless Steel after Exposure to Neutron Flux and Irradiation Temperature Gradients

1. Department of Mechanical Engineering and Materials Science, University of Pittsburgh, Pittsburgh, PA, USA 2. Department of Materials , University of Oxford, Oxford, UK 3. Radiation Effects Consulting, Richland, WA, USA 4. Westinghouse Electric Company LLC, Pittsburgh, PA, USA 5. Westinghouse Electric Japan Ltd, Kobe, Japan 6. Nuclear Fuels Industries Ltd, Osaka, Japan 7. The University of Tokyo, Tokyo, Japan

Implementation of a methodology to analyse cylindrical 5-g sample by neutron activation technique, k 0 method, at CDTN/CNEN, Belo Horizonte, Brazil

Neutron activation analysis (NAA) is routinely applied to geometrical point-source or small samples because there are technical and theoretical difficulties to analyse larger samples weighing more than 0.5 g. The analysis of larger samples is very advantageous, because the analytical procedure will be less time consuming, it may be possible to reach lower detection limit for several elements, it decreases cost and overcomes the difficulties related to the representativeness of the sample when dealing with inhomogeneous volume or several small samples. Thus, increasing the amount of sample is a way to compensate for low flux of neutrons. This paper is about the establishment of a method at Laboratory for Neutron Activation Analysis, CDTN/CNEN, to determine the elemental concentrations in 5 g-samples, 25 times larger than usual samples analysed by neutron activation, k0 method, keeping the current irradiation and gamma spectrometry facilities. To develop this method, several aspects were evaluated such as detector efficiency over the volume source, neutron self-shielding during neutron irradiation, axial neutron flux gradient and gamma ray attenuation within the sample during counting. The results suggest that if an appropriate adjustment of the above mentioned parameters is done, the k0 method of NAA can provide satisfactory results also for larger samples than the samples typically used in NAA. The KayWin software proved to be a robust program analysing the larger samples weighing 5 g and cylindrical geometry as if it were a small cylindrical sample, producing reliable results. It was successfully implemented at Belo Horizonte, Brazil, fully following the basic principles of the k0 standardization method.

Power distribution gradients in WWER type cores and fuel failure root causes

Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs):Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores,Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp.,Neighboring baffle in WWER-1000 type cores, andNeighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part.An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

Re-evaluation of the irradiation conditions in the IFMIF test cell based on the EVEDA phase design

A re-evaluation of irradiation conditions in IFMIF utilizing the new EVEDA geometry model has been carried out. The focus is on the re-assessment of the nuclear responses in the High Flux Test Module (HFTM) as well as other Test Modules. The aim of the present analysis is to provide a reliable and realistic estimate based on the latest design and knowledge. To this end, we have utilized state-of-the-art nuclear data and an updated beam footprint as prepared on the basis of deuteron beam dynamics simulations combined with the Monte Carlo code McDeLicious, an enhancement to MCNP5 developed ad hoc for IFMIF neutronics calculations. The present analysis proves that (1) the averaged DPA rate of 26 dpa/y can be achieved in HFTM with 500cm3 volume, (2) the He/DPA value for HFTM center rigs is similar as the value expected for DEMO first wall, (3) the neutron flux gradient in HFTM center rigs is less than 10%/cm.

The effect of graphite thermal column around the newly designed large irradiation channel of GHARR-1 on neutron flux distribution

Monte Carlo N-Particle code (MCNP-5) was employed to simulate the neutron flux profile in a newly designed irradiation site surrounded by a graphite thermal column for Large Sample Neutron Activation Analysis, in the Ghana Research Reactor-1. The results shows that the average thermal neutron flux in the irradiation channel surrounded with 6cm thick graphite was 5.45×1010ncm−2s−1 as compared with 1.74×1010ncm−2s−1 when there is no graphite around the irradiation channel. This shows an increase in the thermal neutron flux in the irradiation channel by a factor of 3.13. The thermal neutron flux gradient decreased at 7.0×109ncm−2s−1 per cm to 3.90×107ncm−2s−1 per cm when the channel is surrounded by 6cm thick graphite which makes it very suitable for Large Sample Neutron Activation Analysis. The simulation agrees very well with the experimental keff of 1.00400 as compared with the final keff of 1.00390 recorded by this simulation.

Evaluation of measurement uncertainty components associated with results of radiochemical neutron activation analysis for determination of uranium traces

Being aware of the importance to consider every step in the evaluation of the combined measurement uncertainty of the result, the purpose of this work was to evaluate the contribution of the radial thermal neutron flux gradient to the uncertainty budget for trace level uranium determination in biological materials by a radiochemical neutron activation analysis (RNAA). Determination of uranium via the short-lived nuclide 239U was based on solvent extraction with TBP and measurement of the chemical yield from the gamma-ray spectrum of the isolated fraction via 235U. It has been shown previously, that radial neutron flux gradient, could have a relevant effect on the final result obtained by RNAA. In the present work, radial neutron flux gradient within the irradiation assembly generally accepted in our lab (standards tapped beside the sample), varied between 93 and 108% around the mean value and contributes approximately 20% to the combined measurement uncertainty of the result.

Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups diffusion theory. Successively, with the usage of a developed pin-power reconstruction technique capable to account for the innovative fuel assembly design, sub-channel investigations of the individual fuel assemblies have been performed evaluating pin-wise clad temperatures. Obtained results will be discussed giving a detailed insight of the revolutionary HPLWR 3 pass core concept and understanding the physical reasons, which influence the local clad temperatures. The obtained results represent a new quality in core analyses, which takes into full consideration the coupling between neutronics and thermal-hydraulics as well as the spatial effects of the fuel assembly heterogeneity in determining the local pin-power and the associated maximum clad temperature.

Intercalibration of international and domestic 40Ar/39Ar dating standards

Four international standards, Ga1550, MMhb-1, Lp-6, Bern 4M, and one domestic standard BT-1 have been intercalibrated. The repeated measurements on MMhb-1 with different mass demonstrate that MMhb-1 is inhomogeneous in age and its average age is 519.8 Ma. The results of Bern 4M and Lp-6 reflect that they have an invariable value of 40 Ar * / 39 Ar k (F) and the ages we obtained are consensus with their K-Ar age: Lp-6=127.7Ma; Bern 4M=18.2 Ma. Analyses of BT-1 age spectra, Ca/K and Cl/K spectra as well as inverse isochrons indicate that the sample is homogeneous and invariable and keeps close chemically, with its trapped argon isotope composition close to the atmosphere. The dating results show that age values are reproducible and steady, total fusion age, step-heating age, plateau age and isochron age are in accord with each other within the error range (2s). Therefore, we recommend 28.7 Ma as the calibrated age of BT-1. We also discuss the variation in neutron flux gradients of Beijing 49-2 reactor. It was found that the neutron flux gradient varies considerably, and more monitors (standard samples) are needed to fix the trend of variation. The coefficient of the 49-2 reactor that transfers the ratio of production rate of 37 Ar Ca / 39 Ar K into Ca/K ratio is 1.78. This is different from that reported earlier, 2.0, which may be caused by the reconstruction of the reactor.

Standardizing NAA of biological samples of varying size, shape and composition

Neutron activation allows the analysis of wet and dry biological tissues without sample preparation. To eliminate the need to prepare standards of each element to match each sample, we have standardized for a small number of typical sample types and irradiation and counting conditions. Then, for each sample analyzed, the concentration calculation includes corrections to the neutron flux and the counting efficiency, which depend on sample size and composition. The effects corrected include the neutron flux gradient in the irradiation site, the moderation of the neutrons with the hydrogen content of the sample and the variation of counting efficiency with sample volume.