| 1. |
- al-Dbissi, Moad, 1994, et al.
(författare)
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Identification of diversions in spent PWR fuel assemblies by PDET signatures using Artificial Neural Networks (ANNs)
- 2023
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 193
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Tidskriftsartikel (refereegranskat)abstract
- Spent nuclear fuel represents the majority of materials placed under nuclear safeguards today and it requires to be inspected and verified regularly to promptly detect any illegal diversion. Research is ongoing both on the development of non-destructive assay instruments and methods for data analysis in order to enhance the verification accuracy and reduce the inspection time. In this paper, two models based on Artificial Neural Networks (ANNs) are studied to process measurements from the Partial Defect Tester (PDET) in spent fuel assemblies of Pressurized Water Reactors (PWRs), and thus to identify at different levels of detail whether nuclear fuel has been replaced with dummy pins or not. The first model provides an estimation of the percentage of replaced fuel pins within the inspected fuel assembly, while the second model determines the exact configuration of the replaced fuel pins. The two models are trained and tested using a dataset of Monte-Carlo simulated PDET responses for intact spent PWR fuel assemblies and a variety of hypothetical diversion scenarios. The first model classifies fuel assemblies according to the percentage of diverted fuel with a high accuracy (96.5%). The second model reconstructs the correct configuration for 57.5% of the fuel assemblies available in the dataset and still retrieves meaningful information of the diversion pattern in many of the misclassified cases.
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| 2. |
- al-Dbissi, Moad, 1994, et al.
(författare)
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On the use of neutron flux gradient with ANNs for the detection of diverted spent nuclear fuel
- 2024
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 204
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 3. |
- Alhassan, E., et al.
(författare)
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Bayesian updating for data adjustments and multi-level uncertainty propagation within Total Monte Carlo
- 2020
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Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 139
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Tidskriftsartikel (refereegranskat)abstract
- In this work, a method is proposed for combining differential and integral benchmark experimental data within a Bayesian framework for nuclear data adjustments and multi-level uncertainty propagation, using the Total Monte Carlo method. First, input parameters to basic nuclear physics models implemented within the TALYS code, were randomly varied to produce a large set of random nuclear data files. Next, a probabilistic data assimilation was carried out by computing the likelihood function for each random nuclear data file based first on only differential experimental data and then on integral benchmark data. The individual likelihood functions from the two updates were then combined into a global likelihood function. The proposed method was applied for the adjustment of n+Pb-208 in the fast energy region below 20 MeV. The adjusted file was compared with available experimental data as well as evaluations from the major nuclear data libraries and found to compare favourably.
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| 4. |
- Börjesson Sandén, Fredrik, 1995, et al.
(författare)
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Effects of boric acid on volatile tellurium in severe accident conditions
- 2024
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 200
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Tidskriftsartikel (refereegranskat)abstract
- Boric acid is used in light-water nuclear reactors to control the reactor and is expected to be present as part of the chemistry of a severe accident. Therefore, its influence on other prominent species expected in an accident must be investigated. One such species is tellurium. In the present study, tellurium is volatized, and boric acid is dissolved and injected into the system as a means of studying the interaction between it and tellurium. The experiments were evaluated with ICP-MS and XPS. Results suggest that while there is no direct interaction, boric acid still affects the tendency for tellurium to oxidize. In general, less oxidation was detected in the presence of boric acid than in its absence, especially at high temperatures. The species formed upon oxidation was determined to be TeO2. Since tellurium metal is more volatile than TeO2, this may have implication in a wider severe accident context.
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| 5. |
- Chan, Yi Meng, et al.
(författare)
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A deep-learning representation of multi-group cross sections in lattice calculations
- 2024
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 195
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Tidskriftsartikel (refereegranskat)abstract
- To compute few-group nodal cross sections, lattice codes must first generate multi-group cross sections using continuous energy cross-section libraries for each material in each fuel cell. Since the processing cost is significant, we propose representing the multi-group cross sections during lattice calculations using a pre-trained deep-learning-based model. The model utilizes a combination of Principal Component Analysis (PCA) and fully connected Neural Networks (NN). The model is specifically designed to manage extensive multi-group cross-section data sets, which contain data for several dozen nuclides and encompass more than 50 energy groups. Our testing of the trained model on a PWR assembly with a realistic boron letdown curve revealed an average relative error of around 0.1% for both fission and total macroscopic cross sections. The average time required for the model to generate the cross sections was approximately 0.01% of the time needed to execute the cross-section processing module.
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| 6. |
- Chen, Yangli, et al.
(författare)
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Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds
- 2022
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 165
-
Tidskriftsartikel (refereegranskat)abstract
- The cornerstone of severe accident strategy of Nordic BWRs is to flood the reactor cavity for the long-termcoolability of an ex-vessel debris bed. As a prerequisite of the long-term coolability, the hot debris bedformed from fuel coolant interactions (FCI) should be quenched. In the present study, coupling of theMELCOR and COCOMO codes is realized with the aim to analyze the quench process of an ex-vessel debrisbed under prototypical condition of a Nordic BWR. In this coupled simulation, MELCOR performs an integralanalysis of accident progression, and COCOMO performs the thermal–hydraulic analysis of the debrisbed in the flooded cavity. The effective diameter of the particles is investigated. The discussion on thebed’s shape shows a significant effect on the propagation of the quench front, due to different flow patterns.Compared with MELCOR standalone simulation, the coupled simulation predicts earlier cavity poolsaturation and containment venting.
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| 7. |
- Chen, Yangli, et al.
(författare)
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Uncertainty quantification for TRACE simulation of FIX-II No. 5052 test
- 2020
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 143
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Tidskriftsartikel (refereegranskat)abstract
- The Best Estimate Plus Uncertainty approach requires the knowledge of input uncertainties for the uncertainty propagation with best-estimate codes. Inaccurate judgement of some model parameter uncertainties related to the dominant physical phenomena may result in misestimation of the safety margin. This paper presents a framework of inverse uncertainty quantification (UQ) to quantify model parameter uncertainties in order to address this issue. It is applied to TRACE simulation of a large break loss of coolant accident conducted on the FIX-II facility, and peak cladding temperature (PCT) is the simulation output. Sensitivity analysis identifies the parameters of the critical flow model as the most influential to the PCT. The inverse UQ is performed based on Bayesian framework, which adopts Markov Chain Monte Carlo sampling and surrogate modelling algorithms. The quantified uncertainties of the model parameters are the desired results from the inverse UQ process, which are useful in BEPU studies.
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| 8. |
- Dehlin, Fredrik, 1994-, et al.
(författare)
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An analytic approach to the design of passively safe lead-cooled reactors
- 2022
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 169, s. 108971-108971
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Tidskriftsartikel (refereegranskat)abstract
- A methodology to assist the design of liquid metal reactors, passively cooled by natural circulation duringoff-normal conditions, is derived from first principle physics. Based on this methodology, a preliminarydesign of a small LFR is accomplished and presented with accompanying neutronic and reactor dynamiccharacterizations. The benefit of using this methodology for reactor design compared to other availablemethods is discussed.
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| 9. |
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| 10. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Development and test of a novel verification scheme applied to the neutronic modelling of Molten Salt Reactors
- 2022
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 167
-
Tidskriftsartikel (refereegranskat)abstract
- This paper presents the extension of a method to verify transient neutron transport solvers earlier developed for reactors with non-moving fuel, to the case of Molten Salt Reactors (MSRs). This method is based on the extraction of the point-kinetic response of a nuclear reactor excited by a mono-chromatic perturbation and on its subsequent comparison with its expected functional dependence. Whereas a simple expression for this dependence exists for systems with fixed fuel, this is not the case for MSRs, as highlighted in many past studies. A workaround is nevertheless proposed in this work, thus giving the possibility to use a similar verification method to the case of MSRs. The method is applied to a simple dynamic MSR solver, demonstrating the capabilities of the technique. Contrary to other verification methods for which the system has to be simplified so that analytical solutions can be derived, the present method can be applied to any heterogeneous system.
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| 11. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Understanding the neutron noise induced by fuel assembly vibrations in linear theory
- 2022
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 175
-
Tidskriftsartikel (refereegranskat)abstract
- This paper investigates the underlying physical mechanisms involved in the monochromatic vibrations of fuel assemblies and their effects on the induced neutron noise throughout the core of nuclear reactors, in the framework of simplified benchmark configurations. Any vibrating fuel pin introduces noise sources at the frequency of vibrations, as well as at higher harmonics, the first one being the most significant of those. Depending on the harmonics considered, the position of the vibrating fuel pin, the size of the core and its macroscopic cross-sections, different noise responses are observed within the reactor core. Through the lens of a decomposition of the neutron noise into its point-kinetics component and its deviation from it, the spectrum of noise responses is explained and related to the spatial distribution of the amplitude and phase of the noise sources at the considered frequencies. At the frequency of vibration, possible out-of-phase behaviour of the induced neutron noise can be partially or totally shadowed by the in-phase point-kinetics component, the only exception being for central vibrations in symmetrical systems. At the frequency of the first higher harmonics, the structure of the induced neutron noise is more involved. Nevertheless, due to the compensation of the individual responses associated to the different components of the noise source at that frequency, point-kinetics has a significant contribution. The results of this work sheds new light on the complex spatial pattern of the neutron noise computed by realistic core simulators in case of vibrations of fuel assemblies.
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| 12. |
- Deng, Yucheng, et al.
(författare)
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An experimental study on the effect of coolant salinity on steam explosion
- 2024
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 201
-
Tidskriftsartikel (refereegranskat)abstract
- The steam explosion plays an essential role in the safety analysis of light water reactors (LWRs). Some studies have demonstrated that the occurrence of steam explosions is dependent on many factors such as melt and coolant temperatures, melt and coolant properties, non -condensable gases, etc. After the Fukushima accident, seawater as an emergency coolant and its impact on fuel coolant interactions are receiving attention. However, there is still little knowledge on the impact of seawater on steam explosion. The present study is intended to examine the effect of coolant salinity on steam explosion through a series of tests with single molten droplet falling in different coolant pools (DI water, and seawater at different salinities from 7.7 g/kg to 35 g/kg). The experimental results reveal that the salinity of coolant significantly influences the probability of spontaneous steam explosion of molten tin droplets. The probability of steam explosion generally increases with increasing salinity from 0 to 17.5 g/kg. The molten droplet in seawater experiences more pronounced deformation at same depth before the vapor film of the droplet collapses. What's more, the peak pressure generated by steam explosion in seawater is notably higher than that in DI water. The fragmentation of molten tin droplet after the explosion is enhanced accordingly.
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| 13. |
- Dufek, Jan, 1978-, et al.
(författare)
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Optimal time step length and statistics in Monte Carlo burnup simulations
- 2020
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 139
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Tidskriftsartikel (refereegranskat)abstract
- Monte Carlo burnup simulations continue to be seen as computationally very expensive numerical routines despite recent developments of associated methods. Here, we suggest a way of improving the computing efficiency via optimisation of the length of the time steps and the number of neutron histories that are simulated at each Monte Carlo criticality run. So far, users of Monte Carlo burnup codes have been required to set these parameters at will; however, an inadequate choice of these free parameters can severely worsen the computing efficiency. We have tested a large number of combinations of the free parameters on a simplified and fast solver, and we have observed that the computing efficiency was maximized when the computing cost of all Monte Carlo neutron transport calculations (summed over all time steps) was approximately comparable to costs of other procedures (all depletion simulations, the loading and processing of neutron cross sections, etc.). In this technical note, we demonstrate these results, and we also derive a simple theoretical model of the convergence of Monte Carlo burnup simulations that conforms to these numerical results. Here, we also suggest a straightforward way to automatise the selection of the optimal values of the free parameters for Monte Carlo burnup simulations.
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| 14. |
- Espegren, Fredrik, 1989, et al.
(författare)
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Potential tellurium deposits in the BWR containment during a severe nuclear accident
- 2020
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 146
-
Tidskriftsartikel (refereegranskat)abstract
- The release of fission products to the environment is one of the concerns with nuclear power. During an accident, the most likely released are the volatile fission products i.e., tellurium. To evaluate the behavior of tellurium in the event of an accident, it was heated under different conditions (oxidizing, inert, reducing; both dry and humidified). The formed vapor was transported to surfaces (aluminum, copper, zinc) at room temperature that can be found in the BWR-containment. All formed deposits were examined for morphology and species. Moreover, the content of sodium hydroxide liquid traps following the metal surfaces and filter was also investigated. In these traps, the highest amount of tellurium was found under humid-reducing followed by humid-oxidizing conditions. In the deposit removed from the zinc surface acquired under the latter conditions, elemental analysis observed zinc, indicating a possible reaction between tellurium and zinc. The corresponding trap showed significant amounts of zinc. (C) 2020 The Author(s). Published by Elsevier Ltd.
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| 15. |
- Estévez-Albuja, S., et al.
(författare)
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Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 136
-
Tidskriftsartikel (refereegranskat)abstract
- Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.
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| 16. |
- Galushin, Sergey, et al.
(författare)
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Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 135
-
Tidskriftsartikel (refereegranskat)abstract
- Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.
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| 17. |
- Herb, Joachim, et al.
(författare)
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Sensitivity analysis in core diagnostics
- 2022
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 178
-
Tidskriftsartikel (refereegranskat)abstract
- In the CORTEX project, methods to simulate neutron flux oscillations were enhanced and machine-learning based tools to determine the causes of measured neutron flux oscillations were developed, using the results of simulations as training and validation data. For a selected combination of those methods and tools, several sensitivity analyses were performed to assess their robustness and trustworthiness. The neutron flux oscillations were simulated using the tool CORE SIM+. It calculates the three-dimensional field of the neutron flux oscillations, which can be used to determine the response of neutron detectors at given locations. For the sensitivity analysis, the neutron flux oscillations were assumed to be caused by the vibration of one fuel element. It was investigated how selected input parameters like the core loading pattern, the burn up of the fuel elements, the neutronic core data, the geometry details of the vibrating fuel element, the chosen detectors, and other noise source parameters like the amplitude of the fuel element vibrations, affect the simulated neutron flux oscillations. A three dimensional fully convolutional neural network had been developed and trained during the CORTEX project to determine the cause and location of perturbations causing given measurements of in-core detectors in pressurized water reactors. The robustness of this network was tested by applying it to the simulated detector readings created during the sensitivity analysis.
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| 18. |
- Hoseyni, Seyed Mohsen, et al.
(författare)
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Melt infiltration through porous debris at temperatures above Solidification : Validation of analytical model
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 161, s. 108435-
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Tidskriftsartikel (refereegranskat)abstract
- This paper investigates the dynamics of melt infiltration through a preheated porous debris bed which is of importance to severe accident modeling in nuclear power plants. Proper understanding of the flow physics and affecting parameters is needed to define flow regime(s) according to combination of the driving forces, i.e. capillary and gravity. A model development and validation therefore should consider various effects and competing mechanisms. After a careful study of the governing equations and scaling rules, a known analytical model is validated against existing experimental data from REMCOD experiment. The predictions of this model are in good agreement with the experimental data. Furthermore, a global sensitivity analysis identifies the most influential parameters and reveals the need for further experiments with different range of affecting parameters. The results underline the importance of permeability as the most influential parameter.
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| 19. |
- Hou, Yandong, et al.
(författare)
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Effects of rolling motion on helical coil once-through steam generator thermal-hydraulic characteristics
- 2023
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 194, s. 110068-
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Tidskriftsartikel (refereegranskat)abstract
- Steam generators are essential for the safe, economical, and reliable operation of nuclear reactors. Helical Coil Once-through Steam Generators (HCOTSG) offer unique advantages over other types of steam generators for ship reactors or offshore platforms. To investigate the effect of rolling motion due to ocean conditions on the safe and economic operation of HCOTSG, a new model and a new code are developed in this paper. The program is developed and allows numerical simulation of HCOTSG under ocean conditions. The validity of the code was verified by comparing the simulation results with the design parameters of the Marine Reactor X (MRX) steam generator and the simulation results of other dedicated programs. The code was further used to perform the International Reactor Innovative and Secure (IRIS) transient operating conditions under typical rolling motions. Then, the influences of different swing directions, angles, periods, and positions of swing axes from the origin of the system operating parameters are analyzed. In the cases discussed in this paper, the following conclusions are obtained: (a) The direction of the swing significantly affects the system. The most dangerous situation is around the x-axis, and the case around the z-axis is the safest, in which the rolling situation has little effect on the system because the centripetal force is perpendicular to the tube wall, and the gravitational pressure drop is constant. (b) when the swing angle increases, except the fluctuation speed is faster, the fluctuation range of the parameters also increases. (c) when the swing period changes, the parameters’ fluctuation range also change (d) the orientation of the swing axis from the origin affects the magnitudes of the parameters’ changes (e) the distance of the swing axis from the origin affects the magnitude of the parameter variation and the farther the swing axis is from the origin, the greater the parameter fluctuation.
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| 20. |
- Hou, Yandong, et al.
(författare)
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Numerical study on surface corrosion deposition of fuel elements and its influence on flow heat transfer
- 2024
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 201
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Tidskriftsartikel (refereegranskat)abstract
- Corrosion of pressurized water reactors (PWR) in nuclear power plants can lead to serious safety hazards. This study aims to analyze the deposition of corrosion products using FLUENT software. Deposition models and thermal resistance models were developed, and the effects of deposits on the reactor's thermal–hydraulic characteristics were evaluated. Additionally, the impact of various parameters on deposition and thermal–hydraulic characteristics was examined. Results show that deposits accumulate extensively in the inlet section of the fuel cladding, while appearing as spot deposits in the outlet section. For deposit thicknesses below 30 μm, the surface temperature of the cladding gradually increases. However, when the thickness exceeds 30 μm, the surface temperature rapidly rises. Furthermore, the study reveals that the deposition amount decreases with increasing inlet flow velocity, exhibits an upward trend with higher inlet temperature, and increases with a higher wall heat flux density. This research provides important insights for understanding core deposition and thermal–hydraulic characteristics in nuclear reactor systems. It offers valuable guidance for enhancing safety and operational efficiency in nuclear power plants.
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| 21. |
- Hou, Yandong, et al.
(författare)
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Thermal-hydraulic characteristics of helical coiled once-through steam generators in inclining condition of ocean
- 2024
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 200
-
Tidskriftsartikel (refereegranskat)abstract
- This paper develops a one-dimensional thermal–hydraulic analysis program to simulate the main thermal–hydraulic parameter changes of helical coil once-through steam generator (HCOTSG) under inlet thermal–hydraulic parameters perturbation in inclining and vertical conditions. The inclining process is divided into the process of rotating from the vertical position to the inclining position and the stable inclining process. The rotating process is regarded as swaying motion. The swaying and inclining motions are achieved by modifying the momentum equation. What's more, the code was verified through experiments on two-phase heat transfer and friction pressure drop, as well as comparisons with the design parameters of the HCOTSG of the International Reactor Innovative and Secure (IRIS) reactor. The results of the simulation indicate that the direction of incline has an obvious impact on the safety of the HCOTSG, with inclination towards the y-axis having the greatest impact. In the stable inclining process, an increase in the angle of inclining results in a rise in primary-side outlet temperature and a reduction in heat transfer, while the secondary-side pressure drop increases. Furthermore, the HCOTSG underwent testing with eight types of transient perturbations, including four sudden perturbations and four linear perturbations. Results indicated that during sudden perturbations, apart from the secondary-side pressure drop, there were no substantial differences in simulated results between the inclined and vertical states at the same time during the transition process. However, during linear perturbations, due to the slow changes in parameters, the differences between the inclined and vertical states at the same time were more distinct. Regardless of the perturbation type, the inclined state led to a deterioration in the dynamic condition of the system compared to the vertical state.
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| 22. |
- Huang, Zi-Nan, et al.
(författare)
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Analysis of the stress field in the reactor vessel of the China Initiative Accelerator Driven System during postulated ULOF and UTOP transients
- 2023
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 194
-
Tidskriftsartikel (refereegranskat)abstract
- The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The subcritical reactor in CiADS is a liquid Lead Bismuth Eutectic (LBE) cooled fast reactor. When the reactor core is in operation, the LBE coolant will directly contact and corrode the inner surface of reactor vessel. Due to the high temperature, the corrosion will be more severe. If the stress on the reactor vessel exceeds the limit, the plastic deformation will occur, leading to the generation and expansion of defects and cracks, and the safety of the reactor will be affected. Therefore, evaluating the stress field of the reactor vessel under different operating conditions is a very important research project. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS, and the ASME Code was used as stress assessment standards. We can preliminarily prove that the stress assessments of the vessel during the postulated Unprotected Loss of Flow (ULOF) accidents satisfy the requirements of ASME Code. The limit reactivity insertion to protect the vessel from plastic deformation is 0.58$ in the postulated Unprotected Transient over Power (UTOP) accidents based on our current results. Therefore, we can preliminarily conclude that the current material selection and structural design of the reactor vessel in CiADS could survive most of the postulated transient accidents considering the stress effect.
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| 23. |
- Hursin, Mathieu, et al.
(författare)
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Modeling noise experiments performed at AKR-2 and CROCUS zero-power reactors
- 2023
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 194
-
Tidskriftsartikel (refereegranskat)abstract
- CORTEX is a EU H2020 project (2017-2021) devoted to the analysis of ’reactor neutron noise’ in nuclear reactors, i.e. the small fluctuations occurring around the stationary state due to external or internal disturbances in the core. One important aspect of CORTEX is the development of neutron noise simulation codes capable of modeling the spatial variations of the noise distribution in a reactor. In this paper we illustrate the validation activities concerning the comparison of the simulation results obtained by several noise simulation codes with respect to experimental data produced at the zero-power reactors AKR-2 (operated at TUD, Germany) and CROCUS (operated at EPFL, Switzerland). Both research reactors are modeled in the time and frequency domains, using transport or diffusion theory. Overall, the noise simulators managed to capture the main features of the neutron noise behavior observed in the experimental campaigns carried out in CROCUS and AKR-2, even though computational biases exist close to the region where the noise-inducing mechanical vibration was located (the so-called ”noise source”). In some of the experiments, it was possible to observe the spatial variation of the relative neutron noise, even relatively far from the noise source. This was achieved through reduced uncertainties using long measurements, the installation of numerous, robust and efficient detectors at a variety of positions in the near vicinity or inside the core, as well as new post-processing methods. For the numerical simulation tools, modeling the spatial variations of the neutron noise behavior in zero-power research reactors is an extremely challenging problem, because of the small magnitude of the noise field; and because deviations from a point-kinetics behavior are most visible in portions of the core that are especially difficult to be precisely represented by simulation codes, such as experimental channels. Nonetheless the limitations of the simulation tools reported in the paper were not an issue for the CORTEX project, as most of the computational biases are found close to the noise source.
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| 24. |
- Jansson, Peter, 1971-, et al.
(författare)
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A new methodology for thermal analysis of geological disposal of spent nuclear fuel using integrated simulations of gamma heating and finite element modeling
- 2022
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 172
-
Tidskriftsartikel (refereegranskat)abstract
- A new methodology is illustrated, where the evolution of temperature in a geological disposal system for spent nuclear fuel is estimated by integrated calculations of a spatially distributed gamma heating source with conventional finite element thermal transport modeling. A case with one canister loaded with fuel assemblies with a cooling time of 30 years in a KBS-3 type repository illustrates the methodology. For this particular case, the effect of including distributed gamma heating rate in the modeling has a small impact on the temperature distribution compared to the conventional case of heat generated locally in the canister, resulting in a small decrease of the maximum temperature in the canister. A large proportion of gamma heating occurs inside the outer boundary of the copper canister for this case. Other potential consequences of radiation escaping the canister are discussed.
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| 25. |
- Ju, Haoran, et al.
(författare)
-
LES and URANS study on turbulent flow through 3 x 3 rod bundle with spacer grid and mixing vanes using spectral element method
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 161, s. 108474-
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, the turbulent flow in rod bundles with spacer grid and mixing vanes in nuclear reactor core was studied using spectral element method with large eddy simulation (LES) and unsteady RANS (URANS) method. Combining with the radial periodical boundary conditions, the turbulent mixing phenomena in central subchannels were investigated, especially focusing on the flow field after the spacer grid and mixing vanes. Other than setting up long bare rod bundle to establish the fully developed turbulent flow, only a 2-cm-long bare rod bundle was extruded upstream the spacer grid and mixing vanes, while the fully developed laminar velocity profile was applied on the inlet surface for velocity boundary settings. The results indicate that both of the models could predict time-averaged velocity at 0.5Dh, 1Dh, 4Dh downstream mixing vanes properly, while the URANS model underestimate the root-mean-square (RMS) value of fluctuating velocity due to the inlet treatment effect. Furthermore, turbulent kinetic energy contour with transverse velocity vector was combined to investigate the behaviors of vortices formed by split-type mixing vanes, revealing the behavior of vortices for preliminary dynamic and mechanic analysis in design process of fuel assembly.
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| 26. |
- Karkela, T., et al.
(författare)
-
Tellurium retention by containment spray system
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 164
-
Tidskriftsartikel (refereegranskat)abstract
- A containment spray system is used to mitigate the source term from the containment building to the environment as part of the severe accident management actions. Tellurium is one of the volatile fission products and many of the tellurium isotopes decay into iodine, which causes a threat to the public due to its radiotoxicity and build-up in the thyroid gland. The removal efficiency of the containment spray system model against tellurium species formed under severe accident conditions was investigated with experiments and MELCOR simulations. The results indicated efficient removal of tellurium aerosols in the air atmosphere, whereas a decrease in the efficiency was observed in the nitrogen atmosphere. Gaseous tellurium species were not formed in significant amounts during the experiments and therefore, the removal efficiency due to different spray chemistry conditions could not be accurately analysed. However, the alkaline chemicals used in the spray solution seemed to form airborne particles, increasing the overall aerosol transport in the process independently of CsI or Te aerosol transport.
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| 27. |
- Kollias, Stefanos, et al.
(författare)
-
Machine learning for analysis of real nuclear plant data in the frequency domain
- 2022
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 177
-
Tidskriftsartikel (refereegranskat)abstract
- Machine Learning is used in this paper for noise-diagnostics to detect defined anomalies in nuclear plant reactor cores solely from neutron detector measurements. The proposed approach leverages advanced diffusion-based core simulation tools to generate large amounts of simulated data with different types of driving perturbations originating at all theoretically possible locations in the core. Specifically the CORE SIM+ modelling framework is employed, which generates these data in the frequency domain. We train using these vast quantities of simulated data state-of-the-art machine and deep learning models which are used to successfully perform semantic segmentation, classification and localisation of multiple simultaneously occurring in-core perturbations. Actual plant data are then considered, provided by two different reactors, including no labels about perturbation existence. A domain adaptation methodology is subsequently developed to extend the simulated setting to real plant measurements, which uses self-supervised, or unsupervised learning, to align the simulated data with the actual plant data and detect perturbations, whilst classifying their type and estimating their location. Experimental studies illustrate the successful performance of the developed approach and extensions are described that indicate a great potential for further research.
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| 28. |
- Mickus, Ignas, et al.
(författare)
-
Does neutron clustering affect tally errors in Monte Carlo criticality calculations?
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 155
-
Tidskriftsartikel (refereegranskat)abstract
- Monte Carlo criticality calculations of large, loosely-coupled problems are long known to suffer from slow convergence of the tally errors due to cycle-to-cycle fission source correlations. In several recent studies, it was suggested that these correlations could be possibly attributed to the neutron clustering phenomenon that is visible in calculations with a small number of neutrons per iteration cycle (batch size). Nevertheless, other studies have also shown the error convergence rate in such loosely-coupled problems to be batch size-independent during active criticality cycles. Here, we aim to address this inconsistency by studying the error convergence in a large number of test calculations, varying the neutron batch size from small to large. In our tests, we have observed that the presence of visible neutron clusters does not increase the cycle-to-cycle fission source correlations and does not worsen the convergence rate of the tally errors.
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| 29. |
- Mickus, Ignas, et al.
(författare)
-
Stochastic-deterministic response matrix method for reactor transients
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 140, s. 107103-
-
Tidskriftsartikel (refereegranskat)abstract
- Presented is a stochastic-deterministic, response matrix method for transient analyses of nuclear systems. The method is based on the response matrix formalism, which describes a system by a set of response functions. We propose an approach in which these response functions are computed during a set of Monte Carlo criticality calculations and are later used to formulate a deterministic set of equations for solving a space-time dependent problem. Application of the response matrix formalism results in a set of loosely connected equations, which leads to a favourable linear scaling of the problem. The method offers a simplified approach compared to previously proposed response matrix methods by avoiding phase-space expansions in sets of basis functions. We describe the method starting with the fundamental neutron transport considerations, provide a demonstration on two absorber movement transients in a 3 × 3 assembly PWR mini-core geometry, and compare the solutions against time-dependent Monte Carlo simulations.
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| 30. |
- Mylonakis, Antonios, 1987, et al.
(författare)
-
CORE SIM+: A flexible diffusion-based solver for neutron noise simulations
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 155
-
Tidskriftsartikel (refereegranskat)abstract
- The paper presents CORE SIM+, a tool developed for diffusion-based neutron noise simulations. The simulator is based on the 3-dimensional, two-energy group neutron diffusion equation in the frequency domain. The tool includes the necessary solvers to calculate the criticality problems associated with the system and then the response to a variety of perturbations such as absorbers of variable strength, perturbations travelling with the coolant flow and vibrations of core components. Numerical methods suitable for different types of problems have been implemented. A capability that allows to apply non-uniform computational meshes is available so that the discretization of the domain can be optimized with respect to the characteristics of the neutron noise sources. The simulator can generate neutron noise databases for nuclear power reactors via the Green’s function method in a fully automated manner. These databases can be useful when studying the neutron noise response in a reactor and when training machine learning algorithms for core monitoring and diagnostics. As part of the verification process, the CORE SIM+ response of a realistic reactor model to a given perturbation at various frequencies is used to estimate the corresponding point-kinetic component, compared to an exact analytical expression. In addition, the Green’s function generator is used to calculate the response of a system to a fuel cell vibration. The solution is then compared to the one of the direct solver of CORE SIM+. As part of the validation work, a neutron noise experiment with vibrations of fuel rods is simulated, showing a reasonable agreement with the measurements. A representative neutron noise database generated for a generic pressurized water reactor is described. From the database the simulations of fuel assembly and core barrel vibrations are chosen as illustrative examples for demonstrating the capabilities of the tool.
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| 31. |
- Mylonakis, Antonios, 1987, et al.
(författare)
-
Numerical solution of two-energy-group neutron noise diffusion problems with fine spatial meshes
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 140
-
Tidskriftsartikel (refereegranskat)abstract
- The paper presents the development of a strategy for the fine-mesh full-core computation of neutron noise in nuclear reactors. Reactor neutron noise is related to fluctuations of the neutron flux induced by stationary perturbations of the properties of the system. Its monitoring and analysis can provide useful insights in the reactor operations. The model used in the work relies on the neutron diffusion approximation and requires the solution of both the criticality (eigenvalue) and neutron noise equations. A high-resolution spatial discretization of the equations is important for an accurate evaluation of the neutron noise because of the strong gradients that may arise from the perturbations. Considering the size of a nuclear reactor, the application of a fine mesh generates large systems of equations which can be challenging to solve. Then, numerical methods that can provide efficient solutions for these kinds of problems using a reasonable computational effort, are investigated. In particular the power method accelerated with the Chebychev or JFNK-based techniques for the eigenvalue problem, and GMRES with the Symmetric Gauss-Seidel, ILU, SPAI preconditioners for the solution of linear systems, are evaluated with the computation of neutron noise in the case of localized perturbations in 1-D and 2-D simplified reactor cores and in a 3D realistic reactor core.
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| 32. |
- Pan, Qingquan, et al.
(författare)
-
Improved adaptive variance reduction algorithm based on RMC code for deep penetration problems
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 137
-
Tidskriftsartikel (refereegranskat)abstract
- An adaptive variance reduction algorithm based on RMC code was developed for deep penetration problems. The adaptive variance reduction algorithm uses the conservation of penetration rate and a predictor–corrector algorithm to quickly determine exponential importance parameters or equal-gradient importance parameters with fast iterative convergence. The adaptive variance reduction algorithm greatly accelerates one-dimensional deep penetration problems. However, the previous algorithm does not have capacity of energy bias nor does it support calculations on three-dimensional models. The adaptive variance reduction algorithm is improved to allow for energy bias and for three-dimensional models. The improved algorithm was applied to the HBR2 benchmark with the average Figure of Merit (FOM) of RMC code improved 243 fold. Therefore, this improved adaptive variance reduction algorithm can efficiently deal with complex engineering shielding problems.
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| 33. |
- Pasi, Anna-Elina, 1993, et al.
(författare)
-
Gas phase interactions between tellurium and organic material in severe nuclear accident scenarios
- 2024
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 197
-
Tidskriftsartikel (refereegranskat)abstract
- One of the volatile fission products potentially released during a severe nuclear accident is tellurium. Recently, tellurium has been shown to form volatile organic tellurides from paint solvents in aqueous solutions which could further increase the release. However, not much is known about the possible interactions between tellurium and organic material in gas phase. In this study, tellurium was exposed to conditions simulating the containment atmosphere during an accident. Moreover, volatile organic compounds were introduced to the gas phase representing the presence of organic material in the containment during the accident. The results suggest that tellurium aerosols and organic material interact in the gas phase which was observed as an increase in the gaseous fraction and a change in the XPS spectra. Although no exact species were identified, the results raise questions about the behavior of tellurium in severe accident conditions, especially, regarding the reactions involving organic material.
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| 34. |
- Pazsit, Imre, 1948
(författare)
-
Lenard Pal (1925-2019) Obituary
- 2021
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 154
-
Tidskriftsartikel (övrigt vetenskapligt/konstnärligt)
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| 35. |
- Pazsit, Imre, 1948, et al.
(författare)
-
Multiplicity theory beyond the point model
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 154
-
Tidskriftsartikel (refereegranskat)abstract
- Passive methods of nuclear safeguards determine the important parameters of an unknown sample from the statistics of the detection of the neutrons emitted from the item. These latter are due to spontaneous fissions and (α,n) reactions, enhanced by internal multiplication before leaking out. Based on the original work of Böhnel, the methodology of traditional multiplicity counting is based on the first three factorial moments of the number of neutrons, emitted from the sample due to one source event. These “Böhnel moments” were derived in the so-called “point model”, in which no space-dependence is accounted for, rather a uniform first collision probability is assumed for each neutron, irrespective of the position of its birth and its velocity direction, and, more important, it is assumed to be the same for all generations in the fission chain as for the source neutrons. The purpose of the present work is to derive the same factorial moments in a one-speed space-dependent model, in which the position and direction of the neutrons is accounted for, but (similarly to the original Böhnel model), no energy dependence is assumed. The integral equations for the moments are solved numerically with a collision number expansion. It is shown that compared to the space-dependent calculations, the unfolding method using the point model underestimates the fissile mass and the underestimation increases with increasing both of fissile mass and the value of α.
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| 36. |
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| 37. |
- Preston, Markus, et al.
(författare)
-
Analysis of radiation emission from MYRRHA spent fuel and implications for non-destructive safeguards verification
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 163, s. 108525-
-
Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 38. |
- Reale Hernandez, C., et al.
(författare)
-
Dynamic sensitivity and uncertainty analysis of a small lead cooled reactor
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 144
-
Tidskriftsartikel (refereegranskat)abstract
- A sensitivity and uncertainty analysis was performed on a small lead cooled reactor for two types of transients: an unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP). Transients were simulated with the code BELLA, which is a point-kinetics and lumped-parameter model. A Monte Carlo based method was used with 5000 simulations. Input parameters are reactor dimensions, neutronics properties, material properties and thermal hydraulic properties. Outputs are maximum temperatures (clad, coolant and fuel), mass flow disturbance, natural convection mass flow, maximum power and energy deposition. For ULOF, it was found that the most sensitive parameters were the gap between fuel and clad, the flow area in the core, the friction factors in core and steam generator and the pump coastdown time. A deeper analysis recommends increasing pump coastdown time to avoid mass flow disturbances during coastdown. For UTOP, the most sensitive parameters are the gap between fuel and clad, the reactivity feedback coefficients, and to a lesser extent, fuel conductivity and fuel heat capacity. In any case, the uncertainties never bring the reactor beyond safety limits.
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| 39. |
- Robertson, Gustav, Ph.D. student, 1986-, et al.
(författare)
-
Model inadequacy in fuel performance code calibration: Derivative-based parameter uncertainty inflation
- 2024
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 208
-
Tidskriftsartikel (refereegranskat)abstract
- Fuel performance codes are used to forecast fuel behavior and ensure safe operation. These analyses must typically include prediction uncertainties, and fuel performance models need calibration. Consequently, code calibration must derive the best estimates and corresponding uncertainties of model parameters for subsequent propagation.Bayesian calibration is popular for generating the probability distribution of model parameters. However, model inadequacy disrupts these techniques, typically resulting in underestimated uncertainties. Earlier research showcased the incorporation of model inadequacy by model parameter inflation. The method demands cheap code predictions and derivatives, which required further research to develop differentiated Gaussian process surrogates.This work combines those techniques into a complete methodology. We demonstrate it by calibrating Transuranus against fission gas release and cladding oxidation data. The result is model parameter uncertainties that primarily explain the discrepancies between the predictions and corresponding measurements, except when the output behaves highly non-linearly.
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| 40. |
- Robertson, Gustav, Ph.D. student, 1986-, et al.
(författare)
-
Treating model inadequacy in fuel performance model calibration by parameter uncertainty inflation
- 2022
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 179
-
Tidskriftsartikel (refereegranskat)abstract
- The nuclear industry uses fuel performance codes to demonstrate integrity preservation of fuel rods. These codes include a complex system of models with empirical constants that one needs to calibrate for best estimates and uncertainties. However, deriving the appropriate level of uncertainty is often challenging due to model inadequacies.This paper presents a method to address model inadequacies by adapting the mean and covariance of the model parameters so that the propagated uncertainty conforms with the spread of the residuals rather than calibrating the model parameters directly.We demonstrate the method on synthetic data sets from an artificial test-bed containing a cladding oxidation and a hydrogen pick-up model. A repeated validation using many synthetic data sets shows that the method is robust and handles model inadequacies appropriately in most cases. Furthermore, we compare with traditional calibration and show model inadequacy leads to underestimation of uncertainties if not addressed.
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| 41. |
- Senis, Lorenzo, et al.
(författare)
-
Performance evaluation of a novel gamma transmission micro-densitometer for PIE of nuclear fuel
- 2023
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 187
-
Tidskriftsartikel (refereegranskat)abstract
- Collimated Gamma Transmission Micro-Densitometry (GTMD) is a novel technique proposed to investigate local density variations of nuclear fuel in PIE, with a high spatial resolution. In this work, the first experimental tests of a gamma micro-densitometer are presented and the performance is characterized. The experimental procedures are described, including the aligning process and the calibration methodology. The results demonstrated that for the calibration samples with a thickness above 5 mm, a local density was obtained with a maximum discrepancy of about 2% and a spatial resolution of about 280 µm. The setup was used for the first test on an irradiated ADOPTTM fuel pellet slice. From the measurement, an average bulk density of about 9.58 g/cm3 was calculated and local density features were observed, possibly related to rim effects or the presence of local cracks. The information acquired also presented valuable information for possible improvements in the setup’s performance.
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| 42. |
- Siefman, Daniel, et al.
(författare)
-
Development and application of marginal likelihood optimization for integral parameter adjustment
- 2021
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 159
-
Tidskriftsartikel (refereegranskat)abstract
- When adjusting nuclear data with integral experiments, care must be taken that spurious adjustments are not made by assimilating poorly characterized integral parameters. If there are unaccounted for biases or poorly estimated uncertainties in the calculated and experimental values for an integral parameter, the Bayesian data assimilation may adjust the nuclear data in a manner that does not reflect the physics of the integral parameter. To identify and lessen the impact of these inconsistent integral parameters, we present a Marginal Likelihood Optimization algorithm. In a data-driven way, the marginalized likelihood is used to modulate hyperparameter terms that decrease the influence of inconsistent integral parameters on the adjustment. The advantage of this approach over other methods in the literature is that it incorporates correlation information and does not remove an integral parameter from the adjustment. Herein, we present and motivate the algorithm, and apply it to an integral data assimilation case study.
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| 43. |
- Solans, Virginie, et al.
(författare)
-
Spent Nuclear Fuel passive gamma analysis and reproducibility : Application to SKB-50 assemblies
- 2023
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 192
-
Tidskriftsartikel (refereegranskat)abstract
- This work studies the reproducibility of passive gamma spectroscopy measurements for spent nuclear fuels (SNFs). The fifty assemblies used for this study span over a variety of initial enrichments, burnups, and cooling times. These SNFs have been measured in two different gamma axial measurement campaigns. The net peak counts are determined for Cs-137, Eu-154 and Cs-134. Furthermore, a sensitivity analysis of the relative position of the SNF and the detector is performed. Most importantly, this work describes a methodology using an intrinsic self-calibration procedure that can be used to compare the relative activities of the radionuclides without the need for detailed knowledge about the measurement set-up and its properties. The reproducibility of the Cs-137 net peak count rate ranges between 2% and 4%. Systematic reproducibility of the ratio of Eu-154 and Cs-134 to Cs-137 is between 0,4% - 5 % using the intrinsic self-calibration method.
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| 44. |
- Thakre, Sachin, 1984-, et al.
(författare)
-
Experimental investigation of solid particle spreading driven by gas injection into a pool of water
- 2022
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 174, s. 109165-109165
-
Tidskriftsartikel (refereegranskat)abstract
- Motivated from the fuel–coolant interaction phenomena in boiling water reactors, in the present work, effect of natural convection flows, set during the melt dripping, on the nature of formation of debris bed, of solidified particles, at the bottom floor is studied. Standard shape solid particles are used to simulate the dripping melt and their paths are tracked using a particle tracking technique to acquire additional data such as particles velocity, travel time and path. The experimental studies performed on PDS-P facility are designed to study the separate effects and generate the data for codes validation. A novel particle tracking technique allowed quantification of kinetic properties for every particle. The results helped in refinement of the particles distribution on the floor, quantifying the debris bed shape. Higher pool depths and natural convection flows rates are seen effective in enhancing the distribution of debris particles, creating shallow and well spread debris bed.
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| 45. |
- Vidal, Antoni, et al.
(författare)
-
A time and frequency domain analysis of the effect of vibrating fuel assemblies on the neutron noise
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 137
-
Tidskriftsartikel (refereegranskat)abstract
- The mechanical vibrations of fuel assemblies have been shown to give rise to high levels of neutron noise, triggering in some circumstances the necessity to operate nuclear reactors at a reduced power level. This work analyses the effect in the neutron field of the oscillation of one single fuel assembly. Results show two different effects in the neutron field caused by the fuel assembly vibration. First, a global slow variation of the total reactor power due to a change in the criticality of the system. Second, an oscillation in the neutron flux in-phase with the assembly vibration. This second effect has a strong spatial dependence that can be used to localize the oscillating assembly. This paper shows a comparison between a time-domain and a frequency-domain analysis of the phenomena to calculate the spatial response of the neutron noise. Numerical results show a really close agreement between these two approaches.
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| 46. |
- Vidal, Antoni, et al.
(författare)
-
Modelling and simulations of reactor neutron noise induced by mechanical vibrations
- 2022
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 177
-
Tidskriftsartikel (refereegranskat)abstract
- Mechanical vibrations of core internals are among the main perturbations that induce oscillations in the neutron flux field, also known as neutron noise. In this work, different simulation models for the study of the influence of the mechanical vibrations of fuel assemblies on the neutron flux in the reactor core have been discussed. These methodologies employ the diffusion approximation, with or without a previous homogenization model, to simulate the neutron noise in the time or the frequency domain. The diffusion-based approach is expected to be less accurate in the vicinity of the vibrating fuel assemblies, but correct when considering distances larger than a few diffusion lengths away from the perturbation. All methodologies provide consistent results and can reproduce typical features of the neutron noise induced by mechanical vibrations of core components. First, FEMFFUSION can perform simulations in both the time and frequency domains. Second, CORE SIM + can be used to study various neutron noise scenarios in realistic three-dimensional reactor configurations. The third methodology is centred on using commercial codes as CASMO-5, SIMULATE-3 and SIMULATE-3K. This methodology allows time domain simulations of the neutron noise induced by different neutron noise sources in a nuclear reactor. Finally, a model for time-dependent geometry is implemented for the code system ATHLET/QUABOX-CUBBOX employing a cross-section-based approach for encoding water gap width variations at the reflector.
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| 47. |
- Viebach, M., et al.
(författare)
-
Verification of the code DYN3D for calculations of neutron flux fluctuations
- 2022
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 166
-
Tidskriftsartikel (refereegranskat)abstract
- Insufficiently explained magnitudes and patterns of flux fluctuation observed mainly in KWU PWRs are recently investigated by various European institutions. Among the numerical tools used to investigate the neutron flux fluctuations is the time-domain reactor dynamics code DYN3D. As DYN3D and comparable codes have not been developed with the primary intention to simulate low-amplitude neutron flux fluctuations, their applicability in this field has to be verified. In order to contribute to the verification of DYN3D for the simulation of neutron flux fluctuations, two special cases of perturbations of the neutron flux (a localized absorber of variable/oscillatory strength and a travelling oscillatory perturbation) are considered with DYN3D on the one hand and with the frequency-domain neutron noise tool CORE SIM as well as analytical frequency-domain approaches, respectively, on the other hand. The obtained results are compared with respect to the distributions of the amplitude and the phase of the induced neutron flux fluctuations. The comparisons are repeated with varied amplitudes and frequencies of the perturbation. The results agree well both qualitatively and quantitatively for each of the conducted calculations. The remaining deviations between the DYN3D results and the reference results exhibit a dependence on the perturbation magnitude, which is attributed to the neglect of higher-order terms (linear theory) of the perturbed quantities in the calculation of the reference solutions.
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| 48. |
- Vinai, Paolo, 1975, et al.
(författare)
-
On the simulation of neutron noise induced by vibrations of fuel pins in a fuel assembly
- 2023
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 181
-
Tidskriftsartikel (refereegranskat)abstract
- Vibrations of fuel assemblies are an important issue in the safe operation of nuclear reactors, because they can challenge the integrity of the fuel with potential for radioactive releases. Reactor neutron noise-based techniques for monitoring vibrations are valuable for core diagnostic since they are not intrusive and make use of ordinary neutron flux measurements from ex-core and in-core detectors. The application of these techniques involves the solution of inverse problems that require numerical simulations capable of estimating the reactor neutron noise, given a model of the vibrations. For this purpose, several novel reactor neutron noise solvers have been developed in the CORTEX project using either Monte Carlo or deterministic methods, such as the discrete ordinates method, the method of characteristics, and the diffusion approximation. In the current work, these solvers have been scrutinized by computing the neutron noise induced by vibrations of one or multiple fuel pins in a simplified UOX fuel assembly benchmark, via proper variations of macroscopic neutron cross sections. The comparison of these neutron noise solutions obtained from the different methods shows novel insights into the simulation of neutron noise induced by mechanical vibrations, such as the challenges posed by the Monte Carlo method, the impact of the angular discretization on the application of the discrete ordinates method, and the accuracy of the diffusion approximation assessed via the higher-order neutron transport methods.
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| 49. |
- Wallenius, Janne, 1968-
(författare)
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Anomalous reactivity swing in the 238U-233U system
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 139
-
Tidskriftsartikel (refereegranskat)abstract
- An anomalous negative reactivity swing is discovered in the iso-breeding 238U-233U system, resulting from an increase in conversion ratio with burn-up. Based on this discovery, a plutonium-free ternary fuel composition is identified allowing to reach a burn-up of 100 GWd/ton with a reactivity swing of ≃0.3$. Potential applications of such fuels are discussed.
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| 50. |
- Wang, Guan, et al.
(författare)
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Transient analyses for China initiative Accelerator Driven System using the extended BELLA code
- 2023
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 190
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Tidskriftsartikel (refereegranskat)abstract
- China initiative Accelerator Driven System (CiADS) is a 10 MW lead-bismuth eutectic (LBE) cooled subcritical reactor loaded with UO2 fuels in the first phase, which is designed to demonstrate the engineering feasibility of the ADS concept. The transient analyses were performed to show the safety potential and dynamic characteristics of the CiADS subcritical reactor using the extended BELLA code. Typical scenarios, such as beam-trip transients, unprotected beam overpower (UBOP), unprotected and protected loss of flow (ULOF and PLOF), unprotected and protected loss of heat sink (ULOHS and PLOHS) and self-defined station blackout (SBO), were analyzed and simulated. The results indicate that, as a low-power reactor, the CiADS subcritical reactor has a large margin to avoid severe damage in the core. Short-term cladding failure caused by thermal fatigue under frequent beam trips may not happen since the variations of temperatures are relatively small. However, accelerated LBE corrosion combined with accumulated creep tends to be a risk under ULOF and ULOHS.
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