| 1. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Theoretical investigation of the MTC noise estimate in 1-D homogeneous systems
- 2002
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:1, s. 75-100
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Tidskriftsartikel (refereegranskat)abstract
- In this paper, the accuracy of the noise-based determination of the moderator temperature coefficient (MTC) is investigated theoretically and quantitatively. It is known from earlier work that the noise method systematically underestimates the MTC. In this paper, it is found that the main reason for the underestimation lies with the radial incoherence of the temperature fluctuations. The deviation of the reactor response from point-kinetics is another possible reason, but it was found to play a quite insignificant role. The theory of neutron noise, induced by spatially random perturbations is elaborated and by its help the inaccuracy (bias) of the noise based MTC estimation was quantitatively investigated. It was found that a relatively short correlation length of the temperature fluctuations, which is in agreement with experimental evidence, can explain the observed underestimation of the MTC by the noise method.
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| 2. |
- Arzhanov, Vasily
(författare)
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Monotonicity properties of k(eff) with shape change and with nesting
- 2002
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:2, s. 137-145
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Tidskriftsartikel (refereegranskat)abstract
- It was found that, contrary to expectations based on physical intuition, k(eff) can both increase and decrease when changing the shape of an initially regular critical system, while preserving its volume. Physical intuition would only allow for a decrease of k(eff) when the surface/volume ratio increases. The unexpected behaviour of increasing k(eff) was found through numerical investigation. For a convincing demonstration of the possibility of the non-monotonic behaviour, a simple geometrical proof was constructed. This latter proof, in turn, is based on the assumption that k(eff) can only increase (or stay constant) in the case of nesting, i.e. when adding extra volume to a system. Since we found no formal proof of the nesting theorem for the general case, we close the paper by a simple formal proof of the monotonic behaviour of k(eff) by nesting.
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| 3. |
- Arzhanov, Vasily
(författare)
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Multi-group theory of neutron noise induced by vibrating boundaries
- 2002
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:18, s. 2143-2158
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Tidskriftsartikel (refereegranskat)abstract
- The paper extends the one-group analysis of the neutron noise induced by fluctuating boundaries [Ann. Nucl. Energy 27(2000)1385] to the general multi-group non-homogeneous model. The full solution is given through the Green's function of the static problem, the static flux, and a quantity describing the boundary movements. A multi-group absorber model is proposed to represent the perturbation. which turns out to be very useful, for instance, to derive the point reactor and adiabatic approximations of the neutron noise arising from the oscillating boundaries. Finally, an equivalent solution is given in terms of the adjoint function.
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| 4. |
- Eriksson, Marcus, et al.
(författare)
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Inherent Shutdown Capabilities in Accelerator-driven Systems
- 2002
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:14, s. 1689-1706
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Tidskriftsartikel (refereegranskat)abstract
- The applicability for inherent shutdown mechanisms in accelerator-driven systems (ADS) has been investigated. We study the role of reactivity feedbacks. The benefits, in terms of dynamics performance, for enhancing the Doppler effect are examined. Given the performance characteristics of source-driven systems, it is necessary to manage the neutron source in order to achieve inherent shutdown. The shutdown system must be capable of halting the external source before excessive temperatures are obtained. We evaluate methods, based on the analysis of unprotected accidents, to accomplish such means. Pre-concepted designs for self-actuated shutdown of the external source suggested. We investigate time responses and evaluate methods to improve the performance of the safety system. It is shown that maximum beam output must be limited by fundamental means in order to protect against accident initiators that appear to be achievable in source driven systems. Utilizing an appropriate burnup control strategy plays a key role in that effort.
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| 5. |
- Pazsit, I., et al.
(författare)
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Linear reactor kinetics and neutron noise in systems with fluctuating boundaries
- 2000
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 27:15, s. 1385-1398
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Tidskriftsartikel (refereegranskat)abstract
- The general theory of linear reactor kinetics and that of the induced neutron noise is developed for systems with varying size, i.e. in which the position of the boundary fluctuates around a stationary value. The point kinetic and adiabatic approximations are defined by a generalisation of the flux factorisation, and the full solution of the general problem with an arbitrarily fluctuating boundary is given by the Green's function technique. The correctness of the general solution is proven both generally and also by considering the simple case of a 2-D cylindrical reactor with a fluctuating radius, in which case a direct compact solution is possible.
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| 6. |
- Talamo, Alberto, et al.
(författare)
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Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn
- 2006
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:14-15, s. 1176-1188
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Tidskriftsartikel (refereegranskat)abstract
- Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.
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| 7. |
- 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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| 8. |
- 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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| 9. |
- 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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| 10. |
- Alhassan, Erwin, et al.
(författare)
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Benchmark selection methodology for reactor calculations and nuclear data uncertainty reduction
- 2015
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
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Tidskriftsartikel (refereegranskat)abstract
- Criticality, reactor physics and shielding benchmarks are expected to play important roles in GEN-IV design, safety analysis and in the validation of analytical tools used to design these reactors. For existing reactor technology, benchmarks are used for validating computer codes and for testing nuclear data libraries. Given the large number of benchmarks available, selecting these benchmarks for specic applications can be rather tedious and difficult. Until recently, the selection process has been based usually on expert judgement which is dependent on the expertise and the experience of the user and there by introducing a user bias into the process. This approach is also not suitable for the Total Monte Carlo methodology which lays strong emphasis on automation, reproducibility and quality assurance. In this paper a method for selecting these benchmarks for reactor calculation and for nuclear data uncertainty reduction based on the Total Monte Carlo (TMC) method is presented. For reactor code validation purposes, similarities between a real reactor application and one or several benchmarks are quantied using a similarity index while the Pearson correlation coecient is used to select benchmarks for nuclear data uncertainty reduction. Also, a correlation based sensitivity method is used to identify the sensitivity of benchmarks to particular nuclear reactions. Based on the benchmark selection methodology, two approaches are presented for reducing nuclear data uncertainty using integral benchmark experiments as an additional constraint in the TMC method: a binary accept/reject and a method of assigning file weights using the likelihood function. Finally, the methods are applied to a full lead-cooled fast reactor core and a set of criticality benchmarks. Signicant reductions in Pu-239 and Pb-208 nuclear data uncertainties were obtained after implementing the two methods with some benchmarks.
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