| 1. |
- Demaziere, Christophe, 1973, et al.
(författare)
-
Theoretical investigation of the MTC noise estimate in 1-D homogeneous systems
- 2002
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:1, s. 75-100
-
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.
|
|
| 2. |
- Arzhanov, Vasily
(författare)
-
Monotonicity properties of k(eff) with shape change and with nesting
- 2002
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:2, s. 137-145
-
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.
|
|
| 3. |
- Arzhanov, Vasily
(författare)
-
Multi-group theory of neutron noise induced by vibrating boundaries
- 2002
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:18, s. 2143-2158
-
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.
|
|
| 4. |
- Eriksson, Marcus, et al.
(författare)
-
Inherent Shutdown Capabilities in Accelerator-driven Systems
- 2002
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 29:14, s. 1689-1706
-
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.
|
|
| 5. |
- Pazsit, I., et al.
(författare)
-
Linear reactor kinetics and neutron noise in systems with fluctuating boundaries
- 2000
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 27:15, s. 1385-1398
-
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.
|
|
| 6. |
- Talamo, Alberto, et al.
(författare)
-
Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:14-15, s. 1176-1188
-
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.
|
|
| 7. |
- al-Dbissi, Moad, 1994, et al.
(författare)
-
Identification of diversions in spent PWR fuel assemblies by PDET signatures using Artificial Neural Networks (ANNs)
- 2023
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 193
-
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.
|
|
| 8. |
- al-Dbissi, Moad, 1994, et al.
(författare)
-
On the use of neutron flux gradient with ANNs for the detection of diverted spent nuclear fuel
- 2024
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 204
-
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.
|
|
| 9. |
- Alhassan, E., et al.
(författare)
-
Bayesian updating for data adjustments and multi-level uncertainty propagation within Total Monte Carlo
- 2020
-
Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 139
-
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.
|
|
| 10. |
- Alhassan, Erwin, et al.
(författare)
-
Benchmark selection methodology for reactor calculations and nuclear data uncertainty reduction
- 2015
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
-
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.
|
|
| 11. |
- Alhassan, Erwin, et al.
(författare)
-
Selecting benchmarks for reactor simulations : an application to a Lead Fast Reactor
- 2016
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 96, s. 158-169
-
Tidskriftsartikel (refereegranskat)abstract
- For several decades reactor design has been supported by computer codes for the investigation of reactor behavior under both steady state and transient conditions. The use of computer codes to simulate reactor behavior enables the investigation of various safety scenarios saving time and cost. There has been an increase in the development of in-house (local) codes by various research groups in recent times for preliminary design of specific or targeted nuclear reactor applications. These codes must be validated and calibrated against experimental benchmark data with their evolution and improvements. Given the large number of benchmarks available, selecting these benchmarks for reactor calculations and validation of simulation codes for specific or target applications can be rather tedious and difficult. In the past, the traditional approach based on expert judgement using information provided in various handbooks, has been used for the selection of these benchmarks. This approach has been criticized because it introduces a user bias into the selection process. This paper presents a method for selecting these benchmarks for reactor calculations for specific reactor applications based on the Total Monte Carlo (TMC) method. First, nuclear model parameters are randomly sampled within a given probability distribution and a large set of random nuclear data files are produced using the TALYS code system. These files are processed and used to analyze a target reactor system and a set of criticality benchmarks. Similarity between the target reactor system and one or several benchmarks is quantified using a similarity index. The method has been applied to the European Lead Cooled Reactor (ELECTRA) and a set of plutonium and lead sensitive criticality benchmarks using the effective multiplication factor (keffkeff). From the study, strong similarity were observed in the keffkeff between ELECTRA and some plutonium and lead sensitive criticality benchmarks. Also, for validation purposes, simulation results for a list of selected criticality benchmarks simulated with the MCNPX and SERPENT codes using different nuclear data libraries have been compared with experimentally measured benchmark keff values.
|
|
| 12. |
- Alhassan, Erwin, et al.
(författare)
-
Uncertainty and correlation analysis of lead nuclear data on reactor parameters for the European Lead Cooled Training Reactor
- 2015
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 26-37
-
Tidskriftsartikel (refereegranskat)abstract
- The Total Monte Carlo (TMC) method was used in this study to assess the impact of Pb-204, 206, 207, 208 nuclear data uncertainties on reactor safety parameters for the ELECTRA reactor. Relatively large uncertainties were observed in the k-eff and the coolant void worth (CVW) for all isotopes except for Pb-204 with signicant contribution coming from Pb-208 nuclear data; the dominant eectcame from uncertainties in the resonance parameters; however, elastic scattering cross section and the angular distributions also had signicant impact. It was also observed that the k-eff distribution for Pb-206, 207, 208 deviates from a Gaussian distribution with tails in the high k-eff region. An uncertainty of 0.9% on the k-eff and 3.3% for the CVW due to lead nuclear data were obtained. As part of the work, cross section-reactor parameter correlations were also studied using a Monte Carlo sensitivity method. Strong correlations were observed between the k-eff and (n,el) cross section for all the lead isotopes. The correlation between the (n,inl) and the k-eff was also found to be signicant.
|
|
| 13. |
- Andersson, Peter, 1981-, et al.
(författare)
-
A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor
- 2017
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 110, s. 88-97
-
Tidskriftsartikel (refereegranskat)abstract
- The Halden reactor project (HRP) has recently developed a gamma emission tomography instrument dedicated for measurements of irradiated nuclear fuel in collaboration with Westinghouse and Uppsala University. This instrument is now assembled and the first experimental measurements have been performed on fuel assemblies irradiated in the Halden reactor. The objective of the instrument is to map the distribution of radioisotopes of interest in the fuel, e.g. 137Cs or 140La/Ba, and for this purpose, a spectroscopic high-purity Germanium detector has been selected, which enables the identification and tomographic reconstruction of separate isotopes by their characteristic gamma rays.To gain from the analysis of the data from this new instrument, and in the future from other gamma emission tomography instruments for nuclear fuels, various reconstruction methods are available that vary in the accuracy and the amount of detail obtainable in the analysis. This paper presents the details of the working principles of a new code for gamma emission tomography, the UPPREC (UPPsala university REConstruction) code. It is a development in MATLABTM code with the aim to produce detailed quantitative images of the investigated fuel.In this paper, the methods assembled for the analysis of data collected by this novel instrument are described and demonstrated and a benchmark is presented using single rod gamma scanning. It is shown that the UPPREC methodology improves the accuracy of the reconstructions by removing the errors introduced by the presence of highly attenuating fuel and structural material in the fuel assembly. With the introduction of UPPREC, detailed quantitative cross-sectional images of nuclide concentrations in nuclear fuel are now able to be obtained by nondestructive means. This has potential applications in both nuclear fuel diagnostics and in safeguards.
|
|
| 14. |
- Baeten, P., et al.
(författare)
-
Determination of the subcriticality level using the Cf-252 source-detector method
- 2010
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 37:5, s. 740-752
-
Tidskriftsartikel (refereegranskat)abstract
- Measurement and monitoring of reactivity in a subcritical state, e.g. during the loading of a power reactor, has a clear safety relevance. The methods currently available for the measurement of k(eff) in stationary subcritical conditions should be improved as they refer to the critical state. This is also very important in the framework of ADS (accelerator driven systems) where the measurement of a subcritical level without knowledge of the critical state is looked for. An alternative way to achieve this is by mean of the Cf-252 source-detector method. The method makes use of three detectors inserted in the reactor: two "ordinary" neutron detectors and one Cf-252 source-detector which contains a small amount of Cf-252 that introduces neutrons in the system through spontaneous fission. By observing fissions through the detection system and correlating the signals of the three detectors, the reactivity rho (and hence the multiplication factor k) can be determined. Before the actual measurements took place, a suitable data acquisition system was realized in order to process the signals and compute the auto and cross power spectral densities. The measurements were then performed in the VENUS reactor, using the Cf-252 source-detector and two BF3 neutron detectors. The multiplication factor was determined using the Cf source method and compared with measurements using other methods and with computational results (Monte Carlo simulations). The Cf method was benchmarked at a UOX core to other experimental methods that used the critical state as reference and to calculations. Afterwards, the Cf source technique was analyzed in a MOX core to study the possible impact of a significant intrinsic source on the results. This benchmarking gives the possibility to validate the Cf method as a reliable technique for the measurement of subcritical levels in steady state and for cores with an intrinsic source like MOX or burnt fuel cores. (C) 2010 Elsevier Ltd. All rights reserved.
|
|
| 15. |
- Bakardjieva, S., et al.
(författare)
-
Quality improvements of thermodynamic data applied to corium interactions for severe accident modelling in SARNET2
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 74, s. 110-124
-
Tidskriftsartikel (refereegranskat)abstract
- In a severe accident transient, corium composition and its properties determine its behaviour and its potential interactions both with the reactor vessel and in the later phases with the concrete basemat. This, in turn, requires a detailed knowledge of the phases present at temperature and how they are formed. Because it implies mainly the investigation of chemical systems at high temperature, these data are often difficult to obtain or are uncertain if it already exists. Therefore more data are required both to complete the thermodynamic databanks (such as NUCLEA) and to construct accurate equilibrium phase diagrams and to finally contribute to the improvement of the codes simulating these severe accident conditions. The MCCI work package (WP6) of the SARNET 2 Network of Excellence has been addressing these problems. In this framework in large facilities such as VULCANO tests have been performed on the interactions and ablation of UO2-containing melts with concrete. They have been completed by large scale MCCI testing such EPICOR on vessel steel corrosion. In parallel in major EU-funded ISTC projects co-ordinated with national institutes, such as the CORPHAD and PRECOS, smaller, single effect tests have been carried out on the more difficult phase diagrams. These have produced data that can be directly used by databanks and for modelling improvement/validation. From these data significant advances in the melt chemistry and pool behaviour have been made. A selection of experiments from participating institutes are presented in this paper and give hindsight into the major processes and so give clear indications for the future work, especially in light of the Fukushima accident.
|
|
| 16. |
- Bansah, C. Y., et al.
(författare)
-
Theoretical model for predicting the relative timings of potential failures in steam generator tubes of a PWR during a severe accident
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 59, s. 10-15
-
Forskningsöversikt (refereegranskat)abstract
- During certain severe reactor accidents such as station-blackout accidents, countercurrent natural circulation flow could develop within the reactor coolant system. Natural circulation flow is very important because of transfer of decay energy from the core to other parts of the reactor coolant system. The associated heat-ups of the reactor coolant system structures can lead to pressure boundary failures with notable vulnerabilities in the pressurizer surge line, the hot leg nozzles and the steam generator tubes. The potential for a steam generator tube failure has been of particular concern because fission products could be released to the environment through such a failure. To solve the problem of steam generator tube failure, a computer code - Steam Generator Mitigation Program (SGMP), written in FORTRAN 95 computes the recirculation ratio (RR) and the mixing fraction (MF) which are the main parameters used in characterizing natural circulation. In the flow analysis, the RR and MF were respectively found to be 2.4 +/- 0.3 and 0.8 +/- 0.17. The results obtained showed that the natural circulation would delay the failure time of the steam generator tubes and is in good qualitative agreement with results from literature.
|
|
| 17. |
- Basso, Simone, et al.
(författare)
-
Effectiveness of the debris bed self-leveling under severe accident conditions
- 2016
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 95, s. 75-85
-
Tidskriftsartikel (refereegranskat)abstract
- Melt fragmentation, quenching and long term coolability in a deep pool of water under the reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. The success of such strategy is contingent upon the natural circulation effectiveness in removing the decay heat generated in the porous debris bed. The maximum height of the bed is one of the important factors which affect the debris coolability. The two-phase flow within the bed generates mechanical energy which can change the geometry of the debris bed by the "self-leveling" phenomenon. In this work.we developed an approach to modeling of the self-leveling phenomenon. Sensitivity analysis was carried out to rank the importance of the model uncertainties and uncertain input parameters i.e. the conditions of the accident scenario and the debris bed properties. The results provided some useful insights for further improvement of the model and reduction of the output uncertainties through separate-effect experimental studies. Finally, we assessed the self-leveling effectiveness, quantified its uncertainties in prototypic severe accident conditions and demonstrated that the effect of self-leveling phenomenon is robust with respect to the considered input uncertainties.
|
|
| 18. |
- Becares, V., et al.
(författare)
-
Evaluation of the criticality constant from Pulsed Neutron Source measurements in the Yalina-Booster subcritical assembly
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 53, s. 40-49
-
Tidskriftsartikel (refereegranskat)abstract
- The prompt decay constant method and the area-ratio (Sjostrand) method constitute the reference techniques for measuring the reactivity of a subcritical system using Pulsed Neutron Source experiments (PNS). However, different experiments have shown that in many cases it is necessary to apply corrections to the experimental results in order to take into account spectral and spatial effects. In these cases, the approach usually followed is to develop different specific correction procedures for each method. In this work we discuss the validity of prompt decay constant method and the area-ratio method in the Yalina-Booster subcritical assembly and propose a general correction procedure based on Monte Carlo simulations.
|
|
| 19. |
- Becares, V., et al.
(författare)
-
Validation of ADS reactivity monitoring techniques in the Yalina-Booster subcritical assembly
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 53, s. 331-341
-
Tidskriftsartikel (refereegranskat)abstract
- The development of a reactivity monitoring system for subcritical reactors is a major task prior to industrial scale accelerator driven system (ADS) construction. Within the 6th European Framework Program, the IP-EUROTRANS project has performed a series of experiments at the Yalina-Booster subcritical assembly located at the Joint Institute for Power and Nuclear Research (JIPNR) of the National Academy of Sciences of Belarus, using a continuous (D, T) (fusion) neutron source in pulsed and continuous mode with short interruptions (beam trips). In this paper, the implementation and results of three different monitoring techniques intended to operate with continuous neutron sources will be presented, namely the source-jerk technique, the prompt decay constant technique and the current-to-flux technique. The results will be compared with the values of the reactivity obtained using the pulsed source in PNS experiments, discussed in detail in another paper.
|
|
| 20. |
- Bechta, Sevostian, et al.
(författare)
-
On the EU-Japan roadmap for experimental research on corium behavior
- 2019
-
Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 124, s. 541-547
-
Tidskriftsartikel (refereegranskat)abstract
- A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.
|
|
| 21. |
- Berglöf, Carl, et al.
(författare)
-
Auto-correlation and variance-to-mean measurements in a subcritical core obeying multiple alpha-modes
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:2-3, s. 194-202
-
Tidskriftsartikel (refereegranskat)abstract
- Neutron noise measurements based on the Rossi-alpha and Feynman-alpha methodologies have been performed in a heterogeneous subcritical system. It is shown that the traditional single alpha-mode formulations of the Rossi-alpha and Feynman-alpha methods are not applicable due to the presence of higher alpha-modes. Formalisms taking into account multiple alpha-modes are applied resulting in satisfactory results. Three alpha-modes could be identified using the Rossi-alpha method, whereas only two could be obtained using the Feynman-alpha method. In the Feynman-alpha case, the possibility to obtain the fastest decaying alpha-mode was diminished due to detector dead time effects. It was found that the slowest decaying alpha-mode does not exactly correspond to the prompt decay found in pulsed neutron source measurements, which confirms the results of previous studies. Strengths and weaknesses of the multiple alpha-mode Rossi-alpha and Feynman-alpha methods observed in this study are pointed out.
|
|
| 22. |
- Boafo, E.K., et al.
(författare)
-
Utilizing the burnup capability in MCNPX to perform depletion analysisof an MNSR fuel
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 73, s. 478-483
-
Tidskriftsartikel (refereegranskat)abstract
- In this work, we present results of fuel depletion analyses performed for a potential LEU core of Ghana’s Miniature Neutron Source Reactor (GHARR-1) using the Monte Carlo N-particle extended (MCNPX) neutron transport code. Depletion calculation was carried out for the reactor core from the Beginning of Life (BOL) to the End of Life (EOL) which corresponds to 10 years of reactor operation. The amounts of uranium and plutonium actinides were estimated at BOL and EOL of the core. Decay heat removal rate for the MNSR after reactor shut down was investigated due to its significance to reactor safety. Inventory of fission products produced as a result of burnup was also calculated. The results show that a maximum discharge burnup equivalent to 0.568% of U-235 was consumed at EOL equivalent to operating the reactor for 200 Effective Full Power Days (EFPD), while the amount of Pu-239 produced was not significant.Also, the decay heat decreased exponentially after reactor shutdown confirming that decay heat will be removed in the system by natural circulation after shutdown and thus guaranteeing the safety of the reactor.
|
|
| 23. |
- Bortot, Sara, et al.
(författare)
-
BELLA : a multi-point dynamics code for safety-informed design of fast reactors
- 2015
-
Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 85, s. 228-235
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, the multi-point dynamics code BELLA and its benchmarking with respect to SAS4A/SASSYS-1 is described for a small fast reactor cooled with natural convection of lead (ELECTRA). It is shown that BELLA is capable of reproducing the magnitude of mass-flow, reactivity, power and temperature excursions during design extension conditions with an accuracy better than 10%. Hence, the BELLA code can be used for safety-informed design and stability analyses of fast reactor systems, permitting to isolate essential phenomena and trends of significance for their safety assessment. (C) 2015 Elsevier Ltd. All rights reserved.
|
|
| 24. |
- Bortot, Sara, 1983-, et al.
(författare)
-
BELLA : a multi-point dynamics code for simulation of fast reactors
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, the multi-point dynamics code BELLA and its benchmarking with respect to SAS4A/SASSYS- 1 is described for a small fast reactor cooled with natural convection of lead (ELECTRA). It is shown that BELLA is capable of reproducing the magnitude of mass-flow, reactivity, power and temperature excursions during design extension conditions with an accuracy better than 10%. Hence, the BELLA code can be used for safety-informed design and stability analyses of fast reactor systems, permitting to isolate essential phenomena and trends of significance for their safety assessment.
|
|
| 25. |
- Bosland, L., et al.
(författare)
-
Iodine-paint interactions during nuclear reactor severe accidents
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74:C, s. 184-199
-
Tidskriftsartikel (refereegranskat)abstract
- To assess the radiological consequences of a severe reactor accident, it is important to be able to predict the behaviour of iodine in containment. Some interactions between iodine and containment paint (e.g., adsorption) have been well known for a long time. However, in recent years, new phenomena have been identified that can affect the gas phase iodine concentration in the longer term (e.g., the release of molecular iodine and organic iodides from irradiated painted surfaces). Several international collaborations and organizations around the world are currently addressing different aspects of this topic, including laboratory experiments and theoretical studies (ab initio) designed to improve the mechanistic understanding of the phenomena. Knowledge of the underlying mechanisms will provide explanations for behavioural differences observed between paint types, and will support the extrapolation of laboratory results to the safety analyses of nuclear reactors. The purpose of this paper is to present a selection of recent work performed by Severe Accident Research Network (SARNET) members regarding iodine-paint interactions and paint aging in order to improve the common understanding and better define what has still to be done in this area. The Severe Accident Research Network (SARNET) provides a framework within which members can share and discuss results.
|
|
| 26. |
- Börjesson Sandén, Fredrik, 1995, et al.
(författare)
-
Effects of boric acid on volatile tellurium in severe accident conditions
- 2024
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 200
-
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.
|
|
| 27. |
- Calleja, Manuel, 1984, et al.
(författare)
-
Implementation of hybrid simulation schemes in COBAYA3/SUBCHANFLOW coupled codes for the efficient direct prediction of local safety parameters
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 70, s. 216-229
-
Forskningsöversikt (refereegranskat)abstract
- The precise prediction of power generation, heat transfer and flow distribution within a reactor core is of great importance to asses the safety features of any reactor design. The necessity to better describe the most important safety related physical phenomena prevailing in LWRs drive the extensions of current neutronic (N)/thermal-hydraulic (TH) coupled methodologies. Nowadays, several computer codes that solve the time dependent neutron diffusion or transport equations are coupled with TH codes at nodal level. This coarse spatial discretization of both N and TH does not allow direct prediction of local phenomena at pin or subchannel levels. Moreover, pin by pin simulations are currently performed using different strategies and methodologies. The main drawback of these approaches is the considerable computational time needed when addressing whole core solutions. Consequently, new fast running and accurate approaches are needed to simulate reactor cores using multi physics and multi scale methodologies. This type of analysis includes for instance, the use of mixed nodal based solutions with pin level solutions for both N and TH. This paper discusses a methodology implemented to achieve coupled N/TH simulations based on hybrid schemes. First, an overview of the state of the art involving non-conform geometry is presented, followed with the description of the codes used for this purpose and their extensions to perform hybrid simulations. Results for the coupled N/TH scheme are presented for a full size PWR core in steady state.
|
|
| 28. |
- Chan, Yi Meng, et al.
(författare)
-
A deep-learning representation of multi-group cross sections in lattice calculations
- 2024
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 195
-
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.
|
|
| 29. |
- Chen, Yangli, et al.
(författare)
-
Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds
- 2022
-
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.
|
|
| 30. |
- Chen, Yangli, et al.
(författare)
-
Uncertainty quantification for TRACE simulation of FIX-II No. 5052 test
- 2020
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 143
-
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.
|
|
| 31. |
- Chernikova, Dina, 1982, et al.
(författare)
-
Novel passive and active tungsten-based identifiers for maintaining the continuity of knowledge of spent nuclear fuel copper canisters
- 2015
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 219-227
-
Tidskriftsartikel (refereegranskat)abstract
- A new approach to provide a long-term safeguards identification of spent nuclear fuel containers, in particular copper canisters, is presented in this paper. The approach proposes the use of a tungsten insert marked with a binary code and placed inside the container. The insert is read with a combination of two independent techniques, radiation and ultrasonic measurements, in order to get a unique identification of the cask. Passive and active versions of the tag are considered. The passive version makes use of the radiation coming from the spent nuclear fuel itself. The active version of the tag is based on the use of an artificially introduced mixture of α-emitting isotopes, such as 241Am with materials, 11B and 23Na, which easily undergo α-induced reactions with emission of specific γ-lines, 2313 keV and 1809 keV, respectively. The paper discusses results of the radiation and ultrasonic measurements and Monte-Carlo evaluations as the first proof of the concept. The results of the investigations show the strong potential for this concept to maintain the continuity of knowledge of spent nuclear fuel copper canisters for a time scale up to a few thousands years without compromising the environmental safety of the casks.
|
|
| 32. |
- Chernitskiy, S. V., et al.
(författare)
-
Static neutronic calculation of a subcritical transmutation stellarator-mirror fusion-fission hybrid
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 72, s. 413-420
-
Tidskriftsartikel (refereegranskat)abstract
- The MCNPX Monte-Carlo code has been used to model the neutron transport in a sub-critical fast fission reactor driven by a fusion neutron source. A stellarator-mirror device is considered as the fusion neutron source. The principal composition for a fission blanket of a mirror fusion-fission hybrid is devised from the calculations. Heat load on the first wall, the distribution of the neutron fields in the reactor, the neutron spectrum and the distribution of energy release in the blanket are calculated. The possibility of tritium breeding inside the installation in quantities that meet the needs of the fusion neutron source is analyzed. The portion of the plasma column generates fusion neutrons that mainly do not reach the fission reactor core is proposed to be surrounded by a vessel filled with borated water to absorb the flying out neutrons. The flux of the neutrons escaping from the device to surrounding space is also calculated.
|
|
| 33. |
- Chikhi, N., et al.
(författare)
-
Evaluation of an effective diameter to study quenching and dry-out of complex debris bed
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74, s. 24-41
-
Tidskriftsartikel (refereegranskat)abstract
- Many of the current research works performed in the SARNET-2 WP5 deal with the study of the coolability of debris beds in case of severe nuclear power plant accidents. One of the difficulties for modeling and transposition of experimental results to the real scale and geometry of a debris bed in a reactor is the difficulty to perform experiments with debris beds that are representative for reactor situations. Therefore, many experimental programs have been performed using beds made of multi-diameter spheres or non-spherical particles to study the physical phenomena involved in debris bed coolability and to evaluate an effective diameter. This paper first establishes the ranges of porosity and particle size distribution that might be expected for in-core debris beds and ex-vessel debris beds. Then, the results of pressure drop and dry-out heat flux (DHF) measurements obtained in various experimental setups, POMECO, DEBRIS, COOLOCE/STYX and CALIDE/PRELUDE, are presented. The issues of particle size distribution and non-sphericity are also investigated. It is shown that the experimental data obtained in "simple" debris beds are relevant to describe the behavior of more complex beds. Indeed, for several configurations, it is possible to define an "effective" diameter suitable for evaluating (with the porosity) some model parameters as well as correlations for the pressure drop across the bed, the steam flow rate during quenching and the DHF.
|
|
| 34. |
- Dahlfors, Marcus, et al.
(författare)
-
Neutron Cross Section Sensitivity for Minor Actinide Transmutation in Energy Amplifier Systems
- 2007
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 34:10, s. 824-835
-
Tidskriftsartikel (refereegranskat)abstract
- The nuclear data sensitivity in 3D Monte Carlo burnup calculations of minor actinide transmutation in Energy Amplifier Systems is assessed. Ansaldo Nucleare's 80 MWth, Energy Amplifier Demonstration Facility (EADF) design serves as a technical and geometrical platform for the analysis. The accelerator-d riven EADF is a fast, subcritical system based on classical MOX-fuel technology and on molten lead-bismuth eutectic cooling. For Monte Carlo simulations, the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN, is utilised. The code offers treatment of both high-energy particle interactions and low-energy neutron transport with a sophisticated method based on a full Monte Carlo simulation, together with the option of employing different modern nuclear data libraries. In particular, the impact of nuclear data discrepancies on transmutation properties prediction with increasing exposure is examined. Monte Carlo simulation accelerator-driven systems transmutation burnup fast neutron spectrum minor actinides nuclear waste.
|
|
| 35. |
- Davour, Anna, et al.
(författare)
-
Applying image analysis techniques to tomographic images of irradiated nuclear fuel assemblies
- 2016
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 96, s. 223-229
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper we present a set of image analysis techniques used for extraction of information from cross-sectional images of nuclear fuel assemblies, achieved from gamma emission tomography measurements. These techniques are based on template matching, an established method for identifying objects with known properties in images.We demonstrate a rod template matching algorithm for identification and counting of the fuel rods present in the image. This technique may be applicable in nuclear safeguards inspections, because of the potential of verifying the presence of all fuel rods, or potentially discovering any that are missing.We also demonstrate the accurate determination of the position of a fuel assembly, or parts of the assembly, within the imaged area. Accurate knowledge of the assembly position enables detailed modelling of the gamma transport through the fuel, which in turn is needed to make tomographic reconstructions quantifying the activity in each fuel rod with high precision.Using the full gamma energy spectrum, details about the location of different gamma-emitting isotopes within the fuel assembly can be extracted. We also demonstrate the capability to determine the position of supporting parts of the nuclear fuel assembly through their attenuating effect on the gamma rays emitted from the fuel. Altogether this enhances the capabilities of non-destructive nuclear fuel characterization.
|
|
| 36. |
- Degweker, Shashikant, 1956, et al.
(författare)
-
Stochastic equations in the invariant imbedding formulation of particle transport
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1108-1119
-
Tidskriftsartikel (refereegranskat)abstract
- Invariant imbedding theory is an alternative formulation of particle transport theory. Although stochasticfoundations of invariant imbedding have been known from the beginnings, the method itself has so farexclusively been used for calculating first moments, i.e. expectations. The present paper attempts toset up a probability balance equation in the invariant imbedding approach from which equations forthe first and second order densities are derived. It is shown that only the equations for the first order densitiesare non-linear, while subsequent order densities obey linear equations. This is expected to considerablysimplify solution to those problems which involve second order density calculations whereinvariant imbedding techniques may be profitably used. Examples of such quantities are the varianceor correlations between particles detected at two different energies or angles or the higher momentsof the emitted multiplicity distribution such as the variance from a target bombarded by incident particles.One possible area of application of our equations is non-destructive estimation of fissile material bythe active neutron assay technique. Another area is the study of particle cascade development in sputteringand positron backscattering from surfaces. The approach is illustrated by a simple forward–backwardscattering model for these two problems.
|
|
| 37. |
- Dehlin, Fredrik, 1994-, et al.
(författare)
-
An analytic approach to the design of passively safe lead-cooled reactors
- 2022
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 169, s. 108971-108971
-
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.
|
|
| 38. |
|
|
| 39. |
- Demaziere, Christophe, 1973
(författare)
-
CORE SIM: A multi-purpose neutronic tool for research and education
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:12, s. 2698-2718
-
Tidskriftsartikel (refereegranskat)abstract
- This paper deals with the development, validation, and demonstration of an innovative neutronic tool. The novelty of the tool resides in its versatility, since many different systems can be investigated and different kinds of calculations can be performed. More precisely, both critical systems and subcritical systems with an external neutron source can be studied, and static and dynamic cases in the frequency domain (i.e. for stationary fluctuations) can be considered. In addition, the tool has the ability to determine the different eigenfunctions of any nuclear core. For each situation, the static neutron flux, the different eigenmodes and eigenvalues, the first-order neutron noise, and their adjoint functions are estimated, as well as the effective multiplication factor of the system. The main advantages of the tool, which is entirely MatLab based, lie with the robustness of the implemented numerical algorithms, its high portability between different computer platforms and operative systems, and finally its ease of use since no input deck writing is required. The present version of the tool, which is based on two-group diffusion theory, is mostly suited to investigate thermal systems. The definition of both the static and dynamic core configurations directly from the static macroscopic cross-sections and their fluctuations, respectively, makes the tool particularly well suited for research and education. Some of the many benchmark cases used to validate the tool are briefly reported. The static and dynamic capabilities of the tool are also demonstrated for the following configurations: a vibrating control rod, a perturbation traveling upwards with the core flow, and a high frequency localized perturbation. The tool is freely available on direct request to the author of the present paper.
|
|
| 40. |
- Demaziere, Christophe, 1973, et al.
(författare)
-
Development and test of a novel verification scheme applied to the neutronic modelling of Molten Salt Reactors
- 2022
-
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.
|
|
| 41. |
- Demaziere, Christophe, 1973
(författare)
-
Development of a 2-D 2-group neutron noise simulator
- 2004
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 31:6, s. 647-680
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, the development of a so-called neutron noise simulator is reported. This simulator calculates both the direct and the adjoint reactor transfer function between a stationary noise source and its induced neutron noise for any 2-dimensional heterogeneous critical system. The main advantage of this neutron noise simulator is that any realistic core can be modelled, since the simulator is designed to rely on a set of material constants corresponding to the actual reactor operating conditions. The calculations are performed in the 2-group diffusion approximation and in the frequency domain. The spatial discretisation is carried out with respect to the finite difference scheme. The noise source, expressed as an "absorber of variable strength" type, is defined directly from the fluctuations of the macroscopic cross-sections and can be spatially distributed over the core or concentrated in a few discrete nodes. If the noise source is a point-source, the simulator actually estimates the 2-dimensional 2-group discretised Green's function of the system. From the calculated Green's function, the neutron noise induced by a "vibrating absorber" type of noise source can also be determined. Different benchmark cases show that this neutron noise simulator works satisfactorily.
|
|
| 42. |
- Demazière, C., et al.
(författare)
-
Development of three-dimensional capabilities for modelling stationary fluctuations in nuclear reactor cores
- 2015
-
Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 84, s. 19-30
-
Tidskriftsartikel (refereegranskat)abstract
- This paper presents the development of a numerical tool meant at modelling the effect of stationary fluctuations in nuclear cores for systems cooled with either liquid water or boiling water. The originating fluctuations are defined for the variables describing the boundary conditions of the system, i.e. inlet velocity, inlet enthalpy, and outlet pressure. The tool then determines in the frequency domain the three-dimensional distributions within the core of the corresponding fluctuations in neutron flux, coolant density, coolant velocity, coolant enthalpy, and fuel temperature. The tool is thus based on the simultaneous modelling of neutron transport, fluid dynamics, and heat transfer in a truly integrated and fully coupled manner. The modelling of neutron transport relies on the two-group diffusion approximation and a spatial discretization based on finite differences. The modelling of fluid dynamics is performed using the Homogeneous Equilibrium Model, with a void fraction correction based on a pre-computed distribution of the static slip ratio (when two-phase flow conditions are encountered). Heat conduction in the fuel pins is also accounted for, and the heat transfer between the fuel pins and the coolant is modelled also using a pre-computed distribution of the heat transfer coefficient. The spatial discretization of the fluid dynamic and heat transfer problems is carried out using finite volumes. The tool, currently entirely Matlab based, requires minimal input data, mostly in form of the three-dimensional distributions of the macroscopic cross-sections and their relative dependence on coolant density and fuel temperature, the point-kinetic parameters of the core, as well as the three-dimensional distribution of the slip ratio (in case of two-phase flow conditions) and of the heat transfer coefficient. Such data can be provided by any static core simulator that thus needs to be run prior to using the present tool. In addition to briefly summarizing the different test cases used to verify the code, the paper also presents the results of simulations performed for a typical Pressurized Water Reactor and for a typical Boiling Water Reactor, as illustrations of the capabilities of the tool.
|
|
| 43. |
- Demaziere, Christophe, 1973, et al.
(författare)
-
Identification and localization of absorbers of variable strength in nuclear reactors
- 2005
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:8, s. 812-842
-
Tidskriftsartikel (refereegranskat)abstract
- This paper investigates the possibility of localising a noise source of the type &DPRIME; absorber of variable strength&DPRIME; (or reactor oscillator) from as few as five neutron detectors evenly distributed throughout the core of a commercial nuclear reactor. The novelty of this investigation lies with the fact that the calculations are performed for a realistic 2-D heterogeneous reactor in the 2-group diffusion approximation, via the prior determination of the corresponding reactor transfer function. It is first demonstrated that the response of such a reactor to a localized perturbation deviates significantly from point-kinetics. The space-dependence of the induced neutron noise thus carries enough information about the location of the noise source, which makes it possible to determine its position from a few detector readings. The identification of the type of noise source is easily performed from the in-phase behaviour of the induced neutron noise. Different unfolding techniques are finally tested. All these techniques rely on the use of the reactor transfer function. One of these techniques is based on the comparison between the actual measured neutron noise and the neutron noise calculated for every possible location of the noise source. This technique is very reliable and almost insensitive to the contamination of the detector signals by background noise, but also extremely CPU consuming. Another technique, based on the piece-wise inversion of the reactor transfer function and requiring little CPU effort, was developed. Although this technique is much less reliable when background noise is present, this technique is useful to indicate a region of the reactor where a noise source is likely to be located.
|
|
| 44. |
- Demaziere, Christophe, 1973, et al.
(författare)
-
On the possibility of the space-dependence of the stability indicator (decay ratio) of a BWR
- 2005
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:12, s. 1305-1322
-
Tidskriftsartikel (refereegranskat)abstract
- A model is proposed for the explanation of the space-dependence of the so-called decay ratio (DR) which is used to quantify the stability properties of boiling water reactors (BWRs). The study was prompted by the observation of a strongly space-dependent decay ratio in an instability event at the Swedish Forsmark-1 BWR. Prior to that event, the space-dependence of the DR was neither observed, nor assumed possible in the theoretical models of instability. The model proposed here is based on a previous suggestion by one of the authors on how to model the estimation of the DR in case of two different types of oscillations (instabilities) being present in the core simultaneously. The model was earlier only used in a space-independent form, but here its applicability is extended such that space-dependence of the oscillations is also accounted for, by using a noise simulator. The investigations show that the DR, as determined by the individual LPRMs (neutron detectors) at different positions, can be strongly space-dependent if at least two different oscillations with differing DR and space-dependence exist in the core simultaneously. The observed space-dependence of the DR in the Forsmark case can be reconstructed by the model.
|
|
| 45. |
- Demaziere, Christophe, 1973, et al.
(författare)
-
Understanding the neutron noise induced by fuel assembly vibrations in linear theory
- 2022
-
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.
|
|
| 46. |
- Deng, Yucheng, et al.
(författare)
-
An experimental study on the effect of coolant salinity on steam explosion
- 2024
-
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.
|
|
| 47. |
- Dickinson, S., et al.
(författare)
-
Experimental and modelling studies of iodine oxide formation and aerosol behaviour relevant to nuclear reactor accidents
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74, s. 200-207
-
Tidskriftsartikel (refereegranskat)abstract
- Plant assessments have shown that iodine contributes significantly to the source term for a range of accident scenarios. Iodine has a complex chemistry that determines its chemical form and, consequently, its volatility in the containment. If volatile iodine species are formed by reactions in the containment, they will be subject to radiolytic reactions in the atmosphere, resulting in the conversion of the gaseous species into involatile iodine oxides, which may deposit on surfaces or re-dissolve in water pools. The concentration of airborne iodine in the containment will, therefore, be determined by the balance between the reactions contributing to the formation and destruction of volatile species, as well as by the physicochemical properties of the iodine oxide aerosols which will influence their longevity in the atmosphere. This paper summarises the work that has been done in the framework of the EC SARNET (Severe Accident Research Network) to develop a greater understanding of the reactions of gaseous iodine species in irradiated air/steam atmospheres, and the nature and behaviour of the reaction products. This work has mainly been focussed on investigating the nature and behaviour of iodine oxide aerosols, but earlier work by members of the SARNET group on gaseous reaction rates is also discussed to place the more recent work into context.
|
|
| 48. |
- Dokhane, A., et al.
(författare)
-
Analysis of Oskarshamn-2 stability event using TRACE/SIMULATE-3K and comparison to TRACE/PARCS and SIMULATE-3K stand-alone
- 2017
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 102, s. 190-199
-
Tidskriftsartikel (refereegranskat)abstract
- With the goal to enhance the capability to perform best-estimate simulations of Light Water Reactors (LWRs) transients, with strong coupling between core neutronics and plant thermal-hydraulic, a coupling between TRACE and SIMULATE-3K (TS3K) was developed in collaboration between PSI and Studsvik for analyses involving interactions between system and core. In order to verify the coupling scheme and the coupled code capabilities to simulate complex transients, the OECD/NEA Oskarshmn-2 (O-2) Stability benchmark was modeled with the coupled code TS3K. The main goal of this paper is to present TS3K analyses of the Oskarshamn-2 stability event, noting that this constitutes the first reported assessment of this code system for a BWR stability problem. A systematic analysis is carried out using different time-space discretization schemes in order to identify an optimized methodology to simulate correctly the O-2 stability event. In this context, the TS3K results are compared to the available benchmark data both for steady-state and transient conditions. The results show that using a refined model in space and time, the TS3K model can successfully capture the entire behavior of the transient qualitatively, i.e. onset of the instability with growing oscillation amplitudes, as well as quantitatively, i.e. Decay Ratio and resonance frequency. In addition, the results are compared also to those obtained using TRACE/PARCS and S3K stand-alone, which allows a systematic comparison between different codes.
|
|
| 49. |
- Dufek, Jan, et al.
(författare)
-
An efficient parallel computing scheme for Monte Carlo criticality calculations
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1276-1279
-
Tidskriftsartikel (refereegranskat)abstract
- The existing parallel computing schemes for Monte Carlo criticality calculations suffer from a low efficiency when applied on many processors. We suggest a new fission matrix based scheme for efficient parallel computing. The results are derived from the fission matrix that is combined from all parallel simulations. The scheme allows for a practically ideal parallel scaling as no communication among the parallel simulations is required, and inactive cycles are not needed. (C) 2009 Elsevier Ltd. All rights reserved.
|
|
| 50. |
- Dufek, Jan, 1978
(författare)
-
Building the nodal nuclear data dependences in a many-dimensional state-variable space
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:7, s. 1569-1577
-
Tidskriftsartikel (refereegranskat)abstract
- We present new methods for building the polynomial-regression based nodal nuclear data models. Thedata models can reflect dependences on a large number of state variables, and they can consider varioushistory effects. Suitable multivariate polynomials that approximate the nodal data dependences are identifiedefficiently in an iterative manner. The history effects are analysed using a new sampling scheme forlattice calculations where the traditional base burnup and branch calculations are replaced by a largenumber of diverse burnup histories. The total number of lattice calculations is controlled so that the datamodels are built to a required accuracy.
|
|
