| 1. |
- 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.
|
|
| 2. |
- 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.
|
|
| 3. |
- 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.
|
|
| 4. |
- 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.
|
|
| 5. |
- 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.
|
|
| 6. |
- 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.
|
|
| 7. |
- 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.
|
|
| 8. |
- 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.
|
|
| 9. |
- 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.
|
|
| 10. |
- 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.
|
|
| 11. |
- 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.
|
|
| 12. |
- 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.
|
|
| 13. |
- 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.
|
|
| 14. |
- 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.
|
|
| 15. |
- Dufek, Jan, et al.
(författare)
-
Derivation of a stable coupling scheme for Monte Carlo burnup calculations with the thermal-hydraulic feedback
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 62, s. 260-263
-
Tidskriftsartikel (refereegranskat)abstract
- Numerically stable Monte Carlo burnup calculations of nuclear fuel cycles are now possible with the previously derived Stochastic Implicit Euler method based coupling scheme. In this paper, we show that this scheme can be easily extended to include the thermal-hydraulic feedback during the Monte Carlo burnup simulations, while preserving its unconditional stability property. At each time step, the implicit solution (for the end-of-step neutron flux, fuel nuclide densities and thermal-hydraulic conditions) is calculated iteratively by the stochastic approximation; the fuel nuclide densities and thermal-hydraulic conditions are iterated simultaneously. This coupling scheme is derived as stable in theory; i.e.; its stability is not conditioned by the choice of time steps.
|
|
| 16. |
- Dufek, Jan, et al.
(författare)
-
Description of a stable scheme for steady-state coupled Monte Carlo-thermal-hydraulic calculations
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 68, s. 1-3
-
Tidskriftsartikel (refereegranskat)abstract
- We provide a detailed description of a numerically stable and efficient coupling scheme for steady-state Monte Carlo neutronic calculations with thermal-hydraulic feedback. While we have previously derived and published the stochastic approximation based method for coupling the Monte Carlo criticality and thermal-hydraulic calculations, its possible implementation has not been described in a step-by-step manner. As the simple description of the coupling scheme was repeatedly requested from us, we have decided to make it available via this note.
|
|
| 17. |
- Dufek, Jan, et al.
(författare)
-
Numerical stability of the predictor-corrector method in Monte Carlo burnup calculations of critical reactors
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 56, s. 34-38
-
Tidskriftsartikel (refereegranskat)abstract
- Monte Carlo burnup codes use various schemes to solve the coupled criticality and burnup equations. Previous studies have shown that the simplest methods, such as the beginning-of-step and middle-of-step constant flux approximations, are numerically unstable in fuel cycle calculations of critical reactors. Here we show that even the predictor-corrector methods that are implemented in established Monte Carlo burnup codes can be numerically unstable in cycle calculations of large systems.
|
|
| 18. |
- Dufek, Jan, et al.
(författare)
-
The stochastic implicit Euler method - A stable coupling scheme for Monte Carlo burnup calculations
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 60, s. 295-300
-
Tidskriftsartikel (refereegranskat)abstract
- Existing Monte Carlo burnup codes use various schemes to solve the coupled criticality and bumup equations. Previous studies have shown that the coupling schemes of the existing Monte Carlo burnup codes can be numerically unstable. Here we develop the Stochastic Implicit Euler method - a stable and efficient new coupling scheme. The implicit solution is obtained by the stochastic approximation at each time step. Our test calculations demonstrate that the Stochastic Implicit Euler method can provide an accurate solution to problems where the methods in the existing Monte Carlo burnup codes fail.
|
|
| 19. |
- Dufek, Jan, et al.
(författare)
-
Time step length versus efficiency of Monte Carlo burnup calculations
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 72, s. 409-412
-
Tidskriftsartikel (refereegranskat)abstract
- We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy.
|
|
| 20. |
- Dykin, Victor, 1985, et al.
(författare)
-
Investigation of global and regional BWR instabilities with a four heated-channel Reduced Order Model
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 53, s. 381-400
-
Tidskriftsartikel (refereegranskat)abstract
- The development of an advanced Reduced Order Model (ROM) including four heated channels and meant to study global and regional Boiling Water Reactor (BWR) instabilities is described. The ROM contains three sub-models: a neutron-kinetic model (describing neutron transport), a thermal-hydraulic model (describing fluid transport) and a heat transfer model (describing heat transfer between the fuel and the coolant). All these three models are coupled to each other using two feedback mechanisms: the void feedback and the doppler feedback mechanisms. Each of the sub-models is described by a set of reduced ordinary differential equations, derived from the corresponding time- and space-dependent partial differential equations, by using different types of approximations and mathematical techniques that are explained in this paper.One of the novelties of the present ROM is that it takes the effect of the first three neutronic modes into account, namely the fundamental, first, and second azimuthal modes. In order to have a proper representation of both azimuthal modes and of their dependence on the thermal-hydraulic conditions in the heated channels, a four heated channel ROM was constructed. Another novelty of the present work is to develop a special methodology which guarantees the full consistency between the spatial discretization procedures used in the dynamical calculations and the ones implemented in the static case. Accordingly, a re-computation of the static solution based on the CORE SIM tool was embedded into the ROM in such a way that the balance equations expressing the conservation of neutron balance, heat generation, and mass, momentum, enthalpy for the flow, could be fulfilled for the steady-state solution of the coupled neutron-kinetic/thermal-hydraulic problem. Once the static problem is solved, the time-dependent solution in case of a perturbed system can be determined. Moreover, a non-uniform power profile representing different heat production rates in the one- and two-phase regions was introduced into the ROM. Careful attention was paid to the determination of the coupling coefficients for the reactivity effects related to both void fraction and fuel temperature, so that such coefficients correspond to the re-computed static solution. The evaluation of these coefficients was based on the cross-section perturbations estimated by the SIMULATE-3 code, and on the different neutronic eigenmodes of the heterogeneous core determined by the CORE SIM tool.
|
|
| 21. |
- Dykin, Victor, 1985, et al.
(författare)
-
Investigation of local BWR instabilities with a four heated-channel Reduced Order Model
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 53, s. 320-330
-
Tidskriftsartikel (refereegranskat)abstract
- his paper deals with the modeling of Boiling Water Reactor (BWR) local instabilities via so-called Reduced Order Models (ROMs). More specifically, a four-heated channels ROM, which was earlier developed (Dykin et al., submitted for publication), was modified in such a way that the effect of local perturbations could also be accounted for.This model was thereafter used to analyze a local instability event that took place at the Swedish Forsmark-1 BWR in 1996/1997. Such a local instability was driven by unseated fuel assemblies. Comparisons between the results of ROM simulations and actual measurement data demonstrated that the developed ROM was able to correctly reproduce the main features of the event. The ROM has also the ability to give some further physical insights into the phenomena taking place in case of instabilities. For the particular instability event investigated, it was for instance demonstrated that the global and regional oscillation modes were stable, but were excited by the local oscillation acting as an external perturbation. When performing a modal decomposition of the measured neutron flux in case of an instability event driven by a local oscillation, each mode will apparently be excited, whereas in reality such modes might be stable. Such an apparent contradictory behavior is due to the inability of a modal decomposition to catch with only a few modes the spatial dependence of the neutron flux in case of a local oscillation.
|
|
| 22. |
- Fokau, Andrei, et al.
(författare)
-
A source efficient ADS for minor actinides burning
- 2010
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 37:4, s. 540-545
-
Tidskriftsartikel (refereegranskat)abstract
- Taking advantage of the good neutron economy of nitride fuel, a compact accelerator-driven system (ADS) for burning of minor actinide fuels has been designed, based on the fuel assembly geometry developed for the European Facility for Industrial Transmutation (EFIT) within the EUROTRANS project. The small core size of the new design permits reduction of the size of the spallation target region, which enhances proton source efficiency by about 80% compared to the reference oxide version of EFIT. Additionally, adoption of the austenitic steel 15/15Ti as clad material allows to safely reduce the fuel pin pitch, which leads to an increase of fuel volume fraction and therefore makes the neutron energy spectrum faster, consequently increasing minor actinides fission probabilities. Our calculations show that one can dramatically increase neutron source efficiency up to 0.95 without a significant loss of neutron source intensity, i.e. having high proton source efficiency. Consequently, the accelerator current required for operation of the ADS with a fission power of 201 MWth and a burn-up of 27 GW d/t per year (365 EFPD) is reduced by 67%.
|
|
| 23. |
- Gajev, Ivan, et al.
(författare)
-
Sensitivity analysis of input uncertain parameters on BWR stability using TRACE/PARCS
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 67, s. 49-58
-
Tidskriftsartikel (refereegranskat)abstract
- The unstable behavior of Boiling Water Reactors (BWR), which is known to occur at certain power and flow conditions, could cause SCRAM and decrease the economic performance of the plant. For better prediction of BWR stability and understanding of influential parameters, two TRACE/PARCS models of Ringh-als-1 and Oskarshamn-2 BWRs were employed to perform a sensitivity study. Using the propagation of input errors uncertainty method's results, an attempt has been made to identify the most influential parameters affecting the stability. Furthermore, a methodology using the spearman rank correlation coefficient has been used to identify the most influential parameters on the stability parameters (decay ratio and frequency).
|
|
| 24. |
- Gajev, Ivan, et al.
(författare)
-
Space–time convergence analysis on BWR stability using TRACE/PARCS
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 51, s. 295-306
-
Tidskriftsartikel (refereegranskat)abstract
- Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.
|
|
| 25. |
- Hernandéz Solís, Augusto, 1980, et al.
(författare)
-
Uncertainty and sensitivity analyses applied to the DRAGONv4.05 code lattice calculations and based on JENDL-4 data
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 57, s. 230-245
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, multi-group microscopic cross-section uncertainties are propagated through the DRAGON (Version 4.05) lattice code in order to perform uncertainty analysis on k(infinity) and 2-group homogenized macroscopic cross-sections. The test case corresponds to a 17 x 17 PWR fuel assembly segment without poison at full power conditions. A statistical methodology is employed for such purposes, where cross-sections of certain isotopes of various elements belonging to the 172 groups DRAGLIB library format, are considered as normal random variables. This library was based on JENDL-4 data, because JENDL-4 contains a large amount of isotopic covariance matrices among the different major nuclear data libraries. Thus, multi-group uncertainty was computed for the different isotopic reactions by means of ERRORRJ. The preferred sampling strategy for the current study corresponds to the quasi-random Latin Hypercube Sampling (LHS). This technique allows a much better coverage of the input uncertainties than simple random sampling (SRS) because it densely stratifies across the range of each input probability distribution. In order to prove this, the uncertain input space was re-sampled 10 times, and it is shown that the variability of the replicated mean of the different k(infinity) samples is much less for the LHS case, than for the SRS case. The uncertainty assessment of the output space should be based on the theory of non-parametric multivariate tolerance limits, due to the fact that k(infinity) and some of the macroscopic cross-sections are correlated. Therefore, for 10 replicated samples each containing 100 elements, the total output sample is composed by 1000 calculations. This sample size is more than enough to infer that the multivariate output population is covered 95% with a 95% of confidence. On the other hand, statistical sensitivity analysis was performed in order to know which microscopic cross-section has the greatest impact on k(infinity) predictions. It was found that the fission cross-section of Uranium 235 is the dominant input parameter for this particular case, because the computed JENDL-4 variances for such reaction are very high at thermal and resonant regions compared to other variances that for instance, can be computed based on other nuclear libraries such as ENDF/B-VII.1
|
|
| 26. |
- Holcombe, Scott, et al.
(författare)
-
Feasibility of identifying leaking fuel rods using gamma tomography
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 57, s. 334-340
-
Tidskriftsartikel (refereegranskat)abstract
- In cases of fuel failure in irradiated nuclear fuel assemblies, causing leakage of fission gasses from a fuel rod, there is a need for reliable non-destructive measurement methods that can determine which rod is failed. Methods currently in use include visual inspection, eddy current, and ultrasonic testing, but additional alternatives have been under consideration, including tomographic gamma measurements.The simulations covered in this report show that tomographic measurements could be feasible. By measuring a characteristic gamma energy from fission gasses in the gas plenum, the rod-by-rod gamma source distribution within the fuel rod plena may be reconstructed into an image or data set which could then be compared to the predicted distribution of fission gasses, e.g. from the STAV code. Rods with significantly less fission gas in the plenum may then be identified as leakers.Results for rods with low fission gas release may, however, in some cases be inconclusive since these rods will already have a weak contribution to the measured gamma-ray intensities and for such rods there is a risk that a further decrease in fission gas content due to a leak may not be detectable. In order to evaluate this and similar experimental issues, measurement campaigns are planned using a tomographic measurement system at the Halden Boiling Water Reactor.
|
|
| 27. |
- Isotalo, A. E., et al.
(författare)
-
Preventing xenon oscillations in Monte Carlo burnup calculations by enforcing equilibrium xenon distribution
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 60, s. 78-85
-
Tidskriftsartikel (refereegranskat)abstract
- Existing Monte Carlo burnup codes suffer from instabilities caused by spatial xenon oscillations. These oscillations can be prevented by forcing equilibrium between the neutron flux and saturated xenon distribution. The equilibrium calculation can be integrated to Monte Carlo neutronics, which provides a simple and lightweight solution that can be used with any of the existing burnup calculation algorithms. The stabilizing effect of this approach, as well as its limitations are demonstrated using the reactor physics code Serpent.
|
|
| 28. |
- Jareteg, Klas, 1986, et al.
(författare)
-
Fine-mesh deterministic modeling of PWR fuel assemblies: Proof-of-principle of coupled neutronic/thermal–hydraulic calculations
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 68, s. 247-256
-
Tidskriftsartikel (refereegranskat)abstract
- This paper investigates the feasibility of developing a fine mesh coupled neutronic/thermal–hydraulic solver within the same computing platform for selected fuel assemblies in nuclear cores. As a first step in this developmental work, a Pressurized Water Reactor at steady-state conditions was considered. The system being simulated has a finite axial size, but is infinite in the radial direction. The platform used for the modeling is based on the open source C++ library OpenFOAM. The thermal–hydraulics is solved using the built-in SIMPLE algorithm for the mass and momentum fields of the fluid, complemented by an equation for the temperature field applied simultaneously to all the regions (i.e. fluid and solid structures). For the neutronics, a two-group neutron diffusion-based solver was developed, with sets of macroscopic cross-sections generated by the Monte Carlo code SERPENT. The meshing of the system was created by the open source software SALOME. Successful convergence of the neutronic and thermal–hydraulic fields was achieved, thus bringing the solution of the coupled problem to an unprecedented level of details. Most importantly, the true interdependence of the different fields is automatically guaranteed at all scales. In addition, comparisons with a coarse-mesh radial averaging of the thermal–hydraulic variables show that a coarse-mesh fuel temperature identical for all fuel pins can lead to discrepancies of up to 0.5% in pin powers, and of several tens of pcm in multiplication factor.
|
|
| 29. |
- Jonsson, Anders, 1984, et al.
(författare)
-
Two-group theory of neutron noise in Molten Salt Reactors
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:6, s. 1238-1251
-
Tidskriftsartikel (refereegranskat)abstract
- In a previous paper, the kinetics, dynamics and the neutron noise in a Molten Salt Reactor (MSR) was investigated in a simple homogeneous reactor model in one-group diffusion theory. In this paper those investigations are extended to two-group theory in the same reactor model. In addition, unlike in the previous paper where in the quantitative work data from a thermal LWR were used, here material data of a conceptual MSR with a thermal spectrum and thorium fuel are used, along with data from both fast and thermal LWRs. Among other things, the relative weight and the range of the local component is investigated. It is found that the strong neutronic coupling in an MSR, which was pointed out in the preceding paper, diminishes the role of the local component as compared to light water reactors. Some further new features of the noise in MSR, not directly related to the two-group approach, are also found. (C) 2011 Elsevier Ltd. All rights reserved.
|
|
| 30. |
- Klein-Hessling, W., et al.
(författare)
-
Conclusions on severe accident research priorities
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74, s. 4-11
-
Tidskriftsartikel (refereegranskat)abstract
- The objectives of the SARNET network of excellence are to define and work on common research programs in the field of severe accidents in Gen. II-III nuclear power plants and to further develop common tools and methodologies for safety assessment in this area. In order to ensure that the research conducted on severe accidents is efficient and well-focused, it is necessary to periodically evaluate and rank the priorities of research. This was done at the end of 2008 by the Severe Accident Research Priority (SARP) group at the end of the SARNET project of the 6th Framework Programme of European Commission (FP6). This group has updated this work in the FP7 SARNET2 project by accounting for the recent experimental results, the remaining safety issues as e.g. highlighted by Level 2 PSA national studies and the results of the recent ASAMPSA2 FP7 project. These evaluation activities were conducted in close relation with the work performed under the auspices of international organizations like OECD or IAEA. The Fukushima-Daiichi severe accidents, which occurred while SARNET2 was running, had some effects on the prioritization and definition of new research topics. Although significant progress has been gained and simulation models (e.g. the ASTEC integral code, jointly developed by IRSN and GRS) were improved, leading to an increased confidence in the predictive capabilities for assessing the success potential of countermeasures and/or mitigation measures, most of the selected research topics in 2008 are still of high priority. But the Fukushima-Daiichi accidents underlined that research efforts had to focus still more to improve severe accident management efficiency.
|
|
| 31. |
- Kouhia, Virpi, et al.
(författare)
-
Benchmark exercise on SBLOCA experiment of PWR PACTEL facility
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 59, s. 149-156
-
Tidskriftsartikel (refereegranskat)abstract
- The PWR PACTEL benchmark exercise was organized in Lappeenranta, Finland by Lappeenranta University of Technology. The benchmark consisted of two phases, i.e. a blind and an open calculation task. Seven organizations from the Czech Republic, Germany, Italy, Sweden and Finland participated in the benchmark exercise, and four system codes were utilized in the benchmark simulation tasks. Two workshops were organized for launching and concluding the benchmark, the latter of which involved presentations of the calculation results as well as discussions on the related modeling issues. The chosen experiment for the benchmark was a small break loss of coolant accident experiment which was performed to study the natural circulation behavior over a continuous range of primary side coolant inventories. For the blind calculation task, the detailed facility descriptions, the measured pressure and heat losses as well as the results of a short characterizing transient were provided. For the open calculation task part, the experiment results were released. According to the simulation results, the benchmark experiment was quite challenging to model. Several improvements were found and utilized especially for the open calculation case. The issues concerned model construction, heat and pressure losses impact, interpreting measured and calculated data, non-condensable gas effect, testing several condensation and CCFL correlations, sensitivity studies, as well as break modeling. There is a clear need for user guidelines or for a collection of best practices in modeling for every code. The benchmark offered a unique opportunity to test the best practices and solutions in modeling and analyzing tasks as well as a possibility to increase knowledge about the interpretation of test results. The benchmark exercise served as a practical and rewarding forum to discuss the needs, problems and possibilities in the analysis and in producing useful data with an experiment facility. The workshops provided an advantageous site for interaction of the code users and the experimenters.
|
|
| 32. |
- Kozlowski, Tomasz, et al.
(författare)
-
Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 67, s. 4-12
-
Tidskriftsartikel (refereegranskat)abstract
- On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations, and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one.
|
|
| 33. |
- Larsson, Viktor, 1984, et al.
(författare)
-
A coupled neutronics/thermal–hydraulics tool for calculating fluctuations in Pressurized Water Reactors
- 2012
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 43, s. 68-76
-
Tidskriftsartikel (refereegranskat)abstract
- This paper describes a tool for estimating fluctuations in neutron flux, fuel temperature, moderator density and flow velocity in Pressurized Water Reactors by coupling a dynamic thermal–hydraulic module and a dynamic neutron kinetic module. The code calculates the static solution first, giving the profile of the static fuel temperature, moderator density, velocity and neutron flux. The fluctuations (called noise in this work) are the differences between the actual time-dependent values and the corresponding mean values. The fluctuations are in general induced by perturbations in the thermal–hydraulic parameters, e.g. moderator temperature or density, at the inlet of the core. There is also a possibility to directly define the perturbations in the macroscopic cross-sections and to supply them to the neutronic part of the model. The model was validated against two separate calculations using two different commercial tools.
|
|
| 34. |
- Larsson, Viktor, 1984, et al.
(författare)
-
Comparison of the calculated neutron noise using finite differences and the Analytical Nodal Method
- 2012
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 43, s. 176-182
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, a comparison of the calculated neutron noise, i.e. the fluctuation of the neutron flux around its average value assuming that all processes are stationary, is conducted, where the neutron noise is calculated using finite differences alone and with finite differences where the Analytical Nodal Method is used to correct the neutron currents, respectively. It is seen that the lower the frequency of the noise source, the larger difference between the two solutions. The main conclusion from this work is that the gain of calculating the neutron noise using the more sophisticated Analytical Nodal Method compared to the increase of the corresponding computational burden is too little to motivate the use of the ANM.
|
|
| 35. |
- Larsson, Viktor, 1984, et al.
(författare)
-
Neutron noise calculations using the Analytical Nodal Method and comparisons with analytical solutions
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:4, s. 808-816
-
Tidskriftsartikel (refereegranskat)abstract
- In this study, the neutron noise, i.e. the stationary fluctuations of the neutron flux around its mean value, is calculated in 2-group diffusion theory using the Analytical Nodal Method. A brief description of the calculation of the static flux is also included. The static solution is benchmarked against a reference solution in the case of a homogeneous core. The same calculational scheme for the neutron noise as for the static flux is used. As a dynamical benchmark, the calculated neutron noise for a 2D fully homogeneous reactor is compared with the analytical solution of a centered noise source at different frequencies. The numerical solution is also benchmarked to an off-centered source where the analytical solution is determined using the power reactor approximation, extended to two energy groups. The results of the benchmarks are that the numerical calculations using ANM accurately match the analytical solutions.
|
|
| 36. |
- Lau, Cheuk Wah, 1985, et al.
(författare)
-
Conceptual study of axial offset fluctuations upon stepwise power changes in a thorium-plutonium core to improve load-following conditions
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 72, s. 84-89
-
Tidskriftsartikel (refereegranskat)abstract
- The increased share of renewable energy, such as wind and solar power, will increase the demand for load-following power sources, and nuclear reactors could be one option. However, during rapid load-following events, traditional UOX cores could be restricted by the volatile oscillation of the power distribution. Therefore, a conceptual study on stability properties of Th-MOX PWR concerning axial offset power excursion during load-following events are investigated and discussed. The study is performed in SIMULATE-3 for a realistic PWR core (Ringhals-3) at the end of cycle, where the largest amplitude of the axial offset oscillations is expected. It is shown that the Th-MOX core possesses much better stability characteristics and shorter reactor dead time compared with a traditional UOX core, and the main reasons are the lower sensitivity to perturbations in the neutron spectrum, lower xenon poisoning and lower thermal neutron flux.
|
|
| 37. |
- Loberg, John, et al.
(författare)
-
Investigation of axial power gradients near a control rod tip
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:7, s. 1609-1615
-
Tidskriftsartikel (refereegranskat)abstract
- Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, similar to 15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.
|
|
| 38. |
- Meignen, Renaud, et al.
(författare)
-
Status of steam explosion understanding and modelling
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74, s. 125-133
-
Tidskriftsartikel (refereegranskat)abstract
- The main results of the major international activities related to fuel coolant interactions (FCI) of the last 4-year period are presented and a summary of the knowledge gained regarding understanding and the improvements of modelling is provided. At first, the major outcomes of the OECD SERENA-2 program are presented and discussed. Important clarifications were obtained on the so-called material effect and on FCI code capabilities. We then summarise complementary analytical analyses and experimental programs performed in the frame of the SARNET community. The focus was put on the role of melt fragmentation and solidification, the impact of void on the intensity of an explosion and the triggering mechanisms. As a conclusion, tables summarising the improvements are proposed as well as research priorities.
|
|
| 39. |
- Muñoz-Cobo, José-Luis, et al.
(författare)
-
Feynman-alpha and Rossi-alpha formulas with spatial and modal effects
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:2-3, s. 590-600
-
Tidskriftsartikel (refereegranskat)abstract
- Feynman-alpha and Rossi-alpha formulas including multiple alpha-modes are derived for stochastic and continuous neutron sources. The presented formalism is further developed to achieve spatial correction factors for the single alpha-mode point kinetics representations of the Feynman-alpha and Rossi-alpha formulas. As a natural extension of the multiple alpha-mode formalism, delayed neutrons are included in the Feynman-alpha formula. The obtained formulas are validated experimentally in a strongly heterogeneous system obeying multiple alpha-modes, resulting in good agreement with the presented theoretical framework.
|
|
| 40. |
- Noack, Klaus, et al.
(författare)
-
Comments on the power amplification factor of a driven subcritical system
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 59, s. 261-266
-
Tidskriftsartikel (refereegranskat)abstract
- The power amplification factor PAF of a driven subcritical system is defined as the ratio of the fission power output of the blanket to the power which the driver must deliver to sustain its neutron source intensity. This parameter decisively determines the effectiveness of the whole system independent of its special purpose as energy amplifier or as transmutation facility. The present note derives a refined analytical expression for the PAF which reveals more physical details than the expressions given by other authors. Moreover, the traditionally used forms of the static reactor eigenvalue equation and of its adjoint equation are rewritten for subcritical systems and used in the derivation of the expression for the PAF. The derived formula and the modified eigenvalue equations are discussed.
|
|
| 41. |
- Noack, Klaus, et al.
(författare)
-
Neutronic model of a mirror based fusion-fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification
- 2011
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:2-3, s. 578-589
-
Tidskriftsartikel (refereegranskat)abstract
- The Georgia Institute of Technology has developed several design concepts of tokamak based fusion-fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a "fusion transmutation of waste reactor". It was designed for a criticality keff <= 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GWth requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBRcycle = 1.9 and a first wall lifetime of 12-16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead-bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality keff <= 0.97 and produces less fission power of 1.5 GWth. Its main performance parameters turn out substantially better.
|
|
| 42. |
- Oberstedt, S., et al.
(författare)
-
First results on the neutron-induced fission cross-section of Pa-231 for incident neutron energies E-n > 17 MeV
- 2012
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 43, s. 26-30
-
Tidskriftsartikel (refereegranskat)abstract
- First results on the neutron-induced fission cross-section of Pa-231 for incident neutron energies E-n > 17 MeV are presented. The experiments were carried out with quasi mono-energetic neutrons produced in the reaction T(d, n)He-4. Corrections for low-energy neutron background produced in this reaction at incident deuteron energies E-d > 2 MeV are taken into account and based on experimental data obtained by two different techniques. Despite the relatively large error bars at the higher neutron energies, the new cross-section values meet the accuracy requirements set by the IAEA and will allow to remove the hitherto existing large spread between different previously published data. Recent cross-section calculations describe well the new experimental results, which are in consistency with cross-section values obtained in a particle-transfer reaction at excitation energies corresponding to neutron energies E-n < 10 MeV.
|
|
| 43. |
- Pazsit, Imre, 1948, et al.
(författare)
-
Analytical solutions of the molten salt reactor equations
- 2012
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 50, s. 206-214
-
Tidskriftsartikel (refereegranskat)abstract
- The one-group diffusion theory of molten salt reactors in a homogeneous reactor model is revisited. First, the integral terms in the equation for the flux, obtained after the elimination of the delayed neutron precursors, are given a physical interpretation. This gives an understanding of the physical meaning of the concept of infinite fuel recirculation velocity, which eliminates one of the two integral terms, introduced in earlier work in order to find analytic solutions. In the light of the physical interpretation, another approximation, representing a different limiting case can be defined, corresponding to long recirculation times, i.e. no recirculation of the delayed neutron precursors to the core. This approximation incurs the neglecting of the other integral term, and it can also be solved analytically. Finally it is shown that the full equation, without neglecting any of the integral terms, has also a compact analytical solution and it is demonstrated how the case of the infinite velocity can be obtained as a limit case of the full solution. The analytical solutions open up the possibility to study a number of questions in analytical form, such as the calculation of the point kinetic response of the reactor, or the effect of the different boundary conditions. As an illustration, the solutions corresponding to the vanishing of the flux at the extrapolated boundary are compared to those obtained from the logarithmic boundary conditions.
|
|
| 44. |
- Pazsit, Imre, 1948, et al.
(författare)
-
Investigation of the space-dependent noise induced by propagating perturbations
- 2010
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 37:10, s. 1329-1340
-
Tidskriftsartikel (refereegranskat)abstract
- The space-dependent behaviour of the neutron noise, induced by perturbations represented by the fluctuations of the absorption cross sections, propagating with the coolant of a PWR, is investigated in a one-dimensional one-group approach. The general space–frequency dependent problem is solved for this specific noise source with the help of the Green’s function technique. All calculations are made in the frame of first-order perturbation theory. The solution is investigated for different frequencies and system sizes. The limits of point kinetic and space-dependent behaviour were investigated. An interesting interference phenomenon was found between the point kinetic and the pure space dependent components of the noise for certain domains of the frequency and system size. The results bear some significance for the dynamics of Molten Salt Reactors (MSR), which is reported on in a companion paper.
|
|
| 45. |
- Pazsit, Imre, 1948, et al.
(författare)
-
The point kinetic component of neutron noise in an MSR
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 64, s. 344-352
-
Tidskriftsartikel (refereegranskat)abstract
- The point kinetic approximation, and the calculation of the point kinetic component of the neutron noise in Molten Salt Reactors (MSR) is revisited. First, the derivation of the point kinetic equations in an MSR, found in the literature, is discussed. It is shown that to make the equations solvable, some simplifications need to be made whose validity is not justified. Then the point kinetic component of the noise is derived from the full space-frequency dependent solution in an analytical form by a projection to the static adjoint. The solution of the simplified point kinetic equations and the linearly exact point kinetic component obtained by the projection technique are compared quantitatively and it is shown that the solution of the simplified point kinetic equations cannot reconstruct some important features of the true solution. (C) 2013 Elsevier Ltd. All rights reserved.
|
|
| 46. |
- Peltonen, Joanna, et al.
(författare)
-
Effective spatial mapping for coupled code analysis of thermal-hydraulics/neutron-kinetics of boiling water reactors
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 63, s. 461-485
-
Forskningsöversikt (refereegranskat)abstract
- Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal-hydraulics (TH) codes. In order to produce results within a reasonable computing time, the coupled codes use two different spatial description of the reactor core. The TH code uses few, typically 5-20 TH channels, to represent the core. The NK code uses one explicit node for each fuel assembly. Therefore, a spatial mapping of a coarse grid TH and a fine grid NK domain is necessary. However, improper mappings may result in loss of valuable information, thus causing inaccurate prediction of safety parameters. In this article the study of the effectiveness of spatial coupling (channel -refinement and spatial mapping) and developed recommendations for NK/TH mapping are presented. The sensitivity of stability (measured by Decay Ratio and Frequency) to the different types of mapping schemes is analyzed against OECD/NEA Ringhals-1 Stability Benchmark data. Additionally, to increase the efficiency and applicability of spatial mapping convergence, a new mapping methodology is proposed. The new mapping approach is based on hierarchical clustering method; the method of unsupervised learning that is adopted in many different scientific fields, thanks to its flexibility and robustness. The proposed new mapping method is shown to be very successful for spatial coupling problem and can be fully automated allowing for significant time reduction in input preparation and mapping convergence study.
|
|
| 47. |
- Pohlner, G., et al.
(författare)
-
Analyses on ex-vessel debris formation and coolability in SARNET frame
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74, s. 50-57
-
Tidskriftsartikel (refereegranskat)abstract
- The major aim of work in the SARNET2 European project on ex-vessel debris formation and coolability was to get an overall perspective on coolability of melt released from a failed reactor pressure vessel and falling into a water-filled cavity. Especially, accident management concepts for BWRs, dealing with deep water pools below the reactor vessel, are addressed, but also shallower pools in existing PWRs, with questions about partial cooling and time delay of molten corium concrete interaction. The subject can be divided into three main topics: (i) Debris bed formation by breakup of melt, (ii) Coolability of debris and (iii) Coupled treatment of the processes. Accompanied by joint collaborations of the partners, the performed work comprises theoretical, experimental and modelling activities. Theoretical work was done by KTH on the melt outflow conditions from a RPV and on the quantification of the probability of yielding a non-coolable ex-vessel bed by use of probabilistic assessment. IKE introduced a theoretical concept to improve debris bed coolability. A large amount of experimental work was done by partners (KTH, VTT, IKE) on the coolability of debris beds using different bed geometries, particles, heating methods and water feeds, yielding a valuable base for code validation. Modelling work was mainly done by IKE, IRSN, RSE and VTT concerning jet breakup and/or debris bed formation and cooling in 2D and 3D geometries. A benchmark for the DEFOR-A experiment of KTH was performed. Important progress was reached for several tasks and aspects and important insights are given, enabling to focus the view on possible key aspects of future activities.
|
|
| 48. |
- Qvist, Staffan, 1986-, et al.
(författare)
-
The ADOPT code for automated fast reactor core design
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 71, s. 23-36
-
Tidskriftsartikel (refereegranskat)abstract
- The Assembly Design and OPTimization code (ADOPT) is a comprehensive computer code written to automate the process of designing and analyzing fast reactor fuel assemblies and cores. It finds a fuel assembly design that maximizes the fuel volume fraction in the core while adhering to set constraints for all component temperatures, pressure drop, coolant velocity and structural integrity limits, subjected to a specified assembly peak power level. ADOPT can be used very effectively as the first step-in the design process of fast reactor cores that offer the maximum possible breeding ratio, which is proportional to the fuel volume fraction. To design fast reactor cores with different objectives, one can start with a neutronic analysis to find material volume fractions that provide the sought core performance. ADOPT can then reverse-engineer a fuel assembly design with the desired volume fractions that abide by all the thermal-hydraulic and structural constraints. The code provides the necessary input files for a full core analysis to either SERPENT or MCNP neutron transport codes. Power and flux profiles from neutron transport calculations are then used to refine the ADOPT solution until a converged solution, considering thermal-hydraulics, structural mechanics and neutronics is achieved.
|
|
| 49. |
- Suvdantsetseg, Erdenechimeg, 1983-, et al.
(författare)
-
An assessment of prompt neutron reproduction time in a reflector dominated fast critical system : ELECTRA
- 2014
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 71, s. 159-165
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, an accurate method to evaluate the prompt neutron reproduction time for a reflector dominated fast critical reactor, ELECTRA, is discussed. To adequately handle the problem, explicit time dependent Monte Carlo calculations with MCNP, applying repeated time cut-off technique, is used and compared against the σ ∼ 1/v time dependent absorber method, applying artificial cross section data in the Monte Carlo code SERPENT. The results show that when a reflector plays a major role in criticality for fast neutron reactor, the two methods predict different physical parameters (Λ = 69 ± 2 ns and Λ = 83 ± 1 ns for time cut-off and the 1/v method respectively). The reason is explained by applying Avery-Cohn’s two-region prompt neutron model.
|
|
| 50. |
- Tampouratzi, Tatiani, 1965, et al.
(författare)
-
A general regression artificial neural network for two-phase flow regime identification
- 2010
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 37:5, s. 672-680
-
Tidskriftsartikel (refereegranskat)abstract
- Supplementing the collection of artificial neural network methodologies devised for monitoring energy producing installations, a general regression artificial neural network is proposed for the identification of the two-phase flow that occurs in the coolant channels of boiling water reactors. The utilization of a limited number of image features derived from radiography images affords the proposed approach with efficiency and non-invasiveness. Additionally, the application of counter-clustering to the input patterns prior to training accomplishes an 80% reduction in network size as well as in training and test time. Cross-validation tests confirm accurate on-line flow regime identification. (C) 2010 Elsevier Ltd. All rights reserved.
|
|
