| 1. |
- Demaziere, Christophe, 1973, et al.
(author)
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Investigation of the MTC noise estimation with a coupled neutronic/thermal-hydraulic dedicated model – “Closing the loop”
- 2012
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In: Proc. Int. Conf. on Advances in Reactor Physics – Linking Research, Industry, and Education (PHYSOR 2012), Knoxville, TN, USA, April 15-20, 2012, American Nuclear Society. - 9781622763894 ; 1, s. 361-375
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Conference paper (peer-reviewed)abstract
- This paper investigates the reliability of different noise estimators aimed at determining the Moderator Temperature Coefficient (MTC) of reactivity in Pressurized Water Reactors. By monitoring the inherent fluctuations in the neutron flux and moderator temperature, an on-line monitoring of the MTC without perturbing reactor operation is possible. In order to get an accurate estimation of the MTC by noise analysis, the point-kinetic component of the neutron noise and the core-averaged moderator temperature noise have to be used. Because of the scarcity of the in-core instrumentation, the determination of these quantities is difficult, and several possibilities thus exist for estimating the MTC by noise analysis. Furthermore, the effect of feedback has to be negligible at the frequency chosen for estimating the MTC in order to get a proper determination of the MTC. By using an integrated neutronic/thermal-hydraulic model specifically developed for estimating the three-dimensional distributions of the fluctuations in neutron flux, moderator properties, and fuel temperature, different approaches for estimating the MTC by noise analysis can be tested individually. It is demonstrated that a reliable MTC estimation can only be provided if the core is equipped with a sufficient number of both neutron detectors and temperature sensors, i.e. if the core contain in-core detectors monitoring both the axial and radial distributions of the fluctuations in neutron flux and moderator temperature. It is further proven that the effect of feedback is negligible for frequencies higher than 0.1 Hz, and thus the MTC noise estimations have to be performed at higher frequencies.
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| 2. |
- Dykin, Victor, 1985, et al.
(author)
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Development of a fully-consistent Reduced Order Model to study instabilities in Boiling Water Reactors
- 2012
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In: Proc. Int. Conf. on Advances in Reactor Physics – Linking Research, Industry, and Education (PHYSOR 2012), Knoxville, TN, USA, April 15-20, 2012, American Nuclear Society. - 9781622763894 ; 1, s. 332 - 345
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Conference paper (peer-reviewed)abstract
- A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to other existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or C*^{V,D}_{mn} - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good qualitative agreement. The present study provides some insight in a deeper understanding of the physical principles which drive both core-wide and local instabilities.
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| 3. |
- Hernandéz Solís, Augusto, 1980, et al.
(author)
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Statistical uncertainty analysis applied to the DRAGONv4 code lattice calculations and based on JENDL-4 covariance data
- 2012
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In: Proc. Int. Conf. on Advances in Reactor Physics – Linking Research, Industry, and Education (PHYSOR 2012), Knoxville, TN, USA, April 15-20, 2012, American Nuclear Society. - 9781622763894 ; 4, s. 2960-2974
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Conference paper (peer-reviewed)abstract
- In this paper, multi-group microscopic cross-section uncertainty is propagated through the DRAGON (Version 4) lattice code, in order to perform uncertainty analysis on k_inf and 2-group homogenized macroscopic cross-sections predictions. 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 is based on JENDL-4 data, because JENDL-4 contains the largest amount of isotopic covariance matrixes among the different major nuclear data libraries. The aim is to propagate multi-group nuclide uncertainty by running the DRAGONv4 code 500 times, and to assess the output uncertainty of a test case corresponding to a 17x17 PWR fuel assembly segment without poison. The chosen sampling strategy for the current study is Latin Hypercube Sampling (LHS). The quasi-random LHS 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. Output uncertainty assessment is based on the tolerance limits concept, where the sample formed by the code calculations infers to cover 95% of the output population with at least a 95% of confidence. This analysis is the first attempt to propagate parameter uncertainties of modern multi-group libraries, which are used to feed advanced lattice codes that perform state of the art resonant self- shielding calculations such as DRAGONv4.
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| 4. |
- Holcombe, S., et al.
(author)
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Advanced fuel assembly characterization capabilities based on gamma tomography at the halden boiling water reactor
- 2012
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In: Proc. Int. Conf. on Advances in Reactor Physics. - 9781622763894 ; , s. 3478-3489
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Conference paper (peer-reviewed)abstract
- Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Institute for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala University.
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| 5. |
- Larsson, Irina, 1975, et al.
(author)
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Factors influencing helium measurements for detection of control rod failures in BWR
- 2012
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In: International Conference on the Physics of Reactors 2012: Advances in Reactor Physics (PHYSOR 2012), Knoxville, TN;15- 20 April 2012. - 9781622763894 ; 4, s. 3092-3099
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Conference paper (peer-reviewed)abstract
- Much effort has been made to minimize the number and consequences of fuel failures at nuclear power plants. The consequences of control rod failures have also gained an increased attention. In this paper we introduce a system for on-line surveillance of control rod integrity which has several advantages comparing to the surveillance methods available today in boiling water reactors (BWRs). This system measures the helium released from failed control rods containing boron carbide (B4C). However, there are a number of factors that might influence measurements, which have to be taken into consideration when evaluating the measured data. These factors can be separated into two groups: 1) local adjustments, made on the sampling line connecting the detector to the off-gas system, and 2) plant operational parameters. The adjustments of the sample line conditions include variation of gas flow rate and gas pressure in the line. Plant operational factors that may influence helium measurements can vary from plant to plant. The factors studied at Leibstadt nuclear power plant (KKL) were helium impurities in injected hydrogen gas, variation of the total off-gas flow and regular water refill. In this paper we discuss these factors and their significance and present experimental results of measurements at KKL.
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| 6. |
- Lau, Cheuk Wah, 1985, et al.
(author)
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Improvement of the thermal margins in the Swedish Ringhals-3 PWR by introducing new fuel assemblies with thorium
- 2012
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In: Proc. Int. Conf. on Advances in Reactor Physics – Linking Research, Industry, and Education (PHYSOR 2012), Knoxville, TN, USA, April 15-20, 2012, American Nuclear Society. - 9781622763894 ; 5, s. 4041-4055
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Conference paper (peer-reviewed)abstract
- Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium- thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins.
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| 7. |
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| 8. |
- Tran, Hoai Nam, 1981, et al.
(author)
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Neutron noise calculations in a hexagonal geometry and comparison with analytical solutions
- 2012
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In: International Conference on the Physics of Reactors 2012, PHYSOR 2012: Advances in Reactor Physics, Knoxville, TN, USA, April 15-20, 2012, American Nuclear Society. - 9781622763894 ; 5, s. 4107-4119
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Conference paper (peer-reviewed)abstract
- This paper presents the development of a neutronic and kinetic solver for hexagonal geometries. The tool is developed based on the diffusion theory with multi-energy groups and multi-groups of delayed neutron precursors allowing the solutions of forward and adjoint problems of static and dynamic states, and is applicable to both thermal and fast systems with hexagonal geometries. In the dynamic problems, the small stationary fluctuations of macroscopic cross sections are considered as noise sources, and then the induced first order noise is calculated fully in the frequency domain. Numerical algorithms for solving the static and noise equations are implemented with a spatial discretization based on finite differences and a power iterative solution. A coarse mesh finite difference method has been adopted for speeding up the convergence. Since no other numerical tool could calculate frequency-dependent noise in hexagonal geometry, validation calculations have been performed and benchmarked to analytical solutions based on a 2-D homogeneous system with two-energy groups and one-group of delayed neutron precursor, in which point-like perturbations of thermal absorption cross section at central and non-central positions are considered as noise sources.
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