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| Author | : Friederike Bostelmann |
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| Release | : 2020 |
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Mots-clés de l'auteur: sodium-cooled fast reactor ; uncertainty analysis ; sensitivity analysis ; nuclear data ; random sampling ; XSUSA ; SCALE.
| Author | : Jun Shi |
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| Release | : 2016 |
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The 1000 MWt Advanced Burner Reactor (ABR-1000) is a concept of Sodium-cooled Fast Reactor (SFR) developed at Argonne National Laboratory for the study of future reactor designs under the Global Nuclear Energy Partnership (GNEP) program. It was investigated within the OECD/NEA Working Party on Reactor Systems (WPRS) under the Sodium-cooled Fast Reactor core Feed-back and Transient response (SFR-FT) task force benchmark, which was completed in 2014. The results revealed that different nuclear data libraries contribute to the large discrepancies in some calculated neutronic parameters. This task force is followed up by another on-going OECD/NEA WPRS activity entitled as SFR Uncertainty Analysis in Modeling (SFR-UAM). In order to further investigate the properties of the ABR core, the impact of nuclear data uncertainties on the performance of a SFR is analyzed in detail in this master thesis using a "Best Estimate Plus Uncertainty" (BEPU) approach along with the nuclear data from the ENDF/B-VII.1 library. Several computer codes, including MC2-3, TWODANT, DIF3D, REBUS-3, PERSENT, DPT, and SAS4A/SASSYS-1, were used in this study. Significant uncertainties on neutronic parameters (e.g., sodium density coefficient, sodium void coefficient, structure density coefficient, Doppler coefficient) are found due to nuclear data, but thanks to the excellent reactor design, the margins to sodium boiling and fuel melting during the accidents are still large even if these non-negligible nuclear data uncertainties are considered.
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| Release | : 1978 |
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A review of present methods and existing computer codes indicates an enormous capability not only to calculate sensitivity coefficients but also to apply them to a variety of purposes. However, there are still many limitations to our present capabilities. One of these limitations has been our inability to include in a complete and systematic way the effect of methods uncertainties on the determination of adjusted data, which depends, in general, not only on experimental measurements, but also on estimates of covariances associated with the measurements and the methods. Also, the uncertainty in adjusted data contains contributions from uncertainties in covariance estimates which contributions we have heretofore neglected. A new and comprehensive approach to include effects of methods uncertainties is presented here, and all sources which contribute to the uncertainty of the adjusted data are considered. This new approach is demonstrated using rough estimates for the methods uncertainties as applied to a simplified representation of the ZPR-6/7 fast benchmark. The results indicate that it may be essential to include methods uncertainties if integral experiments are to be used for the creation of adjusted nuclear data libraries. A careful evaluation of methods bias and uncertainties must still be performed.
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| Release | : 2004 |
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This research is part of on-going research on the management of uncertainties in simulator predictions for Generation IV systems via optimum experimental design. The focus is on uncertainties originating due to input data. The objective is to devise an algorithmic framework for quantification of uncertainties, identification of their key sources, and ultimately guiding the design of validating experiments for their reduction. An integral part of this research is the development of uncertainty quantification algorithms for models involving many input data and output responses. This represents the focus of the research reported here. Uncertainty Quantification (UQ) in nuclear systems simulation is playing an increasing role in supporting decisions related to the research and development of advanced nuclear energy systems, especially those of interest to the Global Nuclear Energy Partnership (GNEP) and Next Generation Nuclear Plant (NGNP) programs. UQ will help assess the adequacy of existing simulation tools and associated databases, e.g. nuclear cross-section data, and provide guidance to areas of models and/or data where further development and/or measurements should be prioritized. A sensitivity and uncertainty analysis has been conducted to study the effects of neutron microscopic cross-section data uncertainty on macroscopic attributes that influence reactor core design, performance and safety for a Generation IV reactor concept. In the realm of reactor engineering, neutron cross-section data represents the basic physics of neutron interactions with matter and therefore have large impacts on evolution of flux, power, reactivity and other reactor performance attributes. Currently, we focus on uncertainties originating from cross-section data uncertainties, believed to be of primary significance for fast reactor calculations. This thesis presents a recent development of an UQ algorithm for increasing the efficiency of UQ to a level that enables its execution on a r.
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| Release | : 2016 |
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In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE.
| Author | : Jacob Dominic DeWitte |
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| Total Pages | : 179 |
| Release | : 2011 |
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Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a sodium-cooled fast reactor include an energetic core disruptive accident, vessel failure, and a large early release. These consequences can be avoided if unprotected transients are properly defended against, potentially improving the economics of sodium fast reactors. One way to defend against such accidents is to include a highly reliable reactor protection system. The perceived undesirability of the consequences arising from an unprotected transient has led some sodium fast reactor designers to consider incorporating several design modifications to the reactor protection system, including: self-actuated shutdown systems, articulated control rods, and seismic anticipatory scram systems. This study investigates the performance of these systems in sodium fast reactors. To analyze the impact of these proposed design alternatives, a model to analyze plant performance that incorporates uncertainty analysis is developed using RELAP5-3D and the ABR-1000 as the reference design. The performance of the proposed alternatives is analyzed during unprotected loss of flow and unprotected transient overpower scenarios, each exacerbated by a loss of heat sink. The recently developed Technology Neutral Framework is used to contextually rate performance of the proposed alternatives. Ultimately, this thesis offers a methodology for a designer to analyze reactor protection system design efficacy. The principle results of this thesis suggest that when using the Technology Neutral Framework as a licensing framework for a sodium-cooled fast reactor, the two independent scram systems of the ABR- 1000's reactor protection system perform well enough to screen unprotected transients from the design basis. While a regulator may still require consideration of accidents involving the failure of the reactor protection system, these events will not drive the design of the system. However, self-actuated shutdown systems may be called for to diversify the reactor protection system. Of these, the Curie point latch marginally reduces the conditional cladding damage probability for metal cores because of their rapid inherent feedback effects, but is more effective for the more sluggish oxide cores given reasonably long pump coastdown times. Flow levitated absorbers are highly effective at mitigating unprotected loss of flow events for both fuel types, but are limited in response during unprotected transient overpower events. When considered from a risk-informed perspective, a clear rationale and objective is needed to justify the inclusion of an additional feature such as self-actuated shutdown systems. The use of articulated safety rods as one of the diverse means of reactivity insertion and the implementation of an anticipatory seismic scram system may be the most cost-effective alternatives to provide defense in depth in light of the sodium fast reactor's susceptibility to seismic events.
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| Release | : 2008 |
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Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.
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| Release | : 1976 |
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Sensitivity analysis, as applied to both nuclear design and data uncertainty, has developed into a valuable tool for fusion reactor nuclear analysis. Several such studies have been undertaken with the LASL sensitivity system LASS, which includes as its principal modules SENSIT-1D, ONETRAN, and ALVIN. These modules function in a multigroup environment using standard flux and data interface files for communication. The input multigroup cross-section data and uncertainties are obtained primarily from ENDF/B using the NJOY processing system. In particular cases, the input library can be modified by the ALVIN module to improve consistency with available integral experiments. The primary output from LASS is the uncertainty (or change) in important reactor parameters, as calculated in the SENSIT-1D module. Applications of LASS and its component parts have been made to the Tokamak Fusion Test Reactor (TFTR), the Reference Theta-Pinch Reactor (RTPR), and to an Experimental Power Reactor (EPR). This paper emphasizes the initial assessment of cross-section sensitivity for an EPR design. Nucleonic responses examined include neutron and gamma-ray kerma in the toroidal field coils and Mylar superinsulation, displacement damage and transmutation in the copper of the toroidal field coils, and activation of the outboard dewar. These sensitivities are now being used to narrow the range of uncertainty analyses required to quantitatively assess cross-section adequacy for EPR design calculations. Acceptable target uncertainties in nucleonic design parameters are simultaneously being formulated. Experience at LASL with sensitivity and uncertainty analysis techniques incorporated in LASS has provided convincing evidence of their value for fusion reactor studies. Many of these studies are of a shielding nature; e.g., deep penetrations of high-energy neutrons through steel, lead, boron carbide, and graphite, with responses such as activation and kerma.
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| Release | : 1978 |
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The Cross Section Evaluation Working Group (CSEWG) has, in conjunction with the development of the fifth version of ENDF/B, assembled new evaluations for 232Th and 233U. It is the purpose of this paper to describe briefly some of the more important features of these evaluations relative to ENDF/B-4 to project the change in reactor performance based upon the newer evaluated files and sensitivity coefficients for interesting design problems, and to indicate preliminary results from ongoing uncertainty analyses.
