Low-order dynamic modeling of the Experimental Breeder Reactor II

Berkan, R.C.; Upadhyaya, B.R.; Kisner, R.A.

doi:10.2172/6585648

Cited by 3 publications

(16 citation statements)

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“…The primary loop includes the active core, inner and outer blankets, lower and upper reflectors, and piping. The formulation of the whole model and all the parameters are adapted using the models in [4]. The primary loop and IHX models are coupled into one module for the convenience of simulation studies.…”

Section: Erb-ii-core and The Intermediate Heat Exchanger (Ihx) Modelingmentioning

confidence: 99%

“…Many important works were developed to control SFRs with different techniques. [4] performed a linear quadratic (LQR) compensator to control the fractional power, fuel temperature, reactor inlet and outlet temperatures upon -5 cent reactivity insertion. Moreover, [3] designed a traditional PI control and an optimal model predictive control (MPC) as related to core inlet temperature and temperature change across the core under a variety of reactor power levels (30 − 100% power) with external reactivity perturbations of -1 cent.…”

Section: Introductionmentioning

confidence: 99%

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A New Proposal Generalized Predictive Control Algorithm With Polynomial Reference Tracking Applied for Sodium Fast Reactors

et al. 2023

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This paper proposes a generalized predictive control (GPC) with constraints and orthonormal Laguerre functions using the simplified model of the primary system (reactor core and intermediate heat exchanger (IHX)) of a prototypical sodium fast reactor (SFR). This paper develops a multiple-input multiple-output (MIMO) GPC with input constraints able to track polynomial references of any degree applied in coolant temperature difference across the core and fractional power. The manipulated variables of the GPC-SFR are the reactivity and the sodium flow rate of the primary and secondary pipes. Moreover, orthonormal Laguerre functions and step down condition number techniques were also applied to avoid the numerical ill-conditioning issue in quadratic programming of large systems. Thus, a GPC type-2 was designed to control fractional power, coolant temperature difference across the core and sodium tank temperature of the SFR primary system when temperature references change according to a linear ramp after reaching their steady-state operation, sustaining 100% power operation on the reactor. In order to analyze the load tracking capability of the GPC-SFR type-2, the load following from 100% fractional power (FP) to 60% FP at 0.8% FP/min rate is simulated. Constraints on the rate of coolant temperature difference across the core and reactivity were applied for the design safety. For comparison criteria, this paper compares the GPC-SFR type-2 with the GPC-SFR type-1, i.e, standard model predictive control (MPC), to verify the viability and superior performance of the proposal regarding: (a) ramp-tracking capability of temperature and load; (b) the rejections of a reactivity disturbance of -1 cent and a secondary sodium inlet temperature disturbance of +10 o F ; and (c) a simulation with uncertainty in reactor design. The simulations show that the GPC-SFR type-2 overcome the GPC-SFR type-1 robustness and performance.

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Section: Erb-ii-core and The Intermediate Heat Exchanger (Ihx) Modelingmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

A New Proposal Generalized Predictive Control Algorithm With Polynomial Reference Tracking Applied for Sodium Fast Reactors

et al. 2023

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“…The governing equations for each subsystem and definition of variables are presented in the following subsections. The values for model parameters are given in Berkan and Upadhyaya (1988) and Berkan et al (1990).…”

Section: Description Of Ebr-ii Primary Systemmentioning

confidence: 99%

“…The superheater model considers a single-phase heat transfer regime. Dry steam is heated by the primary sodium to 875°F at full power (Berkan et al 1990). The superheater is modeled as a five-node counterflow single-phase heat exchanger, using the same equations as the IHX in the primary system model.…”

Section: Superheater State Equationsmentioning

confidence: 99%

Prototypic Enhanced Risk Monitor Framework and Evaluation - Advanced Reactor Technology Milestone: M3AT-15PN2301054

Ramuhalli¹,

Hirt²,

Veeramany³

et al. 2015

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Enabling real-time decisions on stress-relief for risk-significant equipment susceptible to degradation and damage, thereby supporting optimized lifetime management. As described in previous reports in this series, probabilistic risk assessment (PRA) provides a static representation of risk associated with operation and maintenance (O&M) of nuclear power plants. Technologies for characterizing real-time risk (so called Enhanced Risk Monitors or ERMs) take into account plant-specific normal, abnormal, and deteriorating states of systems, structures, and components (SSCs) in the estimation of current and future risk to safe and economic operation. Additionally, technologies for characterizing real-time risk provide a mechanism for compensating for the relatively small amount of long-term reliability data from AR systems, structures, and components. The ability to monitor performance and characterize changes in operational risk in real-time can reduce the level of dependence on such performance data. Proactively establishing a viable ERM methodology before AR component design specifications are established also supports: (i) building in opportunities for automated monitoring (on-line and off-line) of those components for optimizing performance with respect to anticipated demands on these reactors; and (ii) improving the maintainability of components from the perspective of time-to-repair and component cost. This research report summaries the development and evaluation of a prototypic ERM methodology (framework) that includes alternative risk metrics and uncertainty analysis. This updated ERM methodology accounts for uncertainty in the equipment condition assessment (ECA), the prognostic result, and the PRA model. It is anticipated that the ability to characterize uncertainty in the estimated risk and update the risk estimates in real-time based on ECA will provide a mechanism for optimizing plant performance while staying within specified safety margins. The report provides an overview of the methodology for integrating time-dependent failure probabilities into risk monitors. This prototypic ERM methodology was evaluated using a hypothetical PRA model, generated using a simplified design of a liquid-metal-cooled AR. Component failure data from industry compilation of failures of components similar to those in the simplified AR model were used to initialize the PRA model. By using time-dependent probability of failure (POF) that grows from the initial probability when equipment is in like-new condition to a maximum POF, which occurs before a scheduled maintenance action that restores or repairs the component to "as-new" condition, the changes in core damage frequency (CDF) over time were computed and analyzed. vii Acronyms and Abbreviations AC alternating current AdvSMR advanced small modular reactor AFI aging fractional increase AST aging start time CAFTA Computer Aided Fault Tree Analysis (system) CCF common cause failure CDF core damage frequency CREDO Centralized Reliability Data Organization (component reliabili...

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“…(1987), and Berkan et al (1990). EBR-II contained primary and secondary coolant systems that utilized liquid sodium as the working fluid and a conventional steam system as illustrated in Figures 1 and 2.…”

Section: Ebr-ii Descriptionmentioning

confidence: 99%

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

Davis¹

2006

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SUMMARYThe Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of the ABTR. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR.An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodiumcooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code's calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to better simulate the ABTR. The accuracy of the sodium properties used by the code was also evaluated.The applicability evaluation generally showed that the existing models were adequate or that relatively minor changes were required to improve the code's representation of the important phenomena. However, new models are required to represent electromagnetic pumps and axial conduction in the fluid. Long-term improvements will also be needed to better represent the decay heat associated with the actinides burned by the reactor, the effects of thermal expansion of various core components on reactivity feedback, and the Reynolds-number dependence on the form loss coefficient for orifices. The evaluation of fluid properties indicated that the specific heat capacity of the liquid phase should be changed to better represent the values expected during normal operation and during transients.

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Low-order dynamic modeling of the Experimental Breeder Reactor II

Cited by 3 publications

References 0 publications

A New Proposal Generalized Predictive Control Algorithm With Polynomial Reference Tracking Applied for Sodium Fast Reactors

A New Proposal Generalized Predictive Control Algorithm With Polynomial Reference Tracking Applied for Sodium Fast Reactors

Prototypic Enhanced Risk Monitor Framework and Evaluation - Advanced Reactor Technology Milestone: M3AT-15PN2301054

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

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