GEN-IV LFR development: Status & perspectives

Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of Generation IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to Heavy Liquid Metal (HLM) nuclear reactors. In this frame, ENEA developed one of the larger European experimental fleet of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and at developing components, instrumentations and innovative systems, supported by experiments and numerical tools. The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the frame of the liquid metal technologies for GEN-IV LFR. In particular, an overview on the ongoing R&D experimental program will be depicted considering the actual fleet of facilities: CIRCE, NACIE-UP, LIFUS5, LECOR and HELENA. CIRCE (CIRColazione Eutettico) is the largest HLM pool facility presently in operation worldwide. Full scale component tests, thermal stratification studies, operational and accidental transients and integral tests for the nuclear safety and SGTR (Steam Generator Tube Rupture) events in a large pool system can be studied. NACIE-UP (NAtural CIrculation Experiment-UPgraded) is a loop with a HLM primary and pressurized water secondary side and a 250 kW power Fuel Pin Simulator working in natural and mixed convection. LIFUS5 (lithium for fusion) is a separated effect facility devoted to the HLM/Water interaction. HELENA (HEavy Liquid metal Experimental loop for advanced Nuclear applications) is a pure lead loop with a mechanical pump for high flow rates experiments. LECOR (LEad CORrosion) is a corrosion loop facility with oxygen control system installed. All the experiment actually ongoing on these facilities are described in the paper, depicting their role in the context of GEN-IV LFR development

Exergy analysis of a PWR nuclear steam supply system – Part I, general theoretical model

The paper provides an alternative, novel methodology to perform the exergetic analysis of a Pressurized Nuclear Reactor (PWR) based on the strictest definition of fission temperature to get to a careful evaluation of Exergy Destruction and exergetic Efficiency of the component. Up today, the exegetic analyses of Nuclear Power Plants (NPP) have been based on the assumption that Fission Exergy and Fission Energy are almost the same having assumed Carnot Factor almost equal to 1 as Tfiss >>T0. This assumption is based on some simplified hypotheses concerning fission temperature as applied in the definition of the Fission Exergy itself, whose value, to the best knowledge of the authors, was never modeled. On the contrary, in the first part of the paper, the authors present the results of an ongoing research, just aimed at evaluating the Exergy efficiency of the heat exchange in a PWR reactor, whose first results were already presented in [1], based on the most detailed modeling of Tfiss. The modeling, referring to a steady-state operational mode of the Reactor, takes into account all heat transfer phenomena between nuclear fuel UO2, its Zircaloy clad, cooling water, vessel material and the external environment. In the second part of the paper, the Exergy analysis is extended to all main Reactor Cooling System components (Vertical recirculating type Steam Generator, primary coolant pump and piping) with the aim to compare the Exergy Destructions and exergetic Efficiencies of the RPV with those of the other components of the Nuclear Steam Supply System, NSSS. In the Part II of the same paper,, "Exergy Analysis of a PWR Nuclear Steam Supply System - II part: a case study ", a test case is exemplified with the aim to compare the results obtained applying the methodology in question with those obtained applying the most established methodology adopted by other authors

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

The ARC pilot plant conceptual design study has been extended beyond its initial scope [B. N. Sorbom et al., FED 100 (2015) 378] to explore options for managing ~525 MW of fusion power generated in a compact, high field (B_0 = 9.2 T) tokamak that is approximately the size of JET (R_0 = 3.3 m). Taking advantage of ARC's novel design - demountable high temperature superconductor toroidal field (TF) magnets, poloidal magnetic field coils located inside the TF, and vacuum vessel (VV) immersed in molten salt FLiBe blanket - this follow-on study has identified innovative and potentially robust power exhaust management solutions.Comment: Accepted by Fusion Engineering and Desig

Inverse Dynamics and Control for Nuclear Power Plants

A new nonlinear control technique was developed by reformulating one of the “inverse Problems” techniques in mathematics, namely the reconstruction problem. The theory identifies an important concept called inverse dynamics which is always a known property for systems already developed or designed. Accordingly, the paradigm is called “reconstructive inverse dynamics” (RID) control. The standard state-space representation of dynamic systems constitutes a sufficient foundation to derive an algebraic RID control law that provides solutions in one step computation. The existence of an inverse solution is guaranteed for a limited dynamic space. Outside the guaranteed range, existence depends on the nature of the system under consideration. Derivations include adaptive features to minimize the effects of modeling errors and measurement degradation on control performance. A comparative study is included to illustrate the relationship between the RID control and optimal control strategies. A set of performance factors were used to investigate the robustness against various uncertainties and the suitability for digital implementation in large scale-systems. All of the illustrations are based on computer simulations using nonlinear models. The simulation results indicate a significant improvement in robust control strategies. The control strategy can be implemented on-line by exploiting its algebraic design property. Three applications to nuclear reactor systems are presented. The objective is to investigate the merit of the RID control technique to improve nuclear reactor operations and increase plant availability. The first two applications include xenon induced power oscillations and feed water control in conventional light water reactors. The third application consists of an automatic control system design for the startup of the Experimental Breeder Reactor-II (EBR-II). The nonlinear dynamic models used in this analysis were previously validated against available plant data. The simulation results show that the RID technique has the potential to improve reactor control strategies significantly. Some of the observations include accurate xenon control, and rapid feed water maneuvers in pressurized water reactors, and successful automated startup of the EBR-II. The scope of the inverse dynamics approach is extended to incorporate artificial intelligence methods within a systematic strategy design procedure. Since the RID control law includes the dynamics of the system, its implementation may influence plant component and measurement design. The inverse dynamics concept is further studied in conjunction with artificial neural networks and expert systems to develop practical control tools

Subselenean tunneler melting head design: A preliminary study

The placement of base facilities in subsurface tunnels created as a result of subsurface mining is described as an alternative to the establishing of a base on the lunar surface. Placement of the base facilities and operations in subselenean tunnels will allow personnel to live and work free from the problem of radiation and temperature variations. A conceptual design for a tunneling device applicable to such a lunar base application was performed to assess the feasibility of the concept. A tunneler was designed which would melt through the lunar material leaving behind glass lined tunnels for later development. The tunneler uses a nuclear generator which supplies the energy to thermally melt the regolith about the cone shaped head. Melted regolith is exacavated through intakes in the head and transferred to a truck which hauls it to the surface. The tunnel walls are solidified to provide support lining by using an active cooling system about the mid section of the tunneler. Also addressed is the rationale for a subselenean tunneler and the tunneler configuration and subsystems, as well as the reasoning behind the resulting design

SP-100 nuclear space power systems with application to space commercialization

The purpose of this paper is to familiarize the Space Commercialization Community with the status and characteristics of the SP-100 space nuclear power system. The program is a joint undertaking by the Department of Defense, the Department of Energy and NASA. The goal of the program is to develop, validate, and demonstrate the technology for space nuclear power systems in the range of 10 to 1000 kWe electric for use in the future civilian and military space missions. Also discussed are mission applications which are enhanced and/or enabled by SP-100 technology and how this technology compares to that of more familiar solar power systems. The mission applications include earth orbiting platforms and lunar/Mars surface power

Design innovation for the 1990's

Statement of responsibility on title-page reads: Richard K. Lester, Michael J. Driscoll, Michael W. Golay, David D. Lanning, Lawrence M. Lidsky, Norman C. Rasmussen and Neil E. Todreas"September 1983."Includes bibliographical reference

Control rod monitoring of advanced gas-cooled reactors

The UK’s fleet of Advanced Gas Cooled Reactors (AGR) are approaching, and have in some cases exceeded, their original design lives. Continued operation is under enhanced safety cases based on monitoring, inspection and component condition assessment of the core and related systems. This paper presents an analysis of the regulating control rods of an AGR, which are used to manage the power and reactivity of the core. Current manual analyses attempt to detect possible restrictions in the motion of the rods due to degradation of the graphite core, however the development of an automated intelligent analysis of the control rod data provides a repeatable and auditable method of analyzing the data. It is shown, by means of an example data set, that despite some limitations in the scope of the recorded data, it is possible to estimate the performance of the rods and present this information to the engineer in a way that more easily indicates abnormal behavior than existing analyses. It is also noted that though this work was initially conceived as a method of detecting restrictions in the motion of the regulating control rods, the results are potentially more useful is characterizing control rod performance and has potential application in predictive maintenance

DEVELOPMENT OF INSTRUMENTATION AND CONTROL SYSTEMS FOR AN INTEGRAL LARGE SCALE PRESSURIZED WATER REACTOR

Small and large scale integral light water reactors are being developed to supply electrical power and to meet the needs of process heat, primarily for water desalination. This dissertation research focuses on the instrumentation and control of a large integral inherently safe light water reactor (designated as I2S-LWR) which is being designed as part of a grant by the U.S. Department of Energy Integrated Research Project (IRP). This 969 MWe integral pressurized water reactor (PWR) incorporates as many passive safety features as possible while maintaining competitive costs with current light water reactors. In support of this work, the University of Tennessee has been engaged in research to solve the instrumentation and control challenges posed by such a reactor design. This dissertation is a contribution to this effort. The objectives of this dissertation are to establish the feasibility and conceptual development of instrumentation strategies and control approaches for the I2S-LWR, with consideration to the state of the art of the field. The objectives of this work are accomplished by the completion of the following tasks: Assessment of instrumentation needs and technology gaps associated with the instrumentation of the I2S-LWR for process monitoring and control purposes. Development of dynamic models of a large integral PWR core, micro-channel heat exchangers (MCHX) that are contained within the reactor pressure vessel, and steam flashing drums located external to the containment building. Development and demonstration of control strategies for reactor power regulation, steam flashing drum pressure regulation, and flashing drum water level regulation for steady state and load-following conditions. Simulation, detection, and diagnosis of process anomalies in the I2S-LWR model. This dissertation is innovative and significant in that it reports the first instrumentation and control study of nuclear steam supply by integral pressurized water reactor coupled to an isenthalpic expansion vessel for steam generation. Further, this dissertation addresses the instrumentation and control challenges associated with integral reactors, as well as improvements to inherent safety possible in the instrumentation and control design of integral reactors. The results of analysis and simulation demonstrate the successful development of dynamic modeling, control strategies, and instrumentation for a large integral PWR