Fast Reactor Technology Research Articles

Plutonium Worlds. Fast Breeders, Systems Analysis and Computer Simulation in the Age of Hypotheticality

This article examines the media history of one of the hallmark civil nuclear energy programs in Western Germany – the development of Liquid Metal Fast Reactor (LMFBR) technology. Promoted as a kind of perpetuum mobile of the Atomic Age, the Manhattan Project not only imported big science thinking. In its context, nuclear technology was also put forth as an avantgarde of scientific inquiry, dealing with the most complex and 'critical' technological endeavors. In the face of the risks of nuclear technology, German physicist Wolf Hafele thus announced a novel epistemology of hypotheticality. In a context where traditional experimental engineering strategies became inappropiate, he called for the application of advanced media technologies: Computer Simulations (CS) and Systems Analysis (SA) generated computerized spaces for the production of knowledge. In the course of the German Fast program, such methods had a twofold impact. One the one hand, Hafele emphazised – as the father of the German Fast Breeder – the utilization of CS for the actual planning and construction of the novel reactor type. On the other, namely as the director of the department of Energy Systems at the International Institute for Applied Systems Analysis (IIASA), Hafele advised SA-based projections of energy consumption. These computerized scenarios provided the rationale for the conception of Fast programs as viable and necessary 'alternative energy sources' in the first place. By focusing on the role of the involved CS techniques, the paper thus investigates the intertwined systems thinking of nuclear facilities’s planning and construction and the design of large-scale energy consumption and production scenarios in the 1970s and 1980s, as well as their conceptual afterlives in our contemporary era of computer simulation.

Fast Neutron–Induced Fission Cross Sections of 233, 234, 236, 238U up to 200 MeV

The 233, 234, 236, 238U fission cross sections have been measured relative to 235U(n, f) for incident neutron energies from 200 KeV to 200 MeV using neutron time-of-flight at the Los Alamos Neutron Science Center. The results are generally consistent with the current ENDF/B-VII evaluation, but some discrepancies with previous measurements above 20 to 30 MeV are observed. These measurements are part of a campaign to measure fission cross sections with high precision in support of fast reactor technology.

Resolution of Fission and Fusion Technology Integration Issues: An Upgraded Design Concept for the Subcritical Advanced Burner Reactor

The conceptual design of the subcritical advanced burner reactor (SABR), a 3000-MW(thermal) annular, modular sodium pool–type fast reactor, fueled by metallic transuranic (TRU) fuel processed from discharged light water reactor fuel and driven by a tokamak D-T fusion neutron source based on ITER physics and technology, has been substantially upgraded. Several issues related to the integration of fission and fusion technologies have been addressed, e.g., refueling a modular sodium pool reactor located within the magnetic coil configuration of a tokamak, achieving long-burn quasi-steady-state plasma operation, access for heating and current drive power transmission to a toroidal plasma surrounded by a sodium pool fast reactor, suppression of magnetohydrodynamic effects in a liquid metal coolant flowing in a magnetic field, tritium self-sufficiency in a TRU transmutation reactor, shielding the superconducting magnets from fusion and fission neutrons, etc. A design concept for a SABR that could be deployed within 25 years, based on the IFR/PRISM metal-fuel, sodium pool fast reactor technology and on the ITER fusion physics and technology, is presented. This design concept can be used for realistic fuel cycle, dynamic safety, and other performance analyses of a SABR.

The SABrR Concept for a Fission-Fusion Hybrid 238U-to-239Pu Fissile Production Reactor

The subcritical advanced burner reactor (SABR) concept, which combines IFR-PRISM fast reactor technology and the ITER tokamak fusion physics and technology in a burner reactor for the transmutation of transuranics, has been adapted for a subcritical advanced breeder reactor (SABrR) that produces plutonium. It is found that basically the same fission and fusion technology, geometry, and major parameters as used in SABR can be used to achieve a significant fissile production rate (fissile breeding ratio ≈ 1.3) while maintaining tritium self-sufficiency (tritium breeding ratio >1.15).

FALSTAFF: A New Tool for Fission Fragment Characterization

The future Neutron For Science (NFS) facility to be installed at SPIRAL2 (Caen, France) will produce high intensity neutron beams from hundreds of keV up to 40 MeV. Taking advantage of this facility, data of particular interest to the nuclear community, in view of the development of fast reactor technology, will be measured. The development of an experimental setup called FALSTAFF for a full characterization of actinide fission fragments has been undertaken. Fission fragment isotopic yields and associated neutron multiplicities will be measured as a function of the neutron energy. Based on time-of-flight and residual energy technique, the setup will allow for the simultaneous measurement of the velocity and energy of the complementary fragments. The performance of the time-of-flight detectors of FALSTAFF will be presented and expected resolutions for fragment masses and neutron multiplicities, based on realistic simulations, will be shown.

Effects of Cr and Ni on interdiffusion and reaction between U and Fe–Cr–Ni alloys

Metallic U-alloy fuel cladded in steel has been examined for high temperature fast reactor technology wherein the fuel cladding chemical interaction is a challenge that requires a fundamental and quantitative understanding. In order to study the fundamental diffusional interactions between U with Fe and the alloying effect of Cr and Ni, solid-to-solid diffusion couples were assembled between pure U and Fe, Fe–15wt.%Cr or Fe–15wt.%Cr–15wt.%Ni alloy, and annealed at high temperature ranging from 580 to 700°C. The microstructures and concentration profiles that developed from the diffusion anneal were examined by scanning electron microscopy, and X-ray energy dispersive spectroscopy (XEDS), respectively. Thick U6Fe and thin UFe2 phases were observed to develop with solubilities: up to 2.5at.% Ni in U6(Fe,Ni), up to 20at.%Cr in U(Fe, Cr)2, and up to 7at.%Cr and 14at.% Ni in U(Fe, Cr, Ni)2. The interdiffusion and reactions in the U vs. Fe and U vs. Fe–Cr–Ni exhibited a similar temperature dependence, while the U vs. Fe–Cr diffusion couples, without the presence of Ni, yielded greater activation energy for the growth of intermetallic phases – lower growth rate at lower temperature but higher growth rate at higher temperature.

Framing Prototypes

This paper considers a crucial moment in the innovation process: the shift from a research phase to an industrial phase. The empirical study examines the development in France of Fast Breeder Reactor technology (FBR), from the 1950s to the early closure of the Superphénix plant in 1997. A turning point occurred in the late 1960s, when several European countries judged that the FBR technology was a promising electricity generation technology that would soon be mature for commercialisation, in a context of technological nationalism and future energy scarcity. In this paper, we analyse how the framing of the resulting prototype as “industrial” entailed an impact on decisions during the three decades that the project lasted. Aiming at describing the project actors in action without judging their decision-making processes, we use the ‘framing’ concept preferably to other approaches such as ‘path dependency’. This concept choice is the subject of the discussion.

The reactor ALLEGRO and the sustainable nuclear energy in Central Europe

The Visegrad-4 countries (CZ, HU, PL and SK) would like to use nuclear energy on the long run. The construction of new Generation 3+ nuclear units probably belong in each country to this realm. These reactors will provide safe and cheap electric energy approximately until the end of the 21 st century. In order to use nuclear energy in the 22 nd century, sustainability of fuel supply shall be achieved by applying Generation 4 breeder reactors with fast spectrum. The corresponding research and development is organized now in the framework of the V4G4 Centre of Excellence establshed by the nuclear research institutes of the region with a strong technical support from the French CEA. The most important milestone of these efforts is the start-up of the ALLEGRO reactor that shall demonstrate the viability of the gas cooled fast reactor technology.

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

Indian fast reactor technology: Current status and future programme

The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

In-situ calibration of permanent magnet flow meters in PFBR using noise analysis technique

In Fast Breeder Reactor sodium circuits, permanent magnet flow meters (PMFM) are extensively used to measure the liquid sodium flow rate. The performance of PMFM can degrade with respect to time due to various reasons. This degradation results in reduction in output voltage and affects flow meter stability. The distortion of the magnetic field in the large diameter flow meters makes its characteristics nonlinear. Further, the performance of the flow meter is also affected by vibrations, shocks, temperature and change of reluctance within the magnetic circuit. Hence, it is desirable to calibrate the PMFM at periodic time intervals for ensuring accuracy and stability. However, it is very difficult and almost impossible to calibrate large size flow meters once installed in the system under actual flow conditions in a sodium test loop. Therefore, it is necessary to calibrate the flow meter in-situ without disturbing its normal operating conditions.Experiments were carried out in different sodium loops in Fast Reactor Technology Group (FRTG) and Fast Breeder Test Reactor (FBTR) with permanent magnet flow meters of different sizes to develop an in-situ calibration procedure. Cross-correlation technique is studied and the flow rate is estimated from the transit time with a deviation of ±5.5%, which is comparable with that of calibration of the flow meter in actual sodium test loop. In this paper, in-situ calibration of PMFM is discussed with experimental details, data acquisition, cross-correlation technique and the results obtained.

Sustainable and safe nuclear fission energy: technology and safety of fast and thermal nuclear reactors

Nuclear reactors operating in the world by 2010.- Reactor physics needed for understanding of fission reactors.- Nuclear fuel supply.- Uranium enrichment.- Fission reactors with thermal neutron spectrum.- Breeder reactors with fast neutron spectrum.

Materials development for fast reactor applications

Materials play a crucial role in the economic competitiveness of electricity produced from fast reactors. It is necessary to increase the fuel burn-up and design life in order to realize this objective. The burnup is largely limited by the void swelling and creep resistance of the fuel cladding and wrapping materials. India's 500MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction. The major structural materials chosen for PFBR with MOX fuel are D9 austenitic stainless steel as fuel clad and wrapper material, 316LN austenitic stainless steel for reactor components and piping and modified 9Cr-1Mo steel for steam generator. In order to improve the burnup, titanium, phosphorous and silicon contents in alloy D9 have been optimized for decreased void swelling and increased creep strength and this has led to the development of a modified version of alloy D9 as IFAC-1. Ferritic steels are inherently resistant to void swelling. The disadvantage is their poor creep strength. Creep resistance of 9Cr-ferritic steel has been improved with the dispersion of nano-size yttria to develop oxide dispersion strengthened (ODS) steel clad tube with long-term creep strength, comparable to alloy D9 so as to achieve higher fuel burnup. Improved versions of 316LN stainless steel with nitrogen content of about 0.14wt% having higher creep strength to increase the life of fast reactors and modified 9Cr-1Mo steel with reduced nitrogen content and controlled addition of boron to improve type IV cracking resistance for steam generator application are other developments being perused towards improving the economic competitiveness of fast reactor technology.

Prompt fission neutron emission: Problems and challenges

This paper presents some of the challenges ahead of us even after 75 years of the discovery of the fission process and large progress made since then. The focus is on application orientation, which requires improved measurements on fission cross-sections and neutron and γ -ray multiplicities. Experimental possibilities have vastly improved the past decade leading to developments of highly sophisticated detector systems and the use of digital data acquisition and signal processing. The development of innovative fast nuclear reactor technology needs improved respective nuclear data. Advancements in theoretical modelling also require better experimental data. Theory has made progress in calculating fission fragment distributions (i.e. GEF code) as well as prompt neutron and γ -ray emission to catch up with the improved experiments.

Fast Reactor Technology for Energy Security: Challenges for Materials Development

Fast Reactor Technology for Energy Security: Challenges for Materials Development

FALSTAFF: A new tool for fission studies

The future NFS installation will produce high intensity neutron beams from hundreds of keV up to 40 MeV. Taking advantage of this facility, data of particular interest for the nuclear community in view of the development of the fast reactor technology will be measured. The development of an experimental setup called FALSTAFF for a full characterization of actinide fission fragments has been undertaken. Fission fragment isotopic yields and associated neutron multiplicities will be measured as a function of the neutron energy. Based on time-of-flight and residual energy technique, the setup will allow the simultaneous measurement of the complementary fragments velocity and energy. The performances of TOF detectors of FALSTAFF will be presented and expected resolutions for fragment masses and neutron multiplicities, based on realistic simulations, will be shown.

Selection of sodium coolant for fast reactors in the US, France and Japan

The joint paper presents a common view of fast reactor specific missions in the development of nuclear energy and a cross-analysis of merits and demerits of several Fast Reactors concepts studied worldwide and especially in the Generation-IV International Forum (GIF) framework.The paper provides the context for fast reactors development in the United States, France and Japan and focuses on the comparison on Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), and Lead-cooled Fast Reactor (LFR), i.e. the three fast reactor concepts that have the potential to meet the nuclear fuel cycle sustainability goals.The information provided in the article permits the reader to understand each country's objectives to see that not only the objectives searched for but also the technical orientations are converging. The authors underline that SFR technology evaluation relies significantly on the substantial base technology development programs within each country which is without comparison for the other two fast reactor technologies, e.g., SFR technology has already been developed to commercial or near commercial scale in each country whereas the performance of LFR and GFR technology is still uncertain.The main GFR merits are the potential for high temperatures and the easier possibilities for inspections and repairs. The main challenges are the fuel (fabrication, in-pile behavior), materials for high temperatures, and the implementation of mitigation means to manage severe core degradation. The main LFR merit is the lack of chemical reactivity of the lead coolant with air and water. The main challenges are the development of corrosion resistant structural and cladding materials, the implementation of mitigation means to manage severe core degradation, the density of the lead, and the comparably large core size.The selection of a reference fast reactor concept in view of possible industrialization is made on a national base, taking into account the each countries’ strategic objectives, existing technology base, the proven or expected technical performance, the R&D challenges and technical means to conduct that R&D, the possibility to share development costs and risks, etc. It is important to note that in different contexts, the U.S., French, and Japanese organizations have selected the SFR as their reference fast reactor concept.

MYRRHA – A multi-purpose fast spectrum research reactor

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental Accelerator-Driven System (ADS) currently under development at SCK⋅CEN and will replace the Material Testing Reactor (MTR) BR2. The MYRRHA facility is currently being developed with the aid of the FP7-project “Central Design Team” and will be as a flexible irradiation facility, able to work in both subcritical and critical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications, and Si-doping.MYRRHA will also demonstrate the full concept of Accelerator Driven Systems by coupling the requisite three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow for the study of efficient transmutation of high-level nuclear waste. Since MYRRHA is based on the heavy liquid metal technology, Lead–Bismuth Eutectic, it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) technology. Further, in critical mode, MYRRHA will play the role of European Technology Pilot Plant in the path forward for LFR.In this paper we present the historical perspectives, international and high profile membership within the consortium of the MYRRHA project and the rationale for the design choices are presented. Also, the latest configuration of the reactor system is described together with the different irradiation capabilities. More specifically, the possibilities and performances for fuel irradiations are presented in detail.

Multi-purpose hYbrid Research Reactor for High-tech Applications a multipurpose fast spectrum research reactor

Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) is the flexible experimental accelerator-driven system in development at SCK•CEN. The MYRRHA facility, currently developed with the aid of the FP7-project ‘Central Design Team’ is conceived as a flexible irradiation facility. MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications, and Si-doping. MYRRHA will also demonstrate the accelerator-driven system full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. Because MYRRHA is based on the heavy liquid metal technology, lead–bismuth eutectic, it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) Technology, and in critical mode, MYRRHA will play the role of the European Technology Pilot Plant in the roadmap for LFR. Copyright © 2011 John Wiley & Sons, Ltd.

The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor

The reprocessing actinide materials extracted from spent fuel for use in mixed oxide fuels is a key component in maximizing the spent fuel repository utility. While fast spectrum reactor technologies are being considered in order to close the fuel cycle, and transmute these actinides, there is potential to utilize existing pressurized heavy water reactors such as the CANDU ® 1 1 CANDU is a trademark of Atomic Energy of Canada Limited. design to meet these goals. The use of current thermal reactors as an intermediary step which can burn actinide based fuels can significantly reduce the fast reactor infrastructure needed. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a typical CANDU nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel lattice was created and the two neutron group properties were transferred to RFSP in order to create a 3 dimensional time average full core model. The model was created with typical CANDU limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU reactors as an intermediary step in burning reprocessed spent fuel and reducing actinide burdens within the end repository. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle using existing and proven reactor technologies.