- The Gas-Cooled Fast Reactor (GFR) system features a fast-neutron-spectrum, helium-cooled reactor and closed fuel cycle.
- The Lead-Cooled Fast Reactor (LFR) system features a fast-spectrum lead or lead/bismuth eutectic liquid-metal-cooled reactor and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides.
- The Molten Salt Reactor (MSR) system produces fission power from a molten salt fuel circulating in a fast or epithermal-spectrum reactor and contains an integrated fuel cycle.
- The Sodium-Cooled Fast Reactor (SFR) system features a fast-spectrum, sodium-cooled reactor and a closed fuel cycle for efficient management of actinides and conversion of fertile uranium.
- The Supercritical-Water-Cooled Reactor (SCWR) system is a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water (374 degrees Celsius, 22.1 MPa, or 705 degrees Fahrenheit, 3208 psia).
- The Very-High-Temperature Reactor (VHTR) is a graphite-moderated, helium-cooled reactor with a thermal neutron spectrum.
The Gas-Cooled Fast Reactor (GFR) system features a fast-neutron-spectrum, helium-cooled reactor and closed fuel cycle.
The high outlet temperature of the helium coolant used in the GFR system makes it possible to deliver electricity, hydrogen, or process heat with high efficiency. The reference reactor is a 1200-MWe helium-cooled system operating with an outlet temperature of 850 degrees Celsius using a direct Brayton cycle gas turbine for high thermal efficiency.
Several fuel forms are candidates that hold the potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic-clad elements of actinide compounds. Core configurations may be based on prismatic blocks, pin- or plate-based assemblies. The GFR reference has an integrated, on-site spent fuel treatment and refabrication plant.
The GFR uses a direct-cycle helium turbine for electricity generation, or can optionally use its process heat for thermochemical production of hydrogen. Through the combination of a fast spectrum and full recycle of actinides, the GFR minimizes the production of long-lived radioactive waste. The GFR's fast spectrum also makes it possible to use available fissile and fertile materials (including depleted uranium) considerably more efficiently than thermal spectrum gas reactors with once-through fuel cycles.
This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.
Diagram source: http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf
Advantages and challenges
It is anticipated that GFR systems will minimise the production of long-lived radioactive waste and make it possible to utilize fissile and fertile materials (including depleted uranium) two orders of magnitude more efficiently than thermal spectrum systems. The key challenges associated with this system concern the development of new fuels and materials capable of operating at temperatures of 850°C, the core design and helium turbine. The innovative GFR technologies and design features are intended to overcome the consequences of using a high-pressure gas with poor thermal characteristics to cool down a core with a low thermal inertia during depressurization events.
GIF progress in 2007
Negotiations to put in place GFR research projects in the integration, design and safety of GFR systems, the fast neutron fuel, core materials and fuel cycle processes specific to the GFR system, advanced during 2007 with the finalization of a System Research Plan. The aim is to have an experimental technology demonstration reactor (which will be used as an R&D tool) in place by 2020.
Recent GFR research papers
Garnier, J.C. et al (2005) GFR System: Progress of CEA pre-conceptual design studies, Paper 5305, 2005 International Congress on Advances in Nuclear Power Plants (ICAPP'05).
Mizuno, T., Okano, Y., Aida, T. (2005) Conceptual Core Design Studies of Helium Cooled Fast Reactor with Coated Particle Fuel, Paper 5197, 2005 International Congress on Advances in Nuclear Power Plants (ICAPP'05).
Rouault, J. et al (2003), Fuel Design, Management and Cycles for Generation IV GFRs, Topical Meeting on Advances in Nuclear Fuel Management III, Hilton Head Island, SC, USA, 5-8 October 2003.
Dostal, V., Hejzlar, P., Driscoll, M.J., Todreas, N.E. (2002) A Supercritical CO2 Gas Turbine Power Cycle for Next Generation Nuclear Reactors, 10th International Nuclear Conference on Engineering (ICONE 10), Arlington, Virginia, USA, 14-18 April 2002.
Melese-d'Hospital, G. and Simon, R.H. (General Atomic Company), Status of Gas-Cooled Fast Breeder Reactor Programs, Nuclear Engineering and Design 40 (1977) 5-12.
Related links
DOE Nuclear Energy Research Initiative GFR Program Plan (pdf, 252 kb).
E-mail contact: gfr@gen-4.org
----------------
The Lead-Cooled Fast Reactor (LFR) system features a fast-spectrum lead or lead/bismuth eutectic liquid-metal-cooled reactor and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides.
The LFR system has excellent materials management capabilities since it operates in the fast-neutron spectrum and uses a closed fuel cycle for efficient conversion of fertile uranium. It can also be used as a burner to consume actinides from spent LWR fuel and as a burner/breeder with thorium matrices. An important feature of the LFR is the enhanced safety that results from the choice of molten lead as a relatively inert coolant. In terms of sustainability, lead is abundant and hence available, even in case of deployment of a large number of reactors. More importantly, as with other fast systems, fuel sustainability is greatly enhanced by the conversion capabilities of the LFR fuel cycle.
The LFR was primarily envisioned for missions in electricity and hydrogen production, and actinide management. Given its R&D needs in the areas of fuels, materials, and corrosion control, a two step process leading to industrial deployment of the LFR system has been envisioned: by 2025 for reactors operating with relatively low primary coolant temperature and low power density; and by 2035 for more advanced designs. The preliminary evaluation of the LFR concepts considered by the LFR Provisional System Steering Committee (PSSC) covers their performance in the areas of sustainability, economics, safety and reliability and proliferation resistance and physical protection.
The LFR concepts that are currently being designed are two pool-type reactors (Table1):
- the Small Secure Transportable Autonomous Reactor (SSTAR), developed in the USA and
- the European Lead-cooled System (ELSY), developed by the EC.
Table 1. Key Design data of GIF LFR concepts
Parameters
|
SSTAR
|
ELSY
|
Power [MWe]
|
19.8
|
600
|
Conversion Ratio
|
~1
|
~1
|
Thermal efficiency [%]
|
44
|
42
|
Primary coolant
|
Lead
|
Lead
|
Primary coolant circulation
(at power) |
Natural
|
Forced
|
Primary coolant circulation
for direct heat removal (DHR) |
Natural
|
Natural
|
Core inlet temperature [°C]
|
420
|
400
|
Core outlet temperature [°C]
|
567
|
480
|
Fuel
|
Nitrides
|
MOX, (Nitrides)
|
Fuel cladding material
|
Si-Enhanced Ferretic/Martensitic
Stainless Steel |
T91 (aluminized)
|
Peak cladding temperature [°C]
|
650
|
550
|
Fuel pin diameter [mm]
|
25
|
10.5
|
| Active core dimensions Heigh/equivalent diameter [m] |
0.976/1.22
|
0.9/4.32
|
Primary pumps
|
-
|
N°8, mechanical,
integrated in the SG |
Working fluid
|
Supercritical CO²
at 20 MPa, 552°C |
Water-superheated steam at 18 MPa, 450°C
|
Primary/secondary heat transfer system
|
Four Pb-to-CO² HXs
|
Eight Pb-to-H2O SGs
|
Direct heat removal (DHR)
|
Reactor Vessel Air Cooling System
+ Multiple Direct Reactor Cooling Systems |
Reactor Vessel Air
Cooling System + Four Direct Reactor Cooling Systems + Four Secondary Loops Cooling Systems |
It should be noted that the objective of designing LFR with the high mean core outlet coolant temperatures required for the generation of hydrogen by thermo-chemical processes, could not been addressed simultaneously with the two-track design approach of the systems indicated above, owing to the required longer term R&D necessary for the development of new high-temperature materials that will be needed to provide corrosion resistance with lead as the coolant; this objective will be addressed at a later stage, depending on the success of the nearer term technology demonstration stage, that has been given priority.
The SSTAR is a small factory-built turnkey plant operating on a closed fuel cycle with very long refuelling interval (15 to 20 years or more) cassette core or replaceable reactor module. The current reference design for the SSTAR in the United States is a 20 MWe natural circulation reactor concept with a small shippable reactor vessel (Figure 1). Specific features of the lead coolant, the nitride fuel containing transuranic elements, the fast spectrum core, and the small size combine to promote a unique approach to achieve proliferation resistance, while also enabling fissile self-sufficiency, autonomous load following, simplicity of operation, reliability, transportability, as well as a high degree of passive safety. Conversion of the core thermal power into electricity at a high plant efficiency of 44 % is accomplished utilizing a supercritical carbon dioxide Brayton cycle power converter.
Figure 1. - Conceptual 20 MWe (45 MWt) SSTAR system
The initial design of ELSY is almost complete. The next step in its development is the R&D testing of several design innovations, in order to start with confidence, the detailed engineering design of a reduced-scale demonstration facility.
The ELSY reactor (Figure 2) is rated at 600 MWe. This mid-size rating is the result of the fact that plants of the order of several hundreds MWe are most economically attractive for addition to the European interconnected grids. In addition, a larger plant would require an increase mass of the lead coolant and would entail increased mechanical loads on the reactor vessel and its supporting structure.
The choice of a mid-size reactor power suggested the use of forced circulation to shorten the reactor vessel thereby avoiding excessive coolant mass and alleviating mechanical loads on the reactor vessel.
Thanks to the favorable neutron characteristics of lead, the fuel rods have been spaced further apart than in the case of previous fast-neutron cores. This and the innovative steam generators with flat spirals tube bundle enable the design of a low pressure loss primary loop. The needed pump head, in spite of the higher density of lead, could, therefore, be kept low (on the order of two bars) with reduced requirement of pumping power.
Because of the predicted low primary system pressure loss and the favorable heat transfer properties of lead, decay heat can be removed by natural circulation in the case of loss of station service power (LOSSP).
The proposed thermal cycle of 400 °C at core inlet and only 480 °C at core outlet enable key advantages in terms of use of currently-available structural steels, reduced corrosion and reduced creep thereof, and milder thermal transients.
In terms of efficiency of electricity energy generation, the designers have achieved almost the same thermal efficiency as the Na-cooled SPX1, in spite of the 62 K lower mean core outlet temperature.
The potential of specifying higher operating temperatures of the primary cycle, owing to the low vapour pressure and very high boiling point of lead, depends on the qualification of suitable structural materials, the development of which may prove a long-term task, and has not been included in the near-term development of the LFR. However, the potential for future high temperature operations remains an attractive feature of the LFR.
Priority has been given to the designer’s goal of demonstration of the technical feasibility of the LFR within a relatively short time frame, with features such as a MOX-fuel core self-sustaining because of a conversion ration of about 1 and being adiabatic to (i.e. burner of) the self-generated MA. Development of the LFR to the more ambitious goals of high temperature operation and burning capability of MA beyond the self-generated MA will be considered, but will be pursued in detail in a future stage, depending on R&D and design achievements, and budget.
Figure 2 - ELSY reference configuration. Status at the end of 2008.
Advantages and challenges
The main advantages of the LFR system are its expected fuel efficiency, its capabilities in terms of nuclear materials management (thereby mitigating proliferation risks) and the reduced production of high-level radioactive waste and actinides.
The main features that the members have identified in order to achieve the Generation IV goals are summarized in Table 2. These features are based either on the inherent features of lead as a coolant or on the specific engineered designs.
Table 2 LFR potential performance against the four Goal Areas and the eight Goals for Generation IV.
Generation IV Goal Areas
|
Goals for Generation IV Nuclear Energy Systems
|
Goals achievable via
| |
Inherent features of Lead
|
Specific engineered solutions
| ||
Sustainability
| Resource utilization |
|
|
| Waste minimization and management |
| ||
Economics
| Life cycle cost |
|
|
| Risk to capital (Investment protection) |
| ||
Safety
and Reliability | Operation will excel in safety and reliability |
Lead as:
|
|
| Low likelihood and degree of core damage |
Lead as:
|
| |
| No need for offsite emergency response |
| ||
Profiferation Resistance
and Physical Protection | Unattractive route for diversion of weapon-usable material |
|
|
| Increased physical protection against acts of terrorism |
|
| |
Overview of key challenges for the LFR is provided in table 3.
Table 3. Key challenges of the LFR design.
General issue
|
Specific issue
|
Proposed strategy
|
Corrosion in Lead
|
Tendency for material corrosion with increasing temperature
|
Mean core outlet temperature for the large plant is limited to 480°C¹
Dissolved oxygen provides barrier against corrosion |
Reactor vessel
|
Temperature limited by design to 400°C
| |
Fuel cladding
|
Use of aluminized surface treatment of steels
| |
Reactor internals
|
Dissolved oxygen control
| |
SG tubes
|
Use of aluminized steels to avoid lead pollution and heat transfer degradation
| |
Pump impeller degradation²
|
Use of innovative materials
| |
Seismic design
|
Challenge related to the large mass of lead
|
Use of 2D seismic isolators + short vessel design
|
SGs are installed inside the reactor vessel
with risk of water ingress in lead in case of SGTR accident |
Rupture of the SG collectors in lead
|
Eliminated by design
|
Steam entrainment in the core in case of SGTR
|
Excluded by design
| |
Pressure waves inside the primary system in case of SGTR
|
Harmless by specific design features
| |
DHR
|
Diversification, reliability and passive operation required³
|
Diversification and reliability by means of use of both atmospheric air and stored water
|
Refuelling in lead
|
High temperature makes refuelling difficult⁴
|
Access to fuel assemblies is in cold cover gas
|
¹ The small system operates at a higher temperature but because of the use of natural circulation cooling the erosive effect of lead is reduced
² Pump impeller problem is not characteristic to small system because of the use of natural circulation cooling
³ In case of small system a simple and reliable RV air cooling system is sufficient to remove decay heat
⁴Small system feature sealed core without refueling or complete replacement as a cassette
² Pump impeller problem is not characteristic to small system because of the use of natural circulation cooling
³ In case of small system a simple and reliable RV air cooling system is sufficient to remove decay heat
⁴Small system feature sealed core without refueling or complete replacement as a cassette
Most challenges have been positively addressed by the conceptual ELSY design configuration as of the end of 2008, but the challenge remains of the follow-on design of a very high temperature reactor, operating beyond 550°C, the design of which has not yet been addressed, mainly because of outstanding information about corrosion resistant, high-temperature materials.
GIF progress in 2008
The LFR R&D development plan incorporates two tracks of development leading to a single joint demonstration facility by 2020. Separate designs for a small, transportable LFR with a long core life and a moderate-sized power plant will be researched in the demonstration facility. The LFR system research plan, which sets out the research required in the system design, fuel and lead technology and materials, was updated in the course of 2008.
Recent LFR research papers and links
Cinotti L., et al., “The ELSY Project”, Paper 377, Proceeding of the International Conference on the Physics of Reactors (PHYSOR), Interlaken, Switzerland, 14-19 September, 2008.
L. Cinotti et al, The Potential of the LFR and the ELSY Project, 2007 International Congress on Advances in Nuclear Power Plants (ICAPP '07).
Y. H. Yu, H. M. Son, I. S. Lee, K. Y. Suh, Optimized Battery-Type Reactor Primary System Design Utilizing Lead, Paper 6148, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).
I.S. Hwang, A Sustainable Regional Waste Transmutation System: P E A C E R, Plenary Invited Paper, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).
W. J. Kim, T. W. Kim, M. S. Sohn, K. Y. Suh, Supercritical Carbon Dioxide Brayton Power Conversion Cycle Design for Optimized Battery-Type Integral Reactor System, Paper 6142, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).
A. V. Zrodnikov, G. I. Toshinsky, O. G. Komlev, Yu. G. Dragunov, V. S. Stepanov, N. N. Klimov, I. I. Kpytov, and V. N. Krushelnitsky, Use of Multi-Purpose Modular Fast Reactors SvBR-75/100 in Market Conditions, Paper 6023, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).
L. Cinotti, C. Fazio, J. Knebel, S. Monti, H. Ait Abderrahim, C. Smith, K. Suh, LFR (2006)
“LFR ‘Lead-Cooled Fast Reactor’", Proceedings of FISA 2006, EU Research and Training in Reactor Systems, Luxembourg, 13-16 March 2006
E-mail contact: lfr@gen-4.org
-------------
The Molten Salt Reactor (MSR) system produces fission power from a molten salt fuel circulating in a fast or epithermal-spectrum reactor and contains an integrated fuel cycle.
In a Molten Salt Reactor (MSR), the fuel is dissolved in a fluoride salt coolant. Prior MSRs were mainly considered as thermal-neutron-spectrum graphite-moderated concepts. Since 2005 R&D has focused on the development of fast-spectrum MSR concepts (MSFR) combining the generic assets of fast neutron reactors (extended resource utilization, waste minimization) to those relating to molten salt fluorides as fluid fuel and coolant (favourable thermal-hydraulic properties, high boiling temperature, optical transparency). In addition, MSFRs exhibit large negative temperature and void reactivity coefficients, a unique safety characteristic not found in solid-fuel fast reactors (Mathieu et al, 2009). MSFR systems have been recognized as a long term alternative to solid-fuelled fast neutron systems with unique potential (negative feedback coefficients, smaller fissile inventory, easy in-service inspection, simplified fuel cycle, etc.).
Apart from MSR systems, other advanced reactor concepts are being studied employing liquid salt technology as primary coolant in Fluoride-cooled High-temperature Reactor (FHR), or intermediate coolant as an alternative to secondary sodium in Sodium Fast Reactors (SFR) and to intermediate helium in Very High Temperature Reactors (VHTR).
More generally speaking, the development of higher temperature salts as coolants could bring new nuclear and non-nuclear applications. These salts could facilitate heat transfer for nuclear hydrogen production concepts, concentrated solar electricity generation, oil refineries and shale oil processing facilities amongst other applications (Forsberg et al., 2007).
Fluoride-cooled High-temperature Reactors (FHRs) combine the use of liquid fluoride salt coolants (like MSRs), pool type cores and vessel configurations in common with many sodium reactor designs, and coated particle fuels similar to high temperature gas-cooled reactors (Forsberg et al., 2008). The two most developed FHR designs are the 1200 MWe Advanced High Temperature Reactor (AHTR) that employs prismatic fuel elements and the 410MWe Pebble Bed Advanced High Temperature Reactor (PB-AHTR). The better fluoride salt heat transport characteristics, as compared to helium, enable power densities 4 to 8 times greater as well as power levels over 4000 MWt with passive safety systems. Fuel cycle characteristics are essentially identical to those of the VHTR, while intermediate heat transport, power conversion and balance of plant are essentially identical to those of the “reference” MSR.
This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.
Diagram source: http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf
Advantages and challenges
The renewal and diversification of interests in molten salts led the MSR PSSC in 2008 to shift the R&D aims and objectives promoted in the original Generation IV Roadmap, issued in 2002, in order to include in a consistent body the different applications then envisioned for fuel and coolant salts.
Since then, two baseline concepts are considered which have large commonalities in basic R&D areas, particularly for liquid salt technology and materials behaviour (mechanical integrity, corrosion). These are:
- The MSFR system operated in the thorium fuel cycle. Although its potential has been assessed, specific technological challenges remain and the safety approach has to be established.
- The FHR system, a high temperature reactor with better compactness than the VHTR and passive safety potential for medium to very high unit power (> 2400 MWt).
In addition, opportunities offered by liquid salts for intermediate heat transport in other systems (SFR, LFR, VHTR) are investigated. Liquid salts offer two potential advantages: smaller equipment size, because of the higher volumetric heat capacity of the salts, and the absence of chemical exothermal reactions between the reactor, intermediate loop and power cycle coolants.
Liquid salt chemistry plays a major role in the viability demonstration, with such essential R&D issues as: the physico-chemical behaviour of coolant and fuel salts, including fission products and tritium; the compatibility of salts with structural materials for fuel and coolant circuits, as well as fuel processing material development; the on-site fuel processing; the maintenance, instrumentation and control of liquid salt chemistry (redox, purification, homogeneity), and; safety aspects, including interaction of liquid salts with various elements.
GIF progress in 2009
Significant progress was achieved in 2009. This included:
- Development of MSFR pre-conceptual designs and performance analysis of MSFR potential for starting with plutonium and minor actinides from PWR wastes (France).
- Laboratory scale processing of Ni-W-Cr alloys was recently demonstrated. The alloys were found to have acceptable workability and very good high temperature hardness (France, Auger et al., 2009). The whole potentialities of these kinds of materials as well as Hastelloy N have yet to be tested and characterized over the full range of temperatures and in presence of the fluoride salts.
- Corrosion tests of Ni-based alloys, (France, Fabre et al. 2009) and (ISTC, Ignatiev et al., 2008a).
- Better understanding of the PuF3 solubility in various carrier salts by means of thermochemical modeling (Euratom, Beneš et al., 2009).
- The material property database for molten and liquid salts was extended through experiments (Euratom) and theoretical calculations (Euratom, France). New experimental facilities were and continue to be developed (JRC-ITU).
- Significant improvement of fuel salt clean-up scheme (France).
- A code package for a fast MSR was developed (Hoogmoed, 2009) by coupling the 3-D time-dependent diffusion code DALTON with the thermo-hydraulics code HEAT (The Netherlands, TU-Delft).
- The optimal core configuration and salt composition of a moderated MSR that maximize the power density while keeping the self-breeding capabilities were found (The Netherlands, TU-Delft). New breeding gain definitions were developed (Nagy et al., 2010) that account for the unique behavior of the reactor. Some preliminary studies on the salt composition were published in (Nagy et al., 2008).
- Better understanding of the transmutation capabilities, dynamics and safety-related parameters, for fertile and fertile-free fuel concepts (IAEA, Ignatiev et al., 2008b).
- Demonstration of FHR performance and safety (USA).
- Construction of a fluoride salt test loop was initiated in the USA.
- An FHR component test plan was completed in the USA (Holcomb et al, 2009). The test plan provides a roadmap to the major technical demonstrations required to enable a test scale FHR to be built.
- Construction of a surrogate material compact integral effect test apparatus in support of a test scale FHR was initiated (USA). The new apparatus is intended to demonstrate the coupled thermal hydraulics response of FHRs to transients including loss of heat sink and loss of forced circulation.
- Criticality tests for the assessment of FHR fuel and core behavior (USA, Czech Republic).
A general discussion on these topics can be found in the Generation IV International Forum 2009 Annual Report (pages 52-59). More detailed explanations can be found in the bibliography below.
Recent MSR research papers
Auger T., Cury R., Chevalier J.P. (2009), Development of Ni-W-Cr alloys for Gen IV Nuclear Reactor Applications, TMS annual meeting, 15-19 February 2009, San-Francisco, USA.
Beneš O., et al., (2008), Review Report on Liquid Salts for Various Applications, Deliverable D50, Assessment of Liquid Salts for Innovative Applications, ALISIA project, of the 7th Euratom Framework Programme.
Beneš O., Konings R.J.M., Actinide Burner Fuel: Potential compositions based on the thermodynamic evaluation of the MFX-PuF3 (M=Li, Na, K, Rb, Cs, La) system. J. Nucl. Mater. 377 (2008) 449.
Beneš O., Konings R.J.M., Thermodynamic evaluation of the LiF-NaF-BeF2-PuF3 system, J. Chem. Thermodyn., 41 (2009) 1086-1095.
Delpech S., et al., (2008a), Optimization of fuel reprocessing scheme for innovative molten salt reactor, paper presented at the October 2008 Molten Salts Joint Symposium, Kobe, Japan.
Delpech S., et al., (2008b), Actinides/Lanthanides Separation for the Thorium Molten Salt Reactor Fuel Treatment, paper presented at ATALANTE 2008, Montpellier, France.
Delpech S., et al., (2009a), Reactor Physics and Processing Scheme for Innovative Molten Salt Reactor System, J. of Fluorine Chemistry, 130, Issue 1, p. 11-17.
Delpech S., et al., (2009b), MSFR: Material issues and the Effect of Chemistry Control, 2009 GIF Symposium Paris France (2009).
Fabre S., et al., (2009), Corrosion of metallic materials for molten salt reactors, Proceedings of ICAPP’09, May 10-14 2009, Paper 9309, Tokyo, Japan.
Forsberg C.W., et al., (2007), Liquid Salt Applications and Molten Salt Reactors, presented at ICAPP, 13-18 May 2007, Nice, France.
Forsberg C.W., et al., (2008), Design Options for the Advanced High-Temperature Reactor, Paper presented at ICAPP, 8-12 June 2008, Anaheim, CA, United States.
Holcomb D.E., et al., (2009), An Analysis of Testing Requirements for Fluoride Salt-Cooled High Temperature Reactor Components, ORNL/TM-2009/297, November 2009.
Hoogmoed M.W., (2009), A Coupled Calculation Code System for the Thorium Molten Salt Rector, MSc. Thesis, PNR-131-2009-009, Delft, Netherlands.
Hron M., et al., (2008), Design Reactor Physical Program in the Frame of the MSR-SPHINX Transmuter Concept Development, paper presented at ICAPP, 8-12 June 2008, Anaheim, CA, United States.
Ignatiev V., et al., (2008a), Compatibility of selected Ni-based alloys in molten Li,Na,Be/F salts with PuF3 and tellurium additions, Nuclear Technology, Vol. 164, N°1, pp.130-142, October 2008.
Ignatiev V., et al., (2008b), Main Results of IAEA CRP on Studies of Advanced Options for Effective Incineration of Radioactive Waste: Case for Molten Salt Transmuter Systems, Paper presented at the 10th Information Exchange Meeting on Actinide and Fission Product Partitioning & Transmutation, 6-10 October 2008, Mito, Japan.
Mathieu L., et al., (2009), Possible Configurations for the TMSR and advantages of the Fast Non Moderated Version, Nuclear Science and Engineering 161, pp. 78-89.
Merle-Lucotte E., et al., (2009a), Minimizing the fissile inventory of the molten salt fast reactor, Proceedings of the International Conference Advances in Nuclear Fuel Management IV (ANFM IV), April 2009, Hilton Head Island, USA.
Merle-Lucotte E., et al., (2009b), Optimizing the Burning Efficiency and the Deployment Capacities of the Molten Salt Fast Reactor, Proceedings of the International Conference Global 2009 - The Nuclear Fuel Cycle: Sustainable Options & Industrial Perspectives, September 2009, Paris, France.
Nagy K., et al., (2008), Parametric studies on the fuel salt composition in thermal molten salt breeder reactors, Proceeding of PHYSOR 2008 International Conference, Interlaken, Switzerland, paper 277.
Nagy K., et al., (2010), Definition of breeding gain for molten salt reactors, Proceeding of PHYSOR 2010 International Conference, Pittsburg, USA, to be published.
Renault C., et al., (2009), The Molten Salt Reactor (MSR) in Generation IV - Overview and Perspectives, 2009 GIF Symposium Paris, France (2009).
Salanne M., et al., (2009), Transport in molten LiF-NaF-ZrF4 mixtures: a combined computational and experimental approach, Journal of Fluorine Chemistry, 130, pp. 61-66.
Zherebtsov A., et al., (2008), Experimental Study of Molten Salt Technology for Safe, Low-Waste and Proliferation Resistant Treatment of RadioactiveWaste and Plutonium in Accelerator Driven and Critical Systems, ISTC-1606 Project, Final Report, International Scientific Centre, Moscow, Russian Federation.
E-mail contact: msr@gen-4.org
-----------
The Sodium-Cooled Fast Reactor (SFR) system features a fast-spectrum, sodium-cooled reactor and a closed fuel cycle for efficient management of actinides and conversion of fertile uranium.
The SFR is designed for management of high-level wastes and, in particular, management of plutonium and other actinides. Important safety features of the system include a long thermal response time, a large margin to coolant boiling, a primary system that operates near atmospheric pressure, and intermediate sodium system between the radioactive sodium in the primary system and the power conversion system. Water/steam and carbon-dioxide are being considered as the working fluids for the power conversion system in order to achieve high-level performances in thermal efficiency, safety and reliability. With innovations to reduce capital cost, the SFR can serve markets for electricity.
The fuel cycle employs a full actinide recycle with three major options. The first option is a large size (600 to 1,500 MWe) loop-type sodium-cooled reactor using mixed uranium-plutonium oxide fuel, supported by a fuel cycle based upon advanced aqueous processing at a central location serving a number of reactors. The second option is an intermediate size (300 to 600 MWe) pool-type reactor and the third a small size (50 to 150MWe) modular-type sodium-cooled reactor employing uranium-plutonium-minor-actinide-zirconium metal alloy fuel, supported by a fuel cycle based on pyrometallurgical processing in facilities integrated with the reactor. The outlet temperature is approximately 550 degrees celsius for all the three concepts.
The SFR's fast spectrum also makes it possible to use available fissile and fertile materials (including depleted uranium) considerably more efficiently than thermal spectrum reactors with once-through fuel cycles.
This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.
Diagram source: http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf
Advantages and challenges
The SFR system already benefits from considerable technological experience, but also offers the potential to operate with a high conversion fast spectrum core, with the resulting benefit of increasing the utilization of fuel resources. The envisaged SFR capability to efficiently and nearly completely consume trans-uranium as fuel would reduce the actinide loadings in the high-level radioactive waste it produces. Such reductions would bring benefits in the radioactive waste disposal requirements associated with the system and enhance its non-proliferation attributes. Reducing the capital cost and improving passive safety, especially under transient conditions, are the major challenges for the SFR system.
GIF progress in 2007
Signatories to the SFR System Arrangement (first signed in February 2006) moved research on SFR systems forward during 2007. The most notable achievement was the signing of the first GIF Project Arrangement in March 2007. Signed by the five partners to the SFR System Arrangement, the project agreement sets out a detailed plan for research and development activities in the area of advanced fuel and details the schedule, funding and deliverables expected to achieve this. Other Project Arrangements have since been signed in the areas of component design and balance-of-plant (CD&BOP) and the Global Actinide Cycle International Demonstration (GACID). The CD&BOP aims to develop key components and devices of the plant system and to investigate safe and effective power conversion concepts. The GACID aims to demonstrate on a significant scale that fast neutron reactors can manage the whole actinide inventory and that the associated technologies can satisfy the GIF criteria of safety, economy, sustainability and proliferation resistance and physical protection.
Recent SFR research papers
Chikazawa, Y., Okano, Y., Konomura, M. et al., (2007) A Compact Loop-Type Fast Reactor without Refueling For a Remote Area Power Source, Nuclear Technology, 157, 2, pp. 120-131.
Hahn, D., Kim, Y., Kim, S., Lee, J., Lee, Y. and Jeong H., (2007) Conceptual Design Features of the KALIMER-600 Sodium Cooled Fast Reactor, Global 2007, 9-13 September 2007, Boise, USA.
Niwa, H., Aoto, K. and Morishita, M., Current Status and Perspective of Advanced Loop Type Fast Reactor in Fast Reactor Cycle Technology Development Project, Global 2007, 9-13 September 2007, Boise, USA (2007).
Sienicki, J., Moisseytsev, A., Cho, D., Momozaki, Y., Kilsdonk, D., Haglund, R., Reed, C. and Farmer, M.,(2007) Supercritical Carbon Dioxide Brayton Cycle Energy Conversion for Sodium-Cooled Fast Reactors/Advanced Burner Reactors, Global 2007, 9-13 September 2007, Boise, USA.
Zaetta, A., Dufour, Ph., Pruhliere, G., Rimpault, G., Thevenot, C., Tommasi, J. and Varaine, F. (2007), Innovating Core Design for Sodium Cooled Fast Reactors of Fourth Generation, Paper #7383, ICAPP 2007, 13-18 May 2007, Nice, France.
Chang, Y. (2005), Konomura, M. and Lo Pinto, P., A Case for Small Modular Fast Reactor, Global 2005, 9-13 October 2005, Tsukuba, Japan.
Hahn, D., Kim, Y., Kin, S., Lee, J. and Lee, Y. (2005), Design Concept of KALIMER-600, Global 2005, 9-13 October 2005, Tsukuba, Japan.
Kotake, S., Sakamoto, Y., Ando, M. and Tanaka, T. (2005), Feasibility Study on Commercialized Fast Reactor Cycle Systems/Current Status of the FR System Design, Global 2005, 9-13 October 2005, Tsukuba, Japan.
Mizuno, T., Ogawa, T., Naganuma, M. and Aida, T. (2005), Advanced Oxide Fuel Core Design Study for SFR in the Feasibility Study in Japan, Global 2005, 9-13 October 2005, Tsukuba, Japan.
Related links
DOE Nuclear Energy Research Initiative SFR Program Plan (pdf, 138 kb).
E-mail contact: sfr@gen-4.org
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The Supercritical-Water-Cooled Reactor (SCWR) system is a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water (374 degrees Celsius, 22.1 MPa, or 705 degrees Fahrenheit, 3208 psia).
The supercritical water coolant enables a thermal efficiency about one-third higher than current light-water reactors, as well as simplification in the balance of plant. The balance of plant is considerably simplified because the coolant does not change phase in the reactor and is directly coupled to the energy conversion equipment. The reference system is 1,700 MWe with an operating pressure of 25 MPa, and a reactor outlet temperature of 510 degrees Celsius, possibly ranging up to 550 degrees Celsius. The fuel is uranium oxide. Passive safety features are incorporated similar to those of simplified boiling water reactors.
The SCWR system is primarily designed for efficient electricity production, with an option for actinide management based on two options in the core design: the SCWR may have a thermal or fast-spectrum reactor; the second is a closed cycle with a fast-spectrum reactor and full actinide recycle based on advanced aqueous processing at a central location.
Advantages and challenges
As the system uses existing light water reactor technology, there is already extensive worldwide experience in constructing and operating this sort of reactor. Proposed designs are likely to enjoy high thermal efficiency and a simplified system configuration. A SCWR design could be developed with a fast neutron spectrum. Using fast neutrons with higher kinetic energies would enable the system to produce at least as much fissile material as it consumes (thereby fulfilling the sustainability goal as set out in the Generation IV roadmap). This concept’s tendency to have a positive void reactivity coefficient together with the potential for design basis loss-of-coolant accidents are likely to make this difficult to develop. The other major challenges for the SCWR are to develop a viable core design, accurately estimate the heat transfer coefficient and develop materials for the fuel and core structure that will be sufficiently corrosion-resistant to withstand SCWR conditions.
GIF progress in 2007
The SCWR system research plan was finalised in mid-2007. Project management boards have since been established in the following areas: thermal-hydraulics and safety; materials and chemistry; design and integration. Negotiations to put in place project plans for all of these areas have advanced significantly in the course of 2007.
Recent SCWR research papers and links
J. Yoo, Y. Ishiwatari, Y. Oka, J. Yang, J. Liu, Subchannel analysis of supercritical light water-cooled fast reactor assembly
Nuclear Engineering and Design, 237 (2007), 1096-1105.
Nuclear Engineering and Design, 237 (2007), 1096-1105.
M. Mori, W. Maschek and A. Rineiski, Heterogeneous cores for improved safety performance : A case study: The supercritical water fast reactor, Nuclear Engineering and Design, Volume 236: 14-16 (2006), 1573-1579.
J. Yoo, Y. Ishiwatari, Y. Oka, J. Liu, Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling, Annals of Nuclear Energy, Volume 33 (2006), 945-956.
J. Yoo, Y. Ishiwatari, J. Liu, Composite Core Design of High Power Density Supercritical Water Cooled Fast Reactor, Paper No. 246, GLOBAL 2005, Tsukuba, Japan, 9-13 October 2005.
M. Mori, Core Design Analysis of the Supercritical Water Fast Reactor, Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft, Wissenschaftliche Berichte FZKA 7160 (2005).
M. Mori, Werner Maschek, Eckart Laurien, Koji Morita, Monte-Carlo/Simmer-III Reactivity Coefficients Calculations for the SuperCritical Water Fast Reactor, Paper No. 87753, GLOBAL 2003, New Orleans, Louisiana, 16-21 November 2003.
T. Mukohara, S. Koshizuka, Y. Oka, Core design of a high-temperature fast reactor cooled by supercritical light water, Annals of Nuclear Energy, 26 (1999) 1423-1436.
Y. Oka and S. Kozhizuka, Conceptual Design Study of Advanced Power Reactors, Progress in Nuclear Energy, 32 (1998), 163-177.
J.H. Lee, Y. Oka, S. Koshizuka, Safety System Consideration of Supercritical Water cooled Fast Reactor with Simplified PSA, Reliability Engineering and System Safety, 64 (1999), 327-338.
Y. Oka, S. Koshizuka, Y. Ishiwatari, A. Yamaji, Super light water reactors and Super fast reactors,
ISBN:978-1-4419-6034-4, 416 pages Springer 2010 2.
ISBN:978-1-4419-6034-4, 416 pages Springer 2010 2.
DOE Nuclear Energy Research Initiative SCWR Program Plan (pdf, 272 kb).
Contact: scwr@gen-4.org
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The Very-High-Temperature Reactor (VHTR) is a graphite-moderated, helium-cooled reactor with a thermal neutron spectrum.
The VHTR is designed to be a high-efficiency system, which can supply electricity and process heat to a broad spectrum of high-temperature and energy-intensive processes.
The reference reactor is a 600 MWth core connected to an intermediate heat exchanger to deliver process heat. The reactor core can be a prismatic block core or a pebble-bed core according to the fuel particles assembly. Fuel particles are coated with successive material layers, high temperature resistant, then formed either into fuel compacts embedded in graphite block for the prismatic block-type core reactor, or formed into graphite coated pebbles. The reactor supplies heat with core outlet temperatures up to 1,000 degrees Celsius, which enables such applications as hydrogen production or process heat for the petrochemical industry. As a nuclear heat application, hydrogen can be efficiently produced from only heat and water by using thermochemical iodine-sulfur process, or high temperature electrolysis process or with additional natural gas by applying the steam reformer technology.
Thus, the VHTR offers a high-efficiency electricity production and a broad range of process heat applications, while retaining the desirable safety characteristics in normal as well as off-normal events. Solutions to adequate waste management will be developed. The basic technology for the VHTR has been well established in former High Temperature Gas Reactors plants, such as the US Fort Saint Vrain and Peach Bottom prototypes, and the German AVR and THTR prototypes. The technology is being advanced through near- or medium-term projects lead by several plant vendors and national laboratories, such as: PBMR, GT-HTR300C, ANTARES, NHDD, GT-MHR and NGNP in South Africa, Japan, France, Republic of Korea and the United States. Experimental reactors: HTTR (Japan, 30 MWth) and HTR-10 (China, 10 MWth) support the advanced concept development, and the cogeneration of electricity and nuclear heat application.
This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.
Advantages and challenges
The VHTR offers the potential for the cogeneration of electricity and hydrogen, alongside process heat applications. As the basic technology for VHTR systems has already been established in high temperature gas reactor plants, the design is an evolutionary development. However, the system’s aim of operating above 1000°C presents significant challenges in terms of fuel and materials development, as well as safety under transient conditions.
GIF progress in 2007
With the signing of a VHTR system arrangement by seven GIF members in November 2006, a system research plan has been put in place setting out the VHTR medium-term R&D projects to be pursued. Currently project arrangements to study the following areas are in the final stages of negotiation: the development and validation of materials to be used in VHTR systems; fuels and fuel cycle issues for VHTR systems; and the use of VHTR systems to produce hydrogen. The overall aim of these research efforts is to define the system's baseline concepts by 2010 and to optimize their design and operating features by 2015.
Recent VHTR research papers and links
Oh, C. (2007), Power Cycle and Stress Analyses for High Temperature Gas-Cooled Reactor, 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007).
Sterbentz, J. W. (2007) Low-Enriched Very High Temperature Reactor Core Design, 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007).
R. B. Vilim (2007), Interface Design Studies for the Production of Hydrogen Using the VHTR Coupled to the HTSE Process, 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007).
Billot, P., Hittner, D. and Vasseur, P. (2006) Outlines of the French R&D Program for the development of High and Very High Temperature Reactor, Third International Topical Meeting on High Temperature Reactor Technology, 1-4 October 2006, Johannesburg, South Africa.
Fütterer, M. et al, (2006) Irradiation Results of AVR Fuel Pebbles at Increased Temperature and Burn-Up in the HFR Petten, Third International Topical Meeting on High Temperature Reactor Technology, 1-4 October 2006, Johannesburg, South Africa.
Greyvenstein, R., Correia, M. and Kriel, W. (2006) South Africa's opportunity to maximize the role of nuclear power in a global hydrogen economy, Third International Topical Meeting on High Temperature Reactor Technology, 1-4 October 2006, Johannesburg, South Africa.
Hittner, D. et al (2006), RAPHAEL, A European Project for the development of HTR/VHTR technology for industrial process heat supply and cogeneration, Third International Topical Meeting on High Temperature Reactor Technology, 1-4 October 2006, Johannesburg, South Africa.
Hu, S., Liang, X., Wei, L., (2006) Commissioning and Operation experience and safety experiment at HTR-10, Third International Topical Meeting on High Temperature Reactor Technology, 1-4 October 2006, Johannesburg, South Africa.
Lee, Y-W., Park, J-Y., Kim, YK., Jeong, BG. Kim, YM. (2006), Development of HTGR coated particle fuel technology in Korea, Third International Topical Meeting on High Temperature Reactor Technology, 1-4 October 2006, Johannesburg, South Africa.
Takamatsu, K., Nakagawa, S. Takeda, T. (2006), Development of core dynamics analysis of coolant flow reduction tests of HTTR, Third International Topical Meeting on High Temperature Reactor Technology, 1-4 October 2006, Johannesburg, South Africa.
P. V. Tsvetkov (2006) Coupled Hybrid Monte Carlo - Deterministic Analysis of VHTR Configurations with Advanced Actinide Fuels, Paper ICAPP-6400, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).
Kim, T.K., Taiwo, T.A., Hill, R.N., and Stillman, (2005), Spent Nuclear Fuel Characterization for the VHTR, Paper 67, GLOBAL 2005, Tsukuba, Japan, 9-13 October 2005.
Related links
HTR-10 10MW High Temperature Gas-Cooled Reactor Project
Institute of Nuclear and New Energy Technology, Tsinghua University, People's Republic of China
Institute of Nuclear and New Energy Technology, Tsinghua University, People's Republic of China
RAPHAEL Project
JRC, Euratom
JRC, Euratom
ANTARES
AREVA (France) HTR-VHTR Design
AREVA (France) HTR-VHTR Design
HTTR Project
JAEA, Japan
JAEA, Japan
PBMR Project
Pebble Bed Modular Reactor (Pty) Ltd, South Africa
Pebble Bed Modular Reactor (Pty) Ltd, South Africa
E-mail contact: vhtr@gen-4.org
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