Gen IV Roadmap

GIF-002-00

03-GA50034

ii

A Technology Roadmap for Generation IV Nuclear Energy Systems

DISCLAIMERThis information was prepared as an account of work by the U.S. Nuclear Energy Research Advisory Committee(NERAC) and the Generation IV International Forum (GIF). Neither the NERAC nor any of its members, nor theGIF, nor any of its members, nor any GIF member’s national government agency or employee thereof, makes anywarranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, orusefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringeprivately owned rights. References herein to any specific commercial product, process, or service by trade name,trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, orfavoring by the NERAC or its members, or the GIF or its members, or any agency of a GIF member’s nationalgovernment. The views and opinions of authors expressed herein do not necessarily state or reflect those of theNERAC or its members, or the GIF, its members, or any agency of a GIF member’s national government.

iii


GIF-002-00

A Technology Roadmap forGeneration IV Nuclear Energy Systems

December 2002

Issued by theU.S. DOE Nuclear Energy Research Advisory Committee

and the Generation IV International Forum

iv


v


Contents

AN ESSENTIAL ROLE FOR NUCLEAR ENERGY ..................................................................................................... 1The Long-Term Benefits from Nuclear Energy’s Essential Role ........................................................................... 1Meeting the Challenges of Nuclear Energy’s Essential Role ................................................................................. 2

THE GENERATION IV TECHNOLOGY ROADMAP IN BRIEF ................................................................................. 5An International Effort ........................................................................................................................................... 5Goals for Generation IV ......................................................................................................................................... 6The Generation IV Roadmap Project ..................................................................................................................... 8Evaluation and Selection Methodology .................................................................................................................. 9Generation IV Nuclear Energy Systems ............................................................................................................... 11

FINDINGS OF THE ROADMAP .................................................................................................................................. 13Fuel Cycles and Sustainability ............................................................................................................................. 13Descriptions of the Generation IV Systems.......................................................................................................... 14Missions and Economics for Generation IV ......................................................................................................... 16Safety, Safeguards, and Public Confidence in Generation IV .............................................................................. 18Near-Term Deployment Opportunities and Generation IV .................................................................................. 18

RECOMMENDED R&D FOR THE MOST PROMISING SYSTEMS ........................................................................ 21Introduction........................................................................................................................................................... 21Gas-Cooled Fast Reactor System R&D ................................................................................................................ 22Lead-Cooled Fast Reactor System R&D .............................................................................................................. 27Molten Salt Reactor System R&D ........................................................................................................................ 33Sodium-Cooled Fast Reactor System R&D.......................................................................................................... 38Supercritical-Water-Cooled Reactor System R&D .............................................................................................. 42Very-High-Temperature Reactor System R&D .................................................................................................... 48

RECOMMENDED CROSSCUTTING R&D ................................................................................................................ 53Crosscutting Fuel Cycle R&D .............................................................................................................................. 53Crosscutting Fuels and Materials R&D ................................................................................................................ 60Crosscutting Energy Products R&D ..................................................................................................................... 65Crosscutting Risk and Safety R&D ...................................................................................................................... 69Safety and Reliability Evaluation and Peer Review ............................................................................................. 71Crosscutting Economics R&D .............................................................................................................................. 71Crosscutting Proliferation Resistance and Physical Protection R&D .................................................................. 75

INTEGRATION OF R&D PROGRAMS AND PATH FORWARD ............................................................................... 79Introduction........................................................................................................................................................... 79Overall Advancement of Generation IV ............................................................................................................... 79R&D Programs for Individual Generation IV Systems ........................................................................................ 79Comparison of R&D Timelines ............................................................................................................................ 82Program Implementation ...................................................................................................................................... 82Integration Issues and Opportunities .................................................................................................................... 82

MEMBERS OF THE GENERATION IV ROADMAP PROJECT ................................................................................ 85

ACRONYMS.................................................................................................................................................................. 90

vi


1


The world’s population is expected to expand fromabout 6 billion people to 10 billion people by the year2050, all striving for a better quality of life. As theEarth’s population grows, so will the demand for energyand the benefits that it brings: improved standards ofliving, better health and longer life expectancy, im-proved literacy and opportunity, and many others.Simply expanding energy use using today’s mix ofproduction options, however, will continue to haveadverse environmental impacts and potential long-termconsequences from global climate change. For the Earthto support its population, we must increase the use ofenergy supplies that are clean, safe, and cost-effective.Prominent among these supplies is nuclear energy.

There are currently 438 nuclear power plants in opera-tion around the world, producing 16% of the world’selectricity—the largest share provided by anynongreenhouse-gas-emitting source. This yields asignificant reduction in the environmental impact oftoday’s electric generation. To continue this benefit,new systems will be needed to replace plants as theyretire. In the latter part of this century, the environmen-tal benefits of nuclear energy can expand and evenextend to other energy products besides electricity. Forexample, nuclear energy can be used to generate hydro-gen for use in petroleum refinement and as a transporta-tion fuel to reduce the dependence upon oil, and todesalinate water in areas where fresh water is in shortsupply. To deliver this benefit, new systems will beneeded, requiring near-term deployment of nuclearplants and significant research and development (R&D)on next-generation systems.

Many of the world’s nations, both industrialized anddeveloping, believe that a greater use of nuclear energywill be required if energy security is to be achieved.They are confident that nuclear energy can be used nowand in the future to meet their growing demand forenergy safely and economically, with certainty of long-term supply and without adverse environmental impacts.

To enhance the future role of nuclear energy systems,this technology roadmap defines and plans the necessaryR&D to support a generation of innovative nuclearenergy systems known as Generation IV. Generation IVnuclear energy systems comprise the nuclear reactor andits energy conversion systems, as well as the necessaryfacilities for the entire fuel cycle from ore extraction tofinal waste disposal.

The Long-Term Benefits from NuclearEnergy’s Essential RoleChallenging technology goals for Generation IV nuclearenergy systems are defined in this roadmap in four areas:sustainability, economics, safety and reliability, andproliferation resistance and physical protection. Bystriving to meet the technology goals, new nuclearsystems can achieve a number of long-term benefits thatwill help nuclear energy play an essential role world-wide.

Sustainable Nuclear EnergySustainability is the ability to meet the needs of thepresent generation while enhancing the ability of futuregenerations to meet society’s needs indefinitely into thefuture. In this roadmap, sustainability goals are definedwith focus on waste management and resource utiliza-tion. Other factors that are commonly associated withsustainability, such as economics and environment,a areconsidered separately in the technology roadmap tostress their importance. Looking ahead to the findingsof this roadmap, the benefits of meeting sustainabilitygoals include:

• Extending the nuclear fuel supply into futurecenturies by recycling used fuel to recover its energycontent, and by converting 238U to new fuel

• Having a positive impact on the environmentthrough the displacement of polluting energy andtransportation sources by nuclear electricity genera-tion and nuclear-produced hydrogen

AN ESSENTIAL ROLE FOR NUCLEAR ENERGY

a Internationally, and especially in the context of the recent World Summit on Sustainable Development held in Johannesburg in August 2002,sustainable development is usually examined from three points of view: economic, environmental, and social. Generation IV has adopted anarrower definition of sustainability in order to balance the emphasis on the various goal areas. For a more complete discussion ofsustainability, see NEA News, No. 19.1, available at the http://www.nea.fr/html/sd/welcome.html website.

2


• Allowing geologic waste repositories to accept thewaste of many more plant-years of nuclear plantoperation through substantial reduction in theamount of wastes and their decay heat

• Greatly simplifying the scientific analysis anddemonstration of safe repository performance forvery long time periods (beyond 1000 years), by alarge reduction in the lifetime and toxicity of theresidual radioactive wastes sent to repositories forfinal geologic disposal.

Competitive Nuclear EnergyEconomics goals broadly consider competitive costs andfinancial risks of nuclear energy systems. Lookingahead, the benefits of meeting economics goals include:

• Achieving economic life-cycle and energy produc-tion costs through a number of innovative advancesin plant and fuel cycle efficiency, design simplifica-tions, and plant sizes

• Reducing economic risk to nuclear projects throughthe development of plants built using innovativefabrication and construction techniques, and possi-bly modular designs

• Allowing the distributed production of hydrogen,fresh water, district heating, and other energyproducts to be produced where they are needed.

Safe and Reliable SystemsMaintaining and enhancing the safe and reliable opera-tion is an essential priority in the development of next-generation systems. Safety and reliability goals broadlyconsider safe and reliable operation, improved accidentmanagement and minimization of consequences, invest-ment protection, and reduced need for off-site emer-gency response. Looking ahead, the benefit of meetingthese goals includes:

• Increasing the use of inherent safety features, robustdesigns, and transparent safety features that can beunderstood by nonexperts

• Enhancing public confidence in the safety of nuclearenergy.

Proliferation Resistance and PhysicalProtectionProliferation resistance and physical protection considermeans for controlling and securing nuclear material andnuclear facilities. Looking ahead, the benefits of meet-ing these goals include:

• Providing continued effective proliferation resis-tance of nuclear energy systems through improveddesign features and other measures

• Increasing physical protection against terrorism byincreasing the robustness of new facilities.

Meeting the Challenges of Nuclear Energy’sEssential RoleTo play an essential role, future nuclear energy systemswill need to provide (1) manageable nuclear waste,effective fuel utilization, and increased environmentalbenefits, (2) competitive economics, (3) recognizedsafety performance, and (4) secure nuclear energysystems and nuclear materials. These challenges,described below, are the basis for setting the goals ofnext-generation nuclear energy systems in this roadmap.

Disposition of discharged fuel or other high-levelradioactive residues in a geological repository is thepreferred choice of most countries, and good technicalprogress is being made. Long-term retrievable surfaceor subsurface repositories are also being assessed. Theprogress toward realizing a geologic repository in theUnited States at Yucca Mountain and in other countrieslike Finland and Sweden demonstrates the viability ofrepositories as a solution. However, the extensive use ofnuclear energy in the future requires the optimal use ofrepository space and the consideration of closing the fuelcycle.

Today, most countries use the once-through fuel cycle,whereas others close the fuel cycle by recycling. Recy-cling (using either single or multiple passes) recoversuranium and plutonium from the spent fuel and uses it tomake new fuel, thereby producing more power andreducing the need for enrichment and uranium mining.Recycling in a manner that does not produce separatedplutonium can further avoid proliferation risks. How-ever, recycling has proven to be uneconomical today,given plentiful supplies of uranium at low and stableprices. This will eventually change, and closing the fuelcycle will be favored when the cost of maintaining anopen cycle exceeds that of a closed cycle. With recy-cling, other benefits are realized: the high-level radioac-tive residues occupy a much-reduced volume, can bemade less toxic, and can be processed into a moresuitable form for disposal. In addition, reactors can bedesigned to transmute troublesome long-lived heavyelements. Achieving these benefits, however, willrequire significant R&D on fuel cycle technology.

3


The economic performance of nuclear power has beenmixed: On the positive side, the cost of nuclear powergeneration in many countries is the same as or less thanthe cost of producing electricity from coal, oil, or naturalgas. On the other hand, construction of advancednuclear energy systems must address their economics ina variety of changing markets and overcome theirtraditionally high construction costs. While the currentgeneration of plants generates electricity at competitivecosts, construction costs are not competitive enough, andlicensing needs to be more predictable to stimulatewidespread interest in new nuclear construction. Sig-nificant R&D is needed to reduce capital costs andconstruction times for new plants.

Overall, the safety and environmental record of nuclearpower is excellent. Despite this, public confidence inthe safety of nuclear power needs to be increased. Newsystems should address this need with clear and trans-parent safety approaches that arise from R&D onadvanced systems.

Fissile materials within civilian nuclear power programsare well-safeguarded by an effective internationalsystem. Current-generation plants have robust designsand added precautions against acts of terrorism. Never-

theless, it is desirable for future nuclear fuel cycles andnuclear materials safeguards to design from the start aneven higher degree of resistance to nuclear materialdiversion or undeclared production. Further, questionshave arisen about the vulnerability of nuclear plants toterrorist attack. In response, future nuclear energysystems will provide improved physical protectionagainst the threats of terrorism.

This roadmap has been prepared by many experts fromcountries that have experience developing and operatingnuclear reactors and facilities. These experts brought abroad international perspective on the needs and oppor-tunities for nuclear energy in the 21st century. Theopportunities for advancing Generation IV systems willalso depend on gaining public confidence, which can beenhanced through the openness of the process of devel-oping and deploying Generation IV systems. Thefindings of this roadmap and the R&D plans that arebased on it will be communicated to the public on aregular basis, and opportunities for stakeholder groups toprovide feedback on the plans will be offered.

4


5


THE GENERATION IV TECHNOLOGY ROADMAP IN BRIEF

An International EffortTo advance nuclear energy to meet future energy needs,ten countries—Argentina, Brazil, Canada, France, Japan,the Republic of Korea, the Republic of South Africa,Switzerland, the United Kingdom, and the UnitedStates—have agreed on a framework for internationalcooperation in research for a future generation of nuclearenergy systems, known as Generation IV. The figurebelow gives an overview of the generations of nuclearenergy systems. The first generation was advanced inthe 1950s and 60s in the early prototype reactors. Thesecond generation began in the 1970s in the largecommercial power plants that are still operating today.Generation III was developed more recently in the 1990swith a number of evolutionary designs that offer signifi-cant advances in safety and economics, and a numberhave been built, primarily in East Asia. Advances toGeneration III are underway, resulting in several (so-called Generation III+) near-term deployable plants thatare actively under development and are being consideredfor deployment in several countries. New plants builtbetween now and 2030 will likely be chosen from theseplants. Beyond 2030, the prospect for innovativeadvances through renewed R&D has stimulated interest

worldwide in a fourth generation of nuclear energysystems.

The ten countries have joined together to form theGeneration IV International Forum (GIF) to developfuture-generation nuclear energy systems that can belicensed, constructed, and operated in a manner that willprovide competitively priced and reliable energy prod-ucts while satisfactorily addressing nuclear safety, waste,proliferation, and public perception concerns. Theobjective for Generation IV nuclear energy systems is tohave them available for international deployment aboutthe year 2030, when many of the world’s currentlyoperating nuclear power plants will be at or near the endof their operating licenses.

Nuclear energy research programs around the worldhave been developing concepts that could form the basisfor Generation IV systems. Increased collaboration onR&D to be undertaken by the GIF countries will stimu-late progress toward the realization of such systems.With international commitment and resolve, the worldcan begin to realize the benefits of Generation IVnuclear energy systems within the next few decades.

Gen I Gen II

Commercial PowerReactors

Early PrototypeReactors

Generation I

- Shippingport- Dresden, Fermi I- Magnox

Generation II

- LWR-PWR, BWR- CANDU- AGR

1950 1960 1970 1980 1990 2000 2010 2020 2030

Generation IV

- Highly Economical

- Enhanced Safety

- Minimal Waste

- Proliferation Resistant

- ABWR- System 80+

AdvancedLWRs

Generation III

Gen III Gen III+ Gen IV

Generation III +

Evolutionary Designs Offering Improved Economics for Near-Term Deployment

Gen I Gen II

Commercial PowerReactors


Generation I



Generation I


Generation II

- LWR-PWR, BWR- CANDU- AGR

1950 1960 1970 1980 1990 2000 2010 2020 2030

Generation IV

- Highly Economical

- Enhanced Safety

- Minimal Waste

- Proliferation Resistant

- ABWR- System 80+

AdvancedLWRs

Generation III

AdvancedLWRs

Generation III

Gen III Gen III+ Gen IV

Generation III +


Generation III +


6


Beginning in 2000, the countries constituting the GIFbegan meeting to discuss the research necessary tosupport next-generation reactors. From those initialmeetings a technology roadmap to guide the GenerationIV effort was begun. The organization and execution ofthe roadmap became the responsibility of a RoadmapIntegration Team that is advised by the Subcommittee onGeneration IV Technology Planning of the U.S. Depart-ment of Energy’s Nuclear Energy Research AdvisoryCommittee (NERAC). Roadmapping is a methodologyused to define and manage the planning and execution oflarge-scale R&D efforts. The GIF agreed to support thepreparation of a roadmap, and the roadmap became thefocal point of their efforts. More than one hundredtechnical experts from ten countries have contributed toits preparation.

The scope of the R&D described in this roadmap coversall of the Generation IV systems. However, each GIFcountry will focus on those systems and the subset ofR&D activities that are of greatest interest to them.Thus, the roadmap provides a foundation for formulatingnational and international program plans on which theGIF countries will collaborate to advance Generation IVsystems.

In the United States, the Generation IV TechnologyRoadmap is complemented by an earlier Near-TermDeployment Roadmap.b These roadmaps and otherplanning documents will be the foundation for a set ofR&D program plans encompassing the objectives ofdeploying more mature nuclear energy systems by 2010,developing separations and transmutation technology forreducing existing stores of spent nuclear fuel, anddeveloping next generation nuclear energy systems inthe long term.

Goals for Generation IVAs preparations for the Generation IV TechnologyRoadmap began, it was necessary to establish goals forthese nuclear energy systems. The goals have threepurposes: First, they serve as the basis for developingcriteria to assess and compare the systems in the tech-nology roadmap. Second, they are challenging andstimulate the search for innovative nuclear energysystems—both fuel cycles and reactor technologies.Third, they will serve to motivate and guide the R&D onGeneration IV systems as collaborative efforts getunderway.

Eight goals for Generation IV [see the box below] aredefined in the four broad areas of sustainability, eco-nomics, safety and reliability, and proliferation resis-tance and physical protection. Sustainability goals focuson fuel utilization and waste management. Economicsgoals focus on competitive life cycle and energy produc-

b “A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010, Volume I, Summary Report,” U.S. Department of EnergyNuclear Energy Research Advisory Committee Subcommittee on Generation IV Technology Planning, available at the http://nuclear.gov/nerac/ntdroadmapvolume1.pdf website, accessed September 2002.

Goals for Generation IV Nuclear Energy Systems Sustainability–1 Generation IV nuclear energy systems will

provide sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production.

Sustainability–2 Generation IV nuclear energy systems will minimize and manage their nuclear waste and notably reduce the long-term stewardship burden, thereby improving protection for the public health and the environment.

Economics–1 Generation IV nuclear energy systems will have a clear life-cycle cost advantage over other energy sources.

Economics–2 Generation IV nuclear energy systems will have a level of financial risk comparable to other energy projects.

Safety and Reliability–1 Generation IV nuclear energy systems operations will excel in safety and reliability.

Safety and Reliability–2 Generation IV nuclear energy systems will have a very low likelihood and degree of reactor core damage.

Safety and Reliability–3 Generation IV nuclear energy systems will eliminate the need for offsite emergency response.

Proliferation Resistance and Physical Protection–1 Generation IV nuclear energy systems will increase the assurance that they are a very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism.

tion costs and financial risk. Safety and reliability goalsfocus on safe and reliable operation, improved accidentmanagement and minimization of consequences, invest-ment protection, and essentially eliminating the techni-cal need for off-site emergency response. The prolifera-tion resistance and physical protection goal focuses oncontrolling and securing nuclear material and nuclearfacilities. Each broad goal area is briefly discussedbelow.

7


are adequate, accidents are prevented, and off-normalsituations do not deteriorate into severe accidents. Atthe same time, competitiveness requires a very highlevel of reliability and performance. There has been adefinite trend over the years to improve the safety andreliability of nuclear power plants, reduce the frequencyand degree of off-site radioactive releases, and reducethe possibility of significant plant damage. Lookingahead, Generation IV systems will face new challengesto their reliability at higher temperatures and otheranticipated conditions. Generation IV systems havegoals to achieve high levels of safety and reliabilitythrough further improvements. The three safety andreliability goals continue the past trend and seek simpli-fied designs that are safe and further reduce the potentialfor severe accidents and minimize their consequences.The achievement of these ambitious goals cannot relyonly upon technical improvements, but will also requiresystematic consideration of human performance as amajor contributor to the plant availability, reliability,inspectability, and maintainability.

Proliferation resistance and physical protection are alsoessential priorities in the expanding role of nuclearenergy systems. The safeguards provided by the NuclearNonproliferation Treaty have been highly successful inpreventing the use of civilian nuclear energy systems fornuclear weapons proliferation. This goal applies to allinventories of nuclear materials (both source materialsand special fissionable materials) in the system involvedin enrichment, conversion, fabrication, power produc-tion, recycling, and waste disposal. In addition, existingnuclear plants are highly secure and designed to with-stand external events such as earthquakes, floods,tornadoes, plane crashes, and fires. Their many protec-tive features considerably reduce the impact of externalor internal threats through the redundancy, diversity, andindependence of the safety systems. This goal points outthe need to increase public confidence in the security ofnuclear energy facilities against terrorist attacks. Ad-vanced systems need to be designed from the start withimproved physical protection against acts of terrorism,to a level commensurate with the protection of othercritical systems and infrastructure.

Sustainability is the ability to meet the needs of presentgenerations while enhancing and not jeopardizing theability of future generations to meet society’s needsindefinitely into the future. There is a growing desire insociety for the production of energy in accordance withsustainability principles. Sustainability requires theconservation of resources, protection of the environment,preservation of the ability of future generations to meettheir own needs, and the avoidance of placing unjustifiedburdens upon them. Existing and future nuclear powerplants meet current and increasingly stringent clean airobjectives, since their energy is produced withoutcombustion processes. The two sustainability goalsencompass the interrelated needs of improved wastemanagement, minimal environmental impacts, effectivefuel utilization, and development of new energy productsthat can expand nuclear energy’s benefits beyondelectrical generation.

Economic competitiveness is a requirement of themarketplace and is essential for Generation IV nuclearenergy systems. In today’s environment, nuclear powerplants are primarily baseload units that were purchasedand operated by regulated public and private utilities. Atransition is taking place worldwide from regulated toderegulated energy markets, which will increase thenumber of independent power producers and merchantpower plant owner/operators. Future nuclear energysystems should accommodate a range of plant ownershipoptions and anticipate a wider array of potential rolesand options for deploying nuclear power plants, includ-ing load following and smaller units. While it is antici-pated that Generation IV nuclear energy systems willprimarily produce electricity, they will also help meetanticipated future needs for a broader range of energyproducts beyond electricity. For example, hydrogen,process heat, district heating, and potable water willlikely be needed to keep up with increasing worldwidedemands and long-term changes in energy use. Genera-tion IV systems have goals to ensure that they areeconomically attractive while meeting changing energyneeds.

Safety and reliability are essential priorities in thedevelopment and operation of nuclear energy systems.Nuclear energy systems must be designed so that duringnormal operation or anticipated transients safety margins

8


The Generation IV Roadmap ProjectAs the Generation IV goals were being finalized,preparations were made to develop the Generation IVtechnology roadmap. Theorganization of the roadmapis shown in the figure at theright. The Roadmap Inte-gration Team (RIT) is theexecutive group. Groups ofinternational experts wereorganized to undertakeidentification and evaluationof candidate systems, and todefine R&D to supportthem.

In a first step, an EvaluationMethodology Group wasformed to develop a processto systematically evaluatethe potential of proposedGeneration IV nuclearenergy systems to meet the Generation IV goals. Adiscussion of the Evaluation Methodology Group’sevaluation methodology is included in this report. At thesame time, a solicitation was issued worldwide, request-ing that concept proponents submit information onnuclear energy systems that they believe could meetsome or all of the Generation IV goals. Nearly 100concepts and ideas were received from researchers in adozen countries.

Technical Working Groups (TWGs) were formed—covering nuclear energy systems employing water-cooled, gas-cooled, liquid-metal-cooled, and nonclassi-cal reactor concepts—to review the proposed systemsand evaluate their potential using the tools developed bythe Evaluation Methodology Group. Because of thelarge number of system concepts submitted, the TWGscollected their concepts into sets of concepts withsimilar attributes. The TWGs conducted an initialscreening, termed screening for potential, to eliminatethose concepts or concept sets that did not have reason-able potential for advancing the goals, or were toodistant or technically infeasible.

Following the screening for potential, the TWGs con-ducted a final screening to assess quantitatively thepotential of each concept or concept set to meet theGeneration IV goals. The efforts of the TWGs are

briefly presented in this technical roadmap report. TheTWG Reports are included in their entirety on theRoadmap CD-ROM, along with the reports of the othergroups.

A Fuel Cycle Crosscut Group (FCCG) was also formedat a very early stage to explore the impact of the choiceof fuel cycle on major elements of sustainability—especially waste management and fuel utilization. Theirmembers were equally drawn from the working groups,allowing them to compare their insights and findingsdirectly. Later, other Crosscut Groups were formedcovering economics, risk and safety, fuels and materials,and energy products. The Crosscut Groups reviewed theTWG reports for consistency in the technical evaluationsand subject treatment, and continued to make recom-mendations regarding the scope and priority for cross-cutting R&D in their subject areas. Finally, the TWGsand Crosscut Groups worked together to report on theR&D needs and priorities of the most promising concepts.

The international experts that contributed to thisroadmap represented all ten GIF countries, theOrganisation for Economic Cooperation and Develop-ment Nuclear Energy Agency, the European Commis-sion, and the International Atomic Energy Agency.

NERAC

NERAC Subcommitteeon Generation IV

Technology Planning

Technical Community

• Industry

• Universities

• National Laboratories

DOE-NE

Roadmap Integration Team (RIT)

Evaluation Methodology

Water-Cooled Systems

Gas-Cooled Systems

Liquid-Metal-Cooled Systems

Non-Classical Systems

Generation IV International Forum (GIF)

Argentina Brazil France

S. AfricaKorea Switzerland UK US

Canada Japan

Fu

el C

ycle

Cro

sscu

t

Fu

els

& M

ater

ials

Ris

k &

Saf

ety

Eco

no

mic

s

En

erg

y P

rod

uct

s

9


Evaluation and Selection MethodologyThe selection of the systems to be developed as Genera-tion IV was accomplished in the following steps:

1. Definition and evaluation of candidate systems

2. Review of evaluations and discussion of desiredmissions (national priorities) for the systems

3. Final review of evaluations and performance tomissions

4. Final decision on selections to Generation IV andidentification of near-term deployable designs.

The first step was the collective work of the roadmapparticipants and the NERAC Subcommittee on Genera-tion IV Technology Planning over a one-year period. Itwas concluded with a broad consistency review acrossthe candidate concepts, and reviewed by the Subcommit-tee in early April 2002. The latter three steps continuedto be advised by the Subcommittee but were increas-ingly taken up by the GIF members in a series of meet-ings in the first half of 2002, culminating in the selectionof six Generation IV systems by the GIF. The entireprocess is summarized below, beginning with a detailedexplanation of the evaluation methodology in the firststep.

The use of a common evaluation methodology is acentral feature of the roadmap project, providing aconsistent basis for evaluating the potential of manyconcepts to meet the Generation IV goals. The method-ology was developed by the Evaluation MethodologyGroup at an early stage in the project. The basic ap-proach is to formulate a number of factors that indicateperformance relative to the goals, called criteria, andthen to evaluate concept performance against thesecriteria using specific measures, called metrics.

Two evaluation stages were employed, screening forpotential and final screening. The screening for poten-tial evaluation was designed to eliminate concepts thatlacked sufficient potential, based on the TWG’s judg-ment of their performance against the evaluation criteria.The final screening evaluation was performed forconcepts that passed the screening for potential and wasdesigned to support selection of a small number ofGeneration IV concepts. This final screening employeda more detailed and quantitative set of evaluation criteriathan the screening for potential. Numerical scales wereemployed for a number of the criteria, and weights were

assigned to the criteria associated with each goal. Thescales were established relative to a representativeadvanced light water reactor baseline. To complete theselection process, the GIF members considered theevaluations and eventually selected six to become thebasis for Generation IV. They also considered a numberof plant designs that had good potential for deploymentin the near term, and selected 16 such designs forrecognition as International Near-Term Deployment(INTD). Both lists are presented in the next chapter.

The following figure presents the four goal areas, withthe eight goals arranged under them, and the 15 criteriaand their 24 metrics assigned to the various goals. Thecriteria and metrics are grouped to indicate which goalsthey were assigned to. For example, under thesustainability goal area there are two goals. The firstgoal, “SU1 Resource Utilization,” is evaluated using asingle focused criterion named, “SU1-1 Fuel Utiliza-tion.” The second goal, “SU2 Waste Minimization andManagement” is evaluated using two criteria. It is veryimportant to note that the criteria are only a sampling ofmany factors that could have been evaluated—they werenot selected to be exhaustive but for their ability todiscriminate between concepts on important attributes.

For each criterion, the TWGs evaluated each conceptand specified a probability distribution for its perfor-mance potential to reflect both the expected performanceand performance uncertainty. The Crosscut Groups andthe Roadmap Integration Team reviewed these evalua-tions and recommended changes to make them consis-tent. For a goal evaluated with several criteria, the goalevaluation was combined using criteria weights sug-gested by the Evaluation Methodology Group. Com-parisons of Generation IV candidates were mostly doneat the goal level.

A central feature of the roadmap is that the eight goalsof Generation IV are all equally important. That is, apromising concept should ideally advance each, and notcreate a weakness in one goal to gain strength in another.On the other hand, promising concepts will usuallyadvance one or more of the goals or goal areas morethan others. This will be apparent in the six systemsrecommended below for Generation IV. It should beemphasized that while these numerical evaluation resultswere a primary input to system selection, additionalfactors and judgment were also considered in the selec-tion process, as described below.

10


24 Metrics

Roll Up of Metrics, Criteria, Goals and Goal Areas

ProliferationResistance and Physical Protection PR1-2 Vulnerability of installations • Passive safety features

SR1 Operational Safety and Reliability

EC1 Life Cycle Cost

Safety and Reliability

Sustainability

Economics

SU1 Resource Utilization

PR1 Proliferation Resistance and Physical Protection

SU1-1 Fuel Utilization

EC1-1 Overnight construction costs

SR1-1 Reliability

PR1-1 Susceptibility to diversion or undeclared production

• Use of fuel resources

• Overnight construction costs

• Forced outage rate

• Separated materials• Spent fuel characteristics

SR2-1 Robust safety features • Reliable reactivity control• Reliable decay heat removal

SR1-2 Worker/public - routine exposure

• Routine exposures

SR1-3 Worker/public - accident exposure

• Accident exposures

SR3-2 Robust mitigation features • Long system time constants• Long and effective holdup

SR2-2 Well-characterized models• Dominant phenomena -

uncertainty• Long fuel thermal response time• Integral experiments scalability

SR3-1 Well-characterized sourceterm/energy

• Source term• Mechanis ms for energy release

SR2 Core Damage

SR3 Offsite Emergency Response

SU2 Waste Minimization and Management

SU2-1 Waste minimization• Waste mass• Volume• Heat load• Radiotoxicity

SU2-2 Environmental impact • Environmental impact

EC2 Risk to CapitalEC2-1 Construction duration • Construction duration

EC1-1 Overnight construction costs • Overnight construction costs

EC1-2 Production costs • Production costs

of waste management and disposal

15 Criteria8 Goals4 Goal Areas

EC2-1 Construction duration • Construction duration

24 Metrics

Roll Up of Metrics, Criteria, Goals and Goal Areas

ProliferationResistance and Physical Protection PR1-2 Vulnerability of installations • Passive safety features

SR1 Operational Safety and Reliability

EC1 Life Cycle Cost

Safety and Reliability

Sustainability

Economics

SU1 Resource Utilization

PR1 Proliferation Resistance and Physical Protection

SU1-1 Fuel Utilization

EC1-1 Overnight construction costs

SR1-1 Reliability

PR1-1 Susceptibility to diversion or undeclared production

• Use of fuel resources

• Overnight construction costs

• Forced outage rate

• Separated materials• Spent fuel characteristics

SR2-1 Robust safety features • Reliable reactivity control• Reliable decay heat removal

SR1-2 Worker/public - routine exposure

• Routine exposures

SR1-3 Worker/public - accident exposure

• Accident exposures

SR3-2 Robust mitigation features • Long system time constants• Long and effective holdup

SR2-2 Well-characterized models• Dominant phenomena -

uncertainty• Long fuel thermal response time• Integral experiments scalability

SR3-1 Well-characterized sourceterm/energy

• Source term• Mechanis ms for energy release

SR2 Core Damage

SR3 Offsite Emergency Response

SU2 Waste Minimization and Management

SU2-1 Waste minimization• Waste mass• Volume• Heat load• Radiotoxicity

SU2-2 Environmental impact • Environmental impact

EC2 Risk to CapitalEC2-1 Construction duration • Construction duration

EC1-1 Overnight construction costs • Overnight construction costs

EC1-2 Production costs • Production costs

of waste management and disposal

15 Criteria8 Goals4 Goal Areas

EC2-1 Construction duration • Construction duration

Near the end of the first step, the GIF met to conduct thesecond step of the selection process in February 2002.Leaders from the NERAC Subcommittee participated inthe meeting. The GIF reviewed the preliminary evalua-tion results and discussed additional considerations thatwould be important to their final decision. These

included a review of the important conclusions of thefuel cycle studies, which helped to suggest the variousmissions for Generation IV systems that were of inter-est: electricity and hydrogen production and actinidec

management. These missions are outlined in a sectionbelow.

cThe term actinide refers to the heaviest elements found in used reactor fuel, many of which have long half-lives, including isotopes ofuranium, plutonium, neptunium, americium and curium.

11


A final review of evaluations and performance tomissions by the GIF Experts Group completed the thirdstep in April 2002. The GIF met in May and July 2002to conduct the fourth step. In brief, the candidateconcepts that emerged from the final screening werediscussed. Each was introduced with a presentation ofthe concept in terms of final evaluations, performance ofmissions, and estimated deployment dates and R&Dcosts. The Policy members discussed the concepts untila consensus was reached on six systems found to be themost promising and worthy of collaborative development.

Generation IV Nuclear Energy SystemsThe Generation IV roadmap process described in theprevious section culminated in the selection of sixGeneration IV systems. The motivation for the selectionof six systems is to

• Identify systems that make significant advancestoward the technology goals

• Ensure that the important missions of electricitygeneration, hydrogen and process heat production,and actinide management may be adequatelyaddressed by Generation IV systems

• Provide some overlapping coverage of capabilities,because not all of the systems may ultimately beviable or attain their performance objectives andattract commercial deployment

• Accommodate the range of national priorities andinterests of the GIF countries.

The following six systems, listed alphabetically, wereselected to Generation IV by the GIF:

Generation IV System Acronym

Gas-Cooled Fast Reactor System GFR

Lead-Cooled Fast Reactor System LFR

Molten Salt Reactor System MSR

Sodium-Cooled Fast Reactor System SFR

Supercritical-Water-Cooled Reactor System SCWR

Very-High-Temperature Reactor System VHTR

The six Generation IV systems are summarized in thenext section after a short introduction of the FCCGfindings. The INTD systems are described later in thereport. In addition to overall summaries regarding fuelcycles and overall sustainability, the section describesmissions and economic outlook, approach to safety andreliability, and path forward on proliferation resistanceand physical protection.

12


13


Fuel Cycles and SustainabilityThe studies of the Fuel Cycle Crosscut Group are centralto the development of systems that encompass completefuel cycles. They defined four general classes of nuclearfuel cycle, ranging through (1) the once-through fuelcycle, (2) a fuel cycle with partial recycle of plutonium,(3) a fuel cycle with full plutonium recycle, and (4) afuel cycle with full recycle of transuranic elements.These four general classes were modeled over the nextcentury based on projections of the demand for nuclearenergy developed by the World Energy Council and theInternational Institute for Applied Systems Analysis.The majority of the analyses were based on a projectionthat nuclear energy would only maintain its currentmarket share of electricity, although a number of alterna-tive projections that included the expansion or decline ofnuclear energy’s role were considered to explore thesensitivity of the conclusions.

nuclear energy with the once-through cycle is theavailability of repository space worldwide [see leftfigure]. This becomes an important issue, requiring newrepository development in only a few decades (e.g., atypical repository is of the order of 100 000 tonnecapacity). In the longer term, beyond 50 years, uraniumresource availability also becomes a limiting factor [seeright figure] unless breakthroughs occur in mining orextraction technologies.

Systems that employ a fully closed fuel cycle hold thepromise to reduce repository space and performancerequirements, although their costs must be held toacceptable levels. Closed fuel cycles permit partitioningthe nuclear waste and management of each fraction withthe best strategy. Advanced waste management strate-gies include the transmutation of selected nuclides, cost-effective decay-heat management, flexible interimstorage, and customized waste forms for specific geo-

FINDINGS OF THE ROADMAP

As a reference case, the FCCG determined wastegeneration and resource use for the once-through cycle.While this fuel cycle option is the most uranium re-source-intensive and generates the most waste in theform of used nuclear fuel, the amounts of waste pro-duced are small compared to other energy technologies.In addition, the existing known and speculative eco-nomic uranium resources are sufficient to support aonce-through cycle at least until mid-century. Theyfound that the limiting factor facing an essential role for

logic repository environments. These strategies hold thepromise to reduce the long-lived radiotoxicity of wastedestined for geological repositories by at least an orderof magnitude. This is accomplished by recovering mostof the heavy long-lived radioactive elements. Thesereductions and the ability to optimally condition theresidual wastes and manage their heat loads permit farmore efficient use of limited repository capacity andenhances the overall safety of the final disposal ofradioactive wastes.

0

10

20

30

40

50

2000 2020 2040 2060 2080 2100

Year

Cu

mu

lati

ve N

atu

ral U

(M

illio

n T

on

nes

)

LWR Once Through

Fast ReactorsIntroduced 2050

KnownResources

Speculative Resources

Worldwide Uranium Resource Utilization


0

10

20

30

40

50

2000 2020 2040 2060 2080 2100

Year

Cu

mu

lati

ve N

atu

ral U

(M

illio

n T

on

nes

)

LWR Once Through


KnownResources




0

10

20

30

40

50

2000 2020 2040 2060 2080 2100

Year

Cu

mu

lati

ve N

atu

ral U

(M

illio

n T

on

nes

)

LWR Once Through


KnownResources




0

100

200

300400

500600

700

Worldwide Spent Fuel

2000 2020 2040 2060 2080 2100

Year

Hea

vy M

etal

Mas

s (T

ho

usa

nd

To

nn

es) LWR Once Through

LWR + Fast Reactor

0

100

200

300400

500600

700


2000 2020 2040 2060 2080 2100

Year

Hea

vy M

etal

Mas

s (T

ho

usa

nd

To

nn


LWR + Fast Reactor

0

100

200

300400

500600

700


2000 2020 2040 2060 2080 2100

Year

Hea

vy M

etal

Mas

s (T

ho

usa

nd

To

nn


LWR + Fast Reactor

14


Because closed fuel cycles require the partitioning ofspent fuel, they have been perceived as increasing therisk of nuclear proliferation. The advanced separationstechnologies for Generation IV systems are designed toavoid the separation of plutonium and incorporate otherfeatures to enhance proliferation resistance and incorpo-rate effective safeguards. In particular, to help meet theGeneration IV goal for increased proliferation resistanceand physical protection, all Generation IV systemsemploying recycle avoid separation of plutonium fromother actinides and incorporate additional features toreduce the accessibility and weapons attractiveness ofmaterials at every stage of the fuel cycle.

In the most advanced fuel cycles using fast-spectrumreactors and extensive recycling, it may be possible toreduce the radiotoxicity of all wastes such that theisolation requirements can be reduced by several ordersof magnitude (e.g., for a time as low as 1000 years) after

discharge from the reactor. This would have a beneficialimpact on the design of future repositories and disposalfacilities worldwide. However, this scenario can only beestablished through considerable R&D on recyclingtechnology. This is a motivating factor in the roadmapfor the emphasis on crosscutting fuel cycle R&D.

The studies also established an understanding of theability of various reactors to be combined in so-calledsymbiotic fuel cycles. For example, combinations ofthermal reactors and fast reactors are found to work welltogether. As shown in the figure to the left, they featurethe recycle of actinides from the thermal systems intothe fast systems, and exhibit the ability to reduce ac-tinide inventories worldwide. Improvements in theburnup capability of gas- or water-cooled thermalreactors may also contribute to actinide management in asymbiotic system. Thermal systems also have theflexibility to develop features, such as hydrogen produc-tion in high-temperature gas reactors or highly economi-cal light water reactors, which are part of an overallsystem offering a more sustainable future. This is amotivating factor in the roadmap for having a portfolioof Generation IV systems rather than a single system—realizing that various combinations of a few systems inthe portfolio will be able to provide a desirable symbi-otic system worldwide.

As a final note, the FCCG observed that nuclear energyis unique in the market since its fuel cycle contributesonly about 20% of its production cost. This providesflexibility in separating the approach for meeting theeconomics and safety goals from the approach formeeting sustainability and safeguards goals. That is,adopting a fuel cycle that is advanced beyond the once-through cycle may be achievable at a reasonable cost.

Descriptions of the Generation IV SystemsEach Generation IV system is described briefly, inalphabetical order, below.

GFR – Gas-Cooled Fast Reactor SystemThe Gas-Cooled Fast Reactor (GFR) system features afast-neutron spectrum and closed fuel cycle for efficientconversion of fertile uranium and management ofactinides. A full actinide recycle fuel cycle with on-sitefuel cycle facilities is envisioned. The fuel cycle facili-ties can minimize transportation of nuclear materials andwill be based on either advanced aqueous, pyrometallur-gical, or other dry processing options. The referencereactor is a 600-MWth/288-MWe, helium-cooled systemoperating with an outlet temperature of 850°C using adirect Brayton cycle gas turbine for high thermal effi-

Mining &Enrichment

Fabrication

ThermalReactors

Repository

Fabrication

FastReactors

Symbiotic Fuel Cycles

Mining &Enrichment

Mining &Enrichment

FabricationFabrication

ThermalReactorsThermalReactors

Recycling

RepositoryRepository


FastReactors

FastReactors

Recycling


Depleted UF6Enriched UF6

Fuel Assemblies

U, Pu, & Minor

Actinides

U, Pu & Minor Actinides

Fuel Assemblies

Used Metal/Oxide Fuel

Fission Products & Trace Actinides

Fission Products& Trace Actinides

UsedOxide

Fuel

Mining &Enrichment

Mining &Enrichment





FastReactors

FastReactors


Mining &Enrichment

Mining &Enrichment



Recycling



FastReactors

FastReactors

Recycling


Depleted UF6Enriched UF6

Fuel Assemblies

U, Pu, & Minor

Actinides

U, Pu & Minor Actinides

Fuel Assemblies

Used Metal/Oxide Fuel

Fission Products & Trace Actinides

Fission Products& Trace Actinides

UsedOxide

Fuel

15


ciency. Several fuel forms are being considered for theirpotential to operate at very high temperatures and toensure an excellent retention of fission products: com-posite ceramic fuel, advanced fuel particles, or ceramicclad elements of actinide compounds. Core configura-tions are being considered based on pin- or plate-basedfuel assemblies or prismatic blocks.

The GFR system is top-ranked in sustainability becauseof its closed fuel cycle and excellent performance inactinide management. It is rated good in safety, eco-nomics, and in proliferation resistance and physicalprotection. It is primarily envisioned for missions inelectricity production and actinide management, al-though it may be able to also support hydrogen produc-tion. Given its R&D needs for fuel and recyclingtechnology development, the GFR is estimated to bedeployable by 2025.

LFR – Lead-Cooled Fast Reactor SystemThe Lead-Cooled Fast Reactor (LFR) system features afast-neutron spectrum and a closed fuel cycle for effi-cient conversion of fertile uranium and management ofactinides. A full actinide recycle fuel cycle with centralor regional fuel cycle facilities is envisioned. Thesystem uses a lead or lead/bismuth eutectic liquid-metal-cooled reactor. Options include a range of plant ratings,including a battery of 50–150 MWe that features a verylong refueling interval, a modular system rated at 300–400 MWe, and a large monolithic plant option at 1200MWe. The term battery refers to the long-life, factory-fabricated core, not to any provision for electrochemicalenergy conversion. The fuel is metal or nitride-based,containing fertile uranium and transuranics. The mostadvanced of these is the Pb/Bi battery, which employs asmall size core with a very long (10–30 year) core life.The reactor module is designed to be factory-fabricatedand then transported to the plant site. The reactor iscooled by natural convection and sized between 120–400MWth, with a reactor outlet coolant temperature of550°C, possibly ranging up to 800°C, depending uponthe success of the materials R&D. The system isspecifically designed for distributed generation ofelectricity and other energy products, including hydro-gen and potable water.

The LFR system is top-ranked in sustainability becausea closed fuel cycle is used, and in proliferation resistanceand physical protection because it employs a long-lifecore. It is rated good in safety and economics. Thesafety is enhanced by the choice of a relatively inertcoolant. It is primarily envisioned for missions inelectricity and hydrogen production and actinide man-

agement with good proliferation resistance. Given itsR&D needs for fuel, materials, and corrosion control,the LFR system is estimated to be deployable by 2025.

MSR – Molten Salt Reactor SystemThe Molten Salt Reactor (MSR) system features anepithermal to thermal neutron spectrum and a closedfuel cycle tailored to the efficient utilization of pluto-nium and minor actinides. A full actinide recycle fuelcycle is envisioned. In the MSR system, the fuel is acirculating liquid mixture of sodium, zirconium, anduranium fluorides. The molten salt fuel flows throughgraphite core channels, producing a thermal spectrum.The heat generated in the molten salt is transferred to asecondary coolant system through an intermediate heatexchanger, and then through another heat exchanger tothe power conversion system. Actinides and mostfission products form fluorides in the liquid coolant.The homogenous liquid fuel allows addition of actinidefeeds with variable composition by varying the rate offeed addition. There is no need for fuel fabrication. Thereference plant has a power level of 1000 MWe. Thesystem operates at low pressure (<0.5 MPa) and has acoolant outlet temperature above 700°C, affordingimproved thermal efficiency.

The MSR system is top-ranked in sustainability becauseof its closed fuel cycle and excellent performance inwaste burndown. It is rated good in safety, and inproliferation resistance and physical protection, and it israted neutral in economics because of its large numberof subsystems. It is primarily envisioned for missions inelectricity production and waste burndown. Given itsR&D needs for system development, the MSR is esti-mated to be deployable by 2025.

SFR – Sodium-Cooled Fast Reactor SystemThe Sodium-Cooled Fast Reactor (SFR) system featuresa fast-neutron spectrum and a closed fuel cycle forefficient conversion of fertile uranium and managementof actinides. A full actinide recycle fuel cycle is envi-sioned with two major options: One is an intermediatesize (150 to 500 MWe) sodium-cooled reactor with auranium-plutonium-minor-actinide-zirconium metalalloy fuel, supported by a fuel cycle based on pyrometal-lurgical processing in collocated facilities. The secondis a medium to large (500 to 1500 MWe) sodium-cooledfast reactor with mixed uranium-plutonium oxide fuel,supported by a fuel cycle based upon advanced aqueousprocessing at a central location serving a number ofreactors. The outlet temperature is approximately 550°Cfor both. The primary focus of the R&D is on the

16


recycle technology, economics of the overall system,assurance of passive safety, and accommodation ofbounding events.

The SFR system is top-ranked in sustainability becauseof its closed fuel cycle and excellent potential foractinide management, including resource extension. It israted good in safety, economics, and proliferationresistance and physical protection. It is primarilyenvisioned for missions in electricity production andactinide management. The SFR system is the nearest-term actinide management system. Based on the experi-ence with oxide fuel, this option is estimated to bedeployable by 2015.

SCWR – Supercritical-Water-Cooled ReactorSystemThe Supercritical-Water-Cooled Reactor (SCWR)system features two fuel cycle options: the first is anopen cycle with a thermal neutron spectrum reactor; thesecond is a closed cycle with a fast-neutron spectrumreactor and full actinide recycle. Both options use ahigh-temperature, high-pressure, water-cooled reactorthat operates above the thermodynamic critical point ofwater (22.1 MPa, 374°C) to achieve a thermal efficiencyapproaching 44%. The fuel cycle for the thermal optionis a once-through uranium cycle. The fast-spectrumoption uses central fuel cycle facilities based on ad-vanced aqueous processing for actinide recycle. Thefast-spectrum option depends upon the materials’ R&Dsuccess to support a fast-spectrum reactor.

In either option, the reference plant has a 1700-MWepower level, an operating pressure of 25 MPa, and areactor outlet temperature of 550°C. Passive safetyfeatures similar to those of the simplified boiling waterreactor are incorporated. Owing to the low density ofsupercritical water, additional moderator is added tothermalize the core in the thermal option. Note that thebalance-of-plant is considerably simplified because thecoolant does not change phase in the reactor.

The SCWR system is highly ranked in economicsbecause of the high thermal efficiency and plant simpli-fication. If the fast-spectrum option can be developed,the SCWR system will also be highly ranked insustainability. The SCWR is rated good in safety, and inproliferation resistance and physical protection. TheSCWR system is primarily envisioned for missions inelectricity production, with an option for actinidemanagement. Given its R&D needs in materials com-patibility, the SCWR system is estimated to bedeployable by 2025.

VHTR – Very-High-Temperature Reactor SystemThe Very-High-Temperature Reactor (VHTR) systemuses a thermal neutron spectrum and a once-throughuranium cycle. The VHTR system is primarily aimed atrelatively faster deployment of a system for high-temperature process heat applications, such as coalgasification and thermochemical hydrogen production,with superior efficiency.

The reference reactor concept has a 600-MWth helium-cooled core based on either the prismatic block fuel ofthe Gas Turbine–Modular Helium Reactor (GT-MHR) orthe pebble fuel of the Pebble Bed Modular Reactor(PBMR). The primary circuit is connected to a steamreformer/steam generator to deliver process heat. TheVHTR system has coolant outlet temperatures above1000°C. It is intended to be a high-efficiency systemthat can supply process heat to a broad spectrum of high-temperature and energy-intensive, nonelectric processes.The system may incorporate electricity generationequipment to meet cogeneration needs. The system alsohas the flexibility to adopt U/Pu fuel cycles and offerenhanced waste minimization. The VHTR requiressignificant advances in fuel performance and high-temperature materials, but could benefit from many ofthe developments proposed for earlier prismatic orpebble bed gas-cooled reactors. Additional technologyR&D for the VHTR includes high-temperature alloys,fiber-reinforced ceramics or composite materials, andzirconium-carbide fuel coatings.

The VHTR system is highly ranked in economicsbecause of its high hydrogen production efficiency, andin safety and reliability because of the inherent safetyfeatures of the fuel and reactor. It is rated good inproliferation resistance and physical protection, andneutral in sustainability because of its open fuel cycle. Itis primarily envisioned for missions in hydrogen produc-tion and other process-heat applications, although itcould produce electricity as well. The VHTR system isthe nearest-term hydrogen production system, estimatedto be deployable by 2020.

Missions and Economics for Generation IVWhile the evaluations of systems for their potential tomeet all goals were a central focus of the roadmapparticipants, it was recognized that countries would havevarious perspectives on their priority uses, or missions,for Generation IV systems. The following summary ofmissions resulted from a number of discussions by theGIF and the roadmap participants. The summary defines

17


three major mission interests for Generation IV: elec-tricity, hydrogen (or other nonelectricity products), andactinide management. The table below indicates themission focus of each of the six Generation IV systemswith regard to electricity and hydrogen.

Hydrogen Production, Cogeneration, and otherNonelectricity MissionsThis emerging mission requires nuclear systems that aredesigned to deliver other energy products based on thefission heat source, or which may deliver a combinationof process heat and electricity. Either may serve largegrids, or small isolated grids, or stand alone. Theprocess heat is delivered at sufficiently high tempera-tures (likely needed to be greater than 700°C) to supportsteam-reforming or thermochemical production ofhydrogen, as well as other chemical production pro-cesses. These applications can use the high temperatureheat or the lower temperature heat rejected from thesystem. Application to desalination for potable waterproduction may be an important use for the rejected heat.

In the case of cogeneration systems, the reactor providesall thermal and electrical needs of the production park.The distinguishing characteristic for this mission is thehigh temperature at which the heat is delivered. Besidesbeing economically competitive, the systems designedfor this mission would need to satisfy stringent standardsof safety, proliferation resistance, physical protection,and product quality.

For this mission, systems may again be designed toemploy either an open or closed fuel cycle, and they mayultimately be symbiotically deployed to optimizeeconomics and sustainability.

Actinide ManagementActinide management is a mission with significantsocietal benefits—nuclear waste consumption and long-term assurance of fuel availability. This mission over-laps an area that is typically a national responsibility,namely the disposition of spent nuclear fuel and high-level waste. Although Generation IV systems foractinide management aim to generate electricity eco-nomically, the market environment for these systems isnot yet well defined, and their required economicperformance in the near term will likely be determinedby the governments that deploy them. The table belowindicates that most Generation IV systems are aimed atactinide management, with the exception of the VHTR.

Electricity GenerationThe traditional mission for civilian nuclear systems hasbeen generation of electricity, and several evolutionarysystems with improved economics and safety are likelyin the near future to continue fulfilling this mission. It isexpected that Generation IV systems designed for theelectricity mission will yield innovative improvements ineconomics and be very cost-competitive in a number ofmarket environments, while seeking further advances insafety, proliferation resistance, and physical protection.These Generation IV systems may operate with either anopen or closed fuel cycle that reduces high-level wastevolume and mass. Further, it may be beneficial todeploy these nearer- and longer-term systems symbioti-cally to optimize the economics and sustainability of theensemble. Within the electricity mission, two special-izations are needed:

Large Grids, Mature Infrastructure, DeregulatedMarket. These Generation IV systems are designed tocompete effectively with other means of electricityproduction in market environments with larger, stabledistribution grids; well-developed and experiencednuclear supply, service, and regulatory entities; and avariety of market conditions, including highly competi-tive deregulated or reformed markets.

Small Grids, Limited Nuclear Infrastructure. TheseGeneration IV systems are designed to be attractive inelectricity market environments characterized by small,sometimes isolated, grids and a limited nuclear regula-tory and supply/service infrastructure. These environ-ments might lack the capability to manufacture theirown fuel or to provide more than temporary storage ofused fuel.

Both

– GFR– LFR– MSR

– VHTR– SCWR– SFR

1000˚C500˚C Outlet Temperature

ElectricityProduction

HydrogenProduction

Either

– SCWR – GFR– LFR– MSR– SFR

– VHTR

Once-ThroughFuel Cycle

ActinideManagement

18


Note that the SCWR begins with a thermal neutronspectrum and once-through fuel cycle, but may ulti-mately be able to achieve a fast spectrum with recycle.

The mid-term (30–50 year) actinide managementmission consists primarily of limiting or reversing thebuildup of the inventory of spent nuclear fuel fromcurrent and near-term nuclear plants. By extractingactinides from spent fuel for irradiation and multiplerecycle in a closed fuel cycle, heavy long-livedradiotoxic constituents in the spent fuel are transmutedinto much shorter-lived or stable nuclides. Also, theintermediate-lived actinides that dominate repositoryheat management are transmuted.

In the longer term, the actinide management mission canbeneficially produce excess fissionable material for usein systems optimized for other energy missions. Be-cause of their ability to use recycled fuel and generateneeded fissile materials, systems fulfilling this missioncould be very naturally deployed in symbiosis withsystems for other missions. With closed fuel cycles, alarge expansion of global uranium enrichment isavoided.

Observations on EconomicsThe work of the Economics Crosscut Group is central tounderstanding the limitations and opportunities regard-ing economics in the roadmap. These are discussed inturn.

Many limitations to the evaluation of economics areapparent. Examples are the large uncertainty whenprojecting production and capital costs several decadesinto the future, the uncertainty stemming from theoutcome of R&D on innovative advances for a system,and even the inability to validate the detailed analysesprovided by advocates with a potential bias. As a result,the economics evaluations are very uncertain. Theystrive to indicate a general impression of the futurepotential, having weighed a large amount of information.Of course, all Generation IV systems will need to meetthe economic requirements of the investors. Because ofthis, researchers and designers will need to continuallyaddress system economics as the R&D proceeds. Theeconomic evaluations in the roadmap should be taken asa relative indicator of how much emphasis needs to beplaced on the improvement of economics throughcontinued R&D.

A major opportunity debated among the systems wasbetween the long-established industry trend of larger,monolithic plants that exploit economy of scale, versusthe possibility that smaller, modular plants may be able

to use factory fabrication to exploit economy of volume.The six Generation IV systems feature a range of sizes,as shown in the table below. While the EconomicsCrosscut Group evaluations could not resolve the debate,it underscored the need for crosscutting R&D into theissue of modular plant versus monolithic plant econom-ics and the market/financial conditions under whichthese different types of plants would be preferred.

Safety, Safeguards, and Public Confidencein Generation IVOf all the goal areas, those regarding safety of nuclearenergy systems, protection of nuclear materials andfacilities within the system against acts of terrorism, andnuclear proliferation are most closely linked to publicconfidence in nuclear energy. The roadmap evaluationsof the safety and reliability goals indicated that theselected systems offer significant potential for advances.Most employ passive and active design features to helpavoid accidents in the first place, reduce reliance onoperator action, and mitigate the consequences ofpotential accidents.

While various means to enhance proliferation resistanceand physical protection are implemented in the systems,a standard methodology for their evaluation is not yetdeveloped. A major recommendation of the roadmap isthat R&D in this goal area should be focused on devel-oping a more comprehensive evaluation methodology.This will allow Generation IV systems to optimize theiruse of intrinsic barriers and extrinsic safeguards in thecourse of their development. Public confidence willincrease with enhanced proliferation resistance andphysical protection.

Near-Term Deployment Opportunities andGeneration IVWhile the Generation IV roadmap defined the long-termobjectives and needed R&D on innovative systems,efforts have been underway to define actions for nearer-

SmallModularMid-size

– LFR*

* Range of options

– LFR*– MSR– SFR*– SCWR

– GFR– VHTR– SFR*

LargeMonolithic

19


term deployment of evolutionary nuclear plants. Tobetter appreciate the relationship, the technologyroadmap identified a number of nearer-term systems thatcould have a benefit to the development of GenerationIV systems. These activities are described in turn.

United States Near-Term DeploymentIn the United States, the DOE’s independent NuclearEnergy Research Advisory Committee conducted astudy to identify the actions needed by government andindustry to overcome the technical and regulatorybarriers to new plant construction by 2010. The resultsof this study were documented in the October 2001report titled, A Roadmap to Deploy New Nuclear PowerPlants in the United States by 2010. Eight candidatereactor designs were evaluated with respect to sixcommercialization and regulatory readiness criteria,including advanced boiling water reactors, pressurizedwater reactors, and gas-cooled reactors. Six designswere found to be at least possibly deployable by 2010,provided that generating companies commit to placingnew plant orders by 2003. The list of U.S. Near-TermDeployment (NTD) options areshown in the table with acro-nyms or trade names below:

• ABWR (Advanced BoilingWater Reactor)

• AP1000 (Advanced Pressur-ized Water Reactor 1000)

• ESBWR (European Simpli-fied Boiling Water Reactor)

• GT-MHR (Gas Turbine–Modular High TemperatureReactor)

• PBMR (Pebble Bed ModularReactor)

• SWR-1000 (Siedewasser Reactor-1000).

The recommendations for action involved industry/government collaboration and cost-sharing on genericand plant-specific initiatives in the areas of (1) exercis-ing the new plant regulatory approval process in theUnited States, and (2) completing detailed engineeringand design work for at least one advanced reactor designin each of the water and gas reactor tracks. To accom-plish these tasks, DOE announced in February 2002 itsNuclear Power 2010 initiative, which focuses on deploy-ment of new plants in the United States over the next tenyears.

International Near-Term DeploymentThe Generation IV roadmap effort also identified otherdesigns that could be deployed in the nearer term. TheGIF expressed a strong interest in recognizing thesereactor designs as having this potential. Accordingly,the GIF created a distinct group known as InternationalNear-Term Deployment (INTD), and adopted twocriteria for systems to be included. First, recognizingthe difficulty of deployment by 2010, the GIF decided touse a somewhat later international deployment date of2015 for designs having significant industrial sponsor-ship. Second, the GIF decided to include only thosesystems whose performance is equal to or better than alight water reactor performance baseline representativeof Generation III. The baseline included performancemeasures in the four goal areas. While not described indetail here, they generally represent the Advanced LightWater Reactors (ALWRs) that have been built recently.Beginning with the May 2002 meeting, and working upto the July 2002 meeting, the GIF finalized a list ofsystems to be recognized as INTD designs.

Sixteen designs were found to be probably deployableby 2015 or earlier, and to be equal to or better than theALWR performance baseline. These are shown in thetable below with acronyms or trade names:

Advanced Boiling Water Reactors

• ABWR II (Advanced BoilingWater Reactor II)

• ESBWR (European Simpli-fied Boiling Water Reactor)

• HC-BWR (High ConversionBoiling Water Reactor)

• SWR-1000 (SiedewasserReactor-1000)

Advanced Pressure Tube Reactor

• ACR-700 (AdvancedCANDU Reactor 700)

Advanced Pressurized WaterReactors



• APR1400 (Advanced PowerReactor 1400)

U.S. Near-TermDeployment

(by 2010)

ABWRAP1000ESBWRGT-MHRPBMRSWR-1000

ABWR IIACR-700AP600AP1000APR1400APWR+CAREMEPRESBWRGT-MHRHC-BWRIMRIRISPBMRSMARTSWR-1000

InternationalNear-Term

Deployment(by 2015)

20


• APWR+ (Advanced Pressurized Water Reactor Plus)

• EPR (European Pressurized Water Reactor)

Integral Primary System Reactors

• CAREM (Central Argentina de ElementosModulares)

• IMR (International Modular Reactor)

• IRIS (International Reactor Innovative and Secure)

• SMART (System-Integrated Modular AdvancedReactor)

Modular High Temperature Gas-Cooled Reactors

• GT-MHR (Gas Turbine-Modular High TemperatureReactor)

• PBMR (Pebble Bed Modular Reactor)

Most INTD candidates have R&D needs to address onthe way toward possible deployment. Where the Gen-eration IV roadmap identifies the R&D needs for theselected Generation IV systems, some of the near-termcandidates have similar R&D needs in these areas.Therefore, it is important to recognize that the advance-ment of some candidates could make a beneficialcontribution to the technology development

Generation IV DeploymentThe objective for Generation IV nuclear energy systemsis to have them available for wide-scale deploymentbefore the year 2030. The best-case deployment datesanticipated for the six Generation IV systems are shownin the table to the right, and the dates extend further outthan those for near-term deployment. These datesassume that considerable resources are applied to theirR&D. The specific R&D activities are defined in recom-

mended R&D sections of this roadmap. The integrationand support of those activities is developed in moredetail in the Integration and Path Forward section at theend of this roadmap.

The Generation IV R&D activities are based on theassumption that not all near-term deployable systemswill be pursued by the private sector, but recognizes thatrelevant R&D on the near-term systems may have adirect benefit to the Generation IV program. That is,each one of the six systems has an R&D plan that iscomplete, but the R&D to be undertaken in GenerationIV may be reduced by technology development of arelevant INTD system that is deployed.

The Generation IV program will continually monitorindustry- and industry/government-sponsored R&Dplans and progress in order to benefit from them and notcreate duplicate efforts. Cases where industrial develop-ments are halted or merged may signal needed changesin the Generation IV R&D plans. Likewise, earlyGeneration IV R&D may hold significant advances fornear-term systems.

SFRVHTRGFRMSRSCWRLFR

201520202025202520252025

Generation IVSystem

Best CaseDeployment

Date

21


IntroductionThis section presents a survey of the recommendedsystem-specific R&D for the six Generation IV systems.If the research potentially applies to more than onesystem, it is presented in the next major section ascrosscutting.

The progression of R&D activities is divided intophases. The first is the viability phase, where theprincipal objective is to resolve key feasibility andproof-of-principle issues. The emphasis on the viabilityof the system is intended to yield answers before under-taking large-scale technology development. Earlyinteractions with regulators identifies high-level safetyrequirements. Decisions to proceed with the R&D focuson the feasibility of key technologies. The second phaseis the performance phase, where the key subsystems(such as the reactor, recycling facilities or energyconversion technology) need to be developed andoptimized. Continuing interactions with regulatorsadvances the level of understanding of the safety ap-proach. Decisions to proceed with the R&D now focuson the ability to make progress toward the desiredperformance levels. This phase ends when the system issufficiently mature and performs well enough to attractindustrial interest in large-scale demonstration of thetechnology.

The third phase is the demonstration phase, which has anumber of options as to the nature of the scope, size, andlength of time such a demonstration will have, as well asthe nature of the participation of industry, government,and even other countries in the project. Owing to thenew and innovative technology, it is felt that any Genera-tion IV system will need a demonstration phase. This isgenerally expected to require at least six years, possiblymore, and funding of several billion U.S. dollars. Withsuccessful demonstration, a system may enter a commer-cialization phase, which is an industry action.

The R&D presented in this section is limited to theviability and performance phases. Some recommenda-tions are also included regarding the type of project thatis envisioned to be appropriate for demonstration,although those activities are outside the scope of thetechnology roadmap.

As Generation IV systems advance, the evaluationmethodology will need to develop into broader and morecomprehensive tools for the assessment of the systems.Crosscutting R&D for evaluation methods is found inthe Crosscutting R&D sections on fuel cycles, risk andsafety, economics, and proliferation resistance andphysical protection. Of particular importance to allareas is developing the capability to quantify the uncer-tainty in the evaluations.

Schedules for the recommended R&D and associatedcost estimates are provided at the end of each system-specific R&D and crosscutting R&D sections. Thescope, schedule, and cost of R&D activities described inthe roadmap are conceptual and intended to address themost important of known viability and performanceissues that have been identified by the internationalworking groups. The costs have been estimated throughexpert judgment and comparison, and not throughrigorous program planning. The estimates assumerelatively successful and continuing R&D, and do notproject the effect of major program redirection fromsetbacks and failures. Very importantly, they do notinclude demonstration phase activities. In addition,costs for R&D facilities and infrastructure upgrades,such as the cost of a new materials test reactor, are notincluded.

Crosscutting R&D must be performed in addition to thesystem-specific R&D to support development of asystem. Thus, the complete cost for a system mustinclude an appropriate share of the crosscutting R&Dcosts.

The cost estimates provided in the roadmap are prima-rily for the purpose of comparing the recommendedsystems. They do not reflect the ongoing programs orfuture commitments of the GIF member countries.

The schedules are based on scenarios of successfuldeployment, with ample funding to achieve progress andwith a capable nuclear R&D infrastructure. The esti-mated costs and anticipated schedules for the viabilityand performance R&D are based on the collectivejudgment of the working groups. Large uncertaintiesexist in these costs and schedules. More detailedplanning will be required from the organizations per-forming the R&D.

RECOMMENDED R&D FOR THE MOST PROMISING SYSTEMS

22


Gas-Cooled Fast Reactor System R&DGFR DescriptionThe GFR system features a fast-spectrum helium-cooledreactor [shown below] and closed fuel cycle. Likethermal-spectrum helium-cooled reactors such as theGT-MHR and the PBMR, the high outlet temperature ofthe helium coolant makes it possible to deliver electric-ity, hydrogen, or process heat with high conversionefficiency. The GFR uses a direct-cycle helium turbinefor electricity and can use process heat for thermochemi-

cal production of hydrogen. Through the combination ofa fast-neutron spectrum and full recycle of actinides,GFRs minimize the production of long-lived radioactivewaste isotopes. The GFR’s fast spectrum also makes itpossible to utilize available fissile and fertile materials(including depleted uranium from enrichment plants)two orders of magnitude more efficiently than thermalspectrum gas reactors with once-through fuel cycles.The GFR reference assumes an integrated, on-site spentfuel treatment and refabrication plant.

23


A summary of design parameters for the GFR system isgiven in the following table.

• GFR fuel forms for the fast-neutron spectrum

• GFR core design, achieving a fast-neutron spectrumfor effective conversion with no fertile blankets

• GFR safety, including decay heat removal systemsthat address the significantly higher power density(in the range of 100 MWth/m3) and the reduction ofthe thermal inertia provided by graphite in themodular thermal reactor designs

• GFR fuel cycle technology, including simple andcompact spent-fuel treatment and refabrication forrecycling.

Performance issues for GFR include:

• Development of materials with superior resistance tofast-neutron fluence under very-high-temperatureconditions

• Development of a high-performance helium turbinefor efficient generation of electricity

• Development of efficient coupling technologies forprocess heat applications and the GFR’s hightemperature nuclear heat.

The GFR has several technology gaps in its primarysystems and balance of plant that are in common withthe GT-MHR. Also, the development of very-high-temperature materials with superior resistance to fast-neutron fluence and innovative refractory fuel conceptswith enhanced fission product retention capability are ofgeneric interest to other types of reactors, including theVHTR and water-cooled reactors.

Target values of some key parameters such as powerdensity and fuel burnup are sufficient for reasonableperformance of a first-generation new fuel technology.Because these parameters have a direct impact ontechnical and economical performance, there is strongincentive for additional performance phase R&D, withthe goal of further upgrading the power density tobeyond 100 MWth/m3 and the fuel burnup to the rangeof 15% FIMA.

GFR R&D ScopeAn R&D program is recommended to assess the viabil-ity of the GFR and conduct the performance R&Drequired for successful demonstration of the GFR. Thisdevelopment includes R&D on fuel, fuel cycle processes(treatment and refabrication), reactor systems, balanceof plant, and computer codes needed for design studiesand safety demonstration. A conceptual design of anentire GFR prototype system can be developed by 2019.The prototype system is envisioned as an internationalproject that could be placed in operation by 2025.

Reactor Parameters Reference ValueReactor power 600 MWthNet plant efficiency 48%(direct cycle helium)Coolant inlet/outlet 490°C/850°Ctemperature and pressure at 90 barAverage power density 100 MWth/m3Reference fuel compound UPuC/SiC (70/30%)

with about20% Pu content

Volume fraction, Fuel/Gas/SiC 50/40/10%Conversion ratio Self-sufficientBurnup, Damage 5% FIMA; 60 dpa

Technology Base for the GFRThe technology base for the GFR includes a number ofthermal spectrum gas reactor plants, as well as a fewfast-spectrum gas-cooled reactor designs. Past pilot anddemonstration projects include decommissioned reactorssuch as the Dragon Project, built and operated in theUnited Kingdom, the AVR and the THTR, built andoperated in Germany, and Peach Bottom and Fort StVrain, built and operated in the United States. Ongoingdemonstrations include the HTTR in Japan, whichreached full power (30 MWth) using fuel compacts in1999, and the HTR-10 in China, which may reach 10MWth in 2002 using pebble fuel. A 300-MWth pebblebed modular demonstration plant is being designed byPBMR Pty for deployment in South Africa, and aconsortium of Russian institutes is designing a 300-MWthGT-MHR in cooperation with General Atomics. Thedesign of the PBMR and GT-MHR reactor systems, fuel,and materials are evolutionary advances of the demon-strated technology, except for the direct Brayton-cyclehelium turbine and implementation of modularity in theplant design. The GFR may benefit from developmentof these technologies, as well as development of innova-tive fuel and very-high-temperature materials for theVHTR. A phased development path may be drawn fromthe thermal to the fast-spectrum gas-cooled systems.

Technology Gaps for the GFRDemonstrating the viability of the GFR requires meetinga number of significant technical challenges. Fuel, fuelcycle processes, and safety systems pose the majortechnology gaps:

24


GFR Fuels and Materials R&DCandidate Fuels. A composite ceramic-ceramic fuel(cercer) with closely packed, coated (U, Pu)C kernels orfibers is the best option for fuel development. Alterna-tive fuel options for development include fuel particleswith large (U, Pu)C kernels and thin coatings, or ce-ramic-clad, solid-solution metal (cermet) fuels. Theneed for a high density of heavy nuclei in the fuel leadsto actinide-carbides as the reference fuel and actinide-nitrides with 99.9% enriched nitrogen as the backup.

Initially, the research should focus on studying potentialcandidate fuels and evaluating their technical feasibilitybased on existing information on the structural integrityand radiation resiliency of the coating system and thechemical compatibility among the different materials forthe GFR service conditions (e.g., temperatures up to1400°C, burnup up to 250 GWD/MTHM, and radiationresiliency up to 100 to 150 dpa). This will lead to theestablishment of reference and backup options. Theseoptions will undergo a series of irradiation and high-temperature safety tests in concert with fuel modelingactivities to establish the performance of the fuel type.Irradiations range from small-scale experiments inexisting reactors to large-scale prototype fuel assembliesunder representative GFR conditions. The research isexpected to take nearly 20 years to complete.

Key dates are:

• 2002–2004 Acquisition of basic data on inertmaterials and actinide compounds and definition ofreference and backup fuel concepts

• 2005–2011 Irradiation testing in existing reactors

• 2012–2019 Irradiation of prototype fuel subassem-blies in GFR representative conditions.

Fuel fabrication techniques must be developed to becompatible with on-site processing for actinide recoveryand remote fuel fabrication. Innovative methods such asvapor deposition or impregnation are among the candi-date techniques for on-site manufacturing of compositeceramic fuel (cercer, with cermet as backup). For pin-type fuels, ceramic cladding capable of confining fissionproducts will be considered. Samples of irradiated fuelswill be used to test current and innovative fuel treatmentprocesses likely to be compatible with remote simpleand compact technologies for actinide spent fuel treat-ment and refabrication before recycling.

Candidate Materials. The main challenges are in-vessel structural materials, both in-core and out-of-core,that will have to withstand fast-neutron damage and high

temperatures, up to 1600°C in accident situations.Ceramic materials are therefore the reference option forin-core materials, and composite cermet structures orinter-metallic compounds will be considered as abackup. For out-of-core structures, metal alloys will bethe reference option.

The most promising ceramic materials for core struc-tures are carbides (preferred options are SiC, ZrC, TiC,NbC), nitrides (Zr N, TiN), and oxides (MgO, Zr(Y)O2).Inter-metallic compounds like Zr3Si2 are promisingcandidates as fast-neutron reflector materials. Limitedwork on Zr, V or Cr as the metallic part of the backupcermet option should also be undertaken.

For other internal core structures, mainly the upper andlower structures, shielding, the core barrel and gridplate, the gas duct shell, and the hot gas duct, the candi-date materials are coated or uncoated ferritic-martensiticsteels (or austenitic as alternative solution), other Fe-Ni-Cr-base alloys (Inco 800), and Ni-base alloys. The maincandidate materials for pressure vessels (reactor, energyconversion system) and cross vessel are 21/4 Cr and 9-12 Cr martensitic steels.

The recommended R&D activities include a screeningphase with material irradiation and characterization, aselection of a reference set of materials for core struc-tural materials, and then optimization and qualificationunder irradiation.

The program goal is to select the materials that offer thebest compromise regarding:

• Fabricability and welding capability

• Physical, neutronic, thermal, tensile, creep, fatigue,and toughness properties and their degradation underlow-to-moderate neutron flux and dose

• Microstructure and phase stability under irradiation

• Irradiation creep, in-pile creep, and swellingproperties

• Initial and in-pile compatibility with He (andimpurities).

Recommended R&D activities on out-of-core structuresconsists of screening, manufacturing, and characterizingmaterials for use in the pressure vessel, primary system,and components (pipes, blowers, valves, heat exchangers).

With respect to materials used for the balance of plant,the development program includes screening, manufac-turing, and characterizing heat-resisting alloys orcomposite materials for the Brayton turbomachinery

25


(turbine disk and fins), as well as for heat exchangers,including the recuperator of the Brayton cycle. Like-wise, in the case of nonelectricity energy products,materials development is required for the intermediateheat exchanger that serves to transfer high-temperatureheat in the helium coolant to the process heat applica-tions. R&D recommended for these systems is dis-cussed in the Crosscutting Energy Products R&Dsection.

GFR Reactor Systems R&DThe innovative GFR design features to be developedmust overcome shortcomings of past fast-spectrum gas-cooled designs, which were primarily low thermalinertia and poor heat removal capability at low heliumpressure. Various passive approaches will be evaluatedfor the ultimate removal of decay heat in depressuriza-tion events. The conditions to ensure a sufficient backpressure and to enhance the reliability of flow initiationare some of the key issues for natural convection, theefficiency of which will have to be evaluated for differ-ent fuel types, power densities, and power conversionunit. Dedicated systems, such as semipassive heavy gasinjectors, need to be evaluated and developed. There isalso a need to study the creation of conduction paths andvarious methods to increase fuel thermal inertia and,more generally, core capability to store heat whilemaintaining fuel temperature at an acceptable level.

GFR Balance-of-Plant R&DPerformance R&D is required for the high-temperaturehelium systems, specifically:

• Purification, control of inventory, and in-servicemonitoring of interactions between helium and thematerials it contacts

• Heat transfer and flow pattern through the core, thecircuits, and the heat exchangers

• Dynamics of the circuits and the structures, acous-tics of the cavities.

GFR Safety R&DBecause of the high GFR core power density, a safetyapproach is required that relies on intrinsic core proper-ties supplemented with additional safety devices andsystems as needed, but minimizes the need for activesystems. After in-depth studies have defined the safetycase, safety systems will be demonstrated experimen-tally. Transient fuel testing, of both the developmentaland confirmatory kind, will be conducted. Concurrently,model and code development are required to provide thebasis for the final safety case. An integrated safety

experiment, simulating the safety case of the GFR, willbe prepared. It is expected that the safety experimentswill require an integral helium loop on the order of 20MWth.

GFR Design and Evaluation R&DThe most important issues regarding economic viabilityof the GFR are associated with the simplified andintegrated fuel cycle, and the modularity of the reactor—this includes volume production, in-factory prefabrica-tion, and sharing of on-site resources.

The GFR design and safety analysis will require devel-opment of novel analysis tools capable of modeling thecore with its novel fuel and subassembly forms, unusualfuel composition, and novel safety devices. The analysistools must be validated to demonstrate with sufficientaccuracy the safe behavior of the entire system under alloperational conditions. This requires new neutronics,thermal-gas dynamics, operation, and safety models, orsignificant adaptations of existing codes. Validation ofthe models requires that critical experiments and subas-sembly mockup testing and possibly other qualificationexperiments be conducted.

GFR Fuel Cycle R&DThe range of fuel options for the GFR underscores theneed for early examination of their impacts on thesystem, especially its fuel cycle. Existing fuel cycletechnologies need to be further developed or adapted toallow for the recycling of actinides while preserving theeconomic competitiveness of the nuclear option in themedium and long term. Laboratory-scale processes fortreatment of carbide, nitride, or oxide dispersion fuels inceramic or metal matrices have been evaluated andappear technically feasible. However, extensive experi-mental work is required in order that the process con-cepts can be proven feasible for fuel treatment at pro-duction scale.

Compatibility of Fuel and Fuel Recycling TechnologyOptions. The capabilities of both advanced aqueous andpyrochemical processes for recycling the fuel optionsunder consideration will be assessed, while taking intoaccount the facility requirements associated with on-sitefuel conditioning and refabrication. R&D on the twooptions is discussed in the Crosscutting Fuel Cycle R&Dsection.

The objective for the GFR fuel cycle R&D is to seeksolutions for the separation of its unique materials of thematrices and coatings from actinide compounds that (1)develop the capability to treat cercer fuels, as well as

26


coated particle fuel or cermet as a backup, (2) minimizethe release of gaseous and liquid effluents to the envi-ronment, (3) take into account, starting at the designstage, the management of induced secondary waste fromtreatment and conditioning, (4) simplify the integrationof treatment and fuel manufacturing operations, and (5)allow for integrated in situ treatment.

Both aqueous and pyrochemical processing methods,and combinations of the two processes, will be tested onthe inert-matrix fuels. Hybrid processes may prove to besuperior in the long run. Candidate processes withreasonable expectations of technical feasibility need tobe compared in detail at the conceptual stage. Theevaluations will be based on mass-balance flow-sheetsand estimates of equipment and facility requirementsnecessary to meet established criteria for product qualityand throughput capacity.

Scale Up and Demonstration. An important phase ofthe R&D program will be to demonstrate, at the level ofseveral kilograms of the selected fuel, the treatment andrefabrication of irradiated fuel. The objective is to selectand demonstrate the scientific viability of a process bythe end of 2012. After process screening, mostly withsurrogate materials, more in-depth studies of the se-lected treatment process will be performed in hotlaboratories using irradiated fuel samples provided bythe irradiation program for fuel development. The finalphase of the development program will consist ofdemonstrating the technologies associated with the fuelcycle plant of the GFR prototype system.

GFR R&D Schedule and CostsA schedule for the GFR R&D is shown below, alongwith the R&D costs and decision points (starred).

GAS-COOLED FAST REACTOR SYSTEM (940 M$)Fuels and Materials (300 M$)

Reactor Systems (100 M$)

Balance of Plant (50 M$)

Safety (150 M$)

Design & Evaluation (120 M$)

Fuel Cycle (220 M$)

Core materials screening Core structural material down-selection decision (GFR 2) Core materials fabrication Core materials out-of-pile testing Structural material final selection (GFR 5) Core materials in-pile testing Fuel basic screening Fuel down-selection (GFR 1) Fuel tests

Screening and testing Materials and components He technology test benches Testing and 20 MWth He loop

Turbo machinery technology development Component development Coupling technology to process heat applications

Safety approach and evaluation Safety concept selection (GFR 3) System development and testing

Preconceptual design Viability phase complete Conceptual design Analysis tools

Screening Viability assessment Fuel system viability (GFR 4) Technology and performance testing

decision

decision

decision

decision

2000 2010 2020

27


Lead-Cooled Fast Reactor System R&DLFR DescriptionLFR systems are Pb or Pb-Bi alloy-cooled reactors witha fast-neutron spectrum and closed fuel cycle. One LFRsystem is shown below. Options include a range of plantratings, including a long refueling interval batteryranging from 50–150 MWe, a modular system from300–400 MWe, and a large monolithic plant at 1200MWe. These options also provide a range of energyproducts.

The LFR battery option is a small factory-built turnkeyplant operating on a closed fuel cycle with very longrefueling interval (15 to 20 years) cassette core orreplaceable reactor module. Its features are designed tomeet market opportunities for electricity production onsmall grids, and for developing countries that may notwish to deploy an indigenous fuel cycle infrastructure tosupport their nuclear energy systems. Its small size,

reduced cost, and full support fuel cycle services can beattractive for these markets. It had the highest evalua-tions to the Generation IV goals among the LFR options,but also the largest R&D needs and longest developmenttime.

The options in the LFR class may provide a time-phaseddevelopment path: The nearer-term options focus onelectricity production and rely on more easily developedfuel, clad, and coolant combinations and their associatedfuel recycle and refabrication technologies. The longer-term option seeks to further exploit the inherently safeproperties of Pb and raise the coolant outlet temperaturesufficiently high to enter markets for hydrogen andprocess heat, possibly as merchant plants. LFR holdsthe potential for advances compared to state-of-the-artliquid metal fast reactors in the following:

• Innovations in heat transport and energy conversionare a central feature of the LFR options. Innovationsin heat transport are afforded by natural circulation,

28


• The favorable properties of Pb coolant and nitridefuel, combined with high temperature structuralmaterials, can extend the reactor coolant outlettemperature into the 750–800ºC range in the longterm, which is potentially suitable for hydrogenmanufacture and other process heat applications. Inthis option, the Bi alloying agent is eliminated, andthe less corrosive properties of Pb help to enable theuse of new high-temperature materials. The re-quired R&D is more extensive than that required forthe 550ºC options because the higher reactor outlettemperature requires new structural materials andnitride fuel development.

A summary of the design parameters for the LFRsystems is given in the following table.

Technology Base for the LFRThe technologies employed are extensions of thosecurrently available from the Russian Alpha class subma-rine Pb-Bi alloy-cooled reactors, from the Integral FastReactor metal alloy fuel recycle and refabricationdevelopment, and from the ALMR passive safety andmodular design approach. Existing ferritic stainlesssteel and metal alloy fuel, which are already signifi-cantly developed for sodium fast reactors, are adaptableto Pb-Bi cooled reactors at reactor outlet temperatures of550ºC.

lift pumps, in-vessel steam generators, and otherfeatures. Innovations in energy conversion areafforded by rising to higher temperatures than liquidsodium allows, and by reaching beyond the tradi-tional superheated Rankine steam cycle tosupercritical Brayton or Rankine cycles or processheat applications such as hydrogen production anddesalination.

• The favorable neutronics of Pb and Pb-Bi coolantsin the battery option enable low power density,natural circulation-cooled reactors with fissile self-sufficient core designs that hold their reactivity overtheir very long 15- to 20-year refueling interval. Formodular and large units more conventional higherpower density, forced circulation, and shorterrefueling intervals are used, but these units benefitfrom the improved heat transport and energy conver-sion technology.

• Plants with increased inherent safety and a closedfuel cycle can be achieved in the near- to mid-term.The longer-term option is intended for hydrogenproduction while still retaining the inherent safetyfeatures and controllability advantages of a heattransport circuit with large thermal inertia and acoolant that remains at ambient pressure. Thefavorable sustainability features of fast spectrumreactors with closed fuel cycles are also retained inall options.

Reference ValuePb-Bi Battery Pb-Bi Module Pb Large Pb Battery

Reactor Parameters (nearer-term) (nearer-term) (nearer-term) (far-term)

Coolant Pb-Bi Pb-Bi Pb PbOutlet Temperature (ºC) ~550 ~550 ~550 750–800Pressure (Atmospheres) 1 1 1 1Rating (MWth) 125–400 ~1000 3600 400Fuel Metal Alloy Metal Alloy Nitride Nitride

or NitrideCladding Ferritic Ferritic Ferritic Ceramic coatings

or refractory alloysAverage Burnup ~100 ~100–150 100–150 100(GWD/MTHM)Conversion Ratio 1.0 d>–1.0 1.0–1.02 1.0Lattice Open Open Mixed OpenPrimary Flow Natural Forced Forced NaturalPin Linear Heat Rate Derated Nominal Nominal Drated

29


Technology Gaps for the LFRThe important LFR technology gaps are in the areas of:

• LFR system fuels and materials, with some gapsremaining for the 550ºC options, and large gaps forthe 750–800ºC option, including:

– Nitride fuels development, including fuel/cladcompatibility and performance

– High-temperature structural materials

– Environmental issues with lead.

• LFR system design, including:

– Open lattice heat removal, both forced, andnatural convective

– Neutronic data and analysis tools

– Coolant chemistry control, especially oxygenand 210Po control

– Innovative heat transport methods (such asdesign for natural circulation, lift pumps, in-vessel steam generators)

– Core internals support and refueling machinery

– Seismic isolation.

• LFR balance of plant, adapting supercritical steamRankine or developing supercritical CO2 electricityproduction technology, and crosscutting R&D onhydrogen production technology and heat exchang-ers for process heat applications

• LFR economics, focusing on modularization andfactory fabrication

• LFR fuel cycle technology, including remote fabri-cation of metal alloy and TRU-N fuels.

Important viability and performance issues are found inall areas. Important R&D areas for each option areindicated in the table below.

International economic and regulatory developments arealso needed for the cases where new regional fuel cyclecenters owned by a consortium of clients operatingunder international safeguards close the fuel cycle andmanage the waste.

Major R&D Areas Pb-Bi Battery Pb-Bi Module Pb Large Pb Batter(nearer-term) (nearer-term) (nearer-term) (far-term)

Metal Alloy or Nitride Fuel x x x x(esp. for higher temperature range)High-Temperature Structural Materials xNatural Circulation Heat Transport x x x xin Open LatticeForced Circulation Heat Transport x x x xin Open LatticeCoolant Chemistry Control x x x xInnovative Heat Transport x x x xInternals Support and Refueling x x x xEnergy Conversion:

Supercritical CO2 Brayton x x xSupercritical Water Rankine x xCa-Br Water Cracking xDesalinization Bottoming x x x

Economics:Modularization x x x xModularization & Site Assembly x x x x

Metal Fuel Recycle/Refabrication x xNitride Fuel Recycle/Refabrication x x x x

30


LFR Fuels and Materials R&DThe nearer-term options use metal alloy fuel, or nitridefuel if available. Metal alloy fuel pin performance at550ºC and U/TRU/Zr metal alloy recycle and remoterefabrication technologies are substantially developedalready in Na-cooled systems. Metal alloy fuel andrecycle R&D is discussed in detail in the SFR andCrosscutting Fuel Cycle R&D sections, respectively.

Nitride Fuel. Mixed nitride fuel is also possible for the550ºC options; however, it is clearly required for thehigher-temperature option. New fuel development willrequire a long R&D period, which should begin immedi-ately. It is estimated that 10–15 years will be necessaryto qualify any new fuel for the long-life service condi-tions in Pb or Pb-Bi. During the viability phase, R&Dwill be limited to finding a suitable cladding, developinga property-base for the nitride fuel, and preliminary in-pile testing.

Materials Screening. The top priority viability R&Dareas for higher-temperature starts with materialsscreening for cladding, reactor internals, and heatexchangers. The primary approach will be to adaptmodern materials developments such as composites,coatings, ceramics, and high-temperature alloys fromother fields such as aerospace, and gas turbines. Thegoal is not only long service life but also cost effectivefabrication using modern forming and joining technologies.

For the cladding, compatibility with Pb or Pb-Bi on thecoolant side and mixed nitride fuel on the fuel side isrequired, and radiation damage resistance in a fast-neutron environment is required for a 15–20 yearirradiation period. SiC or ZrN composites or coatingsand refractory alloys are potential options for 800ºCservice, while standard ferritic steel is adequate at550ºC.

For process heat applications, an intermediate heattransport loop is needed to isolate the reactor from theenergy converter for both safety assurance and productpurity. Heat exchanger materials screening is needed for

potential intermediate loop fluids, including moltensalts, He, CO2, and steam. For interfacing with thermo-chemical water cracking, the chemical plant fluid is HBrplus steam at 750ºC and low pressure. For interfacingwith turbomachinery, the working fluid options aresupercritical CO2 or superheated or supercritical steam.

The material screening R&D will take the majority ofthe viability R&D time period and will require corrosionloops, posttest examination equipment, properties testingapparatus, phase diagram development, coolant chemis-try control R&D, fabricability evaluations, and static andflowing in situ irradiation testing.

LFR Reactor Systems R&DChemistry Control. Viability R&D is also needed forchemistry and activation control of the coolant andcorrosion products. Means for oxygen control areneeded for both Pb and Pb-Bi options. Strategies andmeans for control of 210Po, an activation product of Bi,is needed for the Pb-Bi option.

Thermal hydraulics. The heat removal from the fuelpin lattice (and also across intermediate heat exchangertube bundles) uses natural or low-speed forced circula-tion through an open lattice of ductless assemblies. Heattransfer correlations, pressure drop correlations, pressuredrop form factors for plenum flows and transitions, andflow redistribution patterns need to be developed as afunction of geometry and pin linear heat rate both in thelattice and in the overall reactor flow circuit. The effectsof grid spacers, deposits, and clad aging will have to beunderstood to support the long-term viability of naturalcirculation. This requires the availability of loops with aheight useful for natural circulation, and also large-scaleplenum flow facilities.

Neutronics. Neutronic data and computer codes alsoneed to be validated through comparison of calculatedneutronic parameters with measurements from criticalexperiment facilities. The need for improved evalua-tions of lead and bismuth cross sections should beassessed.

31


Reactor Components. Reactor internals supporttechniques and refueling, core positioning, and clampingstrategies are issues because the internals and the fuelwill float (unless restrained) in the dense coolant. In-service inspection technologies have to be developed.

LFR Balance-of-Plant R&DR&D activity is recommended to support the LFRbalance of plant in the areas of Ca-Br water cracking forhydrogen production, and a supercritical CO2 Braytoncycle for energy conversion. These activities are foundin the Crosscutting Energy Products R&D section.

LFR Safety R&DThe assurance of reliable and effective thermostructuralreactivity feedback is key to the passive safety/passiveload following design strategy and will require coordi-nated neutronics/thermal-hydraulics/structural design ofthe core. Preliminary testing of mixed nitride fuel undersevere upset incore temperature conditions should alsobe conducted.

LFR Design and Evaluation R&DEconomics. Viability R&D activities are needed todetermine whether economies can be achieved by plantsimplification and reduced footprint, which is affordedby (1) the coolants being inert in air and water, (2) thehigh conversion efficiency using Brayton cycles orsupercritical steam cycles, (3) the economies of massproduction, modular assembly, and short onsite con-struction startup time, and (4) the production of energyproducts, possibly including the use of waste heat in abottoming cycle.

Modular Construction. Achieving successful econom-ics in the battery and modular options will depend onadaptation of factory-based mass production techniquesfrom industries such as airplane, truck, and auto manu-

facture, and adaptation of modular/rapid site assemblyused for ocean oil rig emplacement and shipbuilding.Life-cycle integrated economics analysis will also beneeded that can address modern techniques in design,fabrication, transport, installation and startup, andmonitoring and maintenance.

Plant Structures. The structural support of the reactorvessel, containing dense Pb or Pb-Bi coolant, willrequire design development in seismic isolation ap-proaches and sloshing suppression. Also, concretesupports, if used, will have to either be cooled or bedesigned for high temperature service.

LFR Fuel Cycle R&DThe preferred option for the LFR fuel cycle ispyroprocessing, with advanced aqueous as an alterna-tive. R&D recommended to generally develop thepyroprocess is found in the Crosscutting Fuel CycleR&D section, although specialization is required tosupport the nitride fuel.

Nitride Fuel Recycle. Specialization anticipated formixed nitride fuel recycle will need to address separa-tions technology, remote refabrication technology, 15Nenrichment technologies, and irradiation testing. Re-cycle and remote refabrication R&D activity in theviability phase should involve an iterative screening ofconceptual recycle and refabrication approaches, benchscale testing, and flow sheet refinements. This work willbuild on existing programs in Japan and Europe, whichare directed to partitioning and transmutation missions.Since 15N enrichment is essential to meetingsustainability goals for waste management (arising fromthe need to control 14C production), fuel cycle R&Dactivity should screen options for 15N enrichment andrecovery and associated bench-scale investigations.

32


LFR R&D Schedule and CostsA schedule for the LFR R&D is shown below, alongwith the R&D costs and decision points.

LEAD-COOLED FAST REACTOR SYSTEM (990 M$)Fuels and Materials (250 M$)



Safety (150 M$)


Fuel Cycle (190 M$)

Ferritic steel out-of-pile corrosion Pb-BiCoolant chemistry monitoring and controlFerritic steel in-pile test in flowing loopScreen materials for higher tempStructural material selection for 550ºC coolant outlet temperature (LFR 1)Develop and evaluate fabrication technologyNitride fuel fabrication approach ( 2)Develop thermophysical propertiesOut-of-pile and drop-in testIn-pile testFeasibility/selection of structural material for 800ºC lead ( 5)Mixed nitride fuel fabricationNitride fuel propertiesIn-pile irradiation testing of nitride fuelAdequacy of nitride fuel performance potential ( 6)

Natural circulation heat transportRefueling approachMaintenance/ISIR technologyNeutronic critical experiments and evaluation

Supercritical CO2 Brayton cycle (R&D/Test)Feasibility of supercritical CO2 Brayton cycle ( 8)IHX development for coupling to H2 productionCa-Br water splittingFeasibility of Ca-Br H2 production ( 7)

SG or IHX tube rupture tests and analysesSeismic isolation development

Modularization/factory fabricationModular installationPreconceptual designViability phase completeConceptual designAnalysis toolsFeasibility of reactor transportFeasibility of transportable reactor/core cartridge ( 3)

N15 enrichment technologyPyro recycle development for nitrideNitride fuel recycle approach (pyro vs. aqueous) ( 4)Advanced aqueous development for nitride

decision

decision LFR

decision LFR

decision LFR

decision LFR

decision LFR

decision LFR

decision LFR

2000 2010 2020

33


Molten Salt Reactor System R&DMSR DescriptionThe MSR produces fission power in a circulating moltensalt fuel mixture [an MSR is shown below]. MSRs arefueled with uranium or plutonium fluorides dissolved ina mixture of molten fluorides, with Na and Zr fluoridesas the primary option. MSRs have the following uniquecharacteristics, which may afford advances:

• MSRs have good neutron economy, opening alterna-tives for actinide burning and/or high conversion

• High-temperature operation holds the potential forthermochemical hydrogen production

• Molten fluoride salts have a very low vapor pres-sure, reducing stresses on the vessel and piping

• Inherent safety is afforded by fail-safe drainage,passive cooling, and a low inventory of volatilefission products in the fuel

• Refueling, processing, and fission product removalcan be performed online, potentially yielding highavailability

• MSRs allow the addition of actinide feeds of widelyvarying composition to the homogenous salt solutionwithout the blending and fabrication needed by solidfuel reactors.

34


There are four fuel cycle options: (1) Maximum conver-sion ratio (up to 1.07) using a Th-233U fuel cycle, (2)denatured Th-233U converter with minimum inventory ofnuclear material suitable for weapons use, (3) denaturedonce-through actinide burning (Pu and minor actinides)fuel cycle with minimum chemical processing, and (4)actinide burning with continuous recycling. The fourthoption with electricity production is favored for theGeneration IV MSR. Fluoride salts with higher solubil-ity for actinides such as NaF/ZrF4 are preferred for thisoption. Salts with lower potential for tritium productionwould be preferred if hydrogen production were theobjective. Lithium and beryllium fluorides would bepreferred if high conversion were the objective. Onlineprocessing of the liquid fuel is only required for highconversion to avoid parasitic neutron loses of 233Pa thatdecays to 233U fuel. Offline fuel salt processing isacceptable for actinide management and hydrogen orelectricity generation missions. To achieve conversionratios similar to LWRs, the fuel salt needs only to bereplaced every few years.

The reactor can use 238U or 232Th as a fertile fuel dis-solved as fluorides in the molten salt. Due to thethermal or epithermal spectrum of the fluoride MSR,232Th achieves the highest conversion factors. All of theMSRs may be started using low-enriched uranium orother fissile materials. The range of operating tempera-tures of MSRs ranges from the melting point of eutecticfluorine salts (about 450°C) to below the chemicalcompatibility temperature of nickel-based alloys (about800°C).

A summary of the reference design parameters for theMSR is given in the following table.

Technology Base for the MSRMSRs were first developed in the late 1940s and 1950sfor aircraft propulsion. The Aircraft Reactor Experiment(ARE) in 1954 demonstrated high temperatures (815°C)and established benchmarks in performance for acirculating fluoride molten salt (NaF/ZrF4) system. The8 MWth Molten Salt Reactor Experiment (MSRE)demonstrated many features, including (1) a lithium/beryllium fluoride salt, (2) graphite moderator, (3) stableperformance, (4) off-gas systems, and (5) use of differ-ent fuels, including 235U, 233U, and plutonium. Adetailed 1000 MWe engineering conceptual design of amolten salt reactor was developed. Under these pro-grams, many issues relating to the operation of MSRs aswell as the stability of molten salt fuel and its compat-ibility with graphite and Hastelloy N were resolved.

Technology Gaps for the MSRThe MSR has a number of technical viability issues thatneed to be resolved. The highest priority issues includemolten salt chemistry, solubility of actinides and lan-thanides in the fuel, compatibility of irradiated moltensalt fuel with structural materials and graphite, and metalclustering in heat exchangers. Specific areas of thisviability research phase include:

• Solubility of minor actinides and lanthanides inmolten fluoride salt fuel for actinide managementwith high actinide concentrations

• Lifetime behavior of the molten salt fuel chemistry,and fuel processing during operation and eventualdisposal in a final waste form

• Materials compatibility with both fresh andirradiated molten salt fuel for higher temperatureapplications

• Metal clustering (noble metals plate-out on of theheat exchanger primary wall)

• Salt processing, separation, and reprocessingtechnology development, including a simplificationof the flowsheet.

The initial viability R&D phase is complemented bystudies to establish conceptual design and preliminarytechnical specifications for the reactor and powergeneration cycle.

The issues in the performance R&D phase include:

• Fuel development, new cross section data, andqualifications to enable selection of the molten saltcomposition

Reactor Parameters Reference Value

Net power 1000 MWePower density 22 MWth/m3

Net thermal efficiency 44 to 50%Fuel-salt – inlet temperature 565°C

– outlet temperature 700°C (850°C forhydrogen production)

– vapor pressure <0.1 psiModerator GraphitePower Cycle Multi-reheat

recuperative heliumBrayton cycle

Neutron spectrum burner Thermal–actinide

35


• Corrosion and embrittlement studies to determinelifetimes of materials and reliability

• Development of tritium control technology

• Molten salt chemistry control, REDOX control,liquid-liquid extraction, and salt purification

• Graphite sealing technology and graphite stabilityimprovement and testing

• Detailed conceptual design studies to develop designspecifications.

MSR Fuels and Materials R&DThe main objective of the fuel characterization researchis to develop a simple and reliable chemistry flowsheetthat is complete from initial fuel loading to the finalwaste form. Fundamental research needs to be con-ducted to determine kinetic and thermodynamic data,fully characterize fission product behavior, and deter-mine the optimum process for separating fission prod-ucts, including lanthanides without removal of minoractinides. Research on solubility of minor actinides andlanthanides will generate critical data needed to designreactors capable of burning minor actinides with mini-mum inventories in the reactor.

Fuel Salt Selection. The fuel salt has to meet require-ments that include neutronic properties (low neutroncross section for the solvent components, radiationstability, negative temperature coefficient), thermal andtransport properties (low melting point, thermal stability,low vapor pressure, adequate heat transfer and viscos-ity), chemical properties (high solubility of fuel compo-nents, compatibility with container and moderatormaterials, ease of fuel reprocessing), compatibility withwaste forms, and low fuel and processing costs. Tooperate the reactor as an actinide burner increases theconcentration of fission and transuranic elements in thecore, which in turn requires a higher solubility than priorart. Thus, new salt compositions such as sodium andzirconium fluorides should be investigated. Sodium hasa higher neutron absorption cross section and is thussomewhat less favorable neutronically. However, thisdrawback can be partially compensated for by increasingthe fuel enrichment. Furthermore, selection of NaF-ZrF4instead of BeF2 increases the solubility of the salt anddecreases the tritium production. Furthermore, NaF-ZrF4 and related salts, with a high percentage of thoriumdissolved in it, are thought to have a better temperaturereactivity coefficient.

Cross Sections and New Fuel Data. Despite thesuccesses of the prototypes, recent neutronics calcula-tions raise questions about the value of the temperaturereactivity coefficient of the fuel salt. To gain confi-dence, new data measurements and qualification areneeded.

Metallic Components. Materials compatibility testingrequires design and operation of a test loop whereaccelerated irradiation testing could be conducted usingfissile and fertile fuel. The primary outcome of thisresearch is to identify and address fission productreactions (if any) and to measure mechanical propertiesand demonstrate lifetime performance of structuralmaterials in the MSR. Test materials should includenickel based alloys with demonstrated performance inMSR test programs of the 1950s and 1960s such asINOR-8, Hastelloy B and N, and Inconel, as well asother promising materials such as niobium-titaniumalloys, for which lifetime performances have not yetbeen demonstrated.

The nickel based alloys have been proven as suitableMSR structural materials. INOR-8 is strong, stable,corrosion-resistant, and has good welding and formingcharacteristics. It is fully compatible with graphite, withnonsodium salts up to 815°C and with sodium salts up to700°C. Modified Hastelloy N, developed for use withfluoride salt at high temperature (up to 800°C), hasproven to be corrosion resistant but requires longer-termtesting. For nongraphite core concepts, it must be notedthat nickel based alloys are sensitive to He-inducedembrittlement under irradiation, resulting in a reductionof the creep ductility of the alloy. Tests show thattitanium addition (up to 2%) solves the embrittlementproblem and increases resistance to tellurium attack,which can also be strongly mitigated by making the saltmore reducing. Additional testing of corrosion effectsdue to molten salt in a thermal gradient, telluriumembrittlement, and irradiation effects on mechanicalproperties are all required to have full confidence in thelifetime performance of these alloys.

Graphite. Graphite’s primary function is to provideneutron moderation. Radiation damage will requiregraphite replacement every 4 to 10 years, similar to therequirements for the VHTR moderator blocks. Longer-lived graphite directly improves plant availabilitybecause the MSR does not need refueling outages. Thisis a driver for research into graphite with improvedperformance.

36


Secondary Coolant Salt Selection. The secondary saltoperates in significantly less damaging conditions thanthe primary system. The temperature is lower, there areno fission products or actinides in the salt, and theneutron fluence is much lower in the secondary system.The secondary circuit metal must resist corrosion by thecoolant salt, which could be the same as the primarycoolant or a fluoroborate (mixture of NaBF4 and NaF).However, additional research is needed to ensure thatthis salt will be satisfactory. The salt selected will partlydepend on the choice of power conversion cycle. Thissalt is more corrosive toward Hastelloy N than the fuelsalt, and additional knowledge of corrosion reactions isrequired.

MSR Balance-of-Plant R&DPower Cycle. Historically, it has been assumed that asteam power cycle would be used to produce electricity.Recent studies indicate that use of an advanced heliumgas turbine for electricity production would increaseefficiency, reduce costs, provide an efficient mechanismto trap tritium, and avoid potential chemical reactionsbetween the secondary coolant salt and the power cyclefluid. Additional research is recommended to confirmthese benefits and develop such systems.

Component Technology. Prior programs demonstratedmolten salt pump operation up to 17 000 hours. Re-search into longer life pumps is required to achieveeconomic performance goals. In addition, shields needto be developed for the motor, seal, and bearings.

Noble metals that plate-out on heat exchanger walls(metal clustering) are an operational issue that scaleswith the power level of the MSR. In the case of loss ofheat sink, the radiation thermal load of the metal clusterscould cause significant damage leading to loss ofintegrity of the MSR intermediate heat exchanger.Bismuth wash, filters, and inclusion of additives to themolten salt are approaches for preventing the metalclustering issue in MSRs. This research should beginwith an out-of-pile test loop using salt with noblemetals.

The main challenge concerning valves, joints, andfittings is to ensure correct mating of surfaces rangingfrom room temperature to 700°C. Avoiding fusionbonding with the molten salt is also a technical chal-lenge for efficient valve operation, and tests will have tobe carried out to improve reliability.

MSR Safety R&DReactor Safety. Prior programs have provided informa-tion to help demonstrate MSR safety. Nevertheless, acomprehensive safety analysis equivalent to those forcurrent reactors remains to be done. Additional technol-ogy demonstration is needed in this area.

MSR Design and Evaluation R&DDetailed design of a MSR has not been done since 1970.An updated design (including design tradeoff studies) isrequired to better understand strengths and weaknessesand allow defensible economic evaluations. The currentregulatory structure is designed for solid fuel reactors,and the MSR design needs to carefully address the intentof current regulations. Work is required with regulatorsto define equivalence in safety for MSRs. Because theMSR shares many features with reprocessing plants, thedevelopment of MSR regulatory and licensing ap-proaches should be coordinated with R&D inpyroprocessing. Under the high radiation and tempera-ture environment, remote and robotic maintenance,inspection, and repair are key technologies that requireR&D.

Fuel Salt In-Line Composition Measurement. Opera-tion of a MSR requires that adequate surveillance bemaintained on the composition of various reactorstreams, such as the redox potential of the salt (which isindicated by the U3+/U4+ ratio). Electroanalyticalmeasurement techniques will need to be developed.

MSR Fuel Cycle R&DSignificant R&D activity is required in salt processingand quality. Earlier work on salt processing developedand demonstrated flowsheets on a laboratory scale toremove radionuclides from the salt and maximize theconversion ratio. The process was divided into multipletiers, which induced large volumes of salt and wastes inthe salt processing. A key need is to develop a simpleprocess with a conversion ratio near one and which isoptimized for transmutation of actinides from otherreactors. This may allow flowsheet simplification andlesser constraints on the recovery rate of fission prod-ucts. In addition, considerable R&D is required todevelop waste forms for the MSR fuel cycle.

R&D activity is also recommended to understandproliferation resistance and physical protection issuesand their impact on the MSR design.

37


MSR R&D Schedule and CostsA schedule for the MSR R&D is shown below, alongwith the R&D costs and decision points.

MOLTEN SALT REACTOR SYSTEM (1000 M$)Materials (200 M$)



Safety (200 M$)


Fuel Cycle (300 M$)

Metal clustering Metallic materials screening Graphite core structures Core materials selection (MSR 1) Materials irradiation testing Materials for separation system Secondary salt selection Secondary salt properties Secondary salt compatibility with working fluid

Power cycle (with tritium control) Power cycle (with tritium control) (MSR 3) Tritium control technology testing Heat exchange testing Critical experiments Transient experiments Chemistry monitoring and control technology Maintenance/inspection

Heat exchanger leakage tests Tritium trapping in secondary coolant

Define accident sequences Formulate test requirements to validate codes Regulatory interactions Dedicated safety testing

Economics Preconceptual design Viability phase complete Conceptual design Analysis tools

Screen salt composition Fuel salt selection (MSR 2) Thermophysical/chemical properties FP solubility and MA solubility Evaluate separation options (screen) Fuel treatment (fission product removal) approach (MSR 4) Chemistry of separations Viability of materials (MSR 6) Separations testing Selection of noble metal management (MSR 5) Management of separation products Waste form development and qualification Immobilization of gaseous fission products

decision

decision

decision

decision

decision

decision

2000 2010 2020

38


Sodium-Cooled Fast Reactor System R&DSFR DescriptionThe Sodium-Cooled Fast Reactor (SFR) system featuresa fast-spectrum reactor [shown below] and closed fuelrecycle system. The primary mission for the SFR ismanagement of high-level wastes and, in particular,management of plutonium and other actinides. Withinnovations to reduce capital cost, the mission canextend to electricity production, given the provencapability of sodium reactors to utilize almost all of theenergy in the natural uranium versus the 1% utilized inthermal spectrum systems.

A range of plant size options are available for the SFR,ranging from modular systems of a few hundred MWe tolarge monolithic reactors of 1500–1700 MWe. Sodium-core outlet temperatures are typically 530–550ºC. Theprimary coolant system can either be arranged in a poollayout (a common approach, where all primary systemcomponents are housed in a single vessel), or in acompact loop layout, favored in Japan. For both options,there is a relatively large thermal inertia of the primarycoolant. A large margin to coolant boiling is achievedby design, and is an important safety feature of thesesystems. Another major safety feature is that the pri-

mary system operates at essentially atmospheric pres-sure, pressurized only to the extent needed to movefluid. Sodium reacts chemically with air, and withwater, and thus the design must limit the potential forsuch reactions and their consequences. To improvesafety, a secondary sodium system acts as a bufferbetween the radioactive sodium in the primary systemand the steam or water that is contained in the conven-tional Rankine-cycle power plant. If a sodium-waterreaction occurs, it does not involve a radioactive release.

Two fuel options exist for the SFR: (1) MOX and (2)mixed uranium-plutonium-zirconium metal alloy(metal). The experience with MOX fuel is considerablymore extensive than with metal.

SFRs require a closed fuel cycle to enable their advanta-geous actinide management and fuel utilization features.There are two primary fuel cycle technology options: (1)an advanced aqueous process, and (2) the pyroprocess,which derives from the term, pyrometallurgical process.Both processes have similar objectives: (1) recovery andrecycle of 99.9% of the actinides, (2) inherently lowdecontamination factor of the product, making it highlyradioactive, and (3) never separating plutonium at anystage. These fuel cycle technologies must be adaptable

39


to thermal spectrum fuels in addition to serving theneeds of the SFR. This is needed for two reasons: First,the startup fuel for the fast reactors must come ulti-mately from spent thermal reactor fuel. Second, for thewaste management advantages of the advanced fuelcycles to be realized (namely, a reduction in the numberof future repositories required and a reduction in theirtechnical performance requirements), fuel from thermalspectrum plants will need to be processed with the samerecovery factors. Thus, the reactor technology and thefuel cycle technology are strongly linked. Consequently,much of the research recommended for the SFR isrelevant to crosscutting fuel cycle issues.

A summary of the design parameters for the SFR systemis given in the following table.

There is an extensive technology base in nuclear safetythat establishes the passive safety characteristics of theSFR and their ability to accommodate all of the classicalanticipated transients without scram events without fueldamage. Landmark tests of two of these events weredone in RAPSODIE (France) in 1983 and in EBR-II(United States) in 1986. Still, there is important viabil-ity work to be done in safety. Key needs are to confirmreliability of passive feedback from heatup of reactorstructures and to establish the long-term coolability ofoxide or metal fuel debris after a bounding case accident.

The options for fuel recycle are the advanced aqueousprocess and the pyroprocess. The technology base forthe advanced aqueous process comes from the long andsuccessful experience in several countries with PUREXprocess technology. The advanced process proposed byJapan, for example, is simplified relative to PUREX anddoes not result in highly purified products. The technol-ogy base for fabrication of oxide fuel assemblies issubstantial, yet further extension is needed to make theprocess remotely operable and maintainable. The high-level waste form from advanced aqueous processing isvitrified glass, for which the technology is wellestablished.

The pyroprocess has been under development since theinception of the Integral Fast Reactor program in theUnited States in 1984. When the program was cancelledin 1994, pyroprocess development continued in order totreat EBR-II spent fuel for disposal. In this latterapplication, plutonium and minor actinides were notrecovered, and pyroprocess experience with thesematerials remains at laboratory scale. Batch size foruranium recovery, however, is at the tens-of-kilogramscale, about that needed for deployment. Remotefabrication of metal fuel was demonstrated in the 1960s.Significant work has gone into repository certification ofthe two high-level waste forms from the pyroprocess, aglass-bonded mineral (ceramic) and a zirconium-stainless steel alloy.

Technology Gaps for the SFRThe important technology gaps for the SFR are in theareas of:

• Ensuring of passive safe response to all design basisinitiators, including anticipated transients withoutscram (a major advantage for these systems)

• Capital cost reduction

• Proof by test of the ability of the reactor to accom-modate bounding events

Technology Base for the SFRSodium-cooled liquid metal reactors are the mosttechnologically developed of the six Generation IVsystems. SFRs have been built and operated in France,Japan, Germany, the United Kingdom, Russia, and theUnited States. Demonstration plants ranged from 1.1MWth (at EBR-I in 1951) to 1200 MWe (at SuperPhenixin 1985), and sodium-cooled reactors are operatingtoday in Japan, France, and Russia. As a benefit of theseprevious investments in technology, the majority of theR&D needs presented for the SFR in this roadmap areperformance-related. With the exception of passivesafety assurance, there are few viability issues withregard to the reactor systems.

The fuel options for the SFR are MOX and metal. Bothare highly developed as a result of many years of workin several national reactor development programs.Burnups in the range of 150–200 GWD/MTHM havebeen experimentally demonstrated for both. Neverthe-less, the databases for oxide fuels are considerably moreextensive than those for metal fuels.


Outlet Temperature 530-550 oCPressure ~1 AtmospheresRating 1000-5000 MWthFuel Oxide or metal alloyCladding Ferritic or ODS ferriticAverage Burnup ~150-200 GWD/MTHMConversion Ratio 0.5-1.30Average Power Density 350 MWth/m3

40


• Scale-up of the pyroprocess with demonstration ofhigh minor actinide recovery

• Development of oxide fuel fabrication technologywith remote operation and maintenance.

The main viability issues for the reactor in the SFRsystem relate to accommodating bounding events.Assurance or verification of passive safety is an impor-tant performance issue. Some consider the acquisitionof irradiation performance data for fuels fabricated withthe new fuel cycle technologies to also be a viabilityissue, rather than a performance issue. Other importantSFR reactor technology gaps are in-service inspectionand repair (in sodium), and completion of the fuelsdatabase.

A key performance issue for the SFR is cost reduction tocompetitive levels. The extent of the technology basefor SFRs is noted above, yet none of the SFRs con-structed to date have been economical to build oroperate. However, design studies have been done, someof them very extensively, in which proponents concludethat both overnight cost and busbar cost can be compa-rable to or lower than those of the advanced LWRs.Ultimately, cost reductions are best if supported byspecific innovations, providing a better measure ofconfidence. In S-PRISM, the key cost reduction is itsmodular construction. In Japanese design studies at theJapan Nuclear Cycle Development Institute, for ex-ample, innovations such as (1) a reduced number ofprimary loops, (2) an integral pump and intermediateheat exchanger, and (3) the use of improved materials ofconstruction are the basis for cost reductions.

With the advanced aqueous fuel cycle, the key viabilityissue is the minimal experience with production ofceramic pellets (using remotely operated and maintainedequipment) that contain minor actinides and traceamounts of fission products. Further, it is important todemonstrate scale-up of the uranium crystallization step.Filling both of these gaps is key to achieving cost goals.

For the pyroprocess, viability issues include lack ofexperience with larger-scale plutonium and minoractinide recoveries, minimal experience with drawdownequipment for actinide removal from electrorefiner saltsbefore processing, and minimal experience with ionexchange systems for reducing ceramic waste volume.

SFR Fuels and Materials R&DThe fuel options for the SFR are MOX and metal alloy.Either will contain a relatively small fraction of minoractinides and, with the low-decontamination fuel cycleprocesses contemplated, also a small amount of fission

products. The presence of the minor actinides andfission products dictates that fuel fabrication be per-formed remotely. This creates the need to verify thatthis remotely fabricated fuel will perform adequately inthe reactor.

These minor actinide-bearing fuels also require furtherproperty assessment work for both fuel MOX and metalfuels, but more importantly for metal fuels. Also formetal fuels, it is important to confirm fuel/claddingconstituent interdiffusion behavior when minor actinidesand additional rare earth elements are present.

SFR Reactor Systems R&DEconomics. As noted, key performance R&D remainsfor sodium-cooled reactors because of the existingknowledge and experience accumulated in this field.The reactor technology R&D that remains is aimed atenhancing the economic competitiveness and plantavailability. For example, development and/or selectionof structural materials for components and piping isimportant to development of an economically competi-tive plant. 12% Cr ferritic steels, instead of austeniticsteels, are viewed as promising structural materials forfuture plant components because of their superiorelevated temperature strength and thermal properties,including high thermal conductivity and low thermalexpansion coefficient.

In-Service Inspection, Maintenance, and Monitoring.Improvement of in-service inspection and repair tech-nologies is important to confirm the integrity of safety-related structures and boundaries that are submerged insodium, and to repair them in place. Motivated by theneed to address sodium-water reactions, it is alsoimportant to enhance the reliability of early detectionsystems for water leaks. New early detection systems,especially those that protect against small leaks, wouldbe adopted to prevent the propagation of tube rupturesand to allow a rapid return to plant operation.

SFR Balance-of-Plant R&DNoting the temperatures at which the SFRs operate,there may be interest in investigating the use of asupercritical CO2 Brayton cycle. This cycle is discussedin the Crosscutting Energy Products R&D section.

SFR Safety R&DA focused program of safety R&D is necessary tosupport the SFR. Worldwide experience with design andoperation of such systems has shown that they can beoperated reliably and safely. The safety R&D challengesfor these systems in the Generation IV context are (1) to

41


verify the predictability and effectiveness of the mecha-nisms that contribute to passively safe response todesign basis transients and anticipated transients withoutscram, and (2) to ensure that bounding events consideredin licensing can be sustained without loss of coolabilityof fuel or loss of containment function.

In-Pile Experiments. Since many of the mechanismsthat are relied upon for passively safe response can bepredicted on a first-principles basis (for example,thermal expansion of the fuel and core grid plate struc-ture), enough is now known to perform a conceptualdesign of a prototype reactor. R&D is recommended toevaluate physical phenomena and design features thatcan be important contributors to passive safety, and toestablish coolability of fuel assemblies if damage shouldoccur. This R&D would involve in-pile experiments,primarily on metal fuels, using a transient test facility.

Accommodation of Bounding Events. The secondchallenge requires analytical and experimental investiga-tions of mechanisms that will ensure passively saferesponse to bounding events that lead to fuel damage.The principal needs are to show that debris resultingfrom fuel failures is coolable within the reactor vessel,

and to show that passive mechanisms exist to precluderecriticality in a damaged reactor. A program of out-of-pile experiments involving reactor materials is recom-mended for metal fuels, while in-pile investigations ofdesign features for use with oxide fuel are now underway.

SFR Design and Evaluation R&DWhile there are design studies in progress in Japan onSFRs, there is little design work in the United States,even at the preconceptual level. Design work is animportant performance issue, and it should accelerategiven the importance of economics for the SFR. R&Dactivity is needed with a focus on the base technologyfor component development.

SFR Fuel Cycle R&DR&D activity is recommended to support the SFR fuelcycle found in the Crosscutting Fuel Cycle R&D section.

SFR R&D Schedule and CostsA schedule for the SFR R&D is shown below, alongwith the R&D costs and decision points. The schedulereflects the advancement of both oxide and metal fueloptions for the SFR.

SODIUM-COOLED FAST REACTOR SYSTEM (610 M$)Fuels and Materials (160 M$)



Safety (160 M$)

Desig n & Evaluation (100 M$)

Oxide

Metal

New materials development (12% Cr ferritic steels)

Severe accident behavior testing

Advanced pelletizing technologyOxide fuel remote fabrication technology selection (SFR 1)ODS cladding (welding)Remote maintenance developmentVibrocompaction alternativeODS MOX fuel pin irradiation

Characterize MA bearing fuelsReduce actinide losses in fabricAdvanced cladding out-of-pile testsIrradiation tests for MA bearing fuels

In-service inspection and repair technology

Increased reliability steam generators

Passive safety confirmationSASS developmentTransient fuel testing and analysis

Debris co-stabilityMolten fuel discharge/dispersal

Evaluate supercritical CO2 turbinePreconceptual designViability phase completeConceptual designAnalysis tools

decision

2000 2010 2020

42


Supercritical-Water-Cooled Reactor SystemR&DSCWR DescriptionSCWRs are high-temperature, high-pressure water-cooled reactors that operate above the thermodynamiccritical point of water (374°C, 22.1 MPa or 705°F, 3208psia). One SCWR system is shown below. Thesesystems may have a thermal or fast-neutron spectrum,depending on the core design. SCWRs have uniquefeatures that may offer advantages compared to state-of-the-art LWRs in the following:

• SCWRs offer increases in thermal efficiency relativeto current-generation LWRs. The efficiency of aSCWR can approach 44%, compared to 33–35% forLWRs.

• A lower-coolant mass flow rate per unit core thermalpower results from the higher enthalpy content of

the coolant. This offers a reduction in the size of thereactor coolant pumps, piping, and associatedequipment, and a reduction in the pumping power.

• A lower-coolant mass inventory results from theonce-through coolant path in the reactor vessel andthe lower-coolant density. This opens the possibilityof smaller containment buildings.

• No boiling crisis (i.e., departure from nucleateboiling or dry out) exists due to the lack of a secondphase in the reactor, thereby avoiding discontinuousheat transfer regimes within the core during normaloperation.

• Steam dryers, steam separators, recirculation pumps,and steam generators are eliminated. Therefore, theSCWR can be a simpler plant with fewer majorcomponents.

43


The Japanese supercritical light water reactor (SCLWR)with a thermal spectrum has been the subject of the mostdevelopment work in the last 10 to 15 years and is thebasis for much of the reference design. The SCLWRreactor vessel is similar in design to a PWR vessel(although the primary coolant system is a direct-cycle,BWR-type system). High-pressure (25.0 MPa) coolantenters the vessel at 280°C. The inlet flow splits, partlyto a downcomer and partly to a plenum at the top of thecore to flow down through the core in special water rods.This strategy provides moderation in the core. Thecoolant is heated to about 510°C and delivered to apower conversion cycle, which blends LWR andsupercritical fossil plant technology; high-, intermediate-and low-pressure turbines are employed with two reheatcycles. The overnight capital cost for a 1700-MWeSCLWR plant may be as low as $900/kWe (about halfthat of current ALWR capital costs), considering theeffects of simplification, compactness, and economy ofscale. The operating costs may be 35% less than currentLWRs.

The SCWR can also be designed to operate as a fastreactor. The difference between thermal and fast ver-sions is primarily the amount of moderator material inthe SCWR core. The fast spectrum reactors use noadditional moderator material, while the thermal spec-trum reactors need additional moderator material in thecore.

A summary of designs parameters for the SCWR systemis given in the following table.

Technology Base for the SCWRMuch of the technology base for the SCWR can befound in the existing LWRs and in commercialsupercritical-water-cooled fossil-fired power plants.However, there are some relatively immature areas.There have been no prototype SCWRs built and tested.For the reactor primary system, there has been very littlein-pile research done on potential SCWR materials ordesigns, although some SCWR in-pile research has beendone for defense programs in Russia and the UnitedStates. Limited design analysis has been underway overthe last 10 to 15 years in Japan, Canada, and Russia. Forthe balance of plant, there has been development ofturbine generators, piping, and other equipment exten-sively used in supercritical-water-cooled fossil-firedpower plants. The SCWR may have some success atadopting portions of this technology base.

Technology Gaps for the SCWRThe important SCW technology gaps are in the areas of:

• SCWR materials and structures, including

– Corrosion and stress corrosion cracking (SCC)

– Radiolysis and water chemistry

– Dimensional and microstructural stability

– Strength, embrittlement, and creep resistance

• SCWR safety, including power-flow stability duringoperation

• SCWR plant design.

Important viability issues are found within the first twoareas, and performance issues are found primarily withinthe first and third areas.

SCWR Fuels and Materials R&DThe supercritical water (SCW) environment is uniqueand few data exist on the behavior of materials in SCWunder irradiation and in the temperature and pressureranges of interest. At present, no candidate alloy hasbeen confirmed for use as either the cladding or struc-tural material in thermal or fast spectrum SCWRs.Potential candidates include austenitic stainless steels,solid solution and precipitation-hardened alloys, ferritic-martensitic alloys, and oxide dispersion-strengthenedalloys.

The fast SCWR design would result in greater doses tocladding and structural materials than in the thermaldesign by a factor of 5 or more. The maximum doses for


Plant capital cost $900/KWUnit power and neutron 1700 MWe,spectrum thermal spectrumNet efficiency 44%Coolant inlet and outlet 280°C/510°C/25 MPatemperatures and pressureAverage power density ~100 MWth/m3

Reference fuel UO2 with austeniticor ferritic-martensiticstainless steel, orNi-alloy cladding

Fuel structural materials Advanced high-strengthcladding structural materials metal alloys are neededBurnup / Damage ~45 GWD/MTHM;

10–30 dpaSafety approach Similar to ALWRs

44


the core internals are in the 10–30 dpa range in thethermal design, and could reach 100–150 dpa in the fastdesign. These doses will result in greater demands onthe structural materials in terms of the need for irradia-tion stability and effects of irradiation on embrittlement,creep, corrosion, and SCC. The generation of helium bytransmutation of nickel is also an important consider-ation in both the thermal and fast designs because it canlead to swelling and embrittlement at high temperatures.The data obtained during prior fast reactor developmentwill play an important role in this area.

To meet these challenges, the R&D plan for the claddingand structural materials in the SCWRs focuses onacquiring data and a mechanistic understanding relatingto the following key property needs: corrosion and SCC,radiolysis and water chemistry, dimensional and microstructural stability, and strength and creep resistance.

Corrosion and SCC. The SCWR corrosion and SCCresearch activities should focus on obtaining the follow-ing information:

• Corrosion rates in SCW at temperatures between280 and 620°C (the corrosion should be measuredunder a wide range of oxygen and hydrogen contentsto reflect the extremes in dissolved gasses)

• Composition and structure of the corrosion films asa function of temperature and dissolved gasses

• The effects of irradiation on corrosion as a functionof dose, temperature, and water chemistry

• SCC as a function of temperature, dissolved gasses,and water chemistry

• The effects of irradiation on SCC as a function ofdose, temperature, and water chemistry.

The corrosion and SCC R&D activities will be organizedinto three parts: an extensive series of out-of-pilecorrosion and SCC experiments on unirradiated alloys,companion out-of-pile corrosion and SCC experimentson irradiated alloys, and in-pile loop corrosion and SCCtests. It is envisioned that at least two and maybe asmany as four out-of-pile test loops would be used, someaddressing the corrosion issues and others addressing theSCC issues. At least two such loops should be builtinside a hot cell in order to study preirradiated material.Facilities to preirradiate samples prior to corrosion andSCC testing will be required. This work should becarried out over a 6–10 year time span for unirradiatedmaterials and the same for irradiated materials. Accel-erators capable of producing high currents of light ionsmay also be utilized to study irradiation effects on

corrosion and SCC in a postirradiation mode at substan-tially lower cost than reactor irradiations.

About mid-way through the out-of-pile work, one or twoin-pile test loops, should start operating under both fastand thermal spectrum irradiation conditions (for a totalof 3 to 4 loops). The in-pile loops will be used to studycorrosion, SCC, and water chemistry control issuesdescribed below. About 10 years of in-pile testing inthese loops will be needed to obtain all the required datato support both the viability and performance phases ofthe development of the thermal spectrum version of theSCWR, and about 15 years to obtain the needed infor-mation for the fast spectrum SCWR. A postirradiationcharacterization and analysis program will accompanythe reactor- and accelerator-based irradiations beginningin year 5 and extending for a 10-year period.

Radiolysis and Water Chemistry. The SCWR waterchemistry research program should focus on obtainingthe following information:

• The complete radiolysis mechanism in SCW as afunction of temperature and fluid density

• The chemical potential of H2, O2, and variousradicals in SCW over a range of temperatures (280–620°C)

• Recombination rates of various radicals, H2, and O2in SCW over a range of temperatures (280–620°C)

• Effect of radiation type: neutrons, gammas, as wellas flux on radiolysis yields

• Formation and reaction of other species by radiolyticprocesses

• Impurities introduced into the primary system.

Two research avenues are envisioned to obtain thisinformation. First, beam ports and accelerators can beused to irradiate SCW chemistries and study the charac-teristics of the recombination processes in some detail.This information will be integrated into a model of thewater radiolysis mechanism. Second, water chemistrycontrol studies can be performed using the in-pile testloops needed for the corrosion and SCC research dis-cussed above.

Dimensional and Microstructural Stability. TheSCWR dimensional and microstructural researchactivities should focus on obtaining the followinginformation:

• Void nucleation and growth, and the effect of Heproduction, on void stability and growth, and He

45


bubble nucleation and growth as a function of doseand temperature

• Development of the dislocation and precipitatemicrostructure and radiation-induced segregation asa function of dose and temperature

• Knowledge of irradiation growth or irradiation-induced distortion as a function of dose and tem-perature

• Knowledge of irradiation-induced stress relaxationas a function of tension, stress, material, and dose.

While many of the test specimens for this work will beirradiated in the corrosion and SCC in-pile loops dis-cussed above, accelerator-based irradiation offers a rapidand low-cost alternative to the handling and analysis ofneutron-irradiated material. Much of the needed infor-mation will be obtained during postirradiation examina-tions over the 15-year period of the corrosion and SCCtests. In addition, some stand-alone capsule irradiationtests in test reactors should be performed in order totimely obtain data on a range of candidate materials. Itmay be possible to utilize some existing SFR data in thisresearch.

Strength, Embrittlement, and Creep Resistance. TheSCWR strength, embrittlement, and creep resistanceresearch activities should focus on obtaining the follow-ing information:

• Tensile properties as a function of dose andtemperature

• Creep rates and creep rupture mechanisms as afunction of stress, dose, and temperature

• Creep-fatigue as a function of loading frequency,dose, and temperature

• Time dependence of plasticity and high-temperatureplasticity

• Fracture toughness as a function of irradiationtemperature and dose

• Ductile-to-brittle transition temperature (DBTT) andhelium embrittlement as a function of dose andirradiation temperature

• Changes in microstructure and mechanical proper-ties following design basis accidents.

The research should aim at high-temperature perfor-mance of both irradiated and unirradiated alloys and alsoat low-temperature performance of irradiated alloys.High-temperature testing will include yield propertydetermination, time dependent (creep) experiments, and

also the effect of fatigue loading with a high mean stress.This R&D should be conducted first on unirradiatedalloys over a period of 8 years. Midway through thework, testing should begin on irradiated materials for aperiod of 10 years. The low-temperature fracturetoughness/DBTT program will require 10 years.

SCWR Reactor Systems R&DA number of reactor system alternatives have beendeveloped for both vessel and pressure tube versions ofthe SCWR. Significant additional work in this area isnot needed. The component development and prooftesting is covered in the SCWR Design and Evaluationsection.

SCWR Balance-of-Plant R&DThe SCWRs can utilize the existing technology from thesecondary side of the supercritical water-cooled fossil-fired plants. Significant research in this area is notneeded.

SCWR Safety R&DAn SCWR safety research activity is recommended,organized around the following topics:

• Reduced uncertainty in SCW transport properties

• Further development of appropriate fuel cladding tocoolant heat transfer correlations for SCWRs undera range of fuel rod geometries

• SCW critical flow measurements, as well as modelsand correlations

• Measurement of integral loss-of-coolant accident(LOCA) thermal-hydraulic phenomena in SCWRsand related computer code validation

• Fuel rod cladding ballooning during LOCAs

• SCWR design optimization studies, includinginvestigations to establish the reliability and systemcost impacts of passive safety systems

• Power-flow stability assessments.

Transport Properties and CorrelationsThe purpose of making additional basic thermal-hydrau-lic property measurements at and near the pseudo-critical temperatures would be to improve the accuracyof the international steam-water property tables. Thiswork could be done over a 3–5 year time frame.

The fuel cladding-to-SCW heat transfer research shouldconsist of a variety of out-of-pile experiments startingwith tubes and progressing to small and then relative

46


large bundles of fuel rods. The bundle tests shouldinclude some variations in geometry (such as fuel roddiameter and pitch, bundle length, channel boxes), axialpower profiles, coolant velocity, pressure, and gridspacer design. The larger bundle tests will requireseveral megawatts of power and the ability to designelectrically heated test rods with appropriate powershapes. This program might take 5–6 years.

The SCW critical flow experiments would be out-of-pileexperiments with variations in hole geometry and waterinventory. This research would take 4–5 years.

LOCA Phenomena and Accident AnalysisThe integral SCWR LOCA thermal-hydraulic experi-ments would be similar to the Semiscale experimentspreviously conducted for the U.S. Nuclear RegulatoryCommission to investigate LOCA phenomena for thecurrent LWRs. A test series and the related computercode development would take about 10 years. It may bepossible to design this facility to accommodate the heattransfer research discussed above as well as the neededLOCA testing, and even some thermal-hydraulic insta-bility testing.

Fuel rod cladding ballooning is an important phenom-enon that may occur during a rapid depressurization.Although considerable work has been done to measureand model the ballooning of Zircaloy clad fuel rodsduring LOCAs, little is known about the ballooningbehavior of austenitic or ferritic-martensitic stainlesssteel or nickel-based alloy clad fuel rods during aLOCA. It is expected that this information could beobtained from out-of-pile experiments using fuel rodsimulators. The research would take 4–6 years.

All of the known accident scenarios must be carefullyevaluated. These include large- and small-breakLOCAs, reactivity insertion accidents (RIAs), loss offlow, main steam isolation valve closure, over coolingevents, anticipated transients without scram, and high-and low-pressure boil off. There may be safety features(e.g., very-high-pressure accumulators) that requirespecial designs. It is estimated that tests can be con-ducted within a period of 3–5 years.

Flow StabilityThe objective of the power-flow stability R&D is betterunderstanding of instability phenomena in SCWRs,identification of the important variables affecting thesephenomena, and, ultimately, the generation of mapsidentifying the stable operating conditions of the differ-ent SCWRs designs. Consistent with the U.S. Nuclear

Regulatory Commission approach to BWRs licensing,the licensing of SCWRs will probably require, at aminimum, demonstration of the ability to predict theonset of instabilities. This can be done by means of afrequency-domain linear analysis.

Both analytical and experimental stability studies needto be carried out for the conditions expected during thedifferent operational modes and accidents. The analyti-cal studies can obviously be more extensive and coverboth works in the frequency domain, as well as directsimulations. These studies can consider the effect ofimportant variables such as axial and radial powerprofile, moderator density and fuel temperature reactiv-ity feedback, fuel rod thermal characteristics, coolantchannel hydraulic characteristics, heat transfer phenom-ena, and core boundary conditions. Mitigating effectssuch as orificing, insertion of control rods, and fuelmodifications to obtain appropriate thermal and/orneutronic response time constants can also be assessedusing analytical simulations. Instability experimentscould be conducted at the multipurpose SCW thermal-hydraulic facility recommended for the safety experi-mentation discussed above. The test section should bedesigned to accommodate a single bundle, as well asmultiple bundles. This will enable studying in-phaseand out-of-phase density wave oscillations. Moreover,the facility will provide a natural circulation flow pathfor the coolant to study buoyancy loop instabilities. Theinstability experiments and related analytical work willrequire about 3 to 4 years. Further work would dependon the issues identified during the experimental program.

SCWR Design and Evaluation R&DMany of the major systems that can potentially be usedin a SCWR were developed for the current BWRs,PWRs, and SCW fossil plants. Therefore, the majorplant design and development needs that are unique forSCWRs are primarily found in their design optimization,as well as their performance and reliability assuranceunder SCWR neutronic and thermo-hydraulic condi-tions. Two major differences in conditions are thestresses due to the high SCWR operating pressure(25 MPa) and the large coolant temperature and densitychange (approximately 280 to 500°C or more, 800 to 80kg/m3, respectively) along the core under the radiationfield.

Examples of design features that need to be optimized toachieve competitiveness in economics without sacrific-ing safety or reliability include the fuel assemblies,control rod drive system, internals, reactor vessel,

47


pressure relief values, coolant cleanup system, reactorcontrol logic, turbine configuration, re-heaters, deaerator,start-up system and procedures, in-core sensors, andcontainment building. This work is expected to takeabout 8 to 10 years.

SCWR Fuel Cycle R&DThe thermal spectrum SCWR option will use conven-tional LEU fuel. The fuel itself is developed; however,new cladding materials and fuel bundle designs will beneeded, as discussed in the Crosscutting Fuels andMaterials R&D section. The designs for the thermal

SUPERCRITICAL-WATER-COOLED REACTOR SYSTEM (870 M$)Fuels and Materials (500 M$)


Balance of Plant (10 M$)Safety (220 M$)


Fuel Cycle (10 M$)

Mechanical properties (unirradiated) Core structural material down-selection (SC 2) Corrosion/SCC (out of pile) Radiolysis and water chemistry (beam ports/accelerators) Irradiation tests (capsule/accelerator/PIE) Core structural material final selection (fast, thermal) (SC 3) In-pile water chem/corrosion/SCC and PIE Adequacy of fuel/cladding system performance potential (SC 5) Mechanical properties (irradiated and PIE)

SCW transport properties Heat transfer in rod bundles

Safety approach specification Safety approach specification (fast, thermal) (SC 1) Critical flow (out of pile, sep. effects measurements) Integral LOCA experiments Cladding ballooning (out of pile experiments) Out-of-pile instability experiments Instability analysis and data verification Severe accident behavior

Preconceptual design Viability phase complete Conceptual design Analysis tools

Fuel cycle Advanced aqueous process applicability for fuel recycle (SC 4)

decision

decision

decision

decision

decision

2000 2010 2020

spectrum SCWR will need significant additional mod-erator, i.e., water rods or solid moderation. The designsfor the fast spectrum SCWRs will require a tight pitch,but high neutron leakage to create a negative densitycoefficient. The fast spectrum SCWR option uses mixedplutonium-uranium oxide fuel with advanced aqueousreprocessing. These fuel cycle technologies are dis-cussed in the Crosscutting Fuel Cycle R&D section.

SCWR R&D Schedule and CostsA schedule for the SCWR R&D is shown below, alongwith the R&D costs and decision points.

48


Very-High-Temperature Reactor SystemR&DVHTR DescriptionThe VHTR is a next step in the evolutionary develop-ment of high-temperature gas-cooled reactors. TheVHTR can produce hydrogen from only heat and waterby using thermochemical iodine-sulfur (I-S) process orfrom heat, water, and natural gas by applying the steamreformer technology to core outlet temperatures greaterthan about 1000°C. A reference VHTR system thatproduces hydrogen is shown below. A 600 MWthVHTR dedicated to hydrogen production can yield over2 million normal cubic meters per day. The VHTR canalso generate electricity with high efficiency, over 50%at 1000°C, compared with 47% at 850°C in the GT-MHR or PBMR. Co-generation of heat and power makesthe VHTR an attractive heat source for large industrialcomplexes. The VHTR can be deployed in refineriesand petrochemical industries to substitute large amounts

of process heat at different temperatures, includinghydrogen generation for upgrading heavy and sour crudeoil. Core outlet temperatures higher than 1000°C wouldenable nuclear heat application to such processes assteel, aluminum oxide, and aluminum production.

The VHTR is a graphite-moderated, helium-cooledreactor with thermal neutron spectrum. It can supplynuclear heat with core-outlet temperatures of 1000°C.The reactor core type of the VHTR can be a prismaticblock core such as the operating Japanese HTTR, or apebble-bed core such as the Chinese HTR-10. Forelectricity generation, the helium gas turbine system canbe directly set in the primary coolant loop, which iscalled a direct cycle. For nuclear heat applications suchas process heat for refineries, petrochemistry, metal-lurgy, and hydrogen production, the heat applicationprocess is generally coupled with the reactor through anintermediate heat exchanger (IHX), which is called anindirect cycle.

49


Technology Base for the VHTRThe VHTR evolves from HTGR experience and exten-sive international databases that can support its develop-ment. The basic technology for the VHTR has been wellestablished in former HTGR plants, such as Dragon,Peach Bottom, AVR, THTR, and Fort St Vrain and isbeing advanced in concepts such as the GT-MHR andPBMR. The ongoing 30-MWth HTTR project in Japanis intended to demonstrate the feasibility of reachingoutlet temperatures up to 950°C coupled to a heatutilization process, and the HTR-10 in China willdemonstrate electricity and co-generation at a powerlevel of 10 MWth. The former projects in Germany andJapan provide data relevant to VHTR development.

Steam reforming is the current hydrogen productiontechnology. The coupling of this technology will bedemonstrated in large scale in the HTTR program butstill needs complementary R&D for market introduction.R&D on thermochemical I-S process is presentlyproceeding in the laboratory-scale stage.

Technology Gaps for the VHTRThe design parameters considered for the VHTR areshown in the table.

Process-specific R&D gaps exist to adapt the chemicalprocess and the nuclear heat source to each other withregard to temperatures, power levels, and operationalpressures. Heating of chemical reactors by helium isdifferent from current industrial practice and needsspecific R&D and demonstration. Qualification of high-temperature alloys and coatings for resistance to corro-sive gases like hydrogen, carbon monoxide, and methanewill be needed.

The viability of producing hydrogen using the iodine-sulfur (I-S) process still requires pilot- and large-scaledemonstration of the three basic chemical reactions anddevelopment of corrosion-resistant materials. Anycontamination of the product will have to be avoided.Development of heat exchangers, coolant gas ducts, andvalves will be necessary for isolation of the nuclearisland from the production facilities. This is especiallythe case for isotopes like tritium, which can easilypermeate metallic barriers at high temperatures.

Performance issues for the VHTR include developmentof a high-performance helium turbine for efficientgeneration of electricity. Modularization of the reactorand heat utilization systems is another challenge forcommercial deployment of the VHTR.

VHTR Fuels and Materials R&DQualification of TRISO Fuel. The increase of thehelium core-outlet temperature of the VHTR results inan increase of the fuel temperature and reduced marginsin case of core heatup accidents. Fuel particles coatedwith silicon-carbide are used in HTGRs at fuel tempera-tures of about 1200°C. Irradiation testing is required todemonstrate that TRISO-coated particles can performacceptably at the high burnup and temperature associ-ated with the VHTR. Following irradiation, high-temperature heating (safety) tests are needed to deter-mine that there is no degradation in fuel performanceunder accident heatup conditions up to 1600°C as aresult of the more demanding irradiation service condi-tions. These fuel demonstration activities would requireabout 5 to 7 years to complete following fabrication ofsamples. Complete fuel qualification would require anadditional 5 to 7 years in which statistically significantproduction scale fuel is irradiated to confirm the perfor-mance of the fuel from the production facility. Irradia-tion facilities and safety test facility exist worldwide,and an integrated coordinated fuel development programcould shorten development times by one-third.

ZrC Coatings for TRISO Fuel. Above a fuel tempera-ture of 1200°C, new coating materials such as zirco-nium-carbide and/or improved coating techniques should


Reactor power 600 MWthCoolant inlet/outlet 640/1000°CtemperatureCore inlet/outlet pressure Dependent on processHelium mass flow rate 320 kg/sAverage power density 6–10 MWth/m3

Reference fuel compound ZrC-coated particles inblocks, pins or pebbles

Net plant efficiency >50%

Demonstrating the viability of the VHTR core requiresmeeting a number of significant technical challenges.Novel fuels and materials must be developed that:

• Permit increasing the core-outlet temperatures from850°C to 1000°C and preferably even higher

• Permit the maximum fuel temperature reachedfollowing accidents to reach 1800°C

• Permit maximum fuel burnup of 150–200 GWD/MTHM

• Avoid power peaking and temperature gradients inthe core, as well as hot streaks in the coolant gas.

50


be considered. Use of ZrC in HTGRs enables an in-crease in power density and an increase in power sizeunder the same coolant outlet temperature and allows forgreater resistance against chemical attack by the fissionproduct palladium. The limited fabrication and perfor-mance data on ZrC indicates that although it is moredifficult to fabricate, it could allow for substantiallyincreased operating and safety envelopes (possiblyapproaching 1800°C). Only laboratory-scale fabricationof ZrC-coated particle fuel has been performed to date.Research into more economical commercial-scalefabrication routes for ZrC-coated particle fuels, includ-ing process development at production scale, is required.Advanced coating techniques or advanced processingtechniques (automation) should be considered. Processdevelopment on production-scale coating is required.Irradiation testing and high-temperature heating (safety)tests are needed to define operation and safety enve-lopes/limits for this fuel, with the goal of high burnup(>10% FIMA and high-temperature (1300–1400°C)operation. The facilities used for TRISO-coated particletesting can also be used for ZrC-coated fuel develop-ment. These activities would require 10 to 15 years tocomplete and could be performed at facilities adaptedfrom those available around the world currently used forSiC-based coated particle fuel.

Burnable Absorbers. Increasing the allowable fuelburnup requires development of burnable absorbers forreactivity control. The behavior of burnable absorbersneeds to be established (e.g., irradiation dimensionalstability, swelling, lifetime) under the design serviceconditions of the VHTR.

Carbon-Carbon Composite Components. Develop-ment of carbon-carbon composites is needed for controlrod sheaths, especially for the VHTR based on a pris-matic block core, so that the control rods can be insertedto the high-temperature areas entirely down to the core.Promising ceramics such as fiber-reinforced ceramics,sintered alpha silicon-carbide, oxide-composite ceram-ics, and other compound materials are also being devel-oped for other industrial applications needing high-strength, high-temperature materials. Planned R&Dincludes testing of mechanical and thermal properties,fracture behavior, and oxidation; post irradiation heat-uptests; and development of models of material behaviorand stress analysis code cases considering anisotropy.The feasibility of using superplastic ceramics in VHTRcomponents will be investigated by studying the effectsof neutron irradiation on superplastic deformationmechanisms. Testing of core internals is envisioned totake 5 to 10 years at any of the test reactors worldwide.

Pressure Vessel Materials. To realize the goal of coreoutlet temperatures higher than 1000°C, new metallicalloys for reactor pressure vessels have to be developed.At these core-outlet temperatures, the reactor pressurevessel temperature will exceed 450°C. LWR pressurevessels were developed for 300°C service, and theHTTR vessel for 400°C. Hasteloy-XR metallic materi-als are used for intermediate heat exchanger and high-temperature gas ducts in the HTTR at core-outlettemperatures up to about 950°C, but further develop-ment of Ni-Cr-W super-alloys and other promisingmetallic alloys will be required for the VHTR. Theirradiation behavior of these superalloys at the serviceconditions expected in the VHTR will need to be charac-terized. Such work is expected to take 8 to 12 years andcan be performed at facilities available worldwide.

An alternate pressure vessel allowing for larger diam-eters and ease of transportation, construction, anddismantling would be the prestressed cast-iron vessel,which can also prevent a sudden burst due to separationof mechanical strength and leak tightness. The vesselcould also include a passive decay heat removal systemwith enhanced efficiency.

Heat Utilization Systems Materials. Internal corestructures and cooling systems, such as intermediateheat exchanger, hot gas duct, process components, andisolation valve that are in contact with the hot heliumcan use the current metallic materials up to about1000°C core-outlet temperature. For core-outlet tem-peratures exceeding 1000°C, ceramic materials must bedeveloped. Piping and component insulation alsorequires design and materials development.

VHTR Reactor Systems R&DCore Internals. Core internal structures containing thefuel elements such as pebbles or blocks are made ofhigh-quality graphite. The performance of high-qualitygraphite for core internals has been demonstrated in gas-cooled pilot and demonstration plants, but recent im-provements in the manufacturing process of industrialgraphite have shown improved oxidation resistance andbetter structural strength. Irradiation tests are needed toqualify components using advanced graphite or compos-ites to the fast fluence limits of the VHTR.

VHTR Balance-of-Plant R&DThe VHTR balance-of-plant is determined by thespecific application, which can be thermochemicalprocesses, dedicated electricity production or cogenera-tion. All components have to be developed for tempera-tures well above the present state of the art and depend

51


on a comprehensive material qualification activity.Failure mechanisms such as creep, fretting, andratcheting have to be studied in detail, precluded withdesign, and demonstrated in component tests. Specificcomponents such as IHX, isolation valves, hot gas ductswith low heat loss, steam reformers, and process-relatedheat exchangers have to be developed for use in themodular VHTR, which mainly uses only one loop. Thisleads to much larger components than formerly devel-oped and a new design approach by modularization ofthe component itself.

Low pressures are necessary or preferable for manyprocesses. Alternate coolants for the intermediate loopsuch as molten salt should be adapted where needed.

Process-specific components will need to be tested.Other applications will require different componentssuch as helium-heated steam crackers, distiller columns,and superheaters.

I-S Process Subsystem. The development and qualifi-cation of an I-S process subsystem is needed. This isdiscussed in the Crosscutting Energy Products R&Dsection.

Analysis Methods. Extension and validation of existingengineering and safety analysis methods is required toinclude new materials, operating regimes, and compo-nent configurations in the models. New models need tobe developed for the VHTR with balance of plantconsisting of thermochemical process and other energyapplications.

VHTR Safety R&DPassive heat removal systems should be developed tofacilitate operation of the VHTR, with a final goal ofsimple operation and transparent safety concepts.Demonstration tests should be performed on the VHTRto verify the system’s passive characteristics, which havea lower margin between operational temperatures andthe limits for fuel and materials.

Analysis and demonstration of the inherent safetyfeatures of the VHTR are needed, and could potentiallydraw on development and demonstration of earlier INTDgas reactors. Additional safety analysis is necessarywith regard to nuclear process heat applications in anindustrial environment. The safe isolation of the reactor

system after failures in the heat delivery system is anessential issue for demonstration of IHX and hot gasvalve tightness after depressurization of the secondarycircuit. Full-scale tests of valves and IHX modules willbe necessary.

Design basis and severe accident analyses for the VHTRwill need to include phenomena such as chemical attackof graphitic core materials, typically either by air orwater ingress. Adequacy of existing models will need tobe assessed, and new models, may need to be developedand validated.

VHTR Fuel Cycle R&DDisposal of Once-Through Fuel and Graphite. TheVHTR assumes a once-through, LEU (<20% 235U) fuelcycle. Like LWR spent fuel, VHTR spent fuel could bedisposed of in a geologic repository or conditioned foroptimum waste disposal. The current HTGR particlefuel coatings form an encapsulation for the spent fuelfission products that is extremely resistant to leaching ina final repository. However, as removed from thereactor, the fuel includes large quantities of graphite, andresearch is required to define the optimum packagingform of spent VHTR fuels for long-term disposal.Radiation damage will require graphite replacementevery 4 to 10 years. An optimized approach for dealingwith the graphite (i.e., recycle, low-level waste, remainintegral with spent fuel) remains to be defined.

Fuel Recycling. Recycling of LWR and VHTR spentfuel in a symbiotic fuel cycle can achieve significantreductions in waste quantities and radiotoxicity becauseof the VHTR’s ability to accommodate a wide variety ofmixtures of fissile and fertile materials without signifi-cant modification of the core design. This flexibilitywas demonstrated in the AVR test reactor in Germanyand is a result of the ability of gas reactors to decouplethe optimization of the core cooling geometry from theneutronics.

For an actinide burning alternative, specific Pu-baseddriver fuel and transmutation fuel containing minoractinides would have to be developed. This fuel canbenefit from the above mentioned R&D on SiC and ZrCcoating but will need more R&D than LEU fuel.

52


VERY-HIGH-TEMPERATURE REACTOR SYSTEM (670 M$)Fuels and Materials (170 M$)



Safety (80 M$)

Desig n & Evaluation (90 M$)

Fuel Cycle (30 M$)

ZrC coated fuel fabricationFuel coating material & design concept (VH 4)ZrC coated fuel irradiation testAdequacy of fuel performance potential (VH 6)Burnup extensionRPV metallic material (T>–450-600˚C)Reactor structural material selection (VH5)RCS metallic material (T>=950˚C)Oxide resistant graphite for core internalsCeramic material for core internals (CR sheath)Ceramic materials for RCS

Passive DHR systemRefueling system

Electricity generation (turbine, compressor, recuperator)High temperature helium turbine (VH 1)Coupling approach and technologyReactor/H2 production process coupling approach (VH 2)Components (IHX, isolation valves, etc.)

Dynamic analysisFP evaluationSafety evaluation/PRAPost-irradiation heat-up test

EconomicsPreconceptual designViability phase completeIdentification of targeted operating temperature (VH3)Conceptual designAnalysis tools

Spent fuel characterizationFuel conditioning/packagingSeparations technology

decision

decision

decision

decision

decision

2000 2010 2020

VHTR R&D Schedule and CostsA schedule for the VHTR R&D is shown below, alongwith the R&D costs and decision points.

53


The crosscutting R&D is organized into the followingareas:

• Fuel cycle

• Fuels and materials

• Energy products

• Risk and safety

• Economics

• Proliferation Resistance and Physical Protection.

Crosscutting Fuel Cycle R&DIntroduction and ApproachA number of options for fuel and recycle technologydevelopment are shared among the six Generation IVsystems. The table below provides an overview of thesesystems, indicating primary and secondary technologyoptions. While this table is organized into four majorfuel categories and two recycle technologies, it isimportant to note that a tight coupling exists in any givensystem between its reactor, fuel, and recycling technol-ogy. These technologies are specialized to a particularsystem through studies and experiments aimed atoptimizing a given system.

RECOMMENDED CROSSCUTTING R&D

The crosscutting fuel cycle R&D is structured recogniz-ing the close coupling of fuel and recycle technologiesfor a given system, but also the value of commontechnology development for Generation IV systems. Inparticular:

• Fuel choice and in-service performance are closelycoupled to, and require specialization for, eachsystem. Therefore, fuel development R&D isdefined for each Generation IV system individually.Relevant developments for different Generation IVsystems will be shared, and effective ways to adapttechnologies will be sought.

• Fuel recycle technology R&D requires substantialinvestment in specialized facilities, so shareddevelopment of recycle technologies and commontest facilities are desirable. Recycle technologyR&D is outlined primarily in terms of the SFRsystem, which is at a comparatively advanced stateof development for both of its selected options (i.e.,oxide fuel with advanced aqueous recycle, andmetal-alloy fuel with pyroprocess recycle). Adapta-tion of the SFR advanced aqueous andpyroprocessing technologies to other Generation IVsystems (e.g., to nitride fuel for the LFR system, or

GenerationIV System Fuel Recycle

Oxide Metal Nitride Carbide Advanced PyroprocessAqueous

GFR1 S P P P

MSR2

SFR3 P P P P

LFR S P P P

SCWR P P

VHTR4 P S S

P: Primary option; S: Secondary option1 The GFR proposes (U,Pu)C in ceramic-ceramic (cercer), coated particles or ceramic-metallic (cermet).2 The MSR employs a molten fluoride salt fuel and coolant, and fluoride-based processes for recycle.3 The SFR has two options: oxide fuel with advanced aqueous, and metal fuel with pyroprocess.4 The VHTR uses a once-through fuel cycle with coated (UCO) fuel kernels, and no need for fuel treatment, as the primary option.

54


reduces the low-level waste. The advanced pelletizingprocess is simplified by eliminating the powder blendingand granulation steps from the conventional MOX pelletprocess.

In the oxide fuel cycle, greater than 99% of U/TRU isexpected to be recycled, and the decontamination factorof the reprocessing product is higher than 100. Fewviability R&D activities are needed, because the mainprocess technology builds heavily on prior light waterand fast reactor fuel cycle technology. Therefore, thisfuel cycle can be rapidly advanced to the demonstrationstage.

To achieve both economic competitiveness and reducedenvironmental impact, the following R&D is recom-mended:

• Determine the crystallization performance ofactinides, the crystallization performance of ura-nium, and the separation efficiency of solids atengineering scale

• Develop the salt-free minor actinide recoveryprocess with high extraction capability for Am andCm, and separation from lanthanides

• Develop compact centrifugal-type contactors toenable a reduction of the facility size

• Establish the fabricability of low- decontaminationfactor minor actinide-bearing pellet fuel (with anemphasis on sinterability), and develop the appara-tus for remote system operability and maintainabil-ity in a hot cell facility

• Extend current studies of the proliferation resistanceof this technology.

Pyroprocess and Remote Metal Fuel Fabrication.Pyroprocessing and refabrication are the preferredrecycle technologies for the metal-fueled SFR option. Aschematic of a closed fuel cycle with pyroprocesingtechnology is shown in the figure on page 56.Pyroprocessing employs molten salts and liquid metalsfor treatment, management, and recycle of spent fuel. Itcan recycle metallic fuel from fast reactors, and withappropriate head end steps to reduce actinide oxides tometals, it can process existing LWR fuel to recovertransuranics for feed to fast reactors. These two useshave many common characteristics and process steps.

Work on the pyroprocessing fuel cycle has been per-formed in the United States, Japan, and Europe. Asignificant portion of the viability R&D and someperformance R&D have already been performed as partof the ongoing EBR-II fuel treatment program in the

to composite fuels for the GFR) will explore keyviability questions at an early stage. These special-izations are presented as system-specific R&D intheir respective sections.

In addition to fuel recycle technology development,crosscutting fuel cycle R&D recommendations are madeto (1) improve the technical and cost performanceachieved in Generation IV fuel cycles, and (2) betterinform the selection of integrated Generation IV fuelcycles by clarifying the advantages and drawbacks oftechnology alternatives and defining the best directionsalong which to proceed. These recommendations aredescribed in the Additional Crosscutting R&D sectionbelow.

The recycle technology R&D addressing advancedaqueous and pyroprocess technology for the SFR ispresented next.

Recycle Technology R&DThe objective of this R&D is to complete the processdevelopment required to initiate the design of commer-cial fuel cycle facilities for both oxide and metal fuels ofthe SFR. The scales of commercial oxide and metalfacilities are different. An oxide treatment facility wouldlikely be centralized with throughput on the order ofabout 1000 MTHM per year for LWR fuel, or about 100MTHM per year for fast reactor fuel. Collocation of thefuel cycle facility and the reactor plant is not excludedhowever. A metal fuel cycle facility would likely belocated with a fast reactor and have a throughput on theorder of 5 MTHM per year.

Advanced Aqueous Process and Remote CeramicFuel Fabrication. Advanced aqueous reprocessing andadvanced pelletizing are the preferred recycle technolo-gies for the MOX-fueled SFR option. Advanced aque-ous technology is also a viable option for processingLWR spent fuel, enabling the production of initial coreloads for fast reactors.

The advanced aqueous reprocessing option consists of asimplified PUREX process with the addition of auranium crystallization step and a minor actiniderecovery process. A schematic of a closed fuel cyclewith advanced aqueous technology is shown in thefigure on the following page. The purification processof U and Pu in the conventional PUREX is eliminated,and U/Pu is co-extracted with Np with reasonabledecontamination factors (DFs) for recycle use. Theuranium crystallization removes most of the bulk heavymetal at the head end and eliminates it from downstreamprocessing. The main process stream is salt-free, which

55


Repository

Plutonium,Fission Productsand Actinides

OrganicSolvent

Used FuelAssemblies

Fuel Rod

Dissolver

CentrifugalContactor

ShieldedCentral Facility

Cladding Hulls

Purification

FuelFabrication

ActinideSeparation

New FuelAssemblies

Uranium

New Fuel

Recycled Fuel

Nuclear Energy Park

FastReactor

Enrichment andFabrication

ThermalReactors

Shear

WasteformFabrication

Fission Productsand Trace Actinides

02-GA50807-08

UraniumMine

Uranium

United States. However, two process steps and high-level waste volume reduction options have not beenpursued beyond laboratory-scale testing. Further, therecovery fraction of the pyroprocess needs to be in-creased. These are the focus of R&D for thepyroprocess option.

The first needed process step is reduction of actinideoxides to metal. Laboratory-scale tests have beenperformed to demonstrate process chemistry, but addi-tional work is needed to progress to the engineeringscale. The second needed step is to develop recoveryprocesses for transuranics, including plutonium. Withregard to volume reduction, additional process R&D

could potentially increase fission product loadings in thehigh-level waste and reduce total waste volumes.

With regard to achieving the high recovery of transuran-ics, pyroprocessing has been developed to an engineer-ing scale only for the recovery of uranium. Recovery ofall transuranics, including neptunium, americium, andcurium, has so far been demonstrated at laboratory scale.Viability phase R&D is recommended to verify that allactinides can be recycled with low losses.

56


Electrorefining

ShieldedCo-Located Facility

Anode Cathode

New Fuel

Recycled Fuel

UraniumMine

Repository

Cathode Processing

Used Fuel New Fuel

Injection Casting

Nuclear Energy Park

SaltElectrolyte

Recycle

Cadmium

WasteformFabrication

02-GA50807-07

FastReactor

Enrichment andFabrication

ThermalReactors

Fission Productsand Trace Actinides

Adaptations for Other Systems and FuelsThe above processes, aimed primarily at the oxide andmetal fuels of the SFR, will be evaluated and adapted forapplication to other Generation IV systems. This isprimarily an issue at the head end of the process (where,e.g., fuels from the GFR or LFR systems would beconverted to oxide or metal and introduced into theprocesses described above), and at the tail end (where

they would be reconverted to fuel feedstock). Feasibilityevaluations and bench-scale testing would enablecomparisons to be made between the advanced aqueousand pyroprocess options. Specific issues are presentedwith the individual systems.

57


Alternative Process DevelopmentUranium Extraction in Aqueous Processing. Theprincipal aim of the uranium crystallization process stepin advanced aqueous reprocessing is the inexpensiveseparation of bulk quantities of low-enriched uraniumfrom spent fuel from LWRs. The motivation for thisapproach is clear: separating the bulk uranium yields anLWR spent fuel process stream that is reduced in heavymetal content by two orders of magnitude, which offerssignificant potential for volume and cost reduction. Theuranium crystallization technique is the favored technol-ogy in Japan, and it shows considerable promise. Othermeans of removing the uranium component of spentLWR fuel are being explored internationally. Principalamong these is the uranium extraction (UREX) process,which is under development in the United States. InUREX, uranium is extracted in a first step of advancedaqueous processing technology, and the plutonium,minor actinides and nonvolatile fission products are sentto the next process step. The relative advantages anddisadvantages of uranium crystallization and UREXshould be established through R&D activities forinternational comparison and development.

Other Dry Processes and Vibropac Fabrication.Alternative nonaqueous, i.e., dry fuel cycle processeshave been investigated in Russia and more recently inJapan. Examples are fluoride volatility and AIROX.These methods also aim to establish remote fuelrefabrication methods that eliminate the need for re-motely operable and maintainable ceramic pellet fabri-cation production lines through vibratory compaction orvibropac. An R&D activity is recommended to betterdevelop these alternatives.

Additional Crosscutting R&DThe fuel cycle preferred for most of the Generation IVsystems is a full actinide recycle fuel cycle, whereplutonium and all minor actinides are recycled. Thisincludes recycling in symbiotic cycles for managementof spent fuel from current and near-term systems.Recycle of all actinides promises to:

• Reduce long-term waste toxicity source term sent toa geologic repository

• Minimize emplacement of nuclear materials suitablefor weapons use in the repository

• Increase repository capacity by reducing long-termdecay-heat generation and emplacement

• Improve repository performance by reducingradiation damage on the final waste forms.

Two alternatives for recycling the minor actinides fromspent fuel may be considered: (1) heterogeneousrecycle, in which most of the minor actinides are sepa-rated from plutonium and incorporated into new fuel forreactor irradiation, or (2) homogeneous recycle, in whichthe minor actinides and plutonium extracted from spentfuel are incorporated together into new fuel. In eithercase, a fast spectrum reactor (or a liquid fueled reactorsuch as the MSR) is required to consume the minoractinides, during subsequent irradiation. Thermalreactors can be used to consume plutonium in the case ofheterogeneous recycle. Achieving viability of fullactinide recycle requires an integrated approach formanaging minor actinides, which is optimized withrespect to the choice of recycle, refabrication method,and reactor system.

Two specific viability phase R&D activities are recom-mended to help decide the best path for developing fullactinide recycle in the performance phase:

Extractant Development. One of the technology gapsfor full actinide recycle is the initial segregation ofuranium contained in the LWR spent fuel from thetransuranics and fission products that are to be furtherprocessed and recycled. Crystallization and lithiumreduction are the reference options for accomplishingthis, and UREX is an alternate technology under devel-opment in the United States.

R&D is recommended to search for a new extractionagent that could extract the uranium from spent nuclearfuel (SNF), leaving the transuranics and fission productsfor further processing and recycle. If such an extractantcould be found, it may offer considerable simplificationand cost advantages.

After the uranium is segregated from the transuranicsand fission products of the LWR spent fuel, then thetransuranics must be separated from fission products andrefabricated for recycle. Current aqueous processingapproaches use a sequence of different processes toextract each transuranic element one at a time. R&D isrecommended to search for new aqueous extractionagents that could remove the Np, Pu, Am, and Cmtransuranics from an aqueous stream in a single step. Ifthis could be achieved, full actinide recycle cost, acci-dent risk, proliferation vulnerability, and developmentrequirements may be dramatically reduced for aqueousprocesses.

58


and institutional requirements for the management ofseparated Cs and Sr, to analyze the costs and benefits,and to determine the preferred decay heat managementoptions.

Integrated Once-Through Fuel Cycles. In the earlyyears of the nuclear power industry, it was thought thaturanium was a scarce resource. The reactors and fuelcycle were developed with the assumption that SNFwould be rapidly processed for recovery of plutoniumand uranium, and the operations at the back end of thefuel cycle, including repository designs, considered thatonly high-level waste would be disposed of. Thisassumption later reversed as many countries changed toa once-through fuel cycle, but some of the back endoperations remained unchanged. If one were to redesignthe once-through fuel cycle, it might be significantlydifferent than the current practice. Further, some of theGeneration IV systems are once-through, which couldbenefit from R&D into new approaches.

For a redesigned once-through fuel cycle, the desiredcharacteristics are as follows:

• Reduced handling of SNF to reduce cost and risks,and improve security and safeguards

• Reduced storage of SNF in reactor pools withenhanced physical security and reduced capital costsfor spent fuel storage in reactors

• Earlier placement of SNF in geological repositories

• Repositories that would allow easy recovery of SNFif conditions were to change. This is termed an openfuture repository; safe disposal is assured andcommitments by future generations to ensure safeSNF disposal are minimized, while at the same timesociety retains an option to retrieve and recycle theSNF if conditions change.

Recent technical developments suggest that once-through systems with such characteristics are possibleand may be more economical than the current system.An element of such a system is a multipurpose self-shielded cask loaded at the reactor with SNF and neverreopened. The cask is used for storage, transport, anddisposal but uses different overpacks during storageversus during disposal—to meet the differing require-ments of storage of SNF after short cooling times versuslong-term disposal. The repository is modified to allowearly placement of SNF.

Some, but not all, of the technology is in existence forsuch a system. R&D is recommended to establish theviability of key technologies: (1) controlling spent fuel

Homogeneous versus Heterogeneous Recycle ofMinor Actinides. Homogeneous and heterogeneousrecycle are introduced in an earlier section.Hetergeneous recycle offers additional flexibility of thetreatment of the streams, yet segregated minor actiniderefabrication and recycle would entail handling of thehighly radioactive minor actinides undiluted by pluto-nium. R&D during the viability phase is recommendedto evaluate the technological and cost implications ofheterogeneous minor actinide recycle using curium asthe example. Curium is a difficult actinide to recyclebecause it produces the highest decay heat and neutronsource per unit mass, and it has a very small criticalmass, which restricts the process batch size. Thisrecommended crosscutting R&D activity seeks toquantify important aspects of the tradeoffs betweenheterogeneous and homogeneous minor actinide man-agement for the case of full-actinide recycle.

Cesium and Strontium Heat Management. For thefirst 50 to 100 years after SNF is discharged from areactor, the cesium and strontium are the primarysources of decay heat, the strontium is the primaryingestion hazard, and the cesium is the primary gammasource. These two fission products decay away withabout a 30-year half life. If these radionuclides, whichare destined ultimately for geologic disposal, wereprocessed and managed separately, several benefitscould accrue:

• A given repository capacity might be increased,because capacity is primarily determined by heatload, and delay in emplacing the main short-termheat source would increase capacity

• Radiation shielding of some process operations,waste transport, and waste disposal would decrease

• A significant short-term hazard from strontiumwould not enter the repository waste stream.

Because of the limited lifetime of cesium and strontium(except 135Cs) and given their high importance to heatloading, inexpensive methods may be developed tohandle these wastes at the fuel cycle back end.

Many alternatives exist for heat management in once-through and recycle fuel cycles. For once-through,interim storage and interim active repository cooling areoptions. For recycle, the waste forms that contain Cs/Srcould be held in interim storage before repositoryemplacement or dual repository designs; one for lowheat and one for high heat waste forms could be consid-ered. R&D activities are recommended to addressscientific, engineering, and geological disposal issues

59


temperatures in large casks with short-cooled spent fuel,(2) components meeting requirements for storage,transport, and disposal, including restrictions on choiceof materials allowed in a repository, (3) behavior ofspent fuel over long periods of time inside a cask, and(4) repository designs that allow placement of shorter-cooled spent fuel without adversely impacting repositorycapacity. The repository becomes a managed facility fora period of time during which it has the characteristicsof a combined storage and disposal facility.

Sustainability Evaluation Methodology. Quantitativemetrics were developed during the roadmap for fuelutilization and waste minimization, but not for environ-mental impacts of the fuel cycle. For fuel utilization andwaste minimization, objective formulas were derived

and can remain the basis for evaluation. In the case ofenvironmental impacts, the methodology that exists forthe preparation of preliminary environmental impactstatements can be adapted for evaluations. Therefore, inprinciple no further development of evaluation methodsis identified in sustainability. However, noting that anumber of countries define sustainability in broaderterms, additional R&D to develop methodology for thesebroader frameworks may be desirable for individualcountries.

Crosscutting Fuel Cycle R&D Schedule andCostsA schedule for the crosscutting fuel cycle R&D is shownbelow, along with the R&D costs and decision points.

FUEL CYCLE CROSSCUT (230 M$)Advanced Aqueous (70 M$)

Pyroprocess (100 M$)

Alternative Process Development (10 M$)Aqueous Group Extractant Development (10 M$)

Systems Evaluation of Homogeneous vs. Heterogeneous Recycle (10 M$)

Cs/Sr Management Strategy (10 M$)

Integrated Once-Through Fuel Cycles (10 M$)

Sustainability Evaluation Methodolgy (10 M$)

Head end process UNH crystallization technology Minor actinide recovery technology Adequacy of actinide recovery fraction (Adv. Aqueous) (FC 1) Main equipment design High level and TRU waste reduction

Process materials selection Oxide SNF reduction (including head end) Applicability of pyro-recycle to LWR spent fuel (FC 2) Electro-refiner development Refabrication process Process waste reduction Adequacy of actinide recovery (pyroprocessing) (FC 3) Waste form development and qualification Material control and accountability

Extractant molecule design campaign Surrogate bench testing Hot cell testing Feasibility of group extraction of actinides in aqueous process (FC 6)

Full-scope life cycle evaluation of Cm management strategies Cm target fabrication option study screening and option selection Hot cell testing

Systems study near-term heat management options and effects Recommendation on separate management of Cs, Sr (FC 4)

System study once-through open-future integrated fuel cycles Approach for integrated management of once-through cycle (FC 5) Design option trade study for variable heat removal cask design Design option trade study of ventilated repository concepts

decision

decision

decision

decision

decision

decision

2000 2010 2020

60


Crosscutting Fuels and Materials R&DIntroduction and ApproachThis section addresses crosscutting R&D on fuels andmaterials. To introduce this area, a few observations arefirst established:

• All Generation IV systems project in-service andoff-normal temperatures that are beyond currentnuclear industry experience, as well as most previ-ous experience with developmental systems. Allrequire relatively long service lifetimes for materialsand relatively high burnup capability for fuels.

• Most systems call for use of fast and epithermalneutron spectra, which will challenge materialsperformance with increased radiation damage.

• Even for systems with different coolants, manyapplications have important similarities, such astemperature, stress, and neutron spectra. Thissuggests the opportunity to survey similar materials,or classes of materials, for use in Generation IVsystems. The following table indicates classes ofmaterials proposed for the systems.

Candidate MaterialsFuels and materials that meet the requirements ofGeneration IV systems must be identified, and databasessufficient to support design and licensing must beestablished. Some applications are similar to nonnuclearapplications, which can provide a basis for identifyingcandidate materials. A summary of the fuels andmaterials options considered for each of the systems isprovided in the table on the next page. The table reflectsinitial suggestions based on experience, but for manyapplications few data are available to support the recom-mendation of a specific alloy or material.

The lack of data for the proposed materials suggests thata broad-based materials R&D program will serve theinitial development of the systems. The proposed R&Dactivities will provide information and property data thatpertain to multiple Generation IV systems. Theseactivities should be crosscutting early in the research,but are expected to become more system specific as thesystems are developed. A broad selection of data needsshould be considered, such as measurements of nucleardata to support the systems design and safety analysis.

Fuel Materials Structural Materials

System

GFR S P P P P P P P

MSR P P P S S

SFR P P P P P

LFR S P P P S S S

SCWR-Thermal P P P S S

SCWR-Fast P S P P S S

VHTR P S P P S P

P: Primary Option

S: Secondary Option

Oxi

de

Met

al

Nitr

ide

Car

bide

Fluo

ride

(liqu

id)

Ferr

itic-

mar

tens

itic

Stai

nles

s St

eel

Allo

ys

Aus

teni

ticSt

ainl

ess

Stee

lA

lloys

Oxi

deD

ispe

rsio

nSt

reng

then

ed

Ni-b

ased

Allo

ys

Gra

phite

Ref

ract

ory

Allo

ys

Cer

amic

s

61


Structural Materials

System Spectrum, Toutlet Fuel Cladding In-core Out-of-coreGFR Fast, 850°C MC/SiC Ceramic Refractory metals Primary Circuit:

and alloys, Ceramics, Ni-based superalloysODSVessel: F-M 32Ni-25Cr-20

Fe-12.5W-0.05CNi-23Cr-18W-0.2CF-M w/ thermalbarriersTurbine:Ni-based alloysor ODS

LFR Fast, 550°C and MN High-Si F-M, High-Si austenitics,Fast, 800°C Ceramics, or ceramics, or

refractory alloys refractory alloys

MSR Thermal, Salt Not Applicable Ceramics, refractory High-Mo Ni-base700–800°C metals, High-Mo alloys (e.g., INOR-8)

Ni-base alloys(e.g., INOR-8),Graphite, Hastelloy N

SFR Fast, 520°C U-Pu-Zr F-M (HT9 or ODS) F-M ducts Ferritics, austenitics(Metal) 316SS grid plate

SFR Fast, 550°C MOX ODS F-M ducts Ferritics, austenitics(MOX) 316SS grid plate

SCWR- Thermal, 550°C UO2 F-M(12Cr, 9Cr, etc.) Same as cladding F-MThermal (Fe-35Ni-25Cr-0.3Ti) options

Incoloy 800, ODSInconel 690, 625,& 718

SCWR Fast, 550°C MOX, F-M (12Cr, 9Cr, etc.) Same as cladding F-M-Fast Dispersion (Fe-35Ni-25Cr-0.3Ti) options

Incoloy 800, ODSInconel 690 & 625

VHTR Thermal, 1000°C TRISO ZrC coating and Graphites Primary Circuit:UOC in surrounding PyC, SiC, ZrC Ni-based superalloysGraphite graphite Vessel: F-M 32Ni-25Cr-20Fe-12.5Compacts; W-0.05CNi-23Cr-18ZrC coating W-0.2CF-M

w/ thermal barriersTurbine: Ni-basedalloys or ODS

Abbreviations:F-M: Ferritic-martensitic stainless steels (typically 9 to 12 wt% Cr)ODS: Oxide dispersion-strengthened steels (typically ferritic-martensitic)MN: (U,Pu)MC: (U,Pu)CMOX: (U,Pu)O2

62


Irradiation Testing of Fuels and MaterialsBased on previous experience with development of fastreactor fuels and based on the range of maturity of theproposed fuel forms, varying needs for fuel developmentexist. In general, a long-term program to develop fuelsentails the following activities: (1) fabrication processdevelopment, (2) property measurement and assessment,(3) irradiation testing and safety demonstration, and (4)modeling and predictive code development.

Four phases of development for the fuel are recom-mended as follows, which include the activities above tovarying degrees:

• Fuel candidate selection

• Fuel concept definition and feasibility

• Design improvement and evaluation

• Fuel qualification and demonstration

The development for structural materials follows asimilar path.

Irradiation Tests. All systems will need irradiationtesting of fuels and materials for in-core components.The similar service conditions for systems and thelimited availability of irradiation test facilities world-wide are two strong reasons to recommend a crosscut-ting irradiation testing program. The availability of fast-spectrum test facilities is a particular concern. Theprogram should comprise tests and experiments atreactors in several countries with needed postirradiationexamination and testing. The recommended R&Dactivities are summarized below:

• Inert environment tests of unirradiated andpreirradiated structural material samples at relevanttemperatures to assess radiation effects on mechani-cal behavior (strength, creep, fracture toughness)over the temperature range of interest and indepen-dent of coolant-induced phenomena.

• Special-effects irradiation tests in laboratoriessimulating the effect of neutrons and fission prod-ucts on material microstructures using ion beams.Such tests would be used as a low-cost means forassessing microstructural evolution in structuralmaterials or in matrix materials proposed for disper-sion fuel concepts. These tests might include swiftion irradiation or fission product and heliumimplantation.

• Irradiation tests of material samples in prototypicneutron spectra and in flowing coolant loops (or

flowing fuel loops, in the case of the MSR) arefundamental to assess the effects of environmentaldegradation (e.g., due to radiolysis-enhancedcorrosion and in situ radiation damage) on materialsproperties and performance. System-specificcorrosion and environment testing of preirradiatedsamples would provide a low-cost means of assess-ing the impact of radiation damage on environmen-tal degradation of performance.

• Preliminary tests of new fuel designs (either newfuel forms or new compositions) in a speciallyconfigured vehicle in a test reactor to identifyirradiation performance issues.

• Irradiation tests of prototypically designed test fuelsto determine fuel lifetime and life-limiting phenom-ena in proposed fuel designs.

• Irradiation tests of reference fuel designs at condi-tions of power and temperature that determine limitsfor safe and reliable operation of fuels. This infor-mation will be essential for supporting a licensingcase for a first-of-a-kind reactor.

Many of the above activities are more fully described inthe R&D recommended for each system. Other cross-cutting R&D is discussed next.

Transient Testing of Reactor FuelsAll Generation IV systems will require transient testing.Fuels that are in initial development stages will requiretransient testing, independent of design-basis accidentissues, to understand transient response and to aid designchanges that ensure required safety-related behavior.Fuels that have matured to the point of reference designswill require transient testing under a range of accidentconditions, including those beyond the design basis, todetermine mechanisms that lead to fuel failure, thresholdconditions at which failure occurs, and the relocation/dispersal behavior of failed fuel under bounding accidentconditions. Fuels with established performance data-bases will require testing at specific design basis acci-dent conditions to verify that behavior in the system is asexpected, which will be an important step in qualifyingthe fuel for licensing. Crosscutting R&D is recom-mended to establish a transient testing capability toserve common needs.

Fuels and Materials Selection and PerformanceBecause many classes of materials are candidates, anactivity to determine the intrinsic properties of materialsand their irradiation-performance is recommended.Major activities are described below.

63


Mechanical Performance and Dimensional Stability.R&D is recommended to study and quantify mechanicalperformance and dimensional stability properties. Forthe range of service conditions expected in GenerationIV systems, including possible accident scenarios, theproposed materials must meet design objectives in thefollowing areas:

• Dimensional stability, including void swelling,thermal creep, irradiation creep, stress relaxation,and growth

• Strength, ductility, and toughness

• Resistance to creep rupture, fatigue cracking, andhelium embrittlement

• Neutronic properties for core internals

• Physical and chemical compatibility with thecoolant

• Thermal properties during anticipated and off-normal operations

• Interactions with other materials in the systems.

For each design objective, the fundamental microstruc-tural features that establish performance (such asdislocation microstructure, void microstructure, precipi-tate microstructure, and radiation-induced segregation)must be understood to allow for further performanceimprovements. The formation and behavior of thesefeatures depend on materials temperature and neutronflux and spectrum. For example, higher-energy neutronspectra induce more radiation damage into the micro-structures of materials, which impacts the formation ofand phenomena associated with microstructural featuresthat degrade properties. At elevated temperatures,radiation damage is more quickly annealed. An addi-tional objective is to limit impacts of neutron activationof components, which can complicate maintenance,handling, and disposal of irradiated components, throughcareful selection of material constituents.

Candidate alloys for the 300–600°C temperature rangeinclude austenitic iron- and nickel-base alloys, ferritic-martensitic alloys and oxide-dispersion strengthenedferritic and austenitic alloys. The primary materialscandidates for 600–900°C range are those with goodstrength and creep resistance at high temperatures, suchas oxide-dispersion strengthened ferritic-martensiticsteels, precipitate-strengthened iron- or nickel-basesuperalloys, coated materials, or refractory alloys ofmolybdenum, niobium, and tantalum. Materials issuesfor applications at temperatures exceeding 900°Cbecome increasingly severe. Of the potential metallic

materials, only tungsten- and molybdenum-basedsystems are believed to have the potential to operate inthis temperature range. However, the potential limita-tions of metallic alloys at higher temperature motivateconsideration of ceramic materials. The extreme tem-peratures also present challenges for conducting experi-ments in existing irradiation facilities.

Materials for Balance-of-Plant. The materials to beselected for balance-of-plant components will be chal-lenged by high operating temperatures and compatibilityissues that are introduced with alternative energy prod-ucts. For example, generation of hydrogen will entailenvironments that are potentially corrosive orembrittling to some materials.

Materials for Fuel Recycle Equipment. Althoughmuch of the emphasis of this section is on fuels andmaterials for reactor systems, the success of the fuelrecycle technologies will depend upon selection ofmaterials that allow fuel processing and fabricationunder harsh environmental conditions, such as hightemperature, radiation fields, and aggressive chemicalenvironments. In addition, the selected materials mustresist interaction with the recycled fuel media, which isessential to achieving low loss of actinides to secondarywaste streams. Therefore, a crosscutting materials R&Dactivity associated with recycle technology is recom-mended.

Dispersion Fuels. Traditional fuel forms appear in mostGeneration IV systems as preferred options. However, itis recommended that less mature fuels forms, dispersionfuels in particular, may be explored as part of thecrosscutting R&D. Specific systems that this fuel formmight benefit include the SCWR, GFR, VHTR, SFR andLFR systems.

Fuels and Materials ModelingThe design of new alloys for Generation IV systems isan extensive undertaking requiring considerable re-sources. Experimental programs will be limited by theamount of available resources, thus limiting the data orperhaps the degree to which prototypic conditions andgeometries can be studied. The capability to modelmaterial properties and performance will be valuable forguiding experimentation, interpreting experiments, andincreasing the understanding of proposed alloy systemproperties and performance. Modeling of microstructureevolution under irradiation is recommended to improvethe understanding of the response of various alloysystems to the higher-temperature and dose conditions.

64


Fabrication Processes and TechniquesJoining Techniques. Little experience with fabricationand joining exists for many of the metallic and ceramiccomponents proposed for Generation IV systems.Therefore, R&D is recommended to assess and developapplicable joining techniques.

Fabrication of Ceramic Fuel. Four of the fuel optionsare ceramic fuels. The R&D plans for the mixed oxide-fueled sodium-cooled reactor include development ofthe simplified pelletizing method, which is intended toprovide a fabrication scheme that is simpler and lesscontamination intensive than the currently used tech-niques. A modest R&D activity to consider whether thesimplified pelletizing method can be extended to mul-tiple ceramic fuels is recommended. Similarly, thevibrational compaction technique of fabricating ceramicfuels is an alternative for fabricating MOX fuel forsodium-cooled reactors. R&D to consider application ofthe vibrational compaction technique to other Genera-

tion IV ceramic fuels is also recommended. R&D intoceramic fabrication process for composite ceramic fuelsshould also be considered to yield new alternatives forthe systems.

Establishment of Standards and Codes. BecauseGeneration IV systems will require deployment ofmaterials and components operating under new condi-tions, codes and standards must be established for theiruse. Materials composition and property data that arecollected during the development of Generation IVtechnologies should be obtained in accordance withquality assurance standards so that they may provide thenecessary basis for codes and standards, and for licensing.

Crosscutting Fuels and Materials R&D Scheduleand CostsA schedule for the crosscutting fuels and materials R&Dis shown below, along with the R&D costs and a deci-sion point.

Conduct measurements of mechanical & corrosion properties of unirradiated alloys Conduct irradiations & measurements to establish microstructural/chemical stability of alloys Conduct mechanical testing & corrosion and stress corrosion cracking experiments on irradiated samples Conduct in-pile tests to assess mechanical & corrosion aspects

Measure fuel properties required for viability R&D Measure fuel properties required for performance R&D

Develop simplified pelletizing method Develop parameters for vi-pac fabrication of ceramic fuels

Adapt or develop, as necessary, joining techniques Establish non-destructive inspection techniques

Define requirements for irradiation and transient testing Requirements for irradiation and transient test facilities decision (FM 1) Design and construct test vehicles for fast-spectrum testing of fuel and/or materials samples Design and construct test loops for use in thermal-spectrum test reactors Irradiate materials samples in a fast neutron test vehicle

FUELS & MATERIALS CROSSCUT (220 M$)Irradiation Testing Preparation (50 M$)

Structural Material Properties & Behavior (50 M$)

Fuel Materials Properties (30 M$)

Reactor Fuel Transient Testing (20 M$)Fuel Fabrication Technique Development (10 M$)

Materials Phenomena Modeling (10 M$)Joining Techniques & Non-destructive Evaluation (30 M$)

Codes and Standards (20 M$)

2000 2010 2020

65


the Ca-Br process as an alternative if it can be developedas a cost-effective method. Nearer-term technology forhydrogen production may be possible through steamreforming of methane or hot electrolysis of water.Nearer-term opportunities for hydrogen use includepetroleum refining. This may require plants with sizesranging from 50 to 500 MWth. However, the mostrecently ordered hydrogen production plants (using steamreforming of natural gas) are systems up to 2000 MWth.

Generation IV systems may potentially be used for avariety of process heat applications: urea synthesis,wood pulp manufacture, recovery and de-sulfurization ofheavy oils, petroleum refining, manufacture of naphtha,ethane and related products, gasification of coal, andmanufacture of iron, cement, or glass. The minimumrequired temperature for some of these applications isabout 600°C. So, the GFR, MSR, LFR, and VHTRsystems could potentially serve them.

Crosscutting Energy Products R&DIntroduction and ApproachMost Generation IV systems are aimed at technologyadvances that enable high operating temperatures. Thehigh temperatures will allow the production of newproducts such as hydrogen and process heat, as well aselectricity production with higher efficiency cycles.This section addresses the crosscutting R&D needed forthese new products and cycles for Generation IV systems.

The table below summarizes the energy productiontechnology options for each Generation IV system. Thechoice among hydrogen production technologies is mostclosely linked with the effective temperature that heatcan be delivered from the reactor, which is a function ofthe outlet temperature and the heat transfer properties ofthe primary coolant, or secondary coolant in the caseswhere an IHX is required.

Generation IV Hydrogen Advanced CyclesSystem (Toutlet) Production Heat Delivery for Electricity Production

I-S Ca-Br Process Desali- Supercritical Water RankineProcess Process Heat nation CO2 Brayton Supercritical Helium Brayton

GFR (850ºC) P S S O P

MSR (700-850ºC) P S S O P

SFR (550ºC) O S

LFR (550ºC) O P S

(800ºC) P S O S1 S1

SCWR (550ºC) O P

VHTR (1000ºC) P S O P

P: Primary option 1 Bottoming cycle using heat at lower temperatures available after higherS: Secondary option temperature heat has been used for hydrogen production.O: Option for all systems

The entries in the table are primarily determined by theoutlet temperature and the choice of coolant. Forexample, the I-S process for hydrogen needs heatdelivered above 800°C, and the process efficiencyimproves above this temperature. The GFR anticipatesoutlet temperatures of 850°C, and the VHTR anticipates1000ºC and the MSR has the potential to reach 850°C.The SCWR, LFR, and SFR deliver heat below 800°C,and therefore do not consider using the I-S process. TheLFR system proposes development of a lower tempera-ture Ca-Br process as an alternative for hydrogenproduction, which may potentially produce hydrogen attemperatures above 700ºC. Others may consider using

The need to provide potable water to the expandingpopulation in arid regions is potentially an emergingapplication for nuclear power. Removal of salt and otherimpurities from seawater or brackish waters generallyuses one of two basic approaches: distillation or pro-cessing through membranes. These methods typicallyrequire heat input at 80–120°C and electricity to operatethe pumps. Nuclear sources may also potentially serveas heat sources for district heating. The temperaturesrequired are typically low, of the order of 80°C. Thus,all six Generation IV systems may consider bottomingcycles that would include desalination and districtheating.

66


For the generation of electricity, the supercritical waterRankine cycle is a central feature of the SCWR systemat 550ºC. The LFR system could also use thesupercritical water Rankine cycle, but this system hasthe potential to improve its efficiency with asupercritical CO2 Brayton cycle. The GFR, MSR, andVHTR could use advanced helium Brayton cycles.

Two generic issues arise for energy products, underscor-ing a need for R&D. These are discussed next, followedby a survey of energy production technology R&Drecommendations.

Product Purity. Three of the potential Generation IVenergy products—hydrogen, fresh water, and districtheat—go directly to consumers. For these, productquality and potential contamination are issues, with themost probable concern being tritium. Two sources oftritium must be considered: it is a ternary fissionproduct and potentially an activation product in theprimary coolant. For desalination and district heating,this is less of an issue because the low temperaturesinhibit tritium diffusion through intermediate heatexchangers. Hydrogen production is in a temperaturerange of concern, where the diffusion of tritium throughhigh-temperature heat exchangers and other componentsis difficult to limit.

The best approach is to avoid tritium generation, whichis primarily accomplished through choice of materials.In addition, R&D is needed to determine how to limittritium diffusion through coatings or barriers or how toseparate tritium at various stages. Tritium can beseparated from hydrogen by using purification systems.However, this may have a significant impact on hydro-gen cost and should be avoided if possible.

Integrated System Safety. R&D is recommended toaddress the integrated safety requirements of a nuclearsource with a hydrogen production or process heat plant.This will require close interaction with the chemical andrefining industries. One R&D approach is to examinehow risk is evaluated in the chemical industry, andintegrate and reconcile it with the risk and safety re-quirements for nuclear installations. In addition, me-chanical systems such as fast acting isolation valvesmust be developed to be placed in the line leading tochemical plants. Other new requirements may emergeconcerning reliability of heat exchangers as well to meetthese integrated plant safety needs. For the chemicalplants, it will be necessary to thoroughly understandenergetic accidents utilizing deterministic and probabi-listic risk assessment (PRA) approaches. For the reactorevents beyond the design basis, accidents must beassessed using PRA methods.

As the requirements for other energy products andapplications are more specifically defined, furthercrosscut issues will emerge. Additional R&D may beneeded to address these emerging needs.

Iodine-Sulfur (I-S) Process Technology R&DThe I-S process involves three component chemicalreactions in a thermochemical water-splitting cycle forthe production of hydrogen. The system creates H2SO4and HI, separates the acids, and carries out reactivedecomposition of HI and concentration and decomposi-tion of H2SO4. The sulfuric acid can be decomposed atabout 825ºC, which defines the temperature of heataddition.

Materials and Database. Currently, the I-S processtechnology requires temperatures in the range of 800–900°C. R&D needs include thermochemical propertymeasurements and databases, rate constant measure-ments for the chemical processes, measurement ofthermodynamic equilibrium data, thermodynamicoptimization, and development of flowsheets. R&Dneeds also include materials compatibility, corrosion,and lifetime. Appropriate materials must be selected andtested. Additional work would involve studies onensuring product quality, investigation of membrane andsubstrate technologies, effects on mechanical properties,and determination of any surface modifications.

Bench Scale and Pilot Scale Testing. Additionalactivities are recommended to design, build, and operatea laboratory-scale, completely integrated, closed-loopexperiment driven by a nonnuclear heat source. Thisscale would produce hydrogen and oxygen at about 1-10liters per hour, and provide proof of principle andverification of the chemical reactions in the closed cycle.

Following bench-scale testing, a pilot plant will need tobe operated using prototypical materials and technolo-gies. The pilot plant would also operate on nonnuclearheat to demonstrate the technologies and materials of afull-size plant.

Calcium-Bromine (Ca-Br) Process TechnologyR&DThe calcium-bromine process has the advantage ofoperating at a lower temperature than the I-S process, inthe range of 725–800°C. However, the Ca-Br processuses four gas-solid reactions that take place in stationarybeds, and is less efficient than the I-S process due to thelower temperatures. The heat necessary to drive hydro-gen generation is supplied to a gas stream that contains alarge excess of high-pressure steam. Hydrogen andoxygen are removed from the gas stream through

67


in coal plants. The application of the supercritical steamRankine cycle to Generation IV systems requiresexamination of several key interfaces, such as thedevelopment of in-vessel steam generators for the LFRsystem.

Process Heat Interface R&DA minimum temperature of 600°C was chosen forprocess heat applications or production of high-qualitysteam for industrial use. Research is underway on anumber of processes other than thermochemical, such asdirect-contact pyrolysis and conversion of agriculturalfeedstock, which may further reduce the temperaturerequirements.

R&D is needed for high-temperature heat exchangersinvolving gas-to-salt, liquid-metal-to-salt, orsupercritical-steam-to-salt. These are an alternative formany of the Generation IV systems, but numerousperformance requirements differentiate them. Forexample, some have large pressure differences across theIHX, high pressures, and challenges in corrosion.

Desalination and District Heating Interface R&DThis area of R&D considers desalination to producefresh water. With regard to desalination, multipleapproaches are possible either through direct use of lowtemperatures heat (120°C) or through optimized reverseosmosis processes. With regard to district heating, anuclear-supplied district heating network has operatedfor almost two decades in Switzerland. This provides avaluable benchmark for evaluating district heatingapplications. Many cities in Eastern Europe, Russia, andthe Former Soviet Union are already equipped with adistrict heating infrastructure.

In the Brayton cycle, coolant temperatures in the heatexchanger range from 150°C down to 30°C and dis-charge heat to the low-temperature heat sink. In thermo-chemical processes such as the I-S process, heat in therange of 100–150°C is available. Thus, the Braytoncycle and thermochemical processes for hydrogenproduction may potentially be combined with desalina-tion, district heating, or numerous other process-heatapplications as a co-generation system without reducingthe thermal efficiency of electricity generation orhydrogen production. R&D is recommended to explorethe impact on the overall plant design and optimization.

semipermeable membranes. The stationary beds arearranged with four sets of cross-over valves to alternatethe gas flow through the CaO/CaBr and FeBr2/Fe3O4beds.

Materials and Database. There are two sets of issuesin the development of the Ca-Br process for furtherresearch. First, the process reactants (steam, hydrogen,and hydrogen bromide) at 600–750°C will requirematerials research to determine corrosion/erosionmechanisms and kinetics. Materials will have to beselected and tested for piping and vessels.

The second set of issues pertain to the reactions and theirkinetics. Support structures for the beds must be devel-oped, and the reaction kinetics as a function of condi-tions and structures must be determined. For the pro-cess, the use of stationary beds with cross-over valveswill require development and pilot-plant operation todetermine whether the alternating flow through the bedswill have an effect on reactor operation. A fluidized bedalternative, which avoids alternating flow, should beinvestigated.

Repeated chemical and thermal cycling of the solidmaterials may also lead to cracking and the formation ofdust in the process stream. Pilot-plant operations arerecommended to develop techniques for avoiding dustformation or needed dust removal. Dust is also an issuein the operation of the semipermeable membranes for H2and O2 separation. Pilot-plant operation is recom-mended to test the membranes in realistic chemical,temperature, and dust operating environments.

Supercritical CO2 Brayton Cycle TechnologyR&DThe supercritical CO2 Brayton cycle offers the potentialfor surpassing 40% energy conversion, even at the moreconventional 550°C coolant temperature. The R&Drequired to show viability of this innovation includes (1)confirmation of materials selections from other indus-tries already using sub- and supercritical CO2, (2)thermodynamic optimization of the cycle, (3) design ofthe recuperator and of the heat exchangers, and (4)design and pilot testing of a small-scale turbine orturbine stage and transient testing of a small integratedpower plant. A limitation of the use of supercritical CO2exists due to dissociation at temperatures above 700°C.

Supercritical Steam Rankine Cycle TechnologyR&DFor the generation of electricity, the supercritical waterRankine cycle is already found in industrial use, notably

68


Crosscutting Energy Products R&D Scheduleand CostsA schedule for the crosscutting energy products R&D isshown below, along with the R&D costs and decisionpoints

Requirements for hydrogen production Requirements for hydrogen production (EP 1) Examine sources of tritium to the final product Determine impact of factory construction Develop ability to operate different duty and reactor cycles Determine optimum size, performance, heat losses Examine use of common facilities Optimize co-generation systems Match reactor parameters to industrial needs

Materials selection Therm/chemical properties measurements & database Rate constant measurements Thermodynamic optimization and flowsheet Bench scale integral test Small scale prototype test Practicality demonstration of H2 thermochemical production (EP 2) Ca support selection (specific for Ca-Br process)

Thermodynamic optimization Materials selections: HX, recuperator, turbine Small scale testing: turbine, recuperator

Review fossil plant experience Monitor work by others on SC steam rankine cycle Economic comparisons

Develop models/adapt IAEA model for nuclear desalination Monitor R&D progress by others on reverse osmosis and multi-effects distillation gers, crud control and brine disposition

Product quality requirements

Economics requirements

Integrated system safety requirementsOperational requirements

decision

decision

ENERGY PRODUCTS CROSSCUT (190 M$)

Product Requirements (10 M$)

Thermochemical Water Cracking (100 M$)

Supercritical CO2 Brayton Cycles (20 M$)

Supercritical Steam Rankine Cycle (20 M$)

Process Heat Interface (30 M$)Desalination/District Heating (10 M$)

Design requirements

Monitor developments by others of multi-stage flash heat exchan

Evaluate commercial opportunities for coupling to product

2000 2010 2020

69


Crosscutting Risk and Safety R&DIntroduction and ApproachCrosscutting R&D for risk and safety is considered inthis section. Recommendations are made for researchactivities that are relevant to the viability and perfor-mance assessments of future nuclear energy systems inmeeting the three Generation IV safety and reliability(SR) goals.

Under SR Goal 1, research focuses on those events ofrelatively high to moderate frequency that affect workersafety, facility reliability and availability, and thefrequency of accident initiating events. Under SR Goal 2,research focuses on those low-probability event se-quences that can lead to core degradation, or in otherfacilities to the release of radionuclides from their mostimmediate confinement, or to nuclear criticality withrisk for undue exposures. Under SR Goal 3, researchfocuses on those very low probability accident se-quences where significant core degradation or otherrelease could occur, and the performance of additionalmitigation measures that reduce and control releasesoutside the facility and doses to the public.

Generally, few viability phase R&D issues exist thatcrosscut multiple systems, primarily because viabilityissues tend to involve unique and less understoodcharacteristics of specific systems. The crosscuttingissues that do emerge arise primarily from SR Goal 3,and from the need to have a consistent methodology forSR viability assessment of systems where detaileddesign information is not available. The opportunity touse common test facilities to conduct crosscuttinginvestigations of fuel transient behavior, including fuelfailure and dispersal mechanisms in accidents beyondthe design basis, is described in the roadmap section onCrosscutting Fuels and Materials R&D. That researchbears directly on SR goals 2 and 3.

Different nuclear energy systems employ differentstrategies to meet the specific SR goals. However, by theend of the viability phase R&D, each system must havea safety case that identifies initiators and strategies forresponse. A standard methodology is needed to provide aconsistent evaluation with respect to the Generation IVSR goals for these different strategies. The capability toaccurately calculate safety margins and their uncertain-ties from all sources will play an important role in theviability and performance evaluations of Generation IVsystems, because it will provide a quantitative basis foroptimization of their designs.

At the time of SR viability evaluation for a givenGeneration IV system, the design of the reactor and fuelcycle facilities must have sufficient detail to allowcomprehensive description of the implementation of thelines of defense that provide defense in depth, includingmeasures available to mitigate the consequences of coreand plant degradation during design extension condi-tions (formerly beyond design basis). The design detailmust also allow use of simplified PRA to identify designbasis accidents and transients as well as the highlyhypothetical sequences. The detail should be sufficientto identify and rank phenomena of importance totransient response and to specify experimental informa-tion required to validate transient models. The table onthe next page summarizes the level of design detailrequired for this evaluation.

Crosscutting SR Viability Phase R&DSystem Optimization and Safety AssessmentMethodology. Generation IV viability evaluations willbe performed with incomplete design information. Forthese evaluations, the deterministic concept of defencein depth needs to be integrated with simplified probabi-listic considerations (e.g., systems reliability and proba-bilistic targets) to provide metrics for acceptability and abasis for additional requirements, and to ensure a well-balanced design. This methodology must explicitlyidentify the assumptions and approximations used in thesimplified process, to ensure that these assumptions andapproximations are addressed during performance R&D.Several Generation IV systems have unique, newassessment issues. For example, many employ passivesafety characteristics and systems to a much greaterextent than current nuclear facilities. The failure ofpassive components requires a complex combination ofphysical and human factor ingredients. This poses anissue for PRA methodology because there is less experi-ence in modeling passive systems compared to activesystems. Moreover, system-specific operating data aresparse and may not provide statistically usefulinformation.

The Code Scaling, Applicability, and Uncertainty(CSAU) method can in principle treat such problems, buthas thus far been applied primarily to LWRs and re-quires more extensive design and modeling informationthan is available during the viability phase. ModelingGeneration IV systems requires improved approaches topredict events of extremely low probability or eventsthat arise from incomplete knowledge of potentialsystem interactions and human factors. Researchfocused on the factors that affect the reliability, and

70


ability to predict reliability, of passive safetycomponents and interactions between compo-nents has the potential to improve the qualityof the viability evaluations. In addition, sucha methodology should take into accountcoupling of Generation IV nuclear systems toalternative energy product plant systems.

Emergency Planning Methods. By virtue oftheir relatively small accident source terms,very slow transient response, low uncertaintyin accident phenomenology, and extremelylow probability for the scenarios resultingsignificant radionuclide release off site,several Generation IV systems could poten-tially benefit from emergency planningtailored to their characteristics. Specifically,it has been proposed that emergency planningzone radii or other planning actions differentthan that used for existing reactors, as well asalternative severe accident mitigation meth-ods such as filtered confinements, could beappropriate for some of the Generation IVsystems.

R&D is recommended to define the technicalbasis underlying existing emergency plan-ning. The technical basis should be used toestablish methods for the design and analysisof Generation IV systems to demonstrate thatall design basis transients, accidents, anddesign extension conditions have beenidentified, that transient analysis has suffi-ciently low uncertainty, and that defense-in-depth hasbeen implemented robustly, so that protective actionguidelines for modified emergency planning require-ments can be met. The approach should be developed incoordination with national regulators and other respon-sible authorities.

Crosscutting SR Performance Phase R&DThere are additional SR technology R&D areas whereadvances have the potential to improve the SR goalperformance of most or all Generation IV facilities.Many of these domains will likely be studied under near-term deployment research for application to near-termsystems. Generation IV facilities should build on suchdevelopments.

Licensing and Regulatory Framework. Many Genera-tion IV systems involve substantial changes in safety-system design and implementation that require licensingimplementation significantly different from currentexperience. Best-estimate and risk-informed bases for

Design Detail for SR Viability Evaluations

For SR viability evaluation, the level of design detail for reactorand fuel cycle facilities should be sufficient to allow:

• Description of the facility design features that implement thefive individual levels of defense as defined by INSAG-10,

• Performance of a simplified PRA to accurately quantify thecontribution to the risk of all the design-basis transients andaccidents resulting from internal and external events, for allfacilities and all operating modes and assess theirapproximate probabilities,

• Identification and ranking of the phenomena that govern thesystem transient response under design basis and designextension conditions,

• Demonstration that separate effects experimental data areavailable, or are planned for, that closely replicates the scaledboundary and initial conditions for the dominant phenomenawith minimal distortion,

• Performance of selected best-estimate design-basis transientand accident analyses demonstrating the quantitativeevaluation of uncertainty, and explicitly identifyingapproximations and assumptions that will be removed bysubsequent performance R&D, and

• Description of the integral test facilities and theirinstrumentation planned to validate system transientresponse models, preferably at prototypical scale.

licensing will play a stronger role, due to the greatersimplicity and improved uncertainty characterization forthe new safety systems. R&D is recommended todevelop more flexible risk-informed regulatory tools forlicensing of these advanced systems, and for increasinginternational consistency in design for licensing.

Radionuclide Transport and Dose Assessment. R&Dis recommended to develop improved phenomenologicaland real-time transport and dose modeling methods tosupport improved real-time emergency response, as wellas optimize emergency planning methods and require-ments.

Human Factors. One of the main objectives of crosscutR&D into human factors should be to identify andcharacterize the plant and systems design features thatinfluence human performance in operation and mainte-nance, and to create quantitative criteria to enableeffective comparison of Generation IV systems andmake design decisions. For example, the decision to

71


maintain humans in an active role in the management offuture plants and decisions that set their actual level ofresponsibility should be based on objective evidence forpositive contributions to plant safety and reliability.R&D is recommended for these objectives.

Additional R&D Areas. Crosscutting R&D during theperformance phase is recommended in the followingareas:

• Operation and maintenance

• Instrumentation, control, and the human machineinterface

• Reactor physics and thermal-hydraulics, includingpossible application of coolants with dispersednano-phase particles for improved performance

• Risk management.

Safety and Reliability Evaluation and PeerReviewDue to the limited information on the detailed design ofGeneration IV systems, reviews in the roadmap havefocused on intrinsic characteristics. These characteris-tics affect the potential performance to the safety andreliability goals, such as the thermal inertia associatedwith reactor cores. Intrinsic characteristics provide astrong foundation but still play only a partial role in thesafety and reliability of nuclear energy systems. Thedetails of the facility designs and the fundamental safetyarchitecture also have a high importance to the evaluation.

Considering the importance of the safety and reliabilityof Generation IV systems, research on systems shouldinclude an effective safety and reliability peer-reviewmechanism. This process should be structured to ensurethat the best design practice is employed in all Genera-tion IV facilities, with a particular focus on the correctimplementation of defense in depth principles.

Crosscutting Risk and Safety R&D Scheduleand CostsA schedule for the crosscutting risk and safety R&D isshown below, along with the R&D costs.

Crosscutting Economics R&DIntroduction and ApproachThis section addresses crosscutting economic researchrelating to Generation IV nuclear energy systems. Asdiscussed in the Observations on Economics sectionearlier in the roadmap, there is a need for crosscuttingR&D to (1) base cost estimates on a robust and compre-hensive methodology addressing uncertainties, and (2)resolve the issue of modular versus monolithic planteconomics. In addition, research is needed into the basisand allocation of costs for nonelectrical products.Researchers and designers will need to continuallyaddress system economics as the R&D proceeds, andtools are needed to guide them. The objective of thesetools is to improve the consistency of economic assess-ments and uncertainties associated with them. With newtools, Generation IV designers can gain a better under-standing of how their designs compare with alternativenuclear systems or other technologies. They can identifyareas deserving specific attention and focus their effortson improving the economic performance.

The innovative nuclear systems within Generation IVwill require unique tools for their economic assessment,because their characteristics are very different fromthose of earlier nuclear power plants. Specifically, thereare five main economic tools that should be refined fromexisting tools or developed as new tools (see figure).

RISK AND SAFETY CROSSCUT (20 M$) Safety assessment methodologySimplified PRA methodologyEmergency planning methodsLicensing and regulatory frameworkRadionuclide transport/dose assessmentHuman factors studiesAdditional R&D areas

2000 2010 2020

Capital and ProductionCost Model

Nuclear FuelCycle Cost Model

Integrated NuclearEnergy Economics Model

Energy Products Model Plant Size Model

72


These consist of four standalone cost models, as well asan overall model that integrates them for the purpose ofexploring uncertainty ranges in the input. These modelsare needed during the viability phase of system R&D togive a preliminary answer to the question of economicviability that is central to all of the Generation IVsystems.

A number of methods and computer models exist thatcan estimate the cost of a reactor under development,i.e., before there is experience constructing and operat-ing it. Most of these models were implemented in theearly stages of nuclear energy deployment (duringGeneration II) and updated on a regular basis during theperiod. However, most nuclear power plants builtrecently are evolutionary, based upon designs andtechnologies that are mature and proven. Therefore,these cost assessment tools have not been updated sincethe early 1990s. Such models can form the basis for twomodels in the figure: the Nuclear Fuel Cycle CostModel (or Fuel Model), and the Capital and ProductionCost Model (or Cost Model).

The fuel and cost models are central to the economicevaluation of nuclear systems. An example of anexisting fuel model is the OECD/NEA model used forpreparing The Economics of the Nuclear Fuel Cycle.dThe fuel model calculates costs associated with both thefront end and back end of the nuclear fuel cycle, andprovides information needed by the cost model. Anexample of an existing cost model is ORNL’s CostEstimate Guidelines for Advanced Nuclear PowerTechnologies.e The cost model inputs the cost of nuclearfuel to a calculation of capital costs, as well as the costsof production. Existing fuel and cost models, however,are not adapted to innovative fuel cycles. For example,minor actinide partitioning and transmutation cannot beanalyzed.

A new model, the Plant Size Model (or Size Model) isneeded to analyze costs and implications of a range ofoptions for innovative systems. The size model needs totreat modular plants and the associated economies ofserial production-construction as well as monolithicplants and the associated economies of scale for largeunits. By itself, the size model may help to determinethe optimal size of the nuclear energy production plantwithin a Generation IV system.

Another new model, the Energy Products Model (orProducts Model) would address the economics ofmultiple energy products. The products model wouldanalyze system tradeoffs between, for example, low costelectricity generation and actinide management and/orhydrogen production.

An Integrated Nuclear Energy Model (or IntegratedModel), combines all of the nuclear-economic modelsdescribed above and provides a robust framework foreconomic optimization. The integrated model would beable to propagate the effects of uncertainties in themodel inputs.

Capital and Production Cost Model (Cost Model)An existing cost model, such as the cited model, shouldbe updated. This, as well as most other production costmodels, uses the lifetime-levelized cost methodology.This methodology calculates costs on the basis of netbus-bar power supplied to the station. Applied togeneration costs, the lifetime-levelized cost methodol-ogy provides costs per unit of electricity generated equalto the ratio of (1) total lifetime expenses and (2) totalexpected generation, both expressed as discountedpresent values. Those costs are equivalent to the averageprice that would have to be paid by consumers to repaythe investor for the capital and the operator for O&Mand fuel expenses, at a discount rate equal to the rate ofreturn. The cost model must include all aspects ofconstruction, including sequencing and duration of plantconstruction or fabrication tasks. Further, capitalexpenditures should include refurbishment (also knownas capital additions) and decommissioning costs. Realescalation rates (nominal escalation rates minus thegeneral level of inflation) for operation and maintenanceand fuel costs are taken into account if applicable.

To assess the economic advantage of nuclear energysystems over alternatives, all costs facing the utility, i.e.,those that would influence its choice of generationoptions, should be taken into account. In particular, thecosts associated with environmental protection measuresand standards, e.g., the cost of safety and radiationprotection measures for nuclear systems, should beincluded in life-cycle costs. On the other hand, externalcosts that are not borne by the utility, such as costsassociated with health and external impacts of residualemissions, are not included. However, if external costs

dOrganisation for Economic Cooperation and Development and the OECD Nuclear Energy Agency, 1994, www.nea.fr/html/ndd/reports/efc,accessed September 2002, This publication is out of print and can be obtained only from this website.eJ. G. Delene and C. R. Hudson, Cost Estimate Guidelines for Advanced Nuclear Power Technologies, ORNL/TM-10071/R3, LockheedMartin Energy Systems, Inc., Oak Ridge National Laboratory, 1993.

73


are borne by the public or the environment, publicagencies should take these costs into account whenchoosing among nuclear technologies. A limitation inthe Lifetime Levelized Cost Methodology is that it isonly relevant for deployment of new nuclear or otherpower plants in traditional cost-of-service regulatedenvironment. The deregulation of electricity markets inmost countries requires traditional cost models to beupdated.

Nuclear Fuel Cycle Cost Model (Fuel Model)Since fuel cycle costs represent about 20% of thelevelized cost of new nuclear electricity generation inmost current nuclear power plants, reducing those costswill help new systems meet the Generation IV econom-ics goals. Further, fuel cycle cost models can play animportant role as a decision tool for optimizing fuelcycle options by taking into account economic tradeoffsbetween design choices in sustainability, safety andreliability, and proliferation resistance and physicalprotection. The model used to prepare the report, TheEconomics of the Nuclear Fuel Cycle (1994), is anexample of existing tools capable of handling the classicfuel cycles.

The assessment of innovative fuel cycle economics isessential and requires updating existing models. Forclassic fuel cycles, the main steps are uranium produc-tion, conversion, enrichment (not needed for naturaluranium fuel cycles), fabrication, and spent fuel disposalfor the once-through option. In the recycle option, theback end of the fuel cycle includes reprocessing,refabrication, and disposal of HLW from reprocessing.Innovative fuel cycles will require the adaptation ofexisting models to include different steps, materials, andservices. For very unique systems, such as the MSR, thedesign and implementation of an entirely new fuelmodel may be required.

An updated fuel model should include recent develop-ments in the understanding of reprocessing and reposi-tory economics. It must provide complete front andback end costs to the cost model.

Energy Products Model (Products Model)The economics of the joint production of electricity andother energy (nonelectrical) products needs to be betterunderstood. For example, the economics of jointelectricity and hydrogen production using nuclearenergy has not yet been fully analyzed, let alone mod-eled. Because most of the Generation IV technologiescan be used to address more than one mission, crosscut-ting economics research must define standards for

accounting for the costs of more than one product.Further, the tradeoff between the use of heat to producehydrogen and residual heat to produce electricity is alsonot well specified. Similarly, the joint production ofelectricity and actinide management services requiresfurther analysis. Standard economic models must bedeveloped to evaluate these tradeoffs under variousregulatory and competitive environments. At the sametime, it is critical to the Generation IV effort to under-stand the supply (industry cost structure) and demand(including alternatives) for hydrogen and actinidemanagement, and how this market will likely evolveduring this century. In particular, using Generation IVtechnologies to manage actinides requires the specifica-tion of the feedback mechanism between the productionof spent nuclear fuel and its life-cycle management.

Plant Size Model (Size Model)An issue that has not yet been resolved in the assessmentof advanced reactor technologies is whether massproduction of small reactors can overtake the costadvantages from scale economies of large units orplants. There are cost factors involved in the construc-tion of a small modular plant that are not encounterednor accounted for in the conventional cost computationof a large monolithic plant. To make a reasoned eco-nomic decision as to which plant to select, it is essentialthat all the cost factors involved are considered. Ingeneral, specific plant capital costs, expressed in cur-rency per installed kWe (e.g., $/kWe) are lower for alarge plant, due to economies of scale. Yet there aresignificant advantages to the early construction comple-tion and start-up of smaller plants (e.g., an early revenuestream) that do not routinely appear in the standard costaccounting system developed for large monolithicplants.

There are several specific cost factors that should beaccounted for when comparing the economic advantagesof large versus small and modular nuclear power plants.Such factors include (1) load management and reliabil-ity, (2) standardization and licensing, and (3) retiringplant replacement possibilities, among others. Eco-nomic models should reflect these factors to ensure afair assessment of the potential economic benefits ofsmall modular systems versus large monolithic systems.More work must be done to properly account for thedifferences between small and large plants. While basicresearch in this area should be inexpensive, developingeconomic-engineering model would require moreresources. For example, research in this area should beextended to developing the conceptual engineeringdesign of fabrication facilities and transportation systems.

74


Integrated Nuclear Energy Model (IntegratedModel)An integrated model, combining all of the modelsidentified above is necessary to compare various Gen-eration IV technologies, as well as to answer optimalconfiguration questions, such as which fuel cycle is mostsuitable for each state of the world and optimal deploy-ment ratios between members of symbiotic set. Thegoal of integrating these models provides incentives tobuild common data interfaces between the models.Also, none of the individual models addresses theproblem of uncertainty, e.g., the uncertainty of cost andparameter estimates. Roadmap evaluations on econom-ics for Generation IV considered ranges, expectedvalues, and probability distributions for constructioncost, construction duration, and production costs. Fromthese, probability distributions for average cost andcapital-at-risk were generated assuming no correlationbetween costs and durations. The integrated modelshould be able to address these type of uncertainties.The integrated model will be able to guide decisionmakers in their assessment of these uncertainties, i.e.,help them to assess the value of reducing uncertaintythrough the allocation or reallocation of research funds.

Model Development StepsThe models should be developed now for use during theviability phase of the Generation IV systems. The figureidentifies the order of these tasks. During the first year:

1. The Cost Model should be created by updating anaccepted model

2. The Fuel Model should be created by updating anaccepted model

3. Reports should define the requirements for the othermodels.

During the next two years, these updated models shouldbe integrated and work should proceed on the creationthe Product and Size Models. During the last two years,all of the models should be integrated with a focus onaddressing uncertainty. Further, the development ofengineering designs of nuclear plant fabrication facilitiesshould begin that would allow further refinement of thesize model. These designs should include expected costsand these costs should be integrated into the integratedmodel. As an integrated set, the models will aid decisionmakers in assessing the viability of Generation IVsystems and technologies.

Economics Evaluation Peer ReviewDue to the limited information on the detailed design ofGeneration IV systems, reviews to this point in theroadmap have primarily considered studies advanced byadvocates. Considering the importance of the econom-ics of Generation IV systems, research on systems mustadopt an effective economics peer-review mechanism.This process should be structured to ensure that thedesigns continually address their progress into competi-tive systems.

Year 1 Years 2–3 Years 4–5

Capital and ProductionCost Model

Nuclear Fuel CycleCost Model

Identify issuesand define requirements

for new models

Integrated Cost Model

Energy Products Model

Plant Size Model

Integrated NuclearEnergy Model

75


Crosscutting Economics R&D Schedule andCostsA schedule for the crosscutting economics R&D isshown below, along with the R&D costs and a decisionpoint.

process is conducted by expert panels using an assess-ment methodology that is established through R&D.This R&D is also presented below.

Overall, the R&D program should be conducted in threeareas. The first area is the safeguards and physicalprotection technology R&D that is carried out in thedevelopment of each Generation IV system. The finaltwo areas are R&D needed for the formulation ofPR&PP criteria and metrics, and their evaluation,respectively.

R&D Supporting the Safeguards and PhysicalProtection Strategy. The following R&D is recom-mended:

1. Determine the type, amount, and location of (1)nuclear materials suitable for weapons use, (2) othernuclear material from which such material could becreated (through enrichment, reprocessing orirradiation followed by reprocessing), and (3)hazardous radioactive material. These should bedefined in the context of each system and theprovisions for its deployment over its entire lifecycle.

2. Identify potential vulnerabilities for all materials inthe fuel cycle for each of the five security threats.For each vulnerability identified, R&D should becarried out to decrease the attractiveness of thematerial for diversion or theft, or to increase thedifficulty of dispersing the material, as appropriate.

3. Determine means to protect key reactor or fuel cyclefacility technology that could be used for prolifera-tion against unintended use, and related systems,equipment, and materials that could be used forproliferation against unauthorized replication.

Crosscutting Proliferation Resistance andPhysical Protection R&DIntroduction and ApproachThe methodology developed during the roadmap pro-vided only a limited evaluation of proliferation resis-tance and physical protection (PR&PP). A substantiallyimproved PR&PP evaluation methodology is needed toprovide a more balanced and complete evaluation. Thissection recommends R&D relevant to this goal area,followed by recommendations on R&D in evaluationmethods.

One of the important endpoints of Generation IV R&Dis a preliminary safeguards and security strategy that isdeveloped during the viability R&D phase. The prelimi-nary strategy will be conceptual and schematic in nature,reflecting the early state of development of the nuclearenergy system. It addresses the vulnerabilities for eachsystem in relation to the following five security threats:

• State-driven diversion or undeclared production offissile materials

• Theft of fissile materials

• Theft of nuclear material for radiation dispersaldevices

• Sabotage of nuclear facilities

• Sabotage of nuclear materials in transport.

During both the viability and performance phases, thestrategy will be reviewed against a set of criteria andmetrics relating to the intrinsic and extrinsic measuresdefined in the strategy to address the five securitythreats. The formulation of the criteria and metricsrequire R&D and are presented below. The evaluation

ECONOMICS CROSSCUT (10 M$) Capital and production cost model cost Nuclear fuel cycle cost model Energy products model Plant size model I Economic viability of modular fabrication and installation technologies (EC 1)

ntegrated nuclear energy model

decision

2000 2010 2020

76


4. For each material of any form in the system, identifyand increase the intrinsic and extrinsic protectionafforded against its diversion, theft, or dispersal.These means may exploit chemical or physicalfeatures, or use radiation barriers to decrease poten-tial vulnerabilities.

5. For solution processing systems involving partialdecontamination, such as the pyroprocess or ad-vanced aqueous process, and for all processesinvolving molten salt fuel, identify potential meansto extract nuclear material suitable for weapons usethrough the misuse of normal plant equipment orthrough the introduction of additional systems thatmight be concealed from discovery by the facilityoperator, the national control authority, or theInternational Atomic Energy Agency (IAEA).

6. Recognizing the importance of an ongoing consulta-tive system, and consistent with the provisions ofapplicable IAEA safeguards agreements, interac-tions with the agency should start during the viabil-ity R&D phase. This effort would identify generalaspects of the safeguards approach, alternativemeasures, and any system specific research anddevelopment needed to facilitate later agreement onthe technical measures to be applied. When suffi-cient information is available about a particularsystem, the interaction with IAEA should lead to acase study by the Safeguards Department of theIAEA. During the performance phase, detailedaspects of the safeguards approach would be speci-fied, developed, and tested. The capabilities of thesafeguards system would be determined and im-provements pursued as needed.

7. Using the simplified PRA for the system, identifythe vulnerability to sabotage that could lead toreleases of radioactive material or theft resultingfrom breaches in containment, and any additionalmeasures appropriate to counter such threats.Specifically, the safety analyses should be reviewedfrom the viewpoint of intentional acts as the initia-tors for the safety sequences identified, taking intoaccount the use of force including armed attack andthe consequent possibilities for the destruction ofcritical safety systems or structures, and the poten-tial acts of knowledgeable insiders to operate thefacility or systems in an intentionally unsafe manner,or to disable or destroy critical safety systems.

8. Determine the potential use of the reactor forclandestine production of plutonium or 233U, theimpact of such use on the safe operation of the

reactor, the detectability of fertile material intro-duced into irradiation positions, and the detectabilityof changes in the neutronic or thermal-hydraulicbehavior of the reactor. For any such potential use,investigate means to minimize the vulnerability.

9. For each step in the fuel cycle, define a concept fordetermining the amounts, locations, and characteris-tics of all material in real time. This would providea foundation for the material protection, control, andaccounting (MPC&A) system, and would providethe basis for the protective system employed by thefacility operator. The foundation should include:

a. Information generated through in-line and off-line monitoring instruments

b. Information from sampling and laboratorymeasurements

c. Development and validation of inventory andflow predictive models for each operation andfacility

d. Information processing algorithms for theestimation of amounts and properties of allmaterials

e. Quality control provisions.

R&D of PR&PP Evaluation Criteria and Metrics.R&D is recommended to produce the set of criteria andmetrics for the evaluation of the intrinsic and extrinsicbarriers that address each of the five security threats. Aswith other criteria and metrics, these are expected to berefined to match the level of detail as the systemsadvance through viability and performance R&D.

R&D of the Assessment Methodology. Deterringproliferation and nuclear terrorism will rely upon thecollective implementation of intrinsic and extrinsicmeasures that are intended to deter such acts. Theselection and implementation of cost-effective combina-tions of such measures is complex, subtle, and involvesmany plausible alternatives. For this reason, efforts toevaluate the risks of proliferation and nuclear terrorismagainst a system of intrinsic and extrinsic barriers havenot yet provided clear and convincing answers. Explicit,comprehensive methods for evaluating the adequacy andrequirements for a safeguards and physical protectionsystem are needed to assess the protection and responsecapabilities it provides.

77


R&D is recommended into the development of practicalassessment methodologies. The research should reflectthe needs of each potential user as a function of time,and the differences in information potentially availableto each. The process of developing this methodology islikely to be iterative in nature, as it strives to encompassthe complexity of the problem.

PROLIFERATION RESISTANCE AND PHYSICAL PROTECTION CROSSCUT (20 M$) Evaluation Criteria and Metrics for PR&PP Assessment Methodology

Development of a PR&PP Strategy for the SCWR Development of a PR&PP Strategy for the VHTR

Development of a PR&PP strategy for the GFR Development of a PR&PP Strategy for the LFR Development of a PR&PP Strategy for the MSR Development of a PR&PP Strategy for the SFR

2000 2010 2020

Crosscutting Proliferation R&D Schedule andCostsA schedule for the crosscutting proliferation resistanceand physical protection R&D is shown below, along withthe R&D cost.

78


79


IntroductionThis section suggests an approach to building a Genera-tion IV program with the necessary and sufficient R&D.Issues and opportunities exist for the program, and theseare explored in a discussion of the path forward.

Overall Advancement of Generation IVProgram Definition and BalanceWith six most promising Generation IV systems and tencountries in the GIF, the approach to building integratedprograms for any of the systems becomes an importantissue. The GIF countries have expressed a stronginterest in collaborative R&D on Generation IV systems.However, it has always been acknowledged that eachcountry will participate only in the systems that theychoose to advance. In light of the considerable resourcesrequired for the development of any Generation IVsystem—roughly 1 billion U.S. dollars each—not all sixsystems are likely to be chosen for collaborations.Those that are will need to assemble the priority R&Dfor the system and the necessary crosscutting R&D, andthen set the desired pace for the program. The technol-ogy roadmap has been structured to allow the indepen-dent assembly of collaborative R&D programs.

With regard to the timing of programs, the figure showsan overall summary of the expected duration of the R&Dactivities for the various systems. It is apparent that thesystems do not complete their viability and performancephases at the same time. As a result, for each of thesystems, the GIF will need to periodically assess itsability to continue. The technology roadmap has takencare to include R&D on evaluation methodology thatwill support the need for these continuing assessments.After the performance phase is complete for eachsystem, at least six years and several US$ billion will berequired for detailed design and construction of ademonstration system.

Cooperation and PartnershipsThe GIF plans to focus their future meetings on thedevelopment of collaborative programs on severalsystems. Of considerable interest is the participation ofindustry in the Generation IV program, and its growth asthe systems advance. While the prospects for demon-stration and entry into commercial markets are a numberof years into the future, the need exists for early involve-ment of industry to provide direction and keep a focuson the requirements for the systems.

R&D Programs for Individual Generation IVSystemsThe technology roadmap has been structured to facilitatethe assembly of larger R&D programs or smallerprojects on which the GIF countries choose to collabo-rate. Programs would consist of all or most of the R&Dneeded to advance a system. Projects would consist ofR&D on specific technologies (either system-specific orcrosscutting) or on subsystems that are needed for aGeneration IV system. In either case, the program orproject should be focused on key technology issues andmilestones. This section highlights the major milestonesand development needs that have been identified in theR&D activities.

INTEGRATION OF R&D PROGRAMS AND PATH FORWARD

System Development Timelines

2000 2010 2020

MSR

LFR

GFR

VHTR

SFR

SCWR

2030

System Development Timelines

2000 2010 2020

ViabilityViability PerformancePerformance

MSR

LFR

GFR

VHTR

SFR

SCWR

2030

DemonstrationDemonstration

80


R&D EndpointsTo better define the viability and performance phaseactivities in the technology roadmap, the tables belowsuggest the objectives and endpoint products of theR&D, or endpoints. The R&D activities in the roadmaphave been defined to support the development to theseendpoints. The specific milestones and technology areasof the R&D are discussed next.

Viability PhaseThe viability phase R&D activities examine the feasibil-ity of key technologies. Examples of these includeadequate corrosion resistance in lead alloys orsupercritical water, fission product retention at hightemperature for particle fuel in the very-high-tempera-ture gas-cooled reactor, and acceptably high recoveryfractions for transuranic actinides for systems employingactinide recycle. The tables on the next page present asummary of the decision milestones and their projecteddates, assuming that the R&D can proceed at a reasonablepace.

Viability Phase Objective:

Basic concepts, technologies and processes are proven out under relevant conditions, with all potential technical show-stoppers identified and resolved.

Viability Phase Endpoints:

1. Preconceptual design of the entire system, with nominal interface requirements between subsystems and established pathways for disposal of all waste streams

2. Basic fuel cycle and energy conversion (if applicable) process flowsheets established through testing at appropriate scale

3. Cost analysis based on preconceptual design

4. Simplified PRA for the system

5. Definition of analytical tools

6. Preconceptual design and analysis of safety features

7. Simplified preliminary environmental impact statement for the system

8. Preliminary safeguards and physical protection strategy

9. Consultation(s) with regulatory agency on safety approach and framework issues

Performance Phase Objective:

Engineering-scale processes, phenomena, and materials capabilities are verified and optimized under prototypical conditions

Performance Phase Endpoints:

1. Conceptual design of the entire system, sufficient for procurement specifications for construction of a prototype or demonstration plant, and with validated acceptability of disposal of all waste streams

2. Processes validated at scale sufficient for demonstration plant

3. Detailed cost evaluation for the system

4. PRA for the system

5. Validation of analytical tools

6. Demonstration of safety features through testing, analysis, or relevant experience

7. Environmental impact statement for the system

8. Safeguards and physical protection strategy for system, including cost estimate for extrinsic features

9. Pre-application meeting(s) with regulatory agency

81


System Viability Phase Decisions DateGFR • Fuel down-selection (GFR 1) 2010

• Core structural materials down-selection (GFR 2) 2010• Safety concept specification (GFR 3) 2010• Fuel recycle viability (GFR 4) 2012• Structural material final selection (GFR 5) 2012

LFR • Structural material selection (550ºC outlet temperature) (LFR 1) 2007• Nitride fuel fabrication method (LFR 2) 2010• Feasibility of transportable reactor/core cartridge (LFR 3) 2010• Feasibility/selection of structural material for 800ºC Pb (LFR 5) 2012• Nitride fuel recycle method (LFR 4) 2014• Adequacy of nitride fuel performance potential (LFR 6) 2014• Ca-Br hydrogen production process (LFR 7) 2014• Supercritical CO2 Brayton cycle (LFR 8) 2014

MSR • Core materials selection (MSR 1) 2006• Fuel salt selection (MSR 2) 2007• Power cycle (with tritium control) (MSR 3) 2009• Fuel treatment (fission product removal) approach (MSR 4) 2012• Noble metal management (MSR 5) 2012• Viability of materials (MSR 6) 2013

SFR • Oxide fuel remote fabrication technology selection (SFR 1) 2006

SCWR • Safety approach specification (SC 1) 2008• Core structural material down-selection (SC 2) 2011• Core structural material final selection (SC 3) 2014• Advanced aqueous process application to recycle (SC 4) 2014• Fuel/cladding system viability (SC 5) 2014

VHTR • High temperature helium turbine (VH 1) 2008• Reactor/hydrogen production process coupling approach (VH 2) 2010• Identification of targeted operating temperature (VH 3) 2010• Fuel coating material and design concept (VH 4) 2010• Adequacy of fuel performance potential (VH 6) 2010• Reactor structural material selection (VH 5) 2010

Crosscut Viability Phase Decisions Date

Fuel Cycle • Adequacy of actinide recovery fraction (advanced aqueous) (FC 1) 2006• Pyroprocess recycle for LWR spent fuel (FC 2) 2006• Adequacy of actinide recovery fraction (pyroprocess) (FC 3) 2006• Recommendation on separate management of Cs, Sr (FC 4) 2007• Integrated management of once-through cycle (FC 5) 2007• Group extraction of actinides in aqueous process (FC 6) 2010

Fuels and Materials • Requirements for irradiation and transient test facilities (FM 1) 2005

Energy Products • Requirements for hydrogen production (EP 1) 2006• Hydrogen thermochemical production demonstration (EP 2) 2011

Economics • Viability of modular fabrication and installation technologies (EC 1) 2008

82


Performance PhaseThe performance phase R&D activities undertakethe development of performance data and optimi-zation of the system. The table to the rightsummarizes the key technology issues for eachsystem. Milestones and dates need to be devel-oped based on the viability phase experience. Asin the viability phase, periodic evaluations of thesystem progress relative to its goals will deter-mine if the system development is to continue.The viability and performance phases will likelyoverlap because some of the performance R&Dactivities may have long lead times that requiretheir initiation as early as possible.

Demonstration PhaseAssuming the successful completion of viabilityand performance R&D, a demonstration phase ofat least six years is anticipated for each system,requiring funding of several billion U.S. dollars.This phase involves the licensing construction andoperation of a prototype or demonstration systemin partnership with industry and perhaps othercountries. The detailed design and licensing ofthe system will be performed during this phase.

Comparison of R&D TimelinesAn R&D timeline has been defined for eachGeneration IV system and crosscutting area. Themore detailed Level 3 timelines are presented inthe recommended R&D section for each of them.A summary of all (less detailed) Level 2 timelines for thesix Generation IV systems is assembled and shown inthe figures to the right for comparison of the overall set.Each timeline identifies the viability and performanceR&D and the cost for each Level 2 task. The timelinefor the crosscutting R&D is shown in the figures onpage 83. The choice of the particular systems and theavailability of resources and partners will affect theactual timeline that is assembled for a Generation IVprogram.

Program ImplementationThe roadmap will be implemented in an internationalframework, with participation by the GIF countries. TheGIF is discussing options on the organization andconduct of its programs. Participation by specialists orfacilities in other countries is desirable.

The GIF expects to implement a set of cooperativeagreements under which multiple countries can partici-pate in research projects. The cooperative agreements

will establish the work scope, obligations, intellectualproperty rights, dispute resolution, amendments, andother necessary items. For each Generation IV systemor crosscut, multiple projects may be defined. Forexample, development of fuel may constitute a singleproject. This structure will allow considerable flexibilityin defining each country’s participation, which is consis-tent with the GIF charter. The GIF has an Experts Groupthat is chartered to oversee and report on programsannually.

Integration Issues and OpportunitiesThe assembly of programs and projects, and the imple-mentation of international collaborations to executethem is the central approach to program integration. Inaddition, there are several important issues that havebeen identified during the roadmap process. Eachpresents an opportunity to more effectively advance aGeneration IV program.

System Prioritized Performance Phase R&D IssuesGFR • Fuel and materials performance

• Safety performance• Recycle performance• Economics performance• Balance-of-plant performance

LFR • Fuel and materials performance• Recycle performance• Economics performance• Balance-of-plant performance• Safety performance• Inspection and maintenance methods

MSR • Fuel treatment performance• Balance-of-plant performance• Safety performance• Materials performance• Reliability performance• Economics performance• Inspection and maintenance methods

SFR • Economics performance• Recycle performance at scale• Passive safety confirmation

SCWR • Fuels and materials performance• Safety performance• Economics performance• Recycle performance

VHTR • Fuels and materials performance• Economics performance• Safety performance• Balance-of-plant performance

83


Viability Performance

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Sodium-Cooled Fast Reactor Systema

$ 100 M

$ 160 M

$ 50 M

$ 140 M

$ 160 M

Total cost = $ 610 M

2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Gas-Cooled Fast Reactor System

$ 220 M

$ 120 M

$ 150 M

$ 50 M

$ 100 M

$ 300 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Molten Salt Reactor System

$ 300 M

$ 100 M

$ 200 M

$ 50 M

$ 150 M

$ 200 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Very-High-Temperature Reactor System

$ 30 M

$ 90 M

$ 80 M

$ 280M

$ 20 M

$ 170 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Safety

Design & Evaluation

Fuel Cycle

Supercritical-Water-Cooled Reactor System

$ 10 M

$ 100 M

$ 220 M

$ 30 M

$ 500 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Lead-Cooled Fast Reactor System

$ 190 M

$ 170 M

$ 150 M

$ 110 M

$ 120 M

$ 250 M


2000 2010 2020

a. Fuel Cycle R&D for the SFR is entirely contained in the Fuel Cycle Crosscut R&D.

Crosscutting R&D Summary2000 2010 2020

Fuel Cycle $ 230 M

Advanced Aqueous $ 70 M

Pyroprocess $ 100 M

Additional R&D $ 60 M

Fuels & Materials $ 220 M


Energy Products $ 190 M

Risk and Safety $ 20 M

Economics $ 10 M

Proliferation Resistance $ 20 M

Balance of Plant $ 10 M

Viability PerformanceViability Performance

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Sodium-Cooled Fast Reactor Systema

$ 100 M

$ 160 M

$ 50 M

$ 140 M

$ 160 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Gas-Cooled Fast Reactor System

$ 220 M

$ 120 M

$ 150 M

$ 50 M

$ 100 M

$ 300 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Molten Salt Reactor System

$ 300 M

$ 100 M

$ 200 M

$ 50 M

$ 150 M

$ 200 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Very-High-Temperature Reactor System

$ 30 M

$ 90 M

$ 80 M

$ 280M

$ 20 M

$ 170 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Safety

Design & Evaluation

Fuel Cycle

Supercritical-Water-Cooled Reactor System

$ 10 M

$ 100 M

$ 220 M

$ 30 M

$ 500 M


2000 2010 2020

Fuels and Materials

Reactor Systems

Balance of Plant

Safety

Design & Evaluation

Fuel Cycle

Lead-Cooled Fast Reactor System

$ 190 M

$ 170 M

$ 150 M

$ 110 M

$ 120 M

$ 250 M


2000 2010 2020

a. Fuel Cycle R&D for the SFR is entirely contained in the Fuel Cycle Crosscut R&D.

Crosscutting R&D Summary2000 2010 2020

Fuel Cycle $ 230 M

Advanced Aqueous $ 70 MAdvanced Aqueous $ 70 M

Pyroprocess $ 100 M

Additional R&D $ 60 MAdditional R&D $ 60 M

Fuels & Materials $ 220 M


Energy Products $ 190 MEnergy Products $ 190 M

Risk and Safety $ 20 M

Economics $ 10 M

Proliferation Resistance $ 20 MProliferation Resistance $ 20 M

Balance of Plant $ 10 M

84


Communications and Stakeholder FeedbackWhile technical advances in Generation IV will contrib-ute to increased public confidence, the degree of open-ness and transparency in program execution may be evenmore important. Accordingly, the findings of thisroadmap and R&D plans based on it will be communi-cated to the public on a continuing basis. Moreover,mechanisms for communicating with interested stake-holder groups should be developed so that their viewsand feedback on the program are considered and, to theextent possible, incorporated into the objectives of theR&D program.

Infrastructure Development and UseGiven the need for substantial R&D on fuel cycles, fuelsand materials, and system conceptual design and analy-sis, it is apparent that existing worldwide infrastructuremay not be sufficient to accomplish the objectives. Anopportunity exists to plan for the shared use of existinginfrastructure, and to undertake the development of newinfrastructure. This is most apparent in the areas of fuelrecycle and refabrication, and fuel and materials irradia-tion and test facilities. Other technology areas maydeserve attention. In addition, the coordinated use ofexisting facilities may offer opportunities, where forexample, irradiation campaigns that support the surveyof candidate fuels and materials may be able to sharefacility space and reduce costs.

Coordinated Licensing ApproachesInteraction with individual regulatory authorities by theR&D programs is essential while the system designsprogress. Such interactions enable the early identifica-tion and resolution of potential licensing issues, becausethey allow the regulator to understand the system designfeatures and technologies and provide feedback. Giventhe emphasis in the Generation IV initiative to enablesystem deployment in larger regions or multiple coun-tries, the opportunity exists for expanding the interac-tions. Beyond this, however, there may be significantopportunity to seek coordinated licensing approachesbetween the authorities. This would be advanced byinteractions of a number of authorities who take up theobjective of exploring a common licensing frameworkfor Generation IV systems.

Institutional Barriers and DevelopmentSome of the Generation IV systems propose deploymentof regional front and back end fuel cycle facilities, andothers propose factory fabrication of modules on a largescale, or connection to future hydrogen supply infra-structures. In the first case, institutional developmentsare needed for regional fuel cycle centers owned by aconsortium of clients and operating under internationalsafeguards oversight. In the other cases, the explorationof barriers and institutional development will presentopportunities for improvement.

Technology Development Interactions withNearer-Term SystemsThe interaction of Generation IV R&D and nearer-termdevelopments such as the U.S. NTD and the INTD willbe beneficial. Near-term development of technologymay offer significant reduction in research needs forGeneration IV systems while expanding the potentialmarket for the developers of a technology. On the otherhand, R&D on Generation IV systems may offer signifi-cant new innovations that could be adopted by nearer-term systems. These benefits point out the opportunityfor the development for collaborations with industry, andfor the coordination of these efforts.

R&D PathwaysMany opportunities are apparent in Generation IV R&Dto sequence the work on its technologies (such as fuelsor fuel cycles) or even entire systems or system options.Some of this has been exploited in the roadmap to thispoint. For example, fuel recycle R&D is on the criticalpath for the SFR and is likely to be first advanced in thatarea. Other systems anticipate this development, andtheir fuel recycle R&D is focused on the specializationof front and back end processes that couple with tech-nology developed for the SFR. Examples are plentiful inthe fuels and materials area, where for example, thedevelopment of nitride fuels by one system will openoptions for several others. With the need to have flex-ibility in program choices and collaborations, however,there has not been a systematic examination of suchpathways. An opportunity exists as the programs andprojects are defined to explore pathways that offerefficiencies and innovation.

85


Generation IVRoadmap NERAC Subcommittee (GRNS)

Salomon Levyg

Levy & AssociatesTed Marston

Electrical Power Research InstituteWilliam Naughton

ExelonNeil Todreasg

Massachusetts Institute of Technology

ROADMAPMEMBERS OF THE GENERATION IV ROADMAP PROJECT

Bobby Abramsf

Duke EngineeringDouglas Chapin

MPR AssociatesB. John Garrick

Independent ConsultantDaniel Kammen

University of California–Berkeley

Roadmap Integration Team (RIT)

Todd AllenArgonne National Laboratory

Ralph Bennettg

Idaho National Engineering and Environmental LaboratoryGian Luigi Fiorini

Commissariat à l’Energie AtomiqueHussein Khalilg

Argonne National Laboratory

John KotekArgonne National Laboratory

John M. RyskampIdaho National Engineering and Environmental Laboratory

Rob VersluisDepartment of Energy – Nuclear Energy

Technical Working Group 1 – Water Cooled

Mario CarelliWestinghouse

John ClevelandInternational Atomic Energy Agency

Michael L. CorradiniUniversity of Wisconsin

Wolfgang DaeuwelEuropean Commission-Framatome

Darío DelmastroComisión Nacional de Energía Atómica

John C. (Jack) Devine, Jr.g

PolestarDavid J. Diamond

Brookhaven National LaboratoryKenneth R. Hedgesg

Atomic Energy of Canada LimitedKazuyoshi Kataoka

ToshibaPhilippe Lauret

Framatome - ANP

Yoon Young LeeDoosan Heavy Industries & Construction Company

Philip E. MacDonaldg

Idaho National Engineering and Environmental LaboratoryYoshiaki Oka

University of TokyoAkira Omoto

Tokyo Electric Power CompanyJong Kyun Park

Korea Atomic Energy Research InstituteNoval A. Smith, Jr.

Dominion Virginia Power CompanyAntonio Teixeira e Silva

IPEN/CNEN/SPAlfredo Vasile

Commissariat à l’Energie AtomiqueGary S. Was

University of MichiganGeorge Yadigaroglu

Swiss Federal Institute of Technology - Zurich

fResigned July 2002gCo-chair or Technical Director

86


Technical Working Group 2 - Gas Cooled

Timothy AbramBritish Nuclear Fuels Limited

Sydney BallOak Ridge National Laboratory

Bernard BallotFramatome - ANP

Arden BementNational Institute of Standards and Technology

Franck Carreg

Commissariat à l’Energie AtomiquePhillip Finck

Argonne National LaboratoryKonstantin Foskolos

Paul Scherrer InstituteKosaku Fukuda

International Atomic Energy AgencyMarco Gasparini

International Atomic Energy AgencyDominique Greneche

COGEMA

Phil Hildebrandtg

Engineering, Management and Technology, Inc.Andrew C. Kadak

Massachusetts Institute of TechnologyShin Whan Kim

Korea Power Engineering CompanyWerner von Lensa

European Commission/FZJ JuelichDavid Moses

Oak Ridge National LaboratoryMasuro Ogawa

Japan Atomic Energy Research InstituteJacques Royen

Organisation for Economic Co-operation andDevelopment-Nuclear Energy Agency

Arkal ShenoyGeneral Atomics

Finis Southworthg

Idaho National Engineering and Environmental Laboratory

Technical Working Group 3 – Metal Cooled

Gerhard BartPaul Scherrer Insitut

Charles BoardmanConsultant

Jean-Louis CarbonnierCommissariat à l’Energie Atomique

Luciano CinottiEuropean Commission-Ansaldo

Jean-Paul GlatzEuropean Commission, Centre Commum de Recherch,Karlsruhe

Orlando J.A. Gonçalves FilhoInstituto de Engenharia Nuclear/CNEN

Masakazu IchimiyaJapan Nuclear Cycle Development Institute

John LeeUniversity of Michigan

Ning LiLos Alamos National Laboratory

Michael Lineberryg


Claes NordborgOrganisation for Economic Co-operation and Development– Nuclear Energy Agency

Ron OmbergPacific Northwest National Laboratory

Stephen Roseng

ConsultantYutaka Sagayamag

Japan Nuclear Cycle Development InstituteAlexander Stanculescu

International Atomic Energy AgencyDo Hee Hahn

Korea Atomic Energy Research InstituteKune Y. Suh

Seoul National UniversityJack Tuohy

Burns & RoeDavid Wade

Argonne National Laborartory

gCo-chair or Technical Director

87


Technical Working Group 4 – Non-classical

Samim Anghaieg

University of FloridaMarc Delpech

Commissariat à l’Energie AtomiqueCharles Forsberg

Oak Ridge National LaboratoryClaude Garzenne

Electricité de FranceJ. Stephen Herring

Idaho National Engineering and Environmental LaboratoryAndrew Klein

Oregon State UniversityTom Lennox

NNC Limited

Maurice LeroyEuropean Commission, Joint Research Center, Karlsruhe

David Lewisg

Argonne National LaboratoryWon Seok Park

Korea Atomic Energy Research InstituteK. Lee Peddicord

Texas A&MPaul Pickard

Sandia National LaboratoryHideki Takano

Japan Atomic Energy Research InstitutePaul Wilson

University of Wisconsin

Economics Crosscut Group (ECG)

Tom LennoxNNC Limited

Keith MillerBritish Nuclear Fuels

Geoffrey Rothwellg

Stanford UniversityJacques Rouault

Commissariat à l’Energie AtomiqueYutaka Sagayama

Japan Nuclear Cycle Development InstituteFinis Southworth

Idaho National Engineering and Environmental LaboratoryJack Tuohy

Burns & Roe

Evelyne Bertelg

Organisation for Economic Co-operation and Development– Nuclear Energy Agency

Chaim BraunAltos Management

Luciano CinottiEC-Ansaldo

Michael L. Corradinig

University of WisconsinKenneth R. Hedges

Atomic Energy of Canada LimitedAndrew C. Kadak

Massachusetts Institute of TechnologyPhilippe Lauret

Framatome –ANP


Energy Products Crosscut Group (EPCG)

Michael GolayMassachusetts Institute of Technology

J. Stephen HerringIdaho National Engineering and Environmental Laboratory

Andrew Kleing

Oregon State UniversityDavid Lewis

Argonne National LaboratoryMichael Lineberry


Masuro Ogawag

Japan Atomic Energy Research InstituteK. Lee Peddicordg

Texas A&MArkal Shenoy

General AtomicsAlfredo Vasile

Commissariat à l’Energie AtomiqueWerner von Lensa

European Commission/FZJ Jülich

88


Evaluation Methodology Group (EMG)


Deborah BennettLos Alamos National Laboratory

Evelyne BertelOrganisation for Economic Co-operation and Development– Nuclear Energy Agency

Dennis BleyButtonwood Consulting

Douglas CrawfordArgonne National Laboratory

Brent DixonIdaho National Engineering and Environmental Laboratory

Michael GolayMassachusetts Institute of Technology

William HalseyLawrence Livermore National Laboratory

Kazuaki MatsuiInstitute of Applied Energy

Keith MillerBritish Nuclear Fuels

Per PetersonUniversity of California-Berkeley

Bill Rasing

ConsultantJordi Roglansg

Argonne National LaboratoryGeoffrey Rothwell

Stanford UniversityThomas Shea

International Atomic Energy AgencyMichel Vidard

Electricité de FranceJean-Claude Yazidjian

Framatome – ANP

Fuel Cycle Crosscut Group (FCCG)

Arden BementNational Institute of Standards and Technology


Bernard BoullisCommissariat à l’Energie Atomique

Rakesh ChawlaPaul Scherrer Institut

Doug CrawfordArgonne National Laboratory

Charles Forsbergg

Oak Ridge National LaboratoryKosaku Fukuda

International Atomic Energy AgencyJean-Paul Glatz

European Commission, Joint Research Center, KarlsruheDominique Greneche

COGEMAWilliam Halsey

Lawrence Livermore National Laboratory

J. Stephen HerringIdaho National Engineering and Environmental Laboratory

Maurice LeroyEuropean Commission, Joint Research Center, Karlsruhe

David LewisArgonne National Laboratory

Hiroshi NodaJapan Nuclear Cycle Development Institute

Per PetersonUniversity of California-Berkeley

Luc van den DurpelOrganisation for Economic Co-operation and Development– Nuclear Energy Agency

David Wadeg

Argonne National LaboratoryMyung Seung Yang

Korea Atomic Energy Research Institute

89


Tim Abramg

British Nuclear FuelsGerhard Bart

Paul Scherrer InstituteDouglas Crawfordg

Argonne National LaboratoryWolfgang Daeuwel

European Commission-FramatomeClaude Garzenne

Electricité de FranceMasakazu Ichimiya

Japan Nuclear Cycle Development InstituteNing Li


Philip E. MacDonaldIdaho National Engineering and Environmental Laboratory

Phillippe MartinCommissariat à l’Energie Atomique

David MosesOak Ridge National Laboratory

Claes NordborgOrganisation for Economic Co-operation and Development– Nuclear Energy Agency

Won Seok ParkKorea Atomic Energy Research Institute

Gary S. WasUniversity of Michigan


Fuels and Materials Crosscut Group (FMCG)

Risk and Safety Crosscut Group (RSCG)

Sydney BallOak Ridge National Laboratory

Bernard BallotFramatome - ANP

Jeff BinderArgonne National Laboratory

Dennis BleyButtonwood


Mario CarelliWestinghouse

Marc Delpechg

Commissariat à l’Energie Atomique

Marco GaspariniInternational Atomic Energy Agency

Kazuyoshi KataokaToshiba

Per Petersong

University of California-BerkeleyPaul Pickardg

Sandia National LaboratoryStephen Rosen

ConsultantJacques Royen

Organisation for Economic Co-operation andDevelopment-Nuclear Energy Agency

Kune Y. SuhSeoul National University

Additional Contributors

Paul BaylessIdaho National Engineering and Environmental Laboratory

Jacopo BuongiornoIdaho National Engineering and Environmental Laboratory

Burrus CarnahanU.S. Department of State

Madeline A. FeltusDepartment of Energy

Robert GottschallDepartment of Energy

Jim KendallInternational Atomic Energy Agency

Hubert LeyArgonne National Laboratory

Luis NunezArgonne National Laboratory

Carter D. SavageDepartment of Energy

Richard SchultzIdaho National Engineering and Environmental Laboratory

Steven SorrellDepartment of Energy

90


ABWR Advanced Boiling Water ReactorABWR II Advanced Boiling Water Reactor IIAIROX Atomics International Reduction

Oxidation ProcessALMR Advanced Liquid Metal ReactorALWR Advanced Light Water ReactorAP1000 Advanced Pressurized Water Reactor

1000AP600 Advanced Pressurized Water Reactor

600APR1400 Advanced Power Reactor 1400APWR+ Advanced Pressurized Water Reactor

PlusARE Aircraft Reactor Experiment (U.S.)AVR Arbeitsgemeinschaft Versuchsreaktor

(Germany)BWR Boiling Water ReactorCANDU Canada Deuterium Uranium, ReactorCAREM Central Argentina de Elementos

ModularesCR control rodCSAU Code Scaling, Applicability, and

Uncertainty MethodDBTT ductile-brittle transition temperatureDF decontamination factorDHR decay heat removalDOE Department of Energy (U.S.)dpa displacements per atomEBR-I Experimental Breeder Reactor I

(U.S.)EBR-II Experimental Breeder Reactor II

(U.S.)EC Economics (Generation IV goal area)EPR European Pressurized Water ReactorESBWR European Simplified Boiling Water

Reactor

ACRONYMS

FCCG Fuel Cycle Crosscut GroupFIMA fissions of initial metal atomsF-M ferritic-martensitic stainless steelsFP fission productGFR Gas-Cooled Fast ReactorGIF Generation IV International ForumGT-MHR Gas Turbine – Modular High-

Temperature ReactorGWD/MTHM gigawatt-days/metric tonne heavy

metalHC-BWR High-Conversion Boiling Water

ReactorHLW high-level wasteHTGR High-Temperature Gas ReactorHTR-10 High-Temperature Reactor 10 (China)HTTR High-Temperature Engineering Test

Reactor (Japan)HX heat exchangerIAEA International Atomic Energy AgencyIHX intermediate heat exchangerIMR International Modular ReactorINTD International Near-Term DeploymentIRIS International Reactor Innovative and

SecureI-S iodine-sulfur processISIR in-service inspection and repairLEU low enriched uraniumLFR Lead-Cooled Fast ReactorLMR Liquid Metal-Cooled ReactorLOCA loss of coolant accidentLWR Light Water ReactorMA minor actinidesMC (U,Pu)C metal carbide fuel formMHTGR Modular High Temperature Gas-

Cooled Reactor

91


MN (U,Pu)N metal nitride fuel formMOX (U,Pu)O2 mixed oxide fuelMPa megapascalsMPC&A material protection, control, and

accountabilityMSR Molten Salt ReactorMSRE Molten Salt Reactor Experiment

(U.S.)MTHM metric tonnes heavy metalMTU metric tonnes uraniumMWe megawatts electricalMWth megawatts thermalNEA Nuclear Energy AgencyNERAC Nuclear Energy Research Advisory

Committee (U.S.)NTD Near-Term DeploymentODS oxide dispersion-strengthened steelsOECD Organisation for Economic

Cooperation and DevelopmentORNL Oak Ridge National LaboratoryPBMR Pebble Bed Modular ReactorPIE postirradiation examinationPR&PP Proliferation Resistance and Physical

Protection (Generation IV goal area);also PR

PRA probabilistic risk assessmentPUREX Plutonium and Uranium Recovery by

ExtractionPWR Pressurized Water Reactorpyro pyroprocessingR&D research and development

RBMK Reactor Bolshoi MoshchnostiKanalnyi

RCS reactor coolant systemREDOX electrochemical reduction- oxidationRIA reactivity-insertion accidentRPV reactor pressure vesselSC supercriticalSCC stress corrosion crackingSCLWR Supercritical Light Water ReactorSCW supercritical waterSCWR Supercritical Water-Cooled ReactorSFR Sodium-Cooled Fast ReactorSG steam generatorSMART System-Integrated Modular Advanced

ReactorSNF spent nuclear fuelS-PRISM Super-Power Reactor Inherently Safe

ModuleSR Safety and Reliability (Generation IV

goal area)SU Sustainability (Generation IV goal

area)SWR-1000 Siedewasser Reactor-1000THTR Thorium High-Temperature Reactor

(Germany)TRU transuranic elementsTWG Technical Working GroupUREX Uranium Recovery by ExtractionVHTR Very-High-Temperature Reactor

Documents

Gen IV Roadmap