An insider's perspective of fast reactor technology

Pergamon Progress in Nuclear Energy, Vol. 34, No. 4. pp. 453-469, 1999

0 1999 Published by Elsevier Science Ltd All rights reserved. Printed in Great Britain

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AN INSIDERS’ PERSPECTIVE OF FAST REACTOR TECHNOLOGY

Harold F. McFarlane and Michael J. Lineberry Argonne National Laboratory

P. 0. Box 2528 Idaho Falls, ID 83403-252

1. INTRODUCTION

We have taken the opportunity of this special edition of Progress in Nuclear Energy honoring Noel Corngold’s 70’ birthday, to offer a personal perspective on the state of development of fast reactor technology. After a quarter- century of effort devoted to its development, our careers and interests have turned to new pursuits, our involvement now relegated only to peripherally related technologies. With this new outlook, but with the echo of hot technical debates still ringing in our ears, what better time to take a step back to assess how far the world has come along this path?

While there are many unique aspects of fast reactor technology, only a limited number today are fundamental to answering the question of how much more remains to be done before a system can be made to be commercially viable. Here we touch on the development status of core design, coolant, plant operability, safety, fuel selection, fuel processing technology, economics, waste management and nonproliferation, as well as some common misconceptions associated with the fast reactor. Almost 10 years ago, Annals published a similar review’ by Charles Till, our mentor in the fast reactor business and the acknowledged leader of the US effort for most of the past two decades.

More in-depth accounts of these topics of course can be found in a number of places. Most recently, at the 1998 American Nuclear Society Winter Meeting in Washington, DC, there was an excellent, but not officially published review2 of these subjects by a panel of experts from France, Japan, Russia and the United States. A comprehensive report of the US development activities prior to their termination was published as a special issue of Progress in Nuclear Energy3 in 1997. Proceedings of the series of International Conference(s) on Future Nuclear

’ C. E. TilI, “Advanced Reactor Development,” Ann. Nucl. Energy, 16-6, p. 301 (1989). ’ “Will Fast Reactors Be Able to Economically Meet Future Energy Needs?” ANS 1998 Winter Meeting, Washington, DC, Nov. 15-19, 1998. 3 W. H. Hannum, Guest Editor, “The Technology of the Integral Fast Reactor and its Associated Fuel Cycle.” Progress in Nuclear Energy, 31-1 (1997).

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454 H. F. McFarlane and M. J. Lineberry

Systems: Global held in Seattle (1993), Versailles (1995) and Yokohama (1997) are excellent references for both the promise and the status of this technology.

Noel Corngold’s Influence: One more strike and I (Harold) was out, or so it seemed. When my second advisor abruptly left the Caltech in the middle of my research, I was handed off to Noel Corngold. It wasn’t an obvious match-the polished Ivy League theorist and the introverted lab rat from west Texas. I had been busy grinding out pulsed neutron experiments in the solitude of the little concrete bunker in the parking lot of Thomas Laboratory. My experience with Noel to that point had been limited to the classroom.

Michael, on the other hand, was under Noel’s guidance from the start of his career at the Institute. Together they were blazing new ground in neutron slowing down theory. Brash and outgoing, Michael was my polar opposite in most respects, so of course we immediately became fast friends. Together we were about as unsophisticated a pair of twenty-somethings as Noel could have ever hoped to mentor. There are stories enough to illustrate the point, but we’ll spare ourselves the embarrassment of repeating them here.

The most disconcerting thing about being a nuclear engineering graduate student at the Institute in the late 1960’s was following in the footsteps of Jim Duderstadt and Bob Conn. But Noel treated us all as individuals and managed to get us through, graduated, and employed with surprising expediency. Along the way this soft-spoken gentleman taught us many things, but nothing more lasting than the importance of relentless inquisitiveness, mainly about scientific pursuits, but really without limit. In Noel, this trait seemed to be a never-ending, almost child-like curiosity that more often than not led to some interesting result. The lesson stuck. Today we continue to do what we can to pass along this simple wisdom to the new generations of scientists and engineers under our charge. In fact, maybe the lesson took too well, because we are both currently enrolled in the University of Chicago’s Executive MBA program, trying to find out everything there is to know about international business.

When Michael left the Institute, he elected to begin his career with Argonne National Laboratory analyzing fast critical experiments in Idaho. One of his first acts was to see about persuading me to come out to Idaho and perform some experiments that would need analyzing. I had spent a year as a junior member of the faculty of New York University, which had coincidentally just decided to pawn off its science and engineering assets to avoid financial free-fall. Trading exile to the backwoods of academia for a job with a future was a tough

An insiders’ perspective 455

sell, but we nevertheless teamed up in 1972 for a couple of years that have stretched into 26.

When we talked Noel into working with us in Idaho for part of one of those first summers, we had little inkling that the domain of our careers would be development of the broad expanse of fast reactor technology, from core physics to the closed fuel cycle. While our participation was centered within the US program, we were also fortunate enough to stay in touch with developments around the globe through participation in joint projects, foreign assignees to our facilities, and of course international conferences and publications. But it was perhaps our forays into the political debates surrounding this controversial technology that sharpened our insights and Noel’s early training that gives us the courage to share our perspectives, and more importantly to face them ourselves.

2. OVERVIEW

We entered the field in 1972, an exciting time in which there was urgency to the task of fast reactor development and deployment. Although several important assumptions that drove the US reactor program didn’t materialize as expected, the fundamental rationale for fast reactor development is as compelling today as it was when first formulated by Enrico Fermi, Walter Zinn and others. The reward for success will be a vast, non-polluting source of electricity capable of serving humankind for several millennia. Without it, nuclear energy is destined to become an historical footnote-a hundred-year blip in energy supply that created a 1 OO,OOO-year radioactive waste legacy.

While the promise is simply stated, its achievement is not. After almost 50 years of development, a commercially viable system has yet to be demonstrated. Such a system would include not only the power plant, but also the closed fuel cycle that the fast reactor requires. Sufficient international progress has been made to conclude that several approaches seem capable of delivering a technical success, the choice of a particular technology being a matter of national preference. Far less consensus can be achieved on the question of when commercialization of such technology might become economically favorable.

Even demonstration of technical and economic viability is not sufficient. Overcoming the political barriers to a plutonium fuel cycle may be the most difficult challenge of all. Created and fissioned within a single cycle of a light water reactor (LWR), plutonium today provides more than five percent of the world’s total electricity supply. Whether this fact is little known or simply ignored, fast reactor development is indefinitely on hold in the US and the UK. The


Japanese program has been set back years by the reaction to public relations snafus following incidents of relatively minor technical importance. Last year’s French elections resulted in replacement of aggressive demonstration projects with a modest R&D program. Only in Russia has enthusiasm for fast reactor development been sustained, but the country’s economic crisis diminishes the significance of its commitment to this capital-intensive technology. The net effect of political setbacks has been to open up the technology to consideration of approaches that have not been in the mainstream of international development. No longer can it be assumed that a commercial fast-reactor system will be based on sodium coolant, mixed-oxide fuel, and Purex reprocessing. We have reached a natural breakpoint in development, and now is perhaps the most opportune time to examine the status of fast reactor technology.

3. TECHNOLOGY EVALUATION

The fast reactor is of course markedly different from today’s established commercial reactors. Its trademark features are a dense, compact core with a high fissile loading and a liquid-metal cooling system. In most respects, the fast reactor has ideal properties for a nuclear power system. Relative to water, sodium has far superior thermal characteristics, some two orders of magnitude higher thermal conductivity and almost an order of magnitude higher boiling point at atmospheric pressure. Using sodium as the coolant allows for a low-pressure, low-activation, non-corrosive cooling system. With its hard spectrum, the fast reactor is capable of producing copious quantities of excess neutrons that can be used beneficially for transmutation of elements. Traditionally fast reactors have been designed to use this feature to breed plutonium within the core and surrounding depleted uranium blankets, hence the often-used name fast breeder reactor (FBR) or liquid-metal fast-breeder reactor (LMFBR). The physics and materials properties of the core also allow the fuel to achieve a factor of four or five higher burnup than LWR oxide fuel. But unlike the LWR, the spent fuel must be reprocessed and recycled to recover its residual high fissile inventory in order for the system to be economic.

Reactor Coolant: For some people, the coolant is the most worrisome aspect of a fast reactor. Water and sodium don’t mix, or more precisely they mix reactively. This gives rise to one of the most persistent misconceptions about the fast reactor, that sodium is somehow a naturally dangerous coolant. To the contrary, sodium is the ideal coolant in most respects. With its very high boiling point (-1 100°C), low pressure, and ability to rapidly remove heat from fuel pins, sodium eliminates the effects of the most troublesome reactor accident initiators. Because it is completely compatible with the metals of construction of the fuel, structures and components, fast reactors don’t have problems with circulating


corrosion products. Impurities are easily removed in cold traps. These features have been demonstrated to result in reduced radiation exposure to operating personnel and an extended lifetime for components.

The assertion about the unsuitability of sodium as a reactor coolant derives from the concern over sodium fires and a misconception about the role of the reactivity coefficient of sodium in reactor safety. The shutdown of the Japanese MONJU plant after some 100 kg of sodium oxidized following a leak fi-om a thermocouple weld highlighted the fire concern. But unlike pressurized water systems, in which high-pressure can quickly eject a large quantity of water and steam, sodium is more likely to drip or weep from a leak site. Such events are readily detected and repaired, as illustrated even in the more significant case of a sodium spill that occurred in the EBR-II plant during the 1960’s. Although comparable in magnitude to the MONJU leak, this event resulted in only a two- month shutdown while lessons learned from the incident were implemented. For some 25 years following, the few minor sodium leaks that did occur were handled as maintenance issues, not emergencies.

Sodium is not the only coolant suitable for use in a fast reactor. Other low-melting-temperature metals or metal alloys have been proposed and tried. In the 1970’s gas coolants were proposed, but offered increased liabilities with few offsetting advantages. Lead and lead-bismuth have received attention more recently because of their compatibility with water and a perceived safety advantage during upset conditions. Lead-bismuth was used successfully in some Russian submarines. This fall Russian Minister of Atomic Energy Yevgeny Adamov announced his nation’s intention to pursue development of a lead-cooled fast reactor concept known as BEST4.

Reactor Physics: The design of fast reactors must be considered well established through the decades of critical experiments, cross section measurement and evaluation, evolution of analysis methods and codes, and of course, computing power. By the mid-1980’s’, most uncertainties in the physics parameters had been reduced to acceptable levels. Today there can be little doubt that the methods and data are adequate to design most core concepts and to specify fuel composition. Originally conceived to produce more plutonium than it consumed while generating power, the fast reactor is still most often thought of as a breeder with a breeding ratio of 1.2 or greater. In fact modern-era design activity has emphasized break-even systems with a breeding ratio near 1.0, or plutonium-consuming

’ “Adamov’s BEST Reactor,” Nuclear News, p. 42, American Nuclear Society, November 1998. 5 L. G. LeSage, M. J. Linebeny and H. F. McFarlane, “Current Status of Reactivity Coefficients,” Progress in Nuclear Energy, 16-3. p. 23 1 (1985).

458 H. F. McFarlane and M. .I. Lineberry

designs with conversion ratios on the order of 0.65. The latter designs have been prompted by the international initiative to reduce the stockpiles of excess weapons plutonium. Such design adaptability is but one example of the versatility of fast- spectrum reactors.

Safety: Perhaps nowhere is the gap between conventional wisdom and the actual status of fast reactor technology greater than in the area of safety. In reality, some of the more impressive achievements have been made in demonstrating the exceptional safety characteristics of properly designed fast reactors. A spectacular test was done at EBR-II in 19866, ironically just three weeks before Chernobyl. The reactivity feedback inherent in EBR-II was sufficient to shut the reactor down from full power without damage in two different “anticipated-transient-without-scram” tests. The combination of the metal fuel and large sodium inventory of the pool arrangement rendered EBR-II immune to such terrors as ATWS events. Analysis has shown that similar results would obtain for larger reactors with the same features. Further, even for hypothetical triple-fault (accident event, failure to scram, and failure of natural reactivity feedback) accident sequences, analysis shows that the damage would be contained within the reactor vessel for a well designed system7. Such events are so unlikely that they are not even considered in the licensing of current generation power plants.

Fuel Performance: In no aspect of the technology is the versatility of the fast reactor better demonstrated than in the success of its fuel development. Measured by burnup capability, several types of fast reactor fuel technologies have already proven to be superior to LWR fuels. The reference fuel type, sintered oxide pellets in stainless steel cladding, has closely followed the development of MOX recycle fuel for LWRs. Satisfactory performance in virtually every participating national program has been achieved using this approach, with 20% (200,000 MW- days/ton) heavy metal burnup demonstrated for test assemblies. Equivalent burnup has been achieved in the US with metal alloy fuel’ and in Russia with fuel made from vibropac oxide granules’. The Russians have included metallic uranium mixed in small percentage with the oxide granules to achieve burnup exceeding 40% in a few test elements. Carbide and nitride fuels have a less-complete database, but they too appear capable of high burnup. Though a fully optimized

’ J. I. Sackett, “Operating and Test Experience with EBR-II, The IFR Prototype,” Progress in Nuclear Energy, 31-l. p. lll(1997).

D. C. Wade, R. A. Wigeland and D. J. Hill, “The Safety of the IFR,” Progress in Nuclear Energy, 31-1. p. 63 (1997). * G. L. Hoffman, L. C. Walters and T. H. Bauer, “Metallic Fast Reactor Fuels.” Progress in Nuclear Energy. 31-I. p. 83 (1997). 9 A. B. Bychkov et al., “Pyroelectrochemical Reprocessing of Irmdiated FBR MOX Fuel III: Experiment on High Bumup Fuel of the BOR-60 Reactor, Proc. Intl. ConjI Future Nuclear Sys., Global ‘97,2, p. 655 (1998).


process has yet to be developed, the certainty of producing the high-density, high- burnup fuels required by fast reactors has been established by unqualified successes in several countries.

Operability: The operation of the prototype and demonstration fast reactors around the world presents a rather uneven picture. We discount the experiences with Fermi I and EBR-I, and other fast reactors of that age as not terribly relevant to a judgment of modern fast power reactor operability. EBR-II was a superb success, operating for thirty years with capacity factor often around 80% in spite of the demands of its experimental workload. Other comparable prototypes-Rapsodie in France, DFR in the UK, KNK-II in Germany, BOR-60 in Russia, and JOY0 in Japan-also met or exceeded expectations. The next (and final), US entry, the Fast Flux Test Facility, also had an excellent record of operation.

Among the demonstration-size fast reactors, the record has varied. The first, the French Phenix plant, has performed well, as have BN-350 in Kazakhstan and BN-600 in Russia. The latter has demonstrated a 73% availability factor. In Scotland PFR performance was not to the same level primarily because of a balky steam generator. SNR-300 in Germany was canceled for political reasons just before startup. The troubles with startup of MONJU in Japan have had more to do with the loss of public trust in the responsible institutions than with any technology failure . The huge Super Phenix plant experienced a variety of operational problems, though now that it is being permanently shut down, the French consider it a successful experiment based on the valuable test data and experience obtained. Finally, the Russians apparently are satisfied with the success of their submarine fast power reactors, about which still little is known in the West.

This hit-or-miss record of power plant operation leaves several open questions on the reliability of operation of a fast-reactor-based generating station. That some plant arrangements have operated with utility-level availability gives some confidence that the technology can be successfully extrapolated to commercial scale. Nevertheless, the top priority of the redirected French R&D program will be in-service inspection and repair (ISIR), a sure sign that more experience with operation of these plants is required to reduce the financial risk of commercialization.

Fuel Cycle: The crux of the arguments against the introduction of fast reactors has focused on the fuel cycle. The thesis of one argument is that it will be decades before uranium prices rise sufficiently for a closed fuel cycle to be economically competitive with the once-through throwaway cycle. The other


argument is that a breeder cycle would be unacceptable from a nonproliferation viewpoint, because of its potential for increasing the world’s supplies of separated plutonium. Because these arguments apply to the less developed elements of fast reactor technology, we have chosen to include more detail in their description. Much of what we have to say was drawn from C. E. Till’s summary10 of the subject at the recent American Nuclear Society meeting in Washington, DC.

The reference fast reactor fuel cycle is based on adaptation of Purex reprocessing and fabrication of pelletized MOX fuel for thermal reactors. With substantial demonstration experience and a synergistic relationship to current commercial MOX recycling, the advantages, limitations and economics of this approach are reasonably defined. Developed initially for LWR fuel, the reference path relies on complete separation of high-purity plutonium and uranium for use in glovebox fabrication of the fuel elements. Alternative approaches are based on taking advantage of the unique features of the fast reactor to reduce both the capital and variable costs of the fuel cycle, while at the same time establishing an improved nonproliferation regime. This direction has been actively pursued in two R&D programs in particular for more than a decade.

Leading the reference development path, France and the UK had mixed oxide FBR demonstration plants by the early 1970’s. FBR fuel reprocessing and fabrication activities carried on simultaneously with reactor commissioning, had the explicit purpose of demonstrating the FBR closed fuel cycle. Both programs had MOX fuel fabrication capability to supply the needs of the reactors. Reprocessing facilities with capacities of 5 tons per year, more than enough to handle the reactors’ spent fuel, were operated during the demonstration period. In the absence of the MOX-fueled demonstration FBRs, the activities in other nations with FBR programs, principally the USSR, Japan, Germany and the US, lacked the impetus to demonstrate the closed fuel cycle.

FBR fuel reprocessing experience in France began in conjunction with Rapsodie, at the small AT1 reprocessing facility at LaHague’ ‘. AT1 operated from 1969 to 1979, with a capacity of 1 kg per day. For Phenix, the Marcoule Pilot Facility (APM) was adapted to FBR oxide fuel in 1973. Modifications (TOR) brought its capacity to 5 tons per year in 1988, a little more than double the actual annual capacity needed. Larger facilities (PURR and MAR 600) were

lo C. E. Till and M. J. Lineberry, “A Review of the Fast Reactor Fuel Cycle,” distributed in the panel session, I* Will Fast Reactors Be Able to Economically Meet Future Energy Needs,” ANS 1998 Winter Meeting, Washington, DC, Nov. 15-19, 1998. ” M. Viala et al., “French Development Program on Fuel Cycle,” Proc. Inntl. Conf: Fast Reactors andRefatedFue1 Qcles, I, p. 1 (1991).


contemplated for Super Phenix and successor reactors before the FBR demonstration program was terminated.

A similar path was followed in the UK12, where reprocessing activities started with facilities constructed at Dounreay for the metal-fueled DFR in the late 1950’s. When the oxide-fkeled 250 MWe PFR was brought into operation in 1974, a major rebuild of the Dounreay reprocessing facility was initiated. The modified facility processed the PFR oxide fuel at a capacity of approximately 5 ton/year. Construction of a 50 ton/year FBR reprocessing plant was completed by 1983. An

intergovernmental agreement was signed in 1984 to bring the UK into the European cooperative program for FBR development. Thereafter, the UK programs began to concentrate on the reprocessing and waste management parts of the fuel cycle. By 1986, design of a 70-ton/year plant had been completed. This was to be the basis for a UK bid on a plant capable of handling the output of the four Super Phenix-size demonstration FBRs planned under the umbrella of the European collaboration agreements. In 1988, however, decisions to slow FBR introduction caused sharp reductions in development activities in the UK, followed retrenchment to generic R&D.

The principal conclusion of the French and UK programs was that the technology was proven sufficiently to move ahead with design of the larger plants that would be necessary for adequate fuel cycle economics. The decontamination factors were satisfactory and the recovery of plutonium was better than 99%. Early issues in the French program with dissolution effkiency for higher Pu content and higher fuel burnup were apparently resolved satisfactorily.

The large LWR fuel reprocessing plants in the UK and France are ongoing commercial concerns today13. However, opponents of plutonium recycle in LWRs have argued that there is no need for reprocessing until uranium prices increase substantially*4, i.e., by an amount that would bring about parity between MOX fuel and the once-through cycle. Using the same model, with slightly different, but reasonable input assumptions, industry and opposition15 analysts have calculated a slight advantage for the once-through cycle16. However, when

I2 H. B. Hickey et al., “Current Status of the UK Fast Reactor Fuel Cycle R&D,” Proc. Ml. Conf Fast Reactors and Related Fuel Cycles, II, p. 15 (1991). ” “EDF Makes Case for Economic Advantage of Reprocessing Over Interim Storage,” Nuclear Fuel, p. 3, Oct. 7. 1996 I4 F. von Hippel, “An Evolutionary Approach to Fission Power,” Proceedings Ml. IConf Future Nuclear Systems, Global ‘95, Versailles, France, 11-14 September 1995, p. 380 (1995). ” B. G. Chow, “Plutonium Economics and the Civilian Nuclear Future,” Proceedings Zntf. Co@ Future Nuclear Systems, Global ‘95, Versailles, France, 11-14 September 1995, p. 400 (1995). ” The Economics of the Nuclear Fuel Cycle, Nuclear Energy Agency, Organization for Economic Cooperation and Development, p. 88, OECD (1994).


the uncertainties are included, the results for the two cases overlap. This narrow economic argument against reprocessing is important to understand because it has implications for the introduction of the breeder, which is even more vigorously opposed by the same people. The additional capital cost of an FBR, generally accepted to be about 20% above a comparable LWR plant, puts even more pressure on the FBR fuel cycle to reduce cost.

For some time it has been recognized that the extensive physical differences between fast and thermal reactors would allow for a completely different approach to reprocessing. FBR spent fuel has at least by a factor of five less bulk than thermal reactor spent fuel for comparable energy production. The residual fissile content is also much higher, by a factor of five to ten. Processes can be more compact, but only if there is little or no moderating material involved, i.e., the advantage is lost with the aqueous Purex process because of criticality safety concerns. The high burnup fuel is proportionately more radioactive, and therefore more of a problem for the organic solvents used in conventional reprocessing schemes. Clean separations are not needed because low concentrations of fission product contaminants do not significantly affect fuel performance in the fast neutron spectrum. Transuranium actinides carried over in the process simply add to the quantity of fissionable material in the fuel matrix. But this process simplicity comes at a premium with regard to fuel fabrication, which must be carried out remotely in heavily shielded hot cells.

Laboratories in the US and Russia independently led the development of alternative FBR fuel cycle technologies that differ radically from the established processes. Both programs relied on electrochemical principles to effect the necessary separations, as opposed to solvent extraction. The Russian process produced superior oxide granules for use in vibratory compaction, while the US product was metal for injection casting. The US development was part of the Integral Fast Reactor (IFR) program that was aimed at developing an advanced FBR with improved characteristics across the board2’.

2o C. E. Till, Y. I. Chaog and W. H. Hanmm, “The Integral Fast Reactor-an Overview,” Progress in Nuclear Energy, 31-1, p. 3 (1997).


In Russia electrochemical process development began to provide enriched uranium-oxide granules for vibropac fuel pins21 in 1977. From 1981 to 1986 MOX granulation experiments were carried out as well. The goal of finding a simpler, more easily automated fabrication technique for high-activity FBR fuel than the reference powder and pellet route was achieved, with eventual production of a dense fuel product with a smear density of about 9 g/cm’. A total of 700 assemblies, two-thirds of which were mixed oxide, were fabricated in this way. The irradiation experience in BOR-60 was reported to be extremely favorable, so much so that substantial numbers of the vibropac assemblies began to be used in BN-600. It is stated that the volume of confirming evidence supporting industrial introduction of the vibropac process is similar to that prevailing at the time of introduction of other accepted technologies such as pelletized processes22.

Pyrochemical processing, which caused some higher actinides to be present in the granulated product, also raised the possibility of their recycle. By the early 1990’s, the Russian program had evolved to planning along these lines23, paralleling then-ongoing development in the US. The fist experiment using irradiated MOX fuel (4.7% burnup) took place in BN-350 during 1991-92. In 1993, BOR-60 began testing fuel at high burnup.

The BOR-60 fuel cycle declads the vibropac column, which has only weak adhesion to the clad, dissolves it by chlorination in molten salt, electrodeposits about a quarter of the UOZ, precipitates the PuO2, electrodeposits the remaining UO 2, and finally removes whatever else is left in the salt by precipitation as phosphates. The PUOZ alone would be used with feed U02 in a closed BOR-60 fuel cycle. Plutonium extraction was almost complete, with 96% in the product, and another percent or so in UO2. More than 3 kg of BOR-60 MOX fuel, some with over 20% burnup, was processed in the experiments. Neptunium mainly went with the uranium. About three-quarters of the americium and most of the curium was in the final waste precipitate. Work on ancillary processes to shift the Am and Cm from the waste to the product stream is planned. As of 1997, larger-scale processing (15kg batch) of BOR-60 spent fuel was also in the works”.

” 0. V. Skiba et al., “Development and Operation Experience of the Pilot Plant for Fuel Pin and Assembly Production Based on Vibropac Uranium-Plutonium Oxide, ” Proc. htl. Conf Fast Reactors and Related Fuel (:&es, Il. p. 15 (1991). ” V. B. Ivanov et al.. “Experimental and Ecological Substantiation of Fuel Cycle Based on Pyroelectrochemical Reprocessing and Vibropac Technology,” Proc. Intl. Conf Future Nuclear Systems, Global ‘97, 2. p. 9 12 (1997). 23 0. V. Skiba et al.. “Nuclear Fuel Cycle Based on Dry Methods for Fuel Reprocessing and Fuel Element Mzmufacture Automated Processes,” Proc. Intl. CTonf: Future Nuclear $vsterns, Global ‘93, 2. p. 934 (1993).

464 H. I?. M cFarlane and M. J. Lineberry

The parallel activities in the US24,2? were part of the Integral Fast Reactor development program that began in 1984. Work continued across a broad front until 1994 when IFR development was terminated following three years of sustained political opposition. At the time of termination, uranium metal processing had been developed at the lo-kg batch scale and plutonium-processing experiments had been conducted at lab scale. In FCF, the irradiated fuel processing facility located next to EBR-II, upgrades to allow plutonium operations under modern standards were almost complete. The intent was to recycle both the uranium and plutonium-bearing spent fuel from EBR-II back into the reactor and demonstrate the closed fuel cycle. At this point in our careers, we had been in charge of modifying the facility and planning experiments for the demonstration, so it was a very exciting time. The history of that work has been exhaustively documented in the open literature26, so there is little need for us to repeat much of it here.

Briefly, the chopped spent fuel was to be submerged in an electrorefiner, where the uranium would be collected by electrotransport and electrodeposition. The empty cladding hulls would be removed, along with some accompanying metallic fission products and a small fraction of residual fuel. After several such operations, plutonium was to be collected by deposition into a liquid cadmium cathode, where it would always be accompanied by at least 30 weight percent of uranium and transuranium actinides, as well as some fraction of the fission products. Following many such cycles and removal of most of the actinide elements, the electrorefiner’s LiCl-KC1 electrolyte would sorbed in zeolite, blended with glass powder, and hot pressed into a ceramic waste form for disposal of the active fission products. The cladding hulls would be blended with zirconium and cast into a metal-alloy waste ingot. After separation of adhering electrolyte by vacuum distillation, the product streams would be blended in the right proportion and cast into fresh fuel pins.

Although the IFR termination decision changed all this, some of the technology was developed for the purpose of treating the EBR-II sodium-bearing metal fuel for safe final disposal. Uranium electrorefining and development of the waste forms continued with little disruption. Research into separating the plutonium-bearing product of course was terminated, as were all fuel fabrication

24 J. J. Laidler et al., Development of Pyroprocessing Technology,” Progress in Nuclear Energy, 31-1, p. 131 (1997). *’ J. P. Ackerman et al., “Treatment of Wastes in the lFR Fuel Cycle,” Progress in Nuclear Energy, 31-1, p. 141 (1997). 26 H. F. McFarlane and M. J. Lineheny, “The IFR Fuel Cycle Demonstration,” Progress in Nuclear Energy, 31-1, p. 155 (1997).


activities for EBR-II. Rather that being recovered, the plutonium and other higher actinides will be immobilized in the ceramic waste form.

Operations with irradiated fuel began in FCF during June 1996. Over a three-year demonstration project, Argonne is processing 100 EBR-II core assemblies, and 25 blanket assemblies2 , At of the end of October 1998, 72 core assemblies had been processed and the first blanket assembly had been introduced for treatment. Two metal waste ingots have been cast, and the equipment is being readied for production of the ceramic waste samples.

The work has demonstrated the viability of the processes at test reactor scale for high-burnup uranium metal fuel. Further, preliminary testing of the two waste forms indicates that their performance compares favorably with the best waste glasses yet produced. However, a number of issues remain in demonstrating the capability of this electrometallurgical technology in a closed fast reactor fuel cycle, as opposed to its use as a nuclear waste processing technology. The process throughput is still low; only 1500 kg of heavy metal will be processed in the three- year demonstration. No practical experience has been gained on the recovery of plutonium from irradiated fuel, or the performance of the recycled fuel during irradiation. Uranium losses due to incomplete dissolution and reactions with crucible coatings have been several percent, too high for a mature recycle technology. An active search for more compatible crucible materials is underway.

Economics: Consideration of the Russian and US experiences with tailored fuel cycles is important to the argument in favor of fast reactor economics. Introduction of fast reactor power plants in the West may require that even the first one be economically competitive with established technology. In an era of cheap uranium, fast reactor fuel cycle costs must be reduced to the extent possible. This could mean that the fuel cycle would have to contribute no more than 10% of the total power generation cost, whereas the best estimates28 today are that the fast reactor fuel cycle costs will be from 10% to 25% of the total. At the lower end of that range, the consumer’s price for electricity would not be very sensitive to marginal variations in fuel cycle costs, so that economic arguments against reprocessing should have little weight.

One way of reducing fuel cycle cost would be to go to very high burnup fuels, which appears to be quite feasible. Another step would be to adapt a

” R. W. Benedict and H. F. McFarlane, “EBR-II Spent Fuel Treatment Demonstration Project Status,” Radwoste, 5. p. 23 (1998). ** B. Barre, “A European View (Em-Based Design and Cost Estimate),” Trans. Am. Nucl. Sot., 79. ANS 1998 Winter Meeting, Washington, DC, Nov. 15-19, 1998.


compact, tailored fuel cycle that did not require the large scale (servicing 20-30 reactors) that seems likely to be necessary for the reference process. Although not specifically for fast reactors, Purex chemists are analyzing single-cycle flowsheets based on centrifugal contactors in an effort to reduce size and cost29. Such innovations could result in a more-compact plant adapted from the reference fuel cycle. The Russians believe that they have sufficient data to estimate their dry processing fuel cost to be no more than 10% of the total power cost. There are too many gaps in the US experience to make a reliable cost estimate, but by the same token, the existing data show promise of a robust, compact fuel cycle.

A number of studies have been performed with the aim of considering ways in which the capital costs of fast reactors could be reduced to, or near the level of commercial LWRs. In general, such studies take advantage of the fast reactor’s superior safety characteristics to justify a reduction in expenditures for components external to the primary system. Given the technology’s level of development, such studies must be considered speculative at this juncture. So long as the reactor must have an intermediate heat transport loop to isolate the primary reactor coolant from the steam generator, it will be difficult to design a robust system with lower capital cost than an LWR. A Japanese scheme to eliminate the intermediate loop is probably well ahead of its time.

Waste Management: The option of significant improvements in radioactive waste management has been a popular theme in the “Global” series of international conferences. Because of its superior ability to transmute elements, the fast reactor has the potential for reducing the performance requirements of geologic repositories. In the US at least, repository performance analysis has become concerned with the very long times, from 10,000 to l,OOO,OOO years. Near the lower end of that time range, technetium and iodine would contribute to most of the dose, while at a million years, neptunium would contribute more than 90% of a peak dose exceeding 0.1 Sv/year3’.

Identification of the isotopes of concern has depended in part on their solubility in water, but if other transport mechanisms such as colloid formation become significant, plutonium will become a major concern. The IFR program planned to keep all of the actinides (uranium, neptunium, plutonium, americium and curium) out of the repository and in the fuel where they would be destroyed by fission. In particular, a fraction of the neptunium would deposit with the uranium,

29 R. J. Taylor et al., “The Development of Chemical Separation Technology for an Advanced Purex Process,” Proc. 5’h Intl. Nucl. ConJ on Recycling, Conditioning and Disposal, RECOD 98, I, p. 417 (1998). ” M. L. Wilson et al., Total-System Performance Assessment for Yucca Mountain - SNL Second Iteration (TSPA- 1993), Volume 2, Samba National Laboratories, Albuquerque, NM. (1994).


while the remainder would carry over with the plutonium. Technetium would be stabilized in the metallic state, in which it is orders of magnitude less soluble than in oxide form. Ideally, only a small volume of high-level waste containing relatively short-lived fission products would be buried.

In reality, some small percentage of the troublesome radionuclides would end up in the repository, but the volume would be small and the waste forms could be tailored to improve stabilization of specific elements. While the IFR program selected this approach as a system design goal Corn the outset, there is some natural separation in the Russian process that could be optimized for waste management. The flowsheet for the reference Purex approach could also be modified to utilize the properties of fast-spectrum reactors for improved high-level waste management. Even for current commercial reprocessing, the volume of high- level waste produced is less than the volume required to directly dispose of the spent fuel. While specific claims have been disputed, no credible challenge has been mounted to the improved waste management regime that could result from the introduction of a fast reactor fuel cycle.

Nonproliferation: Now we turn to the murky world of nonproliferation. It is often hard even to get a dialogue started about whether the introduction of fast reactors would add to the risk of international nuclear proliferation. With its large fissile inventory and ability to double that inventory every 10 to 30 years, the breeder can conjure up such images of uncontrolled plutonium commerce that the answer would seem so obviously to be an unqualified yes. The facts tell quite a different story.

With the introduction of fast reactors, the rate of buildup of plutonium in spent fuel in 31 countries around the world could be reduced, eventually to zero. The inventory of separated plutonium would be drastically reduced, as the large fissile demands of the system would tie up all the material in the reactor core and the fuel cycle. Under the reference fuel cycle, reprocessing and fabrication of new fuel would take place in large central facilities under the auspices of international safeguards. Under the alternative fuel cycles tailored to the fast reactor, plutonium would never be separated into a weapons usable material, always remaining in form protected by high levels of radiation. Although the concerns of proliferation can never be totally eliminated, the versatility of fast reactors is such that safeguards features can be emphasized in their design if that is still a major consideration at the time of their introduction. They need not add to international proliferation concerns, and could reduce such tensions if tailored for an optimized safeguards regime aimed at fast eliminating civilian separated plutonium and then


only adjusting the total inventory of plutonium in fuel as needed to meet power demands.

4. LOOKING AHEAD

After almost 50 years of worldwide research and development, fast reactor technology has reached a state of marginal viability. Using a conservative design, copying the best features of prototypes with a reliable operating history, a power plant could be built. It would cost more to build than a conventional nuclear power station, and initially the operational problems encountered would be anything but routine. Therefore its availability and load factor would probably suffer, and its problems would have to be sorted out in the full glare of international scrutiny. But it would also lack established back-end support for its fuel cycle, because today a new MOX recycle facility capable of supporting a commercial fast reactor would require a commitment to build 20 or more such power stations. At somewhat greater technical risk, but at substantially less expense, a single-unit dry processing facility or tailored Purex facility could be built to service the plant.

While technically feasible, proceeding with a commercial demonstration would not seem to be a rational path today. Of the countries that have contributed to fast reactor development, only Russia and Japan have the national resolve even to consider such a plan. But Japan will move slowly as it builds consensus on every aspect of the decision and which of the various technology options it will actually adopt. Russia perhaps by record of achievement deserves to lead the world in fast reactor implementation, but the country is scarcely in a position to do so while its people are undergoing such terrible economic hardships. The US is only slowly awakening to the reality of a role for nuclear power in cutting carbon emissions, and cannot construct commercial recycle systems under current policies. The UK is focused on nearer-term goals under its recently privatized nuclear facilities. France will focus its R&D mission on the resolution of operational issues identified in its past demonstration programs while it awaits return of a national climate that would support such a bold venture. But perhaps most importantly, few would question that uranium reserves are plentiful for at least the next 20 years, undercutting any urgency for launching into breeder construction.

There is therefore, time enough to complete the R&D needed to develop a fully integrated fast reactor system, rather than one with bits and pieces borrowed from its light water cousin. Primarily these developments should be in the fuel cycle area, learning how to minimize processing and therefore cost, taking


advantage of the fast spectrum rather than paying the economic penalty of adapting thermal reactor fuel processing technology. Waste forms should be developed to meet the needs of the geologic repositories and to the extent practicable troublesome isotopes should be recycled in the fuel. Waste side streams should be reduced to a trickle through aggressive research in relevant areas such as materials compatibility. Separated plutonium should be eliminated from consideration save for using up the existing stockpiles. Safeguards features should be incorporated from the process flowsheets through the plant security to international inspections. New fuels manufactured from the products of the advanced reprocessing schemes should be tested to burnups of 20% and more through international projects conducted in existing reactors such as Phenix, BN-600 and JOYO. The advanced fuels should be subjected to rapid transients to see just how well they stand up to the most unlikely accident conditions. For such a program of R&D to work effectively without excessive expense, international cooperation would be required at a level previously seen in bilateral agreements, but that now would be needed at a multinational level across most of large national nuclear programs.

It is hard for us to imagine a world in which a technology with the unique energy potential of the breeder will never be needed. With its promise of improved safety and high-level waste management, there is justification enough for its introduction long before there is a uranium supply crisis, if only its cost can be reduced sufficiently. A relatively modest level of research, carried out in half a dozen countries with fast reactor technology capabilities, could develop everything needed within a period of 10 years or less.

ACKNOWLEDGMENT

This work was sponsored by the U. S. Department of Energy, Office of Nuclear Energy, Science and Technology, under Contract No. W-3 l-l 09-ENG-38.

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An insider's perspective of fast reactor technology