Chapter 13. Breeder Reactors

The most important feature of the fission process is, of course, the enormous energy release from each reaction. Another significant fact, however, is that for each neutron absorbed in a fuel such as U-235, more than two neutrons are released. To maintain the chain reaction, only one is needed. Any extra neutrons available can thus be used to produce other fissile materials such as Pu-239 and U-233 from the “fertile” materials, U-238 and Th-232, respectively. The nuclear reactions yielding the new isotopes were described in Section 6.3. If losses of neutrons can be reduced enough, the possibility exists for new fuel to be generated in quantities as large, or even larger than the amount consumed, a condition called “breeding.” Several fuel cycles exist, which are distinguished by the amount of recycling. In the once-through cycle, all spent fuel is discarded as waste. Partial recycling makes use of separated plutonium, which can be combined with low-enrichment uranium to form mixed oxide. In the ultimate and ideal breeder cycle, all materials are recycled. As discussed later, there is a revival of interest in some level of recycling to help reduce radioactive waste and to use all fuel energy values.

In this chapter we shall (a) examine the relationship between the reproduction factor and breeding, (b) describe the physical features of the LMFBR, and (c) look into the compatibility of uranium fuel resources and requirements.

13.1. The Concept of Breeding

The ability to convert significant quantities of fertile materials into useful fissile materials depends crucially on the magnitude of the reproduction factor, η, which is the number of neutrons produced per neutron absorbed in fuel. If ν neutrons are produced per fission, and the ratio of fission to absorption in fuel is σ^f/σ^a, then the number of neutrons per absorption is

The greater its excess above 2, the more likely is breeding. It is found that both ν and the ratio σ^f/s^a increase with neutron energy and thus η is larger for fast reactors than for thermal reactors. Table 13.1 compares values of η for the main fissile isotopes in the two widely differing neutron energy ranges designated as thermal and fast. Inspection of the table shows that it is more difficult to achieve breeding with U-235 and Pu-239 in a thermal reactor, because the 0.07 or 0.11 neutrons are very likely to be lost by absorption in structural materials, moderator, and fission product poisons.

Table 13.1. Values of Reproduction Factor η

Neutron Energy
Isotope	Thermal	Fast
U-235	2.07	2.3
Pu-239	2.11	2.7
U-233	2.30	2.45

A thermal reactor that uses U-233 is a good prospect, but the fast reactor that uses Pu-239 is the most promising candidate for breeding. Absorption of neutrons in Pu-239 consists of both fission and capture, the latter resulting in the isotope Pu-240. If the latter captures a neutron, the fissile isotope Pu-241 is produced.

The ability to convert fertile isotopes into fissile isotopes can be measured by the conversion ratio (CR), which is defined as

The fissile atoms are produced by absorption in fertile atoms; the consumption includes fission and capture.

We can compare values of CR for various systems. First is a “burner” fueled only with U-235. With no fertile material present, CR = 0. Second is a highly thermal reactor with negligible resonance capture, in which fuel as natural uranium, 99.28% U-238 and 0.72%U-235, is continuously supplied and consumed. Pu-239 is removed as fast as it is created. Here CR is the ratio of absorption in U-238 and U-235, and because they experience the same flux, CR is simply the ratio of macroscopic cross sections, Σ^a238/Σ^a235. Inserting the cross section ratio 2.7/681 and the atom ratio (ignoring U-234) 0.9928/0.0072, we obtain CR = 0.547. Third, we ask what CR is needed to completely consume both U isotopes in natural U as well as the Pu-239 produced? It is easy to show that CR is equal to the isotopic fraction of U-238 (viz., 0.9928). Fourth, we can derive a more general relationship from the neutron cycle of Figure 11.4. The result for initial operation of a critical reactor, before any Pu is produced, is where η²³⁵ is the value for pure U-235 (i.e., 2.07). For a natural U reactor with = 0.95, p = 0.9. and ϵ = 1.03, we find

It is clear that reducing fast neutron leakage and enhancing resonance capture are favorable to the conversion process. An alternative simple formula, obtained by considerable manipulation as in Exercise 13.6, iswhere ℓ is the total amount of neutron loss by leakage and by non-fuel absorption, per absorption in U-235.

If unlimited supplies of uranium were available at very small cost, there would be no particular advantage in seeking to improve CRs. One would merely burn out the U-235 in a thermal reactor and discard the remaining U-238. Because the cost of uranium goes up as the accessible reserves decline, it is desirable to use the U-238 as well as the U-235. Similarly, the exploitation of thorium reserves is worthwhile.

When the CR is larger than 1, as in a fast breeder reactor, it is instead called the breeding ratio (BR), and the breeding gain (BG) = BR − 1 represents the extra plutonium produced per atom burned. The doubling time (DT) is the length of time required to accumulate a mass of plutonium equal to that in a reactor system, and thus provide fuel for a new breeder. The smaller the inventory of plutonium in the cycle and the larger the BG, the quicker will doubling be accomplished. The technical term “specific inventory” is introduced, as the ratio of plutonium mass in the system to the electrical power output. Values of this quantity of 2.5 kg/MWe are sought. At the same time, a very long fuel exposure is desirable (e.g., 100,000 MWd/tonne) to reduce fuel fabrication costs. BG of 0.4 would be regarded as excellent, but a gain of only 0.2 would be very acceptable.

13.2. Isotope Production and Consumption

The performance of a breeder reactor involves many isotopes of fertile and fissionable materials. In addition to the U-235 and U-238, there is short-lived neptunium-239 (2.355 d), Pu-239 (2.411 × 10⁴ y), Pu-240 (6537 y), Pu-241 (14.4 y), and Pu-242 (3.76 × 10⁵ y), as well as americium and curium isotopes resulting from multiple neutron capture. The idea of a chain of reactions is evident. To find the amount of any of these nuclides present at a given time, it is necessary to solve a set of connected equations, each of the general typewhich is similar to the statement in Section 3.3 except that “removal” is more general than “decay” in that absorption (consumption or burnup) is included.

We can illustrate the approach to solving the balance equations as differential equations. Consider a simplified three-component system of nuclides, using a shorthand for the full names of the isotopes: 1 = U-235, 2 = U-238, and 3 = Pu-239. Because all of their radioactive half-lives are long in comparison with the time of irradiation in a reactor, true decay can be ignored. However, it will be convenient to draw an analogy between decay and burnup. The equation for U-235 isand if we let φ σ^a1 = λ^a1, the equation is the same as that for decay, the solution of which is

A similar solution may be written for U-238,

The growth equation for Pu-239 is where only the capture in U-238 gives rise to Pu-239, not the fission. Assuming that there is already some plutonium present when the fuel is loaded in the reactor, in amount N³⁰, the solution iswhere E³ = exp (−λ^a3t). The first term on the right describes the burnup of initial Pu-239; the second term represents the net of production and consumption. Note the similarity in form of the equations to those in Computer Exercise 3.D related to parent–daughter radioactivity processes.

It is straightforward to calculate the numbers of nuclei, but time-consuming and tedious if one wishes to vary parameters such as the reactor power and neutron flux level or the initial proportions of the different nuclides. To make such calculations easier, refer to Computer Exercise 13.A, in which the programs BREED and BREEDGE are applied.

A one-neutron group model is not adequate to analyze the processes in a fast breeder reactor, where cross sections vary rapidly with energy. The accurate calculation of multiplication requires the use of several neutron energy groups, with neutrons supplied to the groups by fission and removed by slowing and absorption. In Computer Exercise 13.B the analysis is displayed and a simple fast reactor is computed by the program FASTR.

13.3. The Fast Breeder Reactor

LMFBRs have been operated successfully throughout the world. In the United States the Experimental Breeder Reactor I at Idaho Falls was the first power reactor to generate electricity in 1951. Its successor, EBR II, was used from 1963 to 1994 to test equipment and materials. An important feature was its closed fuel cycle, in which used fuel was removed, chemically processed, and refabricated. To accomplish these operations under conditions of high radioactivity, unique handling equipment was devised. In September 1969, the power reached its design value of 62.5 MWt (see References).

The Fermi I reactor built near Detroit was the first intended for commercial application. It was started in 1963 but was damaged by blockage of coolant flow passages and only operated briefly after being repaired.

The 400 MWt Fast Flux Test Facility (FFTF) at Richland, Washington, now shut down, did not generate electricity but provided valuable data on the performance of fuel, structural materials, and coolant (see References). After a number of years of design work and construction, the United States government canceled the demonstration fast power reactor called Clinch River Breeder Reactor Project (CRBRP). There was a great deal of debate in the United States before CRBRP was abandoned. One argument for stopping the project was that increased prices of fuel, being only approximately one fifth of the cost of producing electricity, would not cause converter reactors to shut down or warrant switching to the newer technology except on a long-term basis. This political decision shifted the leadership for breeder development from the United States to other countries.

France took the initiative in the development of the breeder for the production of commercial electric power in cooperation with other European countries. The reactor “Superphenix” was a full-scale pool-type breeder constructed with partial backing by Italy, West Germany, The Netherlands, and Belgium. Because of sodium leaks, and great public opposition, the reactor was shut down permanently. The lower power Phenix was shut down for 6 years but restarted in 2003 with excellent operation.

With the suspension of operation of Superphenix, the lead in breeder reactor development again shifted, this time to Japan, which placed its 280 MWe loop-type sodium-cooled MONJU into operation in 1993. It was part of Japan's long-range plan to construct of a number of breeders starting around 2020. In 1995 the reactor suffered a sodium leak (see References) and was shut down. Renewed interest in breeding prompted a restart of MONJU.

The largest remaining LMFBR in the world is the BN-600 at the Beloyarskiy plant in Russia. Supplying electricity since 1981, it has operated more successfully than any other reactor in that country. Some of its pertinent features are listed in Table 13.2.

Table 13.2. BN-600 Liquid Metal Fast Breeder Reactor, Beloyarskiy Unit #3, Russia From Nuclear Engineering International Magazine

Electric power	560 MW
Sodium coolant temperatures	377 °C, 550 °C
Core fuel height	1.03 m
Core diameter	2.05 m
Vessel height, diameter	12.6 m, 12.86 m
Fuel (w/o U-235)	UO2 (17, 21, 26)
Pin o.d.	6.9 mm
Cladding	stainless steel
Clad thickness	0.4 mm
Pin pitch (triangular)	9.82 mm
Pins per assembly	127
Number of assemblies	369
Number of B4C rods	27
Average power density	413 kWt/1
Cycle length	5 months

India is preparing a 1200 MWt fast breeder reactor to go into operation around 2010. It will complement their proposed Advanced Heavy Water Reactor, a thermal breeder that uses fertile Th-232 to produce fissile U-233.

The use of liquid sodium as coolant ensures that there is little neutron moderation in the fast reactor. The element sodium melts at 208 °F (98 °C), boils at 1618 °F (883 °C), and has excellent heat transfer properties. With such a high melting point, pipes containing sodium must be heated electrically and thermally insulated to prevent freezing. The coolant becomes radioactive by neutron absorption in Na-23, producing the 15-h Na-24. Great care must be taken to prevent contact between sodium and water or air, which would result in a serious fire, accompanied by the spread of radioactivity. To avoid such an event, an intermediate heat exchanger is used, in which heat is transferred from radioactive sodium to nonradioactive sodium.

Two physical arrangements of the reactor core, pumps, and heat exchanger are possible, shown schematically in Figures 13.1 and 13.2. The “loop” type is similar to the thermal reactor system, whereas in the “pot” type all of the components are immersed in a pool of liquid Na. There are advantages and disadvantages to each concept, but both are practical.

Figure 13.1. Loop system for LMFBR.

Figure 13.2. Pot system for LMFBR.

To obtain maximum BRs in the production of new fertile material, more than one fuel zone is needed. The neutron-multiplying core of the breeder reactor is composed of mixed oxide (MOX) fuel as a mixture of U and Pu. Surrounding the core is a natural uranium oxide “blanket” or “breeding blanket.” In early designs, the blanket acted as a reflector for a homogeneous core, but modern designs involve blanket rings both inside and outside the core, rendering the system heterogeneous. The new arrangement is predicted to have enhanced safety as well.

Deployment of breeder reactors demands recycling of the plutonium. This in turn requires reprocessing, which involves physical and chemical treatment of irradiated fuel to separate uranium, plutonium, and fission products. We reserve discussion of reprocessing until Section 22.5, in connection with waste disposal. The United States abandoned commercial reprocessing caused by concerns about the diversion of plutonium and is unlikely to resume the practice for the present generation of power reactors.

13.4. Integral Fast Reactor

After CRBRP was canceled, the United States continued development of breeder reactors. The success of the Experimental Breeder Reactor-II prompted an extension called the Integral Fast Reactor (IFR). It was a fast breeder reactor coupled with a pyrometallurgical process that allowed all fuels and products to remain in the system. The fuel cycle included fuel fabrication, power generation, reprocessing, and waste treatment. Through the isolation of plutonium in a highly radioactive environment, the problem of weapons proliferation was eliminated. The reactor burned all actinides, elements 89 and above. No additional fuel was introduced and the predicted lifetime was 60 y.

Fuel in the IFR was metal, as an alloy of U, Pu, and Zr, with a much higher thermal conductivity than ceramic uranium oxide. Thus in operation the centers of fuel rods were cooler than in conventional reactors at the same power. Coolant in the reactor was liquid sodium that operates at atmospheric pressure and is not corrosive. With the reactor located within a pool, the problem of loss of coolant is eliminated, there being adequate natural convection cooling.

The reactor was found to be inherently safe, as was proved by experiments with EBR-II. In one test, the power to cooling systems was cut off when the reactor was at full power. The core temperature rose slightly as convective cooling in the pool-type vessel took over. The reactor went subcritical and shut itself down.

Use of the IFR with its closed cycle and consumption of plutonium would have eliminated the concern about the spent fuel repositories being “plutonium mines,” with reactor grade plutonium available at some future date for some low-yield explosions. As such, IFR was highly proliferation resistant. The use of uranium was essentially complete, in contrast with a few percent in conventional power plants.

By the recycling of transuranic elements, the waste consisted only of fission products, which would need to be stored for a relatively short time. The need for extra waste repositories would vanish if IFR-type reactors were deployed. The reactor system was expected to be less expensive to build because of its simplicity. Most of the complex systems of water-cooled reactors are not needed, and the cost of basic fuel is essentially zero because of large inventory of uranium-238 in the depleted uranium from years of operation of separation plants.

The reactor serves as a source of experience and data and a starting point for the proposed Advanced Fast Reactor of the 21st century. Finally, it shares the virtue of all nuclear reactors in not releasing gases that could contribute to climate change.

Even though there were many potential benefits to the continuation of R&D on the IFR, Congress chose not to provide funds in 1994. In a Q&A session by George Stanford (see References), “Well-meaning but ill-informed people…convinced so many administrators and legislators that the IFR was a proliferation threat that the program was killed.” There remains a possibility that the concept will be revived, in view of the extensive knowledge base that was developed and the promise that IFR has to solve many of nuclear's perceived problems.

A commercial outgrowth of the IFR was the Advanced Liquid Metal Reactor (ALMR) or PRISM, involving Argonne National Laboratory and General Electric (see References). That design project was also terminated.

Although the principal attention throughout the world has been given to the liquid metal–cooled fast breeder that uses U and Pu, other breeder reactor concepts might someday become commercially viable. The thermal breeder reactor, which uses thorium and uranium-233, has always been an attractive option. Favorable values of CRs are obtained with thorium as fertile material (see References). Neutron bombardment of Th-232 in a thermal reactor yields fissile U-233 as noted in Section 6.3. However, there is a special radiation hazard involved. Two neutron absorptions in U-233 yield 69 y half-life U-232 that decays into Tl-208, a 2.6 MeV gamma emitter. India, which has large reserves of thorium, is especially interested in a U-233 breeder cycle.

One extensive test of that type of reactor was the Molten Salt Reactor Experiment at Oak Ridge, an outgrowth of the aircraft nuclear program of the 1960s (see Section 20.6). The reactor demonstrated the feasibility of the circulating fuel concept with salts of lithium, beryllium, and zirconium as solvent for uranium and thorium fluorides. Other concepts are (a) uranium and thorium fuel particles suspended in heavy water, (b) a high-temperature gas-cooled graphite-moderated reactor containing beryllium, in which the (n,2n) reaction enhances neutron multiplication.

13.5. Breeding and Uranium Resources

From the standpoint of efficient use of uranium to produce power, it is clearly preferable to use a breeder reactor instead of a converter reactor. The breeder has the ability to use nearly all of the uranium rather than a few percent. Its impact can be viewed in two different ways. First, the demand for natural uranium would be reduced by a factor of approximately 30, cutting down on fuel costs while reducing the environmental effect of uranium mining. Second, the supply of fuel would last longer by the factor of 30. For example, instead of a mere 40 y for use of inexpensive fuel, we would have 1200 y. It is less clear, however, as to when a well-tested version of the breeder would actually be needed. A simplistic answer is, “when uranium gets very expensive.” Such a situation is not imminent, because there has been an oversupply of uranium for a number of years, and all analyses show that breeders are more expensive to build and operate than converters. A reversal in trend is not expected until some time well into the 21st century. The urgency to develop a commercial breeder has lessened as the result of slower adoption of nuclear power than anticipated, with the smaller rate of depletion of resources. Another key factor is the availability in the United States and the former U.S.S.R of large quantities of surplus weapons plutonium, which can be used as fuel in the form of MOX.

Uranium resource data as of 2006 from the “Red Book” (see References) are reported by the World Information Source on Energy (WISE) (see References). Recoverable amounts of uranium are given for costs in $/kgU <40, <80, and <130. We show in Table 13.3 the sums by main country sources of two categories: reasonably assured resources (RAR) and inferred resources (IR). The world total is 2,643,340 tonnes. The table also gives the annual uranium requirements by principal country users. The world total per year is 66,877 tonnes. Simple arithmetic tells us that these resources would last 40 y, assuming constant fuel requirements.

Table 13.3. Uranium Demand and Resources (in 1000s of tonnes) From OECD-IAEA Report (see References)

Country	Annual Demand (2006 est.)	Country	Reasonably Assured Resources to $80/kg
United States	22.875	Australia	714.00
Japan	8.670	Kazakhstan	378.29
France	7.185	Canada	345.20
Russia	4.465	Niger	180.47
Korea	3.400	South Africa	177.15
Germany	2.900	Brazil	157.70
Ukraine	2.350	Namibia	151.32
Canada	1.700	Russia	131.75
China	1.565	United States	102.00
United Kingdom	1.500	Uzbekistan	59.74
Belgium	1.455	Ukraine	58.50
Sweden	1.400	Mongolia	46.20
Spain	1.140	China	38.02
Bulgaria	0.840	Jordan	30.37
Taiwan	0.830	Algeria	19.50
Czech Republic	0.700	Malawi	8.77
Finland	0.557	Turkey	7.39
Slovakia	0.450	Portugal	6.00
Brazil	0.450	Bulgaria	5.87
India	0.380	Argentina	4.88
Hungary	0.370	Italy	4.80
Mexico	0.355	Spain	2.46
South Africa	0.280	Congo	1.35
Switzerland	0.270	Zimbabwe	1.35
Lithuania	0.190	Peru	1.22
Slovenia	0.160	Slovenia	1.21
Argentina	0.120	Greece	1.00
Romania	0.100	Czech Republic	0.51
Armenia	0.090	Indonesia	0.32
Pakistan	0.065
Netherlands	0.065
Total	66.877	Total	2643.34

The use of global figures obscures the problem of distribution. In Table 13.3 we list the top countries in the categories demand and resources. Some surprising disparities are seen. The leading potential uranium suppliers, Australia and Kazakhstan, are not on the list of users. On the other hand, the second highest user, Japan, has negligible U resources. Thus there must be a great deal of import/export trade to meet fuel needs. At some time in the future, in place of the Organization of Petroleum Exporting Countries (OPEC), there is the possibility of an “OUEC” cartel. Alternately, it means that for assurance of uninterrupted production of nuclear power, some countries are much more interested than others in seeing a breeder reactor developed.

Some data on United States uranium production are shown in Table 13.4. Not included are byproducts of phosphate and copper mining, or the large stockpile of depleted uranium as tails from the uranium isotope separation process. Such material is as valuable as natural uranium for use in a blanket to breed plutonium. The principal United States deposits in order of size are in Wyoming, New Mexico, Colorado, Texas (coastal plain), and near the Oregon–Nevada border. The greatest concentration of estimated additional resources are in Utah and Arizona. Most of the ores come from sandstone; approximately 30 uranium mills are available. Exploration by surface drilling has tapered off continually since the middle 1970s when nuclear power was expected to grow rapidly.

Table 13.4. United States Uranium (2006) DOE/EIA (see References)

Estimated reserves at $30/lb 74 million tonnes, at $50/lb 424 million tonnes

Annual uranium mine production 4.7 million pounds U3O8

Total mines and sources 11

Employment (person-years) 755

Total expenditures $221 million

There is considerable sentiment in the nuclear community for storing spent fuel from converter reactors rather than burying it as waste, in anticipation of an energy shortage in the future as fossil fuels become depleted. If such a policy were adopted, the plutonium contained in the spent fuel could be recovered in a leisurely manner. The plutonium would provide the initial loading of a new generation of fast breeder reactors, and the recovered uranium would serve as blanket material.

The energy content of uranium is so high that the cost of fuel for nuclear power plants is relatively small, of the order of 5% of the operating cost. However, if shortages of inventory occur because of inadequacy of supply, the price may rise significantly.

Around 1980 a peak price of $46/lb U³O⁸ was reached, but in subsequent decades dropped to a figure in 2000 of only $7/lb. The reason was the appearance of secondary sources such as depletion of inventories and released weapons uranium. Lesser amounts of secondary fuel can come from recycling, re-enrichment of separation tails, and surplus plutonium. These alternatives are expected to decline in the coming years. As the renaissance of nuclear power occurs in the United States and countries such as France, India, South Korea, and China expand their nuclear capabilities, the price of uranium is likely to increase.

Coupled with the increased demand for uranium will be a growth in exploration and mining, but this will be hampered by the lack of qualified personnel. There is said to be ample resources, but investment in retrieval may be slow in coming until a stable price structure is established.

Finally, there is a large delay between discovery of new resources and the availability of nuclear fuel, in part because of a hodgepodge of regulatory oversight and the effect of public opposition. Thorough discussion of these issues appears in articles in Nuclear News, March 2006.

It is not possible to predict the rate of adoption of fast breeder reactors for several reasons. The capital costs and operating costs for full-scale commercial systems are not firmly established. The existence of the satisfactory LWR and the ability of a country to purchase slightly enriched uranium or MOX tends to delay the installation of breeders. It is conceivable, however, that the conventional converter reactors could be replaced by breeders in the present century because of fuel resource limitations. It is possible that the breeder could buy the time needed to fully develop alternative sources such as nuclear fusion, solar power, and geothermal energy. In the next chapter the prospects for fusion are considered.

13.6. Summary

If the value of the neutron reproduction factor η is larger than 2 and losses of neutrons are minimized, breeding can be achieved, with more fuel produced than is consumed. The CR measures the ability of a reactor system to transform a fertile isotope (e.g., U-238) into a fissile isotope (e.g., Pu-239). Complete conversion requires a value of CR of nearly 1. Fast breeder reactors that use liquid sodium with BRs greater than 1 have been built and operated, but several development programs have been canceled. One large-scale breeder continues to operate in Russia. There is a great disparity between uranium resources and uranium use among the countries of the world. Application of the breeder could stretch the fission power option from a few decades to centuries.

13.7. Exercises

1. What are the largest conceivable values of the CR and the BG?
2. An “advanced converter” reactor is proposed that will use 50% of the natural uranium supplied to it. Assuming all the U-235 is used, what must the CR be?
3. Explain why the use of a natural uranium “blanket” is an important feature of a breeder reactor.
4. Compute η and BG for a fast Pu-239 reactor if ν = 2.98, σ^f = 1.85, σ^c = 0.26, and ℓ = 0.41. (Note that the fast fission factor ϵ need not be included.)
5. With a BR = 1.20, how many kilograms of fuel will have to be burned in a fast breeder reactor operating only on plutonium to accumulate an extra 1260 kg of fissile material? If the power of the reactor is 1250 MWt, how long will it take in days and years, noting that it requires approximately 1.3 g of plutonium per MWd?

6. • By use of the neutron cycle, Figure 11.4, find a formula for ℓ as defined in Section 13.1.
   • Calculate the value of ℓ and verify that the alternative formula gives the same answer as in the text, CR = 0.750.

Computer Exercises

1. A breeder reactor is successful if it produces more fissionable material than it consumes. To test that possibility apply computer programs BREED and BREEDGE. The first of these uses cross sections for U-235, U-238, and Pu-239 as deduced from early critical experiments on weapons material assemblies. The second uses more modern cross sections, appropriate to a power reactor design. (a) Run the programs, varying parameters, to explore trends. (b) Use the following common input on both programs: U-235 atom fraction 0.003 (depleted U), plutonium volume fraction 0.123, fast flux 4.46 × 10¹⁵ cm⁻² s⁻¹. (c) Discuss observations of trends and seek to explain in terms of assumed cross section sets.
2. Program FASTR solves the neutron balance equations for a fast reactor with classic 16-group Hansen-Roach cross sections prepared by Los Alamos. Those input numbers are found in the report Reactor Physics Constants, ANL-5800, 1963, page 568 ff. Run the program with the menus, observing input data and calculated results. Compare results for the case of pure U-235 with those obtained in Computer Exercise 11.A, with program CRITICAL.

13.8 References

Wirtz, 1976 Karl Wirtz, Lectures on Fast Reactors 1976 American Nuclear Society La Grange Park, IL

Vendreyes, March 1977 George A. Vendreyes, Superphenix: A Full-Scale Breeder Reactor Scientific American March 1977 W. H. Freeman Co San Francisco26-

Judd, 1981 A.M. Judd, Fast Breeder Reactors: An Engineering Introduction 1981 Pergamon Press Oxford

Waltar and Reynolds, 1981 Alan E. Waltar, Albert B. Reynolds, Fast Breeder Reactors 1981 Pergamon Press New York

Zaleski, 1985 C. Pierre Zaleski, Fast Breeder Reactor Economics Karl O. Ott, Bernard I. Spinrad, Nuclear Energy: A Sensible Alternative 1985 Plenum Press New York and London

Stevenson, 1987 Charles E. Stevenson, The EBR-II Fuel Cycle Story 1987 American Nuclear Society La Grange Park, IL

EBR-II EBR-II: http://www.nuc.berkeley.edu/designs/ifr/ebr.html

Experimental Breeder Reactor at Argonne National Laboratory Experimental Breeder Reactor at Argonne National Laboratory.

Cochran, 1974 Thomas B. Cochran, The Liquid Metal Fast Breeder Reactor 1974 Resources for the Future, Johns Hopkins University Press Baltimore, MD

Liquid Metal Fast Breeder Liquid Metal Fast Breeder

http://www.atomicinsights.com/oct95/LMFBR_oct95.html http://www.atomicinsights.com/oct95/LMFBR_oct95.html

History and status as viewed by Rod Adams in issue of Atomic Energy Insights History and status as viewed by Rod Adams in issue of Atomic Energy Insights.

An Introduction to Argonne National Laboratory's Integral Fast Reactor (IFR) Program An Introduction to Argonne National Laboratory's Integral Fast Reactor (IFR) Program.

http://www.nuc.berkeley.edu/designs/ifr/anlw.html http://www.nuc.berkeley.edu/designs/ifr/anlw.html

George George S. Stanford, “Integral Fast Reactors: Source of Safe, Abundant, Non-Polluting Power.”

http://www.nationalcenter.org/NPA378.html http://www.nationalcenter.org/NPA378.html

A double issue of Progress in Nuclear Energy A double issue of Progress in Nuclear Energy, Vol. 31, Nos. 1–2 (1997) with title “The Technology of the Integral Fast Reactor and its Associated Fuel Cycle.”

http://www.nuc.berkeley.edu/~gav/almr/01.intro.html http://www.nuc.berkeley.edu/~gav/almr/01.intro.html

Description provided by University of California Description provided by University of California.

Status of liquid metal cooled fast reactor technology Status of liquid metal cooled fast reactor technology

http://www.iaea.org/inisnkm/nkm/aws/fnss/fulltext/30023917.pdf http://www.iaea.org/inisnkm/nkm/aws/fnss/fulltext/30023917.pdf

1999 IAEA report 1999 IAEA report (544 pages).

Fast Neutron Reactors Fast Neutron Reactors

http://www.uic.com.au/nip98.htm http://www.uic.com.au/nip98.htm

Comprehensive essay Comprehensive essay.

Uranium 2005 Resources, Production and Demand, 2006 Uranium 2005 Resources, Production and Demand 2006 OECD Nuclear Energy Agency and International Atomic Agency Paris

http://www.nea.fr/html/ndd/reports/2006/uranium2005-english.pdf http://www.nea.fr/html/ndd/reports/2006/uranium2005-english.pdf.

A biennial publication. Executive summary of “Red Book” (the cover is red) A biennial publication. Executive summary of “Red Book” (the cover is red).

Uranium Information Centre (Australia) Uranium Information Centre (Australia)

http://www.uic.com.au http://www.uic.com.au

Select Briefing Papers Select Briefing Papers. More than 100 articles on all topics.

Perry and Weinberg, 1972 A.M. Perry, A.M. Weinberg, Thermal Breeder Reactors Annual Review of Nuclear and Particle Science 22 1972317-

Kang and von Hippel, 2001 Jungmin Kang, Frank N. von Hippel, U-232 and the Proliferation-Resistance of U-233 in Spent Fuel Science & Global Security 9 20011-

Article is also available on the Web at Article is also available on the Web at http://www.princeton.edu/~globsec/publications/pdf/9_1kang.pdf

World Information Source on Energy Uranium Project World Information Source on Energy Uranium Project

http://www.wise-uranium.org http://www.wise-uranium.org

The Web site for WISE includes a Nuclear Fuel Supply Calculator and listings of organizations handling fuel The Web site for WISE includes a Nuclear Fuel Supply Calculator and listings of organizations handling fuel.

United States Uranium Reserves Estimates United States Uranium Reserves Estimates

http://www.eia.doe.gov/cneaf/nuclear/page/reserves/ures.html http://www.eia.doe.gov/cneaf/nuclear/page/reserves/ures.html

Domestic Uranium Production Report Domestic Uranium Production Report

http://www.eia.doe.gov/cneaf/nuclear/dupr/dupr.html http://www.eia.doe.gov/cneaf/nuclear/dupr/dupr.html