Toward a White Paper on a Mass Global Deployment of Nuclear Power: Fuel

It is my intention to present a draft of a White Paper, which will lay out a plan for a global deployment of nuclear power plants in sufficient numbers to insure the goal of an 80% reduction of global CO2 emissions can be accomplished by 2050. The first section of this draft focused on nuclear safety. This section will focus on the role of deploying Generation IV molten salt and sodium cooled reactors in meeting the goal, and the role of Reactor Grade Plutonium drawn from nuclear waste, in starting the breeding process. In addition to the use of RGP, a drawdown of nuclear weapons grade HEU and Pu-239 to start Generation IV reactors is both possible and desirable, but this possibility is not addressed in the body of this paper.)

Finding fuel for a post-carbon energy future

Energy policy should be guided by well thought out goals, and be based on rational and realistic plans. Unfortunately, both American and International energy efforts have not bee based on well thought out goals, or rational and realistic energy plans. The problem is especially serious because of the frequent inclusion of renewable energy goals in the formulation of over all national and international energy goals. In order for goals to meet minimal standards of plausibility, it should be demonstrated that a realistic plan for the implementation of the goal must exist. A realistic plan would be one which would address objections, which could in fact be carried out to successful conclusions.

Climate scientist have argued that it is imperative to lower global Carbon Dioxide emissions by 80% before 2050. In addition, the rapid economic development of nations such as India and China suggest that Global energy demand will increase, at the very time when there appears to be an need to decrease global energy production from carbon based fuels. Shifting global energy production from fossil fuels, while at the same time filling the energy demands of several billion people from developing countries is an enormous and extremely problematic task, which must be done.

Nuclear power is a promising candidate to bridge at least part of the energy gap. However, any account of a massive global deployment of nuclear reactors as a major or the major post carbon energy source faces a number of challenges. Critics can be expected to raise issues related to the cost of nuclear power, the magnitude of the task, the time requirements for nuclear construction, the availability of sufficient nuclear fuel the proliferation issues, the problems related to the management of nuclear waste, nuclear safety issues, and the potential dangers of nuclear terrorism. A successful nuclear global deployment plan should foresee these objections and demonstrate that no single objection or set of objections points to an insurmountable problem. A goal for a nuclear deployment plan would be the inclusion of a demonstration that no set of objections could motivate rational objections to the global nuclear deployment scheme.

Of course, all renewables deployment schemes, should also face the same level of scrutiny, with questions focusing on capacity factors, cost, the use of fossil fuel powered renewable backup systems, grid stability, redundancy, and the cost and efficiency of energy storage schemes. Efficiency advocates need to explain why Jevons Paradox and/or the rebound effect would not apply to efficiency based carbon mitigation plans. Unfortunately to date, renewable/efficiency based post-carbon energy plans to date appear to be singularly lacking in candor in identifying and addressing their problems.

Senior Collell poses a problem for a uranium powered future

Marcel Coderch Collell, a distinguished Spanish technocrat, has reviewed the possibility of replacing much of the worlds fossil fuel generated electricity by 2030. He argues that the limited nuclear fuel supply mans that nuclear power cannot play a predominate role in a post carbon energy order. He is extremely critical of the French Model: Electricity from Nuclear and/or renewable power.

He describes the French Model

The French model: electricity, nuclear or renewable One of the first options to consider would be to follow the French model progressively increasing the reactor park to ensure that by 2030, or maybe a little later, much of the world's electricity is predicted to generate fossil fuels out of nuclear origin, since it does not require, in principle, some technical innovation. Thus, power generation would not produce broadcasts, because it was nuclear or renewable. This would save huge amounts natural gas and coal, and oil-well with a consequent reduction emissions, and could force down, or at least not to contribute upwards of prices of fossil fuels and expand its availability in time.

But there are, according to Senior Collell, problems with the French Model.

But leave aside the logistical difficulties (and financial) would be a nuclear construction program of this size and evaluate how much fuel would be needed to fuel a reactor park of this magnitude, and which could be another limiting factor. Surely it would be, mainly, of building thermal neutron reactors with a Third Generation quasi-cycle Open fuel (MOX fueled with uranium enriched with some plutonium). In the best case, not expected to be operational by 2030 the Fourth Generation of fast neutron reactors with closed fuel cycles (which is expected to reach 60 times the performance thereof, by the massive use of plutonium) and, therefore, in the coming decades should be the main nuclear fuel.

Senior Collell then suggests that in order to follow the nuclear French model by 2030,

4,740 new 1GWe reactors would have to be built and [one] put in operationevery two days for the next 25 years

Senior Collell then suggests that this buildout would be very difficult in a business as usual world.

An optimistic estimate of construction times (five years) would mean having 950 teams of technical specialists, workers and machinery simultaneously working full time. This is hard to imagine, despite talk of standardising designs. In the previous period of nuclear construction (1963-88) only 423 reactors were built, at a rate of 17 per year.

He also argues that fuel shortages would constrict the depolyment of such a large reactor fleet.

A simple calculation suffices to show how an extension of the French model would collide with a scarcity of uranium. This is old news, given the serious doubts that already exist regarding the availability of uranium even to feed a few more reactors than now exist. In 2004, 365 GWe of nuclear capacity consumed about 67 kt of uranium (approximately 180 tons of uranium perGWe per year), of which 36 kt came from currently operating mines, while the rest came from recycled nuclear weapons and other secondary sources (that is, from prior production). Supply forecasts for the reactors currently in operation (plus foreseeable growth) put uranium mining production at 50 kt per year in 2015, with a significant shortfall developing in 2010, by which time Russia's nuclear weapons will have been dismantled and their uranium will have been consumed, . . .

If we assume linear growth from the current 365 GWe to 4,959 GWe in 2030, uranium demand would be around 400 kt in 2015 and 700 kt in 2030. This means multiplying by eight today’s estimates of production capacity in 2015, and multiplying by fifteen for 2030

Senior Collell sees these facts as casting the nuclear build out on the horns of a dilemma.

Let us suppose, however, for argument’s sake, that it were possible to achieve a production capacity of 700 kt/year by 2030. In the context of this analysis, two questions are raised: first, the CO2 emissions that would be generated in this phase of the nuclear cycle. Given the amount of uranium necessary, it would almost certainly be necessary to make use of hard rock deposits and low concentrations ore. There are fortunately multiple flaws in this delima argument. First the rock does not have to be moved. Uranium miners are increasingly adopting a mining technique called in situ leaching. When in situ leaching is practiced on uranium ore, the primarily the uranium is extracted, and the rock is left in place. Let us ignore for the moment these problems.

The problem that Senior Collell is pointing to is the limitation of the Light Water Reactor. Light Water Reactors were first developed as a means of powering American Nuclear submarines. In American Nuclear Submarines LWRs are small, they provide reliable power for 15 years, after which their cores can be replaced. There are also expensive, but nothing can serve as a substitute. Large power reactors can be even more expensive and there are very fuel inefficient. Part of the problem has to do with the flaws in the Uranium cycle. In LWRs as little as 0.3% of the potential fuel gets burned, and the rest falls into a category called "nuclear "waste." The problem is that uranium is relatively cheap, so it cost less to separate out the good stuff, the U-235 and use it for nuclear fuel. In LWRs a tiny fraction of the fuel gets converted to fissionable Pu-239, and a fraction of that gets burned as nuclear fuel. Unfortunately Pu-239 is not very good fuel in LWRs. When enough u-235 and Pu-239 is "burned" in a :WR, the fuel ceases to sustain a chain reaction, and has to be replaced. The now used fuel, still contains a significant amount of U-235 and an even larger amount of plutonium isotopes. Many of the plutonium isotopes are not fissionable, and plutonium produced in Light Water Reactors is not of weapons quality. It is also not very good nuclear fuel in Light Water Reactors.

The new French model of the future from Grenoble

Another French Model, one not contemplated by Senior Collell, is offered by Scientists of the Reactor Physics Group at Laboratoire de Physique Subatomique et de Cosmologie, Grenoble, France, together with other associates, have made important contributions to our understanding of the fuel management problems that would be faced in a Global Nuclear Deployment. The RPG were until very recently among the handful of world scientists who understood the potential of Molten Salt Reactor technology. They are among the most important and far reaching energy thinkers in the world, yet outside of the narrow circle found in the Energy from Thorium Discussion Forum, their work is almost unknown in the United States. Scientists from the RPG wrote a number of papers examining the potential problems of a Global Nuclear Deployment. "Molten Salt Reactors and Possible Scenarios for Future Nuclear Power Deployment offers a good introduction to their work. as a means of further enhancing the intellectual legitimacy of the Energy from Thorium approach to climate change mitigation, and to further enhance American Awareness of the work of the LPSC on Molten Salt technology. LPSC researchers use the term Thorium Molten Salt Reactor. Energy annalists associated with the Energy from Thorium approach to climate change mitigation use the term Liquid Fluoride Thorium Reactor or LFTR. In Future Deployment they write:

The worldwide demand for primary energy is constantly increasing and, if it is to be satisfied, solutions must be thought out and the extent to which the responses are adapted to the issue must be examined. There are not so many options once it is agreed that recourse to fossil energies should be as reduced as possible in order to limit green house gas emissions. Fission based nuclear energy is, along with new renewable energies and, in the longer run, fusion based energy, one of the primary energy sources capable of contributing significantly to satisfying the demand. The scenarios studied in our group show the potential, and limitations, for a worldwide deployment of nuclear power, and demonstrate that the different reactor types are quite complementary. This study shows that fissile matter availability comes as a strong constraint if a fleet of reactors able to breed their own fuel is to be started. In addition, such breeder reactors will not be deployed industrially before the next 20 to 25 years so that any transition towards extensive and sustainable nuclear power production will have to call on second or third generation light water reactors, which will have to be built.

We have considered three main reactor types:

• Pressurized water reactors of the second generation (PWR) and third generation (EPR - European Pressurized Reactor). These reactors do not breed their fuel. PWRs are currently in operation while EPRs will begin production in 2010 in our scenarios.
• Fast neutron reactors with a liquid metal coolant (FNR). These are fourth generation reactors that are based on the ²³⁸U/Pu fuel cycle. Their breeding ratio varies with the scenario considered. FNRs begin production in 2025 in our scenarios.
• Molten salt reactors (MSR). These are fourth generation reactors based on the²³²Th/ ²³³Ufuel cycle, with a neutron spectrum that can be anything from thermal to fast. These begin production in 2030 in our scenarios.

Our studies show that an intensive nuclear power deployment is feasible but that it requires careful handling of fissile matter resources and of nuclear wastes. The scenario that combines the three reactor types is by far the one that gives the most flexibility in the deployment of nuclear power; if necessary, it could accommodate more intensive production than we have set in our scenarios. The three reactor types complement each other strikingly; the use of natural fissile matter is optimized (figure 3); the volume of trans-uranians produced is minimal; the option to stop, then restart nuclear power production remains open so that decisions are not irreversible. Intermediate scenarios, with a greater or lesser contribution of FNRs as compared to MSRs can be considered in order to satisfy regional or other criteria but, with these studies, it appears that the ²³²Th/ ²³³U fuel cycle will be needed early on.

Figure 3: Evolution of natural uranium resources for the three scenarios considered.

Figure 4: Amounts of plutonium and 233U present in the fuel cycles of the reactors for the three scenarios considered.

The French Scientists from the University of Grenoble offered a more detailed analysis of the fuel management problem elsewhere. In "Scenarios with an Intensive Contribution of Nuclear Energy to the World Energy Supply," H.Nifenecker, D.Heuer, S.David, J.M.Loiseaux1, J.M.Martin, O.Meplan, and A.Nuttin, maintain that

If carried out with PWR or BWR reactors, the important nuclear power deployment will make heavy demands on natural Uranium resources. Resources are, presently, estimated to be around 20 Million tons. Assuming PWR or BWR reactors, the cumulative needs in 2050 could reach 16 million tons. This shows that breeding reactors are necessary to meet the needs or, alternately, that Uranium would have to be extracted from sea water, at a significant cost.

These considerations may, however, probably exaggerate the Uranium shortage. It is by no means certain that the cost of extracting Uranium from sea water would exceed the cost of breeding. On the other hand, there are reasons to suspect that the cost of LFTRs and other Molten Salt Reactors might be significantly lower than the cost of Light Water Reactors. Certainly when the large global thorium stock is added to recoverable uranium there will be no shortage of nuclear for a long time to come. Alvin Weinberg relates how the possibility of a future global uranium shortage was understood by the founding fathers of the Nuclear age, including Enrico Fermi, and Eugene Wigner. It was understood even then that nuclear breeding technology would have to be introduced as a part of a global deployment of nuclear power.

In "Intensive Contribution," the French team reviewed

two possible breeding cycles:
* The U-Pu cycle using fast reactors
* The Th-U cycle using thermal reactors

This analysis was expanded with typical French thoroughness in "Scenarios for a Worldwide Deployment of Nuclear Power," if anyone is interested. Both "Intensive Contribution," and "Worldwide Deployment" came to the same conclusion, that a deployment of Light Water Reactors can only be sustained until 2030. Lets call this the conservative case. Conservative, in that it is based on very conservative estimates of global uranium resources. While far more generous estimates of Uranium resources are justifiable, their actual existence is by no means certain. A really plausible plan should make conservative assumptions. If generous assumptions do not pan out, then the plan can be altered in to reflect a better than expected resource picture.

The nuclear intensive plan would assume a nuclear build out to 3387 GWe of electrical generating capacity by 2030. This is, in itself an enormous and extremely daunting build out, and indeed suggests that a major revolution in nuclear manufacturing technology will be required. Fortunately many of the components of that revolution are already understood, and none of them represents a serious impediment to technological change. Factory production of reactor construction kits, together with on site labor saving machines, and new materials savings reactor designs can be expected to improve reactor manufacturing, labor, time and materials efficiencies during the next decade, and to be reinforced by a learning curve. Such a large build out will probably require a shift of many reactor manufacturing activities from the final manufacturing site to factories. The recycling of old steam plant locations as nuclear power stations sites, will also save money and time for the buildout.

Thus while ambitious, the 3387 GWe buildout by 2030 is still not impossible, but the goal must be set soon. Both "Intensive Contribution," and "Worldwide Distribution" then looked at the U-Pu fast reactor cycle. By 2030 an enormous amount of reactor grade plutonium will become available. This RGP can be put to use both in the production of nuclear power and in the breeding of more reactor fuel. Doing so would serve as at least a partial solution to what is commonly seen as a major problem for nuclear power, the so called nuclear waste problem. Indeed the reuse of nuclear fuel turns "nuclear waste," into an asset. "Intensive Contribution," argues that given the supply of plutonium for LWRs and fast breeders, a buildout to 9000 GWe by 2050 is possible.

"Worldwide Deployment" looks at a number of added options including burning recycled RGP in LWRs. This delays, perhaps for a hundred years, but does not prevent the eventual draw down of fissionable materials that are tied to a non-breeding nuclear economy. A better use of the RGP is

Thus the transition to some form of nuclear breeding will be inevitable, if a long term commitment to nuclear power becomes a matter of policy.

Fast sodium cooled reactors are conventionally viewed as the preferred method of nuclear breeding, although various Molten Salt Reactor breeding options exit, and include many attractive features that are more than competitive with what liquid sodium cooled breeder reactors such as the Integral Fast Reactor. IFR backers claim higher breeding ratios, but there are reasons to doubt that high breeding ratios are compatible with optimal safety.

"Worldwide Deployment" also reviews a gas cooled fast reactor option, but did not like it as well as the sodium cooled concept.

There is a significant problem with the start up charge of fast breeders. Fast neutron reactors require much more fissionable material to maintain a chain reaction. If anything the French team underestimated the amount of plutonium required to maintain a high breeding ratio in a Fast Breeder Reactor. A research report from the S-PRISM design team, indicated that the RGP in 40,000 tons of spent nuclear fuel (about 400 tons of RGP) would start 22 IFRs capable of producing 33,440 MWe output. This would suggest that the IFR would be a useful way to use and with a low conversion ration, use up RGP, but not a major adjunct to fighting AGW. A higher breeding ration is possible, but report author, Allen E. Dubberley of GE Nuclear Energy, and his associates, did not discuss the safety problems related to the high breeding ratio.

The problem with the entire sodium cooled fast reactor scheme is stated simply by Lawrence M. Lidsky and Marvin M. Miller of the Massachusetts Institute of Technology,

The LMFBR was chosen over other breeder reactor designs because it was, in theory, capable of very short fuel doubling times, shorter than that of any competing reactor design. The doubling time is the time required to produce an excess of fuel equal to the amount originally required to fuel the reactor itself. In other words, in one doubling time there would be enough fuel available to start up another reactor. In the absence of mined uranium, only a short doubling time would, it was believed, allow nuclear power to grow fast enough to compete with alternative sources of power. Unfortunately, the theoretical advantages of the LMFBR could not be achieved in practice. A successful commercial breeder reactor must have three attributes; it must breed, it must be economical, and it must be safe. Although any one or two of these attributes can be achieved in isolation by proper design, the laws of physics apparently make it impossible to achieve all three simultaneously, no matter how clever the design.

Lidsky and Miller conclude,

Strong support for plutonium recycle, with its associated technical risks and societal costs, in the face of increasing evidence that alternative strategies are superior, is clearly counterproductive.

The Lidsky and Miller superior alternative strategies involve the the employment of Molten Salt Thorium Breeders, that is LFTRs,

Thorium fuel cycles have also been promoted on the basis of lower long-term waste toxicity and greater proliferation resistance, . . . The initial rationale for introduction of the thorium cycle was the perception that it was more abundant than uranium, and that it could be used to breed U-233, an isotope with superior properties for use in thermal reactors.

The principle safety problem with Sodium cooled breeders is the possibility that a bubble in overheated sodium would lead to an uncontrolled increase of power, which interm enlarges the size of the bubble. I should here, in the interest of fairness, note that I have this week come across papers from Argonne National Laboratory which reports findings from simulation studies which suggest that Integral Fast Reactors are safe even in configurations which produce high void worth ratios. If IFRs are in fact safe despite high void worth ratios, then they can probably be pushed to higher breeding rations, but it should also be pointed out that current ANL IFR designs emphasize low breeding ratios, and in fact appear design at a ratio that is equivalent to the anticipated breeding ratio of the LFTR.

The new Indian Fast Breeder is expected to produce new fuel at a 1.12 breeding ratio, far less than its theoretical maximum. Such conservative ratios may at least partially be motivated by safety concerns. It should be noted that a 1.12 to 1 breeding ratio is quite good by LFTR standards, and would be quite satisfactory if a fast breeder could match thermal Molten Salt Breeder's start up charge. In fact as many as 12 MSBRs can be started for every LMFBR, and if the LMFBRs could breed at its theoretical maximum, they could over time produce more fissionable materials than the MSBR, were it not for safety concerns related to higher breeding ratios.

Unlike the sodium cooled LMFBR which the laws of nature appear to frown on, the face of nature positively shines on Molten Salt Nuclear technology, which is an extremely "Green" energy source. Small breeding start up charges, mean rapid scaleability of reactor production and start up. Small start up charges mean that te availability of fissionable materials will not be a factor in determining how many and how fast future MSRs are built.

In addition, low breeding ratios may be viewed as a desirable proliferation control measures. In hugh ratio breeders, breeding surpluses can be viewed a potential proliferation tools through diversion. In a 1 on 1 converter, fuel diversion for weapons purposes will lead to reactor shut down, an undesirable consequence that would decrease the likelihood of proloferation.

Given even conservative estimates of American reactor fleet growth before 2050, the United States will have produced more than enough RGP by then starting up a fleet of breeding capable MSRs capable of supplying 80% of its energy needs. In the absence of Molten Salt reactors, LMFBRs breeding at the reported Indian Breeding ratio of 1.12 to 1 would be helpful, but would probably lead to a somewhat slower nuclear deployment of nuclear power. If the report of IFR safety at higher breeding ratios is correct then the IFR might improve the fast reactor picture. It is likely that ambitious goals such as a 80% reduction of CO2 emissions by 2050 could not be meet by use of conventional nuclear alone. Were MSRs unavailable in 2050 there would likely to be significan consequences in terms of energy prices and availability in a post carbon world. .

(I plan to include an account of the proposed Indian nuclear system as an appendix to this section of the White Paper describing the ambitious and complex long term Indian nuclear program. In addition I plan to discuss Molten salt fast breeder options in a second appendix.)

Progress Toward an American Gas Cooled Reactor and Beyond

I am no longer a big fan of pebble bed reactors. At one time, I was excited about there more advanced features, but then I got a better understanding of their draw backs. The Chinese have a project to build PBRs, quite a lot of them. But as it turns out they are not going to be any cheaper than Chinese built light water reactors. The only advantage of the PBR then would be its ability to operate at a far higher temperature than light water reactors. That heat potentially makes PBRs useful as a source of industrial heat.

Industrial heat is important because there are no renewable post-carbon energy sources that are particularly good at producing the sort of heat that is required by many industrial processes. Very high temperature nuclear reactors turn out to be about the only source of industrial process heat that does not produce CO2. If any nation wants a chemical industry, or even manufacture cement in the post carbon era, it will need access to a very high temperature nuclear technology.

Somewhere in the bowels of the United States Department of Energy there are people who know this. They have reviewed the Generation IV options, and in their somewhat less than great wisdom, have determined that the United States should spend $4,000,000,000 between now and 2021 to produce an American gas cooled pebble bed reactor. Now four billion is not a lot to spend to build a new reactor prototype, but then pebble bed technology has had and continues to have a lot of money spent on its research. So the design of the very high temperature reactor will not exactly start from scratch. There will still be some questions to be asked and answered, however, before the prototype goes critical.

Among the attractive features of the pebble bed reactor design are some nifty passive or natural safety features. Some of the very bad things that might happen with a conventional water cooled reactor just are not going to happen to a pebble bed reactor, because PBR design does not introduce water into its core. Furthermore, as the core of the PBR heats up, the chain reaction will slow down. Make the core hot enough, and the chain reaction will slow down. The core of a well designed PBR is also meltdown proof. In fact you an heat the PBR, turn its heat up, and then turn off the fans that circulate its coolant gas. In a conventional reactor turning off the coolant pump is a sure formula for disaster. Reactor operators at Chernobyl did that, just before their famous reactor disaster. But PBR researchers have already tried turning the coolant gas circulation fans off, and the reactor shut down and cooled off on its own, without operator intervention. So a PBR has some impressively safety features.

Never-the-less, Rainer Moormann has argued that the PBR is not without safety problems and reports that

the primary circuit of the [experimental] AVR is heavily contaminated with dust-bound metallic fission products (strontium-90, caesium-137), which create major problems in the current dismantling effort.

Moormann claims,

If temperature limits for specific metallic fission products are exceeded, they diffuse through fuel kernels, coatings and graphite over the long term.

Not good! Not good at all. Even worse,

Had AVR core temperatures been known from the beginning of its operation, the AVR hot gas temperatures would have been limited to values far below 950°C. Its main advantage, its apparent capability for process heat generation, would not have been available.

Steve Thomas argues that Moormann was wrong,

the high temperatures experienced at AVR were known about 20 years ago.

Dr Albert Koster, responded to the allegations by Moorman and Thomas,

of hiding facts and supposed safety problems in pebble bed reactor . . . that PBMR knew about safety problems all the time and opted to keep quiet about it.

Koster suggests

Moormann makes a few statements on the closure of both the AVR and THTR that are not supported by the literature and personal recollections of the people involved. In effect, the AVR had come to a natural end of life for a research reactor. . . . If safety problems existed at the AVR, continued operation would not have been allowed, nor would significant experiments be approved by a regulator. There is no known document that cites safety reasons for the shutdown, . . .

Koster quotes Prof. Theenhaus, a member of the FZJ board

In more than 20 years of operation the advantages and positive characteristics of this type of reactor have convincingly been demonstrated. Many experiments have been performed, with particular emphasis on safety research…This demonstration reactor and in a certain sense research reactor completely fulfilled its mission.

Koster also pointed to a statement by Dr. Marnett, technical director of the AVR GmbH,

he AVR-Experimental Power Station has operated for 21 years...and was taken out of operation in 1988 for reasons unrelated to the plant itself.

Koster notes,

the technical problems experienced by the THTR were teething troubles that had been anticipated and budgeted for in the risk-sharing agreement between the utilities and the government. The unplanned increased cost in 1989 was due to the updated estimates of the decommissioning costs and the potential delays caused by having to re-license the fuel plant or have no fuel for more than two years

He also points to a paper by Prof. Knizia, the then chairman of the board of VEW (whose subsidiary HKG operated the THTR) and Dr. Baumer, who at the time was station manager for the THTR:

…It was not technical and especially not safety related technical problems in the plant, but external economical factors that caused risks that were outside the influence of the operator, together with a lack of commitment from the political sides to further support the project that caused the eventual early closure of the project THTR...

Koster complained that Moormann

unfortunately . . . has chosen to include so many unsubstantiated (by references) statements

In other words Moormann offered no proof for his charges. Brian Wang has previously covered the controversy, and is due a word of appreciation.

So basically the Very high temperature reactor prototype that Idaho National Laboratory intends to build during the next ten years, will be based on a previous German design, that has been worked over by South African and Chinese researchers. It might be added that the South Africans recently decided against further development of their PBMR design, and reportedly the Chinese have found that their PBR design has no cost advantage over conventional reactors.

We have to ask if the Department of Energy knows what it is doing, by choosing to concentrate on what amounts to a third hand pass me down German reactor design, that has previously rejected by both Germany and South Africa. Is Pebble Bed Technology the best possible use of the $4 billion the DoE is considering spending on advanced nuclear technology over the next 10 years. Several other potential claimants for that money are waiting in the wings. These include, The Pebble Bed Advanced High Temperature Reactor (PB-AHTR), the IFR and the thorium fuel cycle Molten Salt Reactor, the LFTR. In addition a chloride-salt cooled fast reactor is possible, and has attractive features, when compared to sodium-cooled fast reactors. The Pebble Bed Advanced High Temperature Reactor is a hybrid cross between a gas cooled Pebble Bed Reactor and a Molten Salt Reactor.

The most significant advantage of this hybrid is its small core. The use of gas coolants in PBMRs requires large cores proportional to power outputs. The large cores have a negative impact of PBMR costs. In fact the Chinese are expecting that their PBMRs will be at least as expensive as their LWRs. A molten salt cooled Pebble Bed Reactor would have a much smaller core. This would in turn lower construction costs. A Pebble Bed Advanced High Temperature Reactor would take advantage of both research and development histories of both the Pebble Bed Reactor and the Molten Salt Reactor, allowing the rapid development of a new type of very useful, safe and practical reactors. The hybrid reactor would be inexpensive to build in factories, easily transportable at power outputs that would be useful to industries, could be designed to operate in cogeneration modes, with the further use of residual heat for district heat, and fresh water production through desalinization.

The major down sides of the hybrid would be its limitations as a breeder, and the difficulty reprocessing fuel imbedded in pebble bed pebbles. For Molten Salt Reactor advocates, the hybrid offers a route to full MSR development. The Hybrid is in effect a MSR that does not carry fuel and fission products in its liquid salts. It would be a more than half way step to full MSR development.

The hybrid might be more attractive to regulators than a MSR. It looks more like a conventional reactor, and it relies on pebble bed technology, which most have read about in their textbooks. In addition, the use of liquid salt as a core coolant is something that regulators might feel comfortable with. Liquid salts in the hybrid core would function like water in a LWR core, but without the safety problems of LWRs. It would certainly be a safer coolant than liquid sodium. At he same time the liquid salt coolant is compatible with the very high temperature operations required to produce industrial heat. Again once regulators became familiar with the role of liquid salt in hybrid technology, convincing them of the viability and safety of MSR technology would be easier.

For anyone who is interested in learning more about hybrid reactor technology, ORNL is offering a workshop on Fluoride Salt-Cooled High-Temperature Reactor technology on September 20–21, 2010. A meeting of The GIF MSR Provisional Steering Committee will follow on September 22.

The hybrid technology thus offers a very attractive alternative to the gas cooled pebble bed concept, with a potential of producing up to 1200 C heat directly, and even higher heat indirectly through hydrogen production.

The readers of Nuclear Green are aware of the case I have laid out for the potential value of a thorium breeding Molten Salt Reactor - the LFTR. I have argued the LFTR could potentially supply all human energy demands for a billion years without running out of fuel, that it would be safe, potentially could be built at a low cost, and would largely and perhaps completely solve the problem of nuclear waste.

The Integral Fast Reactor (IFR) is a sodium cooled breeder reactor. Advocates claim that it offers a very high breeding ratio, but my review of IFR plans on the The United States Department of Energy information Website, the Energy Bridge, does not substantuate this claim. Statements by IFR designers indicate that the IFR is designed to be a burner of plutonium and other actinides, that can be operated at a low conversion ratio. Current IFR designs are not intended to be high ratio breeders, and in order to performe this role, the IFR would have to be substantually redesigned. Never-the-less, commercial sodium cooled breeders are expected to come into commercial use during the next 10 years in India.

The IFR has considerable built in passive safety features, but the use of sodium coolant will always be a safety worry. Protection against a sodium related accident will probably add to IFR expenses, and it remains open to question whether the IFR with emerge with the sort of cost advantages, which currently seem plausible for the LFTR. In addition the IFR would not operate at a high enough temperature to provide industrial heat for many processes. This is not to say that the IFR reactor is a bad reactor but to say that it potential may be limited, and perhaps far more limited than the LFTR.

Reading the Obama Administration Tea Leaves

The Obama Administration might be accused of sending mixed signals on its attitude toward the future of nuclear power. First we have first the January 28, 2010 Obama Memo to Energy Secretary Steven Chu:

MEMORANDUM FOR THE SECRETARY OF ENERGY

SUBJECT: Blue Ribbon Commission on America’s Nuclear Future

Expanding our Nation’s capacity to generate clean nuclear energy is crucial to our ability to combat climate change, enhance energy security, and increase economic prosperity. My Administration is undertaking substantial steps to expand the safe, secure, and responsible use of nuclear energy. These efforts are critical to accomplishing many of my Administration’s most significant goals.

An important part of a sound, comprehensive, and long-term domestic nuclear energy strategy is a well-considered policy for managing used nuclear fuel and other aspects of the back end of the nuclear fuel cycle. Yet the Nation’s approach, developed more than 20 years ago, to managing materials derived from nuclear activities, including nuclear fuel and nuclear waste, has not proven effective. Fortunately, over the past two decades scientists and engineers in our country and abroad have learned a great deal about effective strategies for managing nuclear material. My Administration is committed to using this advanced knowledge to meet the Government’s obligation to dispose of our Nation’s used nuclear material.

Accordingly, I request that you establish a Blue Ribbon Commission on America’s Nuclear Future (Commission) and appoint its members. Those members should include recognized representatives and experts from a range of disciplines and with a range of perspectives, and may include participation of appropriate Federal officials. The Commission’s business should be conducted in an open and transparent manner.

The Commission should conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle, including all alternatives for the storage, processing, and disposal of civilian and defense used nuclear fuel and nuclear waste. This review should include an evaluation of advanced fuel cycle technologies that would optimize energy recovery, resource utilization, and the minimization of materials derived from nuclear activities in a manner consistent with U.S. nonproliferation goals.

In performing its functions, the Commission should consider a broad range of technological and policy alternatives, and should analyze the scientific, environmental, budgetary, economic, financial, and management issues, among others, surrounding each alternative it considers. Where appropriate, the Commission may also identify potential statutory changes.

The Commission should provide an interim report to you within 18 months of the date of this memorandum, and that report should be made available for public comment. The Commission should provide a final report to you within 24 months of the date of this memorandum. The Department of Energy shall provide funding and administrative support for the Commission, as you determine appropriate, so that it can complete its functions within these time periods. Additionally, all executive departments and agencies shall provide such information and assistance to the Commission as you or the Commission may request for purposes of carrying out the Commission’s functions, to the extent permitted by law. Nothing in this memorandum shall be construed to require the disclosure of classified, proprietary, law enforcement sensitive, or other information protected under governing law. This memorandum shall be implemented consistent with applicable law and subject to the availability of appropriations. This memorandum is not intended to, and does not, create any right or benefit, substantive or procedural, enforceable at law or in equity by any party against the United States, its departments, agencies, or entities, its officers, employees, or agents, or any other person.

You are hereby authorized and directed to publish this memorandum in the Federal Register.

BARACK OBAMA

The list of actual Blue Ribbon Commission appointees appears to be more calculated for political effect than for knowledge and wisdom. Although the Commission is charged with analyze the scientific, budgetary, economic, (and) financial issues involved in solutions to the used nuclear material problem. We have no economists on board, although Geologists Allison Macfarlane, believes herself to be an expert on nuclear costs. Macfarlane, however, appears to view the thorium fuel cycle in a considerably more positive light than she views the U-238 - reactor grade plutonium fuel cycle. Susan Eisenhower serves on the corporate advisory boards of Lightbridge (formerly Thorium Power), that might give her a significant knowledge of the thorium fuel cycle. The weight of the Blue Ribbon Commission lies heavily on the political/bureaucratic/expert of reference axis. The presence of two former congressmen on the commission suggests the Obama administration's desire to manage and even spin its eventual report toward politically acceptable conclusions. The last thing the Obama Administration wants is for a brilliant and charismatic scientist, like Richard Feynman to steal the show, by offering a dramatic demonstration of a politically embarrassing problem.

Thus only one real working scientist is included on the panel, that is Per Peterson. Peterson is well known to the Energy from Thorium community, and is an unabashed supporter of the use of Molten Salt nuclear technology. In addition to his expertise on Generation IV reactor design, and lowering nuclear costs, Per is also a nuclear proliferation expert, with a working knowledge of current thinking about proliferation prevention.

We know that the Yucca Mountain approach to the nuclear waste issue is off the table, and it appears quite likely that the IFR is as well. On January 15, the Defense Daily carried a story by George Lobsenz titled White House Moves To Restrict DoE Nuclear Research. That story stated:

The White House has proposed barring Energy Department research on fast reactor recycling of nuclear waste and technical support for licensing of small, modular light-water reactors, drawing protests from Energy Secretary Steven Chu that such prohibitions will have broad adverse effects, including hurting the U.S. nuclear industry's renaissance; crimping U.S. ability to influence other countries' fast reactor designs to address proliferation concerns; and taking away nuclear waste disposal options that might be considered by the administration's planned blue-ribbon panel on alternatives to the Yucca Mountain repository.

The story went on to discuss what appeared to be a conflict between the Obama White House and Energy Secretary Chu over the exclusion of fast reactor research from the DoE Research program. This is very bad news for the IFR supporters, and might explain some of their recent behavior. Thus there are signs that the Blue Ribbon Commission will seriously consider Molten Salt nuclear technology, and the thorium fuel cycle as potential remedies for the fuel cycle nuclear waste issue. Above all else the panel is clearly expected to be a reliable, unimaginative, and boring extension of the Obama ego, that will reach predictable and unimaginative conclusions.

Fire in Sodium cooled reactors: A Sandia Literature Review

"Metal Fire Implications for Advanced Reactors, Part 1: Literature Review," is an important report for anyone who is interested in the IFR or other fast sodium cooled reactors. The issue of fire in Sodium cooled reactors is an important one, which desirves serious attention. IFR advocates argue that the IFR is highly safe. My own review of the ABTR answered many of my questions about IFR safety, but I am not a nuclear safety expert, and my findings should not be the last word on IFR safety. The good thing about the current Sandia report is that it comes from Sandia rather than Argonne, and therefore the writers cannot be accused of IFR cheerleading. The report is well written, and it is quite approachable by none scientists, who are looking for more information about the problem of IFR/LMFBR safety. The report demonstrates that progress has been made on LMFBR safety, but does not support claims that further LMFBR safety research is unneeded. I do not intend to post all of the post, but rather to call the readers attention to some passages, in the hope that the readers interests will be ignited.

Metal Fire Implications for Advanced Reactors, Part 1: Literature Review

By Tara J. Olivier, Ross F. Radel, Steven P. Nowlen, Thomas K. Blanchat, & John C. Hewson

Abstract
Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

1. INTRODUCTION
The anticipated nuclear power renaissance hinges on public acceptance and a demonstrated treatment of potential safety issues, particularly for advanced reactor designs. The Advanced Burner Reactor (ABR) uses a liquid sodium primary coolant as do certain other advanced reactor concepts. In contrast to today’s Light Water Reactors (LWRs), liquid-metal- cooled reactors present a unique risk; namely, potential metal fires involving the sodium coolant.
Fire is a significant contributor to total nuclear plant risk for current generation LWRs. Given “passively safe” advanced designs, some elements of plant risk will diminish substantially. Fires could represent the dominant risk contributor, especially given the unique characteristics of metal fires such as very high temperatures and fire suppression challenges. Fast breeder reactors all over the world use liquid sodium as a coolant and there has been experimental and analytical research done related to sodium fires as early as the 1950’s. The research has included fundamental studies, work on droplet combustion, pool burning, suppression, and large-scale sodium fire experiments. However, there are gaps in our understanding of the basic combustion behavior and combustion mechanics due to the complexities involved. These gaps have led to little progress in understanding the basic combustion behaviors for sodium. (Makino 2006). Many of these same concerns were noted as far back as 1972 (Newman 1972).

New technologies have substantially improved fire computer modeling capabilities, but to apply these tools to a sodium fire will require some additional model development and validation work. Unfortunately, most of the experiments performed in the past cannot be used to support model development today. Clear definition of the experimental boundary and initial conditions are necessary to create the modeled conditions, and most of the experimental results lack this information. “Reports of precise conditions in experiments are rare in the literature,” so the heat transfer evaluations have almost been impossible (Makino 2006).

This report includes four elements. First, a comprehensive review will define the current state of knowledge for metal fires. This will include actual metals fire experience in various applications. Second, an assessment of advanced reactor concept designs and identification of the unique metal fire safety and hazards was completed. A number of potential safety scenarios exist and will be grouped as to potential importance and representative physics to prioritize the specific research directions that will maximize breadth of applicability to emerging reactor designs. Third, a detailed review of sodium combustion research and potential approaches to the design and conduct of future experiments will be presented. Fourth, Appendix A presents an annotated bibliography of relevant literature identified during extensive literature review.

2. PREVIOUSLY RECORDED SODIUM FIRE ACCIDENTS
This chapter describes past sodium fires at nuclear reactors and other sodium facilities. The incidents discussed in this chapter were chosen to highlight the most significant issues surrounding sodium fires. These issues include design defects at startup (Monju), pipe bursts (BN-600), sodium spray fires (Almeria), and sodium-concrete interactions (ILONA)1.

2.1 Monju Prototype Fast Breeder Reactor

The Monju Prototype Fast Breeder Reactor (FBR) first reached criticality in 1994. Powered operation began in 1995, and a series of power raising tests were performed, with a planned full-power test planned for June 1996. Monju is a loop-type 280 MWe sodium-cooled reactor with mixed oxide fuel (Mikami 1996). During normal operation, the inlet and outlet sodium temperatures in the primary coolant loop are 397 °C and 529 °C, respectively. Sodium temperatures in the secondary coolant loop range between 325-505 °C.

During a scheduled power rating test (40% electrical power) on December 8, 1995, a high sodium temperature alarm sounded at the outlet of the secondary side of the intermediate heat exchanger (IHX) (Mikami 1996). At the same time, smoke detectors sounded in the same area, closely followed by a sodium leak detection alarm. Operators began normal plant shut-down procedures, but after increased smoke was observed 50 minutes later, it was decided to manually trip the reactor. This shutdown occurred approximately 1.5 hours after the initial alarms sounded.

Investigations later confirmed that a sodium leak and fire had occurred, ultimately, the source of the leak was traced to a damaged temperature sensor (pictured in Figure 1). The sensor consists of thermocouple wires housed in a protective well tube. It was found that the tip of the well tube had broken off and the thermocouple was bent at an angle of 45 degrees toward the downstream flow direction.

A microscopic inspection of the flow tube was performed to determine the root cause of the leak. It was concluded that the breakage of the well tube was caused by high cycle fatigue due to flow induced vibration in the direction of sodium flow. It was found that the problems were rooted in the design of the well tube. Although designers applied ASME standards to prevent resonant vibrations, they failed to take into account the sharp taper of the Monju tube design. As a result, the vortex-induced vibration could not be prevented. The design has subsequently been re-evaluated.

In addition to replacing all similarly designed temperature sensors, aspects of sodium fire response and emergency operation procedures were also modified at the Monju site. For example, the reactor will be shut down immediately if a sodium leak is confirmed in the future. A summary of the Monju Improvement Plan is shown in Table 1 (Mikami 1996). As this event was confined to the secondary coolant loop, there was no radiological release that affected either the general public or the plant personnel. However, it has resulted in over a decade of safety reviews in order to re-establish both technical surety and public confidence in the plant. The Monju plant is scheduled to resume operation in mid-2008.

The brief discussion of the ABTR is an excellent indicator of the actual developmental status of the IFR. Once again we see clearly that the IFR is not ready for commercial implementation.

3.1.4 Advanced Breeder Test Reactor

The ABTR is a sodium-cooled, pool-type reactor based on experience gained from the Experimental Breeder Reactor-II (Chang 2006). It is a 95 MWe design with an estimated 38 percent plant efficiency. The ABTR was developed as a test bed for a similar commercial design-the Advanced Breeder Reactor. The ABTR design uses a 20 percent TRU, 80 percent uranium metal fuel clad with HT-9 stainless steel. A summary of plant specifications is provided in Table 8.

There are numerous objectives of the ABTR design, including demonstration of reactor- based transmutation of trans-uranics as part of an advanced fuel cycle, qualification of the trans-uranic-containing fuels and advanced structural materials needed for a full-scale ABR, and supporting the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. ABTR designers also have the following objectives:

• To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in advanced breeder reactors;
• To demonstrate improved technologies for safeguards and security;

• To support development of the U.S. infrastructure for design, fabrication and construction,
testing and deployment of systems, structures and components for the ABR.

3.3 Sodium Fire Consequences

3.3.2 Core Voiding

A fundamental difference between water and sodium-cooled reactors is the void reactivity coefficient. If the water around the core is voided (boiled, drained) in a water-cooled (thermal) reactor during operation, the power level will automatically drop. The reactor is therefore said to have a negative void reactivity coefficient. In contrast, if sodium is voided in certain sodium-cooled fast reactors (particularly large reactors), it will cause the power level of the reactor to rapidly increase. This reactor is said to have a positive void reactivity coefficient. When the reactor power increases, it can lead to additional boiling and voiding until fuel melts. This positive feedback can lead to extremely rapid surges in reactor power, potentially damaging or melting fuel and cladding.

Multiple events can lead to core voiding during operation, and great care is taken in the proposed new reactors to ensure that these events are prevented. They include sodium boiling, loss of coolant accidents (LOCA), and gas bubble entrainment within the sodium. Sodium fires could lead to sodium boiling if an undercooling event is initiated without scram (reactor shutdown). A severe leak in the secondary system, perhaps coupled with cable fires could lead to this situation. A large leak in the primary system could also disrupt flow enough to induce sodium boiling in the core. A sodium leak in the primary system could also lead to either a LOCA or gas bubble entrainment event. A large primary leak could potentially uncover a portion of the core. If gas is pulled back into a leak in the primary system, the resulting bubbles could also reach the core.

3.3.3 Loss of Heat Sink

A loss of heat sink event can be triggered by sodium leaks in the steam generators. As stated above, the standard procedure in response to these leaks is to drain one or both sides of the steam generator. In the event that multiple steam generators are compromised, reactor cooling must be accomplished with backup safety systems. In the case of the new generation of reactors, these safety systems are generally passive in nature (i.e. they require no operator intervention). These systems ultimately rely on natural circulation driven by core decay heat, and so are also independent of cable fires or loss of site power. In addition to these engineered safety features, the inherent high heat capacity of the sodium and structural elements of the reactor will provide valuable time for operators to restore the system to normal.

3.3.4 Loss of Engineered Safety Systems

The inherent mobility of a fire can cause a fire to become a threat to an entire reactor system. Numerous examples exist of cable fires causing serious problems in a nuclear power plant. Perhaps the most famous of these is the 1975 Browns Ferry fire, where all of the normal core-cooling functions were lost due to a cable fire (Nowlen 2001). However, operators were able to maintain core cooling with a control rod drive pump not included in plant procedures. The fire at Greifswald burned for about 92 minutes causing a station blackout and the loss of all active means of cooling the core (Nowlen 2001). As a result, a pressurizer relief valve opened and failed to close. This situation persisted for at least five hours and led to depletion of the secondary and primary side coolant inventories. The plant was ultimately recovered through initiation of low pressure pumps, the recovery of off-site power, and the recovery of one auxiliary feedwater pump.

These and other incidents demonstrate the need for next-generation sodium-cooled reactors to consider the potential impact of fire on safety systems to maintain core cooling, including the passive safety systems. Every adverse situation cannot be anticipated or avoided. However, if the reactor safety systems operate independent of the plant operators and electrical systems, then these systems can likely maintain cooling until plant personnel put out fires and regain control of the situation.

There is one additional factor that is unique to metal fires that may need to be addressed. Conventional (i.e., non-metal) fires are not generally considered a threat to primary plant piping components used in a light-water reactor (Nowlen, Najafi, et al. 2005). This would include the primary piping itself and other piping equipment such as large valves, check valves, and water- filled vessels (e.g., storage tanks). However, sodium fires burn at much higher temperatures than do other types of fires. Hence, metal fires could represent a threat to components and equipment not normally considered fire-vulnerable. For metal-cooled reactors, the performance of plant safety systems and equipment under fire conditions, including the passive safety systems, should be evaluated in this context.

3.4 Summary

Of all the new reactor designs proposed in the Generation IV program, sodium fast reactors have the largest experience base. Thanks in large part to this experience, numerous engineered safety features and safety procedures have been built into the next generation of sodium-cooled reactors. These features are designed to reduce the likelihood and consequence of sodium release and fire.

The risk of sodium release and fire exist during the three stages of a reactor lifetime; startup, day-to-day operation, and refueling and maintenance. Based on past experience, design and manufacturing defects have generated the greatest risk of sodium leakage and fire at reactor startup. Pipes, welds, and steam generator tubes are the most likely components to fail during routine operation. Thermal and mechanical fatigue must be avoided to minimize the chance of these failures. Refueling and maintenance accidents are generally caused by a combination of improper procedures and human error. The experience gained in existing reactors should help to minimize the chance of these leaks.

Sodium fires at any facility can cause serious problems beyond the immediate burn area. However, a sodium fire at a nuclear reactor can have consequences beyond those possible at non- nuclear facilities. The most notable consequences of sodium fire at a nuclear power plant include smoke in the control room, core voiding, reactor under-cooling, loss of heat sink, and loss of engineered safety systems. Sodium fires burn much hotter than other types of fires and might therefore threaten plant equipment, such as piping elements that are not normally considered vulnerable to fire damage. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation of sodium-cooled reactors. However, some unique considerations do come into play with sodium fires.

The report conclusion will serve to demonstrate that IFR safety research still has aways to go.

5. CONCLUDING REMARKS

This report documents the results of the initial stage of the “Metal Fire Implications for Advanced Reactors” Laboratory Directed Research and Development project. Efforts to date have included an extensive literature search to cover the sodium fire recorded accidents, the proposed LMFBR designs and safety concerns and sodium fire combustion experiments and research.

Past experiences/accidents with sodium fires at nuclear and non-nuclear sodium facilities were investigated to identify the types of hazards that must be accounted for when designing the next generation of sodium-cooled nuclear reactors. The risk of sodium release and fire exists primarily during the three stages of a reactor lifetime; startup, day-to-day operation, and refueling and maintenance. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation of sodium-cooled reactors.

A need also exists to improve the state-of-the-art fire modeling codes to include the sodium fire combustion phenomenon. The past experiments did not record the details of the boundary conditions for both pool and spray fire scenarios. A lot of the experiments were small scale compared to the amount of sodium that could be involved in a HCDA. There exists a need to understand the phenomenon of inter-droplet interactions in a spray fire scenario. There has not been any experimental work to address this. Fire is one of the key parameters in a NPP risk analysis. With the GNEP program making progress forward, expertise in metal fires is essential for Sandia National Laboratories.

Argonne Liquid-Metal Advanced Burner Reactor

"Argonne Liquid-Metal Advanced Burner Reactor" is another in a series of Nuclear Green postings that will focus on the developmental status of IFR technology. A secondary function of these posting is to encourage nuclear literacy. I wish people who are not afraid to investigate technology to become more familiar with such documents and to read them and judge them for themselves. The path to enlightenment requires knowledge. A further purpose is to challenge misinformation about the IFR that is currently being spread. It has been widely stated that the IFR technology is more mature than LFTR technology, and that the IFR technology is currently available for commercial development. "Argonne Liquid-Metal Advanced Burner Reactor" demonstrates that Argonne considers IFR technology to be in a prototype stage. Second some IFR advocates have argued that IFR reactors will as soon as a prototype is built, have the capacity to breed significant amounts of fuel. In fact Argonne prototype proposals suggest prototypes which are plutonium burners, not breeders, although the core of he ABTR could be modified to breed with a low positive breeding ratio. Thirdly i would like to call attention to the function of the Argonne Liquid-Metal Advanced Burner Reactor, which is to serve as a research tool for iFR" development. The reader is advised to focus of the questions which the LMABR is expected to answer, before a commercial IFR prototype can be built. I have included both the Abstract and the Introduction to this document in this post, together with a link to the full text.

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

Authors: Kasza, K. Grandy, C. Chang, Y. Khalil, H. Argonne National Laboratory Nuclear Engineering Division

Abstract: This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design application. Capabilities required to establish and operate this center are found primarily in Argonne's Nuclear Engineering and Mathematics and Computer Science Divisions. Funding for the center would be sought from DOE-NE (GNEP/Advanced Burner Reactor and Generation IV programs), DOE-SC/ASCR, and the commercial nuclear industry. Having the above experimental and modeling capabilities at Argonne would constitute a national/international center of excellence for conducting the research and engineering and design tool development needed to support the DOE GNEP/ LM-ABR initiative in developing safe, high-performance reactors.

1.0 Introduction

During the 1970s and 1980s, the U.S. DOE sponsored a substantial effort in the development of sodium-cooled fast nuclear reactors. Initially, these fission reactors were to be breeders with the designation Liquid Metal Fast Breeder Reactor (LMFBR). Later the breeding stipulation was dropped, and the name was changed to Liquid Metal Reactor (LMR) or Advanced Burner Reactor (ABR). The most important feature of the earlier breeder reactor was to significantly extend the useful life of the world’s supply of fissionable uranium by implementing a system of breeder reactors with the goal of producing about 10% more fissionable material each year than consumed in producing electricity. At that time, LMFBRs had only been built in small size, and it was appropriate for DOE to sponsor the development of the commercial-scale technology because the project was too large for private industry and because it was of great national interest and potential benefit. To this end, DOE sponsored a variety of research and development programs to advance this technology.

In response to DOE during the 1970s, Argonne conducted thermal–hydraulic studies of individual LMR components supporting EBR-II, FFTF, and CRBR development. After cancellation of CRBR, DOE in order to begin transferring LMR technology developed under federally funded programs to the U.S. industry funded a design competition between General Electric (GE) and Rockwell International/Combustion Engineering (RI/CE) to design a commercially viable LMR for future deployment. The GE design was called PRISM (Prototype Reactor Inherently Safe Module) and the RI/CE design was called SAFR (Sodium Advanced Fast Reactor). In the mid-1980s and early 1990s, Argonne conducted forced- and natural- convection phenomena studies on complete in-vessel system experimental test models of the GE/PRISM and RI/CE/SAFR designs. These DOE-funded studies were carried out in collaboration with GE and RI/CE. Further development of LMR expertise/design knowledge is vital to the future deployment of safe, reliable, and high-performance LMR Advanced Burner Reactors (ABRs) currently being proposed by DOE under the GNEP initiative for deployment in 2025. This near-term deployment does not involve breeder reactors.

This white paper has been written to summarize the thermal-hydraulic understanding that has been developed over the last 30 plus years, highlight important phenomena that must be factored into future reactor designs, and describe additional developmental efforts still needed. In particular, it describes the need for further LM-ABR technology development support in the form of better testing infrastructure, improved engineering knowledge, and improved/validated computational modeling tools. The paper also addresses the impact of thermal hydraulics on reactor system operation and on the behavior/performance of individual components (thermal duty and structural impact) during normal operation and postulated off-normal low-flow accident events related to safety.

Argonne has been a pioneer in the study of thermal-buoyancy-force governed flows under various important reactor transient conditions, such as the transition from forced to natural convection, instabilities generated by parallel flow paths, and structural thermal stresses caused by thermal stratification and their influence on heat-sink effectiveness.

8Argonne from the early 1970s through the early 1990s was DOE’s lead laboratory for LMR thermal-hydraulic development. During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting EBR- II, FFTF, and CRBR. In the 1980s and into the early 1990s, Argonne conducted studies on forced- and natural-convection (thermal-buoyancy-force) phenomena in complete in-vessel models of GE/PRISM and RI/SAFR. These two reactor initiatives involved Argonne working closely with U.S. industry and the DOE. This paper describes the very important impact of thermal hydraulics on reactor global operation and on individual component behavior/performance (thermal duty, structural impact, and safe operation) during normal operation and postulated off-normal low-flow accident events related to safety. Argonne’s LMR expertise and design knowledge are vital to the further development of a safe, reliable, and high-performance LM-ABR.

In the 1980s Argonne developed/built a large water test facility called the Mixing Components Test Facility (MCTF) for performing steady and thermal-transient experimental simulations of important reactor components under a wide range of operation scenarios. (The MCTF was decommissioned in 1993.) Modeling studies were also conducted by Argonne relative to ascertaining if the thermal-buoyancy phenomena being studied could be effectively addressed through the use of water for testing of both individual LMR components and complete in-vessel system geometries. This modeling is discussed in detail in Section 4.2.1 and in Appendix 1 of this report. These modeling studies also highlighted where water testing was not adequate for addressing certain phenomena.

All of the Argonne studies involved fundamental experimental thermal-hydraulic testing and a strongly integrated component of computational fluid dynamics (CFD) code development and simulation analysis. The CFD analyses were predominately performed with the Argonne COMMIX code, which was augmented by some initial effort at utilization of commercial CFD codes like STAR-CD. One of the first uses of three-dimensional CFD analysis for addressing LMR thermal-hydraulics, which used the COMMIX code, was to address buoyancy-governed reactor flows. This computational modeling was driven and guided by Argonne’s thermal- hydraulic experiments on reactor components such as piping, plenums, steam generators, and heat exchangers.

The contents of this white paper are as follows:

Section 2 briefly describes the new DOE GNEP initiative relative to the pre-conceptual design features of the Argonne proposed LM-ABR, which would be one of the GNEP building blocks. Knowing the general technical features associated with the ABR allowed us to focus on and exploit what has been learned over the last 30 plus years about LMR thermal hydraulics relative to the importance to GNEP. Finally, this section also describes, based on the 1993 GE/PRISM close-out report, what industry ideas were at that time as to what further development and testing were needed to deploy a U.S. LMR. This information helped to further focus the recommendations given in this report regarding a thermal-hydraulic pathway forward.

Section 3 describes Argonne studies and the status of our understanding of thermal buoyancy phenomena occurring in individual reactor components such as:

␣ Piping

␣ Piping/plenum interfaces and thermal plumes

␣ Heat exchangers

␣ Steam generators

␣ Multiple coolant stream thermal mixers

Section 4 describes Argonne’s past studies on reactor in-vessel thermal hydraulics, which initially addressed generic core outlet and plenum flow interactions guided by CRBR needs. These studies in their later stages investigated forced- and natural-convection (thermal- buoyancy-force) phenomena using complete in-vessel models of the GE/PRISM and RI- CE/SAFR designs. Complete in-vessel model experiments were used because the flow and thermal behavior in a given sub-region of the reactor vessel is the result of complex interactions with the rest of the reactor in-vessel components. These interactions are especially important for pool designs under low-flow conditions and the transition to natural convection. They have the potential for strongly affecting reactor:

␣ Thermal-hydraulic performance

␣ Emergency cooling

␣ Structural integrity

␣ Heat-sink effectiveness

These complete in-vessel experimental studies provided GE and RI-CE designers with information vital to the design and assessment of the workability of the various features that were being incorporated into their innovative and inherently safe reactors.

Finally, Section 5 describes a pathway forward regarding further research and development needed to support the GNEP/LM-ABR initiative.