Monday, February 23, 2015

Metal Fires in Fast Reactors: Part I

The origin of this post was from an extensive quotation the first part of of a Sandia NL report on liquid sodium cooled fast reactors.  The question I am addressing is a simple one, how safe is the IFR. Although the backers of the IFR tout its safety, they also advocate for more government investment in IFR r&D.  Although they do not admit it, among the issues requiring futher R&D, as this report clearly shows is safety.

"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.

Saturday, February 14, 2015

Per Petersin's reactor

Per Peterson is not included among the potential builders of Molten Salt technology reactors.  Yet as this video reveals, Peterson is engaged in Research for a Molten Salt cooled Pebble bed reactor.  This concept is not as attractive to chemists, as the conventional MSR. but it does offer significant protection from proliferation risks entailed by the LFTR and Uranium Fueled MSRs.  Thus the PB-MSR offers low cost, very efficient and highly safe energy with low proliferation risk. This form of reactor would lessened the risk involved in increasing the energy input into politically weak societies.  Nations which already posses nuclear weapons, are unlikely to use LFTRs or MSRs to add to their already huge nuclear stockpiles  While many economically advanced, politically stable nations would have little to gain and much to loose by violating anti-proliferation treaties, and thus could be trusted with MSR/LFTR technology.  Here is hoping them that Per or some of his associates launches a project to build PBMSRs.

 

Sunday, February 8, 2015

Alvin Weinberg on the future of Nuclear power

After Alvin Weinberg's death in October 2006, I began a study of his writings that were available on the internet. This was to lead me to several brief papers on the Molten Salt Reactor, and to the Molten Salt Breeder (the LFTR). My interest in Weinberg's Molten Salt Reactor vision, was facilitated by the emergence of Kirk Sorensen's blog Energy from Thorium. Kirk had started his blog about six months before my rediscovery of Weinberg's vision. Kirk had done an amazing job of documenting the history of Molten Salt Reactor Research in Oak Ridge. Kirk's Document archive is perhaps the single most powerful rhetorical tool on nuclear energy available on the Internet. Most people who bother to work their way through Kirk's Documents end up being MSR believers.

In addition to Kirk's Documents a number of important Weinberg papers were posted on the Office of Science and Technology Information web page. One of the most important of those papers was “Towards an Acceptable Nuclear Future" in which Weinberg had
re-examine man's long-term energy options, in particular solar energy and the breeder reactor.
In Acceptable Nuclear Future Weinberg had discussed the cause of the collapse of the first Nuclear Era:
At that time, with Oyster Creek being contracted at a little over $100 per kilowatt of electricity [kW(e)], it seemed plausible to expect nuclear energy to be extremely cheap, as well as inexhaustible. Our dreams of nuclear-powered agro-industrial complexes seemed like legitimate extrapolations from what we thought was demonstrated technology. That the technology turned out to be much more expensive for reasons that few could foresee, or that other sources of energy have also become very expensive, is beside the point: disillusionment with our predictions made it difficult for the nuclear community to retain the confidence of some of the public.
But then Weinberg pointed to a potential low cost nuclear future,
I am unprepared to give up: reactors with an intrinsically low fuel-cycle cost, such as CANDU-Th or molten salt, may yet be realized.
During his OrNL days Weinberg wrote a number of short papers intended to introduce Molten Salt Reactor technology. One was titled WHY DEVELOP MOLTEN-SALT BREEDERS? it was a an introduction too ORNL-TM-1851 (posted in Kirk's document archive). "Why develop" summarized Weinberg's thinking about the Molten Salt reactor, and was offered Weinberg's case to make a persuasive case for the Molten Salt Breeder Reactor:
Nuclear power, based on light-water-moderated converter reactors, seems to be an assured commercial success. This circumstance has placed upon the Atomic Energy Commission the burden of forestalling any serious rise in the cost of nuclear power once our country has been fully committed to this source of energy. It is for this reason that the development of an economical breeder, at one time viewed as a long-range goal, has emerged as the central task of the atomic energy enterprise. Moreover, as our country commits itself more and more heavily to nuclear power, the stake in developing the breeder rises—breeder development simply must not fail. All plausible paths to a successful breeder must therefore be examined carefully.

To be successful a breeder must meet three requirements. First, the breeder must be technically feasible. Second, the cost of power from the breeder must be low; and third, the breeder should utilize fuel so efficiently that a full-fledged-energy economy based on the breeder could be established without using high-cost ores. The molten-salt breeder appears to meet these criteria as well as, and in some respects better than, any other reactor system. Moreover, since the technology of molten-salt breeders hardly overlaps the technology of the solid-fueled fast reactor, its development provides the world with an alternate path to long-term cheap nuclear energy that is not affected by any obstacles that may crop up in the development of the fast breeder.

The molten-salt breeder, though seeming to be a by-way in reactor development, in fact represents the culmination of more than 17 years of research and development. The incentive to develop a reactor based on fluid fuels has been strong ever since the early days of the Metallurgical Laboratory. In 1958 the most prominent fluid-fuel projects were the liquid bismuth reactor, the aqueous homogeneous reactor, and the molten-salt reactor. In 1959 the AEC assembled a task force to evaluate the three concepts. The principal conclusion of their report was that the “molten-salt reactor has the highest probability of achieving technical feasibility.”

This verdict of the 1959 task force appears to be confirmed by the operation of the Molten-Salt Reactor Experiment. To those who have followed the molten-salt project closely, this success is hardly surprising. The essential technical feasibility of the molten-salt system is based on certain thermodynamic realities first pointed out by the late R.C. Briant, who directed the ANP project at ORNL. Briant pointed out that molten fluorides are thermodynamically stable against reduction by nickel-based structural materials; that, being ionic, they should suffer no radiation damage in the liquid state; and that, having low vapor pressure and being relatively inert in contact with air, reactors based on them should be safe. The experience at ORNL with molten salts during the intervening years has confirmed Briant’s chemical intuition. Though some technical uncertainties remain, particularly those connected with the graphite moderator, the path to a successful molten-salt breeder appears to be well defined.

We estimate that a 1000 MWe molten-salt breeder should cost $115 per kilowatt (electric) and that the fuel cycle cost ought to be in the range of 0.3 to 0.4 mill/kWh. The overall cost of power from a privately owned, 1000-MWe Molten-Salt Breeder Reactor should come to around 2.6 mills/kWh. In contrast to the fast-breeder, the extremely low cost of the MSBR fuel cycle hardly depends upon sale of byproduct fissile material. Rather, it depends upon certain advances in the chemical processing of molten fluoride salts that have been demonstrated either in pilot plants or laboratories: fluoride volatility to recover uranium, vacuum distillation to rid the salt of fission products, and for highest performance, but with somewhat less assurance, removal of protactinium by liquid-liquid extraction or absorption.

The molten-salt breeder, operating in the thermal Th-233U cycle, is characterized by a low breeding ratio: the maximum breeding ratio consistent with low fuel-cycle costs is estimated to be about 1.07. This low breeding ratio is compensated by the low specific inventory* of the MSBR. Whereas the specific inventory of the fast reactor ranges between 2.5 to 5 kg/MWe the specific inventory of the molten-salt breeder ranges between 0.4 to 1.0 kg/MWe. The estimated fuel doubling time for the MSBR therefore falls in the range of 8 to 50 years. This is comparable to estimates of doubling times of 7 to 30 years given in fast-breeder reactor design studies.

From the point of view of long-term conservation of resources, low specific inventory in itself confers an advantage upon the thermal breeder. If the amount of nuclear power grows linearly, the doubling time and the specific inventory enter symmetrically in determining the maximum amount of raw material that must be mined in order to inventory the whole nuclear system. Thus, low specific inventory is an essential criterion of merit for a breeder, and the detailed comparisons in the next section show that a good thermal breeder with low specific inventory could, in spite of its low breeding gain, make better use of our nuclear resources than a good fast breeder with high specific inventory and high breeding gain.

The molten salt approach to a breeder promises to satisfy the three criteria of technical feasibility, very low power cost, and good fuel utilization. Its development as a uniquely promising competitor to the fast breeder is, we believe, in the national interest.

It is our purpose in the remainder of this report to outline the current status of the technology, and to estimate what is required to develop and demonstrate the technology for a full-scale thermal breeder based on molten fluorides.
Papers like this sent me to the Internet, where i found Bruce Hoglund's Molten Salt Interest Pages.Bruce was not actively adding to his Pages when i discovered them, so by far the most significant web site on Molten Salt Reactor Technology was Energy from Thorium. My interest in EfT eventually lead to my encounter with Kirk Sorensen.

Weinberg's interest in Molten Salt Reactor technology extended well beyond writing a few brief papers. He was passionately involved in the Progress of the MSR project.

In a brief 1987 talk, R. G. Wymer of the ORNL Chemical Technology Division, described Weinberg's passion for fluid fueled reactors.
At Oak Ridge we were concentrating on the thorium fuel cycle. Alvin Weinberg, for many years the Director of the Oak Ridge National Laboratory, is a great phrase maker. His dream was to "burn the rocks", in "a pot, a pipe, and a pump". By those rather fanciful phrases he meant that the thorium present in many granitic rocks throughout the world could, in a sense, be "burned" by incorporating it in a fluid fuel reactor. The fluid was to be pumped around and around in a loop. The loop was the pipe. As the liquid fuel was pumped through an enlargement in the pipe (the enlargement was the pot), there was enough of it present in the right configuration to go critical. The heat of fission was to be taken off by heat exchangers in the loop and converted to electrical energy. Also, as the fuel circulated around the loop, it was to be reprocessed continuously by equipment located in a side stream through which a fraction of the fuel was diverted. In principle that was a pretty good idea. One of the major reasons for the poor capacity factors of present-day power reactors is that it takes a long time to refuel them. Weinberg's concept avoided that refueling problem. Of course it created a few problems too.

Two major programs were carried out at ORNL based on the pot, pipe, and pump concept. One of them was the Aqueous Homogeneous Reactor, and the other was the Molten Salt Reactor Experiment. Needless to say, these projects were of major importance to Dr. Weinberg. Occasionally he would engage me in conversation. Invariably the conversation would turn to the current ORNL pot, pipe, and pump reactor project. This was perfectly reasonable from his point of view because my division, Chem Tech, had a major role in both of those reactor projects.
The interesting thing about Weinberg's encounters with Wymer was that Wymer was never directly involved in the development of fluid core reactors. Wymer knew what was going on in the Chemical Technology Division and Weinberg was pumping him for information that he was not being given by other division leaders. And Heaven help the division leader who Weinberg was not satisfied with after Weinberg pumped members of their staff for information.

Kirk had posted on EfT a number of short Weinberg papers on Molten Salt Reactors and thorium breeding . Thus my interest in Weinberg lead in turn to an interest in the Molten Salt Reactor, that became foundational to my views on the future of energy. That is, however, another story.

Sunday, January 25, 2015

The electrification of Earth and Thorium

I have been looking at the ThorCon business plan. It is hugely ambitious. Their goalo appears to be nothing less than the electrification of the planet with molten salt technology by 2050 or so. This hugely ambitious goal would not be possible to reach without building thorium cycle breeders or one to one converters. The benefits would be enormous, but as this discussion has identified, there are risks. There are perhaps worse risks if we don't stop global warming, and deal with the massive scale of human poverty that exists today.

Thursday, January 22, 2015

ThorCon moving forward?


"The American ‘skunkworks’ start-up ThorCon, established by shipping pioneer Jack Devanney, announced that Martingale, Inc has completed the pre-conceptual design of itsThorCon Molten Salt Reactor and that further design work is underway at a rapid pace. ThorCon’s MSR is a completely modernised version of Oak Ridge’s successful Molten Salt Reactor Experiment, part-fuelled by thorium and designed, from the ground up, for mass production in shipyards."

If this is true, we have in ThorCon a serious MSR backed by deep pockets.  If their design has already moved past the precopnceptual stage, and they are examining the whers and hows of manufacture, they are very serious.  The word "skunkworks " is important.  So far we have an internet presence, to which is attached the names of several credible people, nut nothing we can look at or touch.  I awate further evidence of a tangible reality with anticipation.

Tuesday, December 30, 2014

NNadir offers brilliant BraveNewClimate post on Climate change, energy poverty and ethics

NNadir is thew best of the pro-nuclear blogger. He and I have some areas of disagreement, but there are far more things we agree on. I would like some day to hear him read this essay, and interrupt him topfrom time to time to comment on I tcoment on things I disagree with, and praise the many passages I agree with and find admirable.



Guest Post by NNadir (who blogs occassionally at Daily Kos, profile here). This is a long but really interesting post. If you'd rather a PDF version, click here. The...
BRAVENEWCLIMATE.COM

Saturday, November 22, 2014

Nuclear Power, Where is China going with it?

According to Brian Wang, China has published its energy plans to 2020, while offering further energy projections to 2100.  The Chinese are estimating 58 GWe of nuclear power complete by 2020 with conventional nuclear output peaking at 200 GWe.

Despite a huge investment in wind and solar, China's recent investment in green electricity will pay off with more energy:
Wind 200 GW (290 TWh)
Solar 100 GW (60-70 TWh)
Nuclear 58 GW (410 TWh)


Among renewables, only nuclear and hydro are dispatchable, and nuclear has a much higher capacity factor.  Of course, of course, unlike nuclear, solar and wind require duplication and backup or energy storage, to make up for thdeir low capacity factor.  

According to Wang, Chinese plans to Top out to 200 GWe of LWR by the Middle of this Century and then start adding Fast Breeder Reactors:

Under previously announced plans, deployment of PWRs is expected to level off at 200 GWe by around 2040, with the use of fast reactors progressively increasing from 2020 to at least 200 GWe by 2050 and 1400 GWe by 2100.  

The fast reactor plan requires 85 years to implament.  This is largely due to the start charge bottleneck.  As I have pointed out it takes a fissionable fuel charge that is at least 10 times larger than the start charge of a thermal breeder, to go critical with a fast breeder.  Thus 10 LFTRs can go into operation for every fast Breeder that starts up.  This is why the Chinese are making a major investment in LFTR R&D.

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