A few posts back I mentioned that the United States Department of Energy had come up with the concept of "generations" of nuclear reactors in order to help explain its current strategy. The DOE is attempting to move forward on two different tracks: building Generation III reactors, which are supposedly improved designs of the light water reactors that were built in the 1970s and 1980s, and conducting research and development into what are hoped to be substantially better reactors to be built starting in the 2020s. The progress on new Gen-III reactors was slow even before the Fukushima Dai-ichi crisis, mostly due to rapid cost escalation. Reactor licensing by the Nuclear Regulatory Commission, which is an independent agency (from the DoE, but perhaps not from the nuclear industry), has also been slower than expected. The crisis in Japan is likely to delay new Gen-III reactors further, and possibly derail the whole program.
Unlike the Gen-III effort, which is concerned mostly with domestic issues of implementation, the Generation IV program is an international effort. Right now it's too early to say whether or not the Gen-IV program is on track - despite having been around for over a decade by now - because a lot of what participants are doing is basic materials research. The effort is also somewhat diffuse, as countries are really coordinating work more than they are working together. That's probably natural, as funding for the research is still being provided by member countries. And finally, it's hard to talk about the status of the effort as the latest GIF annual report hasn't been issued, and the DoE's status page is simply out of date.
The most important achievement of the GIF so far has been to select six conceptual designs for further research. (Note: most of the early GIF documents are no longer available from government websites, but some have been mirrored here.) At least 94 different concepts were submitted for review in early 2001. A number of meetings were then held to classify and evaluate the concepts. The result was a report which laid out the rational for selecting six proposals to be investigated further. Those concepts are: the gas-cooled fast reactor (GFR), the lead-cooled fast reactor (LFR), the molten-salt reactor (MSR), the sodium-cooled fast reactor (SFR), the super-critical water-cooled reactor (SCWR), and the very-high temperature reactor (VHTR). If some of those reactor types sound familiar, that's because they've been built before without much success. But the evaluation team evidently though they were worth a second look. Each of the concepts has a few options for implementation outlined in the report.
You may have noticed that three of the reactors have "fast" in their name, which is a reference to the energy of the neutrons that cause the self-sustaining nuclear reaction in a reactor core. The main advantage of fast neutron reactors is that they are able to "breed" large amounts of new fissionable elements during operation, which could potentially eliminate any uranium shortages for thousands or hundreds of thousands of years. With some configuration changes, fast reactors are also able to "burn" spent fuel from existing thermal neutron reactors. If successfully implemented on large scale, this feature would drastically reduced the amount of high-level waste that needs to be stored in expensive geological repositories. Of the other three, one is a standard thermal neutron reactor, one is a thermal neutron reactor capable of breeding at a low rate, and the other can be configured either to have either thermal or epithermal neutrons.
Below is my short summary of each concept. There are, of course, better summaries elsewhere, but I've added a bit of editorializing that (ahem) you just can't find elsewhere.
Sodium-cooled Fast Reactor - The SFR concept has been previously implemented the most number of times out of the six concepts. It is a fast neutron reactor that uses liquid sodium as the coolant in the primary and secondary cooling loops. In the past these reactors have been called liquid metal fast reactors (LMFR) or just fast breeder reactors (FBR). To date, a total of 20 SFRs have been built and operated, though only four are operating now. Of those still in use, only one produces electricity. The rest exist for research purposes. In addition to being able to breed more fissile material, the SFR has the advantage of providing high outlet temperature without requiring the high pressures found in LWRs. The main disadvantage is that sodium is flammable when exposed to air, and explosive when in contact with water. Most SFRs have used water in the tertiary coolant loop to generate steam for turbines, and water ingress into the secondary coolant loop has been a major problem. Some newer proposals use carbon dioxide in the tertiary loop to avoid the problem of sodium's volatility. My take on this concept is that it useful mainly for breeding in a nuclear power "ecosystem" that includes lots of non-breeder reactors. The difficulties encountered during implementation so far have made the SFR non-economic for electricity generation when compared to LWRs. I think that further research should be done on this concept, but focused on efficient breeding and ease of loading and unloading the fertile material.
Lead-cooled Fast Reactor - This concept is another liquid metal-cooled fast reactor, like the SFR. To date no LFRs have been built, but a closely related design using lead-bismuth eutectic as a coolant was built by the Soviet Union to power some of its submarines. It was not very successful, though that may have had more to do with the Soviet Union's military culture than the design itself. As with the SFR, the main advantage of this concept is high outlet temperatures at low primary loop pressures. Unlike sodium, liquid lead is not explosive when in contact with water, which eliminates the need for an intermediate loop. However, it is highly corrosive to most steels, and activation products (created when neutrons interact with elements in the coolant) remain radioactive for a long time. It also solidifies at a relatively high temperature, which makes handling difficult. And it is very heavy. The best application of this concept seems to be for very small nuclear reactors, on the order of 20-200 MWt, that can be shipped whole to a site and returned to a factory for processing. I think further research on this concept should be done only for small reactors.
Very-High Temperature Reactor - The VHTR is gas-cooled thermal
neutron reactor that uses helium in the primary cooling circuit. It has
previously been implemented a total of eight times, but only two
examples are currently operating. There are also a number of carbon
dioxide-cooled reactors operating, but that gas can't meet the requirements of the VHTR program. The main attraction of
this type of reactor is a very high outlet temperature, which is useful
for generating hydrogen from water and for supplying heat for other
industrial processes. The main disadvantage is very high temperatures in the core,
which many materials can withstand, but not in combination with high
radiation. Another important disadvantage is a once-through fuel cycle, which would leave lots of hot fuel elements that would have to be dealt with for hundreds or thousands of years. Helium is also somewhat tricky to contain, and expensive. My take on this type of reactor is that it was selected
when there was still a lot of buzz about using hydrogen to fuel motor
vehicles. The reasoning behind the so-called hydrogen economy no longer makes
sense, as batteries have improved significantly and people have come to recognize the difficulties of using hydrogen in that way. I think this concept should be dropped. Unfortunately, the DOE has made
the VHTR the first Gen-IV reactor it plans to build. The Next Generation Nuclear Plant (NGNP) program will start soliciting design
proposals from vendors late this year or in 2012.
Gas-cooled Fast Reactor - The GFR is a logical next step from the VHTR, combining both high temperatures and the ability to breed more fissile material. However, it is not as effective at breeding as either the SFR, and it has the same disadvantages as the VHTR. No examples have ever been built. I think this concept should be dropped.
Super-Critical Water Reactor - This concept is the next logical step from existing PWRs and BWRs. It would use light water at pressures above the critical point, beyond which water behaves like both a gas and a liquid. It would have only one coolant loop, like a BWR. The steam handling devices at the top of a BWR pressure vessel would be eliminated, allowing control rods to be inserted from the top as in a PWR. There are coal-fired power plants that use super-critical water already in operation, so the balance-of-plant (BOP) for the SCWR should almost be off-the-shelf. However, it is unknown if a reactor pressure vessel can safely operate at the extreme pressures needed. This concept can either operate as a thermal neutron reactor, or an epithermal neutron reactor. The later has allows for a low level of breeding, but makes loss of coolant accidents (LOCAs) potentially more dangerous. Canada is working on a subtype of this reactor that would leverage its experience with heavy water reactors. Because of its high degree of similarity to existing reactors, I think research on this concept should be continued, with a focus on answering questions about the safety of the pressure vessel as soon as possible. *
Molten Salt Reactor - The MSR is a concept quite unlike any of the others above in that it would not use solid fuel elements. Instead, the fissile material would be dissolved in the coolant, which is a mixture of fluoride salts (salt is used here in the technical sense, not in reference to standard table salt). The nuclear chain reaction would only take place in the reactor core, where the combination a large mass of the fuel-salt fluid and a graphite moderator would bring the mixture to criticality. The hot salt would then be cooled by a secondary loop of salt, which would in turn transfer the heat to a gas or water tertiary loop. The concept could operate as a thermal neutron breeder, producing enough new fissile material to fuel itself for years. The concept is another that would provide high outlet temperatures at low pressures. Despite it's exotic nature, an example of this reactor was built and operated briefly in the 1950s, and another was built and operated in the 1960s for several years. The biggest disadvantages of this concept are that the salts are corrosive, and processing the salt to remove certain fission byproducts currently is expensive. I think an increased pace of research into the MSR is warranted, as it has a number of positive aspects not found in any of the other reactor concepts.
There is one hybrid of these concepts to note, the molten salt-cooled reactor. This reactor would use a core similar to the VHTR but use a molten salt in the primary cooling circuit. I think the concept may be useful an intermediate step towards a fluid-fueled MSR, but only as a research reactor, not as a commercial design.
If resources were more plentiful, funding research into all of the six reactor concepts would be worthwhile. There are commonalities between them all, and scientific research is certainly more productive than blowing up wedding parties in Central Asia. But in the current budgetary environment, I think the focus should be on the SFR, SCWR, and MSR concepts, with a limited additional amount directed towards small LFRs.
* Added 2011/05/14: I should add that I think the answer to the question will be no, a safe RPV can't be created for a SCWR.