FIELD

The present disclosure relates generally to nuclear reactors. More particularly, the present disclosure relates to molten salt nuclear reactors.

BACKGROUND

Molten salt reactors (MSRs) were primarily developed from the 1950s to 1970s but, as of late, there has been increasing world interest in this type of reactor. Older concepts are being re-evaluated and new ideas put forth. This class of nuclear reactor has a great deal of advantages over current nuclear reactors, the advantages including potentially lower capital costs, overall safety, long lived waste profile and resource sustainability.

With MSRs advantages also come some significant technological challenges which lead to difficult basic design decisions. The first and likely foremost is whether and how a neutron moderator may be employed. Graphite has, in almost all cases, been chosen as a moderator as it behaves very well in contact with the fluoride salts used in MSRs. These salts are eutectic mixtures of fissile and fertile fluorides (UF4, ThF4, PuF3, etc.) with other carrier salts such as LiF, BeF2 or NaF. Using graphite as a bulk moderator within the core of the MSR has many advantages. For example, it gives a softer or more thermalized neutron spectrum which provides improved reactor control and a greatly lowered starting fissile inventory. As well, using graphite throughout the core of a MSR allows the ability to employ what is known as an under-moderated outer zone which acts as a net absorber of neutrons and helps shield the outer reactor vessel wall from damaging neutron exposure. The vessel, which contains the nuclear core, has typically been proposed as being made of a high nickel alloy such as Hastelloy® N; however, other materials are possible.

The use of graphite within the core of the MSR (i.e., within the neutron flux of a MSR) can have a serious drawback however. That is, that graphite will first shrink and then expand beyond its original volume as it is exposed to a fast neutron flux. Overall expansion of graphite (graphite core) occurs when the volume of the graphite (graphite core) is larger than its original volume, i.e., the volume preceding any neutron irradiation. An upper limit of total fast neutron fluence can be calculated and operation of the MSR is such that this limit is not exceeded. This limit determines when the graphite would begin to expand beyond its original volume and potentially damage surrounding graphite elements or the reactor vessel itself. How long graphite can be used within the reactor core is thus directly related to the local power density and thus to the fast neutron flux it experiences. A low power density core may be able to use the same graphite for several decades. This is the case for many previous reactors employing graphite such as the British gas cooled Magnox and AGR reactors. They were extremely large and had a low power density for thermohydraulic reasons but, this permitted an extremely long graphite lifetime. However, MSRs would benefit from having a far higher power density and thus graphite lifetime can become an issue.

The scientists and engineers designing MSRs have long been faced with important design options. A first option is to simply design the reactor to be quite large and very low power density in order to get a full 30 year or more lifetime out of the graphite. Thus one can seal all the graphite within the vessel and the graphite can remain in the vessel for the design life of the nuclear plant. Examples of this choice can be found in the studies of Oak Ridge National Laboratories (ORNL) in the late 1970s and early 1980s. For example, ORNL TM 7207 proposes a 1000 MWe reactor which was termed the “30 Year Once Through” design which would have a large reactor vessel of approximately 10 meters in diameter and height in order to avoid the need for graphite replacement. Much of the later work by Dr. Kazuo Furukawa of Japan, on what are known as the FUJI series of reactor designs, also chose this route of large, low power nuclear cores. These very large cores have obvious economic disadvantages in terms of the sheer amount of material required to fabricate the core and reactor vessel, and in the excessive weight of the core. These challenges increase the cost and complexity of the surrounding reactor building as would be understood by those trained in the field. It should be added that a 30 year nuclear plant lifetime was quite acceptable in the 1970s but by today's standards would be thought short. 50 or 60 years is now desired and would mean a still larger core to allow this lifetime without graphite replacement.

A second option often proposed is to employ a much smaller, higher power density core but to plan for periodic replacement of the graphite. This approach was commonly assumed in the work at Oak Ridge National Laboratories (ORNL) in the design of the Molten Salt Breeder Reactor from about 1968 to 1976 before the program was cancelled. This 1000 MWe reactor design had an outer vessel of Hastelloy® N that would contain hundreds of graphite elements fitting together and filling the vessel but with passage channels for the molten salt fuel to flow and exit the core to external heat exchangers. In this second option, the reactor has much smaller dimensions which are of approximately 6 meters in diameter and height. In this case the graphite, particularly in the center of the core with the highest fast neutron flux, only had an expected lifetime of 4 years. Thus the reactor had to be designed to be shut down and opened up every 4 years to replace a large fraction of the graphite elements. This may not sound overly difficult to those not trained in the field but with molten salts, the fission products, some of which are relatively volatile, are in the fuel salt and can also embed themselves onto a surface layer of graphite and, for example, the inner metal surfaces of the reactor vessel. Thus just opening the reactor vessel was known to be an operation that could be difficult to perform without allowing radioactive elements to spread into the surrounding containment zone. As well, the design of the reactor vessel itself is more complex when it needs to be periodically opened. These challenges are why the route of larger, lower power density cores were often chosen.

A third option is to try to omit the use of graphite altogether. This is possible and results in reactors typically with a much harder neutron spectrum. An example of this choice is the Molten Salt Fast Reactor (MSFR) proposed by a consortium of French and other European researchers starting around year 2005. It has very serious drawbacks however. For example it requires upwards of five times the starting fissile load and any accidental exposure of the salt to a moderator, such as water or even hydrogen content in concrete, could lead to criticality dangers.

Beyond the issue of graphite lifetime, there are also the somewhat related issues of the lifetime of the reactor vessel itself and of the primary heat exchangers.

The reactor vessel wall may also have a limited lifetime due to neutron fluence with both thermal and fast neutrons potentially causing problems. The most commonly proposed material being a high nickel alloy, such as Hastelloy® N, with reasonably well understood behaviour and allowed limits of neutron fluence. As such, a great deal of effort goes into core design to limit the exposure of neutrons and/or lower the operating temperature of the vessel wall. As well, adding thickness to the wall may help as strength is lost with increased neutron exposure. This adds both weight and expense. It is thus a challenge to have a 30 to 60 year lifetime of the reactor vessel itself.

Another design challenge is the primary heat exchangers which transfer heat from the radioactive primary fuel salt to a secondary coolant salt. This coolant salt then typically transfers heat to a working media such as steam, helium, CO2 etc. In some cases these heat exchangers are outside or external the reactor vessel itself, which appears to be the case for all 1950s to 1980s ORNL designs. They also may be located within the reactor vessel itself which has its own set of advantages and challenges. One great advantage of internal heat exchangers is no radiation of significance need leave the reactor itself as only secondary coolant salt enters and leaves the vessel.

For both internal and external heat exchangers, the great challenge is in either servicing or replacing them. When a MSR is opened up, it can potentially lead to radioactivity being released into a containment zone or space. ORNL for example proposed common tube in shell heat exchangers external to the core, four heat exchanger units per 1000 MWe reactor. In the case of any tube leaks the operation was not to fix or plug tubes but to open the shell and remove the entire tube bundle and replace with a new bundle. Only after a cooling period would a decision be made on repair and reuse of the bundle or simple disposal. Thus it is clear that primary heat exchanger service and/or replacement techniques are a great challenge in MSR design.

Further, when either graphite or heat exchangers are replaced, then the issue of their safe storage must be also addressed as they will become significantly radioactive during operation. This represents yet another challenge in MSR overall plant design.

It should be further highlighted that the related nuclear design field of Fluoride salt cooled, High temperature Reactors (known as FHRs) has very similar issues. In this work the reactor design can be very similar but instead of the fuel being in the fluoride salt, it is in solid form within the graphite moderator using the fuel form known as TRISO. In this case the limited graphite lifetime is also a function of the lifetime of the solid TRISO fuels; however, all other design issues and challenges are very similar to MSR design work. In FHRs, the primary coolant salt is not nearly as radioactive but does typically contain some radioactive elements such as tritium and a similar set of challenges are present when planning to use solid block TRISO fuels and periodically replace them. A subset of FHR design involves using a pebble fuel form which does ease fuel replacement without opening up the reactor vessel; however, this type of design has its own set of issues

The decay heat that follows the shutdown of a nuclear reactor following the loss of external cooling has been a long-standing industry challenge. The incident at Fukushima Japan indicates the seriousness of the issue. If the decay heat is not removed quickly from the reactor, the temperature in the reactor rises to unacceptable levels. Thus the speed with which the initial decay heat can be removed from the reactor is critical.

Therefore, improvements in nuclear reactors are desirable.

SUMMARY

The present disclosure relates to the integration of the primary functional elements of graphite moderator and reactor vessel and/or primary heat exchangers and/or control rods into a single replaceable unit having a higher and more economic power density while retaining the advantages of a sealed unit. Once the design life of such an Integral Molten Salt Reactor (IMSR) is reached, for example, in the range of 3 to 10 years it is disconnected, removed and replaced as a unit and this unit itself may also potentially function as the medium or long term storage of the radioactive graphite and/or heat exchangers and/or control rods and/or fuel salt itself. The functions of decay heat removal and volatile off gas storage may also be integrated in situ.

The present disclosure also relates to nuclear reactor that has a reactor vessel surrounded by a buffer material. The buffer material can absorb decay heat when external cooling is lost. The absorption of decay heat is effected by the buffer material phase transition latent heat, the phase transition being that of solid phase to liquid phase. The absorption is also effected by convective heat transfer when the buffer material is in the liquid state. The convective heat transfer occurs between the reactor vessel and a heat sink in thermal contact with the buffer material.

In a first aspect of the disclosure, there is provided a method of operating a nuclear power plant, the nuclear power plant comprising a molten salt reactor (MSR) to produce heat, a heat exchanger system, and an end use system, the heat exchanger system to receive heat produced by the MSR and to provide the received heat to the end use system. The method comprises operating the MSR, the MSR comprising a vessel, a graphite moderator core positioned in the vessel, and a molten salt circulating at least in the vessel, the heat exchanger system having an inside portion located inside the vessel and an outside portion located outside the vessel, the inside portion having a plurality of heat exchangers, each heat exchanger having an inlet conduit and an outlet conduit, each inlet conduit and each outlet conduit extending from each respective heat exchanger, through the vessel, toward the outside portion of the heat exchanger system and connecting each respective heat exchanger to the outside portion of the heat exchanger system, the vessel being sealed to permanently integrate the graphite core into the vessel. The method also comprises shutting down the MSR in response to a shutdown event occurring, to obtain a shut-down MSR. The method also comprises severing any operational connection between the inlet conduits and the outlet conduits of the shut-down MSR and the outside portion of the heat exchanger system, to obtain a severed, shut-down MSR; sequestering the severed, shut-down MSR; and operationally connecting a replacement MSR to the outside portion of the heat exchanger system located outside the vessel of the replacement MSR.

In a second aspect of the present disclosure, there is provided a method of operating a nuclear power plant, the nuclear power plant comprising a molten salt reactor (MSR) to produce heat, a heat exchanger system, and an end use system, the heat exchanger system to receive heat produced by the MSR and to provide the received heat to the end use system. The method comprises operating the MSR, the MSR comprising a vessel, a graphite moderator core positioned in the vessel, and a molten salt circulating at least in the vessel, the heat exchanger system to receive heat from the molten salt. The method also comprises shutting down the MSR after a pre-determined duration of operation, to obtain a shut-down MSR; severing any operational connection between the shut-down MSR and any portion of the heat exchanger system located outside the vessel, to obtain a severed, shut-down MSR; sequestering the severed, shut-down MSR; and operationally connecting a replacement MSR to any portion of the heat exchanger system located outside the vessel of the replacement MSR.

Other aspects and features of the present disclosure will become apparent to those ordinarily skilled in the art upon review of the following description of specific embodiments in conjunction with the accompanying figures.

