It should be understood that the purpose of any technological research and development program is tho identify and fix problems related to the implementation of any new technology. Considering the radical and daring nature of the Molten Salt Reactor concept, and the extreme conditions the reactor was expected to operate under, developmental problems were to be expected. The point of the 1966 to 1969 Molten Salt Reactor experiment was to explore potential problems and fix them before the MSR went into development for commercial use. – CB
AN EVALUATION OF THE MOLTEN SALT BREEDER REACTOR
VI. STATUS OF MSBR TECHNOLOGY
A. MSRE - The Reference Point for Current Technology
The Molten Salt Reactor Experiment (MSRE) was begun in 1960 at ORNL as part of the Civilian Nuclear Power Program. The purpose of the experiment was to demonstrate the basic feasibility of molten salt power reactors. All objectives of the experiment were achieved during its successful operation from June 1965 to December 1969. These included the distinction of becoming the first reactor in the world to operate solely on 233U. Some of the more significant dates and statistics pertinent to the MSRE are given in Table II.
In spite of the success of the MSRE, there are many areas of molten salt technology which must be expanded and developed in order to proceed from this small non-breeding experiment to a safe, reliable, and economic 1000 MWe MSBR with a 30-year life. To illustrate this point, some of the most important differences in basic design and performance characteristics between the MSRE and a conceptual 1000 MWe MSBR are given in Table III. Scale-up would logically be accomplished through development of reactor plants of increasing size. Examination of Table III provides an appreciation of the scale-up requirements in going from the MSRE to a large MSBR. Some problems associated with progressing from a small experiment to a commercial, high performance power plant are not adequately represented by the comparison presented in the Table. Therefore it is useful to examine additional facets of MSBR technology in more detail.
[This statement says nothing more than that the MSR is under development, and that problems have been identified and steps to identify and implement corrections are underway. – CB]
B. Continuous Fuel Processing: The Key to Breeding
In order to achieve nuclear breeding in the single-fluid MSBR it is necessary to have an on-line continuous fuel processing system. This would accomplish the following:
VI. STATUS OF MSBR TECHNOLOGY
A. MSRE - The Reference Point for Current Technology
The Molten Salt Reactor Experiment (MSRE) was begun in 1960 at ORNL as part of the Civilian Nuclear Power Program. The purpose of the experiment was to demonstrate the basic feasibility of molten salt power reactors. All objectives of the experiment were achieved during its successful operation from June 1965 to December 1969. These included the distinction of becoming the first reactor in the world to operate solely on 233U. Some of the more significant dates and statistics pertinent to the MSRE are given in Table II.
In spite of the success of the MSRE, there are many areas of molten salt technology which must be expanded and developed in order to proceed from this small non-breeding experiment to a safe, reliable, and economic 1000 MWe MSBR with a 30-year life. To illustrate this point, some of the most important differences in basic design and performance characteristics between the MSRE and a conceptual 1000 MWe MSBR are given in Table III. Scale-up would logically be accomplished through development of reactor plants of increasing size. Examination of Table III provides an appreciation of the scale-up requirements in going from the MSRE to a large MSBR. Some problems associated with progressing from a small experiment to a commercial, high performance power plant are not adequately represented by the comparison presented in the Table. Therefore it is useful to examine additional facets of MSBR technology in more detail.
[This statement says nothing more than that the MSR is under development, and that problems have been identified and steps to identify and implement corrections are underway. – CB]
B. Continuous Fuel Processing: The Key to Breeding
In order to achieve nuclear breeding in the single-fluid MSBR it is necessary to have an on-line continuous fuel processing system. This would accomplish the following:
- Isolate protactinium-233 from the reactor environment so it can decay into the fissile fuel isotope uranium-233 before being transmuted into other isotopes by neutron irradiation.
- Remove undesirable neutron poisons from the fuel salt and thus improve the neutron economy and breeding performance of the system.
- Control the fuel chemistry and remove excess uranium-233 which is to be exported from the breeder system.
1. Chemical Process Development
The Oak Ridge National Laboratory has proposed a fuel processing scheme to accomplish breeding in the MSBR, and the flowsheet processes involve:
At this time, the basic chemistry involved in the MSBR processing scheme has been demonstrated in laboratory-scale experiments. Current efforts at Oak Ridge are being directed toward development of subsystems incorporating many of the required processing steps. Ultimately a complete breeder processing experiment would be required to demonstrate the system with all the chemical conditions and operational requirements which would be encountered with any MSBR.
Not shown on the flowsheet is a separate processing system which would require injecting helium bubbles into the fuel salt, allowing them to circulate in the reactor system until they collect fission product xenon, and then removing the bubbles and xenon from the reactor system. Xenon is a highly undesirable neutron poison which will hamper breeding performance by capturing neutrons which would otherwise breed new fuel. This concept for xenon stripping was demonstrated in principle by the MSRE, although more efficient and controllable stripping systems will be desirable for the MSBR. The xenon poisoning in the MSRE was reduced by a factor of six by xenon stripping; the goal for the MSBR is a factor of ten reduction.
[Xenon is the Achilles heal of reactor physics. Xenon, a fission product, is a noble radioactive gas. It has a high neutron cross section, which means that it captures neutrons which could better be used in promoting chain reactions, or in breeding new fuel. In solid fuel reactors, Xenon remains inside fuel capsules where it poisoned the nuclear process. The presence of Xenon inside solid fueled reactors created control issues, and attempts to compensate for Xenon poisoning, could make reactors unstable and potentially dangerous. However in fluid fueled reactors, Xenon can be removed or stripped. The ability to strip Xenon from reactor fuel was a major accomplishment of ORNL, and clearly demonstrated the importance of the MSR concept. – CB]
2. Fuel Processing Structural Materials
Aside from the chemical processes themselves, there are also development requirements associated with containment materials for the fuel processing systems. In particular, liquid bismuth presents difficult compatibility problems with most structural metals, and present efforts are concentrated on using molybdenum and graphite for containing bismuth. Unfortunately, both molybdenum and graphite are difficult to use for such engineering applications. Thus, it will be necessary to develop improved techniques for fabrication and joining before their use is possible in the reprocessing system.
[This is simply a development issue, but one which does not appear to pose exceptional challenges. - CB]
A second materials problem of the current fuel processing system is the containment for the fluorination step in which uranium is volatilized from the fuel salt. The fluorine and fluoride salt mixture is corrosive to most structural materials, including graphite, and present ORNL flowsheets show a “frozen wall” fluorinator which operates with a protective layer of frozen fuel salt covering a Hastelloy-N vessel wall. This component would require considerable engineering development before it is truly practical for use in on-line, full processing systems.
[Considering the extreme conditions that the fuel processing system operated under, engineering challenges were to be expected. By 1972, ORNL scientists and engineers had made considerable progress in overcoming these challenges, and there were reasons to be optimistic about further progress. – CB]
C. Molten Salt Reactor Design - Materials Requirements
In concept, the molten salt reactor core is a comparatively uncomplicated type of heat source. The MSRE reactor core, for example, consisted of a prismatic structure of unclad graphite moderator through which fuel salt flowed to be heated by the self-sustaining chain reaction which took place as long as the salt was in the graphite. The entire reactor internals and fuel salt were contained in vessels and piping made of Hastelloy-N, a high-strength nickel-base alloy which was developed under the Aircraft Nuclear Propulsion Program. Over the four-year lifetime of the MSRE, the reactor structural materials performed satisfactorily for the purposes of the experiments although operation of the MSRE revealed possible problems with long term use of Hastelloy-N in contact with fuel salts containing fission products.
The MSBR application is more demanding in many respects than the MSRE, and additional development work would be required in several areas of materials technology before suitable materials could become available.
1. Fuel and Coolant Salts
The MSRE fuel salt was a mixture of 7LiF–BeF2–ZrF4–UF4 in proportions of 65.0-29.1-5.0-0.9 mole %, respectively. Zirconium fluoride was included as protection against UO2 precipitation should inadvertent oxide contamination of the system occur. MSRE operation indicated that control of oxides was not a major problem and thus it is not considered necessary to include zirconium in future molten salt reactor fuels. It should also be noted that the MSRE fuel contained no thorium whereas the proposed MSBR fuels would include thorium as the fertile material for breeding. With the possible exception of incompatibilities with Hastelloy-N, the MSRE fuel salt performed satisfactorily throughout the life of the reactor.
The MSBR fuel salt, as currently proposed by ORNL, would be a mixture of 7LiF–BeF2–ThF4–UF4 in proportions of 71.7–16–12–0.3 mole %, respectively. This salt has a melting point of about 930°F (772 K) and a vapor pressure of less then 0.1 mm Hg (13 Pa) at the mean operating temperature of 1150°F (895 K). It also has about 3.3 times the density and 10 times the viscosity of water. Its thermal conductivity and volumetric heat capacity are comparable to water.
The high melting temperature is an obvious limitation for a system using this salt, and the MSBR is limited to high temperature operation. In addition, the lithium component must be enriched in 7Li in order to allow nuclear breeding, since naturally occurring lithium contains about 7.5% 6Li. 6Li is undesirable in the MSBR because of its tendency to capture neutrons, thus penalizing breeding performance.
[The statement about the high melting temperature of salt, is exceedingly strange. One of the disadvantages of water as a reactor coolant is its low boiling point and the high pressure created by heating steam above the boiling point of water. Steam pressure causes major safety problems for light water reactors. In contrast molten salts have extremely high boiling points and vapor pressure is no problem in MSRs. Hence a major safety problem of LWRs is eliminated by using hot molten salt as a coolant. This comment is thus a sort of swift boating. An attempt to point to major strength and argue that it is really a weakness. – CB]
The chemical and physical characteristics of the proposed MSBR fuel mixture have been and are being investigated, and they are reasonably well known for unirradiated salts. The major unknowns are associated with the reactor fuel after it has been irradiated. For example, not enough is known about the behavior of fission products. The ability to predict fission product behavior is important to plant safety, operation, and maintenance. While the MSRE provided much useful information, there is still a need for more information, particularly with regard to the fate of the so-called "noble-metal" fission products such as molybdenum, niobium and others which are generated in substantial quantities and whose behavior in the system is not well understood.
[Here we have more of the same. The MSRE explored numerous issues including the effects of radiation on molten salts, and the behavior of fission products in the circulating reactor salt fuel mixture. ORNL scientist wanted to know more. The desirability of conducting more research is regarded as a liability by WASH-1222. – CB]
A more complete understanding of the physical/chemical characteristics of the irradiated fuel salt is also needed. As an illustration of this point, anomalous power pulses were observed during early operation of the MSRE with 233U fuel which were attributed to unusual behavior of helium gas bubbles as they circulated through the reactor. This behavior is believed to have been due to some physical and/or chemical characteristics of the fuel salt which were never fully understood. Out-of-reactor work on molten fuel salt fission product chemistry is currently under way. Eventually, the behavior of the fuel salt would need to be confirmed in an operating reactor.
[Research is being conducted. But some answers will only come by trying the idea out. - CB]
The coolant salt in the secondary system of the MSRE was of molar composition 66% 7LiF - 34% BeF2. While this coolant performed satisfactorily (no detectable corrosion or reaction could be observed in the secondary svstem), the salt has a high melting temperature (850°F / 728 K) and is relatively expensive. Thus, it may not be the appropriate choice for power reactors for two reasons: (1) larger volumes of coolant salt will be used to generate steam in the MSBR, and (2) salt temperatures in the steam generator should be low enough, if possible, to utilize conventional steam system technology with feedwater temperatures up to about 550°F. The operation of MSRE was less affected by the coolant salt melting temperature since it dumped the 8 MWt of heat via an air-cooled radiator. The high melting temperatures of potential coolant salts remain a problem. The current choice is a eutectic mixture of sodium fluoride and sodium fluoroborate with a molar composition of 8% NaF - 92% NaBF4; this salt melts at 725°F (658 K). It is comparatively inexpensive and has satisfactory heat transfer properties.
However, the effects of heat exchanger leaks between the coolant and fuel salts, and between the coolant salt and steam systems, must be shown to be tolerable. The fluoroborate salt is currently being studied with respect to both its chemistry and compatibility with Hastelloy-N.
[Heat exchange leaks are a major problem with all liquid cooled reactors. Given the endemic problem, molten salt has several advantages over other coolants. It operates under low pressure. High pressure is a major factor in leaks. The molten salts used in the reactor do not interact chemically with some metals. Electro-chemical interactions can be controlled. Finally Hastelloy-N had by 1972, a considerable history of use in molten salt reactors. It was known to perform well under the high tempreture conditions found in heat exchanges. – CB]
2. Reactor Fuel Containment Materials
A prerequisite to success for the MSBR would be the ability to assure reliable and safe containment and handling of molten fuel salts at all times during the life of the reactor. It would be necessary, therefore, to develop suitable containment materials for MSBR application before plants could be constructed.
A serious question concerning compatibility of Hastelloy-N with the constituents of irradiated fuel salt was raised by the post-operation examination of the MSRE in 1971. Although the MSRE materials performed satisfactorily for that system during its operation, subsequent examination of metal which was exposed to MSRE fuel salt revealed that the alloy had experienced intergranular attack to depths of about 0.007 inch (0.2 mm). The attack was not obvious until metal specimens were tensile-tested, at which time cracks opened up as the metal was strained. Further examination revealed that several fission products, including tellurium, had penetrated the metal to depths comparable to those of the cracks. At the present time, it is thought that the intergranular attack was due to the presence of tellurium. Subsequent laboratory tests have verified that tellurium can produce, under certain conditions, intergranular cracking in Hastelloy-N.
Although the limited penetration of cracks presented no problems for the MSRE, concern now exists with respect to the chemical compatibility of Hastelloy-N and MSBR fuel salts when subjected to the more stringent MSBR requirements of higher power density and 30-year life. If the observed intergranular attack was indeed due to fission product attack of the Hastelloy-N, then this material may not be suitable for either the piping or the vessels which would be exposed to much higher fission product concentrations for longer periods of time. Efforts are under way to understand and explain the cracking problem, and to determine whether alternate reactor containment materials should be actively considered.
In addition to the intergranular corrosion problem, the standard Hastelloy-N used in the MSRE is not suitable for use in the MSBR because its mechanical properties deteriorate to an unacceptable level when subjected to the higher neutron doses which would occur in the higher power density, longer-life MSBR. The problem is thought to be due mainly to impurities in the metal which are transmuted to helium when exposed to thermal neutrons. The helium is believed to cause a deterioration of mechanical properties by its presence at grain boundaries within the alloy. It would be necessary to develop a modified Hastelloy-N with improved irradiation resistance for the MSBR, and some progress is being made in that direction. It appears at this time that small additions of certain elements, such as titanium, improve the irradiation performance of Hastelloy-N substantially. Development work on modified alloys with improved irradiation resistance is currently under way.
[Some problems with Hastelloy-N were identified during the course of the MSRE. Their sources were identified, and work on fixing the problem is underway. – The problems would be routine for R&D, and fixing them posed no significant challenge. Indeed they were fixed shortly after WASH-1222 was written. – CB]
3 Graphite
Additional developmental effort on two problems is required to produce graphites suitable for MSBR application. The first is associated with irradiation damage to graphite structures which results from fast neutrons. Under high neutron doses, of the order of 10^22 neutrons/cm2, most graphites tend to become dimensionally unstable and gross swelling of the material occurs.
Based on tests of small graphite samples at ORNL, the best commercially available graphites at this time may be usable to about 3 x 10^21 neutrons/cm2, before the core graphite would have to be replaced. This corresponds to roughly a four-year graphite lifetime for the ORNL reference design. While this might be acceptable, there are still uncertainties about the fabrication and performance of large graphite pieces, and additional work would be required before a four-year life could be assured at the higher MSBR power densities now being. considered. In any event, there would be an obvious economic incentive to develop longer-lived graphites for MSBR application since a four-year life for graphite is estimated to represent a fuel cycle cost penalty of about 0.2 mills/kW-hr relative to a system with 30-year graphite life.
The second major problem associated with graphites for MSBR application is the development of a sealing technique which will keep xenon, an undesirable neutron poison, from diffusing into the core graphite where it can capture neutrons to the detriment of breeding performance. While graphite sealing may not be necessary to achieve nuclear breeding in the MSBR, the use of sealed graphite would certainly enhance breeding performance. The economic incentives or penalties of graphite sealing cannot be assessed until a suitable sealing process is developed.
Sealing methods which have been investigated to date include pyrolytic carbon coating and carbon impregnation. Thus far, however, no sealed graphite that has been tested remained sufficiently impermeable to gas at MSBR design irradiation doses, and research and development in this area is continuing.
[Graphite a form of carbon used in pencils can serve as a reactor moderator, as well as a structural material. Most molten salt concepts envision the use of graphite. However, there are some drawbacks to graphite. Under the high and radiation conditions found inside the MSR, graphite is expected to deteriorate. Readers of the discussion forum of Kirk Sorensen’s blog will find extensive discussions of the advantages and disadvantages of graphite, and proposed methods of solving the graphite problem. The use of graphite as a moderator and structural material inside MSRs is not strictly speaking absolutely required, and other approaches have been explored. It is possible to use heavy water as a moderator, and even unmoderated designs are possible. One approach would be to avoid structural uses of graphite, but moderate the reactor with small graphite sphears that float in the fuel salts. The graphite spears could be floated out of the reactor as they aged, and replaced with new spears. The graphite related issues are perhaps the most difficult material concerns, but since it is possible to forgo using graphite entirely, Grphite problems are hardly fatal for molten salt reactor technology. - CB]
4. Other Structural Materials
In addition to the structural materials requirements for the reactor and fuel processing systems proper, there are other components and systems which have special materials requirements. Such components as the primary heat exchangers and steam generators must function while in contact with two, different working fluids.
At the present time, Hastelloy-N is considered to be the most promising material for use in all salt containment systems, including the secondary piping and components. Research to date indicates that sodium fluoroborate and Hastelloy-N are compatible as long as the water content of the fluoroborate is kept low; otherwise, accelerated corrosion can occur. Additional testing would be needed and is underway.
Hastelloy-N has not been adequately evaluated for service under a range of steam conditions and whether it will be a suitable material for use in steam generators is still not known.
[Here WASH-1222 does not point to significant problems, but to a need for more research and testing. That would be normal for a research & development projects. - CB]
The Oak Ridge National Laboratory has proposed a fuel processing scheme to accomplish breeding in the MSBR, and the flowsheet processes involve:
- Fluorination of the fuel salt to remove uranium as UF6.
- Reductive extraction of protactinium by contacting the salt with a mixture of lithium and bismuth.
- Metal transfer processing to preferentially remove the rare earth fission product poisons which would otherwise hinder breeding performance.
At this time, the basic chemistry involved in the MSBR processing scheme has been demonstrated in laboratory-scale experiments. Current efforts at Oak Ridge are being directed toward development of subsystems incorporating many of the required processing steps. Ultimately a complete breeder processing experiment would be required to demonstrate the system with all the chemical conditions and operational requirements which would be encountered with any MSBR.
Not shown on the flowsheet is a separate processing system which would require injecting helium bubbles into the fuel salt, allowing them to circulate in the reactor system until they collect fission product xenon, and then removing the bubbles and xenon from the reactor system. Xenon is a highly undesirable neutron poison which will hamper breeding performance by capturing neutrons which would otherwise breed new fuel. This concept for xenon stripping was demonstrated in principle by the MSRE, although more efficient and controllable stripping systems will be desirable for the MSBR. The xenon poisoning in the MSRE was reduced by a factor of six by xenon stripping; the goal for the MSBR is a factor of ten reduction.
[Xenon is the Achilles heal of reactor physics. Xenon, a fission product, is a noble radioactive gas. It has a high neutron cross section, which means that it captures neutrons which could better be used in promoting chain reactions, or in breeding new fuel. In solid fuel reactors, Xenon remains inside fuel capsules where it poisoned the nuclear process. The presence of Xenon inside solid fueled reactors created control issues, and attempts to compensate for Xenon poisoning, could make reactors unstable and potentially dangerous. However in fluid fueled reactors, Xenon can be removed or stripped. The ability to strip Xenon from reactor fuel was a major accomplishment of ORNL, and clearly demonstrated the importance of the MSR concept. – CB]
2. Fuel Processing Structural Materials
Aside from the chemical processes themselves, there are also development requirements associated with containment materials for the fuel processing systems. In particular, liquid bismuth presents difficult compatibility problems with most structural metals, and present efforts are concentrated on using molybdenum and graphite for containing bismuth. Unfortunately, both molybdenum and graphite are difficult to use for such engineering applications. Thus, it will be necessary to develop improved techniques for fabrication and joining before their use is possible in the reprocessing system.
[This is simply a development issue, but one which does not appear to pose exceptional challenges. - CB]
A second materials problem of the current fuel processing system is the containment for the fluorination step in which uranium is volatilized from the fuel salt. The fluorine and fluoride salt mixture is corrosive to most structural materials, including graphite, and present ORNL flowsheets show a “frozen wall” fluorinator which operates with a protective layer of frozen fuel salt covering a Hastelloy-N vessel wall. This component would require considerable engineering development before it is truly practical for use in on-line, full processing systems.
[Considering the extreme conditions that the fuel processing system operated under, engineering challenges were to be expected. By 1972, ORNL scientists and engineers had made considerable progress in overcoming these challenges, and there were reasons to be optimistic about further progress. – CB]
C. Molten Salt Reactor Design - Materials Requirements
In concept, the molten salt reactor core is a comparatively uncomplicated type of heat source. The MSRE reactor core, for example, consisted of a prismatic structure of unclad graphite moderator through which fuel salt flowed to be heated by the self-sustaining chain reaction which took place as long as the salt was in the graphite. The entire reactor internals and fuel salt were contained in vessels and piping made of Hastelloy-N, a high-strength nickel-base alloy which was developed under the Aircraft Nuclear Propulsion Program. Over the four-year lifetime of the MSRE, the reactor structural materials performed satisfactorily for the purposes of the experiments although operation of the MSRE revealed possible problems with long term use of Hastelloy-N in contact with fuel salts containing fission products.
The MSBR application is more demanding in many respects than the MSRE, and additional development work would be required in several areas of materials technology before suitable materials could become available.
1. Fuel and Coolant Salts
The MSRE fuel salt was a mixture of 7LiF–BeF2–ZrF4–UF4 in proportions of 65.0-29.1-5.0-0.9 mole %, respectively. Zirconium fluoride was included as protection against UO2 precipitation should inadvertent oxide contamination of the system occur. MSRE operation indicated that control of oxides was not a major problem and thus it is not considered necessary to include zirconium in future molten salt reactor fuels. It should also be noted that the MSRE fuel contained no thorium whereas the proposed MSBR fuels would include thorium as the fertile material for breeding. With the possible exception of incompatibilities with Hastelloy-N, the MSRE fuel salt performed satisfactorily throughout the life of the reactor.
The MSBR fuel salt, as currently proposed by ORNL, would be a mixture of 7LiF–BeF2–ThF4–UF4 in proportions of 71.7–16–12–0.3 mole %, respectively. This salt has a melting point of about 930°F (772 K) and a vapor pressure of less then 0.1 mm Hg (13 Pa) at the mean operating temperature of 1150°F (895 K). It also has about 3.3 times the density and 10 times the viscosity of water. Its thermal conductivity and volumetric heat capacity are comparable to water.
The high melting temperature is an obvious limitation for a system using this salt, and the MSBR is limited to high temperature operation. In addition, the lithium component must be enriched in 7Li in order to allow nuclear breeding, since naturally occurring lithium contains about 7.5% 6Li. 6Li is undesirable in the MSBR because of its tendency to capture neutrons, thus penalizing breeding performance.
[The statement about the high melting temperature of salt, is exceedingly strange. One of the disadvantages of water as a reactor coolant is its low boiling point and the high pressure created by heating steam above the boiling point of water. Steam pressure causes major safety problems for light water reactors. In contrast molten salts have extremely high boiling points and vapor pressure is no problem in MSRs. Hence a major safety problem of LWRs is eliminated by using hot molten salt as a coolant. This comment is thus a sort of swift boating. An attempt to point to major strength and argue that it is really a weakness. – CB]
The chemical and physical characteristics of the proposed MSBR fuel mixture have been and are being investigated, and they are reasonably well known for unirradiated salts. The major unknowns are associated with the reactor fuel after it has been irradiated. For example, not enough is known about the behavior of fission products. The ability to predict fission product behavior is important to plant safety, operation, and maintenance. While the MSRE provided much useful information, there is still a need for more information, particularly with regard to the fate of the so-called "noble-metal" fission products such as molybdenum, niobium and others which are generated in substantial quantities and whose behavior in the system is not well understood.
[Here we have more of the same. The MSRE explored numerous issues including the effects of radiation on molten salts, and the behavior of fission products in the circulating reactor salt fuel mixture. ORNL scientist wanted to know more. The desirability of conducting more research is regarded as a liability by WASH-1222. – CB]
A more complete understanding of the physical/chemical characteristics of the irradiated fuel salt is also needed. As an illustration of this point, anomalous power pulses were observed during early operation of the MSRE with 233U fuel which were attributed to unusual behavior of helium gas bubbles as they circulated through the reactor. This behavior is believed to have been due to some physical and/or chemical characteristics of the fuel salt which were never fully understood. Out-of-reactor work on molten fuel salt fission product chemistry is currently under way. Eventually, the behavior of the fuel salt would need to be confirmed in an operating reactor.
[Research is being conducted. But some answers will only come by trying the idea out. - CB]
The coolant salt in the secondary system of the MSRE was of molar composition 66% 7LiF - 34% BeF2. While this coolant performed satisfactorily (no detectable corrosion or reaction could be observed in the secondary svstem), the salt has a high melting temperature (850°F / 728 K) and is relatively expensive. Thus, it may not be the appropriate choice for power reactors for two reasons: (1) larger volumes of coolant salt will be used to generate steam in the MSBR, and (2) salt temperatures in the steam generator should be low enough, if possible, to utilize conventional steam system technology with feedwater temperatures up to about 550°F. The operation of MSRE was less affected by the coolant salt melting temperature since it dumped the 8 MWt of heat via an air-cooled radiator. The high melting temperatures of potential coolant salts remain a problem. The current choice is a eutectic mixture of sodium fluoride and sodium fluoroborate with a molar composition of 8% NaF - 92% NaBF4; this salt melts at 725°F (658 K). It is comparatively inexpensive and has satisfactory heat transfer properties.
However, the effects of heat exchanger leaks between the coolant and fuel salts, and between the coolant salt and steam systems, must be shown to be tolerable. The fluoroborate salt is currently being studied with respect to both its chemistry and compatibility with Hastelloy-N.
[Heat exchange leaks are a major problem with all liquid cooled reactors. Given the endemic problem, molten salt has several advantages over other coolants. It operates under low pressure. High pressure is a major factor in leaks. The molten salts used in the reactor do not interact chemically with some metals. Electro-chemical interactions can be controlled. Finally Hastelloy-N had by 1972, a considerable history of use in molten salt reactors. It was known to perform well under the high tempreture conditions found in heat exchanges. – CB]
2. Reactor Fuel Containment Materials
A prerequisite to success for the MSBR would be the ability to assure reliable and safe containment and handling of molten fuel salts at all times during the life of the reactor. It would be necessary, therefore, to develop suitable containment materials for MSBR application before plants could be constructed.
A serious question concerning compatibility of Hastelloy-N with the constituents of irradiated fuel salt was raised by the post-operation examination of the MSRE in 1971. Although the MSRE materials performed satisfactorily for that system during its operation, subsequent examination of metal which was exposed to MSRE fuel salt revealed that the alloy had experienced intergranular attack to depths of about 0.007 inch (0.2 mm). The attack was not obvious until metal specimens were tensile-tested, at which time cracks opened up as the metal was strained. Further examination revealed that several fission products, including tellurium, had penetrated the metal to depths comparable to those of the cracks. At the present time, it is thought that the intergranular attack was due to the presence of tellurium. Subsequent laboratory tests have verified that tellurium can produce, under certain conditions, intergranular cracking in Hastelloy-N.
Although the limited penetration of cracks presented no problems for the MSRE, concern now exists with respect to the chemical compatibility of Hastelloy-N and MSBR fuel salts when subjected to the more stringent MSBR requirements of higher power density and 30-year life. If the observed intergranular attack was indeed due to fission product attack of the Hastelloy-N, then this material may not be suitable for either the piping or the vessels which would be exposed to much higher fission product concentrations for longer periods of time. Efforts are under way to understand and explain the cracking problem, and to determine whether alternate reactor containment materials should be actively considered.
In addition to the intergranular corrosion problem, the standard Hastelloy-N used in the MSRE is not suitable for use in the MSBR because its mechanical properties deteriorate to an unacceptable level when subjected to the higher neutron doses which would occur in the higher power density, longer-life MSBR. The problem is thought to be due mainly to impurities in the metal which are transmuted to helium when exposed to thermal neutrons. The helium is believed to cause a deterioration of mechanical properties by its presence at grain boundaries within the alloy. It would be necessary to develop a modified Hastelloy-N with improved irradiation resistance for the MSBR, and some progress is being made in that direction. It appears at this time that small additions of certain elements, such as titanium, improve the irradiation performance of Hastelloy-N substantially. Development work on modified alloys with improved irradiation resistance is currently under way.
[Some problems with Hastelloy-N were identified during the course of the MSRE. Their sources were identified, and work on fixing the problem is underway. – The problems would be routine for R&D, and fixing them posed no significant challenge. Indeed they were fixed shortly after WASH-1222 was written. – CB]
3 Graphite
Additional developmental effort on two problems is required to produce graphites suitable for MSBR application. The first is associated with irradiation damage to graphite structures which results from fast neutrons. Under high neutron doses, of the order of 10^22 neutrons/cm2, most graphites tend to become dimensionally unstable and gross swelling of the material occurs.
Based on tests of small graphite samples at ORNL, the best commercially available graphites at this time may be usable to about 3 x 10^21 neutrons/cm2, before the core graphite would have to be replaced. This corresponds to roughly a four-year graphite lifetime for the ORNL reference design. While this might be acceptable, there are still uncertainties about the fabrication and performance of large graphite pieces, and additional work would be required before a four-year life could be assured at the higher MSBR power densities now being. considered. In any event, there would be an obvious economic incentive to develop longer-lived graphites for MSBR application since a four-year life for graphite is estimated to represent a fuel cycle cost penalty of about 0.2 mills/kW-hr relative to a system with 30-year graphite life.
The second major problem associated with graphites for MSBR application is the development of a sealing technique which will keep xenon, an undesirable neutron poison, from diffusing into the core graphite where it can capture neutrons to the detriment of breeding performance. While graphite sealing may not be necessary to achieve nuclear breeding in the MSBR, the use of sealed graphite would certainly enhance breeding performance. The economic incentives or penalties of graphite sealing cannot be assessed until a suitable sealing process is developed.
Sealing methods which have been investigated to date include pyrolytic carbon coating and carbon impregnation. Thus far, however, no sealed graphite that has been tested remained sufficiently impermeable to gas at MSBR design irradiation doses, and research and development in this area is continuing.
[Graphite a form of carbon used in pencils can serve as a reactor moderator, as well as a structural material. Most molten salt concepts envision the use of graphite. However, there are some drawbacks to graphite. Under the high and radiation conditions found inside the MSR, graphite is expected to deteriorate. Readers of the discussion forum of Kirk Sorensen’s blog will find extensive discussions of the advantages and disadvantages of graphite, and proposed methods of solving the graphite problem. The use of graphite as a moderator and structural material inside MSRs is not strictly speaking absolutely required, and other approaches have been explored. It is possible to use heavy water as a moderator, and even unmoderated designs are possible. One approach would be to avoid structural uses of graphite, but moderate the reactor with small graphite sphears that float in the fuel salts. The graphite spears could be floated out of the reactor as they aged, and replaced with new spears. The graphite related issues are perhaps the most difficult material concerns, but since it is possible to forgo using graphite entirely, Grphite problems are hardly fatal for molten salt reactor technology. - CB]
4. Other Structural Materials
In addition to the structural materials requirements for the reactor and fuel processing systems proper, there are other components and systems which have special materials requirements. Such components as the primary heat exchangers and steam generators must function while in contact with two, different working fluids.
At the present time, Hastelloy-N is considered to be the most promising material for use in all salt containment systems, including the secondary piping and components. Research to date indicates that sodium fluoroborate and Hastelloy-N are compatible as long as the water content of the fluoroborate is kept low; otherwise, accelerated corrosion can occur. Additional testing would be needed and is underway.
Hastelloy-N has not been adequately evaluated for service under a range of steam conditions and whether it will be a suitable material for use in steam generators is still not known.
[Here WASH-1222 does not point to significant problems, but to a need for more research and testing. That would be normal for a research & development projects. - CB]
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