Charles - what is at issue in not your lack of authority as a scientist but your use of 'quotes' from 'distinguished scientists' that are simply cherry picking out of context information to supposedly support your position. Take the following as an example . . .
Did you not read the complete article (a proposal for a project to test and hopefully confirm) the operational and commercial feasibility of MSR's? If you did, why didn't you state this?
3.4 Cost of Electricity Studies
The cost of electricity from an MSR was recently estimated [Moir, 2002], based on
pre-1980 designs. When compared to coal and LWRs of the same era the COE was
lower. The ratio of the COE for the MSR to that of the LWR is 0.93. We will see whether
this cost advantage still holds when the design is analyzed against revised modern
standards. Cost comparisons such as those of [Delene, 1994] should be made.
4. Importance of Proposed Project
Molten salt reactors have the potential of meeting the goals of Generation IV reactors
better than solid fuel reactors. They also have the potential of meeting the goals of the
high-level waste transmutation program better than solid fuel reactors. In fact, they may
enable doing most if not all of the transmutation planned for accelerator-driven
subcritical reactors. The proposed study will give DOE the basis for further planning in
this area of nuclear research.
Energy R&D planners need to know if nuclear energy can be expanded greatly, be
safe, be economical, have adequate resources, have acceptable impacts on waste
repositories and have acceptable risks of weapons proliferation. The molten-salt reactor
holds the promise of better options for the future. This study will thoroughly assess this
option for energy planners to act upon.
Good repository science will result in new robust waste forms and a strategy to
minimize repository needs by a factor of ten to one hundred due to application of
efficient processing coming out of modern physical chemistry science.
Folks, read Charles' links to determine if they, in fact, support his position or he is just 'burying us in BS' and hopes that we won't read his links.
I offer the following quotes from the same link:Beginning with Section 5. Itemized Work Plan.
After the quotes, SkipinBluff added:
"Plus much more... I apologize for the length of this reply but it is probably necessary to convince readers of this site that some contributors are cherry picking - not telling the whole story."
Since this exchange involved extensive quotations from Deep-Burn, I have decided to post most of text of Deep-Burn as a matter of reference for anyone who is interested.
Deep-Burn Molten-Salt Reactors
By R. W. Moir of Lawrence Livermore National Laboratory, T. J. Dolan of Idaho National Engineering and Environmental Laboratory, Sean M. McDeavitt of Argonne National Laboratory. D. F. Williams and C. W. Forsberg of Oak Ridge National Laboratory, and E. Greenspan and J. Ahn of the Department of Nuclear Engineering, University of California, Berkeley
Abstract
The molten salt reactor (MSR)–a fluid fuel reactor, operated successfully at ORNL
30 years ago, demonstrated high burn-up and on-line fueling and found a solution to the corrosion issue, but then the MSR research was discontinued. The thinking at the time was that there were too many reactors types being developed, that the resources and talent were spread too thin and that the efforts should be concentrated on one advanced reactor type, the liquid metal cooled fast breeder reactor (LMFBR). Now, priorities have shifted. Breeding of extra fissile fuel, the primary motivation for the LMFBR development, is no longer as important a goal as it was in the sixties and seventies. Rather, transmutation of the nuclear waste is becoming a major goal. Many fission reactor concepts are being re-examined, with emphasis on passive safety, non-proliferation, waste minimization, resource utilization, and economics. In this light, the demonstrable advantages of a fluid fuel (with extremely high burnup capability, no burnup reactivity swing and on-line removal of poisons and mobile fission products), coupled with recent technical advances in nuclear technology, modeling, and materials science suggest that the timing is right for a fresh look at the MSR. We propose a multi- laboratory/University study of the MSR, with goals to:
• minimize weapons useable material in storage,
• minimize need for high level waste repository space,
• increase the proliferation resistance of nuclear energy
• make beneficial use of spent fuel from LWRs,
• increase resource utilization,
• greatly expand non-carbon based energy (electricity and hydrogen production) at a cost competitive with alternatives.
2. Project Objectives
The Molten Salt Reactor (MSR) was developed in the 1960s at ORNL as a breeder
reactor operating on a thorium-233U fuel cycle. The fuel is dissolved in a molten salt [see Section 3.1: “Description of the Molten-Salt Reactor (MSR).”]. Since uranium resources were considered very limited, the design goal was to maximize the breeding ratio of the reactor. At that time, non-proliferation was not an issue and the reactor used high- enriched uranium. Safety was important but not considered a major issue. Community- wide, waste management was not considered a significant issue at the time; thus, no significant research was conducted on waste management.
National goals have changed. Safety, non-proliferation, waste management, and
economics are major drivers in the design of a modern reactor. The proposed activity will determine whether an MSR can meet these goals and satisfy today’s design criteria. We believe that the MSR could be a superior reactor in each area.
Molten salt reactors (MSR) have the potential of making nuclear energy significantly more competitive than alternative energy sources by virtue of the following characteristics:
1. Relatively low specific capital investment, low fuel cycle cost, no carbon
emissions, and high efficiency for converting thermal to electrical energy. These
factors should facilitate economic competitiveness with alternative energy options
and allow large-scale deployment.
2. Very high fuel utilization and small amount of waste. This facilitates better use of limited natural resources and the burnup of weapons useable materials. It
minimizes the needs for repository space by a factor of ten to one hundred,
making for a more sustainable energy system and a cleaner world. Molten salt
fuel is one of the most suitable forms for refabrication as remote handled-fuel.
3. Superb safety. The option of continually removing fission products and keeping
the fissile inventory at a minimum results in enhanced safety from accidents. The
MSR is designed to be passively safe independent of size, not just in small sizes.
In contrast, for solid fuel reactors to be passively safe they have to be of a
relatively small size (typically up to several hundreds of MWe per unit).
4. Supply of high temperature heat (700 °C with conventional design, and higher
with new materials) improves overall efficiency and may be suitable for efficient
generation of hydrogen. These advantages could stimulate a market place
transition away from carbon based fuels.
The MSR concept deserves to be re-evaluated, because it can satisfy today’s
priorities to:
• minimize weapons useable material in storage,
• minimize need for high level waste repository space,
• increase the proliferation resistance of nuclear energy
• make beneficial use of spent fuel from LWRs,
• increase resource utilization,
• greatly expand non-carbon based energy (electricity and hydrogen production) at a cost competitive with alternatives.
In the proposed NERI project, the objectives are:
• Address and help overcome the potential technical and scientific obstacles to the long-term future use of nuclear energy in the U.S., including the issues involving resistance to proliferation, economics and nuclear waste disposition. The proposed research will make a major contribution to this objective by quantifying the benefits associated with Molten Salt Reactors in each of these areas.
• Advance the state of nuclear technology to maintain a competitive position in
overseas markets and a future domestic market. The proposed MSR design will
provide a commercially viable, safe, proliferation-resistant path for the US nuclear
power industry.
• Promote and maintain a nuclear science and engineering infrastructure to meet future technical challenges. The proposed research will build infrastructure in several laboratories and will also train students in this art. The proposed research fits well within the scope of the NERI solicitation in two categories:
• Advanced Nuclear Energy Systems: This program element includes the
investigation and preliminary development of advanced reactor and power conversion system concepts that offer the prospect of improved performance and operation, design simplification, enhanced safety and reduced overall cost. Proposed projects may involve innovative reactor, system or component designs, alternative power conversion cycles for terrestrial applications, or other important design features and characteristics.
• Advanced Nuclear Fuels/Fuel Cycles: Research and development is needed to
provide measurable improvements in the understanding and performance of nuclear fuel and fuel cycles with respect to safety, waste production, proliferation resistance, and economics to enhance the long-term viability of nuclear energy systems
3. Background
3.1 Description of the molten-salt reactor (MSR)-ORNL Design
In the late 1960s, a conceptual design of a 1000-MW(e) Molten-Salt Breeder Reactor
(MSBR) was developed [Robertson et al. 1971]. The design characteristics are shown in
Table 1, and a schematic of the reactor is shown in Fig 1. The proposed MSR concept is
similar to that reactor except for changes in reactor design and the associated fuel-salt
processing system to change the proliferation-resistance, safety, and waste- generation
characteristics of the reactor. The general plant layout, heat-transfer systems, and power-
generation systems for the proposed MSR are similar. However, we are looking at a
different salt composition to see how much neutron balance degradation occurs by
replacing Li by Na to minimize tritium production and replacing Be by Zr to minimize
chemical toxicity. Fig. 1 shows a loop design. Pool designs are possible. The loop versus
pool debate has yet to take place for the MSR. The major changes in design are:
• Uranium-233. The MSBR fuel used weapons-usable 233U. The MSR fuel will be
denatured 233U; a mixture of thorium and 233U, 238U, and other uranium isotopes.
• LWR actinide wastes burning. The MSR core design and salt-processing system will
be modified to burn other (LWR) fission reactor actinide wastes
• Plutonium. The MSR core design and salt-processing systems will be modified to
suppress plutonium production and inventories.
• Safety. The MSR design will be modified to address modern safety philosophy and
criteria. In particular, it will be designed to be passively safe.
• Waste management
. The MSR design will be modified to minimize the
inventory and toxicity of actinides that need be disposed of in the repository.
• Proliferation resistance. The MSR fuel cycle will be designed to be as proliferation
resistant as practical.
• Alternative molten salt
. We will study the feasibility of using NaF-ZrF4 as the
molten salt; it is free of Be that is chemically toxic material and of Li that generates
tritium.
The fuel is a liquid mixture of lithium-7 fluoride, beryllium fluoride, thorium
fluoride, and uranium fluorides. During operation, various fission products and actinides
also form fluorides in the liquid. Nuclear criticality occurs in the reactor vessel, which
contains unclad graphite. The liquid-fuel salt flows upward through vertical channels in
the graphite. The graphite slows down fast-fission neutrons and creates a thermal neutron
flux. The heat is primarily generated in the liquid fuel. The molten fuel has a high
boiling point; thus, the reactor operates at atmospheric pressure. The liquid-fuel salt
enters the reactor vessel at 565 °C (1050 °F) and exits at 705 °C (1300 °F). The reactor
and primary system are constructed of modified Hastelloy for corrosion resistance to the
molten salt. An inert cover gas is used to prevent unwanted chemical reactions. For a
discussion of the Hastelloy and the strategy to deal with corrosion see p 81-87 of [Engel
et al., 1980]. The strategy was to add 1 to 2% Nb to the Hastello and to keep the salt 4
reducing by maintaining the ratio of UF4/UF3 less than 60. This strategy showed good
corrosion suppression.
(Fig. 1 can be observed in the PDF version of this document:Fig 1 is a schematic of the ORNL molten-salt reactor. Adding 238U to the thorium can make this reactor proliferation resistant.)
(Explanation)
The fuel flows to a primary heat exchanger, where the heat is transferred to a
secondary fluid. The liquid fuel flows back to the reactor core. The secondary fluid
(NaBF4-NaF) provides isolation between the molten fuel and the steam cycle and a
means to trap the small amounts of tritium (~2400 Ci/d) that may be generated in the
primary coolant. The heat-transfer fluid flows to a steam generator to produce steam and
back to the primary heat exchanger. A conventional steam cycle converts the heat to
electricity. The electrical efficiency of the plant is ~44%. The high efficiency, as
compared to that of LWRs, is a consequence of the high reactor operating temperatures
and is a nice advantage of the MSR. The temperatures are determined by the need to
ensure low salt viscosity and a significant margin between the salt melting point and the
system operating temperature. It is a consequence of the selection of the salt composition.
Xenon and other fission-product gases are stripped from the salt in the primary-system circulation pumps. The reactor has control rods for rapid shutdown, however, during
normal operation, the control rods are in the fully withdrawn position. . A very strong
negative temperature coefficient of reactivity was demonstrated during operation of the
Aircraft Reactor Experiment and the Molten Salt Reactor Experiment [Haubenreich,
1970], and this feature is a fundamental characteristic of a molten-salt-fueled reactor. As
salt is heated, it expands and fuel is removed from the reactor.
The 700°C molten salt outlet temperature could be raised somewhat, possibly
permitting gas turbine power conversion systems and hydrogen production, however
neither is the objective of this proposal.
secondary fluid. The liquid fuel flows back to the reactor core. The secondary fluid
(NaBF4-NaF) provides isolation between the molten fuel and the steam cycle and a
means to trap the small amounts of tritium (~2400 Ci/d) that may be generated in the
primary coolant. The heat-transfer fluid flows to a steam generator to produce steam and
back to the primary heat exchanger. A conventional steam cycle converts the heat to
electricity. The electrical efficiency of the plant is ~44%. The high efficiency, as
compared to that of LWRs, is a consequence of the high reactor operating temperatures
and is a nice advantage of the MSR. The temperatures are determined by the need to
ensure low salt viscosity and a significant margin between the salt melting point and the
system operating temperature. It is a consequence of the selection of the salt composition.
Xenon and other fission-product gases are stripped from the salt in the primary-system circulation pumps. The reactor has control rods for rapid shutdown, however, during
normal operation, the control rods are in the fully withdrawn position. . A very strong
negative temperature coefficient of reactivity was demonstrated during operation of the
Aircraft Reactor Experiment and the Molten Salt Reactor Experiment [Haubenreich,
1970], and this feature is a fundamental characteristic of a molten-salt-fueled reactor. As
salt is heated, it expands and fuel is removed from the reactor.
The 700°C molten salt outlet temperature could be raised somewhat, possibly
permitting gas turbine power conversion systems and hydrogen production, however
neither is the objective of this proposal.
(Table 1 can be found in the PDF version of this document.)
3.2 Non-proliferation MSR designs
Several limited studies [Bauman, 1977; Engel, 1978] were undertaken to identify
methods to improve the proliferation resistance of the MSBR. One study [Engel et al.,
1978 and 1980] examined the possibility of an MSBR that operates with isotopically
diluted 233U. The study indicated that isotopic dilution of 233U (<12 style="font-weight:bold;">Table 2 can be found in the PDF version of this document.)
3.3 Molten-Salt Reactor Studies in Japan
The Molten-Salt Reactor Experiment (MSRE) operated successfully for four years in
the late 1960s at 8 MWth [MacPherson, 1985]. Should the MSR development get going
again it might go through a series of steps such as those advanced by a series of studies in
Japan on a small reactor (7 MWe) called mini-Fuji, and a mid-size reactor (155 MWe)
called Fuji-II [Furukawa et al., 1992]. These studies are illustrated in Fig. 2 and 3. The
Japanese designs are strongly based on the ORNL design. We would base our studies on
these designs as well.
(3.3 contains two illustrations found in the PDF version of this document.)
3.4 Cost of Electricity Studies
The cost of electricity from an MSR was recently estimated [Moir, 2002], based on
pre-1980 designs. When compared to coal and LWRs of the same era the COE was
lower. The ratio of the COE for the MSR to that of the LWR is 0.93. We will see whether
this cost advantage still holds when the design is analyzed against revised modern
standards. Cost comparisons such as those of [Delene, 1994] should be made.
4. Importance of Proposed Project
Molten salt reactors have the potential of meeting the goals of Generation IV reactors
better than solid fuel reactors. They also have the potential of meeting the goals of the
high-level waste transmutation program better than solid fuel reactors. In fact, they may
enable doing most if not all of the transmutation planned for accelerator-driven
subcritical reactors. The proposed study will give DOE the basis for further planning in
this area of nuclear research.
Energy R&D planners need to know if nuclear energy can be expanded greatly, be
safe, be economical, have adequate resources, have acceptable impacts on waste
repositories and have acceptable risks of weapons proliferation. The molten-salt reactor
holds the promise of better options for the future. This study will thoroughly assess this
option for energy planners to act upon.
Good repository science will result in new robust waste forms and a strategy to
minimize repository needs by a factor of ten to one hundred due to application of
efficient processing coming out of modern physical chemistry science.
5. Itemized Work Plan
5.1 LLNL
We will evaluate the impact of the MSR on economics, proliferation and national
security issues. For example, one basis is to assume 1000 MSR plants are operating by
the end of the century. What impact on repository needs for processed LWR spent fuel
would be obtained? Much of our findings will be on a per reactor basis. Studies will be
carried out to include the impact of changes in design guidelines since the last major
work published in [Engel et al., 1980] to include the impact of burning actinides,
processing and waste forms. The cost of electricity for MSR of pre-1980 vintage was
estimated to be about 90% that of the same era LWR. We will update this estimate by
updating all aspects that went into these estimates. The impacts of MSR on repositories
(follow-on to Yucca Mountain) under various scenarios will be evaluated.
We will revise the MSR design to meet our project objectives by incorporating new
materials, new or appropriate chemical process technologies anad items required to meet
modern safety guidelines.
Our goals for this study are:
1. Prove that it is possible to design a large MSR to be passively safe
2. Prove that we can design a MSR to be proliferation resistant
3. Prove that we can design MSR to burn most of the actinides from LWR spent
fuel and to need a relatively small repository volume
4. Prove that the MSR has the promise to be economically competitive
5. Prove that the overall environmental impact of the MSR is acceptable (or better
than that of LWRs)
6. Provide possible scenarios for deployment of MSRs
The LLNL contributions will be made by Ralph Moir, Jim Hassberger, William
Halsey and others.
5.2 INEEL -- Safety and Environmental Issues
The design of the molten salt reactor will be evaluated in view of the changes in
safety standards and methodology since the molten salt reactor days of two or more
decades ago and in view of the increased actinide inventories in some of the burning
scenarios.
The Molten Salt Reactor Experiment was shut down about 30 years ago. The
shutdown was not because of safety concerns or significant technical issues, but was to
focus limited available talent and funding on liquid metal reactors. Since then the
emphasis has shifted from high breeding ratio for a plutonium economy to passive safety,
non-proliferation, resource utilization, waste handling, synthetic fuel manufacturing, and
competitive economics. Our understanding of the phenomena has improved greatly, and
better codes have been developed to model neutronics, thermal hydraulics, probabilistic
risk analysis, and environmental impact. After TMI and Chernobyl the operating
procedures, safety regulations, licensing requirements, and environmental requirements
have also changed. In view of the changed emphasis, capabilities, and rules, we need to
re-examine the molten salt reactor concept, to see what role it could play in the new
world scenario. The preliminary answer to these questions appears to be favorable.
We will do the following tasks:
1. Consider the safety and environmental philosophy prevalent in the USA, and how
they have changed since the 1960s, and their impact on future power plant designs
and licensing compared to the pre-1980 MSR design.
2. Study activation of structure and control of dose rates to plant personnel.
3. Estimate the routine emissions to environment of noble gases, tritium, and other
volatile materials, and discuss how to keep them within safe limits.
4. Estimate the volume of routine periodic waste disposal during normal operation.
5. Study a variety of possible accident scenarios, including
• Initiating events
• Failure of vessel or piping
• Passive safety features
• Engineered safety features (if needed)
• Effects of air or water ingress
• Source term issues
• Potential for escape of radionuclides
• Potential offsite consequences of various scenarios
• Potential to avoid the need for an offsite evacuation plan
• Cleanup procedures after an accident
• Time delay to restart.
6. Evaluate the potential utilization of some separated fission products, and the
storage requirements for others.
5. Prove that the overall environmental impact of the MSR is acceptable (or better
than that of LWRs)
6. Provide possible scenarios for deployment of MSRs
The LLNL contributions will be made by Ralph Moir, Jim Hassberger, William
Halsey and others.
5.2 INEEL -- Safety and Environmental Issues
The design of the molten salt reactor will be evaluated in view of the changes in
safety standards and methodology since the molten salt reactor days of two or more
decades ago and in view of the increased actinide inventories in some of the burning
scenarios.
The Molten Salt Reactor Experiment was shut down about 30 years ago. The
shutdown was not because of safety concerns or significant technical issues, but was to
focus limited available talent and funding on liquid metal reactors. Since then the
emphasis has shifted from high breeding ratio for a plutonium economy to passive safety,
non-proliferation, resource utilization, waste handling, synthetic fuel manufacturing, and
competitive economics. Our understanding of the phenomena has improved greatly, and
better codes have been developed to model neutronics, thermal hydraulics, probabilistic
risk analysis, and environmental impact. After TMI and Chernobyl the operating
procedures, safety regulations, licensing requirements, and environmental requirements
have also changed. In view of the changed emphasis, capabilities, and rules, we need to
re-examine the molten salt reactor concept, to see what role it could play in the new
world scenario. The preliminary answer to these questions appears to be favorable.
We will do the following tasks:
1. Consider the safety and environmental philosophy prevalent in the USA, and how
they have changed since the 1960s, and their impact on future power plant designs
and licensing compared to the pre-1980 MSR design.
2. Study activation of structure and control of dose rates to plant personnel.
3. Estimate the routine emissions to environment of noble gases, tritium, and other
volatile materials, and discuss how to keep them within safe limits.
4. Estimate the volume of routine periodic waste disposal during normal operation.
5. Study a variety of possible accident scenarios, including
• Initiating events
• Failure of vessel or piping
• Passive safety features
• Engineered safety features (if needed)
• Effects of air or water ingress
• Source term issues
• Potential for escape of radionuclides
• Potential offsite consequences of various scenarios
• Potential to avoid the need for an offsite evacuation plan
• Cleanup procedures after an accident
• Time delay to restart.
6. Evaluate the potential utilization of some separated fission products, and the
storage requirements for others.
the reaction:
Li2BeF4 + 6 Ca3(PO4)2 --> 4(Li0.5 Be0.25 Ca4.5)(PO4)3F
and the product is a ceramic powder. Further lab-scale tests will be carried out to extend
this approach to the proposed NaF-ZrF4 salt system.
Process flow-sheets will be developed and proof-of-principle experiments will be
performed. The inventory of actinides and fission products ending up in the waste
streams will be quantified and characterized. The strawman path forward for the
immobilization of the loaded SFA powder is to follow the method successfully developed
for sodalite waste forms for chloride salt. That is, the SFA powders would be bonded
with ~25 volume percent glass using a pressureless consolidation method to form a
ceramic monolith. The impact of this approach on processing, waste volume, and fuel
cycle costs will be estimated.
Itemized tasks for this portion of the project are given below:
1. MSR Process Salt Waste Stream Definition
• Develop waste stream composition for process flowsheets in consultation with
other team members.
• Evaluate possible waste fission product separation and immobilization schemes
for inclusion in the waste process flowsheet (with ORNL).
2. Waste Form Processing development
• Lab-scale experiments to validate SFA as a host for NaF-ZrF4 waste salts.
• Waste form fabrication development
3. Waste Processing Equipment Definition and Cost Estimates
• Processing method and equipment selection
• Process throughput, scale, and capital cost estimates
The ANL contributions will be made my S. M. McDeavitt and others.
5.4 ORNL
Feasibility of on-line removal of the fission products will be analyzed and
recommendations for the proper staging of developments will be made. The batchwise
treatment of spent fuel salt to support salt recycle and minimize waste volumes will also
be explored. ORNL will lead this effort with assistance from ANL in the areas of
electrometallurgical treatments and waste form development. We will examine and assess
the processes for removal of fission products with careful attention to low fractional carry
over of actinides in the waste stream. The cost of processing will be roughly estimated
and the needed development program steps outlined. We will look into the feasibility of
using solvents that avoid Li and Beryllium. The MSR produces ~2400 Ci/d of tritium,
mostly from neutron reactions on lithium. It is also desirable to explore options that
eliminate the chemical toxicity of beryllium (if the nuclear performance is not degraded
too much). Alternate fluoride salt systems which meet these objectives will be
recommended on the basis of a physico-chemical screening for acceptable high-
temperature properties. Special attention will be given to solubility characteristics that
can impose burnup limitations (usually trivalent constituents). Minimizing the fission
product inventory in the core will improve the neutron economy, will enable reduction of
the actinide concentration in the MS, will increase the attainable discharge burnup and
will improve the reactor safety. The latter is a direct consequence of a reduction in the
source term, in case of an accident.
Previous MSR fluoride separations chemistry was largely directed to support
breeding in Th-based systems. Most efforts supported maximizing the neutron economy
for this purpose. Spent fuel treatments to support a sustainable disposition of fission
products and actinides were not developed. This state of affairs is undeniable, and the
experience with the spent fuel from the Molten Salt Reactor Experiment is evidence of
the relatively immature state of science in this area relative to the demands of a modern
fuel cycle.
Because of the very limited solubility (and corrosion) of fluoride salts in aqueous
systems, non-aqueous treatments are required. Some areas of non-aqueous treatments are
well developed (fluoride volatility), while others still require considerable research and
development (electrochemical, high-temperature treatments). No single separations
technology can accomplish the goals necessary to achieve the desired outcome. A careful
integration of non-aqueous separations tools will be required. Some new avenues will
need to be explored. The need for actinide/lanthanide separation is such a strategic area,
and it has recently been suggested that the thermodynamics for electrochemical
separation of lanthanides from actinides in fluoride media is more favorable than in
chlorides [Prusakov, 1999]. The development of the fluoride analog to the Russian
chloride electrochemical fuel cycle for vipac-oxide fuels [Bychkov, 1999] will also be
evaluated for its potential to optimize fuel treatments.
Itemized tasks for this portion of the project are given below:
1. On-line Salt Treatment, Fuel Treatment Basis, Alternate Salts
• Analysis of on-line salt treatment – needs & capabilities, including cost.
• Analysis of spent fuel treatment flowsheet options based upon modern technology
(with ANL).
• Physicochemical screening (solubility, melting point, vapor pressure, etc.) of
alternate fuel salt systems that avoid Li (tritium production) and Be (chem.
Toxicity).
2. Non-Traditional Treatments, Definition/Analysis of Reference Flowsheet
• Analysis of non-traditional spent-fuel treatment operations (with ANL).
• Definition and detailed analysis of reference spent fuel treatment flowsheet.
3. Technology Development Plan
• Definition of technology development plan to support commercial spent fuel
treatments (with ANL).
• Definition of technology development plan to support on-line fuel salt treatments.
The ORNL contributions will be made by D. F. Williams and others.
5.5 UC. Berkeley
Tasks by UC Berkeley can be divided into two items:
• Optimization of MSR core design (by Greenspan and others), and
• Repository-capacity analysis (by Ahn and others).
5.5.1 Optimization of MSR Core Design
Optimizing the heterogeneous core design. The optimization goals will be to
maximize the discharge burnup and minimize the volume and toxicity of the high level
waste while being able to maintain criticality and to be below the solubility limit of
actinides and fission products in the MS. At least two different MS materials will be
Previous MSR fluoride separations chemistry was largely directed to support
breeding in Th-based systems. Most efforts supported maximizing the neutron economy
for this purpose. Spent fuel treatments to support a sustainable disposition of fission
products and actinides were not developed. This state of affairs is undeniable, and the
experience with the spent fuel from the Molten Salt Reactor Experiment is evidence of
the relatively immature state of science in this area relative to the demands of a modern
fuel cycle.
Because of the very limited solubility (and corrosion) of fluoride salts in aqueous
systems, non-aqueous treatments are required. Some areas of non-aqueous treatments are
well developed (fluoride volatility), while others still require considerable research and
development (electrochemical, high-temperature treatments). No single separations
technology can accomplish the goals necessary to achieve the desired outcome. A careful
integration of non-aqueous separations tools will be required. Some new avenues will
need to be explored. The need for actinide/lanthanide separation is such a strategic area,
and it has recently been suggested that the thermodynamics for electrochemical
separation of lanthanides from actinides in fluoride media is more favorable than in
chlorides [Prusakov, 1999]. The development of the fluoride analog to the Russian
chloride electrochemical fuel cycle for vipac-oxide fuels [Bychkov, 1999] will also be
evaluated for its potential to optimize fuel treatments.
Itemized tasks for this portion of the project are given below:
1. On-line Salt Treatment, Fuel Treatment Basis, Alternate Salts
• Analysis of on-line salt treatment – needs & capabilities, including cost.
• Analysis of spent fuel treatment flowsheet options based upon modern technology
(with ANL).
• Physicochemical screening (solubility, melting point, vapor pressure, etc.) of
alternate fuel salt systems that avoid Li (tritium production) and Be (chem.
Toxicity).
2. Non-Traditional Treatments, Definition/Analysis of Reference Flowsheet
• Analysis of non-traditional spent-fuel treatment operations (with ANL).
• Definition and detailed analysis of reference spent fuel treatment flowsheet.
3. Technology Development Plan
• Definition of technology development plan to support commercial spent fuel
treatments (with ANL).
• Definition of technology development plan to support on-line fuel salt treatments.
The ORNL contributions will be made by D. F. Williams and others.
5.5 UC. Berkeley
Tasks by UC Berkeley can be divided into two items:
• Optimization of MSR core design (by Greenspan and others), and
• Repository-capacity analysis (by Ahn and others).
5.5.1 Optimization of MSR Core Design
Optimizing the heterogeneous core design. The optimization goals will be to
maximize the discharge burnup and minimize the volume and toxicity of the high level
waste while being able to maintain criticality and to be below the solubility limit of
actinides and fission products in the MS. At least two different MS materials will be
considered: NaF-ZrF4 and 7LiF-BeF2. The fuel feed besides LWR actinides will be
enriched uranium and thorium. The graphite structure will have to be replaced from time
to time due to radiation damage. The graphite lifetime in the MSR will be determined as
part of the core design.
In order to find the best way to burn LWR wastes, we will look at cases with
considerable LWR waste feed and cases with only a little LWR waste feed along with
thorium and 235U feed.
The primary thrust of this task is to find practical and safe core designs that will meet
the project goals. There will be six parts to this task: (1) Development and benchmarking
of computational tools. (2) Establishment of database and of design constraints. (3)
Neutronic parametric studies of cores for the MSR that is to be fuelled with the trans-
uranium isotopes from LWR spent fuel and, possibly, Th, using two different solvent
molten salts: 7LiF-BeF2 and NaF-ZrF4 . (4) Neutronic parametric studies of cores for the
MSR that is to be fuelled with denatured 233U and Th. (5) Reference core designs, taking
into account thermal-hydraulics and safety considerations. (6) Study of the approach to
equilibrium.
5.5.1.1 Computational Tools
Two computer code systems will be used for this study: the SCALE-4.4 code
package [SCALE, 1995] and MOCUP [Moore, at al. 1995]. We have used both code
systems for previous studies of MS reactor cores that are to establish an equilibrium fuel
composition [Hughes, et al., 1993][Lowenthal, et al., 2001]. For the first study [Hughes, et
al., 1993] we worked out a special sequence of selected modules and data libraries of the
SCALE-4.1 computer code package to simulate a MSR core that has a continuous feed in
of fuel and continuous extraction of stable and short lived fission products [Shayer, et al.,
1994]. For the present study we will implement a similar sequence within SCALE-4.4 –
the most updated version of the SCALE code package. The data libraries of SCALE-4.4
use more accurate evaluation of cross-sections (based on ENDF-B/VI) and have a finer
energy group structure. Also to be modified will be the definition of the elements to be
fed-into, and to be extracted from the MS.
The MOCUP code system [Moore, at al. 1995] is presently in use at UCB for the
neutronic analysis of thermal as well as fast reactor cores having fixed fuel. It has been
thoroughly benchmarked [Briesmeister, 1997] and found to be reliable. It is a linkage
code that couples MCNP [Briesmeister, 1997] -- a generalized-geometry, point-energy
Monte Carlo transport code, and ORIGEN-2 [Croff, 1980] -- an isotope generation and
depletion code. This combination of codes is very useful for depletion and transmutation
analysis of systems that have complex geometry or of systems for which ORIGEN-2 does
not have sufficiently accurate effective one-group transmutation cross sections. Such
cross sections can be generated by MCNP starting from the most updated evaluation of
point-energy cross-sections. Before applying MOCUP to the analysis of the MSR we’ll
have to set ORIGEN-2 to handle continuous feed and continuous extraction of selected
elements. Also, special algorithms will have to be added to account for loss of a fraction
of the delayed neutrons due to fuel recycling outside of the core and for the extraction of
233
Pa and feeding back its 233U decay product. Doppler broadened cross sections for
certain actinides will have to be added to the currently available MCNP library.
The two code systems will be benchmarked against each other for infinite unit cells
of MSR having different MS channel diameter, different lattice pitch, different fuels and
different feed and extraction scenarios. The SCALE-based MS reactor analysis code will
be the primary tool for scoping studies; solving the neutron transport equation
deterministically it is significantly faster than MOCUP that solves the neutron transport
equation stochastically. MOCUP, on the other hand, can be used to simulate axially finite
unit cell compositions, 3-D cores, control rods and other 2-D and 3-D geometries.
5.5.1.2 Establishment of Data Base and of Design Constraints
The database to be compiled includes the temperature dependence physical
properties of the molten salts to be considered, of graphite and of the metallic structural
material to be used. Particular attention will be devoted to the search for information on
the solubility limits of different actinides and fission products. Also to be determined is
the fraction of fission products that can be extracted from the MS and the fraction of
actinides that will get into the waste stream in the MS recycling system.
Constraints the study needs to abide by include solubility limits of actinides and
fission products, radiation damage limits of graphite, maximum acceptable MS pumping
power and maximum acceptable MS flow velocity.
5.5.1.3 Parametric Study of MSR Fuelled with Transuranics from LWRs
The parametric study will be similar to that we recently performed [Lowenthal, et al.,
2001] for studying the maximum possible fractional transmutation of different actinides
in MS reactor that operates with a once-through fuel cycle. The MS considered in that
study was NaF-ZrF4. The fuel feed consisted of Pu and Minor Actinides (MA) from
LWR spent fuel. The variable parameters of the study are the pitch of the MS channels,
the diameter of the fuel channels, the volumetric feed and removal rate of the fuel salt,
and the concentration of actinides in the feed salt. The core is modeled as an infinite array
of heterogeneous, hexagonal cells, (see Fig. 4) unit-cells that are finite in the axial
direction; thus accounting for the effect of axial reflectors, shields and axial neutron
leakage.
The search for the equilibrium concentration is done iteratively; for a given
concentration of actinides and fission products in the MS, MCNP [Briesmeister, 1997]
eigenvalue calculation is performed. In addition to the eigenvalue k (multiplication
factor) we extract from the MCNP run the total neutron flux in the MS, effective one-
group spectrum averaged cross sections for all of the core constituents, and the flux of
neutrons having energy above 10 keV in the graphite sleeve. Using these one-group
cross-section and total flux, a set of coupled steady-state rate equations (Eq. 1) are solved
to obtain a new equilibrium concentration of the actinides. If the actinides concentration
significantly differs from that used for the MCNP calculation, a new iteration of MCNP
run followed by solution of the rate equations is performed until convergence is reached.
(A set of equasions follow in the PDF text)
( The PDF text contains Fig. 4: Configuration used to model with MOCUP the graphite-moderated molten salt reactor.)
In Eq. 1, NA is the molar concentration of isotope A in the reactor’s stock of molten
salt, λA is the decay constant for A, σA→B is the spectrum-averaged (one-group)
microscopic cross section for the reaction transmuting a nucleus of A into a nucleus of
isotope B, φ is the total average neutron flux in the fuel, FA is the rate of feed of A per unit
volume of molten salt in the reactor system (perfect mixing is assumed), and R is the
fractional rate of removal of molten salt from the reactor system. The reaction rates in Eq.
1 are divided by 2 to account for the fraction of time the salt is outside of the core while
delivering heat to the heat exchangers
Figure 5 illustrates selected results obtained [Lowenthal, et al., 2001] from such a
parametric study for a NaF-ZrF4 MS. The solubility limit of actinides in this MS is
estimated to be 1.56 mole %, at a MS temperature of 550oC. It is found that it is possible
to reach an equilibrium core composition in which the actinides are below their solubility
limit and keff is close to 1.0. With the particular design parameters considered for the
study summarized in Fig. 5, keff is ~0.97; assuming that the radial non-leakage probability
is ~0.97. Nevertheless, by adjusting the power density, MS flow rate and concentration of
actinides in the MS feed it appears possible to achieve an equilibrium state that is critical.
5.5.1.4 Parametric Study of Denatured 233U-Th Fueled MSR
This study will be similar to that described in Sec. 3 but will consider denatured
233 U-Th fuel cycle. The denatured 233U will contain primarily 238U (≥80%). It is
envisioned that this denatured uranium will be obtained from reactors that burn Pu and
MA in the presence of Th and a suitable amount of 238U.
(Fig. 5 is found in the Pdf version of this text.)
Fig. 5:Dependency of keff and Actinides (Ac) equilibrium concentration on graphite-to molten salt volume ratio for 7-cm fuel channel diameter at a
constant MS feed of 0.8 liters/day having Actinides concentration of 12.87 mol%
5.5.1.5 Reference Core Designs
Focusing on the promising design domains of critical equilibrium MS cores as
identified using the simplified model described in Secs. 2 and 3, we’ll carry-out a detailed
three-dimensional neutronic analysis for a small number of MSR cores. This detailed
analysis will be done using MOCUP. It will account for all the actinides and fission
products. The ORIGEN2 part of MOCUP will be set to simulate continuous feed and
continuous removal of MS.
Among the parameters to be calculated are the power density distribution, flux levels
and radiation damage rates, temperature coefficients of reactivity and void coefficient of
reactivity. The control system of the reactors will be designed to provide adequate
shutdown margin. Also to be estimated is the MS pumping power requirements and the
inventory of actinides and fission products that will end up in the high-level-waste
stream.
Accident scenarios will be identified and preliminary analyzed. A passively safe
design will be attempted that is based on the thermal expansion of the MS. The thermal
expansion will dispel actinides with the MS. A design will be searched in which this MS
expansion will have a significantly negative reactivity effect.
5.5.1.6 Approach to Equilibrium
The preceding tasks consider MSR in which the fuel has reached its equilibrium
composition. In the present task we'll study the evolution of the fuel composition towards
its equilibrium state. Among the questions to be addressed is what is the optimal strategy
for fueling the reactor and how long will it take to approach the equilibrium composition.
The time evolution of the fuel isotopics will be determined.
One of the fuel cycle options to be considered during the approach to equilibrium is
the addition of thorium to the Pu and MA from LWR spent fuel as the fuel feed for the
MS reactor. This will enable to use the excess neutrons in the pre-equilibrium stage to
convert thorium into 233U. The 233U, that will be denatured with 238U, could be used to
fuel MSRs that operate on the denatured 233U-Th fuel cycle.
5.5.2 Repository-Capacity Analysis
Theoretical studies will be performed to confirm that repository capacity required for
the disposal of the waste from MSR is smaller by a factor of 10 to 100, compared with
the conventional LWR spent fuel disposal. This should be performed, based on the waste
compositions and solidification (to be studied by ANL and ORNL). Significant reduction
in repository space requirement is essential for sustainable deployment of the MSR
system at a large scale. It is imperative to demonstrate feasibility of reduction by a factor
of 10 to 100 and its implication to the sustainability of the system.
Repository performance should be evaluated from many viewpoints, such as
radiological health impact to the public, criticality safety, proliferation resistance,
institutional control, cost, public perception, and so on. With the proposed MSR system,
it is expected that the amount of actinides to be included in the waste stream would be
one to two orders of magnitude smaller than that in the conventional LWR spent fuel. In
the past repository performance studies, however, it has been claimed that the mass of
actinides in the waste does not affect the repository performance because of the low
solubility of actinides in groundwater. Recently, it was pointed out by the team of UC
Berkeley researchers [Ahn, et al., 2002] that even under the solubility-limited condition,
spatial configuration of the canister array in a repository could have an important
influence on the concentration of radionuclides leaving the repository area, thus affecting
the amount of radionuclides to be released to the far field around the repository.
In the proposed project, utilizing the waste compositions and solidification
determined by ANL and ORNL, performance assessment and preliminary design study
will be made for a geologic repository. A hypothetical repository is considered. The
computer code, VR, for repository performance assessment is readily available at UC
Berkeley [Tsujimoto, et al., 2000]. The code utilizes the Parallel Virtual Machine (PVM)
technology to perform radionuclide transport calculations for multi-canister
configurations.
For this task, three steps will be taken:
(1) While ANL and ORNL are working to develop a reference-case specification for
the waste from MSR, UC Berkeley will work on necessary extension of the VR code to
accommodate specific features of the MSR waste. This can be completed in the first
fiscal year of the project. Also in the first year, measures for the repository performance
will be developed. Presently, the radiological exposure dose rate to the public is
considered to be the most important measure, which is primarily determined by the
concentration of radionuclides in groundwater. Because the efforts to be made in this
project is to reduce the mass of long-lived actinides in the waste, developing alternative
and/or additional measures for the repository performance that show sensitively the
effects of mass reduction would be essential.
233 U-Th fuel cycle. The denatured 233U will contain primarily 238U (≥80%). It is
envisioned that this denatured uranium will be obtained from reactors that burn Pu and
MA in the presence of Th and a suitable amount of 238U.
(Fig. 5 is found in the Pdf version of this text.)
Fig. 5:Dependency of keff and Actinides (Ac) equilibrium concentration on graphite-to molten salt volume ratio for 7-cm fuel channel diameter at a
constant MS feed of 0.8 liters/day having Actinides concentration of 12.87 mol%
5.5.1.5 Reference Core Designs
Focusing on the promising design domains of critical equilibrium MS cores as
identified using the simplified model described in Secs. 2 and 3, we’ll carry-out a detailed
three-dimensional neutronic analysis for a small number of MSR cores. This detailed
analysis will be done using MOCUP. It will account for all the actinides and fission
products. The ORIGEN2 part of MOCUP will be set to simulate continuous feed and
continuous removal of MS.
Among the parameters to be calculated are the power density distribution, flux levels
and radiation damage rates, temperature coefficients of reactivity and void coefficient of
reactivity. The control system of the reactors will be designed to provide adequate
shutdown margin. Also to be estimated is the MS pumping power requirements and the
inventory of actinides and fission products that will end up in the high-level-waste
stream.
Accident scenarios will be identified and preliminary analyzed. A passively safe
design will be attempted that is based on the thermal expansion of the MS. The thermal
expansion will dispel actinides with the MS. A design will be searched in which this MS
expansion will have a significantly negative reactivity effect.
5.5.1.6 Approach to Equilibrium
The preceding tasks consider MSR in which the fuel has reached its equilibrium
composition. In the present task we'll study the evolution of the fuel composition towards
its equilibrium state. Among the questions to be addressed is what is the optimal strategy
for fueling the reactor and how long will it take to approach the equilibrium composition.
The time evolution of the fuel isotopics will be determined.
One of the fuel cycle options to be considered during the approach to equilibrium is
the addition of thorium to the Pu and MA from LWR spent fuel as the fuel feed for the
MS reactor. This will enable to use the excess neutrons in the pre-equilibrium stage to
convert thorium into 233U. The 233U, that will be denatured with 238U, could be used to
fuel MSRs that operate on the denatured 233U-Th fuel cycle.
5.5.2 Repository-Capacity Analysis
Theoretical studies will be performed to confirm that repository capacity required for
the disposal of the waste from MSR is smaller by a factor of 10 to 100, compared with
the conventional LWR spent fuel disposal. This should be performed, based on the waste
compositions and solidification (to be studied by ANL and ORNL). Significant reduction
in repository space requirement is essential for sustainable deployment of the MSR
system at a large scale. It is imperative to demonstrate feasibility of reduction by a factor
of 10 to 100 and its implication to the sustainability of the system.
Repository performance should be evaluated from many viewpoints, such as
radiological health impact to the public, criticality safety, proliferation resistance,
institutional control, cost, public perception, and so on. With the proposed MSR system,
it is expected that the amount of actinides to be included in the waste stream would be
one to two orders of magnitude smaller than that in the conventional LWR spent fuel. In
the past repository performance studies, however, it has been claimed that the mass of
actinides in the waste does not affect the repository performance because of the low
solubility of actinides in groundwater. Recently, it was pointed out by the team of UC
Berkeley researchers [Ahn, et al., 2002] that even under the solubility-limited condition,
spatial configuration of the canister array in a repository could have an important
influence on the concentration of radionuclides leaving the repository area, thus affecting
the amount of radionuclides to be released to the far field around the repository.
In the proposed project, utilizing the waste compositions and solidification
determined by ANL and ORNL, performance assessment and preliminary design study
will be made for a geologic repository. A hypothetical repository is considered. The
computer code, VR, for repository performance assessment is readily available at UC
Berkeley [Tsujimoto, et al., 2000]. The code utilizes the Parallel Virtual Machine (PVM)
technology to perform radionuclide transport calculations for multi-canister
configurations.
For this task, three steps will be taken:
(1) While ANL and ORNL are working to develop a reference-case specification for
the waste from MSR, UC Berkeley will work on necessary extension of the VR code to
accommodate specific features of the MSR waste. This can be completed in the first
fiscal year of the project. Also in the first year, measures for the repository performance
will be developed. Presently, the radiological exposure dose rate to the public is
considered to be the most important measure, which is primarily determined by the
concentration of radionuclides in groundwater. Because the efforts to be made in this
project is to reduce the mass of long-lived actinides in the waste, developing alternative
and/or additional measures for the repository performance that show sensitively the
effects of mass reduction would be essential.
(2) In the second fiscal year, a thorough performance assessment study will be made
with the MSR waste specifications developed by ANL and ORNL. Effect of arrangement in a hypothetical repository is studied. Through this study, feedback would
be made to the designs of the reactor and the partitioning processes to enhance the
sustainability of the system (i.e., to decrease the repository space requirement).
(3) In the third year, iteration between the front-end and the back-end of the system
will be performed, to tune the system design. We will revisit the performance measures
and check if these are properly set.
6. Collaboration
Efforts will be made to collaborate with other groups interested in MSR. The
Japanese especially have an interest, which will be pursued. Especially K. Furukawa’s
whose work is illustrated by his references and Fig. 2 and 3. He has agreed to collaborate
so his resume is included. The Central Research Institute for Electric Power Industry
(CRIEPI) of Japan has been studying molten-salt dry process for recovery of actinides in
collaboration with UC Berkeley would participate in the proposed project. Formal
agreement is yet to be arranged. French and Russian related work and contacts may
develop.
(Section 7, Project Schedule and Milestones, is omitted. It is mainly composed of scheduling tables., and can be found in the PDF version of the text.
8. Organizations and Qualifications
Lawrence Livermore National Laboratory will be the principal investigating
organization, with responsibilities for planning and coordinating all work performed
under this proposal. Special strengths are in areas of non-proliferation and repository
considerations.
Idaho National Engineering and Environmental Laboratory will have responsibility
for safety.
Argonne National Laboratory will have responsibility for waste form and processing.
Oak Ridge National Laboratory will have responsibility for processing.
University of California, Berkeley will have responsibility for reactor physics and
repository considerations.
Each organization has broad strengths that overlap.
9. Key Personnel
Ralph W. Moir
Education
B.S. 1962--Engineering Physics, University of California, Berkeley
Sc.D. 1967--Nuclear Engineering, MIT, Cambridge, Mass.
Professional Associations, Societies and Honors
Registered Professional Nuclear Engineer in the State of California, NU782.
American Physical Society, Fellow 1981, Plasma Physics Div
American Nuclear Society, Fellow 1989, Fusion Energy Div, Chairman 93 to 94.
Publications
• R. W. Moir, et al., "Tandem Mirror Hybrid Reactor Design Annual Report",
Lawrence Livermore National Laboratory, Livermore, CA, UCID-18808 (1980).
• R. W. Moir, "The Fusion-Fission Fuel Factory, Chapter 15, p. 411-451, in Fusion,
Vol. 1 Part B, edited by E. Teller, Academic Press, New York (1981).
• J. D. Lee and R. W. Moir, "Fission-Suppressed Blanket for Fissile Fuel Breeding
Fusion Reactors", J. Fusion Energy, 1, 299 (1981).
• J. D. Lee, et al., "Feasibility Study of a Fission-Suppressed Tandem-Mirror Hybrid
Reactor", Lawrence Livermore National Laboratory, Livermore, CA, UCID-19327
(1982).
• D. H. Berwald, et al., "Fission-Suppressed Hybrid Reactor--The Fusion Breeder",
Lawrence Livermore National Laboratory, Livermore, CA, UCID-19648 (1982).
• R. W. Moir, et al., "Design of a Helium-Cooled Molten Salt Fusion Breeder", Fusion
Technology, Vol. 8, No. 1 Part 2(A) 465 (1985).
• R. W. Moir, et al., "Study of a Magnetic Fusion Production Reactor", A series of
eight articles on tritium production. J. Fusion Energy, 5, 255-331 (1986) and 6, 3-88
(1987)
• R. W. Moir, R. L. Bieri, X. M. Chen, T. J. Dolan, M. A. Hoffman, P. A. House, R. L.
Leber, J. D. Lee, Y. T. Lee, J. C. Liu, G. R. Longhurst, W. R. Meier, P. F. Peterson,
R. W. Petzoldt, V. E. Schrock, M. T. Tobin, W. H. Williams, "HYLIFE-II: A Molten
Salt Inertial Fusion Energy Power Plant Design-Final Report," Fusion Technology 25
(1994) 5-25.
• B. G. Logan, L. J. Perkins, R. W. Moir and D. D. Ryutov, “The need for research and
development in fusion: economical energy for a sustainable future with low
environmental impact,” Fusion Technology, 28 (1995) 236-239.
• H. Moriyama, A. Sagara, S. Tanaka, R. W. Moir, D. K. Sze, “Molten salts in fusion
nuclear technology,” Fusion Engineering and Design 39-40 (1998) 627-637.
• M. D. Lowenthal, E. Greenspan, R. Moir, W. E. Kastenberg, T. K. Fowler, “Industrial
ecology for inertial fusion energy: selection of high-Z material for HYLIFE-II
targets,” Fusion Technology 34 (1998) 619-628.
• R. W. Moir, “Flibe coolant cleanup and processing in the HYLIFE-II inertial fusion
energy power plant,” UCRL-ID-143228 (2001).
• R. W. Moir, “Cost of electricity from molten salt reactors (MSR),” to be published in
Nuclear Technology April, 2002.
Thomas James Dolan
Consulting Scientist
Idaho National Engineering and Environmental Laboratory
P.O. Box 1625
Idaho Falls, ID 83415-3860, USA
Phone: 208-526-2235
FAX: 208-526-2930
Internet: dolatj@inel.gov
After his BS degree in Engineering Mechanics, Dr. Dolan served two years in the
Navy, then earned a PhD in Nuclear Engineering from the University of Illinois (1970).
He spent 10 months as a post-doctoral student at the Novosibirsk State University and
Institute of Nuclear Physics, USSR. He served as a professor of nuclear engineering at
the University of Missouri-Rolla 1971-1989, where he taught about 20 different courses
and served as Department Head 1985-1987. He had summer research jobs at the
Lawrence Livermore National Laboratory, Oak Ridge National Laboratory, Los Alamos
National Laboratory, and Institut National de la Recherche Scientifique – _nergie,
Université du Quebec. He was a visiting professor at Tsing Hua University, Taiwan
(1977-1978). He was a consultant to Phillips Petroleum Company on the Ohmic Heated
Toroidal Experiment (1981-1988). In 1987 he joined the Idaho National Engineering
Laboratory and worked on physics applications, fusion reactor design studies, space
nuclear power, arms control, and university programs.
From 1995-2001 he served as Head of the Physics Section, International Atomic
Energy Agency (IAEA), Vienna, Austria, doing administration, international research
coordination, and organization of technical meetings on nuclear fusion research,
utilization of research reactors and low-energy accelerators, and nuclear instrumentation.
He organized a coordinated research project involving 11 countries on "Comparison of
Compact Toroid Configurations" (nuclear fusion research). He presented invited lectures
in Iran, Italy, and Japan, and attended or organized technical meetings in Brazil, Canada,
China, Croatia, Egypt, Finland, France, Germany, India, Italy, Japan, Korea, Portugal,
Russia, Slovenia, and the Ukraine. In September 2001 he returned to the Idaho National
Engineering and Environmental Laboratory, where he has been involved with
Generation-4 reactor studies, advanced fuel studies, fission product chemistry, national
security programs, and university programs.
Selected Publications
• T.J. Dolan, Fusion Research, Pergamon Press, 1982 (textbook, now available on CD
free from the IAEA: n.pawel@iaea.org )
• J. K. Hartwell, L. Forman, and T. J. Dolan, “Warhead Demilitarization - Some Pros
and Cons,” Verification Technologies Review, 2, No.6, Nov/Dec 1990.
• C.D. Eshelman, H.K. Tseng, T.J. Dolan, and M.A. Prelas, "Plasma diagnostic x-ray
tomography system," Review of Scientific Instruments 62, 751-754 (1991).
• T.J. Dolan, G.R. Longhurst, and E. Garcia-Otero, "A vacuum disengager for tritium
removal from HYLIFE-II reactor Flibe," Fusion Technology 21, 1949-1954 (1992).
• T.J. Dolan, "Fusion power economy of scale," Fusion Technology 24, 97-111 (1993).
• R. W. Moir, R.L. Bieri, X. M. Chen, T. J. Dolan, et al., "HYLIFE-II: a molten salt
inertial fusion energy power plant design - final report," Fusion Technology 25, 5-25
(1994).
• T. J. Dolan, Review Article, "Magnetic Electrostatic Plasma Confinement," Plasma
Physics and Controlled Fusion 36, 1539-1593 (1994).
• U. Rosengard, T. Dolan, D. Miklush, and M. Samiei, “Neutrons for humanitarian
demining,” IAEA Bulletin, June 2001.
• T. J. Dolan, “Possible generation of self-magnetic fields,” Fusion Science and
Technology 40, 119-124 (September 2001).
Sean M. McDeavitt
Institution: Argonne National Laboratory
Classification: Nuclear Engineer/Section Manager, Materials Development Section
Specialty: Nuclear Materials and Materials Processing
Education:
Degree Discipline University Date
Ph. D. Nuclear Engineering Purdue University 1992
M. S. Nuclear Engineering Purdue University 1990
B. S. Nuclear Engineering Purdue University 1987
Experience Summary:
Dr. McDeavitt has been developing processing methods for nuclear materials for
over ten years. He joined the Chemical Technology Division (CMT) of Argonne
National Laboratory in December of 1992. He developed a stainless steel-15 wt. %
zirconium (SS-15Zr) waste form alloy to immobilize radioactive metallic waste from an
electrometallurgical processing method used to treat spent fuel. He led the SS-15Zr
project from concept to demonstration as a group leader in the Waste Form Development
Section of CMT; this alloy waste form is now being demonstrated in the Fuel
Conditioning Facility at ANL-West (near Idaho Falls, ID). In addition, he is currently
working with an external contractor to develop advanced crucible materials for melting
reactive liquid metal alloys. He also serves a project manager for a 1999 NERI project
for the development of an advanced thorium oxide-based cermet fuel and he has studied
swelling and densification mechanisms in nuclear fuel. Dr. McDeavitt is a member of the
Minerals, Metals, and Materials Society (TMS) (Chairman, Reactive Metals Committee)
and the American Nuclear Society (Member, Materials Science and Technology Division
Executive Committee).
Selected Publications:
• S. M. McDeavitt, “Uranium Processing for the Nuclear Fuel Cycle,” J. Miner. Met.
Mater., 52 (2000) 11.
• S. M. McDeavitt, D. P. Abraham, and J. Y. Park, “Evaluation of Stainless Steel
Zirconium Alloys as High Level Nuclear Waste Forms,” J. Nucl. Mater., 257 (1998)
21.
• S. M. McDeavitt, D. P. Abraham, J. Y. Park, and D. D. Keiser, “Stainless Steel-
Zirconium Waste Forms from Electrometallurgical Treatment of Spent Nuclear Fuel,”
J. Miner. Met. Mater., 49 (1997) 29.
• S. M. McDeavitt and A. A. Solomon, “Hot-Isostatic Pressing of U-10Zr by a Coupled
Grain Boundary Diffusion and Creep Cavitation Mechanism,” J. Nucl. Mater., 228
(1996) 184.
• S. M. McDeavitt, J. Y. Park, and J. P. Ackerman, “Defining a Metal-Based Waste
Form for IFR Pyroprocessing,” in Actinide Processing: Methods and Materials, Eds.
B. Mishra and W. A. Averill, (TMS Publication, Warrendale, PA,1994) 305.
Dave F. Williams
OAK RIDGE NATIONAL LABORATORY
Education
Virginia Institute of Technology B.S. 1978 Chemical and Nuclear
Engineering
University of Tennessee M.S. 1985 Chemical Engineering
University of Washington Ph.D. 1991 Chemical Engineering
David F. Williams has 15 years of professional experience in radiochemical R&D.
His experience has ranged from design work in support of production of sol-gel
particulate nuclear fuel, to development of flowsheet and equipment for the
radiochemical recovery of special isotopes, to more basic chemical research. For the past
three years he has led the basic research studies that established the salt chemistry
necessary for the remediation of the Molten Salt Reactor Experiment at ORNL. He is the
present Group Leader of the Chemistry Research Group in the Chemical Technology
Division.
Selected Publications
• D. F. Williams, A. S. Icenhour, L. D. Trowbridge, G. D. Del Cul, and L. M. Toth,
“Radiolysis Studies in Support of the Remediation of the Molten Salt Reactor
Experiment,” Transactions of the American Nuclear Society (invited paper published
in 1999 Winter Meeting Proceedings, November 14–18, 1999, Long Beach,
California).
• D. F. Williams, G. D. Del Cul, and L. M. Toth , “Molten Salt Fuel Cycle
Requirements for ADTT Applications,” 3rd International Conference on Accelerator
Driven Transmutation Technologies and Applications (ADDTA '99), Prague, Czech
Republic, June 7–11, 1999, (paper We-I-17) in http//www.fjfi.cvut.cz/con_adtt99/).
• D. F. Williams, J. Brynestad, “Evaluation of Fluorine-Trapping Agents for Use
During Storage of the MSRE Fuel Salt,” ORNL/TM-13770, Oak Ridge National
Laboratory, Oak Ridge, Tennessee, May 1999.
• D. F. Williams, L. M. Toth, and G. D. Del Cul, Chemical Interactions During
Melting of the MSRE Fuel Salt, ORNL/M-5506, Oak Ridge National Laboratory, Oak
Ridge, Tennessee, November 1996.
• D. F. Williams and F. J. Peretz, “Characterization of the Molten Salt Reactor
Experiment Fuel and Flush Salts,” American Nuclear Society Meeting Proceedings
(Conference on DOE Spent Nuclear Fuel and Fissile Material Management June 18,
1996, Reno, Nevada).
• D. F. Williams, G. D. Del Cul, and L. M. Toth, A Descriptive Model of the MSRE
after Shutdown Review of FY95 Progress, ORNL/TM-13142, Oak Ridge National
Laboratory, Oak Ridge, Tennessee, January 1996.
Charles Forsberg
OAK RIDGE NATIONAL LABORATORY
Education
University of Minnesota, Minneapolis B.S.—1969, Chemical
Engineering
Massachusetts Institute of Technology, Cambridge M.S.—1971, Nuclear Engineering
Massachusetts Institute of Technology, Cambridge Sc.D —1974, Nuclear Engineering
Professional Activities and Affiliations
Fellow, American Nuclear Society.
Member, American Association for the Advancement of Science
Member, American Institute of Chemical Engineers
Member, Materials Research Society
Member, U.S. Department of Energy 233U Team
Member, U.S. Department of Energy High-level Waste Technical Advisory Panel
Registered Professional Engineer (State of Tennessee)
Principal holder of eight U.S. patents
Highlights
Dr. Charles Forsberg is a senior staff member of ORNL. His research areas are
advanced reactors and fuel Cycles. His doctorate thesis was on uranium enrichment
technologies, and he has done subsequent research on reprocessing, fuel fabrication, and
other fuel-cycle technologies. He has been the program manager for several programs,
including the developmental LWR program, which examined inherently and passively
safe LWRs. He holds eight patents in the areas of passive safety systems for power
reactors, reprocessing, and waste treatment.
At ORNL, he is a member of the DOE 233U multi-site team addressing 233U safety
and storage issues. He directed the technical studies on disposition options for excess
233 U. He participated in the DOE TOPS workshops to examine how to improve
proliferation resistance in the nuclear fuel, is the U.S. molten-salt reactor contact for the
DOE/Russian Proliferation-Resistant Nuclear Technology (PRNT) program, and is a
member of the Non-classical reactor team for the Generation IV road map activity. Dr.
Forsberg led the team that developed the technical basis for defining weapons-usable 233U
(>12 wt % 233U in 238U), which is based on isotopic composition. He also developed the methodology to define waste thresholds for 233U, that is, the concentration of 233U in
waste at which safeguards may be terminated because the 233U is practicably
unrecoverable. He is currently conducting studies on the future uses of 233U for reactors
and other applications. Consequently, reviews of worldwide activities in these areas are
being completed.
Selected Publications (Total list is greater than 150 articles and reports)
• C. W. Forsberg, “Are Chemically Separable Weapons-Usable Fissile Materials a
Characteristic of Nuclear Power Systems”, Science and Global Security, (Submitted)
• C. W. Forsberg, “What is Non-Weapons-Usable Material?,” p. 62 in Trans. 1999
Winter American Nuclear Society Meeting, Long Beach, California, November
14–18, 1999, Vol. 81.
• C. W. Forsberg and L. C. Lewis, Uses For Uranium-233: What Should Be Kept for
Future Needs?, ORNL/TM-6952, Oak Ridge National Laboratory, Oak Ridge,
Tennessee, September 24, 1999.
• C. W. Forsberg, E. C. Beahm, L. R. Dole, A. S. Icenhour, S. N. Storch, L. C. Lewis,
and E. L. Youngblood, Disposition Options for Uranium-233, ORNL/TM-13553,
Oak Ridge National Laboratory, June 1, 1999.
• C. W. Forsberg, “Fissile-Waste Management Constraints: Safeguards and Criticality,”
pp. 91–91 in Trans. 1998 Winter Am. Nuc. Soc. Meeting, Washington, D.C.,
November 15–19, 1998, Am. Nuc. Soc., La Grange Park, Illinois.
• C. W. Forsberg, “Recovery of Fissile Materials From Wastes and Conversion of the
Wastes To Glass,” Nucl. Techno., 123, 341–349, September 1998.
• C. W. Forsberg, “Plutonium Futures,” MIT Nuclear Systems Safety Course,
Department of Nuclear Engineering, Massachusetts Institute of Technology,
Cambridge, Massachusetts, July 23, 1998.
• C. W. Forsberg, S. N. Storch, and L. Lewis, Uranium-233 Waste Definition:
Disposal Options, Safeguards, Criticality Control, and Arms Control, ORNL/TM-
13591, Oak Ridge National Laboratory, Oak Ridge, Tennessee, July 7, 1998.
• C. W. Forsberg, C. M. Hopper, J. L. Richter, and H.C. Vantine, Definition of
Weapons-Usable Uranium-233, ORNL/TM-13517, Oak Ridge National Laboratory,
Oak Ridge, Tennessee, March 1998.
• C. W. Forsberg and J. C. Conklin, “Passive Cooling System with Temperature
Control for Reactor Containments,” Nucl. Technol. 116, 55–65, October 1996.
• C. W. Forsberg, “Passive and Inherent Safety Technologies Applicable to Light-
Water Reactors,” Proc. 3rd Annual Former Soviet Union Nuclear Society Meeting,
St. Petersburg, Russia, September 14–18, 1992.
• C. W. Forsberg, “A Water-Level Initiated Decay Energy Cooling System,” Nucl.
Technol. 96, 229–235 (November 1991).
• C. W. Forsberg and A. M. Weinberg, “Advanced Reactors, Passive Safety, and the
Acceptance of Nuclear Energy,” Annual Rev. of Energy, 15, 133–152, 1990.
• C. W. Forsberg, et al., Proposed and Existing Passive and Inherent Safety-Related
Structures, Systems, and Components (Building Blocks) for Advanced Light-Water
Reactors, ORNL-6554, Oak Ridge National Laboratory, Oak Ridge, Tennessee,
December 1989.
• C. W. Forsberg, “Passive Emergency Cooling Systems for Boiling Water Reactors
(PECOS-BWR),” Nucl. Technol. 76, 185, January 1987.
methodology to define waste thresholds for 233U, that is, the concentration of 233U in
waste at which safeguards may be terminated because the 233U is practicably
unrecoverable. He is currently conducting studies on the future uses of 233U for reactors
and other applications. Consequently, reviews of worldwide activities in these areas are
being completed.
Selected Publications (Total list is greater than 150 articles and reports)
• C. W. Forsberg, “Are Chemically Separable Weapons-Usable Fissile Materials a
Characteristic of Nuclear Power Systems”, Science and Global Security, (Submitted)
• C. W. Forsberg, “What is Non-Weapons-Usable Material?,” p. 62 in Trans. 1999
Winter American Nuclear Society Meeting, Long Beach, California, November
14–18, 1999, Vol. 81.
• C. W. Forsberg and L. C. Lewis, Uses For Uranium-233: What Should Be Kept for
Future Needs?, ORNL/TM-6952, Oak Ridge National Laboratory, Oak Ridge,
Tennessee, September 24, 1999.
• C. W. Forsberg, E. C. Beahm, L. R. Dole, A. S. Icenhour, S. N. Storch, L. C. Lewis,
and E. L. Youngblood, Disposition Options for Uranium-233, ORNL/TM-13553,
Oak Ridge National Laboratory, June 1, 1999.
• C. W. Forsberg, “Fissile-Waste Management Constraints: Safeguards and Criticality,”
pp. 91–91 in Trans. 1998 Winter Am. Nuc. Soc. Meeting, Washington, D.C.,
November 15–19, 1998, Am. Nuc. Soc., La Grange Park, Illinois.
• C. W. Forsberg, “Recovery of Fissile Materials From Wastes and Conversion of the
Wastes To Glass,” Nucl. Techno., 123, 341–349, September 1998.
• C. W. Forsberg, “Plutonium Futures,” MIT Nuclear Systems Safety Course,
Department of Nuclear Engineering, Massachusetts Institute of Technology,
Cambridge, Massachusetts, July 23, 1998.
• C. W. Forsberg, S. N. Storch, and L. Lewis, Uranium-233 Waste Definition:
Disposal Options, Safeguards, Criticality Control, and Arms Control, ORNL/TM-
13591, Oak Ridge National Laboratory, Oak Ridge, Tennessee, July 7, 1998.
• C. W. Forsberg, C. M. Hopper, J. L. Richter, and H.C. Vantine, Definition of
Weapons-Usable Uranium-233, ORNL/TM-13517, Oak Ridge National Laboratory,
Oak Ridge, Tennessee, March 1998.
• C. W. Forsberg and J. C. Conklin, “Passive Cooling System with Temperature
Control for Reactor Containments,” Nucl. Technol. 116, 55–65, October 1996.
• C. W. Forsberg, “Passive and Inherent Safety Technologies Applicable to Light-
Water Reactors,” Proc. 3rd Annual Former Soviet Union Nuclear Society Meeting,
St. Petersburg, Russia, September 14–18, 1992.
• C. W. Forsberg, “A Water-Level Initiated Decay Energy Cooling System,” Nucl.
Technol. 96, 229–235 (November 1991).
• C. W. Forsberg and A. M. Weinberg, “Advanced Reactors, Passive Safety, and the
Acceptance of Nuclear Energy,” Annual Rev. of Energy, 15, 133–152, 1990.
• C. W. Forsberg, et al., Proposed and Existing Passive and Inherent Safety-Related
Structures, Systems, and Components (Building Blocks) for Advanced Light-Water
Reactors, ORNL-6554, Oak Ridge National Laboratory, Oak Ridge, Tennessee,
December 1989.
• C. W. Forsberg, “Passive Emergency Cooling Systems for Boiling Water Reactors
(PECOS-BWR),” Nucl. Technol. 76, 185, January 1987.
Ehud Greenspan
UNIVERSITY OF CALIFORNIA AT BERKELEY
Department of Nuclear Engineering
Berkeley, CA 94720-1730
Phone: (510) 643-9983; Fax: (510) 643-9685; E-mail: gehud@nuc.berkeley.edu
Education:
1957-1961 B.Sc in Mechanical Eng. + Nuclear Option (Cum Laude), Technion -
Israel Institute of Technology.
1961-1963 M.Sc in Nuclear Science & Eng., Technion, Israel. "Optimization of the
Nuclear Design of a 125 MWe Heavy-Water Natural Uranium
Power Reactor."
1963-1966 Ph.D in Nuclear Science & Eng., Cornell University, Ithaca, N.Y., USA.
"Theory and Measurement of Neutron Importance in Nuclear Reactors.”
Relevant Experience:
Ehud Greenspan is a full-time faculty member at UC Berkeley since 1992. He
teaches reactor theory and reactor design & analysis courses. Prior to joining UC
Berkeley he was an Associate Director for Research and Development at the Nuclear
Engineering and Applications Division of the Israeli Atomic Energy Commission. He
has extensive and broad research experience in advanced nuclear reactors and nuclear
fuel-cycle conception and analysis. He was the PI on dozens of advanced nuclear systems
conception and analysis. Among these are studies of molten-salt reactors - see
publications No. 8, 9,20,22 and 25. He has more than 350 publications a sample of which
follows:
1. Ehud Greenspan, "Optimization of the Nuclear Design of a 125 MWe Heavy-Water
Natural Uranium Power Reactor," M.Sc. Thesis, Israel Institute of Technology, 1963.
2. E. Greenspan, K. B. Cady and J.P. Howe, "Economic Potential of Variable
Enrichment Fuel Elements for Power Reactors," Trans. American Nuclear Society, 9,
295-296 (1966).
3. E. Greenspan, "Energy Dependent Fine Structure Effects on the Reactivity Worth of
Resonances," Proc. Advanced Reactors; Physics, Design and Economics, (J. E.
Kallfeltz & R.A. Karam, Ed.) Pergamon Press, pp. 196-205, 1975.
4. E. Greenspan, A. Schneider, D. Gilai and P. Levin, "Natural-Uranium Light-Water
Breeding Hybrid Reactors," Proc. 2nd Topical Meeting on the Technology of
Controlled Nuclear Fusion, CONF-760935-P3, pp. 1061-1072 (1976).
5. E. Greenspan, A. Schneider and A. Misolovin, "On the Feasibility of Plutonium
Separation-Free Nuclear Power Economy with LWHRs," Trans. Am. Nucl. Society,
26, 305-306, (1977).
6. E. Greenspan, A. Schneider and A. Misolovin, "The Physics and Applications of
Subcritical Light Water U-Pu Lattices," Proc. Topical Mtg. on Advances in Reactor
Physics, Gatlinburg, TN., CONF 780401, pp. 411-422 (1978).
7. E. Greenspan and Y. Karni, "Spectral Fine Structure Effects on Material and Doppler
Reactivity Worth," Nuclear Sci. and Eng., 69, 169-190 (1979).
8. J. Hughes, I. Soares, E. Greenspan, W.F. Miller and Z. Shayer, "Molten Salt Critical
Reactors for the Transmutation of Transuranics and Fission Products," Proc. of the
GLOBAL'93 International Conference, Seattle, WA, Sept. 12-17, 1993. pp. 644-651.
9. Z. Shayer, J. Hughes, I. Soares, E. Greenspan and W. Miller, "Modifying SCALE-
4.1 for Transmutation Calculations," Trans. Israeli Nuc. Soc., 18, VII 27-VII 30
(1994).
10. E. Greenspan, “BWR Fuel Assembly Having Oxide and Hydride Fuel,” US Patent
No. 5,349,618, Sept. 20, 1994.
11. J. Vujic, E. Greenspan, S. Slater, T. Postam, L. Leal, Greg Casher and I. Soares,
"Development of Coupled SCALE 4.2/GTRAN2 Computational Capability for
Advanced MOX Fueled Assembly Designs," Proc. Int. Conf. on Math. &
Computations Reactor Physics and Environmental Analysis. Portland, OR. April 30-
May 4, 1995. pp. 1001-1010.
12. N.E. Brown, J. Hassberger, E. Greenspan and E. Elias, “Proliferation Resistant
Fission Energy Systems,” Proc. Global’97: International Conf. On Future Nuclear
Systems, Yokohama, Japan, October 5-10, 1997. pp. 879-884.
13. T.H. Kim, N.Z. Cho. And E. Greenspan, “Fuel-Self-Sufficient Heavy-Water Lattices
for Proliferation Resistant Multiple Fuel Recycling,” Trans. Am. Nucl. Soc., 77, 108-
110 (1997).
14. E. Greenspan, E. Elias, W.E. Kastenberg and N.W. Brown, “Compact Long Fuel-
Life Reactors for Developing Countries,” Proc. 9th International Conference on
Emerging Nuclear Energy Systems, ICENES’98, Herzlia, Israel, June 28 - July 2,
1998. pp. 74-83.
15. E. Greenspan, W.E. Kastenberg, N.Z. Cho, T.H. Kim and S.G. Hong, “Multi-
Recycling of Spent Fuel with Low Proliferation Risk” Proc. 9th International
Conference on Emerging Nuclear Energy Systems, ICENES’98, Herzlia, Israel, June
28 - July 2, 1998. pp. 455-464.
16. N.Z. Cho, S.G. Hong, T.H. Kim, E. Greenspan and W.E. Kastenberg, “Fuel Self-
Sufficient and Low Proliferation Risk Multi-Recycling of Spent Fuel,” Proc. 13th
KAIF/KNS Annual Conf., Seoul, Korea, April 1998. pp. 417-425.
17. A.S. Boloori, M. Frank, E. Greenspan, E. Hill, D.M. Hutchinson, S. Jones, X.
Mahini, M. Nichol, B. H. Park, H. Shimada, N. Stone and S. Wang, “Once-for-Life
Fueled, Highly-Modular, Simple, Super-Safe, Pb-Cooled Reactors,” Proc.
GLOBAL’99: International Conf. On Future Nuclear Systems, Jackson Hole,
Wyoming, Aug. 30-Sept. 2, 1999.
18. E. Greenspan, H. Shimada, D.C. Wade, M.D. Carelli, L. Conway, N.W. Brown and
Q. Hossain, “The Encapsulated Nuclear Heat Source Reactor Concept,” Proc. 8th Int.
Conf. On Nuclear Engineering, Paper ICONE-8750 Baltimore, MD, April 2-6, 2000.
19. E. Greenspan, H. Shimada and K. Wang, “Long-Life Cores with Small Burnup
Reactivity Swing,” Proc. of the 2000 Int. Topical Mtg. On Advances in Reactor
Physics and Math. and Computation into the Next Millennium, PHYSOR2000,
Pittsburgh, PA, May 7-11, 2000.
20. M.D. Lowenthal And E. Greenspan, “Parametric Studies For Optimization of a
Graphite-Moderated Molten Salt Transmuter,” Proc. of GLOBAL-2001, Paris,
France, September 2001. 8 pages.
21. E. Greenspan, N. W. Brown, M. D. Carelli, L. Conway, M. Dzodzo, Q. Hossain, D.
Saphier, J. J. Sienicki, D. C. Wade, “The Encapsulated Nuclear Heat Source Reactor
for Proliferation-Resistant Nuclear Energy,” Proc. of GLOBAL-2001, Paris, France,
September 2001. 8 pages.
22. E. Greenspan, M. Lowenthal, D. Barnes, D. Kawasaki, D. Kimball, H. Matsumoto,
H. Sagara and E. R. Vietez, “Transmutation Capability of a Once-Through Molten-
Salt and Other Transmuting Reactors,” Proc. of Topical Mtg. on Accelerator
Applications, Reno, NV, Nov. 2001. 10 pages.
23. B. Petrovic, M. Carelli, E. Greenspan, M. Milosevic, J. Vujic, E. Padovani and F.
Ganda, “First Core and Refueling Options for IRIS,” Submitted to International
Conference on Nuclear Engineering, ICONE-10, Arlington, VA, April 14-18, 2002.
24. E. Greenspan and the ENHS project team, “ The Long-Life Core Encapsulated
Nuclear Heat Source Generation IV Reactor,” Submitted to the International
Congress on Advanced Nuclear Power Plant, ICAPP, Holywood, FL, June 9-13,
2002.
25. E. Rodriguez-Vieitez, M. D. Lowenthal, E. Greenspan, and J. Ahn, “Optimization of
a Molten-Salt Transmuting Reactor,” to be presented in International Conference on
the New Frontiers of Nuclear Technology : Reactor Physics, Safety and High-
Performance Computing, PHYSOR-2002, Seoul, Korea, October 7-10, 2002.
Joonhong Ahn
Associate Professor
Department of Nuclear Engineering
University of California, Berkeley
Berkeley, CA 94720-1730
ahn@nuc.berkeley.edu
phone: 510-642-5107; fax: 510-643-9685
B.S. Nuclear Engineering, University of Tokyo, 1981
M.S. Nuclear Engineering, University of Tokyo, 1983
Ph.D. Nuclear Engineering, University of California, Berkeley, 1988
D.Eng. Nuclear Engineering, University of Tokyo, 1989
Joonhong Ahn won Junior Scientist Fellowship from the Japan Society for the
Promotion of Science (JSPS) (1988-90). Dr. Ahn joined the faculty in the Department of
Nuclear Engineering, University of Tokyo as an Assistant Professor (1990-1993). He
moved to Department of Nuclear Engineering, Tokai University (1993-1995). In January
1995, he joined the faculty at Berkeley.
His research interests include the performance assessment of deep geological
disposal systems for high-level radioactive wastes (HLW), especially analyses of mass
transport through engineered barriers and the natural geological barrier. He is also
interested in future of nuclear energy in Asia/Pacific region.
Professor Ahn served as a member of the Planning Committee for the Atomic
Energy Society of Japan (1992-1995). He served as the Editor for Radioactive Waste
Research (1994-1996), a journal of the Division of Radioactive Waste Management,
Atomic Energy Society of Japan. The journal was established when he was serving as
the Secretary General for the Division (1993-1995). He serves as a member of the
committees of Technical Journals and Book Publishing of the American Nuclear Society
(since June, 2001)
Selected Publications:
• Ahn, J., Y. Furuhama, Y. Li, and A. Suzuki, Analysis of Radionuclide Transport
Through Fracture Networks by Percolation Theory, Journal of Nuclear Science and
Technology, 28(5), 433––446, 1991.
• Ahn, J. and S. Nakayama, Modeling for Migration of A Redox-Sensitive
Radionuclide in Engineered Barriers, Nucelar Technology, 97(3), 323––335, 1992.
• Ahn, J., and A. Suzuki, Diffusion of the 241Am––> 237Np Decay Chain Limited by
Their Elemental Solubilities in Artificial Barriers of High-Level Radioactive Waste
Repositories, Nuclear Technology, 101(1), 79––91, 1993.
• Ahn, J., S. Nagasaki, S. Tanaka, and A. Suzuki, Effects of Smectite Illitization of
Transport of Actinides Through Engineered Barriers of HLW Repositories, 18th
International Symposium on the Scientific Basis for Nuclear Waste Management,
Materials Research Society, Atomic Energy Society of Japan, October 1994, Kyoto,
Japan, 1995.
• Ahn, J., Transport of Weapons-Grade Plutonium and Boron Through Fractured
Geologic Media, Nuclear Tchnology, 117(3), 316-328), 1997.
• Ahn, J., Integrated Radionuclide Transport Model for an HLW Repository in Water-
Saturated Geologic Formations, Nuclear Technology, 121(1), 24-39, 1998.
• Ahn, J., Criticality Safety Assessment for a Conceptual High-Level-Waste Repository
in Water-Saturated Geologic Media, Nuclear Technology, 126, 303-318, 1999.
• Ahn, J., Preliminary Assessment of the Effects of ATW System Application on YM
Repository Performance, Global ‘99, International Conference on Future Nuclear
Systems, August 29-September 3, 1999, Jackson Hole, Wyoming (1999).
• Ahn, J., E. Greenspan, and P. L. Chambré, A Preliminary Consideration for
Underground Autocatalytic Criticality by Vitrified High-Level Waste in Water-
Saturated Geologic Repository, Journal of Nuclear Science and Technology, 37(5),
465-476, 2000.
• Ahn, J., and P. L. Chambré, Alternative Measure for Performance of HLW Geologic
Repository, Global 2001, International Conference on "Back-end of the fuel cycle:
from research to solutions," Paris, France, 9/9-9/13 (2001).
KAZUO FURUKAWA
Thorium Molten-Salt International Forum
Education
Faculty of science, University of Kyoto B.S.—1951, Chemistry
Institute for Iron, Steel and other Metals, Tohoku University
D. Sc. —1960, Inorganic Liquid Structural Chemistry
Professional Activities [cf. “Five Hundred Leaders of Influence”(1999), American
Biographical Institute]
A visiting fellow at the University of London, Birkbeck College in 1960, he then
took the post of associate professor at Tohoku University. He worked at Japan Atomic
Energy Research Institute from 1962 to 1983, when he became a professor at Tokai
University. He founded the Japanese liquid sodium technology for FBR at JAERIbetween 1962 and 1970, and examined several fluid-fuel reactor concepts relating with
fission, fusion and spallation. In 1980 he invented the Accelerator Molten-Salt Breeder
(AMSB), aiming fissile material breeding, and in 1985 he proposed the simplified small
Molten-Salt Fission Power Station named FUJI, which is fuel self-sustainable without
continuous chemical processing.
Dr. Furukawa was elected as a foreign member of the Ukrainian Academy of
Science in 1991. He was invited by the Elec. de France, President Mr. Bergougnoux in
1987, who decided “no Superphenix No.2”. In 1992 he met with Dr. Allan Bromley,
President Advisor of Science and Technology, who encouraged him during the
development of the THORIMS-NES (Thorium Molten-Salt Nuclear Energy Synergetics)
project, and also gained the support of the following President Advisor of S&T, Dr. John
H. Gibbons, in 1997. Dr. Furukawa’s work on the Thorium Molten-Salt Reactor was
examined by President Clinton in 1997 and was also presented at the International
Conference on Molten-Salt Development at RAND in Santa Monica, California, where
he submitted a paper entitled “Conclusions and Recommendations” with 24 peoples from
Japan, Russia, Belarus, France, Czech, Turkey and the U.S.A. Three Agency Study
(OECD/NEA, OECD/IEA, IAEA) is recommending his work for international
developmental item.
Selected Publications (Total list is more than 150 articles and reports)
• K. Furukawa, H. Ohno: ”Molten LiF-BeF2 (Flibe) System”, DATA-BOOKS FOR
MOLTEN MATERIALS I, Japan Nuclear-Energy Information Center, pp.114
(1980)
• K. Furukawa, Y, Kato, T. Ohmichi, H. Ohno:”The Combined System of Accelerator
Molten-Salt Breeder(AMSB) and Molten-Salt Converter Reactor(MSCR)”[Japan-US
Seminar on "Th Fuel Reactors"(Oct.1982,Nara)],”Thorium Fuel
Reactors",(1985)P.271-281, Atom.Ene.Soc.Japan.
• [Russian Trans.:"Atomnaja Texnika za Rubezom", 1983 [6]P.23-29 (1983)]
• K.Furukawa:”Symbiotic Molten-Salt Systems coupled with Accelerator Molten-Salt
Breeder(AMSB) or Inertial-Confined Fusion Hybrid Molten-Salt Breeder(IHMSB)
and their Comparison,”[3rd_ICENES, Helsinki,1983] Atomkernenergie/Kerntechnik,
44 [1] P.42-45 (1984)
• K. Furukawa, K. Minami, T. Oosawa, M. Ohta, N. Nakamura, K. Mitachi, Y. Kato:”
Design Study of Small Molten-Salt Fission Power Station suitable for Coupling with
Accelerator Molten-Salt Breeders”, [4th:ICENES, Madrid, 1986] "Emerging Nuclear
Energy Systems" P.235-239 (1987) World Sci.
• K. Furukawa, K. Minami, K. Mitachi, & Y. Kato:”Compact Molten-Salt Fission
Power Stations (FUJI-series) and their Developmental Program”, Proc.Joint
Int.Sympo.Molten-Salts, Proc.Vol. 87-7 P.896-905(1987) Electrochem.Soc.
• K. Furukawa, K. Mitachi, K. Minami, & Y. Kato:”High-Safety and Economical
Small Molten- Salt Fission Power Stations and their Develop.Program [Th Molten-
Salt Nuclear Energy Synergetics: THORIMS-NES]”, 8th Miami
Int.Conf.Alter.Ene.Sources:(Decem.14-16,1987)"Alternative Energy Sources VIII",
Vol.2, P.3-22 (1989)Hemisph.Pub.
• K. Furukawa & A. Lecocq :”Preliminary Examination on The Next Generation
Nuclear Reactors in Comparison with the Small Thorium Molten-Salt Reactor”,
sponsored by Electricite de France (EdF), France [Tokai University
(xiii+100pp)](Dec.,1988)
• K. Furukawa, A. Lecocq, Y. Kato & K. Mitachi :”Summary Report : Thorium
Molten-Salt Nuclear Energy Synergetics”
, J.Nucl. Sci. & Tech.,Vol.27, No.12,
P.1157-1178 (1990)
• K. Furukawa, K. Mitachi, Y. Kato & A. Lecocq :”Global Nuclear Energy System --
Thorium Molten-Salt Nuclear Energy Synergetics—“,Proc.Indo-Japan Seminar on
"Thorium Utilization" (Bombay, India, Dec. 10-13, 1990) P.21-28(1991)
• K. Furukawa, A. Lecocq, Y. Kato, & K. Mitachi :”Radiowaste Management in
Global Application of Thorium Molten-Salt Nuclear Energy Synergetics with
Accelerator Breeders”LA-12205-C Conf. SKN Report No.54 ; UC-940[November
1991] P.686-697.
• K. Furukawa, K. Mitachi & Y. Kato :”Small Molten-Salt Reactors with a Rational
Thorium Fuel- Cycle”,Nuclear Engineering and Design, 139, P.157-165 (1992)
• Y.Kato, K.Furukawa, K.Mitachi & S.E. Chigrinov:”Fuel Trajectory in Acceretator
Molten-Salt Breeding Power Reactor System including Pu Burning”,"Emerging
Nuclear Energy Systems ICENES'93", Ed.H.Yasuda, (1994) P.439-443, World
Sci.Pub.
• K.Furukawa: [COMPILED BOOKS] “Important Papers concerning on Thorium
Molten-Salt Nuclear Energy Synergetics[THORIMS-NES]”, Vol.I (Oct.,1994),
pp.272; Vol.II(Oct.,1994), pp.279. Inst.of Research & Development, Tokai
University,
• K.Furukawa, K.Kato & S.E.Chigrinov:”Plutonium (TRU) Transmutation and 233U
Production by Single-Fluid type Accelerator Molten-Salt Breeder (AMSB)”.AIP
CONFERENCE PROCEEDINGS 346 (1995)p.745-751, Am.Inst.Physics.
• K.Furukawa, K.Mitachi, Y.Kato, S.E.Chigrinov, A.Lecocq & L.B.Erbay :”Rational
Pu-Dispo- sition for 233U-Production by THORIMS-NES” : IAEA-TECDOC-840,
P.169-181(1995)
• K.Mitachi, Y.Yamana, T.Suzuki, & K.Furukawa :”Neutronic Exami. on Plutonium
Transmutation by a Small Molten-Salt Fission Power Station” : IAEA-TECDOC-840,
P.183-195(1995)
• K.Mitachi, K.Furukawa, T.Suzuki & A.Namekata:”Pu Burning Molten-Salt Power
Station (FUJI- Pu3) for Prevailing Th Nucl.Industry”,ICENES’96:Vol.I(1997)P.180-
190, Inst.Phys.Power Eng.
• S.Chigrinov, A.Kievitskaia, C.Rutkovskaia, I.Rakhno, K.Furukawa & A.Lecocq
:”Accelerator Molten-Salt Breeder as Fissile Producing Component of THORIMS-
NES Concept for Energy Produc.and Transmut.of Plutonium”.ICENES’96:
Vol.II(1997) P.564-571, Inst.Phys.Power Eng.
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10. Facilities and Resources
No special facilities and physical resources are required for the proposed work.
However, immediate access to the staffs of each of the participating organizations are
very valuable strengths of the proposal, as each is a leader in the fields of nuclear
engineering and nuclear weapons.
11. Budget
Omitted, can be found in the PDF text.
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Posts
3 comments:
charles,
I haven't studied the entire document. Can you point out where the supposed examples are of cherry picking?
thx
Basically I quoted the first sentence of 2.1, 1.2, 2.3, and 2.4. I was accused of cherry picking because I pointed to a limited number of sentences in the text.
Yes, according to the transcript of the Google talk, there is the fact that the three alphas produced in p B11 reaction are not equal: the first one has 3.76 MeV, and the other two a median energy of 2.46 MeV. If we want to harness the full energy of the first one, we'll need a grid charged to 1.88 MV.
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