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PostPosted: Feb 14, 2007 11:07 pm 
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VIII. CONCLUSIONS

The Molten Salt Breeder Reactor, if successfully developed and marketed, could provide a useful supplement to the currently developing uranium-plutonium reactor economy. This concept offers the potential for:
  • Breeding in a thermal spectrum reactor;
  • Efficient use of thorium as a fertile material;
  • Elimination of fuel fabrication and spent fuel shipping;
  • High thermal efficiencies.
Notwithstanding these attractive features, this assessment has reconfirmed the existence of major technological and engineering problems affecting feasibility of the concept as a reliable and economic breeder for the utility industry. The principal concerns include uncertainties with materials, with methods of controlling tritium, and with the design of components and systems along with their special handling, inspection and maintenance equipment. Many of these problems are compounded by the use of a fluid fuel in which fission products and delayed neutrons are distributed throughout the primary reactor and reprocessing systems.

The resolution of the problems of the MSBR will require the conduct of an intensive research and development program. Included among the major efforts that would have to be accomplished are:
  • Proof testing of an integrated reprocessing system;
  • Development of a suitable containment material;
  • Development of a satisfactory method for the control and retention of tritium;
  • Attainment of a thorough understanding of the behavior of fission products in a molten-salt system;
  • Development of long-life moderator graphite, suitable for breeder application;
  • Conceptual definition of the engineering features of the many components and systems;
  • Development of adequate methods and equipment for remote inspection, handling, and maintenance of the plant.
The major problems associated with the MSBR are rather difficult in nature and many are unique to this concept. Continuing support of the research and development effort will be required to obtain satisfactory solutions to the problems. When significant evidence is available that demonstrates realistic solutions are practical, a further assessment could then be made as to the advisability of advancing into the detailed design and engineering phase of the development process including that of industrial involvement. Proceeding with this next step would also be contingent upon obtaining a firm demonstration of interest and commitment to the concept by the power industry and the utilities and reasonable assurances that large-scale government and industrial resources can be made available on a continuing basis to this program in light of other commitments to the commercial nuclear power program and higher priority energy development efforts.


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IX. REFERENCES
  1. US Atomic Energy Commission, “The 1967 Supplement to the 1962 Report to the President on Civilian Nuclear Power” USAEC Report, February 1967.
  2. US Atomic Energy Commission, “The Use of Thorium in Nuclear Power Reactors” USAEC Report WASH-1097, 1969.
  3. US Atomic Energy Commission, “Potential Nuclear Power Growth Patterns,” USAEC Report WASH-1098, December 1970.
  4. US Atomic Energy Commission, “Cost-Benefit Analysis of the US Breeder Reactor Program,” USAEC Report WASH-1126, 1969.
  5. US Atomic Energy Commission, “Updated (1970) Cost-Benefit Analysis of the US Breeder Reactor Program,” USAEC Report WASH-1184, January 1972.
  6. Edison Electric Institute, “Report on the EEI Reactor Assessment Panel,” EEI Publication No. 70-30, 1970.
  7. Annual Hearings on Reactor Development Program, US Atomic Energy Commission FY 1972 Authorizing Legislation, Hearings before the Joint Committee on Atomic Energy, Congress of the United States p. 820-830, US Government Printing Office
  8. Nuclear Applications and Technology, Volume 8, February 1970.
  9. Robertson, R. D. (ed) “Conceptual Design Study of a Single-Fluid Molten Salt Breeder Reactor,” ORNL-4541, June 1971.
  10. Rosenthal, M. W., et al.; “Advances in the Development of Molten-Salt Breeder Reactors,” A/CONF-49/P-048, Fourth United Nations International Conference on the Peaceful Uses of Atomic Energy, Geneva, September 6 - 16, 1971.
  11. Trinko, J. R. (ed.), “Molten-Salt Reactor Technology,” Technical Report of the Molten-Salt Group, Part I, December 1971.
  12. Trinko, J. R. (ed), “Evaluation of a 1000 MWe Molten-Salt Breeder Reactor," Technical Report of the Molten Salt Group, Part II, November 1971.
  13. Ebasco Services Inc., “1000 MWe Molten-Salt Breeder Reactor Conceptual Design Study,” Final Report Task I, Prepared under ORNL subcontract 3560, February 1972.
  14. “Project for Investigation of Molten-Salt Breeder Reactor,” Final Report, Phase I Study for Molten Salt Breeder Reactor Associates, September 1970.
  15. Cardwell, D. W. and Haubenreich, P. N., “Indexed Abstracts of Selected References on Molten-Salt Reactor Technology,” ORNL-TM-3595, December 1971.
  16. Kasten, P. R., Bettis, E. S. and Robertson, R. C., “Design Studies of 1000 MWe Molten-Salt Breeder Reactors,” ORNL-3996, August 1966.
  17. Molten Salt Reactor Program Semiannual Reports beginning in February 1962.


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PostPosted: Feb 24, 2008 2:48 pm 
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Wow--Charles Barton,on his blog "Nuclear Green", has given WASH-1222 his usual excellent treatment with historical background and found it severely wanting.

It's really impressive how Charles was able to tie together his first-hand knowledge of Alvin Weinberg's life and circumstances, along with his research on Milton Shaw's background and tenure at the AEC, to put the production of WASH-1222 in context.

Please read!

WASH-1222 with Comments: Part 1

WASH-1222 with Comments: Part 2

WASH-1222 with Comments: Part 3

WASH-1222 with Comments: Part 4

WASH-1222 with Comments: Part 5

WASH-1222 with Comments: Part 6

Some concluding remarks about WASH-1222


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PostPosted: Feb 24, 2008 10:55 pm 
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Kirk asked me to cross post my annotated WASH-1222. I am of course happy to comply. - Charles Barton

Kirk Sorensen, in response to my posting on Milton Shaw, suggested that I review WASH-1222, a document by which Shaw hopped to bury the MSR. I have begun to do that. The purpose of this review will be to assess the extent to which the decision to shut down the development of the MSR in the late 1960's and early 1970's was an irrational political decision. I am going to post my results on section at a time.

Quote:
AN EVALUATION OF THE MOLTEN SALT BREEDER REACTOR

I. INTRODUCTION


The Division of Reactor Development and Technology, USAEC, was assigned the responsibility of assessing the status of the technology of the Molten Salt Breeder Reactor (MSBR) as part of the Federal Council of Science and Technology Research and Development Goals Study. In conducting this review, the attractive features and problem areas associated with the concept have been examined; but more importantly, the assessment has been directed to provide a view of the technology and engineering development efforts and the associated government and industrial commitments which would be required to develop the MSBR into a safe, reliable and economic power source for central station application.

The MSBR concept, currently under study at the Oak Ridge National Laboratory (ORNL), is based on use of a circulating fluid fuel reactor coupled with on-line continuous fuel processing. As presently envisioned, it would operate as a thermal spectrum reactor system utilizing a thorium-uranium fuel cycle. Thus, the concept would offer the potential for broadened utilization of the nation's natural resources through operation of a breeder system employing another fertile material (thorium instead of uranium).

The long-term objective of any new reactor concept and the incentive for the government to support its development are to help provide a self-sustaining, competitive industrial capability for producing economical power in a reliable and safe manner. A basic part of achievement of this objective is to gain public acceptance of a new form of power production. Success in such an endeavor is required to permit the utilities and others to consider the concept as a viable option for generating electrical power in the future and to consider making the heavy, long-term commitments of resources in funds, facilities and personnel needed to provide the transition from the early experimental facilities and demonstration plants to full-scale commercial reactor power plant systems.

Consistent with the policy established for all power reactor development programs, the MSBR would require the successful accomplishment of three basic research and development phases:
  • An initial research and development phase in which the basic technical aspects of the MSBR concept are confirmed, involving exploratory development, laboratory experiment, and conceptual engineering.
  • A second phase in which the engineering and manufacturing capabilities are developed. This includes the conduct of in-depth engineering and proof testing of first-of-a-kind components, equipment and systems. These would then be incorporated into experimental installations and supporting test facilities to assure adequate understanding of design and performance characteristics, as well as to gain overall experience associated with major operational, economic and environmental parameters. As these research and development efforts progress, the technological uncertainties would need to be resolved and decision points reached that would permit development to proceed with necessary confidence. When the technology is sufficiently developed and confidence in the system was attained, the next stage would be the construction of large demonstration plants.
  • A third phase in which the utilities make large-scale commitments to electric generating plants by developing the capability to manage the design, construction, test and operation of these power plants in a safe, reliable, economic, and environmentally acceptable manner.
Significant experience with the Light Water Reactor (LWR), the High-Temperature Gas-cooled Reactor (HTGR) and the Liquid Metal-cooled Fast Breeder Reactor (LMFBR) has been gained over the past two decades pertaining to the efforts that are required to develop and advance nuclear reactors to the point of public and commercial acceptance. This experience has clearly demonstrated that the phases of development and demonstration should be similar regardless of the energy concept being explored; that the logical progression through each of the phases is essential; and that completing the work through the three phases is an extremely difficult, time consuming and costly undertaking, requiring the highest level of technical management, professional competence and organizational skills. This has again been demonstrated by the recent experience in the expanding LWR design, construction and licensing activities which emphasize clearly the need for even stronger technology and engineering efforts than were initially provided, although these were satisfactory in many cases for the first experiments and demonstration plants. The LMFBR program, which is relatively well advanced in its development, tracks closely this LWR experience and has further reinforced this need as it applies to the technology, development and engineering application areas.


This paragraph reflects Milton Shaw's views, but Shaw clearly over-estimated the relative maturity of the reactor technologies referred to by the paragraph. Developmental problems with light-water reactor reactor technology were to cost reactor owners tens of billions of dollars during the next two decades. Reactor scientists had told Shaw about the problems, but Shaw discounted the warnings. Again Shaw's belief that the LMFBR had reached an advanced stage of development was far from reality in 1972, and remains questionable in 2008. Shaw's demonstrably mistaken beliefs thus appear to lie at the heart of the WASH-1222 assessment of the potential of MSR technology.

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It should also be kept in mind that the large backlog of commitments and the shortage of qualified engineering and technical management personnel and proof-test facilities in the government, in industry and in the utilities make it even more necessary that all the reactor systems be thoroughly designed and tested before additional significant commitment to and construction of, commercial power plants are initiated.


In fact this was not the case when Shaw joined the AEC in 1964. Shaw immediately proceeded to destroy the the research and development units that were needed to carry out such a project. Hence "the shortage of qualified engineering and technical management personnel and proof-test facilities" was a problem which Shaw had created. Thus as we shall see, not only does WASH-1222 commit egregious errors in logic, as well as misstatement of facts, it covers up the fact that Shaw himself had destroyed the resources that were required to complete the development of the MSR. Statements like this must be counted as duplicitous.

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With regard to the MSBR, preliminary reactor designs were evaluated in WASH-1097 (“The Use of Thorium in Nuclear Power Reactors”) based upon the information supplied by ORNL. Two reactor design concepts were considered—a two-fluid reactor in which the fissile and fertile salts were separated by graphite and a single fluid concept in which the fissile and fertile salts were completely mixed. This evaluation identified problem areas requiring resolution through conduct of an intensive research and development program.


The two-fluid MSR was an auto-breeder. That meant it produced at least enough U233 to keep working until it ran out of thorium to breed. As long as a reactor produces as much fuel as it consumes, it is a successful breeder. Thus WASH-1222 should have considered the advantages of 2 fluid MSRs.

Quote:
Since the publication of WASH-1097, all efforts related to the two-fluid system have been discontinued because of mechanical design problems and the development of processes which would, if developed into engineering systems, permit the on-line reprocessing of fuel from single fluid reactors. At present, the MSBR concept is essentially in the initial research and development phase, with emphasis on the development of basic MSBR technology. The technology program is centered at ORNL where essentially all research and development on molten salt reactors has been performed to date. The program is currently funded at a level of $5 million per year. Expenditures to date on molten salt reactor technology both for military and civilian power applications have amounted to approximately $150 million of which approximately $70 million has been in support of central station power plants. These efforts date back to the 1940's.


ORNL chose a one-fluid approach, because Shaw demanded a higher breeding ratio than the two-fluid approach could achieve in order to bump up theoretical breeding ratios. This choice was made to meet Shaw's demands. These sums of $150 million and $70 million seem quite paltry by the standards of 2008. Even if dollars from the 1950’s and 1960’s are translated into 2008 terms, the amount spent seems trivial in comparison to say the cost of military weapons systems. In retrospective we can say that ORNL provided a whole lot of information about a promising technology very inexpensively.

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In considering the MSBR for central station power plant application, it is noted that this concept has several unique and desirable features; at the same time, it is characterized by both complex technological and practical engineering problems which are specific to fluid-fueled reactors and for which solutions have not been developed. Thus, this concept introduced major concerns that are different in kind and magnitude from those commonly associated with solid fuel breeder reactors. The development of satisfactory experimental units and further consideration of this concept for use as a commercial power plant will require resolution of these as well as other problems which are common to all reactor concepts.


This paragraph shifts from obvious facts, to unwarranted conclusions. The facts involve "complex technological," and "practical engineering problems" for which "solutions have not been developed." Now had solutions been developed already, then there would be no purpose for the development program which has been proposed. The next statement does not follow from the stated issues. "Thus, this concept introduced major concerns that are different in kind and magnitude from those commonly associated with solid fuel breeder reactors." Why are concerns about the developmental problems of the MSR different in kind and magnitude? Given what we know today, the AEC had not only under-estimated the problems associated with the development of the LMFBR, they had seriously underestimated the developmental problems associated with the LWR, a technology which Shaw and the AEC in 1972 incorrectly believed to be mature.

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As part of the AEC's Systems Analysis Task Force (AEC report WASH-1098) and the "Cost-Benefit Analysis of the U.S. Breeder Reactor Program" (AEC reports WASH-1126 and WASH-1184), studies were conducted on the cost and benefit of developing another breeder system, "parallel" to the LMFBR. The consistent conclusion reached in these studies is that sufficient information is available to indicate that the projected benefits from the LMFBR program can support a parallel breeder program. However, these results are highly sensitive to the assumptions on plant capital costs with the recognition, even among concepts in which ample experience exists, that capital costs and especially small estimated differences in costs are highly speculative for plants to be built 15 or 20 years from now. Therefore, it is questionable whether analyses based upon such costs should constitute a major basis for making decisions relative to the desirability of a parallel breeder effort. Experience in reactor development programs in this country and abroad has demonstrated that different organizations, in evaluating the projected costs of introducing a reactor development program and carrying it forward to the point of large-scale commercial utilization, would arrive at different estimates of the methods, scope of development and engineering efforts, and the costs and time required to bring that program to a stage of successful large scale application and public acceptance.


The statement of risk considerations in the proceeding paragraph is sound. Future cost estimates for projects in developmental stages represent risky conjectures. This would seem to be an argument for rather than against parallel programs. Given the cost uncertainties attendant to taking a single-line approach to a technological development, it is always wise to have an alternative solution at hand, in case cost start to run away. Developmental costs for the LMFBR "Clinch River Breeder Reactor" project did run away in the 1970’s and early 1980’s. Since all of the contentions about cost risk apply equally to both the LMFBR and the MSR, the argument in the last paragraph is incoherent. That is it supports contradictory conclusions. We are being set up by this paragraph for an attempt to block further development of the MSR on the basis of cost. WASH-1222 has already made the judgment that development of the LMFBR would proceed. It appears to have assessed that the AEC’s 1970’s LMFBR project could fail, as it did. The possibility of project failure is a risk. Any comparative cost/benefits study, should assess relative risks of failure.

Quote:
Based upon the AEC's experience with other complex reactor development programs, it is estimated that a total government investment up to about 2 billion dollars in undiscounted direct costs could be required to bring the molten salt breeder or any parallel breeder to fruition as a viable, commercial power reactor. A magnitude of funding up to this level could be needed to establish the necessary technology and engineering bases, obtain the required industrial capability, and advance through a series of test facilities, reactor experiments, and demonstration plants to a commercial MSBR, safe and suitable to serve as a major energy option for central station power generation in the utility environment.


Looking at this statement today with the benefit of hindsight, I would have to say that a development cost of $2 billion 1972 dollars was trivial. The Apollo Moon program cost $25.4 Billion 1969 dollars, arguably the nation would have been far better off if 10% of that money had been diverted to MSR development. In 1984 the GAO reviewed the Clinch River Breeder Reactor project, which was the AEC’s LMFBR project. In 1971 the AEC had estimated that the project would cost $400 million, of which $257 was to have come from private sources. By 1972, when WASH-1222 was written the cost estimate had risen to $700 million. By 1981 after $1 billion had been spent, the estimated cost of completion was 3 to 3.2 Billion more, with an estimated further project cost of $1 billion for a plutonium processing facility. By 1984 project cost had risen to $8 billion. And this was only a proof of concept reactor. Other proof LMFBR proof of concept reactors have had a very mixed history. Even today, a good case can be made that LMFBR technology has not been proven either safe, reliable or cost effective.


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PostPosted: Feb 24, 2008 10:56 pm 
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I have, in three previous posts discussed the effects that Milton Shaw's beliefs, managerial style, and policies had both on nuclear research at some and probably all National Laboratories. Shaw's ruthless methods of imposing his views even lead to the firing of Alvin Weinberg over disagreements about nuclear safety. In my last post, on the introduction to WASH-1222, I suggested that not only did Shaw have mistaken beliefs about the maturity of reactor technology, but that these beliefs cost reactor owners tens of billions of dollars. I may elaborate on this at a further time. I also argued that Shaw's mistakes about technology extended to breeder reactors, He held the mistaken belief that the maturity of LMFBR's had reached a level of maturity similar to that of LWRs.

In my last post, I began a review of a WASH-1222, a document prepared under Shaw's direction. In the introduction to the document I found evidence that Shaw's mistaken beliefs, coupled with the consequences of his own bureaucratic decisions for which he failed to acknowledge responsibility and shocking errors in logic, led Shaw to discount a promising new reactor technology, the Molten Salt Reactor. At the same time, Shaw was implicitly pushing other technologies, even though many the shared many of the problems of Molten Salt Reactor technology, while for other problems with technology Shaw favored, were solved by using the molten salt approach. Shaw claimed that unanticipated costs might incurred during the course of development. My intention at the moment is to to post the next three sections of WASH-1222. There are a few points that might require further comments, so I might add them tomorrow.

Quote:
II. SUMMARY

The MSBR concept is a thermal spectrum, fluid-fuel reactor which operates on the thorium-uranium fuel cycle and when coupled with on-line fuel processing, has the potential for breeding at a meaningful level. The marked differences in the concept as compared to solid-fueled reactors make the MSBR a distinctive alternate. Although the concept has attractive features, there are a number of difficult development problems that must be resolved; many of these are unique to the MSBR while others are pertinent to any complex reactor system.

The technical effort accomplished since the publication of WASH-1097 and WASH-1098 has identified and further defined the problem areas; however, this work has not advanced the program beyond the initial phase of research and development. Although progress has been made in several areas (e.g., reprocessing and improved graphite), new problems not addressed in WASH-1097 have arisen which could affect the practicality of designing and operating a MSBR. Examples of major uncertainties relate to materials of construction, methods for control of tritium, and the design of components and systems along with their special handling, inspection and maintenance equipment. Considerable research and development efforts are required in order to obtain the data necessary to resolve the uncertainties.

Assuming that practical solutions to these problems can be found, a further assessment would have to be made as to the advisability of proceeding to the next stage of the development program. In advancing to the next phase, it would be necessary to develop a greatly expanded industrial and utility participation and commitment along with a substantial increase in government support. Such broadened involvement would require an evaluation of the MSBR in terms of already existing commitments to other nuclear power and high priority energy development efforts.

III. RESOURCE UTILIZATION

It has long been recognized that the importance of nuclear fuels for power production depends initially on the utilization of the naturally occurring fissile 235U; but it is the more abundant fertile materials, 238U and 232Th, which will be the major source of nuclear power generated in the future. The basic physics characteristics of fissile plutonium produced from 238U offer the potential for high breeding gains in fast reactors, and the potential to expand greatly the utilization of uranium resources by making feasible the utilization of additional vast quantities of otherwise uneconomic low grade ore. In a similar manner, the basic physics characteristics of the thorium cycle will permit full utilization of the nation's thorium resources while at the same time offering the potential for breeding in thermal reactors.

The estimated thorium reserves are sufficient to supply the world's electric energy needs for many hundreds of years if the thorium is used in a high-gain breeder reactor. It is projected that if this quantity of thorium were used in a breeder reactor, approximately 1,000,000 quad (1 quad = 1 quadrillion Btu) would be realized from this fertile material. It is estimated that the uranium reserves would also supply 1,000,000 quads of energy if the uranium were used in LMFBRs. In contrast, only 20,000 quads would be available if thorium were used as the fertile material in an advanced converter reactor because the reactor would be dependent upon 235U availability for fissile inventory make-up. (Note: a conservative estimate is that between 20,000 and 30,000 quads will be used for electric power generation between now and the year 2100.)

IV. HISTORICAL DEVELOPMENT OF MOLTEN SALT REACTORS

The investigation of molten salt reactors began in the late 1940's as part of the U.S. Aircraft Nuclear Propulsion (ANP) Program. Subsequently, the Aircraft Reactor Experiment (ARE) was built at Oak Ridge and in 1954 it was operated successfully for nine days at power levels up to 2.5 MWt and fuel outlet temperatures up to 1580ºF (1133 K). The ARE fuel was a mixture of NaF, ZrF4, and UF4. The moderator was beryllium oxide and the piping and vessel were constructed of Inconel.

In 1956, ORNL began to study molten salt reactors for application as central station converters and breeders. These studies concluded that graphite moderated, thermal spectrum reactors operating on a thorium-uranium cycle were most attractive for economic power production. Based on the technology at that time, it was thought that a two-fluid reactor in which the fertile and fissile salts were kept separate was required in order to have a breeder system. The single-fluid reactor, while not a breeder, appeared simpler in design and also seemed to have the potential for low power costs.

Over the next few years, ORNL continued to study both the two-fluid and single-fluid concepts, and in 1960 the design of the single-fluid 8 MWt Molten Salt Reactor Experiment (MSRE) was begun. The MSRE was completed in 1965 and operated successfully during the period 1965-1969. The MSRE experience is treated in more detail in a later section.

Concurrent with the construction of the MSRE, ORNL performed research and development on means for processing molten salt fuels. In 1967 new discoveries were made which suggested that a single-fluid reactor could be combined with continuous on-line fuel processing to become a breeder system. Because of the mechanical design problems of the two-fluid concept and the laboratory-scale development of processes which would permit on-line reprocessing, it was determined that a shift in emphasis to the single-fluid breeder concept should be made; this system is being studied at the present.


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PostPosted: Feb 24, 2008 10:57 pm 
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In the first section of my exploration of WASH-1222, I pointed to what might be considered errors in logic or reasoning, in which argument went from true assumptions to false conclusions. But in the WASH-1222 discussion of the actual state of Molten Salt Reactor technology, we clearly encounter repeated a dishonest statements of facts. I will presently demonstrate the depth of this dishonesty.

Quote:
V. MOLTEN SALT BREEDER REACTOR CONCEPT DESCRIPTION

The breeding reactions of the thorium cycle are:

232Th +n --> 233Th --> 233 Pa --> 233U

Because of the number of neutrons produced per neutron absorbed and the small fast-fission bonus associated with 233U and 232Th in the thermal spectrum, a breeding ratio only slightly greater than unity is achievable. In order to realize breeding with the thorium cycle it is necessary to remove the bred 233Pa and the various nuclear poisons produced by the fission process from the high-flux region as quickly as possible. The Molten-Salt Breeder Reactor concept permits rapid removal of 233Pa and the nuclear poisons (e.g. 135Xe and the rare earth elements). The reactor is a fluid-fueled system containing UF4 and ThF4 dissolved in LiF - BeF2. The molten fuel salt flows through a graphite moderator where the nuclear reactions take place. A side stream is continuously processed to remove the Pa and rare earth elements, thereby permitting the achievement of a calculated breeding ratio of about 1.06.

The MSBR is attractive because of the following:
  1. Use of a fluid fuel and on-site processing would eliminate the problems of solid fuel fabrication and the handling and shipping and reprocessing of spent fuel elements which are associated with all other reactor types under active consideration.
  2. MSBR operation on the thorium-uranium fuel cycle would help conserve uranium and thorium resources by utilizing thorium reserves with high efficiency.
  3. The MSBR is projected to have attractive fuel cycle costs. The major uncertainty in the fuel cycle cost is associated with the continuous fuel processing plant which has not been developed.
  4. The safety issues associated with the MSBR are generally different from those of solid fuel reactors. Thus, there might be safety advantages for the MSBR when considering major accidents. An accurate assessment of MSBR safety is not possible today because of the early state of development.
  5. Like other advanced reactor systems such as the LMFBR and HTGR, the MSBR would employ modem steam technology for power generation with high thermal efficiencies. This would reduce the amount of waste heat to be discharged to the environment.


Statement 4 is particularly egregious because safety is one of the towering strengths of the MSR. By 1972 safety issues involving the MSR had been well thought out, and solutions identified. The characteristic features of the MSR, its molten core, and salt fluid, conferred enormous safety advantages on the MSR concept. Statement 4 simply overlooks what was known about the safety advantages of MSR technology in 1972.

To give an idea of the extent to which Statement 4 subverts the truth I will compare it with Eric H. Ottewitte sometime latter listing of the safety advantages of the MSR:

  1. Already being a molten fuel, further "meltdown" cannot occur
  2. Fluid fuel has inherently a strong negative temperature coefficient of reactivity due to expansion, greatly inhibiting boiling
  3. Elimination of pressurized and pressure-evolving components inside the containment
  4. Elimination of the possibility of gas and vapor evolution, especially the release of free hydrogen and attendant fire hazard
  5. Reduced risk of radioactivity release outside the containment due to
    1. reduced risk of failure of the containment, and
    2. two orders of magnitude reduction in the FP decay heat source relative to conventional solid-fuel reactors, due to continuous on-site chemical processing
  6. Reduced FP inventory improves the capability for emergency heat removal by natural convection, thereby greatly reducing the designated evacuation area
  7. Fluidity facilitates removal from the reactor to ever-safe containers
  8. High heat capacity of fuel restricts temperature rise on loss of normal cooling
  9. Low salt vapor pressure minimizes the effect of any temperature rise

Ottewitte's list would have been based on knowledge that the writers of WASH-1222 in 1972. The list should not be regarded as comprehensive. Thus WASH-1222 appears to be deliberately minimizing the numerous and well-established safety advantages of the MSR.

Quote:
Selected conceptual design data for a large MSBR, based primarily on design studies performed at ORNL, are given in Table I.

There are, however, problem areas associated with the MSBR which must be overcome before the potential of the concept could be attained. These include development of continuous fuel processing, reactor and processing structural materials, tritium control methods, reactor equipment and systems, maintenance techniques, safety technology, and MSBR codes and standards. Each of these problem areas will now be evaluated in some detail, using as a reference point the technology which was demonstrated by the Molten Salt Reactor Experiment (MSRE) during its design, construction and operation at Oak Ridge and the conceptual design parameters presented in Table I and in Appendix A. A conceptual flowsheet for this system is shown in Figure 1.


I will comment on the problems alluded too in this paragraph as WASH-1222 discusses them in greater detail.


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PostPosted: Feb 24, 2008 10:59 pm 
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Assessments of technologies under development are problematic. At best they are like a snapshot of where the technology is, coupled with some educated guesses about where it might be headed. It is the educated guess that creates the biggest problems. Overly optimistic assessments of the potential for progress may be based on unrealistic expectations for program progress. An example of this was Milton Shaw’s assessments of the degree of difficulty involved in the development of commercial LMFBR. Conversely, bureaucratic manipulators like Shaw, could use the assessments to kill promising technologies that competed with pet projects.

It should be understood that the purpose of any technological research and development program is to identify and fix problems related to the implementation of any new technology. Considering the radical and daring nature of the Molten Salt Reactor concept, and the extreme conditions the reactor was expected to operate under, developmental problems were to be expected. The point of the 1966 to 1969 Molten Salt Reactor Experiment was to explore potential problems and fix them before the MSR went into development for commercial use.

Quote:
VI. STATUS OF MSBR TECHNOLOGY

A. MSRE - The Reference Point for Current Technology


The Molten Salt Reactor Experiment (MSRE) was begun in 1960 at ORNL as part of the Civilian Nuclear Power Program. The purpose of the experiment was to demonstrate the basic feasibility of molten salt power reactors. All objectives of the experiment were achieved during its successful operation from June 1965 to December 1969. These included the distinction of becoming the first reactor in the world to operate solely on 233U. Some of the more significant dates and statistics pertinent to the MSRE are given in Table II.

In spite of the success of the MSRE, there are many areas of molten salt technology which must be expanded and developed in order to proceed from this small non-breeding experiment to a safe, reliable, and economic 1000 MWe MSBR with a 30-year life. To illustrate this point, some of the most important differences in basic design and performance characteristics between the MSRE and a conceptual 1000 MWe MSBR are given in Table III. Scale-up would logically be accomplished through development of reactor plants of increasing size. Examination of Table III provides an appreciation of the scale-up requirements in going from the MSRE to a large MSBR. Some problems associated with progressing from a small experiment to a commercial, high performance power plant are not adequately represented by the comparison presented in the Table. Therefore it is useful to examine additional facets of MSBR technology in more detail.


This statement says nothing more than that the MSR is under development, and that problems have been identified and steps to identify and implement corrections are underway.

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B. Continuous Fuel Processing: The Key to Breeding

In order to achieve nuclear breeding in the single-fluid MSBR it is necessary to have an on-line continuous fuel processing system. This would accomplish the following:
  • Isolate protactinium-233 from the reactor environment so it can decay into the fissile fuel isotope uranium-233 before being transmuted into other isotopes by neutron irradiation.
  • Remove undesirable neutron poisons from the fuel salt and thus improve the neutron economy and breeding performance of the system.
  • Control the fuel chemistry and remove excess uranium-233 which is to be exported from the breeder system.
1. Chemical Process Development

The Oak Ridge National Laboratory has proposed a fuel processing scheme to accomplish breeding in the MSBR, and the flowsheet processes involve:
  1. Fluorination of the fuel salt to remove uranium as UF6.
  2. Reductive extraction of protactinium by contacting the salt with a mixture of lithium and bismuth.
  3. Metal transfer processing to preferentially remove the rare earth fission product poisons which would otherwise hinder breeding performance.
The fuel processing system shown in Fig. 2 is in an early stage of development at present and this type of system has not been demonstrated on an operating reactor. By comparison, the MSRE required only off-line, batch fluorination to recover uranium from fuel salt.

At this time, the basic chemistry involved in the MSBR processing scheme has been demonstrated in laboratory-scale experiments. Current efforts at Oak Ridge are being directed toward development of subsystems incorporating many of the required processing steps. Ultimately a complete breeder processing experiment would be required to demonstrate the system with all the chemical conditions and operational requirements which would be encountered with any MSBR.

Not shown on the flowsheet is a separate processing system which would require injecting helium bubbles into the fuel salt, allowing them to circulate in the reactor system until they collect fission product xenon, and then removing the bubbles and xenon from the reactor system. Xenon is a highly undesirable neutron poison which will hamper breeding performance by capturing neutrons which would otherwise breed new fuel. This concept for xenon stripping was demonstrated in principle by the MSRE, although more efficient and controllable stripping systems will be desirable for the MSBR. The xenon poisoning in the MSRE was reduced by a factor of six by xenon stripping; the goal for the MSBR is a factor of ten reduction.


Xenon is the Achilles heel of reactor physics. Xenon, a fission product, is a noble radioactive gas. It has a high neutron cross section, which means that it captures neutrons which could better be used in promoting chain reactions, or in breeding new fuel. In solid-fuel reactors, xenon remains inside fuel capsules where it poisons the nuclear process. The presence of xenon inside solid-fueled reactors created control issues, and attempts to compensate for xenon poisoning could make reactors unstable and potentially dangerous. However in fluid-fueled reactors, xenon can be removed or stripped. The ability to strip xenon from reactor fuel was a major accomplishment of ORNL, and clearly demonstrated the importance of the MSR concept.

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2. Fuel Processing Structural Materials

Aside from the chemical processes themselves, there are also development requirements associated with containment materials for the fuel processing systems. In particular, liquid bismuth presents difficult compatibility problems with most structural metals, and present efforts are concentrated on using molybdenum and graphite for containing bismuth. Unfortunately, both molybdenum and graphite are difficult to use for such engineering applications. Thus, it will be necessary to develop improved techniques for fabrication and joining before their use is possible in the reprocessing system.


This is simply a development issue, but one which does not appear to pose exceptional challenges.

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A second materials problem of the current fuel processing system is the containment for the fluorination step in which uranium is volatilized from the fuel salt. The fluorine and fluoride salt mixture is corrosive to most structural materials, including graphite, and present ORNL flowsheets show a “frozen wall” fluorinator which operates with a protective layer of frozen fuel salt covering a Hastelloy-N vessel wall. This component would require considerable engineering development before it is truly practical for use in on-line, full processing systems.


Considering the extreme conditions that the fuel processing system operated under, engineering challenges were to be expected. By 1972, ORNL scientists and engineers had made considerable progress in overcoming these challenges, and there were reasons to be optimistic about further progress.

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C. Molten Salt Reactor Design - Materials Requirements

In concept, the molten salt reactor core is a comparatively uncomplicated type of heat source. The MSRE reactor core, for example, consisted of a prismatic structure of unclad graphite moderator through which fuel salt flowed to be heated by the self-sustaining chain reaction which took place as long as the salt was in the graphite. The entire reactor internals and fuel salt were contained in vessels and piping made of Hastelloy-N, a high-strength nickel-base alloy which was developed under the Aircraft Nuclear Propulsion Program. Over the four-year lifetime of the MSRE, the reactor structural materials performed satisfactorily for the purposes of the experiments although operation of the MSRE revealed possible problems with long term use of Hastelloy-N in contact with fuel salts containing fission products.

The MSBR application is more demanding in many respects than the MSRE, and additional development work would be required in several areas of materials technology before suitable materials could become available.

1. Fuel and Coolant Salts

The MSRE fuel salt was a mixture of 7LiF–BeF2–ZrF4–UF4 in proportions of 65.0-29.1-5.0-0.9 mole %, respectively. Zirconium fluoride was included as protection against UO2 precipitation should inadvertent oxide contamination of the system occur. MSRE operation indicated that control of oxides was not a major problem and thus it is not considered necessary to include zirconium in future molten salt reactor fuels. It should also be noted that the MSRE fuel contained no thorium whereas the proposed MSBR fuels would include thorium as the fertile material for breeding. With the possible exception of incompatibilities with Hastelloy-N, the MSRE fuel salt performed satisfactorily throughout the life of the reactor.

The MSBR fuel salt, as currently proposed by ORNL, would be a mixture of 7LiF–BeF2–ThF4–UF4 in proportions of 71.7–16–12–0.3 mole %, respectively. This salt has a melting point of about 930°F (772 K) and a vapor pressure of less then 0.1 mm Hg (13 Pa) at the mean operating temperature of 1150°F (895 K). It also has about 3.3 times the density and 10 times the viscosity of water. Its thermal conductivity and volumetric heat capacity are comparable to water.

The high melting temperature is an obvious limitation for a system using this salt, and the MSBR is limited to high temperature operation. In addition, the lithium component must be enriched in 7Li in order to allow nuclear breeding, since naturally occurring lithium contains about 7.5% 6Li. 6Li is undesirable in the MSBR because of its tendency to capture neutrons, thus penalizing breeding performance.


The statement about the high melting temperature of salt is exceedingly strange. One of the disadvantages of water as a reactor coolant is its low boiling point and the high pressure required to generate steam above the normal boiling point of water. Steam pressure causes major safety problems for light-water reactors. In contrast molten salts have extremely high boiling points and vapor pressure is no problem in MSRs. Hence a major safety problem of LWRs is eliminated by using hot molten salt as a coolant. This comment is thus a sort of swift-boating. An attempt to point to major strength and argue that it is really a weakness.

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The chemical and physical characteristics of the proposed MSBR fuel mixture have been and are being investigated, and they are reasonably well known for unirradiated salts. The major unknowns are associated with the reactor fuel after it has been irradiated. For example, not enough is known about the behavior of fission products. The ability to predict fission product behavior is important to plant safety, operation, and maintenance. While the MSRE provided much useful information, there is still a need for more information, particularly with regard to the fate of the so-called "noble-metal" fission products such as molybdenum, niobium and others which are generated in substantial quantities and whose behavior in the system is not well understood.


Here we have more of the same. The MSRE explored numerous issues including the effects of radiation on molten salts, and the behavior of fission products in the circulating reactor salt fuel mixture. ORNL scientists wanted to know more. The desirability of conducting more research is regarded as a liability by WASH-1222.

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A more complete understanding of the physical/chemical characteristics of the irradiated fuel salt is also needed. As an illustration of this point, anomalous power pulses were observed during early operation of the MSRE with 233U fuel which were attributed to unusual behavior of helium gas bubbles as they circulated through the reactor. This behavior is believed to have been due to some physical and/or chemical characteristics of the fuel salt which were never fully understood. Out-of-reactor work on molten fuel salt fission product chemistry is currently under way. Eventually, the behavior of the fuel salt would need to be confirmed in an operating reactor.


Research is being conducted. But some answers will only come by trying the idea out.

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The coolant salt in the secondary system of the MSRE was of molar composition 66% 7LiF - 34% BeF2. While this coolant performed satisfactorily (no detectable corrosion or reaction could be observed in the secondary svstem), the salt has a high melting temperature (850°F / 728 K) and is relatively expensive. Thus, it may not be the appropriate choice for power reactors for two reasons: (1) larger volumes of coolant salt will be used to generate steam in the MSBR, and (2) salt temperatures in the steam generator should be low enough, if possible, to utilize conventional steam system technology with feedwater temperatures up to about 550°F. The operation of MSRE was less affected by the coolant salt melting temperature since it dumped the 8 MWt of heat via an air-cooled radiator. The high melting temperatures of potential coolant salts remain a problem. The current choice is a eutectic mixture of sodium fluoride and sodium fluoroborate with a molar composition of 8% NaF - 92% NaBF4; this salt melts at 725°F (658 K). It is comparatively inexpensive and has satisfactory heat transfer properties.

However, the effects of heat exchanger leaks between the coolant and fuel salts, and between the coolant salt and steam systems, must be shown to be tolerable. The fluoroborate salt is currently being studied with respect to both its chemistry and compatibility with Hastelloy-N.


Heat exchange leaks are a major problem with all liquid-cooled reactors. Given the endemic problem, molten salt has several advantages over other coolants. It operates under low pressure. High pressure is a major factor in leaks. The molten salts used in the reactor do not interact chemically with some metals. Electro-chemical interactions can be controlled. Finally Hastelloy-N had, by 1972, a considerable history of use in molten salt reactors. It was known to perform well under the high temperature conditions found in heat exchanges.

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2. Reactor Fuel Containment Materials

A prerequisite to success for the MSBR would be the ability to assure reliable and safe containment and handling of molten fuel salts at all times during the life of the reactor. It would be necessary, therefore, to develop suitable containment materials for MSBR application before plants could be constructed.

A serious question concerning compatibility of Hastelloy-N with the constituents of irradiated fuel salt was raised by the post-operation examination of the MSRE in 1971. Although the MSRE materials performed satisfactorily for that system during its operation, subsequent examination of metal which was exposed to MSRE fuel salt revealed that the alloy had experienced intergranular attack to depths of about 0.007 inch (0.2 mm). The attack was not obvious until metal specimens were tensile-tested, at which time cracks opened up as the metal was strained. Further examination revealed that several fission products, including tellurium, had penetrated the metal to depths comparable to those of the cracks. At the present time, it is thought that the intergranular attack was due to the presence of tellurium. Subsequent laboratory tests have verified that tellurium can produce, under certain conditions, intergranular cracking in Hastelloy-N.

Although the limited penetration of cracks presented no problems for the MSRE, concern now exists with respect to the chemical compatibility of Hastelloy-N and MSBR fuel salts when subjected to the more stringent MSBR requirements of higher power density and 30-year life. If the observed intergranular attack was indeed due to fission product attack of the Hastelloy-N, then this material may not be suitable for either the piping or the vessels which would be exposed to much higher fission product concentrations for longer periods of time. Efforts are under way to understand and explain the cracking problem, and to determine whether alternate reactor containment materials should be actively considered.

In addition to the intergranular corrosion problem, the standard Hastelloy-N used in the MSRE is not suitable for use in the MSBR because its mechanical properties deteriorate to an unacceptable level when subjected to the higher neutron doses which would occur in the higher power density, longer-life MSBR. The problem is thought to be due mainly to impurities in the metal which are transmuted to helium when exposed to thermal neutrons. The helium is believed to cause a deterioration of mechanical properties by its presence at grain boundaries within the alloy. It would be necessary to develop a modified Hastelloy-N with improved irradiation resistance for the MSBR, and some progress is being made in that direction. It appears at this time that small additions of certain elements, such as titanium, improve the irradiation performance of Hastelloy-N substantially. Development work on modified alloys with improved irradiation resistance is currently under way.


Some problems with Hastelloy-N were identified during the course of the MSRE. Their sources were identified, and work on fixing the problem is underway. – The problems would be routine for R&D, and fixing them posed no significant challenge. Indeed they were fixed shortly after WASH-1222 was written.

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3 Graphite

Additional developmental effort on two problems is required to produce graphites suitable for MSBR application. The first is associated with irradiation damage to graphite structures which results from fast neutrons. Under high neutron doses, of the order of 10^22 neutrons/cm2, most graphites tend to become dimensionally unstable and gross swelling of the material occurs.

Based on tests of small graphite samples at ORNL, the best commercially available graphites at this time may be usable to about 3 x 10^21 neutrons/cm2, before the core graphite would have to be replaced. This corresponds to roughly a four-year graphite lifetime for the ORNL reference design. While this might be acceptable, there are still uncertainties about the fabrication and performance of large graphite pieces, and additional work would be required before a four-year life could be assured at the higher MSBR power densities now being. considered. In any event, there would be an obvious economic incentive to develop longer-lived graphites for MSBR application since a four-year life for graphite is estimated to represent a fuel cycle cost penalty of about 0.2 mills/kW-hr relative to a system with 30-year graphite life.

The second major problem associated with graphites for MSBR application is the development of a sealing technique which will keep xenon, an undesirable neutron poison, from diffusing into the core graphite where it can capture neutrons to the detriment of breeding performance. While graphite sealing may not be necessary to achieve nuclear breeding in the MSBR, the use of sealed graphite would certainly enhance breeding performance. The economic incentives or penalties of graphite sealing cannot be assessed until a suitable sealing process is developed.

Sealing methods which have been investigated to date include pyrolytic carbon coating and carbon impregnation. Thus far, however, no sealed graphite that has been tested remained sufficiently impermeable to gas at MSBR design irradiation doses, and research and development in this area is continuing.


Graphite (a form of carbon used in pencils) can serve as a reactor moderator, as well as a structural material. Most molten salt concepts envision the use of graphite. However, there are some drawbacks to graphite. Under the high and radiation conditions found inside the MSR, graphite is expected to deteriorate. Readers of the discussion forum of Kirk Sorensen’s blog will find extensive discussions of the advantages and disadvantages of graphite, and proposed methods of solving the graphite problem. The use of graphite as a moderator and structural material inside MSRs is not strictly speaking absolutely required, and other approaches have been explored. It is possible to use heavy water as a moderator, and even unmoderated designs are possible. One approach would be to avoid structural uses of graphite, but moderate the reactor with small graphite spheres that float in the fuel salts. The graphite spheres could be floated out of the reactor as they aged, and replaced with new spheres. The graphite related issues are perhaps the most difficult material concerns, but since it is possible to forgo using graphite entirely, graphite problems are hardly fatal for molten salt reactor technology.

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4. Other Structural Materials

In addition to the structural materials requirements for the reactor and fuel processing systems proper, there are other components and systems which have special materials requirements. Such components as the primary heat exchangers and steam generators must function while in contact with two, different working fluids.

At the present time, Hastelloy-N is considered to be the most promising material for use in all salt containment systems, including the secondary piping and components. Research to date indicates that sodium fluoroborate and Hastelloy-N are compatible as long as the water content of the fluoroborate is kept low; otherwise, accelerated corrosion can occur. Additional testing would be needed and is underway.

Hastelloy-N has not been adequately evaluated for service under a range of steam conditions and whether it will be a suitable material for use in steam generators is still not known.


Here WASH-1222 does not point to significant problems, but to a need for more research and testing. That would be normal for research and development projects.


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PostPosted: Feb 24, 2008 11:00 pm 
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This section of WASH-1222 focuses on the ORNL MSBR design of 1972. I must confess that I am not an admirer of the MSBR design. The flaws of the design were in part a direct result of the requirements of Milton Shaw. A better research and development plan would have involved a evolutionary follow up to the MSRE. The MSRE was a one-fluid design. Scientists like my father felt that a two-fluid design held greater promises for breeding. The fluoride-salt based reactor us not a high output breeder. It is a 1-for-1 breeder. It only breeds as much fuel as it uses, or slightly more. I call this type of breeder an auto-breeder.

A second form of molten salt breeder, the chloride reactor, offers far more promise as a breeder of surplus reactor fuel. The 1972 MSBR was in many ways a mistake and not a fair test of molten salt reactor technology. I invite my readers to contemplate the distinction between a flawed concept, and a flawed design. The Molten Salt Reactor is a promising concept, but the 1972 MSBR was a flawed design.

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E. Reactor Equipment and Systems Development

While the MSBR would utilize some existing engineering technology from other reactor types, there are specific components and systems for which additional development work is required. Such work would have to take into account the induced activity that those components would accumulate in the MSBR system, i.e., special handling and maintenance equipment would also need to be developed. The previous discussion has already dealt with a number of these, such as fuel processing components and systems, but additional discussion is appropriate.


True.

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1. Components

As indicated in the Table 3, a number of components must be scaled up substantially from the MSRE sizes before a large MSBR is possible. The development of these larger components along with their special handling and maintenance equipment is probably one of the most difficult and costly phases of MSBR development. However, reliable, safe, and maintainable components would need to be developed in order for any reactor system to be a success.

The MSBR pumps would likely be similar in basic design to those for the MSRE, namely, vertical shaft, overhung impeller pumps.

Substantial experience has been gained over the years in the design, fabrication and operation of smaller salt pumps, but the size would have to be increased substantially for MSBR application. The development and proof-testing of such units along with their handling and maintenance equipment and test facilities are expected to be costly and time consuming.

The intermediate heat exchangers for the MSBR must perform with a minimum of salt inventory in order to improve the breeding performance by lowering the fuel inventory. Special surfaces to enhance heat transfer would help achieve this, and more studies would be in order. Based on previous experience with other reactor systems, it is believed that these units would require a difficult development and proof testing effort.

The steam generator for MSBR applications is probably the most difficult large component to develop since it represents an item for which there has been almost no experience to date. It is believed that a difficult development and proof-testing program would be needed to provide reliable and maintainable units. As discussed previously, the high melting temperatures of candidate secondary coolants, such as sodium fluoroborate, present problems of matching with conventional steam system technology. At this time, central station power plants utilize feedwater temperatures only up to about 550ºF (560 K). Therefore, coupling a conventional feedwater system to a secondary coolant which freezes at 725ºF (658 K) presents obvious problems in design and control. It might be necessary to provide modifications to conventional steam system designs to help resolve the problems. Because of these factors, a study related to the design of steam generators hash been initiated at Foster-Wheeler Corporation.

Control rods and drives for the MSBR would also need to be developed. The MSRE control rods were air-cooled and operated inside Hastelloy-N thimbles which protruded down into the fuel salt. The MSBR would require more efficient cooling due to the higher power densities involved. Presumably rods and drives would be needed which permit the rods to contact and be cooled by the fuel salt.

The salt valves for large MSBR's represent another development problem, although the freeze valve concept which was employed successfully in the MSRE could likely be scaled up in size and utilized for many MSBR applications. Mechanical throttling valves would also be needed for the MSBR salt systems, even though no throttling valve was used with the MSRE. Mechanical shutoff valves for salt systems, if required, would have to be developed.

Other components which would require considerable engineering development and testing include the helium bubble generators and gas strippers which are proposed for use in removing the fission product xenon from the fuel salt. Research and development in this area is currently under way as part of the technology program at ORNL.


MSR technology is poorly matched to steam generation but well-matched to Brayton-cycle gas turbines. ORNL molten-salt research began with an attempt to power aircraft jet engines with the output heat from reactors. The Aircraft Nuclear Propulsion program envisioned small molten-salt reactors heated red-hot by nuclear energy, channeling its coolant fuel salts into the jet engines of the atomic powered aircraft. Thus the molten-salt reactor was known to be well matched to Brayton-cycle turbines. Using a MSR for heat to convert water into steam presents difficulties.

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2. Systems

The integration of all required components into a complete MSBR central station power plant would involve a number of systems for which development work is still required. It should be noted that some components, such as pumps and control rod drives, would require their own individual system for functions such as cooling-and lubrication.

Given the required components and materials of construction, the basic reactor primary and secondary flow systems can be designed. However, the primary flow system would require supporting systems for continuous fuel processing, on-line fuel analysis and control of salt chemistry, reactor control and safety, handling of radioactive gases, fuel draining from every possible holdup area in components and equipment, afterheat control, and temperature control during non-nuclear operations.

The continuous fuel processing systems proposed to date are quite complicated and include a number of subsystems, all of which would have to operate satisfactorily within the constraints of economics, safety, and reliability. The effects of off-design conditions on these systems would have to be understood so that control would be possible to prevent inadvertent contamination of the primary system by undesirable materials.

The fuel drain system is important to both operation and safety since it would be used to contain the molten fuel whenever a need arises to drain the primary system or any component or instrument for maintenance or inspection. Thus, additional systems would be required, each with its own system for maintaining and controlling temperatures. The fuel-salt drain tank would have to be equipped with an auxiliary cooling system capable of rejecting about 18 MWt of heat should the need arise to drain the salt immediately following nuclear operation. The secondary coolant system would also require subsystems for draining and controlling of salt chemistry and temperature. In addition, the secondary loop might require systems to control tritium and to handle the consequences of steam generator or heat exchanger leaks.

The steam system for the MSBR might require a departure from conventional designs due to the unique problems associated with using a coolant having a high melting temperature. Precautions would have to be taken against freezing the secondary salt as it travels through the steam generator; suitable methods for system startup and control would need to be incorporated. ORNL has proposed the use of a supercritical steam system which operates at 3500 psia (240 bar) and provides 700ºF (640 K) feedwater by mixing of supercritical steam and high pressure feedwater. This system would introduce major new development requirements because it differs from conventional steam cycles.


The argument here is that there were many developmental challenges confronting ORNL’s MSBR. Had the same criteria been applied to the AEC’s proposed LMFBR, many of the same concerns would have emerged. Since the current MSR concepts envision the use of closed-circulation gas turbines, rather than steam generation, so while the inclusion of concern about the steam generation system in 1972 was valid, it is not still valid today. However, the design and construction of primary and secondary heat exchange systems in any MSR power recovery system, do present technical challenges.

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F. Maintenance - A Difficult Problem for the MSBR

Unlike solid-fueled reactors in which the primary system contains activation products and only those fission products which may leak from defective fuel pins, the MSBR would have the bulk of the fission products dispersed throughout the reactor system. Because of this dispersal of radioactivity, remote techniques would be required for many maintenance functions if the reactor were to have an acceptable plant availability in the utility environment.

The MSRE was designed for remote maintenance of highly radioactive components; however, no major maintenance problems (removal or repair of large components) were encountered after nuclear operation was initiated. Thus, the degree to which the MSRE experience on maintenance is applicable to large commercial breeder reactors is open to question.

As has been evident in plant layout work on nuclear facilities to date, this requirement for remote maintenance will significantly affect the ultimate design and performance of the plant system. The MSBR would require remote techniques and tools for inspection, welding and cutting of pipes, mechanical assembly and disassembly of components and systems, and removing, transporting and handling large component items after they become highly radioactive. The removal and replacement of core internals, such as graphite, might pose difficult maintenance problems because of the high radiation levels involved and the contamination protection which would be required whenever the primary system is opened.

Another potential problem is the afterheat generation by fission products which deposit in components such as the primary heat exchangers. Auxiliary cooling might be required to prevent damage when the fuel salt is drained from the primary system, and a requirement for such cooling would further complicate inspection and maintenance operations.

In some cases, the inspection and maintenance problems of the MSBR could be solved using present technology and particularly experience gained from fuel reprocessing plants. However, additional technology development would be required in other areas, such as remote cutting, alignment, cleaning and welding of metal members. Depending to some degree on the particular plant arrangement, other special tools and equipment would also have to be designed and developed to accomplish inspection and maintenance operations.

In the final analysis, the development of adequate inspection and maintenance techniques and procedures and hardware for the MSBR hinges on the success of other facets of the program, such as materials and component development, and on the requirement that adequate care be taken during plant design to assure that all systems and components which would require maintenance over the life of the plant are indeed maintainable within the constraints of utility operation.


The observations about potentially difficult maintenance problems are valid. The removal and replacement of core graphite was a realistic possibility with the MSBR, and indeed its graphite core might have requited frequent replacement. The best approach to maintenance is to eliminate or simplify maintenance problems as much as possible during the design phase. If pumps can be replaced with thermal siphoning this would eliminate problems attendant on the design of costly parts, and the problems and cost of maintaining them. Having said this, it still should be noted that requirement for automated maintenance of the MSBR was by no means beyond the capacity of 1972 technology.

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G. Safety - Different Issues for the MSBR

The MSBR concept has certain characteristics which might provide advantages relating to safety, particularly with respect to postulated major types of accidents currently considered in licensing activities. Since the fuel would be in a molten form, consideration of the core meltdown accident is not applicable to the MSBR. Also, in the event of a fuel spill, secondary criticality is not a problem since this is a thermal reactor system requiring moderator for nuclear criticality.

Other safety features include the fact that the primary system would operate at low-pressure with fuel salt that is more than 1000ºF (550 K) below its boiling point, that fission product iodine and strontium form stable compounds in the fluoride salts, and that the salts do not react rapidly with air or water. Because of the continuous fuel processing, the need for excess reactivity would be decreased and some of the fission products would be continuously removed from the primary system. A prompt negative temperature coefficient of reactivity is also a characteristic of the fuel salt.

Safety disadvantages, on the other hand, include the very high radioactive contamination which would be present throughout the primary system, fuel processing plant, and all auxiliary primary systems such as the fuel drain and off-gas systems. Thus, containment of these systems would have to be assured. Also, removal of decay heat from fuel storage systems would have to be provided by always ready and reliable cooling systems, particularly for the fuel drain tank and the 233Pa decay tank in the reprocessing plant where megawatt quantities of decay heat must be removed. The tritium problem, already discussed, would have to be controlled to assure safety.

Based on the present state of MSBR technology, it is not possible to provide a complete assessment of MSBR safety relative to other reactors. It can be stated, however, that the safety issues for the MSBR are generally different from those for solid-fuel reactors, and that more detailed design work must be done before the safety advantages and disadvantages of the MSBR could be fully evaluated.


This last statement is quite disingenuous. The problem lies with the word complete. By 1972 the safety advantages of the MSR were well known. If there were missing components of the picture the limits of the safety problems were obvious. Uri Gat and H. L. Dodds were later to discuss the safety of the MSR in a paper titled MOLTEN SALT REACTORS - SAFETY OPTIONS GALORE. Gat also discusses MSR safety in another paper. This discussion was based on of information that would have been available to the authors of WASH-1222. It is clear from Gat and Dodds that WASH-1222 could and should have discussed the relative safety of the MSR. In summarizing the MSR’s safety features, they observe:

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“The molten-salt reactor with fuel processing can be designed to be almost as safe as desirable. The basic features of fluoride based molten salts allow for a high temperature, and thus efficient, operation at low pressures. The molten salts are inert and well compatible with selected structural materials. The MSR is not subject to safety concerns from chemical or mechanical violent reactions or explosions. External cooling results in a simple design with few structural requirements that permits optimization of the design for safety-eliminating compromises. The on-line processing results in an equilibrium fuel that requires no excess reactivity for burn-up or poison compensation. The fission product inventory, and therefore the source term, is held low. The severe accidents of uncontrolled super-criticality or loss-of-cooling that fails to remove the after-heat can become a hypothetical accident.

The dreaded meltdown looses all its meaning in a fluid-fuel reactor. In an MSR, a spill may be self-containing by the freezing of the fuel upon cooling. Freeze valves are one more feature that can make an MSR PINT (passive, inherent, non-tamperable) safe.”


In addition Gat and Dodds believe that the MSR concept is capable of refinement into an reactor that is absolutely and ultimately safe, if that degree of safety is desired.

It is clear then, that WASH-1222 has largely discounted the known and highly desirable safety features of the MSR. This was a reflection of the problem that Milton Shaw brought to the AEC. As I have demonstrated in earlier posts, Milton Shaw believed that LWRs were completely safe. His anger caused by Alvin Weinberg insistence that LWR safety issues had not been resolved eventually lead to Weinberg’s firing. Shaw was determine to prevent adverse evaluations of the relative safety of his pet reactors to become a part of AEC project evaluations. Thus the safety related comments of WASH-1222 ware written to further Shaw’s agenda, and in no way reflected the state of reactor technology. This was part of the path which lead to Three Mile Island. In addition the safety features of the LMFBR compared quite unfavorably with those of the MSR. Shaw was protecting another pet project by withholding a true safety evaluation of the MSR.


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PostPosted: Feb 24, 2008 11:01 pm 
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H. Codes, Standards, and High Temperature Design Methods

Codes and standards for MSBR equipment and systems must be developed in conjunction with other research and development before large MSBR's can be built. In particular, the materials of construction which are currently being developed and tested would have to be certified for use in nuclear power plant applications.

The need for high-temperature design technology is a problem for the MSBR as well as for other high temperature systems. The AEC currently has under way a program in support of the LMFBR which is providing materials data and structural analysis methods for design of systems employing various steel alloys at temperatures up to 1200ºF (920 K). This program would need to be broadened to include MSBR structural materials such as Hastelloy-N and to include temperatures as high as 1400ºF (1030 K) to provide the design technology applicable to high-temperature, long-term operating conditions which would be expected for MSBR vessels, components, and core structures.


No comment.

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VII. INDUSTRIAL PARTICIPATION IN THE MSBR PROGRAM

Privately funded conceptual design studies and evaluations of MSBR technology were performed in 1970 by the Molten Salt Breeder Reactor Associates (MSBRA), a study group headed by the engineering firm of Black & Veatch and including five midwest utilities. The MSBRA concluded that the economic potential of the MSBR is attractive relative to light-water reactors, but they recognized a number of problems which must be resolved in order to realize this potential. Since that time the MSBRA has been relatively inactive.

A second privately funded organization, the Molten Salt Group, is headed by Ebasco Services, Incorporated and includes five other industrial firms and fifteen utilities. In 1971 the Group completed an evaluation of the MSBR concept and technology and concluded that existing technology is sufficient to justify construction of an MSBR demonstration plant although the performance characteristics could not be predicted with confidence. Additional support for further studies has recently been committed by the members of this group.

In addition to these studies, manufacturers of graphite and Hastelloy-N have been cooperating with ORNL to develop improved materials.

There has been little other industrial participation in the MSBR Program aside from ORNL subcontractors. At the present time, there are two ORNL subcontracts in effect. Ebasco Services, Inc., utilizing the industrial firms who are participants in the Molten Salt Group is performing a design and evaluation study. Foster-Wheeler Corporation is currently performing design studies on steam generators for MSBR application.


Another disingenuous argument. We are pointed to evidence of industrial interest in the MSR concept. Then we are told that there is little industrial participation in MSR beyond ORNL contractors. In fact, interested industries in the early 1970's included, Babcock & Wilcox, Byron Jackson, Cabot Corp., Continental Oil, and Union Carbide. Other industries had approached ORNL informally. Because the viability of the MSR concept had not been established prior to the MSRE, industrial interest was premature prior to the late 1960’s. The MSRE had changed that. The Black & Veatch study had demonstrated that the MSR was potentially economically attractive to utilities relative to light-water reactors. There is little doubt that Shaw would have regarded this as a threat to his intention to direct US nuclear development towards the light water reactor. The LMFBR would have complimented the LWR, while the MSR had the potential of competing successfully against it.

Quote:
A number of factors can be identified which tend to limit further industrial involvement at this time, namely:
  1. The existing major industrial and utility commitments to the LWR, HTGR, and LMFBR.
  2. The lack of incentive for industrial investment in supplying fuel cycle services, such as those required for solid-fuel reactors.
  3. The overwhelming manufacturing and operating experience with solid-fuel reactors in contrast with the very limited involvement with fluid-fueled reactors.
  4. The less advanced state of MSBR technology and the lack of demonstrated solutions to the major technical problems associated with the MSBR concept.


The formula of industrial involvement is used to object to the MSR project. Here the term refers to manufacturer involvement. No one is yet interested in building a commercial MSR. Yet the AEC has not encouraged manufactures to do so, indeed as WASH-1222 demonstrates, quite the opposite was the case. Because the AEC was in effect discouraging manufacturer interest in the MSR, manufactures were largely withholding interest. The lack of manufacture interest was then used to justify the AEC’s attempt to suppress MSR technology.

Yet if there was customer interest in MSR technology, as an earlier paragraph of this section reports, there was potentially manufacturer interest.

The second point of this argument at first seem to be obvious. The MSR, by using fuel far more efficiently would seem to create disincentives for industrial investment in supplying fuel cycle services. But this could be a case of Jevon’s paradox. By greatly increasing the efficiency of nuclear fuel use, the MSR might actually create as situation in which more rather than less fuel was in demand. Thus the objection is short sighted, and assumes shortsightedness on the part of suppliers as well.

The third objection could be raised against any new technology. In 1952 manufacturers had had very limited involvement Light Water Reactors. Demand had led them to explore the new technology.

The fourth objection is circular. In effect, “the less advanced state of MSBR technology and the lack of demonstrated solutions to the major technical problems associated with the MSBR concept,” is used as an argument against advancing the state of MSBR technology, and finding solutions to the major technical problems.

Quote:
It should be noted that these factors are also relevant considerations in establishing the level of governmental support for the MSBR program which in turn, to some extent, affects the interest of the manufacturing and utility industries.


False premises lead to false conclusions. This is the money statement as far as Milton Shaw was concerned. Given the weakness and contradictions of the arguments WASH-1222 offered against “industrial interest” in the development of MSR technology, the case most certainly had not been made against continuing government support of research directed at the development of MSR technology. Milton Shaw had thus erected his case in favor of discontinuing development of MSR technology of the most insubstantial grounds.

Quote:
VIII. CONCLUSIONS

The Molten Salt Breeder Reactor, if successfully developed and marketed, could provide a useful supplement to the currently developing uranium-plutonium reactor economy. This concept offers the potential for:
  • Breeding in a thermal spectrum reactor;
  • Efficient use of thorium as a fertile material;
  • Elimination of fuel fabrication and spent fuel shipping;
  • High thermal efficiencies.


Although WASH-1222 slights the advantages of MSR technology, it points to a singular advantage of Molten Salt Reactor technology, the elimination of the problem of “nuclear waste.” The problem of spent fuel was one of the aspects of the LWR that Milton Shaw was trying to sweep under the rug in 1972.

H. G. MacPherson listed the advantages of the MSR as:

1. The fuel handling system will be much simpler.
2. The molten salts have a much higher heat capacity per unit volume than sodium, so that the physical size of pumps and piping will be smaller.
3. There is no threat of a "core disruptive accident" with the MSCR, so that safety-related equipment can be simpler.
4. The molten salts have a much lower thermal conductivity than sodium, so that sudden coolant temperature changes will provide less thermal shock to system components.
5. The coolant is more compatible with water than is sodium, so that there should be fewer problems in the design and maintenance of steam generators.

MacPherson did what WASH-1222 did not do. He compaired the MSR to the LMFBR and demonstrated that the MSR had decided advantages in a number of areas. MacPherson's comments are very brief, and are his comparison is far from exhaustive.

Eric Ottewitte listed some salient advantages of an unusual type of MSR, the Molten Chloride Fast Reactor (MCFR) as:

1. Simplicity: no control rods, fuel handling mechanisms, fuel elements or associated structures. Very uncluttered: should maximize test space and facilitate access thereto. Fluid fuel can be transferred remotely by pumping through pipes connecting storage and reactor. 

2. MSRs don't refuel or reprocess, just add fuel and process out wastes. Continuous processing and refueling would minimize reactor downtime. Can usefully consume all fuel forms, simplifying fuel supply while simultaneously solving other people's problems. 

3. MSR is the safest concept of all due to very strong negative temperature coefficient. No gaseous hydrogen can possibly evolve from fuel or primary coolant. Fuel already molten and handled by system. Simple design technique makes boiling impossible. Continuous removal of fission products reduces their heat source by two orders of magnitude; consequently, natural circulation suffices for emergency cooling, thereby greatly reducing the designated evacuation area. Also, under any off-normal conditions, the liquid fuel can be channeled to a continuously cooled drain tank, in a short time. 

4. Very fast neutron spectrum in an annular core engenders high neutron fluxes, driving inner and outer thermal neutron flux traps, each variable in size and neutron energy spectrum by means of molten salt composition. Elimination of fuel cladding and structural material significantly improves the neutron economy of the reactor: more neutrons are available for applications. 

5. Elimination of pressurized and pressure-evolving components inside the containment, reducing risk of containment failure. 

6. Potential additional missions for an MCFR BATR could include
A. Sr and Cs waste transmutation because of very high neutron flux
B. Useful consumption of fissile fuel from dismantled weapons because of the flexibility in fuel form
C. Process heat R&D due to high temperature capability
D. A 6LiD or 6LiOD shell for generation of a 14 MeV fusion neutron trap.

Ottewitte also notes some disadvantages of MSRs and MCFRs.

Quote:
Notwithstanding these attractive features, this assessment has reconfirmed the existence of major technological and engineering problems affecting feasibility of the concept as a reliable and economic breeder for the utility industry. The principal concerns include uncertainties with materials, with methods of controlling tritium, and with the design of components and systems along with their special handling, inspection and maintenance equipment. Many of these problems are compounded by the use of a fluid fuel in which fission products and delayed neutrons are distributed throughout the primary reactor and reprocessing systems.


Here we have Shaw's attempt to nail the lid on the coffin of the MSR. The claim that "this assessment has reconfirmed the existence of major technological and engineering problems affecting feasibility of the concept as a reliable and economic breeder for the utility industry" is demonstrably untrue. The assessment had not established anything other that MSR technology requires further development. WASH-1222 does not established that MSR developmental problems are unusually difficult and certainly has not established that developmental problems are in any way insurmountable. Many of the so called principal concerns were close to resolution when WASH-1222 was written, as the author must have known. The final sentence provides us with yet another example of Swift boating, turning the MSR primary strength - its fluid core, into a weakness.

Quote:
The resolution of the problems of the MSBR will require the conduct of an intensive research and development program. Included among the major efforts that would have to be accomplished are:
  • Proof testing of an integrated reprocessing system;
  • Development of a suitable containment material;
  • Development of a satisfactory method for the control and retention of tritium;
  • Attainment of a thorough understanding of the behavior of fission products in a molten-salt system;
  • Development of long-life moderator graphite, suitable for breeder application;
  • Conceptual definition of the engineering features of the many components and systems;
  • Development of adequate methods and equipment for remote inspection, handling, and maintenance of the plant.


The argument here begins with some partially true statements of premises. But the resolution 2 and 3 was already well in hand, which Milton Shaw and the author of WASH-1222 must have known. A more honest assessment might have taken note of progress in some area. Milton Shaw was not interested in an honest assessment. The other problems were what could be considered a normal part of research and development. Problem 5 was the most difficult, but in fact graphite could be dispensed with entirely without compromising the MSR’s safety, stability and ability to breed. The basic argument continues to be that the MSR is in the development stage, therefore the MSR should not be developed.

But not all of the premises to the argument are stated, and unstated premises are false. WASH-1222 has already told us that other nuclear technologies, including the LWR and the LMFBR, are so mature that they were beyond the research and development stage requirements which are outlined above. In 1972 this was disputed by scientist at AEC Labs, who correctly argued that LWR safety issues had not been adequately addressed. Many scientist at AEC facilities had well founded doubts about the maturity of LMFBR technology. In addition, in 1972 there was growing public concern about both the problems of “nuclear waste,” and “nuclear proliferation.” Both of these issues were unresolved problems for LWR technology. “Nuclear proliferation,” was and is a significant issue for LMFNR technology. In contrast, the MSR offers attractive solutions to both the “nuclear waste,” and “nuclear proliferation” issues. Thus the hidden premises of the conclusions reached by WASH-1222, were basically false. False assumptions lead to false conclusions, and this was most certainly the case with WASH-1222.

Quote:
The major problems associated with the MSBR are rather difficult in nature and many are unique to this concept. Continuing support of the research and development effort will be required to obtain satisfactory solutions to the problems. When significant evidence is available that demonstrates realistic solutions are practical, a further assessment could then be made as to the advisability of advancing into the detailed design and engineering phase of the development process including that of industrial involvement. Proceeding with this next step would also be contingent upon obtaining a firm demonstration of interest and commitment to the concept by the power industry and the utilities and reasonable assurances that large-scale government and industrial resources can be made available on a continuing basis to this program in light of other commitments to the commercial nuclear power program and higher priority energy development efforts.


This last paragraph requires very careful examination:

1. “The major problems associated with the MSBR are rather difficult in nature and many are unique to this concept.” In fact WASH-1222 has not shown that there were exceptional difficulties involved in the development of MSR technology.
2. “Continuing support of the research and development effort will be required to obtain satisfactory solutions to the problems.” This is to state the obvious.
3. “When significant evidence is available that demonstrates realistic solutions are practical, a further assessment could then be made as to the advisability of advancing into the detailed design and engineering phase of the development process including that of industrial involvement.” Which would have been the case had not WASH-1222 not been used to justify curtailing of MSR research.
4. “Proceeding with this next step would also be contingent upon obtaining a firm demonstration of interest and commitment to the concept by the power industry and the utilities and reasonable assurances that large-scale government and industrial resources can be made available on a continuing basis to this program in light of other commitments to the commercial nuclear power program and higher priority energy development efforts.” This last statement is the crowning hypocrisy of WASH-1222, a document intended to discourage further industrial interest in the MSR concept, and to lay down a smokescreen for Shaw’s moved to destroy the MSR project.


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PostPosted: Feb 24, 2008 11:04 pm 
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Some concluding remarks about WASH-1222

I have argued that WASH-1222 was a bureaucratic hatchet job. It was designed as part of Milton's Shaw's concerted program to kill off the Molten Salt Reactor. I would also like to note that WASH-1222 should be read in light of a closely related event, the firing of Alvin Weinberg as Director of ORNL. I believe that the WASH-1222 and the firing of Weinberg were part of a single bureaucratic move by Shaw to gain control of the American nuclear establishment, and to control the future direction of nuclear technology and of the nuclear industry in the United States. Indeed I believe that despite his 1973 firing by Dixie Lee Ray, Shaw largely achieved his objectives. The United States Nuclear Industry still largely carries Milton Shaw's imprint. Unfortunately the legacy of problems left by Shaw's fundamentally flawed vision is still with us. The mistrust of nuclear safety and of the governmental regulation of the nuclear industry is still widespread in American society, and represents a significant handicap in the fight against global warming.

The 1970 Bureau of Mines Mineral Yearbook reported:

"[The] AEC also requested proposals for a design study of a 1,000-megawatt molten-salt
breeder reactor (MSBR). There was also a significant increase in private efforts involving this concept. The Molten Salt Breeder Reactor Associates, an association of five electric utility companies and a consulting engineering firm, completed Phase I of their study of the MSBR. In addition, 15 utility companies and six major industrial companies formed the Molten Salt Group, which will jointly study MSBR technology, including the feasibility of thorium
as a fuel.12"

The footnote cited "Wall Street Journal. V. 176, No. 29, Aug. 10, 1970, p. 17."

Contrary to WASH-1222 there was by 1970 considerable industrial interest in MSR technology.

Among the reports on the MSR by private industrial groups and their consultants were:

Molten-Salt Breeder Reactor Associates Staff, Final Report, Phase I Study—Project for
Investigation of Molten-Salt Breeder Reactor, Black & Veatch Consulting Engineers,
Kansas City, Mo. (1970).

Evaluation of a 1000-MWe Molten-Salt Breeder Reactor, Technical Report of the
Molten-Salt Group, Part II, Ebasco Services, Inc., October 1971.

Molten-Salt Reactor Technology, Technical Report of the Molten-Salt Group, Part I,
Ebasco Services, Inc., December 1971.

1000-MW(e) Molten-Salt Breeder Reactor Conceptual Design Study, Final Report—
Task I, Ebasco Services, Inc., New York, February 1972.

Shaw had managed to abort an important industrial development and his destructive action had profoundly negative implications for the energy future of the United States.

Shaw's vision was also flawed by his failure to recognize that problems like "nuclear waste" and "nuclear proliferation" could and should be solved by a radical change in reactor design. The MSR possessed the potential to resolve these issues. Failure to move forward on MSR technology meant that the best chance to address major public concerns about nuclear technology was ignored.

In 2008 MSR technology remains potentially the best single tool for responding to the challenge posed by global warming, and peak fossil fuel energy. Yet the molten salt reactor is today little known and almost entirely ignored by decision makers. This should not be its fate, considering the potential that it brings.


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PostPosted: Feb 25, 2008 11:58 am 
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Charles,

Thanks so much for this absolutely excellent review. I`ve only skimmed through so far, I look forward to looking closer when I have a chance.

David L.


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PostPosted: Jul 25, 2008 7:45 pm 
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charlesB

Quote:
Although WASH-1222 slights the advantages of MSR technology, it points to a singular advantage of Molten Salt Reactor technology, the elimination of the problem of “nuclear waste.” The problem of spent fuel was one of the aspects of the LWR that Milton Shaw was trying to sweep under the rug in 1972.

H. G. MacPherson listed the advantages of the MSR as:

1. The fuel handling system will be much simpler.
2. The molten salts have a much higher heat capacity per unit volume than sodium, so that the physical size of pumps and piping will be smaller.
3. There is no threat of a "core disruptive accident" with the MSCR, so that safety-related equipment can be simpler.
4. The molten salts have a much lower thermal conductivity than sodium, so that sudden coolant temperature changes will provide less thermal shock to system components.
5. The coolant is more compatible with water than is sodium, so that there should be fewer problems in the design and maintenance of steam generators.

MacPherson did what WASH-1222 did not do. He compaired the MSR to the LMFBR and demonstrated that the MSR had decided advantages in a number of areas. MacPherson's comments are very brief, and are his comparison is far from exhaustive.


Where can I find MacPherson's evaluation?


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PostPosted: Jul 25, 2008 7:51 pm 
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CharlesH
Quote:
Where can I find MacPherson's evaluation?

http://www.energyfromthorium.com/pdf/MSadventure.pdf


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PostPosted: Nov 08, 2008 9:08 am 
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I've started a wiki article on WASH-1222:

WASH-1222


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PostPosted: Jan 19, 2011 4:42 pm 
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In one of the Google Tech talk videos one of the expressed concerns for LFTR was the graphite moderator. This report states:
Quote:
Based on tests of small graphite samples at ORNL, the best commercially available graphites at this time may be usable to about 3 x 10^21 neutrons/cm2, before the core graphite would have to be replaced. This corresponds to roughly a four-year graphite lifetime for the ORNL reference design.

The moderator should hold for 9.5 years if a flux of 10^13 neutrons/(sec*cm2) is assumed. The flux in the small research reactor of the Technical University of Delft is 10^13 hence the assumption. I also vaguely remember that the neutron flux is higher for small and lower for large reactors. Does anyone know more about this? I couldn't find the flux for today's operating energy producing reactors on google. I should have the number in a report at home somewhere, but I'm not at home...

Also note that the (imo quite irrational) evaluation was written in the beginning of the 70's. What amount of neutron radiation would today's high quality graphite be able to withstand?


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