•V. \ ORNL-3708 Contract No. W-7405-eng-26 MOLTEN-SALT REACTOR PROGRAM SEMIANNUAL PROGRESS REPORT For Period Ending July 31, 1964 R. B. Briggs, Program Director NOVEMBER 1964 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION MARTIN MARIETTA ENERGY SYSTEMS LIBRARIES 3 M45b DDSDD3S fl Xll CONTENTS INTRODUCTION 1 THE PLACE OF MOLTEN-SALT REACTORS IN THE AEC PROGRAM - E. E. Sinclair 1 MOLTEN-SALT POWER REACTORS AND THE ROLE OF THE MSRE IN THEIR DEVELOPMENT - R. B. Briggs 3 MSRE DESIGN AND CONSTRUCTION - W. B. McDonald 22 NUCLEAR CHARACTERISTICS OF THE MSRE - R. J. Engel 83 INSTRUMENTATION AND CONTROL OF THE MSRE - J. R. Tallackson 115 PUMP DEVELOPMENT - A. G. Grindell and P. G. Smith 147 COMPONENT DEVELOPMENT IN SUPPORT OF THE MSRE - Dunlap Scott, Jr 167 REMOTE MAINTENANCE OF THE MSRE - Robert Blumberg 190 FUEL PROCESSING FACILITY - R. B. Lindauer 201 PLANS FOR OPERATION OF THE MSRE - P. N. Haubenreich 205 CHEMICAL BASIS FOR MOLTEN-SALT REACTORS - W. R. Grimes 214 EFFECTS OF RADIATION ON THE COMPATIBILITY OF MSRE MATERIALS - F. F. Blankenship 252 PREPARATION OF MSRE FUEL, COOLANT, AND FLUSH SALTS - J. H. Shaffer 288 FUTURE CHEMICAL DEVELOPMENT - H. F. McDuffie 304 ANALYTICAL CHEMISTRY FOR THE MOLTEN-SALT REACTOR - J. C. White.... 320 METALLURGICAL DEVELOPMENTS - A. Taboada 330 MSRE GRAPHITE - W. H. Cook 373 INTRODUCTION This semiannual report is a collection of papers that were presented at a general information meeting, on the Molten-Salt Reactor Experiment at the Oak Ridge National Laboratory, August 18 and 19, 1964. It describes the design and construction of the Experiment, the related research and development, and is intended to bring the reader up to date on the status of the general technology of molten-salt thermal-breeder reactors. Previous progress reports of the Molten-Salt Reactor Program are listed below: ORNL-2474 Period Ending January 31, 1958 ORNL-2626 Period Ending October 31, 1958 ORNL-2684 Period Ending January 31, 1959 ORNL-2723 Period Ending April 30, 1959 ORNL-2799 Period Ending July 31, 1959 ORNL-2890 Period Ending October 31, 1959 ORNL-2973 Periods Ending January 31 and April 30, 1960 0RNL-3014 Period Ending July 31, 1960 0RNL-3122 Period Ending February 28, 1961 0RNL-3215 Period Ending August 31, 1961 0RNL-3282 Period Ending February 28, 1962 ORNL-3369 Period Ending August 31, 1962 ORNL-3419 Period Ending January 31, 1963 0RNL-3529 Period Ending July 31, 1963 ORNL-3626 Period Ending January 31, 1964 THE PLACE OF MOLTEN-SALT .REACTORS IN THE AEC PROGRAM E. E. Sinclair This meeting was organized so that interested parties may learn about the status of molten-salt reactor technology and inspect the Molten-Salt Reactor Experiment (MSRE) before it goes into operation. As you know, the MSRE is just about ready for prenuclear operation, and the timing for this meeting was selected with that fact in mind. As a representative of the Atomic Energy Commission, which is sup porting the development of molten-salt reactor technology, I have agreed to describe "the place of molten-salt reactors in the AEC program." As a matter of fact, this is very easily done by reference to the 1962 AEC Report to the President, which delineates the objectives of this Nation's reactor development programs. In discussing the program for the long- range future (i.e., the development of breeders which will fully utilize nuclear resources), the Report states in part: "Although breeding in the thorium-uranium 233 cycle can build upon experience gained with less ad vanced reactors ... vigorous and specific efforts will be required to at tain breeding on a significant scale. Both fuel and blanket systems must be pushed. Attention should be directed at methods of continuous removal of fission products, including the use of fluid fuels (such as fused ura nium salts) and blanket materials. Experimental reactors designed to breed must be built and operated. Hopefully, within the next several years, the program will achieve the stage where operating prototypes will be appropriate." From this statement and the AEC support of the molten-salt program here at ORNL, it is evident that the AEC looks to molten-salt reactor technology for the near-term accomplishment of breeding on a significant scale in the Th-233U cycle. There is, of course, no obvious reason why molten-salt technology could not be- developed for fast breeding in the uranium-plutonium cycle, and perhaps this will someday be done. The salt system that has been developed here, however, quite naturally lends itself to thermal breeding in the Th-233U cycle; and it is the objective of this program to develop and demonstrate a reactor system which will exploit this capability. Although the forthcoming operation of the MSRE will not achieve this objective alone, it is an exceedingly important and necessary step toward the goal. The MSRE, being a small single-region machine, is not designed to breed, but it does tie all the pertinent technology developed thus far together into a relatively simple operating system. Its operation is ex pected to supply the information' and confidence needed to proceed with the next and, perhaps final, step of the demonstration program: the de sign, construction, and operation of a two-region thermal breeder suffi ciently large to permit scale-up to commercial sizes^ It would be wishful thinking to suppose that there will be no "bugs" in the MSRE as designed and built. The history of reactor development suggests that there will be bugs or minor difficulties unforeseen in de sign or caused by errors in construction. These are unwanted but are more or less expected. It is also possible that operation of the MSRE will uncover some truly unexpected major problems. I personally do not expect any, because the technology base that has been worked out for the MSRE ' is very broad and thorough. In my view the MSRE is much better off in this respect than has been the case for some past reactor experiments. Today and tomorrow, this technological base which has been developed for the MSRE will be explained in detail. I hope you are as favorably impressed as I am, by the broad coverage, yet thoroughness, of the devel opments that have been conducted here at ORNL. I venture to predict that there will be another information meeting here in the not too distant future to introduce the operation of a molten-salt breeder prototype. I hope to see you all here at that time. MOLTEN-SALT POWER REACTORS AND THE ROLE OF THE MSRE IN THEIR DEVELOPMENT R. B. Briggs Realization of a system that makes full use of the potential energy in thorium to produce cheap electricity is the primary mission of reactor development at the Oak Ridge National Laboratory. That system must be an efficient breeder system. An advanced converter may be a worthwhile step in the development, but an advanced converter does not reach the goal. No matter how good the conversion ratio, if it is significantly less than 1, the amount of uranium that must be mined to make up the def icit in fissionable material is greater than the amount of thorium that must be mined to compensate for the thorium converted to 233U and burned. For example, if the conversion ratio is 0.90, the 235U from 20 tons of natural uranium will be burned with each ton of thorium consumed. Even with a conversion ratio of 0.99, the 235U from 2 tons of uranium must be supplied with each ton of thorium. One feature of the Th-233U fuel system is that breeding should be possible with thermal and intermediate reactors as well as with fast re actors. Many technical and economic factors combine to favor the thermal breeders; so we have chosen to emphasize them in our program. Our studies lead us to believe that molten-salt reactors are the most promising of the several possible thermal breeder reactors for achieving a satisfactory breeding gain and producing cheap electricity. Our estimate1 of the breeding performance of a two-fluid, graphite- moderated thermal breeder is shown in Table 1. The neutron yields and losses are known with less accuracy than the numbers imply, but the table was prepared to show some of the smaller losses and to balance. We see that the uranium and thorium concentrations and the ratio of uranium to carbon atoms can be adjusted to obtain a good balance between the neutron yield and the parasitic absorptions in moderator and in car rier salts. This leaves a net of about 12 neutrons per 100 neutrons ab sorbed in fuel for production of excess 233U. The losses to 233Pa can Table 1. Performance Estimate for Molten-Salt Breeder Reactor Fuel salt composition, mole $ 0.«FF4-63LiP-36. 6BeF2 Blanket salt composition, mole $ 15ThF4-67LiF-l8BeF2 Moderator-fuel ratio, N(C)/n(U) 5100 Volume fraction of fuel salt in core 0.16 Volume fraction of fertile salt in core 0.071 0.070 0.069 0.068 0.068 0.066 0.065 0.066 Thorium inventory, metric tons for 270 140 1000 Mw (electrical) Fertile stream cycle time, days 35 35 35 50 100 . 200 200 50 Fuel stream cycle time, days 11 23 31 55 55 72 84 50 Total 233U, 235U, and 233Pa inventory, 880 860 860 900 1020 1220 1220 880 kg Net neutron yield, ^e 2.2137 2.2131 2.2128 2.2120 2.2115 2.2103 2.2097 2.2095 Neutron losses: 233- U absorptions 0.9172 0.9150 0.9139 0.9111 0.9095 0.9053 0.9034 0.9021 235- U absorptions 0.0828 0.0850 0.0861 0.0889 0.0905 0.0947 0.0966 0.0979 > 232Th fission 0.0019 0.0019 0.0018 0.0018 0.0018 0.0018 0.0018 0.0018 233Pa absorptions (x2) 0.0120 0.0117 0.0116 0.0113 0.0112 0.0108 0.0106 0.0214 236U absorptions 0.0106 0.0115 0.0119 0.0132 0.0140 0.0164 0.0176 0.0165 237Np absorptions 0.0004 0.0010 0.0010 0.0008 0.0008 0.0010 0.0011 0.0010 135Xe absorptions 0.0050 0.0050 0.0050 0.0050 0.0050 0.0050 0.0050 0.0050 Sm absorptions 0.0001 0.0001 0.0001 0.0002 0.0003 0.0005 0.0005 0.0003 Other fission product absorptions 0.0218 0.0267 0.0317 0.0422 0.0442 0.0532 0.0631 0.0478 Corrosion product absorptions 0.0008 0.0008.