between 500° and 600°, and steam pressure of about 600 pounds per square inch. The PWR core will be about 6 feet in diameter and 712 feet high and will require a pressure vessel about 9 feet in diameter and 28 feet high. A fuel charge will consist of 15 to 20 tons of slightly enriched uranium, that is, uranium containing 1.5 to 2 percent of the 235 isotope rather than natural uranium which contains only 0.7 percent of this isotope. A pressure of 2,000 pounds per square inch will keep the cooling water from boiling. Other preliminary specifications include reactor power, 264,000 kilowatts of heat; maximum heat flux, 350,000 British thermal units per square foot per hour; average power density, 45 kilowatts of heat per liter; and average specific power, 1,000 kilowats of heat per kilogram of fissionable material. EXPERIMENTAL BOILING WATER REACTOR Some years ago use of a reactor as the direct source of steam for the turbine was suggested as an attractive way of making power. This arrangement would eliminate the need for a heat exchanger (steam boiler) outside the reactor and permit appreciable reduction in pumping power, and hence should lead to lower capital costs. However, it was thought that boiling in the core would cause continual changes in reactivity and might result in unstable operation. In the summer of 1953, experiments with a small temporary reactor, conducted by Argonne National Laboratory at the National Reactor Testing Station in Idaho, demonstrated that these fears are not justified. It may be possible to design boiling reactors which will operate in a stable, self-regulating manner, and which in the event of trouble will shut themselves down without serious damage. These developments constitute a major contribution toward safe power reactors. An experimental boiling water reactor of about 20,000 kilowatts of heat and 5,000 kilowatts of electricity is planned to explore further the possibilities revealed by the investigations in 1953. Specifications are being established by the Argonne Laboratory. The boiling reactor will be fueled with enriched uranium and moderated and cooled with ordinary (light) or heavy water. The uranium enrichment is necessary to make any natural uranium-light water reactor critical. Enrichment is also needed for a heavy water reactor of small size, like the EBWR, but not for a large boiling reactor. An important purpose of the experimental boiling water reactor is to determine whether it can be operated without significant deposit of radio-activity in the turbine, the condenser, and the feed water pumps. Such deposits might cause major maintenance problems in case of equipment failure. Assuming success with this boiling reactor, tentative specifications for a full-scale central station plant of this type have been estimated. However, EBWR is not expected to provide nuclear data on the critical mass of the large reactor or on the proper spacing of its fuel. Such information would be obtained from critical experiments. The Atomic Energy Commission is selecting an architect-engineering contractor and a site for the facilities required for experimental boiling water reactor. The schedule calls for completion of the reactor and facilities during the latter part of calendar year 1956. The program for advancing the technology of the sodium-cooled, graphite-moderated type of reactor centers about a preliminary design for a full-scale power plant. This design uses metallic fuel elements of either slightly enriched uranium or a combination of thorium and uranium 233. With slightly enriched uranium fuel, the full-scale reactor is expected to have a regeneration ratio of about 0.9, producing plutonium as a byproduct. Charged with uranium 233 and thorium, the reactor should have a ratio slightly greater than one and thus operate as a power breeder, producing more uranium 233 than it consumes. Although concepts incorporated in the full-scale design have been determined to be feasible by North American Aviation, Inc., the chief contractor developing this type of reactor, many features of the proposed plant and its operating procedure have not been tested in reactor practice. Neither are the upper limits known for fuel and coolant temperatures, burnup, and other operating variables. Moderate changes in some of these variables, such as increasing maximum uranium metal temperature from 1,200° to 1,400° F. and maximum coolant temperature from 1,000° to 1,250° F., will have appreciable effect on the cost of power. A small sodium reactor experiment is planned to obtain information needed for evaluating the possibilities. This unit will have a heat power level of about 20,000 kilowatts, but it will not be equipped for generating electricity. The heat produced will be exhausted to the atmosphere through a relatively inexpensive sodium-to-air heat exchanger. The sodium reactor experiment will resemble the design for a fullscale plant in important respects. For example, both designs call for tank-type reactors, both have the entire reactor structure below ground level, and both use similar fuel arrangements. Tests possible with the SRE include fuel element performance, maximum permissible fuel element and structure temperature, and corrosion and radio-active transfer. The reactor's temperature and specific power will be increased gradually to determine performance limitations. Test "loops" circulating sodium can be installed in the SRE to determine the effect of radiation on aspects of sodiumgraphite technology. The schedule for the sodium reactor experiment calls for completion of fabrication and beginning of experimental operation in calendar year 1956. EXPERIMENTAL BREEDER REACTOR NO. 2 Two years of operating experience with the experimental breeder reactor at the National Reactor Testing Station provides the basis for scaling up to a larger unit, called experimental breeder reactor No. 2. The scale up planned is from 1,400 to 62,500 kilowatts in heat power output and from 170 to 15,000 kilowatts in electrical generating capacity. Fuel and coolant temperatures will be substantially higher and steam pressure will be correspondingly greater. In fact, the temperatures and steam pressure will be the same for EBR No. 2 as are now visualized for a full-scale power-breeder reactor. The new reactor will also be similar to a large central station unit in power density, control, and fuel-handling features. It will include pumps, heat exchangers, valves, flow meters, and other "hardware" of sizes suitable for a full-scale powerplant. In addition, as an experimental reactor, EBR No. 2 will test advanced ideas for longrange application such as variations in core and blanket concentration, in fuel-handling techniques, in power-cycle conditions, and in component design. Operation of the first EBR, now designated No. 1, will continue to contribute to the fast-breeder research and development program valuable information on physics, radiation damage, chemical processing, and fabrication techniques. A neutron source reactor is in operation at Argonne National Laboratory to provide neutrons for reactor physics measurements on subcritical core arrangements. Results are being used to design a critical assembly for construction at the testing station in Idaho during the current calendar year. The critical assembly will provide information on such factors as critical mass, breeding ratio, and power distribution in the core. Plans call for a mechanical mockup to be built at Argonne during fiscal year 1955 to test heat transfer and mechanical components under simulated operating conditions. Components to be tested include loading and unloading devices, control mechanism, heat exchangers, boilers, and superheaters. EBR No. 2 probably will be loaded first with uranium 235 and later with plutonium. The blanket in each case will consist of natural or depleted uranium whose uranium 238 will be transmuted into plutonium. Because of physical constants, greater production of plutonium is expected with plutonium fuel than with uranium 235 fuel. However, the purpose of this experimental reactor is to test engineering features rather than produce a maximum amount of fissionable material. Facilities must be built for developing, manufacturing, and processing partly used uranium and plutonium fuel elements and the irradiated uranium blanket containing plutonium. Startup of experimental breeder reactor No. 2 is planned for calendar year 1958. HOMOGENEOUS REACTOR EXPERIMENT NO. 2 An experimental reactor designated homogeneous reactor experiment No. 2 is to be the next major step in developing homogeneous type reactors which have their fuel and moderator in a water solution. Potential advantages of this type include low-cost chemical processing, elimination of fuel element fabrication and handling, and simplified reactor design. The homogeneous reactor experiment, now designated No. 1, at Oak Ridge National Laboratory, has demonstrated that a 1,000-kilowatt reactor, circulating uranyl sulfate fuel solution at nearly 500° F. under 1,000 pounds per square inch pressure and at a power density of 30 kilowatts of heat per liter, will operate with stable power output. HRE No. 1 has also shown that the reactor can be operated and maintained safely after its fuel solution becomes highly radioactive from fission products and while a mixture of hydrogen and oxygen is formed by irradiation-produced decomposition of the solution water. HRE No. 1 will be dismantled early in calendar year 1954 and HRE No. 2 assembled in the same building. The new reactor should be in operation by the summer of 1956. Homogeneous reactor experiment No. 2 will have a heat output of about 3,000 kilowatts as compared with 1,000 for its predecessor. Its primary purpose is to produce a simplified, mechanically reliable plant which will demonstrate operability and reliability over a long period under conditions closely simulating those of a full-scale reactor. The plant will include chemical processing equipment for the purification of the fuel solution by removal of fission products. The homogeneous development also seeks more information on the effect of irradiation on the corrosion of materials and on the chemical stability of the fuel solution. A long series of corrosion tests without irradiation has demonstrated the compatibility at elevated temperatures of a dilute uranyl sulfate solution with a number of materials. Quantitative data of the effects of varying temperature, salt concentration, acidity, and solution velocity have also been obtained. Corrosion and stability tests under irradiation will utilize the low intensity testing reactor at Oak Ridge and the materials testing reactor at the testing station. The equipment required for the tests includes closed "in-pile loops" of piping, pumps, and instruments which will circulate fuel solution past samples of different materials while they are under intense neutron bombardment in the test reactors. HOMOGENEOUS THORIUM REACTOR As the next step in developing homogeneous reactors, scale up to about 65,000 kilowatts of heat, of which about 16,000 will be converted into electricity, is planned for the homogeneous thorium reactor. This reactor is also aimed at demonstrating the production of uranium 233 from a blanket of thorium. The physical constants of the uranium isotopes and of thorium and plutonium make the generation of uranium 233 from thorium attractive for thermal reactors in which fission is primarily by slow, or thermal, neutrons, whereas the production of plutonium from uranium 238 can be accomplished most readily in a fast reactor. Though the core of the homogeneous thorium reactor will not be as large in diameter as that of a full-scale plant, it is planned to have the same blanket thickness and concentration of thorium as a large central station reactor of this type. Two chemical plants, one for removing fission products from the fuel solution and the other for separating the uranium 233 from the thorium blanket, are contemplated as integral parts of the nuclear powerplant. Following development and design, construction is tentatively scheduled to begin during fiscal year 1958 with completion in fiscal year 1959. The HTR probably will start operating with uranium 235, which eventually can be replaced with uranium 233 produced in the blanket. UNITED STATES ATOMIC ENERGY COMMISSION, Hon. STERLING COLE, Chairman, Joint Committee on Atomic Energy, DEAR MR. COLE: Attached is the statement of Dr. Zinn's which Mr. Strauss said you had asked about. Dr. Zinn brought the statement with him to Washington in response to a request from a member of your staff, with a view to having it declassified. The statement has been revised to permit its publication. Sincerely yours, K. D. NICHOLS, General Manager. STATEMENT BY DR. WALTER H. ZINN CONCERNING AN EXPERIMENT USING THE BOILING REACTOR PRINCIPLE The Argonne National Laboratory has carried out for the Atomic Energy Commission certain experiments which it is expected will have a vital bearing on the question of the safety of operation of industrial power reactors. Since safety is an important factor in the evaluation of the use of atomic energy for the generation of economically competitive electricity, it has been decided to make this information generally available. The experiments were done during the summer of 1953 at the National Reactor Testing Station, Arco, Idaho. A team of scientists and engineers from the Argonne Laboratory at Lemont, Ill., and the staff of the experimental breeder reactor project in Idaho carried out the work. Principal members of the team were S. Untermyer, J. R. Dietrich, D. C. Layman, H. V. Lichtenberger, and W. C. Lipinski, working under the laboratory director, Walter H. Zinn, as leader. The experiment consisted of setting up and operating a nuclear reactor and then imposing conditions on the reactor which would make it "run away." This means that the power of the machine was caused to rise precipitously and was allowed to continue to rise indefinitely. Under such conditions it had been assumed in the past that the core of the reactor would melt and that this would permit the escape of radio-active fission products. It is this particular assumed circumstance which has governed decisions concerning locations of nuclear reactors and which has required an uninhabited, restricted area surrounding them of an acreage which is determined by the power of operation. This particular reactor was moderated by water and cooled by water. The experiment showed that power excursions of very large magnitude and which took place quite rapidly did not produce melting of the fuel and no radioactive contamination of the surroundings whatsoever resulted. The favorable effects observed were anticipated and are due to the particular design of the reactor, 15 |