tions. Methods will be evolved for charging and discharging fuel elements through the pressure shell.

The Westinghouse Electric Corp. is responsible for the development, design, and construction of the nuclear portion of the PWR project. Stone and Webster Engineering Corp., Boston, Mass., was selected to perform architect-engineering services for the same portion of the project as a subcontractor under Westinghouse.

The plant will be built in the vicinity of Shippingport, Pa., on the Ohio River, 25 miles northwest of Pittsburgh. It will be operated by the Duquesne Light Co. of Pittsburgh, whose proposal to participate in the project was accepted by the Commission as the most favorable to the Government of 9 major offers submitted. The terms of Duquesne's proposal are to: Furnish a site for the entire project and build and operate a new electric generating plant at no cost to the Government; operate the reactor part of the plant and bear the labor costs thus entailed; assume $5 million of the cost of research, development, and construction of the reactor portion of the plant; buy steam from the Commission, and waive any reimbursement by the Government of costs incident to termination of the contract.

It is estimated that-along with revenues from the sale of steam generated by the reactor-the company's proposal would reduce by an estimated $30 million the expenditures the Government would have had to make during the period of construction and 5 years of operation if it had undertaken the full cost of the project.

Experimental Boiling Water Reactor

It was formerly believed that boiling within the core of a watercooled reactor would cause unstable operation. However, experiments conducted last year by the Argonne National Laboratory at the National Reactor Testing Station in Idaho, have indicated this to be untrue. In addition, these experiments may have an important bearing on the safety of industrial power reactors, for it is possible that boiling reactors may be designed to operate in a stable self-regulating manner, shutting themselves down without serious damage in case of trouble.

The tests at NRTS during the summer of 1953 consisted of setting up and operating a water-cooled and moderated reactor and imposing conditions normally expected to cause a "run away." It was previously assumed that under these circumstances the core would melt and allow the escape of fission products.

However, these things did not happen in last summer's tests. Experimental results with a small, temporary water-cooled and moder

ated reactor showed that power excursions could take place quite rapidly and neither produce melting of the fuel nor radioactive contamination of the surroundings. In experiments which allowed the power to rise to several thousand kilowatts in a fraction of a second, the steam cut off the nuclear reaction completely before a dangerous temperature was induced. More boiling experiments are to be conducted by Argonne at NRTS in the summer of 1954.

Perhaps this mechanism applied to power reactors could be a safety device making it impossible for the nuclear reaction to create dangerously high temperatures. Further, a reactor used as a direct source of steam in a power plant might reduce capital costs by eliminating the heat exchanger (steam boiler) outside the reactor, by reducing the pressure of the primary coolant system, and the pumping power required for the system.

Plans called for starting construction during the year ahead of an Experimental Boiling Water Reactor designed to produce 20,000 kilowatts of heat and 5,000 kilowatts of electricity, to be completed in the winter of 1956. The reactor will make use of normal uranium fuel plus enriched uranium 235 and will be moderated and cooled by ordinary (light) water. As with the PWR, the slight enrichment of fuel is necessary to bring a light water moderated reactor to criticality.

An important aspect of tests with the Experimental Boiling Water Reactor will be to determine whether it can be operated without troublesome or hazardous deposit of radioactivity in the turbine, condenser, feedwater pumps, or other equipment outside the reactor. The deposited radioactivity might cause major maintenance problems in the event of failure of the equipment.

Sodium Graphite Reactor

Although a proposed sodium-cooled, graphite-moderated reactor has been determined to be feasible in principle, there are still numerous features of such a plant significant to the economies of the system that have not been tested in actual practice. For example, these include the upper limits for fuel and coolant temperatures, burn-up, and other operating variables.

North American Aviation, Inc., the major contractor exploring the sodium-graphite approach, continued its investigation in this field. The firm has agreed to contribute $2.5 million of the $10 million required for development, constructio., and operation of a graphitemoderated Sodium Reactor Experiment (SRE) during the period 1954-58. The SRE is planned to produce 20,000 kilowatts of heat and would not be equipped with a turbogenerator.

A full-scale reactor, using slightly enriched uranium fuel, would be expected to have a regeneration ratio of 0.9. If charged with uranium 233 and thorium, it is believed it would have a ratio slightly greater than one and operate as a power breeder producing more uranium 233 than it burns.

Tests with the Sodium Reactor Experiment are planned to include fuel element performance, maximum permissible fuel element and structure temperatures, and corrosion and radioactive transfer. The reactor's temperature and specific power will be increased gradually to determine performance limitations.

Completion of construction and the beginning of experimental operation are scheduled for 1956.

Homogeneous Reactors

The potential advantages of homogeneous reactors include: Low cost due to simplified design, economical chemical processing, and elimination of fuel element fabrication.

Oak Ridge National Laboratory continued the operation of Homogeneous Reactor Experiment No. 1 during the early months of 1953-54. This first power reactor of its type circulated uranyl sulfate fuel solution at nearly 500° F. under a pressure of 1,000 pounds per square inch and at a power density of 30 kilowatts per liter. It had a heat output of 1,000 kilowatts and a small turbogenerator.

In testing the self-regulation of HRE No. 1, the reactor's power was purposely increased as rapidly as possible, but the reactor did not "run away." Its power rose quickly to an abnormally high but very short peak and then leveled off at a moderately high value as the expanding fuel solution caused the core to lose reactivity. This reactor also demonstrated safe operation after its fuel solution became highly radioactive from fission products, and while a mixture of hydrogen and oxygen was formed by irradiation-produced decomposition of the solution water.

The HRE No. 1 was dismantled in the spring of 1954 to make way for the construction-in the same building at Oak Ridge-of Homogeneous Reactor Experiment No. 2, which should be completed and put in operation early in 1956.

The HRE No. 2 will have a heat output of 5,000 kilowatts. Its primary purpose is to produce a simplified, mechanically reliable plant demonstrating operability and reliability over a long period under conditions more closely simulating those of a full-scale reactor. It will furnish steam for a small turbogenerator, as did HRE No. 1, and will dissipate excess steam. The plant will include chemical processing

equipment for purification of the fuel solution by removing fission products. It will also seek additional information on the effect of irradiation on the corrosion of materials and on chemical stability of the fuel solution.

In the development of the homogeneous type, corrosion tests without irradiation demonstrated the compatibility of a dilute uranyl sulfate solution with a number of materials at high temperatures. Data on the effects of varying temperatures, salt concentration, acidity, and solution velocity were also obtained. The Oak Ridge National Laboratory completed a full-scale model of an "in-pile loop," which consists of an assembly of piping, pumps, and instruments for the purpose of circulating a fuel solution past samples of different materials while they are under intense neutron bombardment in the Low Intensity Test Reactor at ORNL and the Materials Testing Reactor. New information obtained from these and other investigations will be applied to the design of a Homogeneous Thorium Reactor now in the planning stage. The HTR is to produce about 65,000 kilowatts of heat, of which some 16,000 kilowatts will be converted into electricity. It will have a blanket of thorium from which uranium 233 will be produced.

Although the core diameter of the Homogeneous Thorium Reactor will not be as large as that of a full-scale plant, the thickness of the thorium blanket and the concentration of fertile material will be the same as for a central power station of this type. In addition, two chemical plants-one for removing fission products from the fuel solution and another for separating the uranium 233 from the thorium blanket-are to be an integral part of the plant.

Following development and design of the HTR by ORNL, construction is expected to begin during fiscal year 1957 and be completed in fiscal year 1959. The reactor will probably start operating with uranium 235 and change later to uranium 233, as this synthetic fissionable material is produced in the blanket.

Experimental Breeder Reactor No. 2

After 2 years of operating experience with the Experimental Breeder Reactor No. 1 at the National Reactor Testing Station in Idaho, the AEC plans to build a much larger unit, the Experimental Breeder Reactor No. 2. The EBR No. 1, developed and operated by Argonne National Laboratory, was the first reactor to produce electricity and to demonstrate breeding.

The EBR No. 2 is planned as a scale-up to 62,500 kilowatts in heat power and 15,000 kilowatts in electrical generating energy. The

EBR No. 1 supplied 1,400 kilowatts of heat and 170 kilowatts of electricity.

The EBR No. 2 will closely resemble a large central station powerbreeder reactor in power, control, fuel handling, and other features. Much of the equipment for a full-scale powerplant will also be usedsuch as pumps, heat exchangers, valves, and flow meters. In addition to providing information on engineering features, the EBR No. 2 will be operated to develop fuel-handling techniques, power-cycle conditions, core and blanket concentration, and component designs.

Heat transfer and mechanical components under simulated operating conditions will be tested in mechanical mockups of the EBR No. 2, scheduled to be built at the ANL during 1955. The startup of the reactor itself is planned for 1958.

Plutonium will be used as fuel in EBR No. 2 although it may be necessary to supplement it with uranium 235 for metallurgical reasons. However, the fuel charge will be changed to plutonium later. The blanket will be composed of natural or depleted uranium 238 which will be transmuted into piutonium. A greater production of plutonium will be achieved by using this element as a fuel than by using uranium 235. The objective will not be to produce a large quantity of fissionable material, but rather to test engineering features of the reactor and its auxiliaries.

The Experimental Breeder Reactor No. 1 will continue in operation. Its contributions in fast neutron reactor physics and radiation damage are expected to continue.

INDUSTRIAL PARTICIPATION

Consistent with its policy of encouraging industrial participation in the development of economic nuclear power, the Commission approved 5 new study group agreements with commercial firms during the first 6 months of 1954. This brings to 13 the total number of participating study teams, each of which represents one or more contractors.

Since the initiation of this program in 1951, industrial contractors, through their teams or groups, invested (by the end of 1953) an estimated $3 million surveying reactor technology, making preliminary designs and economic studies, and carrying on research and development. It is believed that this figure will reach $8 million by the end of 1954.