The objective of this test is to determine the DR-Reactor effluent systems characteristics under 95 degrees Celsius bulk temperature operation. This proposed bulk temperature increase from 93.5 to 95 degrees represents a 33% decrease in the bulk temperature suppression below the boiling point. A major aim of this test will be to evaluate the degree of increased maintenance at this higher temperature operation. The basis and justification, test preparation and instrumentation, procedure, costs, outage time, hazards, standards, and responsibilities are discussed in this document.

A short-term study was made to investigate the removal of radioiodine vapor from air-steam mixtures as could occur immediately following a nuclear incident in a pressurized water reactor. Activated charcoal traps, designed to simulate a small section of a commercial charcoal canister, were tested at gas velocities of 23.9 to 74.9 ft/min over the temperature range of 75 to 118°C. The iodine removal efficiency was found to range from 99.80 to 99.4% with an average of 99.9%. One test performed at a temperature of 105°C and with gas velocity of 290 ft/min indicated that the iodine removal efficiency was reduced to 99.54%. Activated charcoal exhibits a high efficiency for iodine vapor removal under these test conditions and appears suitable for application in the decontamination of air-steam mixtures.

Supersedes ITR-1421. Twelve large-diameter buried conduit sections of various shapes were tested in the 60- to l49-psi overpressure region of Burst Priscilla to make an empirical determination of the degree of personnel protection afforded by commercially available steel and concrete conduits at depths of burial of 5, 7.5, and 10 feet below grade. Essentially, it was desired to assure that Repartment of Defense Class I, 100psi and comparable radiations, and Class II, 50-psi and comparable radiations, protection is afforded by use of such conduits of various configurations. Measurements were made of free-field overpressure at the ground surface above the structure; pressure inside the structures; acceleration of each structure; deflection of each structure; dust inside each structure; fragmentary missiles inside the concrete structures; and gamma and neutron radiation dose inside each structure. All buried conduit sections tested provided adequate Class I protection for the conditions under which the conduits were tested. Standard 8-foot concrete sewer pipe withstood 126-psi overpressure without significant damage, minor tension cracks observed; standard 10-gage corrugated-steel 8-foot circular conduit sections withstood 126- psi overpressure without significant damage; and standard 10-gage corrugated- steel cattle-pass conduits withstood 149-psi overpressure without significant damage. Durations of positive pressure were from 206 to …

The fixation of radioactive waste in a stable solid media by means of calcination of these aqueous solutions has been the subject of considerable-effort throughout the U. S. Atomic Energy Commission and by atomic energy organizations in other countries. Several methods of doing this on a continuous or semi-continuous basis have been devised, and a fev have been demonstrated to be feasible for the handling of non-radioactive, or low-activity, simulated wastes. Notable among methods currently under development are: (a) batch-operated pot calcination of waste generated from reprocessing stainless steel clad fuel elements (Darex process) and Purex waste, (b) combination rotary kiln and ball mill calcination of aluminum nitrate (TBP-25 and Redox process), and (c) fluidized bed calcination of TBP-25 and Purex wastes. Although a considerable amount of engineering experience has been obtained on the calcination of dissolved salts in a fluidized bed, and the other methods have been the subjects of a great deal of study, none of them have been developed to-the extent which would rule out the desirability of further investigation of other possible methods of calcination.

This report describes the work carried out in 321 Building semiworks equipment, to define the factors contributing to high corrosion-product contamination and presents recommendations for reducing the impurity level to meet current specifications (maximum of 200 parts total metals per million parts U).

A technical and economic evaluation of a mercury-cooled fast breeder reactor is presented. The objectives of the program were to establish the technical feasibility of a fast breeder reactor cooled with boiling mercury and to evaluate the long-range potential of such a reactor power plant for production of economic power. Details of the conceptual design of a 100-Mw(e) reactor and system are discussed. The power cost from a mercury cooled fast breeder reactor was estimated as 21.4 mills/kwh which is competitive with the power cost for the initial Enrico Fermi plant. It was concluded that this reactor concept is technically feasible and has promising long-range economic potential. (M.C.G.)

The current status of the NPR fuel fabrication and testing program has been reviewed; suggestions have been made concerning areas requiring more information. This report is divided into a series of sections, each of which is a complete entity so a specific problem area can be considered apart from the entire NPR Fuel process. The final section incorporating conclusions and recommendations reaffirms the factors of major importance as well as evaluating those areas which are interrelated. Process problems have been considered from the aspect of manufacturing, not engineering.

Conclusions from the study: "Radioactive dusts are carcinogenic, and can cause cancer of the lung. complete dose response curves have not yet been determined. All the parameters that render this demonstrably toxic material (radioactive dust) have not yet been evaluated. It is strongly suggested by the experimental data that duration of radiological insult to the lung is an important factor in eliciting lung cancer. The atmospheric tolerance concentrations now in use seem to afford little margin of safety."

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An investigation was conducted to establish an operating procedure for flushing water through a purification system demineralizer which was out of service for an extended period, and to determine the demineralizer serviceability. (J.R.D.)

A controlled radiation experiment was performed to determine the stability of fuel containing 28 wt.% UN dispersed in and clad with Type 318 stainless steel as compared with fuel containing 30 wt.% UO/sub 2/ dispersed in the same material. The specimens were prepared by hot rolling the fuel coupons in Type 318 stainless steel using the picture-frame technique for initial bonding and reduction. Final dimensions were obtained by cold rolling. A special radiation capsule was designed which contained heat control and enough thermocouples to ensure a good continuoustemperature history throughout the test. This capsule and the method by which the specimens were loaded are discussed in some detail. Because of the capsule instrumentatton. the known MTR position into which the capsule was placed. and the dostmeters placed tn the capsule it was possible to obtain a complete flux and temperature history of the capsule during the irradiation. When it was estimated that the specimen burnup was about 7.2 at.% of uranium-235 the capsule was removed from the reactor and returned to the Battelle Hot-Cell Facility. The postirradiation examination which consisted of fission-gas analysis, density and dimensional determinations, radiochemical and isotropic burnup analysis, and metallography is discussed completely in the report. …

Tests in which aluminum-jacketed, Al-Si bonded uranium fuel elements were baked at various temperatures have shown there is a time-temperature relationship for Al-Si layer decomposition. For heat transfer and secondary coolant barrier considerations, the extent of bonding layer deterioration during fuel element irradiation is important. Under present reactor operating conditions, Al-Si bonded fuel elements show evidence of internal bond deterioration, and to a lesser degree, external bond deterioration following irradiation. Such evidence has aroused concern for the ability of the Al-Si bonding layer to withstand future reactor operating conditions. Currently, several potential uranium fabrication and canning process improvements are being developed to further advance fuel element stability and performance. Optimization of process conditions based on these improvements may provide the necessary Margin of safety for good bond layer integrity during irradiation. Before a decision can be made to continue improvement of the present process or convert to a new canning process, more information on the stability of the present fuel element bond is needed. The purpose of this report is to summarize data derived from past and present testing and to recommend courses of action to more fully evaluate Al-Si bond integrity under anticipated future reactor operating conditions.

Report issued by the Brookhaven National Laboratory discussing fuel reactor test loops. The "design, construction, operation, and experimental results" (p. 1) are presented. This report includes tables, illustrations, and photographs.

This report summarizes investigations taken between October 1, 1959, to September 30, 1960 with the objectives of developing high-tungsten alloys which are readily fabricable into massive and complex shapes and which possess adequate strength for operation at temperatures up to 2000 F. Within this report, studies are divided into two main categories: alloy development and material fabrication techniques

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Summary statements are given of significant findings related to the distribution characteristics, and biological availability of fall-out debris originating from testing programs at the Nevada Test Site during the past decade. The delineation of fall-out patterns has been accomplished by the use of aenial and ground monitoring surveys. Only about 25% of the total amount of fission products produced by tower-supported detonations were deposited within distances corresponding to fall-out time of H + 12 hr; a much smaller quantity was deposited by halloon-supported detonations. Fall-out particles less than 44 (For in diameter are presumed to be of the greatest biological significance. About 30% of the fall-out radioactivity from tower-supported detonations was contained in the 0 to 44 (For particles as compared to almost 70% for balloon-supported detonations. Fall-out debris from halloon- supported detonations was also much more water and acid soluble than was the debris from towel-supported detonations. The <44 (For fallout particles contained a higher percentage of Sr/sup 89/, Sr/sup 90/, Ru/sup 10/ / sup 3/, and Ru/sup 106/ than did larger sized particles. There was a higher percentage of these radioelements in the particles from balloon-supported detonations. Within distances corresponding to H + 12 hr fall-out time, balloon- …

Summary statements are given of significant findings related to the distribution, characteristics, and biological availability of fall-out debris originating from testing programs at the Nevada Test Site during the past decade. The delineation of fall-out patterns has been accomplished by the use of aerial and ground monitoring surveys. Only about 25% of the total amount of fission products produced by tower-supported detonations were deposited within distances corresponding to fall-out time of H + 12 hr; a much smaller quantity was deposited by balloon-supported detonations. Most of the fall-out debris that was redistributed by various environmental factors after original deposition consisted of particles < 44 ..mu.. in diameter; the particles in this size range also represented the predominant contamination on plant foliage. /sup 90/Si levels in surface soil ranged from 31.9 to 142 mc/sq mile in virgin areas near known fall-out pattern midlines and from 7.5 to 22.7 mc/sq mile in agricultural areas that in some cases did not coincide with fall-out pattern midlines. The accumulation of radioiodine by native animals was observed to be a function of distance from GZ. /sup 140/Ba, /sup 91/Y, /sup 89/Sr and /sup 90/Sr were major bone contaminants. Post-series sampling of native animals indicated that …

A development program is being carried out in the pilot plant to improve the equipment involved in monitoring, metering and pumping process streams. The equipment under development includes an alpha monitor, a gamma monitor, a waste water monitor, a liquid flow rate meter, a canned rotor pump, a diaphragm pump, air lifts and a pressure tank pump. This report presents the present status of the testing program.

To permit the irradiation of one dummy fuel element assembly for one operating period and to permit, during a subsequent operating period, the irradiation of one defected, four-rod-cluster UO{sub 2} fuel element assembly, in a KE front-to-rear test hole. The fuel material is natural UO{sub 2} of 95 per cent theoretical density; the cladding is zircaloy. The defect in the assembly is artificial and will be made before irradiation by drilling a .005in. diameter hole through the cladding near the mid-point of two of the rods.

To permit the irradiation of one dummy fuel assembly for one operating period and to permit, during a subsequent operating period, the irradiation of one defected, four-rod-cluster, UO{sub 2} fuel element assembly, in a KE front-to-rear test hole. The fuel material is natural UO{sub 2} of 95 per cent theoretical density; the cladding is zircaloy. The defect in the assembly is artificial and will be made before irradiation by drilling a .005in. diameter hole through the cladding near the mid-point of two of the rods.

A study was made to determine the structural and functional adequacy of the EGCR graphite core design. Maximum stress and distortion of the core and the probable operating time before graphite cracking were determined. The major cause of stress is non-uniform fast-neutron flux, which causes non-uniform shrinkage of graphite components. The critical stress was found to be tensile. The criterion for determining the time of cracking of graphite columns appears to be the point at which the creep strain results in rupture. Column life before cracking may also be related to the maximum allowable stress at rupture. It was concluded that non-uniform shrinkage is the major cause of distortion of graphite components of the core. In general the amount by which a column tends to bow during 20-year core life exceeds the available free space. The combined internally generated and bowing stresses result in an average period before cracking of all core columns of 14 years, with a minimum period of 6 years. (M.C.G.)

The reactivity control capacity of the N-reactor control systems has been calculated by methods which take into account the absorption of epithermal as well as thermo neutrons. Earlier calculations were made using a method which is satisfactory in control calculations for the existing Hanford reactors but which largely neglects the epithermal absorption. The new calculations were undertaken because of recent evidence that the epithermal absorption is more important to the NPR control strength and because of some concern regarding the margin between previously calculated control strength and anticipated control requirements. The calculational methods used in the present study and the resulting calculated control capacities are reported in this document.

In recognition of the need for an integrated system plan for future development requirements for radiological instrumentation, this information report was written to provide the initial step concerning such a planning. The report is to be considered a basic information analysis of the present in-use plant radiological instrumentation, of the instrumentation now undergoing active development, and of the future development work contemplated.