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A program was initiated on the spectrophotometric study of aqueous solution chemistry. The goal is operation at temperatures up to at least 330 deg C and at pressures up to 200 atm, and to near the critical point if this appears to be feasible. A spectrometer capable of operation under these extreme conditions is being designed. (W.L.H.)

The predicted behavior of German underground personnel shelters, equipment, and certain instrumentation was investigated. Data obtained will be used for evaluation and improvement of present design criteria. Nine reinforcedconcrete underground shelters, designed by German engineers, were tested at the 170-, 155-, 110-, 78-, 26-, 11.5-, and 7.2-psi overpressure ranges as determined from average blast-line instrumentation measurements. Reinforcing steel, doors, and ventilation equipment were received and incorporated in the shelters. Preshot and postshot precise location surveys were made to determine the total lateral and vertical motions of the structure as a result of the blast. Blast instrumentation used in the shelters and entranceways consisted of pressure gauges, earthpressure gauges, self-recording pressure gauges, and dynamic pressure gauges. Free-field measurements were recorded along the blast line using U. S. self-recording and electronic pressure gauges and German self- recording pressure gauges. Structural response was recorded by deflection and acceleration gauges, strain gauges, and scratch gauges. Radiation measurements were taken using U. S. gamma-radiation film dosimeters, gamma-radiation chemical dosimeters, neutron detectors, telemetering gamma dosimeters, and German gamma chemical dosimeters. Mice were used as biological specimens in environmental tests in seven of the nine structures tested. In addition to the environmental tests, a series of tests …

High-temperature x-ray diffraction experiments provided confirming evidence regarding the origin of the metastable gamma phase and of the general features of the zirconiumhydrogen phase system. These experiments also showed that the reaction alpha + delta -- beta (the eutectoid reactlon on heating) proceeds at an extraordinarily slow rate, although the reverse reaction proceeds so rapidly that beta phase cannot be retained to room temperature by quenching. The sluggishness of the eutectoid reaction on heating undoubtedly resulted in erroneous interpretation of certain physical and mechanical properties determined of the mechanical properties of zirconium hydride containing various minor alloy additions is continuing. Although these data are still quite sketchy, alloy additions of 5 at.% Sc or 2.2 at.% Cu show the most promise at this time. (auth)

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 …

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A calculation is made of the interaction of a beam of particles in an accelerator with the radio-frequency cavity that provides the accelerating mechanism of the machine. A Hamiltonian for synchrotron motion is employed that makes possible the simultaneous solution of Maxwell's equations and the Vlasov equation, so that a self-consistent distribution of particles in synchrotron phase space is determined. The effective voltage on the cavity due to the beam of charged particles is of the order of magnitude of the product of the total circulating current in the accelerator and the shunt impedance of the rf cavity. It has the net effect of producing a total voltage on the cavity which is both less than the applied voltage, and shifted in phase with respect to it. The increase in the stable phase angle required so the particles will remain in phase with the accelerating radio frequency is calculated. The decrease in total voltage and increase in stable phase angle result in a decrease in stable phase space available for acceleration, and convenient expressions are given for these quantities in terms of parameters of the accelerator. It is shown that the consequences of the induced voltage may be alleviated by …

The iron thermal shields of the Hanford reactors are cooled by means of water flow through thermal cooling tubes embedded in the shield blocks. The flow rate, temperature rise, and allowable pressure in the tubes and the conditions under which some of the tubes may be out of service are specified in the process standards. Shield heat formation and heat transfer calculations are necessarily based on broad assumptions and are therefore usually reliable only for order of magnitude and trend predictions. In order to specify condition of shield cooling under which the reactors may be safely and economically operated data must be available regarding operating temperature as related to the flow of cooling water through the shield. Some of the data on which the current standards are based have been determined by extrapolation to present conditions of measurements taken several years ago. The purpose of this test was to establish current data that may be used in updating the thermal shield coolant standards. Measurements were taken of the operating temperatures experienced in an uncooled thermal shield cooling tube in relation to the specific power of the adjacent process tube. Conditions were varied by adjusting the flow in the thermal shield …

In 1955 the Fuels Development Operation began irradiation testing of fuel elements in high temperature water. It was assumed that if a new reactor were built at Hanford, it would be cooled by high-temperature, pressurized water. Corrosion tests showed that aluminum-clad production fuel elements could not be used in high-temperature water. Therefore, while work to improve the resistance of aluminum to high-temperature water proceeded, the Fuel Design Operation began irradiation of stainless steel- and Zircaloy-2-clad fuel elements. During 1956 and 1957, stainless steel-clad elements were tested in the Materials Testing Reactor (MTR), Hanford H Reactor Loop, and the KE Reactor Recirculating (KER) Loops. During 1957, a coextrusion method for cladding uranium rods with Zircaloy-2 was developed. The first irradiation of Zircaloy-2-clad fuel from an off-site supplier began in late 1958. The objective of the irradiation was to study the dimensional stability of the fuel rods and a seven-rod fuel assembly. Two coextruded, seven-rod elements were irradiated in KER Loop l.

To date a great deal of emphasis has been placed on flattening the side-to-side and top-to-bottom flux distribution with only minor effort to improve the front-to-rear distribution. Minor variations in the front-to-rear distribution have been achieved by horizontal control rod and Supplemental control positioning. It has-been reasonably well established that the rupture potential for one tube charge increases markedly with higher specific power and temperature; thus there is a great deal of incentive to flatten in the front-to-rear dimension. Although flattening in this dimension will caure increased neutron leakage out of the reactor, this is compensated by increased conversion efficiency resulting from a more uniform exposure distribution within the tube charge. The purpose of this document is to describe the basic analytical methods and the techniques, of flattening front-to-rear through the integrated use of enrichment and poison material in combination with natural uranium, and to point out the requirements to insure that total control criteria is satisfied in the event of a water loss with this loading. For the purpose of this survey report an old reactor, 32-piece charge length, and a symmetrical front-to-rear distribution were considered; however, the methods given can be extended quite easily to different length and …

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.

Use of X-8001 Al alloy as cladding for Hanford reactors was initiated because of superior (laboratory) resistance to intergranular corrosion over that of C-64 alloy. However, since severe pitting attack was observed intermittently, an evaluation was carried out on X-8001 alloy fuel element cladding.

The loop was charged October 30, 1959 with seven 12-inch Natural Uranium Zircaloy-2 clad 7-rod clusters. The test was primarily for the new hot-headed method of end closure used on these elements. The loop was pressure tested at 4000 psi after it was charged. Initially the loop was held at low-temperature to study the buildup of oxygen (presumably from radiolytic decomposition of water). After startup the neutron activity held at slightly above normal but the strainer gamma activity was exceptionally low. Frequent additions of LiOH bombs were necessary to maintain the pH at 10.0 (previously it was 4--5 pH but was raised to pH 10 for this test.) After the temperature was raised to operating conditions the pH held nicely at 10. On November 16, 1959, the heat exchanger exit temperature thermocouple blew out resulting in depressurization of the loop and a reactor scram. Repairs were made during the outage and the loop was returned to normal operation. Eleven scrams occurred during the test period caused by 105-KE or 1706-KER.

Since startup in January 1945 through December 1959, 4 {times} 10{sup 9} gallons of radioactive liquid wastes have been discharged to cribs and trenches at HAPO by the Chemical Processing Department facilities in the 200 Areas. These wastes contained approximately 2.5 {times} 10{sup 6} curies of beta emitters. The Scavenged Waste Recovery Program was completed in 1957 and Redox Plant process changes were made during the latter part of 1958. These changes resulted in significant reduction in the amount of radioactive materials that have been discharged to the ground in subsequent years, from a maximum of 8.2 {times} 10{sup 5} curies in 1955 to 9 {times} 10{sup 3} curies in 1959. Although these large amounts of radioactive materials have been discharged to the ground, periodic waste report inventories include no reduction due to radioactive decay. The estimated depletion by radioactive decay is the basis of this 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.

The program`s objective is to develop the technical foundation for use of projection fuel elements through an accelerated testing program. Two types of projection fuel elements are being considered (1) Self-supported fuel elements for use in ribless aluminum or zirconium process tubes; and (2) Bumper fuel elements for use in ribbed process tubes. A further objective of the program is to determine the most desirable method of fabrication and type of support for projection fuel elements. The program must develop the self-supported concept in time to meet the proposed ribless zirconium process tube installation schedule. Parallel development of the bumper fuel element is an interim program to realize immediate gains in reduced hot spot rupture tendency, during the change over to zirconium tubes. The appended charts indicate the proposed accelerated testing and development schedules being followed. Testing of the self-support concept has indicated that the collapsible supports are adequate to maintain fuel centering, do not introduce additional corrosion problems resulting from the supports, and do in fact greatly reduce the incidence of hot-spot flow patterns. Reductions in rupture potential of at least a factor of 30 have been demonstrated at the 95% confidence level. Bumper fuel testing has indicated that …

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The purpose of this memorandum is to transmit revisions to the goal exposure plans of reference (1). The plans, herein, supersede the previous recommendations. Similar to the plans of reference (1), the exposures calculated from these plans are designed to maximize plant return on the basis that the only restraint on metal usage is that imposed by the economics of the plutonium production process and associated uranium cycle, as described and defined in references (2) and (3). When metal throughput is limited by ex-reactor physical capabilities to a level lower than the unrestrained requirements, exposures higher than those recommended in this document would be indicated, as explained in reference (4). In essence then, these plans represent minimum economical exposures for current production fuel types, on the basis of attempting to maximize plant returns. The adjustments recommended herein, stem solely from revised estimates of the metal performance level indices (C{sub R} values) of the various metal types.

The objective of this report is to summarize all in-reactor operating data associated with PT-288-A. This document will serve as a ready reference to the irradiation history of the fuel elements, and will record any anomalies encountered in the performance of this test.

One of the three major categories of HAPO fuel element failures is the side corrosion type rupture. The majority of side-corrosion failures has been characterized by oval or tear-drop shaped flow patterns containing evidence of accelerated corrosion. Thorough examination of many of these so-called `hot spot` failures has indicated the failure was caused by poor heat transfer associated with misalignment, dimensional distortion or poor jacket-to-core bonding. It has been postulated that misalignment of the fuel element is a necessary condition for formation of hot spots under the present reactor operating conditions. Neither tru-line contours nor X-8001 alloy are successful in the prevention of misalignment and associated ruptures; therefore, it has been proposed to test the effectiveness of projections on the side of the fuel element toward preventing fuel misalignment in ribbed process tubes. A previous test of this element termed the `bumper fuel element` was encouraging; however, it failed to provide the conclusive proof required to justify a large-scale demonstration loading. Supplement A to the basic test was written to obtain necessary preliminary data. This report presents an outline of further testing required to accelerate evaluation of the bumper concept.

This document details activities of the Fuels Preparation Department during the month of June 1960. (FI)

This document details activities of the irradiation processing department during the month of June 1960. A general summary is included at the start of the report, after which the report is divided into the following sections: research and engineering operations; production and reactor operations; facilities engineering operation; employee relations operation; and financial operation.

The thermocouple train is modified by the substitution of one eighteen-inch enriched and four twenty-four inch natural uranium tubular elements for the two thirty-six inch enriched tubular elements used on the original thermocouple train. The operating limits for the loading have been changed because of the change in charge power. This supplement also authorizes the addition of fuel elements upstream of the thermocouple train if the thermocouple element and heater elements do not provide enough heat to operate the loop at the desired temperature.