Structures are subjected to a 50-kt blast, in order to obtain fundamental data on structures subjected to blast loading, to observe the response of the structures under this loading, and to determine the relative blast-resistance merits of several structural types. Modes of failure are determined. Shaped structures are found to be superdor to rectangular structures. Earth cover for the structures is also found to increase the blast resistance. It is found that standard Navy heavy bomb-proof structures with modifications can withstand a near-surface atomic burst at ground zero. (T.F.H.)

In order to determine the effects of atomic explosions on small civil defense shelters for family use, 29 simple structures were built along an arc 1200 ft from the target point and exposed to Buster Bursts B, C, and D. The structures were of four basic types; covered-trench, metal-arch, wood-arch, and basement lean-to. Because of poor cohesive properties of the soil, much of the earth cover on the shelters was removed by the first shot. Since test procedures prevented restoration of structures and replacement of cover after each blast, test results were materially affected. Partly above-grade cover-trench shelters provided less protection against blast than belowgrade cover-trench shelters and were much less desirable as protection against gamma radiation. Metal-arch shelters set in concrete appeared to have good potentialities with minor design modifications. Wood-arch shelters as designed, proved to be unsuitable substitutes for metalarch shelters. No worthwhile information was obtained on basement lean-to shelters. Unusual conditions disclosed design deficiencies in entrance construction, front, and end sections, and effective earth cover. Small shelters are potentially capable of meeting requirements of civil defense. (auth)

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Ducks, geese, coots, and other wild birds, both migratory and resident, have been swimming on the Redox waste water pond and eating the herbage and mud on the bottom. The pond is caused by the discharge of five to ten million gallons of water a day from the heating and cooling coils of reactor vessels in the Redox plant. At least twice in recent months leaks in the coils have permitted solutions of radioactive material to escape to the swamp. Once the leak was in the waste concentrator tank and once in the first oxidation tank, so that the material in the swamp would include significant amounts of every radioisotope found in irradiated uranium, including plutonium and all fission products. This paper discusses the hazards to humans from handling or eating wildfowl and to risk of both internal and external exposure.

It is considered appropriate occasionally to make independent checks on the effectiveness of the plant laundry in removing radioactive contamination from plant-issue protective clothing. Previous surveys have offered constructive criticism resulting in improved handling of high level and soft beta contaminated clothing and incorporation in new designs of ventilating and air sampling recommendations. Recurrently the adequacy of laundry reject limits is questioned, and only recently an accurate, special study resulted in relaxed limits for Metal Preparation area clothing. A current question concerns the advisability of determining the reject level on the beta-gamma monitor more frequently than once a day. 2 tabs.

Increases in power levels of the Hanford Piles have resulted in increased demands for process cooling water to the extent that the present filter plants are now operating at or above their design capacity. Further power level increases requiring even more cooling water may demand the operation of the filters considerably above design capacity. It has been proposed that by the use of aluminum sulfate, or filter alum, as a process water coagulant and activated silica as a coagulation aid, the present filter plant capacities could be increased appreciably. In order to investigate the effects of alum-treated water on pile operation, a full-pile production test was authorized in which alum was substituted for the standard ferric sulfate coagulant. This test was started in the 100-F Area on October 10, 1951. The results of the first thirty days of operation were presented in a previous report and served as a basis of the decision to proceed with installation of the activated silica addition facility. This report presents the data, results and conclusions obtained from the start of the test until its termination of December 28, 1951, when the addition of activated silica began.

This report is a comprehensive study of the cooling water systems for Hanford production reactors. It includes: film formation study, water plant flow improvement study, water quality tests, sodium dichromate elimination tests, corrosion in aluminium coagulated water study, and water composition, film formation, and effluent activity relationships. A section on special water studies is also included. 22 figs., 25 tabs.

Monthly activities for the following studies are given: (1) pile graphite monitoring; (2) graphite burnout and chemical studies; (3) graphite physical properties studies; (4) controlled temperature exposure; (5) thermal conductivity of gases; (6) damage mechanism study; (7) special exposures; and (8) experimental graphite program.

This study soughs to determine the x-ray activity of two samples of stainless steel prepared from tubing used in the ANL-140, in-pile experiment (P-13). This information is necessary in order to predict the activity of the P-13 rig at the time of removal. The primary contribution to the over-all activity will be from Ta{sup 181} and Co{sup 59} impurities.

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The test established that a practical orifice selecting device can be made and used to control the water flow rate to a process tube at either of two accurately calibrated rates with no loss of accuracy in flow measurement. With the selector tested, there was no flow restriction during the orifice transition. The pressure fluctuations during the change were small in magnitude, and caused no operational problem. The test demonstrated that orifice selection is feasible from operating other factors justify their use.

The factors responsible for the advances of 100-X compared with the older areas are: Simplification of the process, such as elimination of separate process water clearwells, by having the filtered water reservoirs perform that function. Combination of separate buildings into one building, such as combining filter pump house and process pump house. Use of electric standby. Use of higher capacity pumps and filter basins, and so fewer number of units. Centralization of control and operation. More compact arrangement of plant components. Use of waste heat for space heating, recovered from reactor effluent, backed up by steam plant.

An examination of the process scheme for handling present and future 231 Building supernatant solutions is necessary to forecast the effects of placing one of the processing facilities (224B) in a standby status and to ascertain if rescheduling of any of the supernatant solutions to 202S will be necessary. The examination indicates the 231 Building process supernates and cell cleanouts together with the 221, 224T Buildings acid washes and the 234-5 Building returns can be processed through the 224 T Building as additions to regular runs or master recycle runs when T Plant is processing as much as 41.6 tons/month of 600 MWD/T material and S Plant is processing 120 tons/month of similar material. The processing rate through, 224 T Building to satisfy these conditions is seventy-six runs per month. This total is divided into 63 regular runs plus 13 meter master recycle runs. The number of master recycle runs are primarily dependent upon the T Plant rate in as much as only one master recycle is directly required for each 35 tons processed through S Plant.

The lack of control rod capacity in the reactors with nine rods is causing significant production losses. The present control system, which functioned satisfactorily at a 300 MW equilibrium power level, is unable to counteract the larger startup reactivity transient effects at the current high power levels. This lack of strength causes the power level to be restricted during startup. The control system is also not flexible enough to prevent outages for the discharge of temporary poison. (1) A method of obtaining the required supplementary control strength and flexibility has been conceived. (2) and equipment developed for this purpose. The purpose of this test. (3) was to demonstrate the operational feasibility of the special front and rear face equipment which was designed to permit the routine charging and discharging of poison slugs at selected process tubes during periods of reactor operation. This report presents the data, results and conclusions obtained from the trial use of this equipment.

The purpose of this report is to present a basis for the application of slug corrosion rate data to corrosion limitations on pile power levels.

Members of the Working Committees for RDA DC-4 and RDA DC-7 (KE and KW Reactors) have reviewed and approved the design basis being used for the preparation of the 200 Area portion of the Program X preliminary project proposal. This design basis includes the following: plant capacity of 275 tons per month, average flow rate; the basic chemical process used will be the Purex process; the plant will be located in the Southeast Quadrant of the 200 East Area; the plant will contain a double line of canyon process equipment, each line being so designed as to operate independently of the other line; all facilities in the 202-A Area will be designed for earthquake (Zone 2) resistance; in the design of Building 202-A, 276-A, 291-A, and 203-A, consideration will be given to achieving the highest degree of blast protection compatible with economical design, availability of data on the effects of bomb blast, and design and construction schedules; and building 211-A will include chemical storage capacity required for operation of the plant at full capacity for sixty days; building 203-A will include uranyl nitrate hexahydrate storage capacity of three 60,000 gallon stainless steel tanks; Building 241-A, Waste storage tank farm, will …

Since the issuance of HW-17398, Separations Section Operating Standards - Part I relating to the Bismuth Phosphate Process on March 1, 1951, several changes in the process have been found desirable and have been reduced to practice. This document presents these changes.

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The basis for the safety control system for RDA-DC-3 is set forth in HDC-2462 with revisions as noted here. These conditions are satisfied by an array of safety control components comprising 37 vertical safety rods and 51 ball hoppers (51 total vertical graphite slots) as shown on drawing H-1-21403.

There were twenty slug jacket failures during the month of January 1952. Of these, fourteen were end cap failures, five were split slugs and one has not been removed. A total of 311.4 hours of outage time was required for removal of these ruptured slugs from the reactors. The detection, removal and radiation aspects, along with the slug data, are shown on the attached sheets.

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

This memorandum provides a discussion of material lost upon rupture of a fuel element.

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