A proposed modification to the experiment off-gas system for the EGCR-LITR Capsule Experiment is evaluated by comparing the atmospheric radiation doses to laboratory personnel that could result in the event of an experiment failure while utilizing the present off-gas system to those doses that would result if various modifications to the existing system were effected. The modifications considered include the addition of an iodine vapor adsorber and a krypton-xenon adsorber. The addition of an iodine vapor adsorber to the off-gas system, which presently includes an absolute filter, reduces the atmospheric dose to the thyroid from 510 mr to 1.4mr. The total body dose is also reduced from 2.2 to 1.3 mr. The total body dose can be further reduced to 0.7 mr if a large krypton-xenon adsorber is included. Basd upon these calculated atmospheric radiation doses it is concluded that the addition of an activated charcoal iodine vapor adsorber will reduce the atmospheric hazard to an acceptable level.

Abstract. A series of paired explosions in a salt mine near Winnfield, Louisiana, has been conducted to test a theory by Dr. A. L. Latter concerning seismic decoupling by underground cavities. The theory predicted a decoupling of about 100. Free-field and surface measurements from an explosion in either a 6-ft- or a 15-ft-radius spherical cavity were compared with similar measurements from a completely tamped explosion. Shot sixes were from a 10 lb up to a few tons. Surface measurements were made out to 100 km and covered the frequency range from 0.05 to 100 cpa. The experiment confirmed that decoupling does occur. The actual decoupling factor as a function of frequency is presented and compared with the Latter theory.

This technical report represents the final design for a sodium to sodium intermediate heat exchanger and a sodium to water steam generator. The intermediate heat exchanger is designed for operation in a nuclear power plant using liquid sodium as the primary and secondary coolant. Since the primary fluid coming fro the reactor is radioactive, the purpose of the IHX is to transfer heat to a nonradioactive fluid which then goes to the steam generator. Because of radioactivity the unit will be enclosed in a concrete pit and will not be accessible during periods of operation. Immediately after shutdown it will be necessary to allow time for radioactive decay before the unit will be accessible to personnel. Because of inaccessibility and possible long periods allowed for decay time, it is imperative that the unit give trouble free operation. During periods of shutdown, the internals should have easy access for inspection and repair if necessary so that down time is held to a minimum. The steam generator is designed to generate superheated steam using liquid sodium from the intermediate heat exchanger as the heat source. Its basic design is a shell and tube unit made up of three difference sections: (1) a …

Introduction: This volume deals principally with the chemical analysis and the stress analysis for a sodium to sodium intermediate heat exchanger and a sodium to water steam generator. The work presented is an extension and modification of the analysis presented in the preliminary design report. The chemical analysis covers the sodium cover gas system and the effects of sodium-water reactions in the event of a leak in the steam generator. Considerable design work was done in an effort to maintain the integrity of the steam generator vessel under maximum leak conditions. The method of sizing relief valves for each unit under varying leak rates is presented in this text and operation of the unit for the various leak rates is resented in the Operation and Maintenance volume. The stress analysis section covers those thermal transients which would be physically possible with this intermediate heat exchanger and steam generator design. Attention has been given to methods of operation which would minimize the magnitude and frequency of thermal shocks. Certain areas have been studied in detail where thermal stresses appear high. This report also includes a structural design basis for handling stress analysis of combined mechanical, hydrostatic and thermal stresses and conditions …

Introduction: Sodium Components Material Specifications. Twenty-three material, inspection and welding specification are presented for the various parts of both the intermediate heat exchanger and steam generator. Tables indicate the applicable parts and assemblies to which these specifications shall apply. For other parts, where the material requirements are not severe, the ASTM or other indicated specifications shall apply.

This technical report contains the operation and maintenance specifications for the intermediate heat exchanger and the steam generator. The report contains eight sections: (1) General Information, (2) Shipping and Installation, (3) Operation Procedures, (4) Scram and Casualty Shutdowns, (5) Leaks, (6) Instrumentation and Control, (7) Maintenance, and (8) four Appendixes (a) Boiler Water Chemistry Recommendations, (b) Final Concept Drawings, (c) Industrial Nucleonics Literature on Liquid Level Detector, and (d) Sodium Purity Control Recommendations.

An estimate is made of the illumination required for the 80-in. bubble chamber, based on the absolute sensitivity of the film. The required intensity is estimated to be 23.4 tube is calculated to be 145 cm/sup 2/. The results are compared with these obtained from 72- and 20-inch chambers. (D.L.C.)

This report summarizes and documents radioactive waste released to the environs of the Savannah River Plant from startup through 1959. During this period, the quantity of beta-emitting radioisotopes released was determined by a total or ``gross`` analysis. However, advanced instrumentation and technology now permit an economical determination of most individual radionuclides. Therefore, future waste audit reports, beginning with January 1960, will record the quantity of specific radioisotopes released rather than gross amounts.

Calculations of critical size, reactivity coefficients, and delayed neutron fractions are described for mockups of a plutonium-fueled EBR-I (Mark IV) and five possible metal-fueled power reactors. The calculations for critical size were performed using the SNG transport-theory code. The reactivity coefficients were estimated by SNG calculations or by one-group perturbation- theory desk calculations. The delayed-neutron fractions were estimated by an approximate method, which starts with a calculation of beta for a bare core and then makes a correction for the small effect of the blanket. The reactivity effects studied were the effects of fuel-plate and whole-core expansion and control-rod worths. (auth)

The ORNL analog computer was used to simulate four widely different experiment conditions in the small EGCR in-pile loops. Various control and safety systems were evaluated in each case. Curves are included which show the response to postulated accidents and component failures. The results indicate that one standard control and safety system will probably be adequate for a wide variety of test conditions.

The scintillation and Cherenkov properties of UVA (ultraviolet absorbing) and UVT (ultraviolet transmitting) Plexiglas specimens were tested by analyzing cosmic-rayinduced pulses traversing the materials. It was found that the UVT sample has less than 1/6 the scintillation effect of UVA and is preferred for use in counters. (J.R.D.)

An exact treatment of the structure of tracks in nuclear research emulsion recently was attempted with some success. Many of the proofs also appear to be generally valid for bubble tracks in Wilson chambers. This is a summary of the results that are applicable to tracks of charged particles in bubble chambers.

With the exception of a brief period of slightly elevated chloride level in the secondary blowdown, water-chemistry conditions during the period were satisfactory. During the period, the reactor was shut down for end-of-core-life testing and rearrangement. A set of specifications covering all electronic and electromechanical mechanisms required to control the SM-1 reactor through the rod- drive motors and clutches was prepared and issued. Installation of instrumentation for plant response and system performance was virtually completed. Work on the interpretation of long-lived radiochemical data obtained at the SM-1 during core lifetime was continued. Analysis of all fissionproduct data collected during Core I life has started. Thirty-eight stationary and seven control subassemblies from SM-1 Core II were checked for alpha contamination by a gas-flow proportional-counting technique. The work on the final design of a waste-disposal system for SM-lA was stopped and an investigation of an interim system containing a bypass sampling system was undertaken. Work continued on tests 202, 203, and 204 in the activitybuildup phase of Test Series 200. Core- physics measurements were taken at end of Core I life to complete the series of measurements made throughout the lifetime of the core. (W.L.H.)

A digital comparator has been designed to flag data which goes above a maximum limit or below a minimum limit.

A training course was held to familiarize members of radiological assistance teams from various parts of the U. S. with the origin and nature of situations that might, by the event of an unusual accident. release radioactive materials to a populated environment. The course consisted of a series of lectures and a tour of some of the radiation monitoring, source handling, and transportation facilities at NRTS. A summary of the lecture material is presented. (W.D.M.)

From introduction: This report is a summary of the lecture material in the training course held at Idaho Falls, Idaho, to familiarize members of radiological assistance teams with information helpful in the response to an unusual accident including release of radioactive materials to a populated environment.

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From Introduction: "The report is organized into three major sections and three appendixes. In first major section are presented a brief historical account of the work done by other organizations prior to November, 1957, and a detailed description of the experiments conducted by Battelle with the No.4 and No. 5 stills. The second major section summarizes the results of cost studies. The third major section gives data, calculations, and derivations pertaining to the heat-transfer aspects of the Hickman still."

A mass flow measurement technique is described which has the capability of measuring homogeneous flow, slurries, highly corrosive fluids, and multiphase fluids. The device features ruggedness and reliability and has the ability to measure external to the flow. In the proposed system, the fluid is made to pass through a U-shaped tube wherein measurements of the augular momentum and density yield mass flow directly. As the fluid traverses the 180 deg bend, a radial force is generated, which can be measured with a force transducer. Density is determined by measurement of the radiation absorption in the fluid. (W.D.M.)

The following report describes a mass flow system that can measure homogeneous flow, slurries, highly corrosive fluids and multiphase fluids, with the addition of considering pressure drops, measuring external to the flow, ruggedness and reliability.

As set forth in the CPD Waste Management Program (HW-63958), new and improved methods of waste processing are planned to be initiated at Hanford. A major part of this program is concerned with the high-level wastes from the Purex Plant, with major objectives being the control of long-term hazards, the insurance of production continuity, and the attainment of positive waste accountability at reasonable over-all costs. Specific plans for the disposal of Purex wastes are based on the use of high temperature calcination for immobilization and volume reduction of the solvent extraction raffinates (1WW). Four major steps are involved, each of which is of basic importance to the program objectives. These are: (1) interim 1WW storage and removal for calcination, (2) calcination, (3) primary containment of the calcined waste, and (4) long-term storage. Immobilization of coating wastes and organic wash wastes is also planned, either by tank solidification methods or possibly by inclusion in the calcination feed stream. This document was dated September 15, 1960.

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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."

The scale deposit on the Experimental Boiling Water Reactor fuel plates can be removed by slurry honing the plates with an abrasive-water mixture. Problems inherent in any production operation of this type are discussed. Areas of continued investigation of the method are suggested. (auth)