This report addresses the calculational models for beta dose from spherical particles.

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A fuel jacket failure that occurred in May 1961 in the Type 304 stainless steel clad fuel element exposed in the Vallecitos Boiling Water Reactor superheated steam loop (SADE) was attributed to chloride stress corrosion cracking. In order to better understand the failure, a test program was carried out to try to reproduce the rapid stress corrosion attack in the simulated superheat reactor environment of the CL-1 superheat facility. The methods of corrosion testing under heat transfer conditions reported previously were modified: to apply a longitudinal stress on the test sheaths to produce a 0.1 per cent elongation in 1000 hours; to increase the chloride content of the moisture carryover with the steam by increasing the chloride in the recirculating water to 1.5 ppm; and to expose the solids deposits to various metal temperatures. After 1000 hours of exposure, no significant attack was noted on the test sheaths. The test procedures were further altered to simulate the significant amount of SADE fuel element exposure to saturated steam at varying temperatures with little to no superheat being generated. A 776-hour total exposure was carried out with the test conditions cycled several times. The entrance heater (calculated metal temperature during normal operation …

The graphite temperature limits specified in the process standards-are being closely approached or have limited power levels at B, D, DR, and reactors. An increase of approximately 50--100 C in graphite temperature limits at these reactors would permit year-around operation on bulk outlet temperature limits. There has been considerable recent interest in extending to all reactors the use of nitrogen as a reactor gas constituent. With present graphite temperature limits, significant benefit from nitrogen usage will not be obtained during the winter months at reactors which are graphite temperature limited with approximately 100 per cent helium. However, during the summer months when bulk outlet temperature limitations result in graphite temperatures considerably below the maximum permissible graphite temperatures, the substitution of nitrogen for carbon dioxide could be of value. Under these circumstances, both a reduction in helium usage and a reduction in enrichment costs could result. With an increase in permissible graphite temperature limits, year-around benefit from reduced helium usage and reduced enrichment costs would be possible. To meet the Pu-240 specification with higher graphite temperatures however, would require a reduction in current goal discharge exposures with resulting increased fuel and burnout costs. Additionally, the incentives for higher graphite temperatures are …

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A study was made to determine the optimum arrangement of the fringe flow zones at KE Reactor. Also considered, was the possibility of converting part or all of the existing fringe zone (actually three flow zones) from solid metal to I & E metal to decrease rupture potential in these low flow zones. The necessity for this study was indicated by high tube outlet temperatures in the fringe flow zones and the recent occurrence of two solid metal ruptures in process tubes located in the fringe zone of the reactor. Two additional solid ruptures occurred during the period that conversion to I & E in zone 2, as recommended below, was being completed.

In developing a method of numerical analysis for the solution of thermal- stress problems special emphasis was given to fuel elements with internal coolant channels. Numerical techniques ior reducing the partial differential equation system te a form suitable for numerical solution and a new iteratlve method of solving large systems of linear algebraic equations were employed. Computer codes were devised to obtain the numerical solution of the thermal-stress problems and were used to obtain numerical results for single-hole and seven-hole hexagonal elements and plate-type elements. Comparisons were made between analytical results and numerical results for the case of t:.ie simple annulus shape. (auth)

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The NIGHTMARE program calculates the gamma-dose rate at any point in or near a reactor by means of the NDA buildup-factor method. The source distributton is obtained from a two-group, two-dimensional diffusion code (EQUIPOISE-2). Seven gamma-energy groups are used. As many as nine different concentric cylindrical regions may be considered, but only radial variation of attenuation and buildup properties is permdtted. Two schemes of estimating buildup factors through a succession of materials are built into the code, and results may be obtained by using either or both methods. An IBM 7090 computer wtth seven tape untts is required. Computations are performed at the approximate rate of l0,000/n mesh points per minute, where n is the number of axial, radial, and angular increments. The program is run under the control of the IBM 7090 MONITOR system. (auth)

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The development of equations used for determination .. starting weights of the individual mix batch constituents is presented. Experimental data resulting from batches compounded in this manner are shown for uranium loading, carbon density, and composite density of as graphitized natural uranium loaded fuel elements.

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Currently the Irradiation Processing Department is working on a program which has as its goal the reduction of radioactive discharge to the Columbia River. The most significant radioactive isotopes of concern are the relatively long lived isotopes P{sup 32}, Zn{sup 65}, As{sup 76}, N{sub p}{sup 239} and Cr{sup 5l}. Because of their relatively long life they can be found in the Columbia River well below the Hanford Plant. P{sup 32}, because it is absorbed and concentrated by the river biota is of prime concern because the dose which it contributes to man through the eating of fish is not known for certain. Included in this reactor effluent decontamination study program are studies of process water recirculation systems, process water pre and post treatment methods, and effluent disposal to an artificial lake. This report reviews the artificial lake concept to determine its feasibility and possible effects on the concentration of radioactive isotopes in the Columbia River downstream of the Hanford Plant. The findings are based on the study and evaluation of available data. Meterological, hydrological, and biological absorption aspects of the artificial lake concept were studied several years ago. Little or no scientific work has been done in recent years to …