An analytical trajectory and aerothermodynamic analysis of a satellite containing a Task 2 thermoelectric generator was completed. A 300-statute mile circular polar orbit was used for this analysis and the launch was assumed to be from Vandenberg Air Force Base. Results of this study show that upon natural decay from a successful mission, the radio-cerium fuel will burn up in space at high altitude, thus only a very minor amount of radio cerium will be released to the stratosphere. A complete analyses of the fate of the radio-cerium fuel following various aborted launching attempts also was carried out. Charts summarizing the various assumed failures and locations of the fuel following failure are shown. A technical discussion of the methods used in performing the analysis is included in the report. (auth)

The EGCR was designed to accommodate up to four gascooled experimental loops plus several experimental fuel elements in the open core. Two of the loops will utilize 51/2-in.-O.D. stainless steel tubes passing through the core along an axis which is about 17 in. from the central axis of the core. The other two loops will utilize 91/2-in.-o.d. tubes about 68 in. from the central axis. Inherent safety in the design, facility design, primary loop design, auxiliary systems and equipment design, primary and secondary containment design, instrumentation and controls, and special operations are discussed. (M.C.G.)

Channel-to-channel flow distribution within Type 3 (SM-2, stationary and control rod fuel elements modified for use in the SM-1, SM1-1A, and PM-2A core support structures and control rod tubes was measured in single element flow testing. Plots of channel-to-channel flow distribution and element pressure drop at various element flow rates are given. Flow distribution for the top-orificed SM-1A and PM-2A stationary elements was within the plus or minus 12% deviation from element average utilized in previous thermal analyses of these cores. Testing of the bottom-orificed SM-1 stationary element and the SM-1, SM-1A, and PM-2A control rod assemblies showed flow distribution exceeded plus or minus 12% devation from average. Simple modifications to the SM-1 stationary element indicated the possibility of improvng fiow distribution in that element. (auth)

Thermal stress tests on Type 1, SM-1A Core II fuel ele-ment sections were performed to study plate distortion and determine its dependency on temperature distribution, temperature differential, initial flatness, and ripple length. Test results will be correlated with the analytical model and used to predict ripple growth in other plate-type fuel elements. The tests showed that ripple growth is dependent on initial flatness of the plate and that the characteristic shape of ripples is maintained at all temperature differentials: The tests also showed that the ripple growth rate for a ripple of 5 mil initial magnitude is approximately 0.12 mils/ deg F for a peak temperature differential of 103 deg F and that the apparent relationship between ripple net growth and length is 1.3 mil/in. of ripple for a peak temperature differential of 103 deg F. A permanent distortion of 2 mils for a complete temperature cycle from 0 to 103 to O deg F differential was found. The temperature profile across the plate width was found to affect the magnitude of ripple growth. (auth)

Editor please delete 26456.<><DSN>16:026457<ABS>Work done in the P122 reactor control actuator area is summarized. Actuators were required to radially position the absorber blades in the core of the reactor. The P122C1 was a subsonic power plant and temperatures were low enough to permit the use of hydraulics in the actuator area. The program was reoriented and the power plant designated P122C3 which was a supersonic version of the folded flow power plant. The ambient temperature at maximum power was high enough to require pneumatic actuation of the control blades. The program was reoriented after two design iterations of the subsonic power plant. A test model of the actuating equipment and the entire linkage assembly was on hand and completed when the program was cancelled. The linkage was being redesigned for the supersonic application and special bearings were ordered for fabrication into the lower temperature rig. The actual mechanical concepts of the pneumatic actuator were under study when the program was cancelled. (auth)

The postulate that the average number of lattice displacements is directly proportional to the available energy is carried one step further; it is assumed that damage to steel (particularly in regard to brittle fracture) is proportional to the number of lattice vacancies that occur. The model, although crude, permits estimation of the relative damage resulting from differences in neutron spectra. The results can be used as a rough method of correcting damage data for the effect of the neutron-energy spectrum. Radiation damage calculations for steel, relative to those for a fission spectrum, were made for neutron spectra that result from fission neutrons penetrating water or graphite. The results were plotted as a function of effective distance from the fission source. From this plot it is possible to make a conservative estimate of the correction factor to apply to damage data obtained with different neutron spectra. (auth)

In a prevlous paper, some inequalities occurring in the one-veloclty theory of neutron transport with lsotropic scattering were derived. Some of the previous results are generalized to the case of linearly anisotropic scattering. (auth)

Effective energy cutoffs have been calculated on an IBM7090 computer for Cd, Gd, Sm, and B filters as functions of filter geometry, the ratio of Maxwellian to epithermal flux (assumed to be 1/E), the lower energy limit of the 1/E flux, the energy corresponding to the Maxwellian most probable (modal) velocity and filter thickness. The geometrical configurations were spherical (which on the assumptions madc is equivalent to a beam flux case), cylindrical and slab. By the use of two or three different filters (Cd and Gd and perhaps Sm) it should be possible to detect resonances in the thermal to cutoff energy regions, in addition to measuring resonance integrals and thermial cross sections of unknown nuclides. (auth)

Process and product requirements to be met in the fabrication of Type 3 fuel elements are presented. The fuel elements specified consist of thin plates of a dispersion of highly enriched UO/sub 2/ and ZrB/sub 2/ in a stainless steel matrix which is clad with stainless steel on all surfaces. Quality assurance provisions are discussed. Process and material specifications and packaging and packing for shipment are described. Sample calculations and drawings are included. (M.C.G.)

The temperature coefficient of the Reactivity Measurement Facility was found to be 49 plus or minus 1 mu k/ deg C (1 mu k = 10/sup -6/ DELTA k/k) in the range 15.4 to l7.8 deg C. The change in the net reactivity of a standard sample was --0.48 plus or minus 0.02, --0.66 plus or minus 0.03, and --0.78 plus or minus 0.02 mu k/ deg C in three measuring positions. These low values generally make temperature corrections insignificant. The above results are compared with previous determined values. This information developed in the RMF should be generally applicable to flux-trap-type reactors such as the Advanced Reactivity Measurement Facility (ARMF) and ARMF-ll, now under construction. RMF was dismantled in April 1962. (auth)

This report addresses the comparison of SNAP-50/SPUR System with MHD, Brayton, boiling reactor and thermionic systems.

Submitted to Univ. of Tennessee, Knoxville. The fast neutron-induced fission cross section of U/sup 234/ was measured from threshold to 4-Mev neutron energy. A maximum of 1.26 barns was found at 850 kev followed by a minimum of 1.10 barns at 8050 kev. The angular ani-sotropy of the fragment distribution was measured for neutron energies from 400 kev to 4 Mev. Extrema in the ratio sigma /sub f//( sigma /sub f(90 deg ) were found at 500, 850, and 1050 kev; the distribution at 500 kev showing a maximum in the direction normal to the beam (side-wise peaking) while that at 850 kev showed a maximum along the beam direction. The distribution at 8050 kev showed forward peaking but to a lesser extent than for energies immediately higher or lower. The behavior was analyzed according to the theories of Bohr and Wheeler. The dip in cross section between 850 and 1050 kev is consistent with the suggestion of Wheeler that neutron competition in the decay of the compound nucleus enters with increased strength in this area. Vibration-rotational levels in U/sup 234/ beginning at 790 kev are known to exist and inelastic neutron scattering to these levels serves to depress the …

The failure of a fuel element cladding in one of the water-cooled plutonium production reactors permits the erosion of irradiated uranium metal by the cooling water which is normally disposed to the Columbia River. Monitoring systems at both the reactors and at their effluent basin outlets to the river continuously monitor these streams, and if major fission-products release occurs the coolant can be held in retention basins. In addition to these monitoring systems, a river monitor is located at the 300 area which continuously monitors the gross gamma activity of the Columbia River; however, its sensitivity to rupture debris is limited because of the relatively high background'' from the short-lived (n, {gamma}) produced radionuclides in the river. In making hazard assessments and in providing adequate monitoring techniques and equipment at down-river locations, it is essential to know what changes occur in fission and (n, {gamma}) produced radionuclides following release of rupture products to the river.

Cerium sulfate purified by D2EHPA in Semiworks can be precipitated by adjusting pH to between 1 and 2 in tank 6 with 50% caustic. The solution can then be transferred through tank 1 to tank 67, where sodium bisulfate is added to make the solution 0.5M sulfate. A stoichiometric amount (mole for mole) of 50% caustic is added to just neutralize the sodium bisulfate. The precipitate is digested one hour at 60 C, then filtered.

This report discuss the design of the Plutonium Reclamation Facility the capacity of which will be 300 kilograms per month or 3600 kilograms per year or plutonium. The subject facility, as the name implies, must be extremely flexible in its ability to handle a wide variety of feed materials. The new facility will be operating on a three-shift day, five-day week, 40-week year with an overall efficiency of 75 percent; twelve weeks per year will be required for ``turnaround`` time to enable campaign operation for segregation of feed plutonium by isotopic content.

Since it is now possible to obtain a heat-treated U core that is randomly oriented and has a finer average grain size, it is necessary to irradiate measured, transformed fuel cores over the full range of grain sizes, in order to compare relative dimensional stabilities. An improved ultrasonic tester, Fuel Core Tester-UT-2, is used to test all fuel cores.

Reactor operations at Hanford are to be conducted in the future under a new contract clause on nuclear health and safety, which sets up a form of AEC control somewhat akin to the licensing and regulatory function of the Commission with respect to commercial nuclear plants. As this transition in functional responsibilities approaches, it seems appropriate to review the record of more than 100 reactor-years of General Electric operation at Hanford, and to give an accounting of stewardship in regard to the all-important matter of nuclear safety in the production reactor operations. For the purposes of this review the events since General Electric assumed responsibility for plant operations in 1946 may be grouped into three periods. The first of these might be termed {open_quotes}The Early Years{close_quotes} and covers the period 1946 through 1952. During this time three additional reactors were placed on the line to supplement the original three war-built reactors. Power levels about doubled over the span, but there were relatively few changes from the original concepts of reactor design and operation, including the approach to reactor safeguards. A second period, 1953 through 1958, saw substantial advances in technology which led to new reactors, new loadings, and sharply increased …

Fuel element misalignment is an apparent cause of ``hot-spot`` ruptures. Several methods of eliminating hot spots have been tried, however, none appear to have completely solved the hot-spot problem. Attachment of projections to the side of the fuel elements appears to offer a means of minimizing misalignment since they act as bumpers against the side of the process tubes. Preliminary data from tests now in progress or recently completed indicate excessive corrosion rates are not to be expected on fuel elements with projection attached. In fact, no hot spots were found on 39 columns of normal self-supported I&E fuel elements discharged at normal goal or on four columns of enriched self-supported I&E elements irradiated to 850 mwd/t exposure.

Reference (2) requested that we review the possibilities that classified information might be revealed by isotopic composition of fission products, and suggest methods of making any such compositions. These questions have been reviewed thoroughly by Hanford Laboratories experts; their analysis confirms earlier conclusions (Reference 1). Based on current classification guides, isotopic compositions of cesium, promethium, and probably other rate earths could reveal secret information. Unless classification rules are changed, we see no way to declassify these fission products except by blending them with unclassified products from other sources.

The steadily increasing graphite stack distortion in the K Reactors has caused serious operating problems with the existing horizontal control rods. To compound the seriousness of the problems, the high level of reactor operation today and the anticipated higher level of operation in the future demands a reliable control rod system. A flexible control rod has been designed by Reactor Design, IPD, to facilitate reliable operation of the HCR system in spite of channel bowing arising from graphite stack distortion. This flexible control rod design is radically different from the existing K Reactor control rods and in fact, is more closely aligned to the control rods now in use at the older Hanford Reactors. The major difference of the new rod is the elimination of intimate contact between the poison-containing section of the rod and the cooling water. Such a change in design as described above could result in significant changes in the operating temperatures of the rod proper. This study was undertaken to provide a calculated indication of the temperature changes and the relative magnitude of such changes relative to reactor power levels, graphite temperatures, coolant temperatures, etc. In addition to this basic information, the scope of the study …

This production test authorizes the irradiation of full length charges of solid aluminum in the KER loops while on single-pass operation with process water cooling. Calculations show that the irradiations will not result in any foreseen reactor or personnel hazards and can be safely performed in the KER facilities.

McCormack and Schwendiman issued a report in 1959 which estimated the amount of fission products from fuel element failures entering the Columbia River during the period 1952--1958 inclusive. Since that time there have been no attempts made to publish similar information for the subsequent years. It is the purpose of this report to review the fuel element rupture experience of 1964 as somewhat typical of the last few years, and to estimate the amount of significant fission products which entered the Columbia River. The routine measurements of fission products both in the reactor effluent streams and in the Columbia River will be reviewed for comparison purposes.

This document is explanatory in nature and is intended to clarify certain questions about reactor neutron flux spectra in various AEC production facilities. Simplified models are used to illustrate neutron ``temperature,`` spectral ``hardening,`` and the so-called ``Westcott R.``

At the request of IPD Personnel, calculations on neutron and gamma attenuation were made for the NPR shield. The calculations were made using a new shielding computer code developed for the IBM 7090. The calculations show the thermal neutron flux, total neutron dose rate, and gamma dose rate distribution through the entire shield assembly. The calculations show that the side and top primary shield design is adequate to reduce the radiation level below design tolerances. The radiation leakage through the front shield was higher than the design tolerances. Two alternate biological shield materials were studied for use on the front face. These two materials were iron serpentine concrete mixtures with densities of 245 lb/ft{sup 3} and 265 lb/ft{sup 3} (designated by I-S-245-P and I-S-265-P, respectively). Both of these concretes reduced the radiation below design tolerances. It is recommended that the present front face biological shield be changed from I-S-220-P to I-S-245-P. With this change the NPR shield is adequate according to these calculations. The calculations reported here do not include leakage through penetration in the shield.