During August 1976, a total of 876 natural waters and 861 bottom sediments were collected at a nominal density of one location each 23 km/sup 2/ from streams and small lakes throughout the Noatak NTMS quadrangle, the southern two-thirds of the Baird Mountains NTMS quadrangle, and in the southwest corner of the Ambler River NTMS quadrangle. These samples were collected as part of the National Uranium Resource Evaluation program in Alaska being conducted by the Los Alamos Scientific Laboratory (LASL). The field collection and treatment of the samples were performed following strict LASL specifications. Total uranium was measured in the waters by fluorometry and in the sediments by delayed-neutron counting, using stringent quality assurance controls at the LASL. The uranium contents of the waters ranged from below the detection limit of 0.02 parts per billion (ppB) to a high of 8.38 ppB, and the uranium contents of the sediments ranged from a low of 0.3 parts per million (ppM) to a high of 34.0 ppM. In general, the locations of waters containing relatively high uranium contents were found to occur in clusters, and particularly in the headwaters of streams draining the southern slopes of the Baird Mountains. Few sediments contained …

Progress is reported on the following studies: chemistry studies designed to achieve a more complete understanding of the fundamental chemistry of technetium in order to facilitate the design of future radiopharmaceuticals incorporating the radionuclide /sup 99m/Tc; the development of new radiopharmaceuticals intended to improve image quality and lower radiation doses by the use of short-lived radionuclides and disease-specific agents; the development of short-lived positron-emitting radionuclides which offer advantages in transverse section imaging of regional physiological processes; and studies of the toxic effects of particulate radiation.

Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to potential nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each protective measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment falure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

The laboratory cyclone incinerator, which started operation in January, 1978, has been developed to study the effects of burning feed containing radionuclides other than /sup 238/Pu. These isotopes must be removed from the off-gas during incineration. Preliminary data from early runs show that the combustion efficiency of the incinerator can still be improved. Controlled experiments were conducted on the pressed ash-cement pellet matrix to define compressive strength, mechanical stability, density, and effect of curing environment (wet cure and dry cure). Leachability studies were initiated on pressed sludge/cement matrix in distilled water at ambient temperature. Compressive strengths of sludge/cement pressed matrix samples were investigated. Physical and chemical attributes of incinerated ash were evaluated in relationship to the ash/cement matrix. 7 figs., 6 tables.

The plastic deformation of crystals is usually accomplished through the motion of dislocations. The glide of a dislocation is impelled by the applied stress and opposed by microstructural defects such as point defects, voids, precipitates and other dislocations. The planar glide of a dislocation through randomly distributed obstacles is considered. The objective of the present research work is to calculate the critical resolved shear stress (CRSS) for athermal glide and the velocity of the dislocation at finite temperature as a function of the applied stress and the nature and strength of the obstacles. Dislocation glide through mixtures of obstacles has been studied analytically and by computer simulation. Arrays containing two kinds of obstacles as well as square distribution of obstacle strengths are considered. The critical resolved shear stress for an array containing obstacles with a given distribution of strengths is calculated using the sum of the quadratic mean of the stresses for the individual obstacles and is found to be in good agreement with the computer simulation data. Computer simulation of dislocation glide through randomly distributed obstacles containing up to 10/sup 6/ obstacles show that the CRSS decreases as the size of the array increases and approaches a limiting value. …

A number of materials in and within the proximity of buildings housing fusion energy experiments (FEE) were analyzed for their potential fire hazard. The materials used in this study were mostly: electrical and thermal insulations. The fire hazard of these materials was assessed in terms of their ease of ignition, heat release rate, generation of smoke, and the effect of thermal environment on the combustion behavior. Several fire protection measures for buildings housing the (FEE) projects are analyzed and as a result of this study are found to be adequate for the near term.

The PARC (plutonium accident resistant container) project resulted in the design, development, and certification testing of a crashworthy air-transportable plutonium package (shipping container) for certification by the USNRC. This PAT-1 (plutonium air transportable) package survives a very severe sequential test program of impact, crush, puncture, slash, burn, and water immersion. There is also an individual hydrostatic pressure test. The package has a payload mass capacity of 2 kg of PuO2 and a thermal capacity of 25 watts. The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonalby small air-transportable package, advancing the packaging state-of-art. Optimization design iterations were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to radioactive materials containment acceptance criteria, shielding and criticality standards. (14 figs, 7 tables)

The study results indicate that blindly selecting the ''worst-possible-case assumptions'', and then postulating a radioactive material (RAM) release without identifiable mechanisms compounds unrealism and adds confusion. In order for a RAM release to occur, an unlikely series of events must first occur which breach the multiple containment barriers surrounding the fuel. The overall safety margin provided by the packaging equipment increases geometrically beyond the already adequate margin provided by each containment barrier. A case evaluation of the NL 10/24 packaging system illustrates this contention by showing that: (1) the accident events which must occur before a release of RAM, other than gases, is possible are in themselves incredible; and (2) the biological effects of a release of fission gases will in all liklihood be nil.

Objective is the qualification of a PLBR-size primary loop flexible piping joint to the ASME Band PVC rules. Progress and activities are reported for: Class 1 flexible joint code approval support, engineering and design, material development, component testing, and manufacturing development. (DLC)

This study had a threefold purpose: to examine the atomic absorption gamma spectroscopy, emission spectroscopy, and neutron activation data accumulated by Atlantic Richfield Hanford Company through 1976 (predecessor company to Rockwell Hanford Operations), and evaluate the precision of the data and their calibration reliability; to determine from these chemical data those elements that were best for characterizing basalt flows for correlation purposes; and, to use these chemical data to establish statistical correlations among type locality reference groups and the basalt flows penetrated by deep core holes in the Pasco Basin or exposed on the surface at certain field section locations.

Laser induced fusion is the forerunner of a class of inertial confinement schemes in which hydrogen isotopes are heated to thermonuclear conditions in a very short period. The process is characterized by such short time scales that fuel confinement is achieved through its' own finite mass and expansion velocity, approaching 1 ..mu..m/psec for ignition temperatures of order 10 keV (10/sup 8/ /sup 0/K). With current laser powers limited to several terrawatts one readily estimates, on the basis of energy conservation, target mass, and expansion velocity, that target size and laser pulse duration are on the order of 100 ..mu..m and 100 psec, respectively. Within these constraints, targets have been heated and confined to the point where thermonuclear conditions have been achieved. This paper describes a sampling of diagnostic techniques with requisite resolution (microns and picoseconds) to accurately describe the dynamics of a laser driven compression. As discussed in each case cited, these in turn provide insight to and quantitative measure of, the physical processes dominating the implosion. The success of the inertial confinement fusion program is strongly dependent on the continued development of such diagnostics and the understanding they provide.

The report presents refined methods for calculation of fuel temperatures in PuO/sub 2/-UO/sub 2/ fuel in Fast Breeder Reactor (FBR) fuel pins. Of primary concern is the calculation of the temperature changes across the fuel-to-cladding gap of pins with fuel burnups that range from 60 to 10,900 MWd/MTM (0.006 to 1.12 at.%). Described in particular are: (1) a proposed set of heat transfer formulations and corresponding material properties for modeling radial heat transfer through the fuel and cladding; and (2) the calibration of a fuel-to-cladding gap conductance model, as part of a thermal performance computer code (SIEX-M1) which incorporates the proposed heat transfer expressions, using integral thermal performance data from two unique in-reactor experiments. The test data used are from the HEDL P-19 and P-20 experiments which were irradiated in the Experimental Breeder Reactor Number Two (EBR-II), for the Hanford Engineering Development Laboratory (HEDL).

Analyses of integral power-to-melt tests, HELD P-19 and P-20, are presented. These tests contained mixed-oxide, UO/sub 2/-PuO/sub 2/, fuel prototypic of fuel to be used in the driver fuel assemblies of the Fast Flux Test Facility, FFTF, which could be used in initial Fast Breeder Reactors, FBR's. The experiments were irradiated in the Experimental Breeder Reactor No. 2, EBR-II, for the Hanford Engineering Development Laboratory, HELD. The primary test variables are fabricated fuel-to-cladding gap size and accumulated fuel burnup to 1.1 at.%. The observed improved thermal performance at higher burnup is analyzed based on results and conclusions drawn from sibling pins irradiated during the preconditioning phases of the P-20 test. These pins were not included in the overpower phase of the P-20 test. The primary cause of the improved thermal performance was found to be maturing fuel restructuring and rapid fuel-to-cladding gap closure. The favorable impact of these observations on the design and operation of FFTF and future FBR's is also reviewed.

Control of temperature swings is a major consideration in design of passive solar heated buildings - especially so as the designer seeks to achieve most of the building heat from the sun. Observations of temperature swings in several passive buildings are cited. Methods of temperature control are discussed, both by means of control intervention such as using of auxiliary backup heating, ventilation, and blowers, and by means of building design. The design approach is preferred as the main course with the intervention techniques used for fine tuning.

Feedback stabilization of the Scyllac 120/sup 0/ toroidal sector is reported. The confinement time was increased by 10-20 ..mu..s using feedback to a maximum time of 35-45 ..mu..s, which is over 10 growth times of the long-wavelength m = 1 instability. These results were obtained after circuits providing flexible waveforms were used to drive auxiliary equilibrium windings. The resultant improved equilibrium agrees well with recent theory. It was observed that normally stable short-wavelength m = 1 modes could be driven unstable by feedback. This instability, caused by local feedback control, increases the feedback system energy consumption. An instability involving direct coupling of the feedback l = 2 field to the plasma l = 1 motion was also observed. The plasma parameters were: temperature, T/sub e/ approximately equal to T/sub 1/ approximately equal to 100 eV; density, n/sub e/ approximately equal to 2 x 10/sup 16/ cm/sup -8/; radius, a approximately equal to 1 cm; and ..beta.. approximately equal to 0.7. Beta decreased significantly in 40 ..mu..s, which can be accounted for by classical resistivity and particle loss from the sector ends.

An investigation is presented of the role which trapped particles might play in the drift wave stability of ELMO Bumpy Torus (EBT). The model adopted consists of a bounce-averaged drift kinetic equation with a Krook collision operator. Care has been taken to model, at least in an elementary way, the features which distinguish the physics of EBT from that of tokamaks, namely the large magnitude and velocity space dependence of the poloidal drift frequency ..cap omega.., the relatively small collisionality ..nu../..cap omega.., the enhancement of ..nu../sub eff/ for passing particles, and the closed nature of the field lines. Instabilities are found which have a somewhat dissipative character, however the precessional drift is found to be a significant stabilizing influence. In most cases, the modes are completely stabilized when ..omega../sub *//l..cap omega.. approximately equal to 1 for normal gradients. For reversed gradients (..omega../sub *//l..cap omega.. < 0), stability is greatly enhanced.

The successful development of the neutral beam injection method of heating tokamaks has opened up a new range of theoretically predicted phenomena to be explored. This article, intended for the nonspecialist, reviews the existing experimental observations and theoretical understanding of tokamak equilibrium and large scale stability. Then a survey is presented of the new phenomena, such as flux conserving sequences of equilibria and pressure-driven ballooning modes, that are expected to accompany the significantly enhanced plasma pressure to be produced in tokamaks now under construction.

These four papers on high-..beta.. tokamak instabilities, fundamental time scales for flux conserving tokamak heating, intense tokamak heating in the adiabatic approximation, and the Elmo Bumpy Torus were presented at the Workshop on High Beta Tokamaks held at Varenna, Italy on September 1--9, 1977. They will be included in the proceedings of that meeting, to be published by the Department of Energy.

This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary …

One element in the Physical Protection of Nuclear Material in Transit Program is a determination of the number of local law enforcement agency (LLEA) officers available to respond to an attack upon a special nuclear material (SNM) carrying convoy. A computer model, COPS, has been developed at Sandia Laboratories to address this problem. Its purposes are to help identify to the SNM shipper areas along a route which may have relatively low police coverage and to aid in the comparison of alternate routes to the same location. Data bases used in COPS include population data from the Bureau of Census and police data published by the FBI. Police are assumed to be distributed in proportion to the population, with adjustable weighting factors. Example results illustrating the model's capabilities are presented for two routes between Los Angeles, CA, and Denver, CO, and for two routes between Columbia, SC, and Syracuse, NY. The estimated police distribution at points along the route is presented. Police availability as a function of time is modeled based on the time-dependent characteristics of a trip. An example demonstrating the effects of jurisdictional restrictions on the size of the response force is given. Alternate routes between two locations …

Effects of internal alpha radiation on borosilicate glass, a perspective matrix for long-term storage of Savannah River Plant (SRP) radioactive waste, were evaluated in samples containing 45 wt % simulated waste (Fe(OH)/sub 3/--MnO/sub 2/) and either 0.5 wt % /sup 244/Cm or 1 wt % /sup 238/Pu. A glass containing /sup 238/Pu without waste was also studied for comparison. The glasses were examined for changes in physical stability, leachability, and dilatation. Alpha dose rates in the test glasses ranged from 4.5 x 10/sup 14/ to 1.3 x 10/sup 15/ alpha dis/(g-day). After 420 days, microcracks had formed; however, no macrostructural damage to the glasses was observed. Leachabilities for /sup 244/Cm and /sup 238/Pu were <7 x 10/sup -8/ g/(cm/sup 2/-day) and were not affected by the radiation. Continuous leaching by water for 5 days removed <10/sup -5/% of the isotopes. Alpha radiolysis caused expansion of the simulated-waste glasses in proportion to dose. Application of these results to glass containing radioactive Savannah River Plant waste indicated that internal alpha radiolysis will not cause detrimental effects during long-term storage (>10/sup 6/ years) of the waste glass.

Evaluations are made as to recommended maximum allowable cladding temperatures of spent nuclear fuels for extended dry storage. A number of near-surface dry storage concepts have been evaluated, as well as concepts for geologic storage. Work is continuing to upgrade and revise maximum temperature evaluations as the program proceeds.

The use of Nondestructive Assay (NDA) instrumentation at a nuclear waste terminal storage facility for purposes of Special Nuclear Material (SNM) accountability is evaluated. Background information is given concerning general NDA techniques and the relative advantages and disadvantages of active and passive NDA methods are discussed. The projected characteristics and amounts of nuclear wastes that will be delivered to a waste terminal storage facility are presented. Wastes are divided into four categories: High Level Waste, Cladding Waste, Intermediate Level Waste, and Low Level Waste. Applications of NDA methods to the assay of these waste types is discussed. Several existing active and passive NDA instruments are described and, where applicable, results of assays performed on wastes in large containers (e.g., 55-gal drums) are given. It is concluded that it will be difficult to routinely achieve accuracies better than approximately 10--30% with ''simple'' NDA devices or 5--20% with more sohpisticated NDA instruments for compacted wastes. It is recommended that NDA instruments not be used for safeguards accountability at a waste storage facility. It is concluded that item accountability methods be implemented. These conclusions and recommendations are detailed in a concurrent report entitled ''Recommendations on the Safeguards Requirements Related to the Accountability of …

A stress analysis of the LOFT Diesel Generator ''A'' Exhaust Stack was performed to determine its reaction to Safe-Shutdown Earthquake loads. The exhaust stack silencer and supporting foundation was found to be inadequate for the postulated seismic accelerations. Lateral support is required to prevent overturning of the silencer pedestal and reinforcement of the 4'' x 0.5'' silencer base straps is necessary. Basic requirements for this additional support are discussed.