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- 100-N Decontamination Facility Design Guide.
- Space has been reserved near the southeast corner of the 100-N Area for the 122-N Decontamination Facility. Previous correspondence between Burns and Roe, Inc and General Electric bae discussed various facilities which might be needed in the building. The concepts of the decontamination processes are under active development by research groups at Hanford. At present, there are several workable processes known; each one has one or more fairly serious drawbacks.
- 6144-Channel Time-of-Flight Analyzer
- Report describing a 6144-channel analyzer designed and built for the purpose of analyzing time-of-flight during studies of slow neutron scattering at Hanford Laboratories. This includes descriptions of the analyzer, its logic and circuits, and its test mode.
- An Alpha Scintillation Tester for Uranium Surface Contamination of N-Reactor Fuel
- Report that "describes a nondestructive tester and some of its applications in measuring 10 to 100 µg of uranium surface contamination on unirradiated, low enrichment, uranium fuel elements" (p. ii).
- Aluminum Alloy Work at Hanford
- Investigation of aqueous and steam corrosion at elevated temperature has been carried on at Hanford for several tears. A number of reports have been published on various phases of the corrosion program but, except for monthly reports, little has been written on the progress of the Corrosion and Coatings Operation alloy development program. This interim report will outline the more important current phases of the aluminum alloy program.
- Analog Computer Simulation of a Pulsed Extraction Column
- Report describing "computer techniques used in the simulation and parameter determination, how the data were prepared for inclusion in the computer simulation, (and) methods used in presenting the results" (p. 3-4).
- Analog Simulation of the Hanford N-Reactor Plant, Part 1: Description of the Overall Simulation Model
- Report that describes Hanford Laboratories' overall N-Reactor plant simulation model and its use to study the characteristics of the plant, plant operational procedures, and the effects of plant, operator, and control system malfunctions.
- APDAC-I, A PCTR Data Analysis Code for the IBM 709
- A flexible foil data processing program is described. Raw data on foil radioactivity are the basic input information required. Output may consists of relative activities, saturated activities, and/or cadmium ratio and flux spectrum data. A statistical analysis of the data is executed with the direct calculation, and errors estimated for the output data.
- An Automatic Monitor for As76 Concentrations in Reactor Cooling Water
- Report that describes the development, design, and operation of a prototype instrument meant to monitor the As76 concentration in reactor cooling water streams.
- Beta-Gamma Dose Rates from U232 in U233
- This report defines in detail the source of the dose rate of U233 and describes a method by which they may be predicted.
- The Blast Cleaning Process as an Aid to Visual Weld Inspection
- Late in 1958 it became apparent that some fuel elements were failing in the Hanford reactors as a result of water entering through the weld. The mode of entry appeared to be first through a void in the weld, then through a non-wet area or a train of voids in the braze, and finally to the uranium core. Defective closures of a similar nature were also typical of many fuel elements which have failed in the autoclaving operation as shown in Figure 1.
- Calculated Costs of Fabrication of Plutonium-Enriched Fuel Elements
- Report regarding the study of the calculated costs of fabricating plutonium-enriched fuel elements for Hanford Laboratories' Plutonium Recycle Program. Topics include descriptions of the process, design study of plants that use U235 and plutonium enriched fuel, and estimates of costs.
- Calculated Equilibrium Distributions for the Uranyl Nitrate - Tributyl Phosphate - Dilute Nitric Acid System for Temperatures Between 25 and 75 C
- Report discussing the "equilibrium uranium distribution between an aqueous nitric acid solution and a 30 per cent by volume solution of tributyl phosphate in a hydrocarbon diluent" (p. 2). This includes the necessary equations.
- Classification of PuO2 Particles
- The Biology Operation, Hanford Laboratories, required plutonium dioxide particles in five different micron and sub-micron size ranges. These were to be supplied as a water suspension in small containers. It was necessary to design equipment, suitable for hood operation, which would give a satisfactory separation of these particles.
- Coagulant Aids as Filter Aids
- The Hanford water treatment plants were operated for a number of years with alum as a coagulant and activated silica as a coagulant and a polyacrylamide as a filter conditioning agent. (other polyelectrolytes may be used.) Sufficient time has elapsed to make a comparison of the two methods of operation valid and useful. Such a comparison is the purpose of this paper.
- Completion Report An Automatic Fuel Element Stamping Machine
- This document will be concerned with the design, operation and maintenance of an automatic fuel element stamping machine. In order to take advantage of automatic unloading and length checking at the Acme-Gridley lathes in the 313 Building, automatic stamping of fuel elements is desirable. This, along with the need for a permanent canning line identification, resulted in a request for a machine to automatically stamp HAPO fuel elements on a production basis.
- Construction Completion Report: CAI-816, 100-N Reactor Plant
- Report from Hanford Laboratories concerning "the design and construction of the 100-N Reactor and heat dissipation plant complete with the necessary auxiliaries" (p. 2). Details of its construction and the plant's systems and instrumentation are described as well as economic considerations.
- Contraction Losses and High Temperature Pressure Drop Determinations for Tube Bundles
- In some engineering applications it has become necessary to operate equipment containing small diameter rods or tubes oriented parallel to flow stream. In the case of several nuclear reactors such as the Plutonium Recycle Test Reactor and the S. S. Savannah Maritime Reactor, bundles of small diameter rods are used as the fuel elements. The situation also has application to heat exchangers. A method for accurately predicting the pressure drop characteristics of various bundle configurations would be helpful in the design and selection of equipment. Some progress has been made toward gaining a greater knowledge of the bundle characteristics under low temperature conditions. This report includes the development of relationship for the effect of temperature on pressure drop.
- Corrosion Studies in Simulated N-Reactor Secondary System Water Environment
- Report containing the procedures and results of tests made to determine the corrosion resistance of materials in a simulation of Hanford Laboratories' N-Reactor secondary system water environment.
- Critical Mass Studies of Plutonium Solutions
- Chain reacting conditions for plutonium nitrate in water solution have been examined experimentally for a variety of sizes of spheres and cylinders.
- Critical Pressure Ratio for a Nozzle with Two-Phase Fog Flow
- In many cases of analysis of two-phase flow in systems, considerable computation or program time could be saved if the critical pressures ratio were known. If a reservoir or plenum pressure is fixed, the usual computational procedure involves the assumption of several critical pressures and the generation of several momentum terms to find the applicable critical pressure ratio and thereby the critical flow. The formulation of an equation of state make it possible to compute critical pressure ratios directly.
- Critical Tests for PRT Reactor
- This document authorizes the performance in accordance with the specifications noted, the PRTR Critical Tests described herein. The experiments described have the following objectives:
- "Crud" Scrubbing with a Purex-Type Zebra Cartridge
- Entrained solids have in the past been one of the major factors limiting solvent extraction fission product decontamination. These solids are known to be surface active, to absorb fission products, and to form interfacial films or "cruds". A major step in eliminating these cruds was taken when the Purex columns were modified to operate with the interfaces at the waste ends of the columns rather than the product ends. This type of operation tended to flush the solids out with the waste rather than the product streams.
- Decontamination of the Irradiated Rupture Prototype with Peroxide-Carbonate and Apace
- Considerable work has been done on decontamination procedures for both corrosion products and fission products. Testing of promising procedures under typical rupture conditions is an important phase of this work. The Irradiated Rupture Prototype (IRP) has been used since August in this evaluation work. This document is one of a series reporting these data.
- Density and Hydrogen Content of Uranium Oxide Cakes and Slurries
- The work described was undertaken to provide data for nuclear safety studies concerning NPF reprocessing equipment. The original objective was to determine the uranium density and water (hydrogen) content of UO2-H2O mixtures ranging from compact centrifuge cakes to dilute slurries. The scope was later expanded to include mixtures of UO2 with hydrocarbon oil and mixtures of UO3-H2O.
- Depassivation of 304-L Stainless Steel in Sulfex Decladding Solutions
- In the Sulfex process, stainless steel cladding, generally 304-L, is removed from fuel elements by dissolution in boiling four to six molar sulfuric acid. When cladding removal is complete, the decladding solution is removed is removed to waste. The exposed fuel cores are rinsed and dissolved in nitric acid.
- Description of the Thermal Hydraulics Laboratory at Hanford
- Report describing the thermal hydraulics facilities at Hanford Laboratories as well as its operation, equipment, and systems.
- Development and Demonstration of an Ion-Exchange Process for Kilogram-Scale Production of High Purity Promethium
- Report regarding "the development of a process suitable for large-scale separation of highly purified promethium from the trivalent fission product rare earth mixture" (p. 3).
- Development of a Welding Process for End Closures on NPR and KER Fuel Element
- With the development of the zircaloy-2 clad uranium fuel element, it was necessary to develop a method of closing the end of the element after the extruded tubes were cut to the desired length. Regardless of the ultimate design of the closure, a satisfactory zircaloy-2 fusion welding technique had to be worked out. The first fully successful welding was performed with electron beam welding in a vacuum chamber. To make satisfactory welds, the electron beam welding involved time and equipment, plus undue maintenance on the equipment.
- Development of a Welding Process for Spire-Can Fuel Elements
- The components for the present aluminum clad, Al-Si bonded, internally and externally cooled (I & E), uranium fuel elements are composed of impact extruded cans and spire caps as shown in Figure 1. This type of component requires two impact extrusions; however, in December, 1957, J. E. Ruffin proposed another design of component in which there was only one impact extrusion. For this component, Figure 2, the spire was impact extruded as a part of the can.
- Development of Pressures Tubing for the Plutonium Recycle Test Reactor
- Pressurized water nuclear reactors may be designed based upon either of two concepts: (1) pressure vessel, wherein the entire core is placed in a large, high strength fuel channels within a low pressure container. The Plutonium Recycle Test Reactor is a pressure tube type reactor. Selection of this basic type of pressurized water reactor depended to an appreciable extent upon the availability of suitable pressure tubing.
- A Device for Recording Fuel Element Dimension on Punched Paper Tape
- Report that "describes a system for translating the position of a strip-chart recorder potentiometer shaft into decimal digital form" (p. ii).
- Disposition of Contaminated Processing Equipment at Hanford Atomic Products Operation 1958-1959
- Operating plant experiences in restoring to useful service failed equipment contaminated through exposures to radioactive processes is of international interest to the nuclear industry. In accordance with a request from the Hanford Operations Office of the United States Atomic Energy Commission this report of the various measures taken in rehabilitating for re-use expensive engineered equipment has been compiled by the General Electric Company. This report is a review and summary of the disposition of contaminated major processing equipment at Hanford Atomic Products Operation during 1958-1959 and a listing of 1) equipment buried, 2) equipment stored, to be buried or decontaminated, and 3) equipment decontaminated and returned to service.
- Dissolution of Power Reactor Fuel Cores
- Report discussing the dissolution of simulated fuel cores. Equipment and procedures are discussed along with significant characteristics of the core materials.
- Economic Evaluation of a 300-Mw(e) Supercritical Pressure Power Reactor
- Report describing a 300-Mw(e) Supercritical Pressure Power Reactor's facilities, physics, economics, and problems encountered during its development. Appendix begins on page 103.
- Economic Evaluation of a 300 Mwe Fast Supercritical Pressure Power Reactor
- Report regarding Hanford Laboratories' 300 Mwe fast supercritical pressure power reactor. This includes descriptions of the plant and reactor, their operations, safety systems, and plant economics. Appendices begin on page 107.
- The Effect of Deionized Water on the Reactor Effluent Activities: Part 2 - Zirconium and Aluminum Clad Fuel
- Report containing experimental data testing the hypothesis that aluminum reactor tubes were "the site of most radioisotope production in the deionized water cooled system" and that zirconium tubes provided "only nominal participation" (p. 2).
- Effect of Gamma Radiation on the Physical Properties of Elastics
- This article presents data on the effects of gamma radiation on the physical properties of a number of different classes of plastics materials of construction which were screened to determine their capabilities and limitations for service in applications where ionizing radiation exists. The irradiations were carried out in air at ambient temperature at normal atmospheric pressure to various gamma radiation doses up to 1*10^9 r depending upon the stability of the materials to radiation exposure. The different classes of materials and materials within a class are compared.
- Effect of the PRTR Fuel Elemental Rupture Test Facility on Plutonium Recycle Program Objectives.
- To insure a full evaluation of the effects of in-reactor loops with respect to all phases of the Plutonium Recycle Program, a separate study of each loop has been undertaken. An initial study was carried out which analyzed the effects of in-reactor loops using the design criteria for the gas loop as a basis. As soon as the design criteria for the H2O high pressure loop became available, a more detailed evaluation was completed for that loop. Recent completion of the scope description of the PRTR fuel element rupture test loop now permits an individual evaluation of this loop.
- EGCR Lattice Radial and Angular Power Distribution
- One of the problems encountered in the design of the EGCR has been the possibility of bowing in the outer six rods of the seven rod cluster. The power distribution in an outer rod is asymmetrical with respect to the center of the rod. This causes a variation in the cladding temperature, and might result in unven thermal expansion, or bowing.
- EGCR Lattice Radial and Angular Power Distribution 2.6 w/o Enrichment
- The measurements reported here are companion measurements to those reported earlier in HW-63585. The only significant difference between the measurements is that 1.8 w/o enrichment UO2 fuel was used for the first set, and 2.6 w/o enrichment UO2 fuel was used for the measurements described in this report. The new results will be presented graphically, and for completeness, the details of the measurement will be included here as well as in HW-63585.
- Equations of State for Stream-Water Mixtures and Some Representative Applications Analysis
- The majority of two-phase flow problems involving equations of state are solved by use of point-wise utilization steam table values. In this manner, problems involving the use of the various flow equations of continuity, momentum and energy are generally forced into iterative solutions. Considerable effort towards the development of an analytical expression for the state equation seems indicated so as to simplify the analysis of two-phase problems, particularly the transient cases. The question of instability of state and mixture condition is particularly apparent in the analysis of systems undergoing phase transformation as demonstrated by the significant difference between simple theory and experimental critical flow determinations. The assumption of homogeneous, equilibrium mixtures is indicated as a first attack upon the problem.
- Evaluation of Aluminum for use in Reactors Cooled by High-Temperature Recirculating Water
- aluminum has been considered as a material for cladding fuel elements to be used in reactors cooled by high-temperature recirculating water. Aluminum is abundant, chap, has a relatively low cross section for thermal neutrons, and is readily fabricated into complex shapes. For these reasons, it appears that the use of aluminum rather than steel, Zircaloy or other candidate materials for fuel element cladding, would result in considerable savings in operating expense. The actual economics will of course depend on specific factors such as core materials and design of the reactor/. Aluminum appears particularly attractive for cladding the cores of aluminum-plutonium alloy proposed for the Plutonium Recycle Test Reactor since, in this case, the physical characteristics of the core and cladding would be practically identical.
- Evaluation of Fretting Corrosion of ZR-2
- Although the corrosion of resistance of Zircaloy--2 clad fuel elements in high temperature water is excellent, some concern has been expressed in regard to the resistance against fretting corrosion. Since this form of corrosion can proceed quite rapidly it could induce fuel element failure through penetration of the cladding. Fretting corrosion if it occurred, would take place either where the fuel element contacted the process tube or where portions of the fuel element contacted each other, e. g. wire-wrapping around a cluster of cylindrical fuel rods. This form of corrosion is defined as corrosion occurring at contact areas and as such, fretting of Zircaloy-2 could take place in many reactor concepts. This report details the results of testing performed to induce fretting corrosion of Zircaloy-2 and the direction of the future tests.
- Evaluation of Iron- and Nickel-Base Alloys for Medium and High Temperature Reactor Applications, Part 1
- Report discussing the evaluation of the "behavior of austenitic and ferritic stainless steels, iron-chromium-aluminum base alloys and nickel-base superalloys, in medium and high temperature simulated reactor environments" (p. 2).
- Evaluation of Properties of Irradiated Zircaloy-2 Pressure Tube from KER Loop 1
- Report describing the evaluation of the effects of irradiation on a high pressure water loop that ruptured while testing fuel elements. Includes photographs of the damaged tube as well as chemical analyses.
- Failure Test of a Double Chambered NaK-Filled Irradiation Capsule
- Report describing tests "in which a double chambered irradiation capsule containing NaK (sodium-potassium alloy) in its inner chamber was deliberately perforated to allow NaK (the heat transfer medium) and water to react within the capsule chambers" (p. 2). The report includes descriptions of the materials used in the tests, test procedures, and results.
- Final Design Report: DR-1 Gas Loop
- Report describing the performance, fission product tolerance, design, and costs of the DR-1 Gas Loop, which is an in-reactor test facility.
- Final Report A CG-791 Containment Test
- This report describes and evaluates the Hanford 105 reactor building structures' ability to withstand an internal pressure increase. The means by which their roof and wall surface would contain a pressure buildup 0.3 psi, and prevent contaminant release which might accompany a nuclear incident are discussed. Prototypes of the B, D, DR, F, and H reactor block wall configuration, the corrugated transite roof of the K reactors, and the corrugated transite walls of the K and C reactors are evaluated. Methods of securing certain building components are described, and a comparison of several candidate sealant coatings presented for consideration. These tests were performed at the request of the Design Operation, as part of CG-791, an existing reactor containment program. This series of tests represents only a part of the overall modifications program. A study of the reactor building containment design criteria is available in a Hanford document, HW-59236, by T. O. Brown.
- Final Report of Piping Components for Organic Coolants
- Organic compounds with high boiling points and low corrosion qualities have been considered for use as reactor coolants and moderators. Those being considered, however, have a greater tendency to leak than does water. A program to evaluate mechanical closures to be used for a leak-tight system for organic fluids has been established. Report HW-59446 covered descriptions of the facility used in this study and the primary results of tests covering two liquids, fittings, and instrumentation. This report will cover final results and other pertinent information.
- Fluidized-Bed Calcination Studies with Stimulated ICPP Waste Solution
- At the present time, high-radioactivity-level wastes at Hanford are neutralized and stored as liquid in underground tanks lined with mild steel. This method of storage is relatively inexpensive and is satisfactory on a short-term basis. However, on a long term basis, liquid storage is less desirable than solid storage because of the greater mobility of the liquid. In addition, storage as aa solid would significantly reduce the volume of waste stored. Consequently, various research and development studies have been undertaken in an attempt to develop a practical waste solidification.