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Accurate Nuclear Fuel Burnup Analyses; Eighth Quarterly Progress Report, (September - November 1963)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analyses; First Quarterly Report, (December 1961 - February 1962)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analyses: Fourth Quarterly Progress Report September-November, 1962
Work has continued on the development of accurate nuclear fuel burnup analysis. Work performed during the fourth quarter is summarized here.
Accurate Nuclear Fuel Burnup Analyses; Ninth Quarterly Progress Report, (December 1963 - February 1964)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analyses; Second Quarterly Progress Report, (March - May 1962)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analyses: Third Quarterly Progress Report June - August, 1962
Work has continued on the development of accurate nuclear fuel burnup analysis. Work performed by the third quarter of 1962 is summarized.
Accurate Nuclear Fuel Burnup Analysis Quarterly Progress Report: Seventh Quarter, June 1963 - August 1963
Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.
Accurate Nuclear Fuel Burnup Analysis Quarterly Progress Report: Sixth Quarter, March 1963 - May 1963
Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.
AEC Fuel Cycle Program Design and Fabrication of Special Assembly 9-L : Irradiation Performance Test of UO2-Cermet Fuel
Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter of the fuel rods is 1.308 cm (0.515 inch), and the clad wall thickness if 0.051 cm (0.020 inch). the cladding material is Type-304 stainless steel. The fuel pellets were all centerless ground to achieve a uniform outside diameter and thereby control the pellet-to-clad diametral clearance within a range of 0.076 to 0.102 mm (0.003 to 0.004 inch). Operation of the fuel rods will be at high specific power levels with surface heat fluxes of about 157 W/cm(2) (~500,000 Btu/h-ft(2)). The assembly was designed for a lifetime of 4.1 x 10(20) fission/cc (15,000 MWD/T) exposure.
Fuel Cycle Program Progress Report: Thirteenth Quarter, July-September 1963
Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities
Amendment No. 1 to Preliminary Hazards Summary Report For The Dresden Nuclear Power Station
Commonwealth Edison Company for the purpose of supplementing the license application for the Dresden Nuclear Power Station submits herewith Amendment No.1 to the portion of the Final Hazards Summary Report entitled Preliminary Hazards Summary Report.
Amendment No. 2 To License Application For Nuclear Test Reactor
GE is amending its application of 6/5/57 to construct and operate the Nuclear Test Reactor in order to incorporate changes in procedure and equipment.
Amendment No. 2 To Preliminary Hazards Summary Report For The Dresden Nuclear Power Station
This report is the second amendment to the Preliminary Hazards Summary Report for the Dresden Nuclear Power Station (GEAP-1044) submitted to the United States Atomic Energy Commission on September 3, 1957.
Amendment No. 4 to Hazards Summary Report For The Dresden Nuclear Power Station
This report is an amendment to the Preliminary Hazards Summary Report (1) and the Operating Procedures and Emergency Plans (5) for the Dresden Nuclear Power Station, submitted to the United States Atomic Energy Commission on September 3, 1957, and June 5, 1958, respectively.
Analysis of UO2 Grain Growth Data From "Out of Pile" Experiments
Summary: Data on equlaxed UO2 grain growth from "out of pile" experiments have been gathered from all known sources and analyzed to determine the relationship between the grain size developed and annealing temperature and between grain size and the time at temperature. On the basis of the analysis, an equation relating gain size to time and temperate has been selected that appears to best describe the data considered as a whole. The coefficients in this grain growth equation have been evaluated to indicate the variance between different investigations and/or different UO2 sinters. The general applicability and limitations of "out of pile" grain growth data for the determination of temperatures in the microstructures of irradiated UO2 are discussed. Specific recommendations are made for the conditions under which grain size can be reliably employed as a temperature indicator. The objective in undertaking this analysis on UO2 grain growth was to obtain a working relationship between UO2 grain size and annealing time and/or temperature, and also a measure of the potential variation in the relationship. The intended use of the results was the determination of temperatures based on the grain sized observed in the post-irradiation metallographic examination of AEC-Euratom High Performance UO2 Program fuel capsules. The results are being reported in the belief that they will be of use in the analysis of other fuel experiments.
Application of Boron Carbide Nickel Dispersion to a Prototype Control Rod
Previously reported results on the testing of small samples of boron carbide dispersed in nickel by electrolytic codeposition were adequately encouraging to lead to the development of a prototype control rod for operation in the Vallecitos Boiling Water Reactor. The operation of the control rod has been entirely satisfactory.
Applications of Strain Cycling Considerations to Superheat Fuel Design
A potential performance limitation of superheat fuel is the susceptibility of the fuel cladding to low cycle fatigue failure. Two simplified analytical methods are presented to estimate the cyclic lifetime of circular superheat fuel cladding. One failure relation is based on a displacement method. The other failure relation is based on a stress method. These relations were compared with data from the literature, and with data involving damage obtained by Reynolds. A recommended design procedure involving the relations is presented. The technique was applied to the SADE 4B experiment with moderate success. These cycling relations involve only mechanical damage imposed by cycling, with a modification for additional damage caused by radiation; they do not include any other potential performance limiting mechanisms, such as stress corrosion, which are normally factored into the over-all fuel design. This work work done under Task C (Materials Development) of the Nuclear Superheat Project, AEC Contract AT(04-3)-189 - Project Agreement 13.
Automatic Control of T7 Tanker Boiling Water Reactor Propulsion System: Preliminary Design and Economic Evaluation
From introduction: "This report sets forth the results of a technical and economic analysis of automatic propulsion system control as a possible design improvement in the direct cycle boiling water reactor propulsion system in a T7 tanker."
Bounce III
BOUNCE III is a program which was written for the IBM-704 as part of a study of the parameters of the neutron distribution in a large thermal column. The program calculates the eigenvalues and corresponding eigenvectors of the matrix resulting from a diffusion-theory, multigroup description of the thermal neutron spectrum.
Burnout Conditions for Nonuniformly Heated Rod in Annular Geometry, Water at 1000 PSIA
Tests were run at the General Electric Company, Atomic Power Equipment Department, to determine the burnout conditions for a non-uniformly heated rod in an annular geometry.
Calculated Scattering Kernels For Light Water at 23C, 42C, 61C, and 82C
This report contains a listing of the bound-proton kernels for neutron scattering which were calculated in conjunction with the USAEC Control Rod Materials Program.
Collected Methods for Analysis of Sodium Metal
Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine the following impurities: Calcium, Carbon, Chromium, Iron, Lithium, Nickel, Oxygen, Potassium, and Zirconium. Spectrographic analysis was used to determine many other impurities. Sodium samples obtained from experimental apparatus operated as part of the work being conducted for Atomics International were analyzed by these methods.
Compact Control Rod Drive Study For a Boiling Water Reactor in a T7 Tanker
The reason for initiating the compact drive study for the T7 tanker was to investigate control rod drive size, location, and removal space requirement factors and select the control rod drive mechanism which would allow optimization of the over-all size of the containment vessel. Approximately twelve mechanical/hydraulic control rod drive arrangements were considered during this study.
Compilation of Techniques Used By Vallecitos Radioactive Materials Laboratory
Equipment and techniques for remote examination of irradiated fuel assemblies applicable to the Maritime Program are described. The following subjects are covered: visual and photographic examination, dimensional measurements, gamma activity scanning, fission gas release and fuel rod void volume determinations, density measurements, metallographic examination, and radiochemical burnup analysis.
Consumers Baffle Two-Phase Air-Water Flow Tests in a One-Fifth Scale Model
Tests in a one-fifth scale clear plastic model were conducted to investigate the flow characteristics of the asymmetrical riser configuration in the Consumers Big Rock power plant.
Consumers Big Rock Point Nuclear Power Reactor Stability Analysis
This report presents the results of an analysis which was undertaken to investigate the power stability of the Consumers Big Rock Point Nuclear Power Reactor.
Control Worth of B4C Rods
This report considers the theoretical evaluation of a system for gaining increased control strength and increased control lifetime and presents a theoretical model which is applicable to conventional multigroup diffusion theory.
A Controlled-Environment Steam Corrosion Facility
Abstract; Technical report describing a low-flow autoclave system developed for out-of-pile corrosion testing of materials in controlled environment steam up to 500 C. The system has been set up in triplicate to provide for the exposure of various zirconium alloys to steam at 300, 400, and 500 C. The oxygen and hydrogen of the steam were controlled at 25 ppm and 3 ppm, respectively, to simulate the gas conditions from radiolytic water decomposition found in a boiling water reactor. The autoclave internals were so designed to result in a temperature variation between specimens under test of less than 2C.
Design and Fabrication of Coextruded Stainless Steel Clad UO2 Fuel Rods
A process was developed in which stainless steel-clad UO2 fuel rods are fabricated by high-temperature coextrusion. The process has a potential of being a more economical method for the preparation of stainless steel-clad UO2 fuel rods than the conventional pellet process. Consequently, it was considered advantageous to evaluate the irradiation characteristics of fuel rods fabricated in this manner. Therefore, 24 coextruded fuel rods were manufactured for evaluation in a reactor. The required amounts of UO2 and clad were soaked in separate containers at 1875 and 760 degree C, respectively. The containers were removed from their respective furnaces and were coextruded in one pass. A force of 450 to 475 tons was used, and a reduction ratio of 18 to 1 was obtained. The coextruded rods were cut to the approximate length, and the ends were sealed with an acid-resistant tape. The carbon steel can covering the stainless steel clad was removed by immersion in 1:1 nitric acid for 20 minutes. The rods were visually inspected, the specified lengths of clad and fuel were obtained by machining, and the correct diameter was obtained by belt sanding. The fabrication of the fuel rods was completed by inserting the plenum support tubes and welding in the end plugs. Nineteen of these fuel rods were sent to the Atomic Power Equipment Department (APED) for irradiation in the Vallecitos Boiling Water Reactor (VBWR). The irradiation of 12 of these rods was begun in August 1961, while irradiation of 3 rods was begun in July 1962. The irradiations will continue until an average burnup of at least 10,000 Mwd/t is achieved by some of the fuel rods.
Design and Fabrication of Fuel Rods Containing Sintered UO2 Extrusions - Assembly 11L
The extrusion forming of ceramic powders may be economically interesting in the field of nuclear fuel fabrication. When applied to the forming of rod-type uranium dioxide fuel, extrusion processes have been able to produce cylindrical bodies with length-to-diameter ratios much greater than those of the conventional die-pressed pellets. Furthermore, after being sintered, the extrusions have exhibited densities at least as high as those of sintered pellets. Thus, extrusion forming may offer reductions in handling during fabrication and, at the same time, provide a fuel with improved performance characteristics by decreasing the number of discontinuities in the fuel column. This report reviews the production of these extrusions, sets forth some of their characteristics, describes the materials and processes employed in cladding them, and records the pre-irradiation data pertaining to the finished fuel rods and fuel assembly. Irradiation of the fuel assembly in the VBWR was initiated on July 17, 1962.
Design and Fabrication of Pellet Fuel Rods Clad With Thin Wall Stainless Steel
Summary: Stainless steel clad nuclear fuel cycle costs can be reduced to those associated with Zircaloy clad fuel or potentially lower by reducing the thickness of the clad tube wall until performance penalties offset the savings associated with the reduction in parasitic neutron absorption. To demonstrate the feasibility and investigate performance capabilities of thin clad fuel rods for power reactor application an assembly was fabricated with 0.0127 cm (5 mil) thick stainless steel cladding tubes for irradiation testing in the Vallecitos Boiling Water Reactor (VBWR). The fuel bundle was placed in the VBWR and irradiation was begun in November, 1961. The irradiation is scheduled to continue until the target exposure of 2.74 x 10(20) fissions/cc (10,000 MWD/T of uranium) average burnup is reached. Destructive examinations of fuel rods will be performed at regular intervals throughout life to determine fuel rod performance.
Design and Operating Experience of the ESADA Vallecitos Experimental Superheat Reactor (EVESR)
Summary: "The various design features significant to superheat are described for the 12-1/2 MW., 960 psig, 1050° F, steam-cooled, low-enriched, annular-fueled, experimental superheat reactor built by the General Electric Company at the Vallecitos Atomic Power Laboratory. Results obtained during the first six months of full-power operation, on the emergency cooling system, core thermal performance, and pressure vessel temperatures are presented and compared with predictions. Operating experience with over-all reactor system is also discussed."
Design Report: Superheat Strain - Cycle Capsule
In order to investigate the low frequency strain cycle fatigue for tubular sheath geometries an apparatus was designed and fabricated for laboratory and reactor experiments. The design of this apparatus is described herein.
The Determination of Fission Product Gamma Doses
In this paper arbitrary limits of the general fission source gamma problem are set. Then, by assuming cooling of at least one day, it is shown that only twelve different fission product gamma sources need ever be considered.
The Development of a Scheduling Computer For The Big Rock Plant
The basic work for development of a scheduling computer for the Consumers' Big Rock Plant is outlined in this report. The computer's purpose is to make feasible higher power densities by operation closer to fuel element burnout limits, and to maximize fuel burnup.
Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Fifth Quarterly Progress Report, April 1-June 30, 1963
Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. The development of pulsed neutron source techniques for large power reactors has led to a new theoretical model recently developed by E. Garelis and J.L. Russell, Jr. The theory is presently based on a bare, one-group model with m-delayed precursors and takes all spatial modes into account. Results indicate, however, that the application of this model is much broader. Experiments were designed and carried out to both verify this new theory and to demonstrate the performance of the experimental hardware in a large power reactor.
Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Fourth Quarterly Progress Report, January 1-March 31, 1963
Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. The experimental work centers around aspects of detecting neutrons in the presence of 10/sup 7/ r/hr gamma fields. Boric acid experiments and Humboldt Bay experiments are reported.
Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Seventh and Eighth Quarterly Progress Report, October 1, 1963-March 31, 1964
Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. In the course of this program, a new theory was suggested and an experimental apparatus was designed and built. Experiments were carried out to test the new model. This present report contains additional data and information extracted from the experiments at PG&E Humboldt Bay Power Reactor at Eureka, California. During the last days of 1963 a number of control rod and fuel bundle worth measurements were made in the ESADA Vallecitos Experimental Superheat Reactor (EVESR) using the (k[beta]/[script l] technique. A description of the experiments is given in the text of the report and some results are reported. A computer program was written to perform the data analysis of the pulsed neutron experiments and the code is discussed in the Appendix.
Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Sixth Quarterly Progress Report, July 1-September 30, 1963
Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. The development of pulsed neutron source techniques for large power reactors has led to a new theoretical model recently developed by E. Garelis and J.L. Russell, Jr. The theory is presently based on a bare, one-group model with m-delayed precursors and takes all spatial modes into account. Results indicate, however, that the application of this model is much broader. Experiments were designed and carried out to both verify this new theory and to demonstrate the performance of the experimental hardware in a large power reactor.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 2
The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 3
The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 4
The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 5
The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 6
The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 7
The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). One task is in progress: Task I - Data Logging and Computer System. The work on the other tasks is being planned and initiated.
Economic Evaluation of Control Rod Materials and Fabrication Processes
Control rod materials, designs, and fabrication processes are compared for their relative economies. Control rod lifetime data are calculated with a simple approximation of nuclear worth depreciation. These data are used in conjunction with the estimated fabrication costs to determine the cost of using several absorber materials in a typical power reactor on a cost-per-year basis. The effect the control system has on core power density and fuel lifetime is included.
The Effects of Non-Uniform Flow and Concentration Distributions and the Effect of the Local Relative Velocity on the Average Volumetric Concentration in Two-Phase Flow
Abstract: A general expression which can be used either for predicting the average volumetric concentration or for analyzing and interpreting experimental data is derived. The analysis takes into account both the effect of non-uniform flow and concentration profiles as well as the effect of the local relative velocity between phases. The first effect is taken into account by a distribution parameter, whereas the latter is accounted for by the weighted average drift velocity.
Enclosure Pressure Calculation Method
A method of determining enclosure pressure in the event of a reactor rupture is presented and a sample calculation is shown. This method was used in calculating the design pressure of the Dresden Nuclear Power Station enclosure.
Enclosure Section of the Hazards Summary Report for the Dresden Nuclear Power Station
The General Electric Company is designing and building a 180,000 kilowatt nuclear power plant for the Commonwealth Edison Company at a site near the confluence of the Kankakee and Des Plaines Rivers in Grundy County, Illinois, about 47 miles southwest of Chicago. The plant will be known as the Dresden Nuclear Power Station, and will employ a nuclear reactor of the dual-cycle boiling water type.
Environmental Testing of a B4C-Ni Prototype Control Rod
Summary: A prototype control rod containing absorber plates made from an electro- deposited dispersion of boron carbide in nickel was tested in the VBWR. It was exposed to the reactor environment of 545 degree F boiling water and thermal neutron fluxes (perturbed) which ranged from 0.6 to 1.1 x 10/sup 13/ nv for 2236 hours over a period of six months. The maximum B/sup 10/ burnup achieved during the test period was 1.8 percent. After irradiation, the rod was examined. The results of the examination are summarized below: (1) The B/sub 4/C-- Ni plate assembly did not undergo significant dimensional changes during irradiation. (2) Numerous blisters developed on both the outer and inner surfaces of three of the four plates. Blistering was more severe on the outer surface than on the inner, and was most severe in a large region located in the lower half of plate 4. Metallographic examination revealed that the blisters were located only in the 2- mil protective nickel overlay covering the B/sub 4/C-- Ni dispersion. It was concluded that they formed from the buildup of gas pressure at the Ni: Ni-- B/sub 4/C interfaces, rather than from corrosion attack. Helium from the B/sup 10/(n alpha )Li/sup 7/ reaction probably contributed to this pressure. However it is conjectured that the major gas was very likely hydrogen, possibly generated and dissolved in the nickel during electroplating and then released to defects at the Ni: Ni--B/sub 4/C interface during reactor exposure. The variation in the degree of blistering among the four plates of the prototype indicated that the blistering was related to variations in the fabrication process. Failure of the nickel overlay was not observed in any of the blisters examined metallographically, and the underlying B/sub 4/C-- Ni appeared to be in good condition. (3) Evidence of corrosion …
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