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.

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.

Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.

Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.

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 …

Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities

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 …

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.

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.

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 …

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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 …

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.