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Progress reports for Gen IV sodium fast reactor activities FY 2007.

Description: An important goal of the US DOE Sodium Fast Reactor (SFR) program is to develop the technology necessary to increase safety margins in future fast reactor systems. Although no decision has been made yet about who will build the next demonstration fast reactor, it seems likely that the construction team will include a combination of international companies, and the safety design philosophy for the reactor will reflect a consensus of the participating countries. A significant amount of experience in the design and safety analysis of Sodium Fast Reactors (SFR) using oxide fuel has been developed in both Japan and France during last few decades. In the US, the traditional approach to reactor safety is based on the principle of defense-in-depth, which is usually expressed in physical terms as multiple barriers to release of radioactive material (e.g. cladding, reactor vessel, containment building), but it is understood that the 'barriers' may consist of active systems or even procedures. As implemented in a reactor design, defense-in-depth is classed in levels of safety. Level 1 includes measures to specify and build a reliable design with significant safety margins that will perform according to the intentions of the designers. Level 2 consists of additional design measures, usually active systems, to protect against unlikely accidental events that may occur during the life of the plant. Level 3 design measures are intended to protect the public in the event of an extremely unlikely accident not foreseen to occur during the plant's life. All of the design measures that make up the first three levels of safety are within the design basis of the plant. Beyond Level 3, and beyond the normal design basis, there are accidents that are not expected to occur in a whole generation of plants, and it is in this class that severe accidents, ...
Date: October 4, 2007
Creator: Cahalan, J. E. & Tentner, A. M.
Partner: UNT Libraries Government Documents Department

Assessment of a small pressurized water reactor for industrial energy

Description: An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton.
Date: October 4, 1977
Creator: Klepper, O. H.; Fuller, L. C. & Myers, M. L.
Partner: UNT Libraries Government Documents Department

Review of FFTF and CRBRP control rod systems designs

Description: The evolution of the primary control rod system design for FFTF and CRBR, beginning with the initial choice of the basic concepts, is described. The significant component and systems tests are reviewed together with the test results which referenced the development of the CRBR primary control rod system design. Modifications to the concepts and detail designs of the FFTF control rod system were required principally to satisfy the requirements of CRBR, and at the same time incorporating design refinements shown desirable by the tests.
Date: October 4, 1977
Creator: Pitterle, T. A. & Lagally, H. O.
Partner: UNT Libraries Government Documents Department

Stress analysis of cylindrical pressure vessels with closely spaced nozzles by the finite-element method. Volume 1. Stress analysis of vessels with two closely spaced nozzles under internal pressure. [BWR; PWR; MULT-NOZZLE code]

Description: A finite-element computer program, MULT-NOZZLE, was developed for the stress analysis of cylindrical pressure vessels with two or three closely spaced reinforced nozzles. MULT-NOZZLE consists of two modules which may be operated independently. The first module, FEMG, automatically prepares a finite-element mesh including the nodal point coordinates, finite-element connectivities, mesh options, and boundary value specifications for input to the finite-element solution module SAP3M. SAP3M, which is a modified and improved version of the SAP3 computer program, computes the nodal point displacements and stress tensor components, and prints and/or stores the results for later postprocessing. The accuracy of the SAP3M module is demonstrated by comparison studies of two classical theory-of-elasticity problems: a simply supported beam and a thick-walled ring under internal pressure loading. A complete discussion of MULT-NOZZLE is presented in four volumes. Volume develops the finite-element idealization for pressure vessels with two idential radially attached closely spaced nozzles for internal pressure loading. The nozzles may be unreinforced or fully reinforced according to the rules of the ASME Boiler and Pressure Vessel Code and may be located in either a longitudinal or a transverse plane of the vessel. Validation of the program for analyzing this type of structure is demonstrated by the analysis of three two-nozzle pressure vessel models and comparison of results with experimental data. In general, quite satisfactory results were obtained.
Date: October 4, 1977
Creator: Tso, F.K.W.; Bryson, J.W.; Weed, R.A. & Moore, S.E.
Partner: UNT Libraries Government Documents Department

Radiation damage study on the lithium hydride SNAP shield

Description: Radiation damage may occur to the lithium hydride shields as a result of the reaction Li{sup 6}(n, {alpha})H{sup 3}. There is evidence in the literature indicating both the existence and absence of radiation damage to the SNAP shields. It is believed that there is a high probability that there will be damage and that it will adversely affect the properties of the shield. This damage may take the form of: (1) volume expansion of the hybrids, (2) void formation within the hybrids, and (3) gas pressure build-up in the shield container. Based upon the results of experiments with lithium fluoride, which may serve as a model for the hydride, there appears to be a threshold neutron dose which volume expansion effects can not be removed by annealing. Similarly, above the threshold dose, intercrystalline voids, formed as a result of radiation damage, appear to increase in size with increasing temperature. It has been established that at the SNAP shield operating conditions, essentially all of the hydrogen formed will recombine with free lithium. The helium atoms, however, remain trapped interstitially, in intercrystalline voids, or along subgrain boundaries. Appreciable amounts of helium gas are not released until the melting point of the hydride is approached. An insignificant portion of the hydrogen in the shield is lost by permeation of the stainless steel shield container at the SNAP 10 operating conditions. 23 refs.
Date: October 4, 1961
Creator: Doctor, R.D.
Partner: UNT Libraries Government Documents Department

Compilation of data on 51 ruptured slugs

Description: The following tabulation includes information on all uranium slug failures which have occurred through September 26, 1951. The four suspect slugs which are listed were discharged from tubes which gave strong indication of containing ruptures. Although no obvious rupture could be found among the slugs from these tubes, the listed pieces exhibited defects which may be incipient ruptures.
Date: October 4, 1951
Creator: O'Keefe, D.P.
Partner: UNT Libraries Government Documents Department