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Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

Description: The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results are related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.
Date: October 12, 1990
Creator: Ruggles, A.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

Reliability of leak detection systems in LWRs

Description: In this paper, NRC guidelines for leak detection will be reviewed, current practices described, potential safety-related problems discussed, and potential improvements in leak detection technology (with emphasis on acoustic methods) evaluated.
Date: October 1, 1986
Creator: Kupperman, D.S.
Partner: UNT Libraries Government Documents Department

Oxygen suppression in boiling water reactors. Quarterly report 2, January 1--March 31, 1978

Description: Boiling water reactors (BWR's) generally use high purity, no-additive feedwater. Primary recirculating coolant is neutral pH, and contains 100 to 300 ppB oxygen and stoichiometrically related dissolved hydrogen. However, oxygenated water increases austenitic stainless steel susceptibility to intergranular stress-corrosion cracking (IGSCC) when other requisite factors such as stress and sensitization are present. Thus, reduction or elimination of the oxygen in BWR water may preclude cracking incidents. One approach to reduction of the BWR coolant oxygen concentration is to adopt alternate water chemistry (AWC) conditions using an additive(s) to suppress or reverse radiolytic oxygen formation. Several additives are available to do this but they have seen only limited and specialized application in BWR's. The objective of this program is to perform an in-depth engineering evaluation of the potential suppression additives supported by critical experiments where required to resolve substantive uncertainties.
Date: October 1, 1978
Creator: Burley, E.L.
Partner: UNT Libraries Government Documents Department

FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback

Description: FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600.
Date: October 1, 1978
Creator: Shober, R.A.; Daly, T.A. & Ferguson, D.R.
Partner: UNT Libraries Government Documents Department

Precipitates in irradiated Zircaloy

Description: Precipitates in high-burnup (>20 MWd/kg U) Zircaloy spent-fuel cladding discharged from commercial boiling- and pressurized-water reactors have been characterized by TEM-HVEM. Three classes of primary precipitates were observed in the irradiated Zircaloys: Zr3O (2 to 6 nm), cubic-ZrO2 (greater than or equal to 10 nm), and delta-hydride (35 to 100 nm). The former two precipitations appears to be irradiation induced in nature. Zr(Fe/sub x/Cr/sub 1-x/)2 and Zr2(Fe/sub x/Ni/sub 1-x/) intermetallics, which are the primary precipitates in unirradiated Zircaloys, were largely dissolved after the high burnup. It seems, therefore, that the influence of the size and distribution of the intermetallics on the corrosion behavior may be quite different for the irradiated Zircaloys.
Date: October 1, 1985
Creator: Chung, H.M.
Partner: UNT Libraries Government Documents Department

Special hazards report - I E fuel loads

Description: This report has been prepared in answer to the request from the AEC contained in the letter of October 1, 1957, from A. T. Gifford, HOO to A. B. Greninger. As requested, the report is of a summary nature and a more complete discussion of many of the points considered will be found in the references listed. The report is directed primarily at C reactor but some discussion of the other reactors is also included. A description of the proposed utilization of I E slugs in C reactor together with the associated power increase schedule is presented below. The reasons for changing to the I E element are presented together with a comparison of solid and I E slugs in the C reactor. The changes being made in C reactor under CG 600 are described. The operational characteristics of the C reactor using solid and I E elements are compared and finally the nuclear safety status of all of the Hanford reactors assuming I E loadings is reviewed.
Date: October 15, 1957
Creator: Brown, J.H.; Fullmer, G.C.; Trumble, R.E. & VanWormer, F.W.
Partner: UNT Libraries Government Documents Department

(Environmental impact of radionuclide release during the Kyshtym, Windscale, and Chernobyl accidents)

Description: The traveler attended the conference, Comparative Assessment of the Environmental Impact of Radionuclides Released During Three Major Nuclear Accidents: Kyshtym, Windscale, and Chernobyl and presented an invited paper giving a western perspective of the Kyshtym (Chelyabinsk-40) high-level waste explosion that took place in 1957. Papers of interest to several ORNL and DOE programs were presented. These covered the topics of accident source terms, atmospheric dispersion, resuspension, chemical and physical forms of contamination (e.g., hot'' particles), environmental contamination and transfer, radiological effects on humans and the environment, and countermeasures. The traveler also made valuable contacts with Soviet and other scientists related to an ongoing assessment sponsored by the International Union of Radioecologists of releases from the Chelyabinsk-40 site. This included an agreement in principle for direct participation by key Soviet scientists.
Date: October 22, 1990
Creator: Trabalka, J.R.
Partner: UNT Libraries Government Documents Department

Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

Description: The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.
Date: October 1, 1978
Creator: Philbin, J.S.; Nelson, W.E. & Rosenstroch, B.
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

Safety margins associated with containment structures under dynamic loading. [BWR Mark I]

Description: A technical basis for assessing the true safety margins of containment structures involved with MARK I boiling water reactor reevaluation activities is presented. It is based on the results of a plane-strain, large displacement, elasto-plastic, finite-element analysis of a thin cylindrical shell subjected to external and internal pressure pulses. An analytical procedure is presented for estimating the ultimate load capacity of the thin shell structure, and subsequently, for quantifying the design margins of safety for the type of loads under consideration. For defining failure of structures, a finite strain failure criterion is derived that accounts for multiaxiality effects.
Date: October 30, 1978
Creator: Lu, S.C.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor safety studies for the Division of Accident Evaluation quarterly progress report, January 1-March 31, 1985

Description: Modeling, code development, and analyses of the modular High-Temperature Gas-Cooled Reactor (HTGR) continued with work on the side-by-side design. Fission-product release and transport experiments were completed. A description and assessment report on Oak Ridge National Laboratory HTGR safety codes was issued.
Date: October 1, 1985
Creator: Ball, S.J.; Cleveland, J.C.; Harrington, R.M.; Weber, C.F. & Wilson, J.H.
Partner: UNT Libraries Government Documents Department

Studies of aged cast stainless steel from the Shippingport reactor

Description: Charpy-impact and tensile tests were conducted on several cast stainless steel materials from the Shippingport reactor. Baseline mechanical properties for unaged material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550{degree}C and water-quenched, or material from the cooler region of the component. The materials indicate relatively modest decreases in impact energy. The results show good agreement with estimations based on accelerated laboratory-aging studies. Correlations for estimating thermal-aging degradation of cast stainless steels indicate that the degree of embrittlement of the Shippingport materials is low. The minimum room-temperature impact energies that would ever be achieved after long-term aging are >75 J/cm{sup 2} (>45 ft{center dot}lb) for all materials. The estimated activation energies for embrittlement range from 150 to 230 kJ/mole. The estimated fracture toughness J-R curves for the materials are also presented. 14 refs., 16 figs.
Date: October 1, 1990
Creator: Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Irradiation-induced sensitization of austenitic stainless steel in-core components

Description: High- and commercial-purity specimens of Type 304 SS from BWR absorber rod tubes, irradiated during service to fluence levels of 6 {times} 10{sup 20} to 2 {times} 10{sup 21} n{center dot}cm{sup {minus}2} (E > 1 MeV) in two reactors, were examined by Auger electron spectroscopy to characterize irradiation-induced grain boundary segregation and depletion of alloying and impurity elements, which have been associated with irradiation-assisted stress corrosion cracking (IASCC) of the steel. Ductile and intergranular fracture surfaces were produced by bending of hydrogen-charged specimens in the ultra-high vacuum of Auger microscope. The intergranular fracture surfaces in high-fluence commercial-purity material were characterized by relatively high levels of Si, P, and In segregation. An Auger energy peak at 59 eV indicated either segregation of an unidentified element or formation of an unidentified compound on the grain boundary. In contrast to the commercial-purity material, segregation of the impurity elements and intergranular failure in the high-purity material were negligible for a similar fluence level. However, grain boundary depletion of Cr was more significant in high-purity material than in commercial-purity material, which indicates that irradiation-induced segregation of impurity elements and depletion of alloying elements are interdependent. 7 refs., 10 figs., 2 tabs.
Date: October 1, 1990
Creator: Chung, H.M.; Sanecki, J.E.; Ruther, W.E. & Kassner, T.F.
Partner: UNT Libraries Government Documents Department

The PEGASUS Drive: A nuclear electric propulsion system for the space exploration initiative

Description: The advantages of using electric propulsion for propulsion are well-known in the aerospace community. The high specific impulse, lower propellant requirements, and lower system mass make it a very attractive propulsion option for the Space Exploration Initiative (SEI), especially for the transport of cargo. One such propulsion system is the PEGASUS Drive (Coomes et al. 1987). In its original configuration, the PEGASUS Drive consisted of a 10-MWe power source coupled to a 6-MW magnetoplasmadynamic (MPD) thruster system. The PEGASUS Drive propelled a manned vehicle to Mars and back in 601 days. By removing the crew and their associated support systems from the spacecraft and by incorporating technology advances in reactor design and heat rejection systems, a second generation PEGASUS Drive can be developed with an alpha less than two. Utilizing this propulsion system, a 400-MT cargo vehicle, assembled and loaded in low Earth orbit (LEO), could deliver 262 MT of supplies and hardware to Mars 282 days after escaping Earth orbit. Upon arrival at Mars the transport vehicle would place its cargo in the desired parking orbit around Mars and then proceed to synchronous orbit above the desired landing sight. Using a laser transmitter, PEGASUS would provide 2-MWe on the surface to operate automated systems deployed earlier and then provide surface power to support crew activities after their arrival. The additional supplies and hardware, coupled with the availability of megawatt levels of electric power on the Mars surface, would greatly enhance and even expand the mission options being considered under SEI. 9 refs., 1 fig., 1 tab.
Date: October 1, 1990
Creator: Coomes, E.P. & Dagle, J.E.
Partner: UNT Libraries Government Documents Department

Historical perspective, economic analysis, and regulatory analysis of the impacts of waste partitioning-transmutation on the disposal of radioactive wastes

Description: Partitioning-transmutation, sometimes called actinide burning, is an alternative approach to high-level radioactive waste management. It consists of removing long-lived radionuclides from wastes and destroying those radionuclides, thus reducing the long-term hazards of radioactive waste. It was studied in detail in the 1970's. New developments in technology and other factors are resulting in a reexamination of this waste management option. This report consists of three papers which summarize the historical work, update the analysis of the costs of waste disposal, and describe current regulatory requirements which might be impacted by P-T. The papers provide a starting point for future research on P-T. 152 refs., 2 figs., 19 tabs.
Date: October 1, 1990
Creator: Forsberg, C.W.; Croff, A.G. & Kocher, D.C.
Partner: UNT Libraries Government Documents Department

Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

Description: A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.
Date: October 1, 1985
Creator: Adams, S.R.
Partner: UNT Libraries Government Documents Department

Theory and application of a quasi-Eulerian fluid element for the STRAW code. [LMFBR]

Description: Two-dimensional finite-element models for the treatment of the nonlinear, transient response of fluids and structures are described. The fluid description is quasi-Eulerian, so that the mesh can move independently of the material, and it includes a new finite-element upwinding scheme. The structural description is based on a corotational formulation in which the coordinate system is embedded in the elements, which is applicable to arbitrarily large rotations. The interface between the fluid and structure permits relative sliding, but because of the description of the quasi-Eulerian fluid, the nodes of the fluid and structure can be allowed to remain contiguous. Modeling procedures for treating the various aspects of subassemblies, such as the narrow fluid channels, the fuel bundles which are immersed in the coolant, and the axial flow are developed. Calculations are made for a symmetric 7-subassembly cluster and compared to experimental results. In addition, the application to a 19-subassembly cluster is described.
Date: October 1, 1978
Creator: Kennedy, J.M. & Belytschko, T.B.
Partner: UNT Libraries Government Documents Department

High-temperature, radiation-tolerant electronics for the MMW (Multi-megawatt) Space Reactor Program

Description: One of the objectives of the Multi-Megawatt (MMW) space reactor program is to determine, within the next five years, what types of power electronic devices would be suitable for MMW space power applications. Suitable devices must be able to withstand high temperatures and high radiation fields. After investigating the literature on solid state device and miniature vacuum tube technologies, we have concluded that the miniature vacuum tube technology is, currently, the most promising. The main reason for choosing this technology, is because miniature vacuum tubes can operate at very high temperatures (775 K or potentially higher) and are tolerant to very high neutron fluence and gamma dose. Although there are still problems to be solved before miniature vacuum tubes can be used, the time required for their development will be much shorter than the five year period required by the MMW space reactor program. 13 refs., 3 figs., 3 tabs.
Date: October 17, 1986
Creator: Yee, J.H.; Orvis, W.J.; McConaghy, C. & Ciarlo, D.R.
Partner: UNT Libraries Government Documents Department

Status of FFTF startup program and future FFTF utilization

Description: A brief FFTF project description is provided which includes general plant siting information, general layout, plant design parameters, description of principal systems and components, and description of support facilities. The current status of the FFTF project is provided, including status of plant construction, overall status of the plant checkout and test program, status of operating authorization and plant operating procedures and personnel, and status of reactor core components and experiments. Specific information on the acceptance test program and early program results is discussed. The role of FFTF in the future breeder program is described, including its objectives for verification of plant system and components designs and operability and use as an irradiation test facility.
Date: October 1, 1978
Creator: Hard, E. N.; Bliss, R. J. & Olson, O. L.
Partner: UNT Libraries Government Documents Department

Secondary side photographic techniques used in characterization of Surry steam generator

Description: Characterization of the generator's secondary side prior to destructive removal of tubing presents a significant challenge. Information must be obtained in a radioactive field (up to 15 R/h) throughout the tightly spaced bundle of steam generator tubes. This report discusses the various techniques employed, along with their respective advantages and disadvantages. The most successful approach to nondestructive secondary side characterization and documentation was through use of in-house developed pinhole cameras. These devices provided accurate photographic documentation of generator condition. They could be fabricated in geometries allowing access to all parts of the generator. Semi-remote operation coupled with large area coverage per investigation and short at-location times resulted in significant personnel exposure advantages. The fabrication and use of pinhole cameras for remote inspection is discussed in detail.
Date: October 1, 1984
Creator: Sinclair, R.B.
Partner: UNT Libraries Government Documents Department