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OSMOSE experiment representativity studies.

Description: The OSMOSE program aims at improving the neutronic predictions of advanced nuclear fuels through measurements in the MINERVE facility at the CEA-Cadarache (France) on samples containing the following separated actinides: Th-232, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, Cm-244 and Cm-245. The goal of the experimental measurements is to produce a database of reactivity-worth measurements in different neutron spectra for the separated heavy nuclides. This database can then be used as a benchmark for integral reactivity-worth measurements to verify and validate reactor analysis codes and integral cross-section values for the isotopes tested. In particular, the OSMOSE experimental program will produce very accurate sample reactivity-worth measurements for a series of actinides in various spectra, from very thermalized to very fast. The objective of the analytical program is to make use of the experimental data to establish deficiencies in the basic nuclear data libraries, identify their origins, and provide guidelines for nuclear data improvements in coordination with international programs. To achieve the proposed goals, seven different neutron spectra can be created in the MINERVE facility: UO2 dissolved in water (representative of over-moderated LWR systems), UO2 matrix in water (representative of LWRs), a mixed oxide fuel matrix, two thermal spectra containing large epithermal components (representative of under-moderated reactors), a moderated fast spectrum (representative of fast reactors which have some slowing down in moderators such as lead-bismuth or sodium), and a very hard spectrum (representative of fast reactors with little moderation from reactor coolant). The different spectra are achieved by changing the experimental lattice within the MINERVE reactor. The experimental lattice is the replaceable central part of MINERVE, which establishes the spectrum at the sample location. This configuration leads to a uniform well-behaved system so that the reactor configuration is in the fundamental mode. In fact, an important ...
Date: October 10, 2007
Creator: Aliberti, G. & Klann, R.
Partner: UNT Libraries Government Documents Department

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

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report.

Description: A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.
Date: October 13, 2009
Creator: Tentner, A.
Partner: UNT Libraries Government Documents Department

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

Description: Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.
Date: October 1, 1977
Creator: Esparza, V. & Sackett, K.E.
Partner: UNT Libraries Government Documents Department

End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

Description: Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.
Date: October 1, 1987
Creator: Richardson, K.D.
Partner: UNT Libraries Government Documents Department

Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

Description: Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment.
Date: October 1, 1979
Creator: Kang, C.S. & Torri, A.
Partner: UNT Libraries Government Documents Department

Safety Evaluation Report Restart of K-Reactor Savannah River Site

Description: In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues necessary to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied at the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. In August 1991, DOE issued DOE/DP-0090T, Safety Evaluation Report Restart of K-Reactor Savannah River Site Supplement 1.'' That document was the first Supplement to the April 1991 SER, and documented the resolution of 62 of the open items identified in the SER. This document is the second Supplement to the April 1991 SER. This second SER Supplement documents the resolution of additional open times identified in the SER, and includes a complete list of all remaining SER open items. The resolution of those remaining open items will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed.
Date: October 1, 1991
Partner: UNT Libraries Government Documents Department

Characterization of an energy source for modeling hypothetical core disruptive accidents in nuclear reactors. First interim report. [LMFBR]

Description: The expansion characteristics of the detonation products of a high-explosive energy source used to simulate the pressure-volume change relationships for sodium-vapor expansions during hypothetical core disruptive accidents in a Fast Test Reactor were determined experimentally. Rigid cylinder-piston experiments performed at two scales (ratio 1:3) were undertaken to determine a pressure-volume relationship as a function of source mass and expansion environment. Some of these measurements were compared with code calculations for the source.
Date: October 1, 1972
Creator: Cagliostro, D J & Florence, A L
Partner: UNT Libraries Government Documents Department

Potential seismic structural failure modes associated with the Zion Nuclear Plant. Seismic safety margins research program (Phase I). Project VI. Fragilities

Description: The Zion 1 and 2 Nuclear Power Plant consists of a number of structures. The most important of these from the viewpoint of safety are the containment buildings, the auxiliary building, the turbine building, and the crib house (or intake structure). The evaluation of the potential seismic failure modes and determination of the ultimate seismic capacity of the structures is a complex undertaking which will require a large number of detailed calculations. As the first step in this evaluation, a number of potential modes of structural failure have been determined and are discussed. The report is principally directed towards seismically induced failure of structures. To some extent, modes involving soil foundation failures are discussed in so far as they affect the buildings. However, failure modes involving soil liquefaction, surface faulting, tsunamis, etc., are considered outside the scope of this evaluation.
Date: October 1, 1979
Partner: UNT Libraries Government Documents Department

Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

Description: A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.
Date: October 1, 1979
Partner: UNT Libraries Government Documents Department

SCTI chemical leak detection test plan

Description: Tests will be conducted on the CRBRP prototype steam generator at SCTI to determine the effects of steam generator geometry on the response of the CRBRP chemical leak detection system to small water-to-sodium leaks in various regions of the steam generator. Specifically, small injections of hydrogen gas (simulating water leaks) will be made near the two tubesheets, and the effective transport times to the main stream exit and vent line hydrogen meters will be measured. The magnitude and time characteristics of the meters' response will also be measured. This information will be used by the Small Leak Protection Base Program (SG027) for improved predictions of meter response times and leak detection sensitivity.
Date: October 12, 1981
Partner: UNT Libraries Government Documents Department

High-temperature process heat. Interim design and cost status report, FY 1981

Description: Studies conductd on HTGR systems in FY 1980 were concluded in Application Study Reports to describe the preconceptual system designs to that point and discuss possible applications for three variations of the systems; the steam cycle/cogeneration plant, the higher temperature reformer plant, and the gas turbine concept. The HTGR-Reformer Application Study was conceived and directed to evaluate the HTGR-R with a core outlet temperature of 850/sup 0/C as a near-term Lead Project and as a vehicle to long-term HTGR Program Objectives. The scope of this effort included evaluations of the HTGR-R technology, evaluation of potential HTGR-R markets, assesment of the economics of commercial HTGR-R plants, and the evaluation of the program scope and expenditures necessary to establish HTGR-R technology through the completion of the Lead Project.
Date: October 1, 1981
Partner: UNT Libraries Government Documents Department

Evaluation of improved light water reactor core designs. Final progress report, September 1979. LWRCD-20

Description: The work conducted under this research project has developed information which supports in all respects the U.S. position evolved under the NASAP/INFCE programs with respect to the near and intermediate term potential for ore conservation in LWRs on the once-through fuel cycle. Moreover, in the even longer term, it has been confirmed that contention by Edlund and others that tight-pitch Pu/UO/sub 2/ PWR cores can achieve conversion ratios which may allow these reactors to provide a competitive energy source far into the ore-scarce post-2000 era.
Date: October 31, 1979
Partner: UNT Libraries Government Documents Department

Compendium of computer codes for the safety analysis of fast breeder reactors

Description: The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available.
Date: October 1, 1977
Partner: UNT Libraries Government Documents Department

Organizing for nuclear power facility development. Final draft report

Description: Centralized power development concepts have been of interest for some years and have been given considerable study in the past, e.g., the Congressionally-directed 1975 NRC studies. In general, while all such studies have concluded that such Centers did offer potential benefits and were feasible, in the mid-1970's, when most of these studies were done, the advantages did not appear to make use of Energy Centers on balance, preferable to continued conventional or dispersed siting. The DOE recognized that more recent circumstances, particularly the TMI accident, and the new imperatives which have been defined since that event for the proper conduct of the nuclear power ''enterprise'' may well have changed that balance. Centralized siting may today offer important benefits, but clearly those benefits can only be realized if the Center is effectively organized and if the institutional problems of organization (i.e., financial, political and jurisdictional) can be dealt with. Thus the Department of Energy asked the S.M. Stoller Corporation (SMSC) to outline the institutional factors and the organizational considerations to be taken into account in the establishment of nuclear power energy centers in the United States.
Date: October 1, 1980
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

TMI-2 Lessons Learned Task Force. Final report

Description: In its final report reviewing the Three Mile Island accident, the TMI-2 Lessons Learned Task Force has suggested change in several fundamental aspects of basic safety policy for nuclear power plants. Changes in nuclear power plant design and operations and in the regulatory process are discussed in terms of general goals. The appendix sets forth specific recommendations for reaching these goals.
Date: October 1, 1979
Partner: UNT Libraries Government Documents Department

Cover-gas seal program. Test report - sodium dip-seal wetting study. [at 450/sup 0/F]

Description: This report documents the tests conducted to find a reliable surface preparation method of treating the CRBRP dip seal blade (SA508 Class 2 steel) to insure its sodium wettability at 450F or less. Two techniques were established which depressed the sodium wetting temperature of SA 508, Class 2 dip seal blade material to 375F. These techniques were depositing an approx. 60 x 10/sup -6/ inch layer of tin on the blade surface by a brush-on plating process, and, by cleaning the blade surface with ultrasonics while it is immersed in sodium. The tin plating technique is recommended as the initial and primary surface preparation method and ultrasonics as a rewetting and backup technique. This work was conducted in support of the Sodium Dip Seal Feature Test, DRS 32.05.
Date: October 20, 1977
Creator: Carnevali, R.
Partner: UNT Libraries Government Documents Department

Characterization of actinide physics specimens for the US/UK joint experiment in the Dounreay Prototype Fast Reactor

Description: The United States and the United Kingdom are engaged in a joint research program in which samples of the higher actinides are irradiated in the Dounreay Prototype Fast Reactor in Scotland. The purpose of the porogram is (1) to study the materials behavior of selected higher actinide fuels and (2) to determine the integral cross sections of a wide variety of the higher actinide isotopes. Samples of the actinides are incorporated in fuel pins inserted in the core. For the fuel study, the actinides selected are /sup 241/Am and /sup 244/Cm in the form of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanides. For the cross-section determinations, the samples are milligram quantities of actinide oxides of /sup 248/Cm, /sup 246/Cm, /sup 244/Cm, /sup 243/Cm, /sup 243/Am, /sup 241/Am, /sup 244/Pu, /sup 242/Pu, /sup 241/Pu, /sup 240/Pu, /sup 239/Pu, /sup 238/Pu, /sup 237/Np, /sup 238/U, /sup 236/U, /sup 235/U, /sup 234/U, /sup 233/U, /sup 232/Th, /sup 230/Th, and /sup 231/Pa encapsulated in vanadium. Coincident with the irradiations, neutron flux and energy spectral measurements are made with vanadium-encapsulated dosimeter materials located within the same fuel pins.
Date: October 1, 1983
Creator: Walker, R.L.; Botts, J.L.; Cooper, J.H.; Adair, H.L.; Bigelow, J.E. & Raman, S.
Partner: UNT Libraries Government Documents Department

Acoustic leak detection and ultrasonic crack detection

Description: A program is under way to assess the effectiveness of current and proposed techniques for acoustic leak detection (ALD) in reactor coolant systems. An ALD facility has been constructed and tests have begun on five laboratory-grown cracks (three fatigue and two thermal-fatigue and two field-induced IGSCC specimens. After ultrasonic testing revealed cracks in the Georgia Power Co. HATCH-1 BWR recirculation header, the utility installed an ALD system. Data from HATCH-1 have given an indication of the background noise level at a BWR recirculation header sweepolet weld. The HATCH leak detection system was tested to determine the sensitivity and dynamic range. Other background data have been acquired at the Watts Bar Nuclear Reactor in Tennessee. An ANL waveguide system, including transducer and electronics, was installed and tested on an accumulator safety injection pipe. The possibility of using ultrasonic wave scattering patterns to discriminate between IGSCCs and geometric reflectors has been explored. Thirteen reflectors (field IGSCCs, graphite wool IGSCCs, weld roots, and slits) were examined. Work with cast stainless steel (SS) included sound velocity and attenuation in isotropic and anisotropic cast SS. Reducing anisotropy does not help reduce attenuation in large-grained material. Large artificial flaws (e.g., a 1-cm-deep notch with a 4-cm path) could not be detected in isotropic centrifugally cast SS (1 to 2-mm grains) by longitudinal or shear waves at frequencies of 1 MHz or greater, but could be detected with 0.5-MHz shear waves. 13 figures.
Date: October 1, 1983
Creator: Kupperman, D.S.; Claytor, T.N. & Groenwald, R.
Partner: UNT Libraries Government Documents Department

Chemical and Radiochemical Constituents in Water from Wells in the Vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1996

Description: The U.S. Geological Survey, in response to a request from the U.S. Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office (IBO), samples water from 13 wells during 1996 as part of a long-term project to monitor water quality to the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility (NRF), Idaho National Engineering and Environmental Laboratory, Idaho. The IBO requires information about the mobility of radionuclide- and chemical-waste constituents in the Snake River Plain aquifer. Waste-constituent mobility is determined principally by (1) the rate and direction of ground-water flow; (2) the locations, quantities, and methods of waste disposal; (3) waste-constituents chemistry; and (4) the geochemical processes taking place in the aquifer. The purpose of the data-collection program is to provide IBO with water-chemistry data to evaluate the effect of NRF activities on the water quality of the Snake River Plain aquifer. Water samples were analyzed for naturally occurring constituents and man-made contaminants.
Date: October 1, 1999
Creator: Knobel, L. L.; Bartholomay, R. C.; Tucker, B. J. & Williams, L. M.
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

Description: Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated. (FS)
Date: October 31, 1978
Creator: Marvin, M.D.
Partner: UNT Libraries Government Documents Department