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The Fast Flux Test Facility built on safety

Description: No other high-tech industry has grown as fast as the nuclear industry. The information available to the general public has not kept pace with the rapid growth of nuclear data---its growth has outpaced its media image and the safety of nuclear facilities has become a highly debated issue. This book is an attempt to bridge the gap between the high-tech information of the nuclear industry and its understanding by the general public. It explains the three levels of defense at the Fast Flux Test Facility (FFTF) and why these levels provide an acceptable margin to protect the general public and on-site personnel, while achieving FFTF's mission to provide research and development for the US Department of Energy (DOE).
Date: January 1, 1989
Partner: UNT Libraries Government Documents Department

Boilup threshold for the bottled-up transition phase pool. [LMFBR]

Description: Since the inception of the hypothesized transition phase, for the late stages of a postulated LMFBR accident, there has been a continual effort to characterize the anticipated conditions of such a hypothetical state. To date, several techniques and methods have been employed to analyze the potential for energetic criticality. As part of this effort, an arbitrary criterian of monotonical dispersiveness has been employed as the measure of diminished recriticality potential. The various attempts to demonstrate monotonic dispersiveness have included experimental demonstrations, theoretical approaches, and integrated analysis using both. As part of this treatment, flow regime maps have been devised as a convenient method for inferring the state of dispersiveness. They included bubbly, churn turbulent, foam and drop fluidized regimes. Of these, foam and drop fluidized regimes were considered the most dispersive. The main thrust of the analysis to date, including flow regime maps, relates primarily to the open pool configuration. However, the bottled configuration may be the pertinent geometry. To date, no reliable escape path has been demonstrated for the advanced stages of core disruption, although strong potential escape mechanisms have been identified and are currently being analyzed. The bottled pool is examined in this paper.
Date: October 1, 1978
Creator: Martin, F. J.
Partner: UNT Libraries Government Documents Department

Breeder Spent Fuel Handling Program multipurpose cask design basis document

Description: The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref.
Date: September 1, 1985
Creator: Duckett, A.J. & Sorenson, K.B.
Partner: UNT Libraries Government Documents Department

As-built description of the EBR-II, Run 97 dosimetry experiment

Description: A dosimetry experiment has been designed and fabricated for inclusion in the Experimental Breeder Reactor-II (EBR-II) during Run 97 in a Row 2 position. Various types of dosimeter material have been included in the single B-7c pin from 60 cm below midplane to 60 cm above. This report contains the as-built description of irradiation hardware and a detailed description of the dosimetry.
Date: December 1, 1978
Creator: Long, C.L.; Ulseth, J.A. & Lippincott, E.P.
Partner: UNT Libraries Government Documents Department

Achieving higher count rates with EDX

Description: An automated energy-dispersive x-ray spectrometer (EDX) was developed for the close-coupled analysis of mixed U, Pu oxide fuel pellets. (Close-coupled means the analytical glove box is closely adjacent to the production line.) The L ..cap alpha.. fluorescences of U and Pu were chosen. The advantages and disadvantages of the system are given. (DLC)
Date: January 1, 1980
Creator: Lambert, M.C.
Partner: UNT Libraries Government Documents Department

Acoustic emission: who needs it - and why

Description: Acoustic emission (AE) is an emerging NDT method that offers attractive capabilities for monitoring structural integrity and characterizing materials behavior. Although its limitations are such that it should not be regarded as a panacea, AE is proving to be a viable complement to the other NDT methods. The paper examines the extent and reasons for the growing industrial interest in AE. Some of the inherent limitations of conventional NDT methods are discussed, and several surveys of defects found during the manufacture and operation of pressure boundary components are reviewed. Although welds and weld-affected areas are the most likely locations for significant defects, very little experience is available to date to describe the AE response during impending pressure vessel failures due to weld associated defects. Acoustic emission offers potential for providing increased assurance of both initial, and continued, structural integrity. Furthermore, if AE is properly applied in conjunction with recently proposed fitness-for-purpose criteria, it may be possible to reduce present manufacturing costs without compromising actual structural adequacy. This technology is exhibiting definite signs of increasing industrial maturity, as evidenced by the recent availability of industrial standards, and the activities of various AE related technical groups throughout the world.
Date: May 1, 1979
Creator: Spanner, J. C.
Partner: UNT Libraries Government Documents Department

Aerosol behavior during sodium pool fires in a large vessel: CSTF tests AB1 and AB2

Description: Two large-scale aerosol behavior tests, using sodium pool fires as the aerosol source, were performed in the Containment Systems Test Facility (CSTF). The tests were conducted to characterize the properties and behavior of sodium aerosol particles formed and aged in a large containment vessel. The 20-m high, 850-m/sup 3/ CSTF containment building in regard to parameters that affect agglomeration and gravitational settling. In both tests, sodium burned for one hour in a 4.38-m/sup 2/ pool, and the only difference between them was that steam was injected during the second test, simulating the release of water vapor from heated concrete.
Date: June 1, 1979
Creator: Hilliard, R.K.; McCormack, J.D. & Postma, A.K.
Partner: UNT Libraries Government Documents Department

Analyses of eigenvalue bias and control rod worths in FFTF (Fast Flux Test Facility)

Description: The Fast Flux Test Facility (FFTF) core loading during its ninth operating cycle was significantly different from that of previous cycles because of the presence of the Core Demonstration Experiment (CDE). The CDE consists of a number of axially blanketed fuel assemblies and internal blankets prototypic of advanced oxide cores in Liquid Metal Reactors (LMR). In preparation for the Cycle 9 reload design effort, a careful assessment of control rod worth and reactivity calculations for Cycles 1 through 8 was made. The goal of this study was to establish calculational biases and reduce uncertainties factored into the reload design calculations. These analyses helped assure that the operational objectives for Cycle 9 were met.
Date: January 1, 1987
Creator: Nelson, J.V.; Dobbin, K.D.; Wootan, D.W. & Campbell, L.R.
Partner: UNT Libraries Government Documents Department

Analysis of piping systems with nonlinear supports subjected to seismic loading

Description: An analytical study of effects of nonlinearities in piping supports on response to seismic excitation is presented. Response calculations for simplified single degree of freedom nonlinear models are used to illustrate sensitivity to stiffness variations, lost motion and impact damping. Seismic responses of typical spans of various sizes of piping supported by both linear and nonlinear constraints are compared to assess the support load magnifications due to impacting. The idealized nonlinear piping support models are integrated with a finite element model of a large piping system. Time domain seismic responses of the nonlinear piping system are compared to loads determined by a standard linearized seismic response spectra technique.
Date: March 1, 1979
Creator: Barta, D. A.
Partner: UNT Libraries Government Documents Department

Analytical validation of the CACECO containment analysis code. [LMFBR]

Description: The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code.
Date: August 1, 1979
Creator: Peak, R.D.
Partner: UNT Libraries Government Documents Department

Computer generated movies at Hanford

Description: The message contained in the results of a large computer program is often difficult to present to large groups of people. This difficulty may be overcome by using 16mm color movie techniques. This presentation shows the results of directly using computer output to explain a story about fuel behavior during a power transient.
Date: October 1, 1979
Creator: Lewis, C.H. & Fox, G.L.
Partner: UNT Libraries Government Documents Department

CONACS: the DOE safety analysis system

Description: The CONtainment Analysis Code System (CONACS) is a large, comprehensive scientific simulation system for predicting conditions in an LMR facility following the occurrence of a postulated accident. It has now been developed to a stage of completion that can be referred to as a limited operational version. This version forms a permanent portion of the ultimate system. Because CONACS was developed with change in mind, it is now possible to draw on this strength to respond to changing requirements arising from advanced design concepts. The generalized design applications in the nuclear and non-nuclear fields and the quality assurance applied to the project make those adaptations reliable. In this paper the results of prototype tests and the implications of limited version tests are presented along with a brief description of CONACS and its relationship to LMR design optimization and cost reduction.
Date: March 1, 1985
Creator: Martin, F.J.; Armstrong, G.R. & Niccoli, L.G.
Partner: UNT Libraries Government Documents Department

Containment air cleaning for LMFBRs

Description: A variety of air cleaning concepts was evaluated for potential use in future sodium-cooled breeder reactors. A 3-stage aqueous scrubber system was selected for large-scale demonstration testing under conditions similar to those postulated for containment venting and purging during reactor melt-through accidents. Two tests were performed in the Containment Systems Test Facility using a quench tank, a jet venturi scrubber and a high efficiency fibrous scrubber in series. The results of two tests with Na/sub 2/0/sub 2/ and Na0H aerosol and NaI vapor are presented showing >99.9% removal of Na/sub 2/0/sub 2/ and Na0H and >99.7% for NaI.
Date: January 1, 1979
Creator: Hilliard, R. K.; McCormack, J. D.; Postma, A. K. & Owen, R. K.
Partner: UNT Libraries Government Documents Department

Controlled biaxial strain-rate test results from unirradiated 20% CW 316 stainless steel cladding at constant temperature

Description: Constant temperature controlled biaxial strain-rate (CBSR) tests were performed on unirradiated 20% CW 316 stainless steel reactor cladding. Tests were made at hoop strain-rates of 1.2 x 10/sup -5//s, 1.2 x 10/sup -4//s, 6 x 10/sup -4//s, and 1.2 x 10/sup -3//s. For each of these strain rates, tests were performed at 425/sup 0/C, 540/sup 0/C, and 650/sup 0/C. The data from these tests are examined and compared with previously reported tensile data. A simple relationship between tensile and CBSR tests is developed, allowing easy correlation of the CBSR results.
Date: January 1, 1979
Creator: Cannon, N. S.
Partner: UNT Libraries Government Documents Department

Cost reduction through improved seismic design

Description: During the past decade, many significnt seismic technology developments have been accomplished by the United States Department of Energy (USDOE) programs. Both base technology and major projects, such as the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR) plant, have contributed to seismic technology development and validation. Improvements have come in the areas of ground motion definitions, soil-structure interaction, and structural analysis methods and criteria for piping, equipment, components, reactor core, and vessels. Examples of some of these lessons learned and technology developments are provided. Then, the highest priority seismic technology needs, achievable through DOE actions and sponsorship are identified and discussed. Satisfaction of these needs are expected to make important contributions toward cost avoidances and reduced capital costs of future liquid metal nuclear plants. 23 references, 12 figures.
Date: January 1, 1984
Creator: Severud, L.K.
Partner: UNT Libraries Government Documents Department

Crack propagation in irradiated B/sub 4/C induced by swelling and thermal gradients

Description: Irradiation testing of hot-pressed boron carbide pellets in a fast neutron reactor flux has demonstrated its susceptibility to cracking. Two characteristic mechanisms of pellet cracking are distinguished. The first is commonly observed in ceramics, i.e., thermal stress cracking. A second mechanism is associated with internal strains resulting from swelling gradients within the boron carbide pellets. From the results presented here, it appears that thermal and swelling gradient stresses are opposite in direction, and are reflected in the crack patterns that are produced. Pellet cracking effects have not been identified as the cause of a cladding rupture. Large diameter pellets propagate stable cracks at low burnup levels from the circumference radially inward as the result of tensile thermal stress on the outside of the pellet. Smaller pellets with swelling tended to crack in the center of the pellet as a result of radial swelling gradients.
Date: January 1, 1979
Creator: Hollenberg, G.W. & Basmajian, J.A.
Partner: UNT Libraries Government Documents Department

Cross sections required for FMIT dosimetry

Description: The Fusion Materials Irradiation Test (FMIT) facility, currently under construction, is designed to produce a high flux of high energy neutrons for irradiation effects experiments on fusion reactor materials. Characterization of the flux-fluence-spectrum in this rapidly varying neutron field requires adaptation and extension of currently available dosimetry techniques. This characterization will be carried out by a combination of active, passive, and calculational dosimetry. The goal is to provide the experimenter with accurate neutron flux-fluence-spectra at all positions in the test cell. Plans have been completed for a number of experimental dosimetry stations and provision for these facilities has been incorporated into the FMIT design. Overall needs of the FMIT irradiation damage program delineate goal accuracies for dosimetry that, in turn, create new requirements for high energy neutron cross section data. Recommendations based on these needs have been derived for required cross section data and accuracies.
Date: May 2, 1980
Creator: Gold, R.; McElroy, W.N.; Lippincott, E.P.; Mann, F.M.; Oberg, D.L.; Roberts, J.H. et al.
Partner: UNT Libraries Government Documents Department

Damage analysis and fundamental studies. Quarterly progress report, April-June 1980

Description: The DAFS program element is a national effort composed of contributions from a number of National Laboratories and other government laboratories, universities, and industrial laboratories. It was organized by the Materials and Radiation Effects Branch, Office of Fusion Energy, DOE, and a Task Group on Damage Analysis and Fundamental Studies which operates under the auspices of that Branch. The purpose of this series of reports is to provide a working technical record of that effort for the use of the program participants, for the fusion energy program in general, and for the Department of Energy. This report is organized along topical lines in parallel to a Program Plan of the same title so that activities and accomplishments may be followed readily, relative to that Program Plan.
Date: August 1, 1980
Partner: UNT Libraries Government Documents Department

Cesium behavior and control in sodium systems. [LMFBR]

Description: A series of capsule tests were performed to screen candidate packing materials for a Cs trap. Specimens of medium density graphite rho = 1.8 to 1.7 gm/cm/sup 3/ and a low density amorphous carbon foam, Reticulated Vitreous Carbon (RVC) rho = 0.06 gm/cm/sup 3/ were tested. X-ray diffraction verified the hexagonal structure of the graphite specimens and the amorphous structure of the RVC. The test capsules contained approximately 0.1 mCi (3.7 x 10/sup 3/ Kbq)/sup 137/Cs dissolved in 30 gm of sodium. The behavior of Cs in a circulating sodium system was studied in the Fission Product Transport Loop (FPTL). Tracer isotopes, 1 mCi (3.7 x 10/sup 4/KBq)/sup 137/Cs, 1 mCi (3.7 x 10/sup 4/KBq)/sup 134/Cs and 0.5 mCi/sup 22/(1.85 x 10/sup 4/KBq)/sup 22/Na were added to the 19 kg sodium inventory in the loop. The tracer distributions in the loop were monitored with a collimated GeLi detector and multichannel analyzer.
Date: March 1, 1980
Creator: Colburn, R.P. & Maffei, H.P.
Partner: UNT Libraries Government Documents Department

Chemical decontamination of metals

Description: A metal decontamination process based upon removal of contamination by treatment with a cerium (IV)-nitric acid solution (or other redox agent in nitric acid) is feasible and highly promising. The technique is effective in dissolving the surface layer of stainless steel. Dissolution rates of approximately 1.5 mils/h were demonstrated with cerium (IV)-nitric acid solutions. Removal of plutonium contamination from stainless steel was demonstrated in laboratory tests, in which activity levels were reduced from greater than 5 x 10/sup 5/ counts per minute to nondetectable levels in approximately one hour at 90/sup 0/C. Removal of paint from stainless steel surfaces was also demonstrated. Advantages of this process over other chemical solutions include: (1) The solutions are not high salt systems; therefore, there is potentially less waste generated. (2) Cerium(IV) in nitric acid is a good dissolution agent for plutonium oxide. (3) Regeneration of Ce(IV) during the decontamination is accomplished by electrolysis. (4) The process should be effective for irregularly shaped equipment. (5) It could be effective as a spray or a flow-through system. 13 figures.
Date: October 1, 1979
Creator: Partridge, J.A. & Lerch, R.E.
Partner: UNT Libraries Government Documents Department

Combining within and between instrument information to estimate precision

Description: When two instruments, both having replicated measurements, are used to measure the same set of items, between instrument information may be used to augment the within instrument precision estimate. A method is presented which combines the within and between instrument information to obtain an unbiased and minimum variance estimate of instrument precision. The method does not assume the instruments have equal precision.
Date: January 1, 1980
Creator: Jost, J.W.; Devary, J.L. & Ward, J.E.
Partner: UNT Libraries Government Documents Department

Comparison of aerosol behavior during sodium fires in CSTF with the HAA-3B code. [LMFBR]

Description: Four large-scale tests using sodium fire aerosol sources have been carried out in the Containment System Test Facility (CSTF). Two of the tests employed pool fires and two used spray fires as the aerosol source. Because the CSTF containment vessel is approximately half-scale (20.3 m in height) of a typical reactor building, the CSTF results have provided a large-scale proof test of the HAA-3B Code. For the two pool fire tests, the measured and predicted airborne concentrations were in good agreement when the aerosol source term was based on post-test measurements of aerosol formation, accounting for water vapor uptake.
Date: March 1, 1980
Creator: Postma, A.K. & Owen, R.K.
Partner: UNT Libraries Government Documents Department

Correlation of macroscopic material properties with microscopic nuclear data

Description: Two primary irradiation-induced changes occur during neutron irradiation: the displacement of atoms forming crystal defects and the transmutation of atoms into either gaseous or solid products. The material scientist studying irradiation damage to material by fusion-produced neutrons is faced with several questions: Is the nature of high-energy (14-MeV) displacement damage the same as or different from that caused by fission neutrons (< 2 MeV). How do the high helium concentrations expected in a fusion environment affect the material properties. What effects do solid transmutation products have on the behavior of the irradiated materials. In the past few years, much work has been done to answer these questions. This paper reviews recent work in this area.
Date: December 18, 1981
Creator: Simons, R.L.
Partner: UNT Libraries Government Documents Department

Cover gas seals. 11 - FFTF-LMFBR seal-test program, January-March 1974

Description: The objectives of this program are to: (1) conduct static and dynamic tests to demonstrate or determine the mechanical performance of full-size (cross section) FFTF fuel transfer machine and reactor vessel head seals intended for use in a sodium vapor - inert gas environment, (2) demonstrate that these FFTF seals or new seal configuration provide acceptable fission product and cover gas retention capabilities at LMFBR Clinch River Plant operating environmental conditions other than radiation, and (3) develop improved seals and seal technology for the LMFBR Clinch River Plant to support the national objective to reduce all atmospheric contaminations to low levels.
Date: January 1, 1974
Creator: Kurzeka, W.; Oliva, R. & Welch, F.
Partner: UNT Libraries Government Documents Department