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Effect of decontamination on aging processes and considerations for life extension

Description: The basis for a recently initiated program on the chemical decontamination of nuclear reactor components and the possible impact of decontamination on extended-life service is described. The incentives for extending plant life beyond the present 40-year limit are discussed, and the possible aging degradation processes that may be accentuated in extended-life service are described. Chemical decontamination processes for nuclear plant primary systems are summarized with respect to their corrosive effects on structural alloys, particularly those in the aged condition. Available experience with chemical cleaning processes for the secondary side of PWR steam generators is also briefly considered. Overall, no severe materials corrosion problems have been found that would preclude the use of these chemical processes, but concerns have been raised in several areas, particularly with respect to corrosion-related problems that may develop during extended service.
Date: October 1, 1987
Creator: Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Development of internal reforming carbonate fuel cell stack technology

Description: Activities under this contract focused on the development of a coal-fueled carbonate fuel cell system design and the stack technology consistent with the system design. The overall contract effort was divided into three phases. The first phase, completed in January 1988, provided carbonate fuel cell component scale-up from the 1ft{sup 2} size to the commercial 4ft{sup 2} size. The second phase of the program provided the coal-fueled carbonate fuel cell system (CGCFC) conceptual design and carried out initial research and development needs of the CGCFC system. The final phase of the program emphasized stack height scale-up and improvement of stack life. The results of the second and third phases are included in this report. Program activities under Phase 2 and 3 were designed to address several key development areas to prepare the carbonate fuel cell system, particularly the coal-fueled CFC power plant, for commercialization in late 1990's. The issues addressed include: Coal-Gas Related Considerations; Cell and Stack Technology Improvement; Carbonate Fuel Cell Stack Design Development; Stack Tests for Design Verification; Full-Size Stack Design; Test Facility Development; Carbonate Fuel Cell Stack Cost Assessment; and Coal-Fueled Carbonate Fuel Cell System Design. All the major program objectives in each of the topical areas were successfully achieved. This report is organized along the above-mentioned topical areas. Each topical area has been processed separately for inclusion on the data base.
Date: October 1, 1990
Creator: Farooque, M.
Partner: UNT Libraries Government Documents Department

Fuel cells for extraterrestrial and terrestrial applications

Description: The fuel cell is a nineteenth century invention and a twentieth century technology development. Due to the high power and energy density, high efficiency, reliability, and production of pure water, hydrogen-oxygen fuel cell systems have no competition as auxiliary power sources for space vehicles. The alkaline fuel cell system is a well developed and proven technology for this application. The solid polymer electrolyte system may be its future competitor. The energy crisis of 1973 stimulated research, development and demonstration of the phosphoric acid, molten carbonate, solid oxide and solid polymer electrolyte fuel cell systems using natural gas, petroleum or coal derived hydrogen (and carbon monoxide for the high temperature systems) for terrestrial applications. The direct methanol-air fuel cell is still an electrochemist's dream. Though considerable technological advances have been made, the present price of crude oil, and the high capital costs and limited lifetime of fuel cell systems impede their terrestrial applications in the developed countries. Conversely, the potential for lower capital costs of labor intensive manufacturing processes and the relatively higher fossil fuel prices make these systems more attractive for such applications in the developing countries. 11 refs.
Date: January 1, 1987
Creator: Srinivasan, S.
Partner: UNT Libraries Government Documents Department

Isotec final report

Description: General Atomic (GA) developed processes to prepare NdSe/sub 1.5-x/. The partial pressure of Se/sub 2/ above the NdSe/sub 1.5-x/ was found to be a function of x. The thermoelectric properties and the friability of the hot pressed element were also a function of x. Process modification changed the value of x, providing a method to control the properties of N-type NdSe/sub 1.5-x/ elements. A method of joining NdSe/sub 1.5-x/ to a nickel hot cap was developed using a gold foil intermediate. NdSe/sub 1.5-x/ was diffusion-bonded to PbTe by the same process developed for Gd/sub 2/Se/sub 3/. Couples were fabricated from (Cu,Ag)/sub 2/Se/Fe/(Bi,Sb)/sub 2/Te/sub 3/ and NdSe/sub 1.5-x/ PbTe elements. Limited couple test data were obtained. General Atomic examined methods to develop high-efficiency couples made from segmented elements containing (Cu,Ag)/sub 2/Se P-type material and NdSe/sub 1.5-x/ N-type material. Techniques were developed to reproducibly prepare and hot press (Cu,Ag)/sub 2/Se segments and elements with a metallurgically joined MoRe hot cap. This hot cap and a W-26/Re wire mesh used to make the hot cap junction were found to be stable in contact with (Cu,Ag)/sub 2/Se at 750/sup 0/C. Stability of (Cu,Ag)/sub 2/Se to current flow in a thermal gradient was examined. This report describes specifications for vapor suppression to improve the stability of P-type (Cu,Ag)/sub 2/Se in thermal gradient life tests and presents life test information on match-loaded (Cu,Ag)/sub 2/Se and (Cu,Ag)/sub 2/Se/Fe/(Bi,Sb)/sub 2/Te/sub 3/ elements.
Date: November 1, 1981
Creator: Elsner, N.B.; Chin, J.; Reynolds, G.H.; Norman, J.H.; Bass, J.C. & Staley, H.G.
Partner: UNT Libraries Government Documents Department

Assessment of martensitic steels as structural materials in magnetic fusion devices

Description: This manuscript documents the results of preliminary experiments and analyses to assess the feasibility of incorporating ferromagnetic martensitic steels in fusion reactor designs and to evaluate the possible advantages of this class of material with respect to first wall/blanket lifetime. The general class of alloys under consideration are ferritic steels containing from about 9 to 13 percent Cr with some small additions of various strengthening elements such as Mo. These steels are conventionally used in the normalized and tempered condition for high temperature applications and can compete favorably with austenitic alloys up to about 600/sup 0/C. Although the heat treatment can result in either a tempered martensite or bainite structure, depending on the alloy and thermal treatment parameters, this general class of materials will be referred to as martensitic stainless steels for simplicity.
Date: January 1, 1980
Creator: Rawls, J.M.; Chen, W.Y.K.; Cheng, E.T.; Dalessandro, J.A.; Miller, P.H.; Rosenwasser, S.N. et al.
Partner: UNT Libraries Government Documents Department

Progress report No. 31 for a program of thermoelectric generator testing and RTG degradation mechanisms evaluation

Description: Thermal conductivity measurements of the p-type and n-type selenide alloys after 19,800 hours and 6,500 hours, respectively, continue to show good agreement with the 3M Co. published data. Ingradient testing of n-legs after 8900 hours show comparable performance to the reported 3M data. Ingradient testing of p-legs at accelerated test conditions has completed 5432 hours and performance shows reasonable correlation with accelerated temperature and current gradients. This test has been discontinued to accommodate ingradient performance testing of the new design p-type legs with the Astroquartz/slurry wrap. Weight loss testing of the new design p-type TPM217 (no partition) with the Astroquartz/slurry wrap has been continued at both ingradient and isothermal test conditions. A check of one sample after 1000 hours at T/sub H/900/sup 0/C and T/sub C/150/sup 0/C shows a very low weight loss rate, comparable to that reported by the 3M Co. and these tests are continuing. A post-mortem examination of the RCA Reference Generator, which operated for 62,339 hours prior to failure, has been initiated and the metalligraphic analysis was completed during this report period. The remaining MHW generator, Q1-A, on test has accumulated 14,191 hours and performance remains stable. An evaluation of the performance of the Voyager 1 and 2 RTGs after 14,661 hours and 15,160 hours of operation, respectively, versus the DEGRA prediction code, continues to show good agreement.
Date: July 1, 1979
Creator: Stapfer, G. & Garvey, L.
Partner: UNT Libraries Government Documents Department

TFE Verification Program semiannual report for the period ending March 31, 1990

Description: The objective of the semiannual progress report is to summarize the technical results obtained during the latest reporting period. The information presented herein will include evaluated test data, design evaluations, the results of analyses and the significance of results. The program objective is to demonstrate the technology readiness of a TFE suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full-power life of 7 years. The TFE Verification Program builds directly on the technology and data base developed in the 1960s and early 1970s in an AEC/NASA program, and in the SP-100 program conducted in 1983, 1984 and 1985. In the SP-100 program, the attractive features of thermionic power conversion technology were recognized but concern was expressed over the lack of fast reactor irradiation data. The TFE Verification Program addresses this concern. 6 refs., 67 figs., 37 tabs.
Date: July 1, 1990
Partner: UNT Libraries Government Documents Department

CO sub 2 emissions from coal-fired and solar electric power plants

Description: This report presents estimates of the lifetime carbon dioxide emissions from coal-fired, photovoltaic, and solar thermal electric power plants in the United States. These CO{sub 2} estimates are based on a net energy analysis derived from both operational systems and detailed design studies. It appears that energy conservation measures and shifting from fossil to renewable energy sources have significant long-term potential to reduce carbon dioxide production caused by energy generation and thus mitigate global warming. The implications of these results for a national energy policy are discussed. 40 refs., 8 figs., 23 tabs.
Date: May 1, 1990
Creator: Keith, F.; Norton, P. & Brown, D.
Partner: UNT Libraries Government Documents Department

TFE Verification Program

Description: The objective of the semiannual progress report is to summarize the technical results obtained during the latest reporting period. The information presented herein will include evaluated test data, design evaluations, the results of analyses and the significance of results. The program objective is to demonstrate the technology readiness of a TFE suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full-power life of 7 years. The TF Verification Program builds directly on the technology and data base developed in the 1960s and 1970s in an AEC/NASA program, and in the SP-100 program conducted in 1983, 1984 and 1985. In the SP-100 program, the attractive but concern was expressed over the lack of fast reactor irradiation data. The TFE Verification Program addresses this concern. The general logic and strategy of the program to achieve its objectives is shown on Fig. 1-1. Five prior programs form the basis for the TFE Verification Program: (1) AEC/NASA program of the 1960s and early 1970; (2) SP-100 concept development program;(3) SP-100 thermionic technology program; (4) Thermionic irradiations program in TRIGA in FY-86; (5) and Thermionic Technology Program in 1986 and 1987. 18 refs., 64 figs., 43 tabs.
Date: March 1, 1990
Partner: UNT Libraries Government Documents Department

The Integral Fast Reactor

Description: Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.
Date: January 1, 1990
Creator: Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)) & Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department

A probabilistic approach to the evaluation of the PTS issue

Description: An integrated probabilistic approach for the evaluation of the pressurized-thermal-shock (PTS) issue was developed at the Oak Ridge National Laboratory (ORNL) at the request of the Nuclear Regulatory Commission (NRC). The purpose was to provide a method for identifying dominant plant design and operating features, evaluating possible remedial measures and the validity of the NRC PTS screening criteria, and to provide an additional tool for estimating vessel life expectancy. The approach was to be integrated in the sense that it would include the postulation of transients; estimates of their frequencies of occurrence; systems analyses to obtain the corresponding primary-system pressure, down-comer coolant temperature, and fluid-film heat-transfer coefficient adjacent to the vessel wall; and a probabilistic fracture-mechanics analysis, using the latter data as input. A summation of the products of frequency of transient and conditional probability of failure for all postulated transients provides an estimate of frequency of vessel failure. In the process of developing the integrated pressurized-thermal-shock (IPTS) methodology, three specific plant analyses were conducted. The results indicate that the NRC screening criteria may not be appropriate for all US pressurized water reactor (PWR) plants; that is, for some PWRs, the calculated mean frequency of vessel failure corresponding to the screening criteria may be greater than the maximum permissible value in Regulatory Guide 1.154. A recent view of the ORNL IPTS study, which was completed in 1985, indicates that there are a number of areas in which the methodology can and should be updated, but it is not clear whether the update will increase or decrease the calculated probabilities. 31 refs., 2 tabs.
Date: January 1, 1991
Creator: Cheverton, R.D. & Selby, D.L.
Partner: UNT Libraries Government Documents Department

Benchmark field study of deep neutron penetration

Description: A unique benchmark neutron field has been established at the Lawrence Livermore National Laboratory (LLNL) to study deep penetration neutron transport. At LLNL, a tandem accelerator is used to generate a monoenergetic neutron source that permits investigation of deep neutron penetration under conditions that are virtually ideal to model, namely the transport of mono-energetic neutrons through a single material in a simple geometry. General features of the Lawrence Tandem (LATAN) benchmark field are described with emphasis on neutron source characteristics and room return background. The single material chosen for the first benchmark, LATAN-1, is a steel representative of Light Water Reactor (LWR) Pressure Vessels (PV). Also included is a brief description of the Little Boy replica, a critical reactor assembly designed to mimic the radiation doses from the atomic bomb dropped on Hiroshima, and its us in neutron spectrometry. 18 refs.
Date: June 10, 1991
Creator: Morgan, J.F.; Sale, K. (Lawrence Livermore National Lab., CA (USA)); Gold, R.; Roberts, J.H. & Preston, C.C. (Metrology Control Corp., Richland, WA (USA))
Partner: UNT Libraries Government Documents Department

Severe accident testing of electrical penetration assemblies

Description: This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.
Date: November 1, 1989
Creator: Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))
Partner: UNT Libraries Government Documents Department

SCDAP/RELAP5/MOD2 code manual

Description: This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs.
Date: February 1, 1990
Creator: Hohorst, J. K. (ed.) (EG and G Idaho, Inc., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department

Overview of the Savannah River reactor surveillance program

Description: Preliminary radiation effects studies, now nearing completion, are based upon short-term irradiations in the High Flux Isotope Reactor (HFIR) and University of Buffalo Reactor (UBR). The purpose of these studies is to assess the impact of radiation effects on the performance and service life of Savannah River Site (SRS) reactor components. The SRS reactor surveillance program was simultaneously established in order to validate the results of these short-term irradiations under conditions more applicable to the SRS reactor components. The main conditions of interest include: temperature, moderator chemistry and neutron spectrum. The materials employed in both the surveillance and short-term irradiation programs were taken from the SRS R-reactor process piping and included base, weld, and heat-effected zone metal samples in both the C-L and L-C orientations. Test specimen types include tensile, Charpy V-notch, compact tension, constant extension rate test, and wedge-opening-load. A program schedule has been established for the irradiation and testing of surveillance specimens. The surveillance capsules are currently undergoing irradiation near the center of the SRS K-rector core and will remain in this location until their total fast fluence is equal to the current maximum reactor tank wall exposure. The bulk of the specimens will then be moved to blanket capsules at the edge of the core in order to track the wall exposure. A small portion of the specimens will be removed and tested, and a third group of specimens will be removed from the central core position after they have accumulated a thermal fluence 50% greater than the current maximum tank wall exposure. Specimen groups will be removed from the blanket capsules and tested periodically. Additional capsules may be inserted in order to generate flow fluence or high thermal to fast flux ratio data. 17 refs.
Date: January 1, 1991
Creator: Sindelar, R.L.; Thomas, J.K. & Baumann, N.P.
Partner: UNT Libraries Government Documents Department