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Water chemistry of breeder reactor steam generators. [LMFBR]

Description: The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed.
Date: August 1, 1980
Creator: Simpson, J.L.; Robles, M.N.; Spalaris, C.N. & Moss, S.A.
Partner: UNT Libraries Government Documents Department

Real-time LMFBR steam generator analyzer

Description: A model was developed for steam generator analysis in real-time. This model is based on a movable boundary formulation, the Gear method for the solution of stiff differential equations, and the use of analytic relationships. (DLC)
Date: January 1, 1985
Creator: Tzanos, C.P.
Partner: UNT Libraries Government Documents Department

Concept and preliminary design of double tubesheet connector region for design and steady-state conditions. [LMFBR]

Description: In this analysis the structural integrity of the double tubesheet connector for the LMFBR demonstration plant steam generator is investigated for design and steady state conditions. Although the evaporator and superheater are to be interchangeable, only the steam outlet end of the evaporator is investigated. This was selected because the mean temperature differences between the tubesheets, tubes and connector are largest here at steady state, thus yielding the highest thermal loads. Combined with the thermal loads are superheater pressures in order to also use the highest mechanical loads. Although this is a conservative approach for the evaporator, the superheater must be analyzed separately at a later date in order to assure interchangeability. Evaluation of the results is based upon ASME Code Case 1331-8.
Date: June 1, 1974
Creator: Rinne, W.A.
Partner: UNT Libraries Government Documents Department

Commercial LMFBR steam generator design comparison. Final report for period from 1 October 1977 through 30 September 1978

Description: This report presents results obtained from the commercial LMFBR Steam Generator Design Comparison Study from 1 October 1977 through 30 September 1978 relative to selecting the preferred steam generator design for a commercial-size plant using a Benson, Sulzer, or saturated steam cycle. The primary emphasis was placed on identifying potential problem areas in each design for each steam cycle. The study indicates the hockey stick design as the preferred concept for each steam cycle.
Date: September 30, 1978
Creator: Newburn, F.
Partner: UNT Libraries Government Documents Department

AI reference LMFBR steam-generator development

Description: The Design Data Sheets summarize the key parameters being used in the design and analysis of the AI Prototype LMFBR Steam Generator. These Data Sheets supplement SDD-097-330-002, Steam Generator System, 1450 psi Steam Conditions. This document will serve as the baseline design data control until a GE/RRD approved steam generator specification with ordering data is received.
Date: October 12, 1973
Creator: Anderson, T.L.
Partner: UNT Libraries Government Documents Department

Feasibility of leak-detection instrumentation for duplex-tube steam generator. [LMFBR]

Description: A literature search has been carried out to determine if current state-of-the-art for sodium vapor and water vapor detectors are feasible as leak detection instrumentation for the Westinghouse duplex-tube steam generator. A commercially available probe-type water vapor detector has been identified and a thermal ionization type sodium vapor detector, currently being developed by Westinghouse, has been selected as the reference sodium-vapor leak detector. Recommendations are made concerning the experimental studies required to adapt the selected instrumentation to steam-generator plant applications. Proposed future instrumentation development programs are also identified.
Date: January 1, 1974
Creator: Berkey, E. & Witkowski, R.E.
Partner: UNT Libraries Government Documents Department

Comments on US LMFBR steam generator base technology

Description: The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects.
Date: January 1, 1984
Creator: Simmons, W.R.
Partner: UNT Libraries Government Documents Department

Shutdown and post-test examination plan for the small steam generator model (SSGM)

Description: A detailed working plan to terminate and to examine the Westinghouse duplex tube small steam generator model (SSGM) is defined and presented. Following completion of the current phase (DNB Corrosion Testing in Off-Normal Water Chemistry) of testing, the Advanced Reactors Division GPL-1/SWL-1 Loops will be shut down on both the water side and the sodium side. The SSGM will be cleaned prior to removal of the model from the test silo. During disassembly and following removal of the model to a dissection area at ARD, visual and photographic records of SSGM will be made. Secondary cleaning and initial dissection into seven sections will be conducted with protective oils and plastic bags used to inert the outer shell and duplex tube samples. The major sampling and detailed microscopic examination of the duplex tubing, tube-to-tubesheet welds and tube-support pads will be made following shipment of the dissected model to Tampa. A 72-inch section containing the region over which the DNB interface moved during the SSGM testing will be inerted and shipped to General Electric Company for their examination.
Date: December 1, 1975
Creator: York, J.W. & Sessions, C.E.
Partner: UNT Libraries Government Documents Department

Evaluation of exposure conditions for the water-side corrosion test of a sodium heated steam generator evaporator model employing a duplex tube (2160 hours at critical heat flux - phase III SSGM tests). [LMFBR]

Description: This report describes the specialized corrosion test water steam loop, test procedures, test conditions, and test results. A complete water chemistry and thermal hydraulic performance history is given and evaluated for the Phase III test program. The movement of the dryout location and the heat flux variations in pre- and post-critical heat flux regions are documented and analyzed. On seven occasions during the course of the test program to date, the operating conditions drifted from the CHF reference Phase III operation in the dryout regime into the DNB regime. The corresponding corrosion mechanism experienced differing exposure conditions on these occasions. CHF sensitivity to this apparent drift behavior is evaluated.
Date: December 1, 1975
Creator: Waszink, R.P.; Hwang, J.Y. & Efferding, L.E.
Partner: UNT Libraries Government Documents Department

Transient simulation of a helical-coil sodium/water steam generator

Description: The MINET (Momentum Integral Network) code heat exchanger model was used to analyze transient test data provided by PNC of Japan. Testing of the MINET model is part of a larger effort to facilitate and validate the use of the SSC/MINET code for MONJU plant transient analysis. In MINET, a heat exchanger is modeled using one or more representative tubes, with each tube consisting of the fluid inside the tube, the tube wall, and the fluid outside that is associated with the tube. The heat exchanger tube is divided into one or more axial nodes of equal length. Five time dependent equations are utilized per node, including the wall heat conduction equation and donor-cell differenced conservation of mass and energy equations for the fluids on both sides of the tube. These nodal equations are used to constrain the tube wall temperature, fluid mass flow rates, and fluid enthalpies.
Date: January 1, 1982
Creator: Van Tuyle, G.J. & Iwashita, T.
Partner: UNT Libraries Government Documents Department

Ultrasonic inspection for wastage in the LMFBR steam generator due to sodium--water reactions

Description: As part of a program to study the results of large sodium-water reactions in the LMFBR Steam Generator, a boreside ultrasonic inspection device was developed to measure the wall thickness and diameter of the 2-/sup 1///sub 4/Cr-1 Mo, 0.397 in. I.D. steam tubes. The reaction was created in a near prototype steam generator by guillotine-type rupture of a steam tube, while the generator was at operating conditions. Wastage occurred on the surrounding tubes due to the high temperature reaction. The UT test instrument was designed to operate with a 15 MHz transducer in the pulse-echo shear-wave mode, with a sampling rate of 10/sup 4//sec. System outputs are diameter, wall thickness, attitude and axial position of the transducer. All are displayed digitally and may be recorded. Measurements are fed into a computer for later retrieval, and/or cascaded outputs into an x-y recorded displaying either out-of-limit or thickness data. The UT data taken in this experiment were consistent with physical measurements on a tube which was removed from the generator after the test. A machined flat /sup 1///sub 8/-inch long and 0.002-inch deep could readily be detected.
Date: January 1, 1977
Creator: Neely, H.H. & Renger, L.
Partner: UNT Libraries Government Documents Department

Multidimensional numerical modeling of heat exchangers. [LMFBR]

Description: A comprehensive, multidimensional, thermal-hydraulic model is developed for the analysis of shell-and-tube heat exchangers for liquid-metal services. For the shellside fluid, the conservation equations of mass, momentum, and energy for continuum fluids are modified using the concept of porosity, surface permeability and distributed resistance to account for the blockage effects due to the presence of heat-transfer tubes, flow baffles/shrouds, the support plates, etc. On the tubeside, the heat-transfer tubes are connected in parallel between the inlet and outlet plenums, and tubeside flow distribution is calculated based on the plenum-to-plenum pressure difference being equal for all tubes. It is assumed that the fluid remains single-phase on the shell side and may undergo phase-change on the tube side, thereby simulating the conditions of Liquid Metal Fast Breeder Reactor (LMFBR) intermediate heat exchangers (IHX) and steam generators (SG).
Date: January 1, 1982
Creator: Sha, W.T.; Yang, C.I.; Kao, T.T. & Cho, S.M.
Partner: UNT Libraries Government Documents Department

Small-leak behavior: summary report on scoping tests. [LMFBR]

Description: A scoping phase has been completed on a study of small steam leaks in Liquid Metal Fast Breeder Reactor steam generators. Methods were developed to fabricate leaks in the size range of 10/sup -10/ to 10/sup -5/ lb/sec H/sub 2/O in 2 1/4 Cr-1 Mo steam generator tubing of prototypic wall thickness (approx. 0.110 inch). A test system was modified to supply flowing superheated steam at 1525 psig to as many as three specimens immersed in flowing sodium. Test results are presented.
Date: June 1, 1975
Creator: Sandusky, D.W.
Partner: UNT Libraries Government Documents Department

Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

Description: An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.
Date: April 1, 1977
Creator: Wolf, S. & Holmes, D.H.
Partner: UNT Libraries Government Documents Department

Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

Description: The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM.
Date: March 11, 1981
Creator: Impellezzeri, J.R.; Camaret, T.L. & Friske, W.H.
Partner: UNT Libraries Government Documents Department

Report on steam-generator concept development. [LMFBR]

Description: This report presents a summary of the conceptual design work on commercial steam generators by the General Electric Company for the ERDA Steam Generator Development Program. Three conceptual steam generator designs are included and discussed in some detail; the Hockey Stick, the J-Module and the Straigth Tube with Expansion Bellows, all referenced to a 1500 MWe 4 loop commercial LMFBR design. Single and double tube concept development is summarized and the effect of each upon steam generator reliability and plant availability is discussed. Double and single tubesheet concepts are discussed and possible ways of introducing leak detection systems for increased reliability are presented. A conceptual design of a Hockey Stick evaporator, for use in a 1200 MWe LMFBR, using double tubes and double tubesheets with a leak detection system is presented and discussed.
Date: June 1, 1975
Creator: Scott, W.B. (comp.)
Partner: UNT Libraries Government Documents Department

Field radiography using 100 Ci of Co-60 without interrupting adjacent manufacturing operations

Description: Radiography is the primary method of Nondestructive Examination recognized by the ASME B and PV Code as providing objective evidence of volumetric examination of the pressure boundary welds that are present in the Clinch River Breeder Reactor Program (CRBRP) Steam Generator. In order to support the steam generator production schedule, the radiographic examinations must be performed without interrupting any other manufacturing or inspection operations taking place within a 20-ft radius from the source. This condition imposes rigorous radiation safety requirements since the gamma radiation sources chosen to be used for examination of the pressure boundary welds, Co-60 and Ir-192, are very energetic. Co-60 gamma ray energy is 1.17 and 1.33 MeV and Ir-192 gamma ray is .6 MeV. The hazard of using such a high energy sources in the immediate vicinity of working personnel has necessitated the need for a thorough evaluation of methods of protection. Personnel protection from penetrating radiation, both x-ray and gamma ray, is accomplished by ingenious use, singly or in combination, of two factors which reduce radiation intensity. These factors are distance and shielding. In all radiographic operations the primary consideration is for personnel safety. The maximum radiation dose rate limit will be 0.002 rem/hr. This document describes how the factors of distance and shielding have been used to assure personnel safety at all times.
Date: January 1, 1979
Creator: Donnelly, C.W.
Partner: UNT Libraries Government Documents Department

Test and evaluation of Alco/BLH prototype sodium-heated steam generator. Final report

Description: A 30-Mwt prototype sodium-to-sodium intermediate heat exchanger and a 30-Mwt prototype sodium-heated steam generator were tested in combined operation in its Sodium Components Test Installation. This report contains the results of test and evaluation of the steam generator. During plant performance tests, performance degradation was observed, which resulted in the initiation of a diagnostic test series. This test series revealed that under certain operating conditions, the thermohydraulic characteristic of the steam generator changed either suddenly or gradually, resulting in overall performance degradation. A structural failure, requiring retirement of the unit, occurred before the diagnostic test series and analytical support effort were completed. This report describes the thermohydraulic and structural performance, including the structural failures, and related evaluation analyses of the Alco/BLH prototype steam generator performed prior to termination of the test and evaluation program. In addition, the report presents a post-test examination plan to obtain data that could possibly explain the cause of performance anomalies and structural failures experienced during testing.
Date: January 31, 1971
Creator: Kaplan, C.J.; Auge, L.J.; Cho, S.M.; Hanna, R.W.; Prevost, J.R.; Steger, N.A. et al.
Partner: UNT Libraries Government Documents Department

Test of the second model of the GCFR steam generator cavity closure plug

Description: As part of its participation in the program to develop a 300 MW(e) gas-cooled fast reactor (GCFR) demonstration power plant, the Oak Ridge National Laboratory (ORNL) is conducting a series of structural model tests. One of the tasks of the ORNL effort provides for the design, analysis and/or testing of small-scale models of the PCRV penetration closure plugs. The principal objectives are to determine structural response of the plugs to normal operating pressures and to substantial overpressurization. The tests also provide reliable data for verification of analysis methods. Included in the testing program are models of both the steam generator cavity closure and central core cavity closure plugs. Thus far, two models of the steam generator cavity closure series have been tested. The first model which represented the original GA design was pressurized to 75.8 MPa (11,000 psi) without sustaining cracking of the concrete or failure of the plug although yielding of the metallic components was recorded by attached strain gages.
Date: January 1, 1979
Creator: Callahan, J.P.; Robinson, G.C. & Dodge, W.G.
Partner: UNT Libraries Government Documents Department

Prototype steam generator test at SCTI/ETEC. Acoustic program test plan. [LMFBR]

Description: This document is an integrated test plan covering programs at General Electric (ARSD), Rockwell International (RI) and Argonne National Laboratory (CT). It provides an overview of the acoustic leak detection test program which will be completed in conjunction with the prototype LMFBR steam generator at the Energy Technology Engineering Laboratory. The steam generator is installed in the Sodium Components Test Installation (SCTI). Two acoustic detection systems will be used during the test program, a low frequency system developed by GE-ARSD (GAAD system) and a high frequency system developed by RI-AI (HALD system). These systems will be used to acquire data on background noise during the thermal-hydraulic test program. Injection devices were installed during fabrication of the prototype steam generator to provide localized noise sources in the active region of the tube bundle. These injectors will be operated during the steam generator test program, and it will be shown that they are detected by the acoustic systems.
Date: October 1, 1981
Creator: Greene, D.A.; Thiele, A. & Claytor, T.N.
Partner: UNT Libraries Government Documents Department

Model of a once-through steam generator with moving boundaries and a variable number of nodes

Description: A model of a once-through steam generator (OTSG) with moving boundaries and variable nodes was developed for liquid-metal fast breeder reactor (LMFBR) applications. The main advantage of the current model is its ability to generate a variable number of nodes. The OTSG model consists of four regions identified as subcooled, nucleate-boiling, film-boiling, and superheated. The number of nodes within each region is variable, which allows the user to change the number of nodes and assess this effect on the resulting solution. The model also permits the elimination of regions (during a cool-down transient) or the formation of regions (start-up, or reheat after a cool-down transient). The model can also be used to simulate a boiler by eliminating the film-boiling and superheated regions. The model permits the user to select the appropriate heat-transfer correlations or use the default selections. The model has a separate section where steady-state values are determined before the code enters the transient section. The paper describes the formation of the model, and the steady-state and transient solution methods. Typical results are presented.
Date: January 1, 1983
Creator: Berry, G.
Partner: UNT Libraries Government Documents Department

Use of a steam leak simulator in EBR-II

Description: A steam leak simulator has been installed on EBR-II to periodically test and calibrate the steam-generator leak detection system. Measured amounts of molten anhydrous sodium hydroxide are injected at controlled rates simulating leaks in the range of 0.024 to 0.16 g H/sub 2/O/s. Experience with 11 injections over an 18 month period is described.
Date: January 1, 1984
Creator: McKee, J.M. & Osterhout, M.M. Batten, R.L.
Partner: UNT Libraries Government Documents Department