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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

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

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

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

Intermediate leak protection/automatic shutdown for B and W helical coil steam generator

Description: The report summarizes a follow-on study to the multi-tiered Intermediate Leak/Automatic Shutdown System report. It makes the automatic shutdown system specific to the Babcock and Wilcox (B and W) helical coil steam generator and to the Large Development LMFBR Plant. Threshold leak criteria specific to this steam generator design are developed, and performance predictions are presented for a multi-tier intermediate leak, automatic shutdown system applied to this unit. Preliminary performance predictions for application to the helical coil steam generator were given in the referenced report; for the most part, these predictions have been confirmed. The importance of including a cover gas hydrogen meter in this unit is demonstrated by calculation of a response time one-fifth that of an in-sodium meter at hot standby and refueling conditions.
Date: January 1, 1981
Partner: UNT Libraries Government Documents Department

CRBRP modular steam generator tube-to-tubesheet and shell-closure welding

Description: The original Modular Steam Generator (MSG), whiand inh was designed, built, and tested by the Energy Systems Group (ESG) of Rockwell International, was a departure from conventional boilers or heat exchangers. The design was a hockeystick concept - the upper section of the generator is curved 90/sup 0/. Factors affecting operating parameters were considered and incorporated in the original MSG design. The MSG was fully instrumented and functionally tested at the Energy Technology Engineering Center at Rockwell. The MSG steamed continuously for over 4000 h, and at the conclusion of the 9000-h test cycle, it was systematically dismantled and examined for wear to critical components. This paper explains the solutions to several manufacturing challenges presented by the unique design of the MSG.
Date: January 1, 1982
Creator: Viri, D.P.
Partner: UNT Libraries Government Documents Department

CRBRP steam-generator design evolution

Description: The overall design of the CRBRP Steam Generator is briefly discussed. Two areas of particular concern are highlighted and considerations leading to the final design are detailed. Differential thermal expansion between the shell and the steam tubes is accommodated by the tubes flexing in the curved section of the shell. Support of the tubes by the internals structure is essential to permit free movement and minimize tube wear. Special spacer plate attachment and tube hole geometry promote unimpeded axial movement of the tubes by allowing individual tubes to rotate laterally and by providing lateral movement of the spacer plates relative to the adjacent support structure. The water/steam heads of the CRBRP Steam Generator are spherical heads welded to the lower and upper tubesheets. They were chosen principally because they provide a positively sealed system and result in more favorable stresses in the tubesheets when compared to mechanically attached steamheads.
Date: January 1, 1983
Creator: Geiger, W.R.; Gillett, J.E. & Lagally, H.O.
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) is part of an extensive testing program being 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. 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 are described.
Date: September 13, 1979
Creator: Impellezzeri, J.R. & Camaret, T.L.
Partner: UNT Libraries Government Documents Department

Characterization of ESR and VAR 2-1/4Cr--1Mo alloy tubing. [LMFBR]

Description: Commercial tubing for LMFBR sodium-water steam generator applications was produced from ESR and VAR 2/sup 1///sub 4/Cr--1Mo alloy (Grade T22) and characterized regarding chemical composition, microstructure, physical characteristics, and short and long term mechanical properties. These results demonstrate that tubing meeting RDT Standard RDT M3-33 can be produced from either the ESR or VAR process and is acceptable to ASME Code Section III requirements. Metallurgical and mechanical properties are similar for both ESR and VAR material, indicating that either remelting practice is acceptable.
Date: September 1, 1976
Creator: Bubsoom, H.J.
Partner: UNT Libraries Government Documents Department

Analysis of the U. S. large leak test program Series I sodium--water reaction test results. [LMFBR]

Description: A computer code for modeling Liquid Metal Fast Breeder Reactor (LMFBR) Steam Generator (SG) leaks has been partially validated. This was accomplished by comparing code results with data from large leak Sodium/Water Reaction (SWR) tests conducted at the Energy Technology Engineering Center (ETEC). In each of six tests, a Double-Ended Guillotine (DEG) rupture of an SG tube was produced in a sodium-filled SG at typical operating conditions. The primary purpose of these tests was to provide data for validation of the TRANSWRAP (Transient Sodium Water Reaction Analysis Program) code. Primary areas of concern are the leaksite pressure history, pressures at various locations throughout the system, and expulsion of reaction products through the relief system. Agreement with TRANSWRAP results was found to be remarkably good.
Date: January 1, 1979
Creator: Sane, J.O.; Regimbal, J.J.; Fairbairn, J.A. & Meyer, R.A.
Partner: UNT Libraries Government Documents Department

Hockey-stick steam generator for LMFBR

Description: This paper presents the criteria and evaluation leading to the selection of the Hockey Stick Steam Generator Concept and subsequent development of that concept for LMFBR application. The selection process and development of the Modular Steam Generator (MSG) is discussed, including the extensive test programs that culminated in the manufacture and test of a 35 MW(t) Steam Generator. The design of the CRBRP Steam Generator is described, emphasizing the current status and a review of the critical structural areas. CRBRP steam generator development tests are evaluated, with a discussion of test objectives and rating of the usefulness of test results to the CRBRP prototype design. Manufacturing experience and status of the CRBRP prototype and plant units is covered. The scaleup of the Hockey Stick concept to large commercial plant application is presented, with an evaluation of scaleup limitations, transient effects, and system design implications.
Date: January 1, 1981
Creator: Hallinan, G.J. & Svedlund, P.E.
Partner: UNT Libraries Government Documents Department

History of the water chemistry for the few tube test model

Description: The water chemistry activities carried out in support of the Few Tube Test are described. This test was conducted to provide design confirmation data for the Clinch River Breeder Reactor Project (CRBRP) steam generators. Proposed CRBRP chemistry was followed; all volatile treatment (AVT) of water was carried out with on-line monitoring capability.
Date: September 1, 1979
Creator: Moss, S.A. & Simpson, J.L.
Partner: UNT Libraries Government Documents Department

Vibration tests of a full scale water model of a Clinch River steam generator module

Description: Cross and parallel flow induced vibrations were measured in a shortened full scale model of the CRBRP steam generators using water as the working fluid. M easurements were made on five selected tubes, some of which fit through oversize tube support holes. Power spectral density plots were made of vibration parameters at selected locations. Maximum displacement was about 3 mils and occurred at the lower frequencies. The maximum vibrational stress was calculated to be 2000 psi. Measurements of vortex shedding frequencies showed that vortices are not a dominant source of tube excitation. In general, no adverse effects on the CRPRP steam generators were identified which affect service life or structural integrity.
Date: unknown
Creator: Durand, R.E.; Gabler, M.J.; Weinberg, L. & Yang, T.M.
Partner: UNT Libraries Government Documents Department

Air-water tests in support of LLTR series II Test A-4. [Large Leak Test Rig]

Description: A series of tests injecting air into a tank of stagnant water was conducted in June 1980 utilizing the GE Plenum Mixing Test Facility in San Jose, California. The test was concerned with investigating the behavior of air jets at a submerged orifice in water over a wide range of flow rates. The main objective was to improve the basic understanding of gas-liquid phenomena (e.g., leak dynamics, gas bubble agglomeration, etc.) in a simulated tube bundle through visualization. The experimental results from these air-water tests will be used as a guide to help select the leak size for LLTR Series II Test A-4 because air-water system is a good simulation of water-sodium mixture.
Date: July 1, 1980
Creator: Chen, K.
Partner: UNT Libraries Government Documents Department

Steam ingress reactivity effects in the GCFR

Description: Because of the higher pressure on the steam side of the steam generator in the gas-cooled fast breeder reactor (GCFR), the potential exists for steam-generator tube leaks or ruptures which might introduce water vapor into the primary coolant and thence into the core region. This potential for core steam ingress requires that a satisfactory understanding of the reactivity effects due to various concentrations of hydrogeneous material in the core be achieved. The reactivity effects resulting from an increase in hydrogen density are two-fold: a decrease in neutron leakage from the core, and a moderation of the neutron energy spectrum. The leakage effect is always positive and is predominant in smaller systems with attendant higher leakage fractions. Also, the cold-to-hot operating core temperature transition produces a negative reactivity effect (to a large extent due to the Doppler effect on absorption in U-238), which is substantially increased by progressive softening of the spectrum due to hydrogen addition. Research programs studying these reactivity effects are described.
Date: July 1, 1979
Creator: Hess, A.L. & Hamilton, C.J.
Partner: UNT Libraries Government Documents Department

Interim report on the feasibility study of acoustic leak location for WTD steam generators. [LMFBR]

Description: This interim report summarizes the current status of a feasibility study of an acoustic leak location system for two Westinghouse-Tampa Division (WTD) double-wall steam generators. The primary WTD design requirement is that a leak of helium gas into sodium be located to within a cluster of seven tubes at hot standby conditions. The report documents the results of an extensive analytical assessment, outlines areas of concern resulting from the analyses which will require experimental validation, and presents results of experiments thus far completed. The conclusion of the study at this time is that acoustic leak location is feasible, pending the results of the remaining experimental tests. The acoustic leak location system operates on the same principles as the acoustic leak detection system being developed under SG027, Subtask X2. Analyses indicate that, in general, the location system will use similar design parameters to the acoustic leak detection system. Much of the hardware can be indentical, in particular the accelerometers. It is recommended that a minicomputer/software approach be employed rather than the custom-designed hardware approach used for the detection system.
Date: March 1, 1979
Creator: Greene, D.A. & McMurtrie, K.A.
Partner: UNT Libraries Government Documents Department

Dynamic stability experimental/analytical program results on a multiple tube sodium heated steam generator model employing double wall tubes. [LMFBR]

Description: Experimental results are described of a test program which induced water/steam dynamic instability in the multiple tube circuits of a sodium heated, double-wall tube, steam generator model whose water/steam side is designed to operate under once-through conditions with exit superheat (457/sup 0/C/855/sup 0/F; 15.8 MPa/2290 psia). The instability thresholds were determined experimentally with off-nominal operation in order to establish the stability margin that exists with and without the damping orifice configuration employed at the inlet of each tube circuit. Two classes of experiments were performed; one with low magnitudes of water/steam mass velocity resulting in high exit superheat (Benson Cycle), and the other with high magnitudes of mass velocity and no bulk exit steam superheat (Sulzer Cycle). Analytical prediction of the instability inception conditions were compared with experiment and are included along with an evaluation of the applicability of the analysis method to plant scale stability evaluations.
Date: January 1, 1983
Creator: Efferding, L.E.
Partner: UNT Libraries Government Documents Department

Preliminary thermal/hydraulic sizing calculations for duplex tube evaporator/superheater (interchangeable units). Revision 1

Description: This is a preliminry thermal/hydraulic report reflecting work under Subtask 6.2 of Ref. 1.1. This report is an extension of the previous thermal/hydraulic design report. Parts of this report have been transmitted to GE. The detailed design basis, listed by source, is given. Additional details are discussed.
Date: June 1, 1974
Creator: Waszink, R.P.; Hwang, J.Y. & Efferding, L.E.
Partner: UNT Libraries Government Documents Department

Simulation of fluid thermal fluctuations in the CRBRP steam generator using model testing

Description: Testing has been performed using two 1/6 scale models of the Clinch River Breeder Reactor Plant Steam Generators to simulate fluid temperature fluctuations in the sodium. Water was used as the working fluid and the models were fabricated from Plexiglas to facilitate flow visualization. Hydraulic scaling was achieved through Richardson Number and Euler Number similarity. Fluid temperature fluctuations were recorded on movable thermocouples in both the upper and lower semi-stagnant regions of the Steam Generator Module. The magnitude of the temperature fluctuations (peak to peak) were as high as 95% of the maximum potential and were used to assess the amount of fatigue damage of the steam tubes.
Date: April 1, 1981
Creator: Garner, D.C. & Novendstern, E.H.
Partner: UNT Libraries Government Documents Department