WorldWideScience

Sample records for sassys lmfbr systems

  1. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  2. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  3. Damping in LMFBR pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

    1983-06-01

    LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

  4. A new balance-of-plant model for the SASSYS-1 LMR systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1991-01-01

    In this paper, a balance-of-plant model is developed for the SASSYS code. This model represents the balance of plant as a network of components. It interfaces with the existing SASSYS code through the water side of the steam generator. The network representation provides a discretization of the mass, momentum, and energy equations and the equation of state and allows a simultaneous solution for the changes in pressure, flow, and enthalpy throughout the waterside system. The model has been tested for several types of transients and been found to perform both accurately and efficiently

  5. A user's guide to the SASSYS-1 control system modeling capability

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1987-06-01

    This report describes a control system modeling capability that has been developed for the analysis of control schemes for advanced liquid metal reactors. The general class of control equations that can be represented using the modeling capability is identified, and the numerical algorithms used to solve these equations are described. The modeling capability has been implemented in the SASSYS-1 systems analysis code. A description of the card input, a sample input deck and some guidelines for running the code are given

  6. Coupling the System Analysis Module with SAS4A/SASSYS-1

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-30

    SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced reactors, with an emphasis on sodium fast reactors. SAS4A/SASSYS-1 has been under development and in active use for nearly forty-five years, and is currently maintained by the U.S. Department of Energy under the Office of Advanced Reactor Technology. Although SAS4A/SASSYS-1 contains a very capable primary and intermediate system modeling component, PRIMAR-4, it also has some shortcomings: outdated data management and code structure makes extension of the PRIMAR-4 module somewhat difficult. The user input format for PRIMAR-4 also limits the number of volumes and segments that can be used to describe a given system. The System Analysis Module (SAM) is a fairly new code development effort being carried out under the U.S. DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM is being developed with advanced physical models, numerical methods, and software engineering practices; however, it is currently somewhat limited in the system components and phenomena that can be represented. For example, component models for electromagnetic pumps and multi-layer stratified volumes have not yet been developed. Nor is there support for a balance of plant model. Similarly, system-level phenomena such as control-rod driveline expansion and vessel elongation are not represented. This report documents fiscal year 2016 work that was carried out to couple the transient safety analysis capabilities of SAS4A/SASSYS-1 with the system modeling capabilities of SAM under the joint support of the ART and NEAMS programs. The coupling effort was successful and is demonstrated by evaluating an unprotected loss of flow transient for the Advanced Burner Test Reactor (ABTR) design. There are differences between the stand-alone SAS4A/SASSYS-1 simulations and the coupled SAS/SAM simulations, but these are mainly

  7. A new balance-of-plant model for the SASSYS-1 LMR systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1989-01-01

    A balance-of-plant model has been added to the SASSYS-1 liquid metal reactor systems analysis code. Until this addition, the only waterside component which SASSYS-1 could explicitly model was the water side of a steam generator, with the remainder of the water side represented by boundary conditions on the steam generator. The balance-of-plant model is based on the model used for the sodium side of the plant. It will handle subcooled liquid water, superheated steam, and saturated two-phase fluid. With the exception of heated flow paths in heaters, the model assumes adiabatic conditions along flow paths; this assumption simplifies the solution procedure while introducing very little error for a wide range of reactor plant problems. Only adiabatic flow is discussed in this report. 3 refs., 4 figs

  8. The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-01-01

    The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

  9. Dynamic simulation of LMFBR systems

    International Nuclear Information System (INIS)

    Agrawal, A.K.; Khatib-Rahbar, M.

    1980-01-01

    This review article focuses on the dynamic analysis of liquid-metal-cooled fast breeder reactor systems in the context of protected transients. Following a brief discussion on various design and simulation approaches, a critical review of various models for in-reactor components, intermediate heat exchangers, heat transport systems and the steam generating system is presented. A brief discussion on choice of fuels as well as core and blanket system designs is also included. Numerical considerations for obtaining system-wide steady-state and transient solutions are discussed, and examples of various system transients are presented. Another area of major interest is verification of phenomenological models. Various steps involved in the code and model verification are briefly outlined. The review concludes by posing some further areas of interest in fast reactor dynamics and safety. (author)

  10. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  11. Verification and implications of the multiple pin treatment in the SASSYS-1 LMR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1994-01-01

    As part of a program to obtain realistic, as opposed to excessively conservative, analysis of reactor transients, a multiple pin treatment for the analysis of intra-subassembly thermal hydraulics has been included in the SASSYS-1 liquid metal reactor systems analysis code. This new treatment has made possible a whole new level of verification for the code. The code can now predict the steady-state and transient responses of individual thermocouples within instrumented subassemlies in a reactor, rather than just predicting average temperatures for a subassembly. Very good agreement has been achieved between code predictions and the experimental measurements of steady-state and transient temperatures and flow rates in the Shutdown Heat Removal Tests in the EBR-II Reactor. Detailed multiple pin calculations for blanket subassemblies in the EBR-II reactor demonstrate that the actual steady-state and transient peak temperatures in these subassemblies are significantly lower than those that would be calculated by simpler models

  12. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  13. Computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

    1977-01-01

    Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

  14. Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F.

    1992-09-01

    This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

  15. Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F.

    1992-01-01

    This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

  16. A new balance-of-plant model for the SASSYS-1 LMR [liquid metal reactor] systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1989-01-01

    A balance-of-plant (BOP) model has been developed for use within the SASSYS-1 liquid-metal reactor systems analysis code. This model expands the scope of SASSYS-1 so that the code can explicitly model the waterside components of a nuclear power plant; previously, only the water side of the steam generators could be modeled, with the remainder of the water side represented by boundary conditions on the steam generator. The model represents the BOP a set of flow paths and path junctions; the mass and energy equations are solved at the junctions, and the momentum equation is solved along the flow paths. The junctions are thus mass and energy cells, and the paths are momentum cells. The various waterside component models (pumps, valves, etc.) are specialized types of energy or momentum cells, as appropriate. The solution scheme implicitly couples the energy cells through the momentum cells and solves simultaneously for pressures and enthalpies within the energy cells and for flows within the momentum cells

  17. Confirmatory simulation of safety and operational transients in LMFBR systems

    International Nuclear Information System (INIS)

    Guppy, J.G.; Agrawal, A.K.

    1978-01-01

    Operational and safety transients that may originate anywhere in an LMFBR system must be adequately simulated to assist in safety evaluation and plant design efforts. This paper describes an advanced thermohydraulic transient code, the Super System Code (SSC), that may be used for confirmatory safety evaluations of plant wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions, and presents results obtained with SSC illustrating the degree of modelling detail present in the code as well as the computing efficiency. (author)

  18. Confirmatory simulation of safety and operational transients in LMFBR systems

    International Nuclear Information System (INIS)

    Guppy, J.G.; Agrawal, A.K.

    1978-01-01

    Operational and safety transients (anticipated, unlikely, or extremely unlikely) that may originate anywhere in a liquid-metal fast breeder reactor (LMFBR) system must be adequately simulated to assist in safety evaluation and plant design efforts. An advanced thermohydraulic transient code, the Super System Code (SSC), is described that may be used for confirmatory safety evaluations of plant-wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions. Results obtained with SSC illustrating the degree of modeling detail present in the code as well as the computing efficiency are presented. A version of the SSC code, SSC-L, applicable to any loop-type LMFBR design, has been developed at Brookhaven. The scope of SSC-L is to enable the simulation of all plant-wide transients covered by Plant Protection System (PPS) action, including sodium pipe rupture and coastdown to natural circulation conditions. The computations are stopped when loss of core integrity (i.e., clad melting temperature exceeded) is indicated

  19. Emergency air cleaning system development for LMFBR containments

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.; Muhlestein, L.D.

    1975-01-01

    Criteria for evaluating the various types of Emergency Air Cleaning Systems which may be used in LMFBR plants have been established for both single containment and containment-confinement arrangements. These two plant arrangements have quite different air cleaning requirements for postulated design base accident conditions. Work is currently in progress to select from a list of candidate air cleaning systems those which best meet the criteria requirements. By means of a weighted rating system, areas of strength or weakness can be found and the conceptual system design then optimized. The final system arrangements will be ranked and several of the most promising systems selected for large-scale tests in the former CSE vessel at Hanford. 8 references. (U.S.)

  20. LMFBR self-activated shutdown systems

    International Nuclear Information System (INIS)

    Sowa, E.S.; Barthold, W.P.; Eggen, D.T.; Huebotter, P.R.; Josephson, J.; Pizzica, P.A.; Turski, R.B.; van Erp, J.B.

    1976-01-01

    Self-actuated shutdown systems (SASSs), fully contained within the dimensions of a fuel subassembly and installed in the core in judiciously chosen locations, can provide an important additional safety feature for LMFBRs. If actuated by phenomena inherent to the system and its immediate environment, these systems can contribute considerably to the total reliability of the overall plant protection system, in particular as regards protection against human error. It was shown that this type of shutdown system is capable of inserting a substantial amount of negative reactivity into the core with a relatively small impact on plant performance. Furthermore, it was shown that a coolable geometry can be maintained in LMFBRs of current design for a wide spectrum of accident initiators, and for a range of response times and insertion rates which appear to be achievable within practical design limits. Experiments showed that Curie-point-operated devices have considerable promise for application in self-actuated shutdown systems, in particular as regards meeting the requirements of testability and resettability

  1. Development of a graphical user interface allowing use of the SASSYS LMR systems analysis code as an EBR-II interactive simulator

    International Nuclear Information System (INIS)

    Garner, P.L.; Briggs, L.L.; Gross, K.C.; Ku, J.Y.; Staffon, J.D.

    1994-01-01

    The SASSYS computer program for safety analyses of liquid-metal- cooled fast reactors has been adapted for use as the simulation engine under the graphical user interface provided by the GRAFUN and HIST programs and the Data Views software package under the X Window System on UNIX-based computer workstations to provide a high fidelity, real-time, interactive simulator of the Experimental Breeder Reactor Number II (EBR-II) plant. In addition to providing analysts with an interactive way of performing safety case studies, the simulator can be used to investigate new control room technologies and to supplement current operator training

  2. Refractory metal carbide coatings for LMFBR applications: a systems approach

    International Nuclear Information System (INIS)

    Gotschall, H.L.; Ople, F.S.; Riccardella, P.C.

    1975-01-01

    The selection, testing and improvement of high density, tightly bonded plasma and detonation gun coatings designed to meet LMFBR core component criteria are described. The process descriptions include a review of the important developments in substrate surface preparation which were required to ensure strong bonding and to minimize interface contamination. Coating finishing techniques which were developed to optimize friction behavior are also described

  3. LARA: Expert system for acoustic localization of robot in a LMFBR

    International Nuclear Information System (INIS)

    Lhuillier, C.; Malvache, P.

    1986-12-01

    The expert system LARA (Acoustic Localization of Autonomic Robot) has been developed to show the interest of introducing artificial intelligency for fine automatic positioning of refuelling machine in a LMFBR reactor. LARA which is equipped with an acoustic detector gives rapidly a good positioning on the fuel [fr

  4. Evaluation of air cleaning system concepts for emergency use in LMFBR plants

    International Nuclear Information System (INIS)

    Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

    1976-12-01

    Nineteen different air cleaning concepts are arranged into twenty-four systems and evaluated for use as accident mitigating systems in LMFBR plants. Both single, low-leakage containment plants and once-through operation applicable to containment/confinement plants are considered. Plant characteristics affecting air cleaning requirements are defined for 1000 MW(e) plants and a sodium and radiological release term is postulated. The accident conditions under which the emergency air cleaning system (EACS) must function is established by use of SOFIRE-II and HAA-3B computer codes. Criteria are developed for evaluating the various systems and for assigning comparative ratings. The numerical ratings are combined with information on cost and development potential to arrive at recommendations for the most promising systems. The conclusion is made that reliable and effective systems are feasible for use as engineered safety features for LMFBR plants, but that development effort is required for all the air cleaning concepts evaluated

  5. Comprehensive method of common-mode failure analysis for LMFBR safety systems

    International Nuclear Information System (INIS)

    Unione, A.J.; Ritzman, R.L.; Erdmann, R.C.

    1976-01-01

    A technique is demonstrated which allows the systematic treatment of common-mode failures of safety system performance. The technique uses log analysis in the form of fault and success trees to qualitatively assess the sources of common-mode failure and quantitatively estimate the contribution to the overall risk of system failure. The analysis is applied to the secondary control rod system of an early sized LMFBR

  6. LMFBR system-wide transient analysis: the state of the art and US validation needs

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.

    1982-01-01

    This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants

  7. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

    Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

  8. System seismic analysis of an innovative primary system for a large pool type LMFBR plant

    International Nuclear Information System (INIS)

    Pan, Y.C.; Wu, T.S.; Cha, B.K.; Burelbach, J.; Seidensticker, R.

    1984-01-01

    The system seismic analysis of an innovative primary system for a large pool type liquid metal fast breeder reactor (LMFBR) plant is presented. In this primary system, the reactor core is supported in a way which differs significantly from that used in previous designs. The analytical model developed for this study is a three-dimensional finite element model including one-half of the primary system cut along the plane of symmetry. The model includes the deck and deck mounted components,the reactor vessel, the core support structure, the core barrel, the radial neutron shield, the redan, and the conical support skirt. The sodium contained in the primary system is treated as a lumped mass appropriately distributed among various components. The significant seismic behavior as well as the advantages of this primary system design are discussed in detail

  9. Transference of advanced LMFBR control technology to the aerospace power system program

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Much recent R and D has been devoted to the safety of liquid metal fast breeder reactors (LMFBR's). Part of the resulting technology, especially advanced control systems, appears to be directly transferable to the space nuclear power program. Some of the ideas described herein have been already culminated in successful products that are available for application, e.g. analytical redundancy and fault-tolerant computers. Others, in various stages of R and D, are being developed as elements to support the design goals outlined in the following section, e.g. automated software verification, automated hardware verification, and system validation

  10. Component design for LMFBR's

    International Nuclear Information System (INIS)

    Fillnow, R.H.; France, L.L.; Zerinvary, M.C.; Fox, R.O.

    1975-01-01

    Just as FFTF has prototype components to confirm their design, FFTF is serving as a prototype for the design of the commercial LMFBR's. Design and manufacture of critical components for the FFTF system have been accomplished primarily using vendors with little or no previous experience in supplying components for high temperature sodium systems. The exposure of these suppliers, and through them a multitude of subcontractors, to the requirements of this program has been a necessary and significant step in preparing American industry for the task of supplying the large mechanical components required for commercial LMFBR's

  11. Sodium-NaK engineering handbook. Volume III. Sodium systems, safety, handling, and instrumentation. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Foust, O J [ed.

    1978-01-01

    The handbook is intended for use by present and future designers in the Liquid Metals Fast Breeder Reactor (LMFBR) Program and by the engineering and scientific community performing other type investigation and exprimentation requiring high-temperature sodium and NaK technology. The arrangement of subject matter progresses from a technological discussion of sodium and sodium--potassium alloy (NaK) to discussions of varius categories and uses of hardware in sodium and NaK systems. Emphasis is placed on sodium and NaK as heat-transport media. Sufficient detail is included for basic understanding of sodium and NaK technology and of technical aspects of sodium and NaK components and instrument systems. Information presented is considered adequate for use in feasibility studies and conceptual design, sizing components and systems, developing preliminary component and system descriptions, identifying technological limitations and problem areas, and defining basic constraints and parameters.

  12. Test system to simulate transient overpower LMFBR cladding failure

    International Nuclear Information System (INIS)

    Barrus, H.G.; Feigenbutz, L.V.

    1981-01-01

    One of the HEDL programs has the objective to experimentally characterize fuel pin cladding failure due to cladding rupture or ripping. A new test system has been developed which simulates a transient mechanically-loaded fuel pin failure. In this new system the mechanical load is prototypic of a fuel pellet rapidly expanding against the cladding due to various causes such as fuel thermal expansion, fuel melting, and fuel swelling. This new test system is called the Fuel Cladding Mechanical Interaction Mandrel Loading Test (FCMI/MLT). The FCMI/MLT test system and the method used to rupture cladding specimens very rapidly to simulate a transient event are described. Also described is the automatic data acquisition and control system which is required to control the startup, operation and shutdown of the very fast tests, and needed to acquire and store large quantities of data in a short time

  13. Large-scale tests of aqueous scrubber systems for LMFBR vented containment

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.

    1980-01-01

    Six large-scale air cleaning tests performed in the Containment Systems Test Facility (CSTF) are described. The test conditions simulated those postulated for hypothetical accidents in an LMFBR involving containment venting to control hydrogen concentration and containment overpressure. Sodium aerosols were generated by continously spraying sodium into air and adding steam and/or carbon dioxide to create the desired Na 2 O 2 , Na 2 CO 3 or NaOH aerosol. Two air cleaning systems were tested: (a) spray quench chamber, educator venturi scrubber and high efficiency fibrous scrubber in series; and (b) the same except with the spray quench chamber eliminated. The gas flow rates ranged up to 0.8 m 3 /s (1700 acfm) at temperatures to 313 0 C (600 0 F). Quantities of aerosol removed from the gas stream ranged up to 700 kg per test. The systems performed very satisfactorily with overall aerosol mass removal efficiencies exceeding 99.9% in each test

  14. Collection and evaluation of salt mixing data with the real time data acquisition system. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Glazer, S.; Chiu, C.; Todreas, N.E.

    1977-09-01

    A minicomputer based real time data acquisition system was designed and built to facilitate data collection during salt mixing tests in mock ups of LMFBR rod bundles. The system represents an expansion of data collection capabilities over previous equipment. It performs steady state and transient monitoring and recording of up to 512 individual electrical resistance probes. Extensive real time software was written to govern all phases of the data collection procedure, including probe definition, probe calibration, salt mixing test data acquisition and storage, and data editing. Offline software was also written to permit data examination and reduction to dimensionless salt concentration maps. Finally, the computer program SUPERENERGY was modified to permit rapid extraction of parameters from dimensionless salt concentration maps. The document describes the computer system, and includes circuit diagrams of all custom built components. It also includes descriptions and listings of all software written, as well as extensive user instructions.

  15. Microcomputer-controlled ultrasonic data acquisition system. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, W.A. Jr.

    1978-11-01

    The large volume of ultrasonic data generated by computer-aided test procedures has necessitated the development of a mobile, high-speed data acquisition and storage system. This approach offers the decided advantage of on-site data collection and remote data processing. It also utilizes standard, commercially available ultrasonic instrumentation. This system is controlled by an Intel 8080A microprocessor. The MCS80-SDK microcomputer board was chosen, and magnetic tape is used as the storage medium. A detailed description is provided of both the hardware and software developed to interface the magnetic tape storage subsystem to Biomation 8100 and Biomation 805 waveform recorders. A boxcar integrator acquisition system is also described for use when signal averaging becomes necessary. Both assembly language and machine language listings are provided for the software.

  16. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  17. Evaluation of clamp effects on LMFBR piping systems

    International Nuclear Information System (INIS)

    Jones, G.L.

    1980-01-01

    Loop-type liquid metal breeder reactor plants utilize thin-wall piping to mitigate through-wall thermal gradients due to rapid thermal transients. These piping loops require a support system to carry the combined weight of the pipe, coolant and insulation and to provide attachments for seismic restraints. The support system examined here utilizes an insulated pipe clamp designed to minimize the stresses induced in the piping. To determine the effect of these clamps on the pipe wall a non-linear, two-dimensional, finite element model of the clamp, insulation and pipe wall was used to determine the clamp/pipe interface load distributions which were then applied to a three-dimensional, finite element model of the pipe. The two-dimensional interaction model was also utilized to estimate the combined clamp/pipe stiffness

  18. Heat transfer performance of multilayer insulation system under roof slab of pool-type LMFBR

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Naohara, Nobuyuki; Uotani, Masaki

    1986-01-01

    To cope with thermal expansion of stainless steel plate, about 90 insulation structures are installed under the roof-slab of pool-type LMFBR. The objective of this study is to evaluate from heat transfer experiment and visualized experiment, the effect of distance between each thermal insulation structure on heat transfer characteristics of insulation system under roof-slab. Two types of insulation structures are selected, one is open type and the other is closed type. Distance between each thermal insulation structure and hot surface temperatures are varied as a parameter. Furthermore, heat flux of the roof-slab insulation system of reactor are estimated from the results of heat transfer experiment. (author)

  19. On the computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.-W.; Fistedis, S.H.

    1977-01-01

    A two-dimensional coupled hydrodynamic-structural response analysis of piping systems is described. Implicit Continuous-Fluid Eulerian (ICE) technique is utilized in the hydrodynamics while a finite-element technique is used in the structural analysis. Different piping components such as elbows, valves, reducers, expansions, heat exchangers, and tees are modelled and coupled with the straight pipe model. An axisymmetric general component model that can be used in modelling valves, reducers, expansions, and heat exchangers is described. At the inlet and outlet region of such component the cross-sectional area may change suddently or gradually, or many not change at all. Among the options available in this model are deformable exterior walls, interior rigid wall simulation, and tube bundle effect. Exterior walls of pipes and components are treated as thin axisymmetric shell. A convected coordinate explicit finite-element scheme for large displacement small strain, elastic-plastic material behavior in which membrane and bending strengths are accounted for is employed. The strains are linearly related to the displacement of the element relative to its convective coordinates, and similarly, the nodal forces are linearly related to the elements stresses. The coupling of the hydrodynamics and structural problems is done in such a way that the hydrodynamics supplies the structure with a pressure loading and the structure supplying the hydrodynamics with a moving boundary condition. Because of the difficulties of handling interior walls that may occupy partial zones, the walls are assumed rigid and limited in their orientation to be parallel to the radial or axial directions, their position to zone boundaries, and their thickness to zero

  20. Single-phase pump model for analysis of LMFBR heat transport systems

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.

    1978-05-01

    A single-phase pump model for transient and steady-state analysis of LMFBR heat transport systems is presented. Fundamental equations of the model are angular momentum balance to determine transient impeller speed and mass balance (including thermal expansion effects) to determine the level of sodium in the pump tank. Pump characteristics are modeled by homologous head and torque relations. All regions of pump operation are represented with reverse rotation allowed. The model also includes option for enthalpy rise calculations and pony motor operation. During steady state, the pump operating speed is determined by matching required head with total load in the circuit. Calculated transient results are presented for pump coastdown and double-ended pipe break accidents. The report examines the influence of frictional torque and specific speed on predicted response for the pump coastdown to natural circulation transient. The results for a double-ended pipe break accident indicate the necessity of including all regions of operation for pump characteristics

  1. SASSYS-1 modelling of RVACS/RACS heat removal in an LMR

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1987-01-01

    The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This air-side RVACS/RACS model is coupled with the sodium-side primary loop thermal hydraulics model in SASSYS-1 to give a complete treatment of the problem. Application of this model to an unprotected loss-of-flow event in the PRISM rector shows that in the long run the RVACS cooling is sufficient to prevent unacceptably high system temperatures in this case

  2. Corrosion critique of the 2 1/4 Cr--1 Mo steel for LMFBR steam generation system applications

    International Nuclear Information System (INIS)

    Zima, G.E.

    1977-07-01

    The unstabilized ferritic steel of nominal composition, 2 1 / 4 Cr-1Mo, has been proposed for critical structural assignments in LMFBR powerplants, specifically: the tubing, tubesheet and shell of the evaporator and superheater components. The interest in this steel has been based on a presumably favorable general corrosion property spectrum, acceptable mechanical properties and fabricability, and certain economies associated with the low alloy content. This report is an attempt at a general corrosion assessment for the 2 1 / 4 Cr-1Mo steel and an identification of corrosion problem areas potential to this steel from the sodium and water/steam systems of the proposed working environment. There is a considerable area of uncertainty in the sodium-side response of 2 1 / 4 Cr-1Mo steel, centered in the loss and redisposition of carbon during long-term exposure to sodium of various impurity backgrounds. It is submitted that present evidence relating to the water/steam-side corrosion behavior of the 2 1 / 4 Cr-1Mo steel, under nominal and conceivable perturbed environmental conditions, constitutes the principal concern for the proposed LMFBR powerplant applications of this steel. It is suggested that this unfavorable corrosion aspect represents an inherent limitation of the low alloy content of this steel, probably largely independent of melting and processing recourses, and it is a sufficient basis to question the incentive for a continuation of the collateral studies of this steel for the proposed LMFBR steam generation system assignments

  3. LMFBR: safety aspects

    International Nuclear Information System (INIS)

    Natta, M.

    1990-01-01

    This presentation of LMFBR safety is limited at Super Phenix reactor. After a brief description of the reactor, some details on safety systems, in normal or accidental conditions, are given. The main functions studied are: chain reaction trip, residual power evacuation, reactor containment. In heavy accident the behaviour of Super Phenix is studied which its particular characteristics and the possibilities of operators reactions. The probability of appearance and the maximum consequences of heavy accidents are given [fr

  4. Small sodium-to-gas leak behavior in relation to LMFBR leak detection system design

    International Nuclear Information System (INIS)

    Hopenfeld, J.; Taylor, G.R.; James, L.A.

    1976-01-01

    Various aspects of sodium-to-gas leaks which must be considered in the design of leak detection systems for LMFBR's are discussed. Attention is focused primarily on small, weeping type leaks. Corrosion rates of steels in fused sodium hydroxide and corrosion damage observed at the site of small leaks lead to the conclusion that the sodium-gas reaction products could attack the primary hot leg piping at rates up to 0.08 mils per hour. Based on theoretical considerations of the corrosion mechanism and on visual observations of pipe topography following small sodium leak tests, it is concluded that pipe damage will be manifested by the formation of small detectable leaks prior to the appearance of larger leaks. The case for uniform pipe corrosion along the pipe circumference or along a vertical section of the pipe is also examined. Using a theoretical model for the gravity flow of sodium and reaction products along the pipe surface and a mass transport controlled corrosion process, it is shown that below sodium leak rates of about 30 g/hr for the primary piping corrosion damage will not extend beyond one radius distance from the leak site. A method of estimating the time delay between the initiation of such leaks and the development of a larger leak due to increased pipe stresses resulting from corrosion is presented

  5. The role of SASSYS-1 in LMR [Liquid Metal Reactor] safety analysis

    International Nuclear Information System (INIS)

    Dunn, F.E.; Wei, T.Y.C.

    1988-01-01

    The SASSYS-1 liquid metal reactor systems analysis computer code is currently being used as the principal tool for analysis of reactor plant transients in LMR development projects. These include the IFR and EBR-II Projects at Argonne National Laboratory, the FFTF project at Westinghouse-Hanford, the PRISM project at General Electric, the SAFR project at Rockwell International, and the LSPB project at EPRI. The SASSYS-1 code features a multiple-channel thermal-hydraulics core representation coupled with a point kinetics neutronics model with reactivity feedback, all combined with detailed one-dimensional thermal-hydraulic models of the primary and intermediate heat transport systems, including pipes, pumps, plena, valves, heat exchangers and steam generators. In addition, SASSYS-1 contains detailed models for active and passive shutdown and emergency heat rejection systems and a generalized plant control system model. With these models, SASSYS-1 provides the capability to analyze a wide range of transients, including normal operational transients, shutdown heat removal transients, and anticipated transients without scram events. 26 refs., 16 figs

  6. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  7. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  8. Analysis of pressure wave transients and seismic response in LMFBR piping systems using the SHAPS code

    International Nuclear Information System (INIS)

    Zeuch, W.R.; Wang, C.Y.

    1985-01-01

    This paper presents some of the current capabilities of the three-dimensional piping code SHAPS and demonstrates their usefulness in handling analyses encountered in typical LMFBR studies. Several examples demonstrate the utility of the SHAPS code for problems involving fluid-structure interactions and seismic-related events occurring in three-dimensional piping networks. Results of two studies of pressure wave propagation demonstrate the dynamic coupling of pipes and elbows producing global motion and rigorous treatment of physical quantities such as changes in density, pressure, and strain energy. Results of the seismic analysis demonstrate the capability of SHAPS to handle dynamic structural response within a piping network over an extended transient period of several seconds. Variation in dominant stress frequencies and global translational frequencies were easily handled with the code. 4 refs., 10 figs

  9. Modeling developments for the SAS4A and SASSYS computer codes

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Wei, T.Y.C.

    1990-01-01

    The SAS4A and SASSYS computer codes are being developed at Argonne National Laboratory for transient analysis of liquid metal cooled reactors. The SAS4A code is designed to analyse severe loss-of-coolant flow and overpower accidents involving coolant boiling, Cladding failures, and fuel melting and relocation. Recent SAS4A modeling developments include extension of the coolant boiling model to treat sudden fission gas release upon pin failure, expansion of the DEFORM fuel behavior model to handle advanced cladding materials and metallic fuel, and addition of metallic fuel modeling capability to the PINACLE and LEVITATE fuel relocation models. The SASSYS code is intended for the analysis of operational and beyond-design-basis transients, and provides a detailed transient thermal and hydraulic simulation of the core, the primary and secondary coolant circuits, and the balance-of-plant, in addition to a detailed model of the plant control and protection systems. Recent SASSYS modeling developments have resulted in detailed representations of the balance of plant piping network and components, including steam generators, feedwater heaters and pumps, and the turbine. 12 refs., 2 tabs

  10. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1979-03-01

    This document contains up-to-date data on existing or firmly decided prototype or demonstration LMFBR reactors (Table I), on planned commercial size LMFBR according to the present status of design (Table II) and on experimental fast reactors such as BOR-60, DFR, EBR-II, FERMI, FFTF, JOYO, KNK-II, PEC, RAPSODIE-FORTISSIMO (Table III). Only corrected and revised parameters submitted by the countries participating in the IWGFR are included in this document

  11. A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

    International Nuclear Information System (INIS)

    Yoshihisa Shindo; Kazuo Haga

    2005-01-01

    Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

  12. An improved steam generator model for the SASSYS code

    International Nuclear Information System (INIS)

    Pizzica, P.A.

    1989-01-01

    A new steam generator model has been developed for the SASSYS computer code, which analyzes accident conditions in a liquid-metal-cooled fast reactor. It has been incorporated into the new SASSYS balance-of-plant model, but it can also function as a stand-alone model. The model provides a full solution of the steady-state condition before the transient calculation begins for given sodium and water flow rates, inlet and outlet sodium temperatures, and inlet enthalpy and region lengths on the water side

  13. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  14. Application of 2-1/4 Cr-1 Mo as a structural material in saturated steam cycle LMFBR systems. Final report

    International Nuclear Information System (INIS)

    Licina, G.J.; Busboom, H.J.; Ring, P.J.; Roy, P.; Schmidt, C.G.; Spalaris, C.N.

    1982-02-01

    The suitability and incentives were examined for using 2-1/4Cr-1Mo steel as a structural material for the entire primary and secondary sodium systems in a 1000 MWe pool-type Liquid Metal Fast Breeder Reactor. The critical properties, advantages and disadvantages of 2-1/4Cr-1Mo, and data needed for design were described for each major component in the reactor. The relative importance of alloy properties to the successful use of ferritics in LMFBR was identified. Licensing issues, likely to surface if ferritic alloys were to be used for critical reactor components, were discussed

  15. Implementation, verification, and validation of the FPIN2 metal fuel pin mechanics model in the SASSYS/SAS4A LMR transient analysis codes

    International Nuclear Information System (INIS)

    Sofu, T.; Kramer, J.M.

    1994-01-01

    The metal fuel version of the FPIN2 code which provides a validated pin mechanics model is coupled with SASSYS/SAS4A Version 3.0 for single pin calculations. In this implementation, SASSY/SAS4A provides pin temperatures, and FPIN2 performs analysis of pin deformation and predicts the time and location of cladding failure. FPIN2 results are also used for the estimates of axial expansion of fuel and associated reactivity effects. The revalidation of the integrated SAS-FPIN2 code system is performed using TREAT tests

  16. Comparison of the SASSYS/SAS4A radial core expansion reactivity feedback model and the empirical correlation for FFTF

    International Nuclear Information System (INIS)

    Wigeland, R.A.

    1987-01-01

    The present emphasis on inherent safety for LMR designs has resulted in a need to represent the various reactivity feedback mechanisms as accurately as possible. The dominant negative reactivity feedback has been found to result from radial expansion of the core for most postulated ATWS events. For this reason, a more detailed model for calculating the reactivity feedback from radial core expansion has been recently developed for use with the SASSYS/SAS4A Code System. The purpose of this summary is to present an extension to the model so that it is more suitable for handling a core restraint design as used in FFTF, and to compare the SASSYS/SAS4A results using this model to the empirical correlation presently being used to account for radial core expansion reactivity feedback to FFTF

  17. Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system

    International Nuclear Information System (INIS)

    Moisseytsev, A.; Sienicki, J.J.

    2012-01-01

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO 2 ) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO 2 cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO 2 . It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO 2 heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO 2 -to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO 2 turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the calculations reveal that the

  18. Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Allard, L.T.; Elerath, J.G.

    1976-02-01

    This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

  19. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    International Nuclear Information System (INIS)

    Tippets, F.E.

    1975-01-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  20. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Tippets, F E

    1975-07-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  1. FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-30

    The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variable lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.

  2. Welding development for LMFBR applications

    International Nuclear Information System (INIS)

    Slaughter, G.M.; Edmonds, D.P.; Goodwin, G.M.; King, J.F.; Moorhead, A.J.

    1976-01-01

    High-quality welds with suitable properties for long-time elevated-temperature nuclear service are among the most critical needs in today's welding technology. Safe, reliable, and economic generation of future power depends on welded construction in systems such as Liquid Metal Fast Breeder Reactors (LMFBRs). Rapid thermal transients in LMFBR systems at coolant temperatures around 590 to 650 0 C (1000 to 1200 0 F) could cause creep and creep-fatigue damage that is not encountered in lower temperature reactor systems. The undesirable consequences of interaction between the two working fluids - sodium and steam - in the steam generators are also of major concern. Thus sound welds that have excellent reliability over a 30-year service life are essential. Several programs are actively underway at ORNL to satisfy this critical need and selected portions of three of these programs are discussed briefly

  3. Studies on modeling to failed fuel detection system response in LMFBR

    International Nuclear Information System (INIS)

    Miyazawa, T.; Saji, G.; Mitsuzuku, N.; Hikichi, T.; Odo, T.; Rindo, H.

    1981-05-01

    Failed Fuel Detection (FFD) system with Fission Products (FP) detection is considered to be the most promissing method, since FP provides direct information against fuel element failure. For designing FFD system and for evaluating FFD signals, some adequate FFD signal response to fuel failure have been required. But few models are available in nowadays. Thus Power Reactor and Nuclear Fuel Development Corporation (PNC) had developed FFD response model with computer codes, based on several fundamental investigations on FP release and FP behavior, and referred to foreign country experiences on fuel failure. In developing the model, noble gas and halogen FP release and behavior were considered, since FFD system would be composed of both cover gas monitoring and delayed neutron monitoring. The developed model can provide typical fuel failure response and detection limit which depends on various background signals at cover gas monitoring and delayed neutron monitoring. According to the FFD response model, we tried to assume fuel failure response and detection limit at Japan experimental fast reactor ''JOYO''. The detection limit of JOYO FFD system was estimated by measuring the background signals. Followed on the studies, a complete computer code has been now made with some improvement. On the paper, the details of the model, out line of developed computer code, status of JOYO FFD system, and trial assumption of JOYO FFD response and detection limit. (author)

  4. Pipe supports and anchors - LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.

    1983-06-01

    Pipe design and support design can not be treated as separate disciplines. A coordinated design approach is required if LMFBR pipe system adequacy is to be achieved at a reasonable cost. It is particularly important that system designers understand and consider those factors which influence support train flexibility and thus the pipe system dynamic stress levels. The system approach must not stop with the design phase but should continue thru the erection and acceptance test procedures. The factors that should be considered in the design of LMFBR pipe supports and anchors are described. The various pipe support train elements are described together with guidance on analysis, design and application aspects. Post erection acceptance and verification test procedures are then discussed

  5. Status of the LMFBR development

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, J J

    1975-01-01

    The development of any new power generation system which can make a major contribution to our energy needs is a multi-faceted task involving the utilization of major human and material resources. The LMFBR development, which has the potential for supplying abundant energy for generations, is therefore a large, multi-faceted program. This summary will cover (1) the need for the liquid metal fast breeder reactor, (2) an overall perspective of its development throughout the world, (3) a brief look at the in-depth technological development program in the United States, (4) a description and status of the two major projects now under way in the program, the Fast Flux Test Facility and the Clinch River Breeder Reactor Plant, and (5) a review of the plans for continued development to achieve a reliable, safe and economic power generation system for practical commercial use on the utility networks of the country.

  6. Self-actuated shutdown system for a commercial size LMFBR. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dupen, C.F.G.

    1978-08-01

    A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility and reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.

  7. Preliminary considerations on developing IAEA technical safeguards for LMFBR power systems

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1980-09-01

    Nuclear fuel cycles safeguards should be considered in the dynamic context of a world deployment of various reactor types and varying availability of fuel-cycle services. There will be a close interaction between thermal-reactor cycles and the future deployment of fast breeders. The quantitites of plutonium and the reprocessing, conversion, fabrication, and storage methods of the fuel for the fast breeders will have a significant impact on safeguards techniques. The approach to the fast breeder fuel cycle safeguards follows the general safeguards system approach proposed by the IAEA. Objective of IAEA safeguards is the detection of diversion of nuclear material and deterrence of such diversion. To achieve independent verification of material balance accountancy requires the capability to monitor inventory status and verify material flows and quantities of all nuclear materials subject to safeguards. Containment and surveillance measures are applied to monitor key measurement points, maintain integrity of material balance, and complement material accountancy. The safeguards study attempts to develop a generic reference IAEA Safeguards System and explores various system options using containment/surveillance and material accountancy instrumentation and integrated systems designs

  8. Self-actuated shutdown system for a commercial size LMFBR. Final report

    International Nuclear Information System (INIS)

    Dupen, C.F.G.

    1978-08-01

    A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility and reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power

  9. Unique Capabilities of the Situational Awareness Sensor Suite for the ISS (SASSI) Mission Concept to Study the Equatorial Ionosphere

    Science.gov (United States)

    Habash Krause, L.; Gilchrist, B. E.; Minow, J. I.; Gallagher, D. L.; Hoegy, W. R.; Coffey, V. N.; Willis, E. M.

    2014-12-01

    We present an overview of a mission concept named Situational Awareness Sensor Suite for the ISS (SASSI) with a special focus here on low-latitude ionospheric plasma turbulence measurements relevant to equatorial spread-F. SASSI is a suite of sensors that improves Space Situational Awareness for the ISS local space environment, as well as unique ionospheric measurements and support active plasma experiments on the ISS. As such, the mission concept has both operational and basic research objectives. We will describe two compelling measurement techniques enabled by SASSI's unique mission architecture. That is, SASSI provides new abilities to 1) measure space plasma potentials in low Earth orbit over ~100 m relative to a common potential, and 2) to investigate multi-scale ionospheric plasma turbulence morphology simultaneously of both ~ 1 cm and ~ 10 m scale lengths. The first measurement technique will aid in the distinction of vertical drifts within equatorial plasma bubbles from the vertical motions of the bulk of the layer due to zonal electric fields. The second will aid in understanding ionospheric plasma turbulence cascading in scale sizes that affect over the horizon radar. During many years of ISS operation, we have conducted effective (but not perfect) human and robotic extravehicular activities within the space plasma environment surrounding the ISS structure. However, because of the complexity of the interaction between the ISS and the space environment, there remain important sources of unpredictable environmental situations that affect operations. Examples of affected systems include EVA safety, solar panel efficiency, and scientific instrument integrity. Models and heuristically-derived best practices are well-suited for routine operations, but when it comes to unusual or anomalous events or situations, there is no substitute for real-time monitoring. SASSI is being designed to deploy and operate a suite of low-cost, medium/high-TRL plasma sensors on

  10. EAC european accident code. A modular system of computer programs to simulate LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Wider, H.; Cametti, J.; Clusaz, A.; Devos, J.; VanGoethem, G.; Nguyen, H.; Sola, A.

    1985-01-01

    One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat transfer together with coolant thermohydraulics in single- and two-phase flow. Temperature variations in fuel, coolant and neighbouring structures induce, in fact, thermal reactivity feedbacks which are added up and put in the neutronics calculation to predict the neutron flux and the subsequent heat generation in the reactor. At this point a whole-core analysis code is necessary to examine for any hypothetical transient whether the various feedbacks result effectively in a negative balance, which is the basis condition to ensure stability and safety. The European Accident Code (EAC), developed at the Joint Research Centre of the CEC at Ispra (Italy), fulfills this objective. It is a modular informatics structure (quasi 2-D multichannel approach) aimed at collecting stand-alone computer codes of neutronics, fuel pin mechanics and hydrodynamics, developed both in national laboratories and in the JRC itself. EAC makes these modules interact with each other and produces results for these hypothetical accidents in terms of core damage and total energy release. 10 refs

  11. LMFBR thermal-striping evaluation

    International Nuclear Information System (INIS)

    Brunings, J.E.

    1982-10-01

    Thermal striping is defined as the fluctuating temperature field that is imposed on a structure when fluid streams at different temperatures mix in the vicinity of the structure surface. Because of the uncertainty in structural damage in LMFBR structures subject to thermal striping, EPRI has funded an effort for the Rockwell International Energy Systems Group to evaluate this problem. This interim report presents the following information: (1) a Thermal Striping Program Plan which identifies areas of analytic and experimental needs and presents a program of specific tasks to define damage experienced by ordinary materials of construction and to evaluate conservatism in the existing approach; (2) a description of the Thermal Striping Test Facility and its operation; and (3) results from the preliminary phase of testing to characterize the fluid environment to be applied in subsequent thermal striping damage experiments

  12. LMFBR with booster pump in pumping loop

    International Nuclear Information System (INIS)

    Rubinstein, H.J.

    1975-01-01

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

  13. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1985-07-01

    This document has been prepared on the basis of information compiled by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains parameters of 25 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBR). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA) are presented because its design was nearly finished and most of the components were fabricated at the time when the project was terminated. Three reactors (RAPSODIE (France), DFR (UK) and EFFBR (USA)) have been shut down. However, they are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. The first LMFBRs (CLEMENTINE (USA), EBR-1 (USA), BR-2 (USSR), BR-5 (USSR)) and very special reactors (LAMPRE (USA), SEFOR (USA)) were not recommended by the members of the IWGFR to be included in the report

  14. A risk-based evaluation of LMFBR containment response under core disruptive accident conditions

    International Nuclear Information System (INIS)

    Hartung, J.; Berk, S.

    1978-01-01

    Probabilistic risk methodology is utilized to evaluate the failure modes and effects of LMFBR containment systems under Core Disruptive Accident (CDA) conditions. First, the potential causes of LMFBR containment failure under CDA conditions are discussed and categorized. Then, a simple scoping-type risk assessment of a reference design is presented to help place these potential causes of failure in perspective. The highest risk containment failure modes are identified for the reference design, and several design and research and development options which appear capable of reducing these risks are discussed. The degree to which large LMFBR containment systems must mitigate the consequences of CDA's to achieve a level of risk (for LMFBR's) comparable to the already very low risk of contemporary LWR's is explored. Based on the results of this evaluation, several suggestions are offered concerning CDA-related design goals and research and development priorities for large LMFBR's. (author)

  15. Airborne effluent control for LMFBR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-01-01

    A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including 131 I, krypton, tritium, 129 I, and most recently 14 C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems

  16. Replaceable LMFBR core components

    International Nuclear Information System (INIS)

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  17. SASSYS-1 balance-of-plant component models for an integrated plant response

    International Nuclear Information System (INIS)

    Ku, J.-Y.

    1989-01-01

    Models of power plant heat transfer components and rotating machinery have been added to the balance-of-plant model in the SASSYS-1 liquid metal reactor systems analysis code. This work is part of a continuing effort in plant network simulation based on the general mathematical models developed. The models described in this paper extend the scope of the balance-of-plant model to handle non-adiabatic conditions along flow paths. While the mass and momentum equations remain the same, the energy equation now contains a heat source term due to energy transfer across the flow boundary or to work done through a shaft. The heat source term is treated fully explicitly. In addition, the equation of state is rewritten in terms of the quality and separate parameters for each phase. The models are simple enough to run quickly, yet include sufficient detail of dominant plant component characteristics to provide accurate results. 5 refs., 16 figs., 2 tabs

  18. The Snapshot A Star SurveY (SASSY)

    Science.gov (United States)

    Garani, Jasmine I.; Nielsen, Eric; Marchis, Franck; Liu, Michael C.; Macintosh, Bruce; Rajan, Abhijith; De Rosa, Robert J.; Jinfei Wang, Jason; Esposito, Thomas M.; Best, William M. J.; Bowler, Brendan; Dupuy, Trent; Ruffio, Jean-Baptiste

    2018-01-01

    The Snapshot A Star Survey (SASSY) is an adaptive optics survey conducted using NIRC2 on the Keck II telescope to search for young, self-luminous planets and brown dwarfs (M > 5MJup) around high mass stars (M > 1.5 M⊙). We present the results of a custom data reduction pipeline developed for the coronagraphic observations of our 200 target stars. Our data analysis method includes basic near infrared data processing (flat-field correction, bad pixel removal, distortion correction) as well as performing PSF subtraction through a Reference Differential Imaging algorithm based on a library of PSFs derived from the observations using the pyKLIP routine. We present the results from the pipeline of a few stars from the survey with analysis of candidate companions. SASSY is sensitive to companions 600,000 times fainter than the host star withint the inner few arcseconds, allowing us to detect companions with masses ~8MJup at age 110 Myr. This work was supported by the Leadership Alliance's Summer Research Early Identification Program at Stanford University, the NSF REU program at the SETI Institute and NASA grant NNX14AJ80G.

  19. An improved steam generator model for the SASSYS code

    International Nuclear Information System (INIS)

    Pizzica, P.A.

    1989-01-01

    A new steam generator model has been developed for the SASSYS computer code, which analyzes accident conditions in a liquid metal cooled fast reactor. It has been incorporated into the new SASSYS balance-of-plant model but it can also function on a stand-alone basis. The steam generator can be used in a once-through mode, or a variant of the model can be used as a separate evaporator and a superheater with recirculation loop. The new model provides for an exact steady-state solution as well as the transient calculation. There was a need for a faster and more flexible model than the old steam generator model. The new model provides for more detail with its multi-mode treatment as opposed to the previous model's one node per region approach. Numerical instability problems which were the result of cell-centered spatial differencing, fully explicit time differencing, and the moving boundary treatment of the boiling crisis point in the boiling region have been reduced. This leads to an increase in speed as larger time steps can now be taken. The new model is an improvement in many respects. 2 refs., 3 figs

  20. SASSYS-1 computer code verification with EBR-II test data

    International Nuclear Information System (INIS)

    Warinner, D.K.; Dunn, F.E.

    1985-01-01

    The EBR-II natural circulation experiment, XX08 Test 8A, is simulated with the SASSYS-1 computer code and the results for the latter are compared with published data taken during the transient at selected points in the core. The SASSYS-1 results provide transient temperature and flow responses for all points of interest simultaneously during one run, once such basic parameters as pipe sizes, initial core flows, and elevations are specified. The SASSYS-1 simulation results for the EBR-II experiment XX08 Test 8A, conducted in March 1979, are within the published plant data uncertainties and, thereby, serve as a partial verification/validation of the SASSYS-1 code

  1. LMFBR plant parameters 1991

    International Nuclear Information System (INIS)

    1991-03-01

    The document has been prepared on the basis of information provided by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains updated parameters of 27 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBRs). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA), PEC (Italy), RAPSODIE (France), DFR (UK) and EFFBR (USA) are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. Two more reactors appeared in the list: European Fast Reactor (EFR) and PRISM (USA). Parameters of these reactors included in this publication are based on the data from the papers presented at the 23rd Annual Meeting of the IWGFR. All in all more than four hundred corrections and additions have been made to update the document. The report is intended for specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors

  2. Assessment of accident energetics in LMFBR core-disruptive accidents

    International Nuclear Information System (INIS)

    Fauske, H.K.

    1977-01-01

    An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

  3. Seismic response and damping tests of small bore LMFBR piping and supports

    International Nuclear Information System (INIS)

    Barta, D.A.; Anderson, M.J.; Severud, L.K.; Lindquist, M.R.

    1984-01-01

    Seismic testing and analysis of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described. Measured responses to simulated seismic excitations are compared with analytical predictions based on NRC Regulatory Guide 1.61 and measured system damping values. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps

  4. Italian guidelines for energy performance of cultural heritage and historical buildings: the case study of the Sassi of Matera.

    Science.gov (United States)

    Negro, Elisabetta; Cardinale, Tiziana; Cardinale, Nicola

    2016-04-01

    researches is that these buildings are able to reduce the temperature oscillations of the external environment as a result of their high thermal mass. These traditional settlements are typical of the Mediterranean area that has summers with high temperature and daily thermal oscillations. So we can conclude that these buildings could be considered as bioclimatic. Regarding the monuments, as the Matera Cathedral, the evaluation of the indoor microclimate during and after the restoration works shows excellent results and ensures the optimal preservation of artistic heritage from the thermo hygrometric point of view. The plant solution adopted (installation of floor heating system) is in line with the Italian guidelines because this type of system at low temperature allows high energy savings as it enables the use of combustion systems with high-efficiency (condensing boilers) and/or renewable energy installations (heat pumps, solar thermal collectors). A complete knowledge of historical heritage and energy performance of Sassi building is a strategy indicated by Italian guidelines in order to preserve the identity of their inhabitants. So it is necessary to conduct a complete mapping of the entire heritage of this city and develop specific guidelines which combine technical and economic feasibility, appropriate landscaping and architectural integration and environmental sustainability within a proper building lifecycle.

  5. Studies of LMFBR: method of analysis and some results

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.; Nascimento, J.A. do.

    1983-01-01

    Some results of recent studies of LMFBR characteristics are summarized. A two-dimensional model of the LMFBR is taken from a publication and used as the base model for the analysis. Axial structures are added to the base model and a three-dimensional (Δ - Z) calculation has been done. Two dimensional (Δ and RZ) calculations are compared with the three-dimensional and published results. The eigenvalue, flux and power distributions, breeding characteristics, control rod worth, sodium-void and Doppler reactivities are analysed. Calculations are done by CITATION using six-group cross sections collapsed regionwise by EXPANDA in one-dimensional geometries from the 70-group JFS library. Burnup calculations of a simplified thorium-cycle LMFBR have also been done in the RZ geometry. Principal results of the studies are: (1) the JFS library appears adequate for predicting overall characteristics of an LMFBR, (2) the sodium void reactivity is negative within - 25 cm from the outer boundary of the core, (3) the halflife of Pa-233 must be considered explicitly in burnup analyses, and (4) two-dimensional (RZ and Δ) calculations can be used iteratively to analyze three-dimensional reactor systems. (Author) [pt

  6. SASSYS validation with the EBR-II shutdown heat removal tests

    International Nuclear Information System (INIS)

    Herzog, J.P.

    1989-01-01

    SASSYS is a coupled neutronic and thermal hydraulic code developed for the analysis of transients in liquid metal cooled reactors (LMRs). The code is especially suited for evaluating of normal reactor transients -- protected (design basis) and unprotected (anticipated transient without scram) transients. Because SASSYS is heavily used in support of the IFR concept and of innovative LMR designs, such as PRISM, a strong validation base for the code must exist. Part of the validation process for SASSYS is analysis of experiments performed on operating reactors, such as the metal fueled Experimental Breeder Reactor -- II (EBR-II). During the course of a series of historic whole-plant experiments, EBR-II illustrated key safety features of metal fueled LMRs. These experiments, the Shutdown Heat Removal Tests (SHRT), culminated in unprotected loss of flow and loss of heat sink transients from full power and flow. Analysis of these and earlier SHRT experiments constitutes a vital part of SASSYS validation, because it facilitates scrutiny of specific SASSYS models and of integrated code capability. 12 refs., 11 figs

  7. Application of the SASSI soil structure interaction method to CANDU 6 NPP seismic analysis

    International Nuclear Information System (INIS)

    Ricciuti, R.A.; Elgohary, M.; Usmani, S.A.

    1996-01-01

    The standard CANDU 6 NPP has been conservatively qualified for a Design Basis Earthquake (DBE) peak horizontal ground acceleration of 0.2 g. Currently there are potential opportunities for siting the CANDU 6 at higher seismicity sites. In order to be able to extend the use of a standardized design for sites with higher seismicity than the standard plant, various design options, including the use of the SASSI Soil Structure Interaction (SSI) analysis method, are being evaluated. This paper presents the results of a study to assess the potential benefits from utilization of the SASSI computer program and the use of more realistic damping ratios for the structures

  8. Status of U.S. LMFBR programme

    International Nuclear Information System (INIS)

    Yevich, J.

    1978-01-01

    The determents of the decision for deterrence of commercial reprocessing and further demonstration of the plutonium breeder were based on two premises: time is needed to establish the programme for non-proliferating fuel cycle and there is a lessened sense of urgency for the USA to establish a commercial breeder in the near future. A strong, well funded base technology effort remains and will continue until institutional and technical solutions can be found to minimize or eliminate the proliferation risk. An LMFBR option will be maintained. The FFTF will be coming on line providing a powerful tool in breeder fuel and materials development and a baseline from which to scale up heat transfer systems and components. Sodium system hardware development and testing will continue to have high priority

  9. Nuclear welding, application for an LMFBR

    International Nuclear Information System (INIS)

    Patriarca, P.; Goodwin, G.M.

    1975-01-01

    Fabrication of an LMFBR system is discussed, with emphasis on areas where joint welding innovations have been introduced. Each major component of the system, including reactor vessel, intermediate heat exchanger, steam generator, and sodium-containment piping, is treated separately. Developmet of special filler metals to avoid the low elevated-temperature creep ductility obtained with conventional austenitic stainless steel weldments is reported. Bore-side welding of steam generator tube-to-tubesheet joints with and without filler metal is desirable to improve inspectability and eliminate the crevice inherent with face-side weld design, thus minimizing corrosion problems. Automated welding methods for sodium-containment piping are summarized which iminimize and control distortion and ensure welds of high integrity. Selection of materials for the various components is discussed for plants presently under construction, and materials predictions are made for future concepts. (U.S.)

  10. Microprocessor-based integrated LMFBR core surveillance

    International Nuclear Information System (INIS)

    Gmeiner, L.

    1984-06-01

    This report results from a joint study of KfK and INTERATOM. The aim of this study is to explore the advantages of microprocessors and microelectronics for a more sophisticated core surveillance, which is based on the integration of separate surveillance techniques. Due to new developments in microelectronics and related software an approach to LMFBR core surveillance can be conceived that combines a number of measurements into a more intelligent decision-making data processing system. The following techniques are considered to contribute essentially to an integrated core surveillance system: - subassembly state and thermal hydraulics performance monitoring, - temperature noise analysis, - acoustic core surveillance, - failure characterization and failure prediction based on DND- and cover gas signals, and - flux tilting techniques. Starting from a description of these techniques it is shown that by combination and correlation of these individual techniques a higher degree of cost-effectiveness, reliability and accuracy can be achieved. (orig./GL) [de

  11. Design approaches to achieve competitive LMFBR capital costs

    International Nuclear Information System (INIS)

    Arnold, W.H.; Ehrman, C.S.; Sharbaugh, J.E.; Young, W.H.

    1982-01-01

    Through analysis of the essential functional elements of an LMFBR, numerous ways were found to simplify system design, reduce the size of components and equipment, and eliminate some components and systems. The projected capital cost per net kW of this design is competitive with that of current PWRs. RandD programs and the construction and operation of CRBRP now are needed to prove out the features of this new design

  12. Standpipe-bubbler pump level control study: 1E-3862, controlled clearance pump/system Model W. P. 7373. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    1976-01-01

    Computer simulation results of the standpipe-bubbler pump level control system/controlled clearance pump configuration 1E-3862 using specific plan elevations are presented. Fluid level control system behavior is presented in graphical form for normal plant loading, unloading, and trip transients. Satisfactory fluid level control can be attained for this configuration with a standpipe-bubbler system.

  13. Development of acidic processes for decontaminating LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    Hill, E F [Rockwell International, Atomics International Division, Canoga Park (United States); Colburn, R P; Lutton, J M; Maffei, H P [Hanford Engineering Development Laboratory, Richland (United States)

    1978-08-01

    The objective of the DOE decontamination program is to develop a well characterized chemical decontamination process for application to LMFBR primary system components that subsequently permits contact maintenance and allows requalification of the components for reuse in reactors. The paper describes the subtasks of deposit characterization, development of requalification and process acceptance criteria, development of process evaluation techniques and studies which led to a new acidic process for decontaminating 304 stainless steel hot leg components.

  14. LMFBR operational safety: the EBR-II experience

    International Nuclear Information System (INIS)

    Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

    1978-01-01

    The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

  15. LMFBR technology. FFTF cover-gas leakage calculation

    International Nuclear Information System (INIS)

    Deboi, H.

    1974-01-01

    The FFTF LMFBR is intended to have a near zero release of radioactive gases during normal reactor operation with 1% failed fuel. This report presents calculations which provide an approximation of these cover gas leakages. Data from ongoing static and dynamic seal leak tests at AI are utilized. Leakage through both elastomeric and metallic seals in all sub-assemblies and penetrations comprising the reactor cover gas containment during reactor operation system are included

  16. Development of acidic processes for decontaminating LMFBR components

    International Nuclear Information System (INIS)

    Hill, E.F.; Colburn, R.P.; Lutton, J.M.; Maffei, H.P.

    1978-01-01

    The objective of the DOE decontamination program is to develop a well characterized chemical decontamination process for application to LMFBR primary system components that subsequently permits contact maintenance and allows requalification of the components for reuse in reactors. The paper describes the subtasks of deposit characterization, development of requalification and process acceptance criteria, development of process evaluation techniques and studies which led to a new acidic process for decontaminating 304 stainless steel hot leg components

  17. LMFBR core design analysis

    International Nuclear Information System (INIS)

    Cho, M.; Yang, J.C.; Yoh, K.C.; Suk, S.D.; Soh, D.S.; Kim, Y.M.

    1980-01-01

    The design parameters of a commercial-scale fast breeder reactor which is currently under construction by regeneration of these data is preliminary analyzed. The analysis of nuclear and thermal characteristics as well as safety features of this reactor is emphasized. And the evaluation of the initial core mentioned in the system description is carried out in the areas of its kinetics and control system, and, at the same time, the flow distribution of sodium and temperature distribution of the initial FBR core system are calculated. (KAERI INIS Section)

  18. Use of reliability in the LMFBR industry

    International Nuclear Information System (INIS)

    Penland, J.R.; Smith, A.M.; Goeser, D.K.

    1977-01-01

    This mission of a Reliability Program for an LMFBR should be to enhance the design and operational characteristics relative to safety and to plant availability. Successful accomplishment of this mission requires proper integration of several reliability engineering tasks--analysis, testing, parts controls and program controls. Such integration requires, in turn, that the program be structured, planned and managed. This paper describes the technical integration necessary and the management activities required to achieve mission success for LMFBR's

  19. Structural and containment response to LMFBR accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Fistedis, S.H.; Baker, L. Jr.; Stepnewski, D.D.; Peak, R.D.; Gluekler, E.L.

    1978-01-01

    The results of current developments in analysing the response of reactor structures and containment to LMFBR accidents are presented. The current status of analysis of the structural response of LMFBR's to core disruptive accidents, including head response, potential missile generation and the effects of internal structures are presented. The results of recent experiments to help clarify the thermal response of reactor structures to molten core debris are summarized, including the use of this data to calculate the response of the secondary containment. (author)

  20. SASSIE: A program to study intrinsically disordered biological molecules and macromolecular ensembles using experimental scattering restraints

    Science.gov (United States)

    Curtis, Joseph E.; Raghunandan, Sindhu; Nanda, Hirsh; Krueger, Susan

    2012-02-01

    A program to construct ensembles of biomolecular structures that are consistent with experimental scattering data are described. Specifically, we generate an ensemble of biomolecular structures by varying sets of backbone dihedral angles that are then filtered using experimentally determined restraints to rapidly determine structures that have scattering profiles that are consistent with scattering data. We discuss an application of these tools to predict a set of structures for the HIV-1 Gag protein, an intrinsically disordered protein, that are consistent with small-angle neutron scattering experimental data. We have assembled these algorithms into a program called SASSIE for structure generation, visualization, and analysis of intrinsically disordered proteins and other macromolecular ensembles using neutron and X-ray scattering restraints. Program summaryProgram title: SASSIE Catalogue identifier: AEKL_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKL_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: GNU General Public License v3 No. of lines in distributed program, including test data, etc.: 3 991 624 No. of bytes in distributed program, including test data, etc.: 826 Distribution format: tar.gz Programming language: Python, C/C++, Fortran Computer: PC/Mac Operating system: 32- and 64-bit Linux (Ubuntu 10.04, Centos 5.6) and Mac OS X (10.6.6) RAM: 1 GB Classification: 3 External routines: Python 2.6.5, numpy 1.4.0, swig 1.3.40, scipy 0.8.0, Gnuplot-py-1.8, Tcl 8.5, Tk 8.5, Mac installation requires aquaterm 1.0 (or X window system) and Xcode 3 development tools. Nature of problem: Open source software to generate structures of disordered biological molecules that subsequently allow for the comparison of computational and experimental results is limiting the use of scattering resources. Solution method: Starting with an all atom model of a protein, for example, users can input

  1. Expansion joints for LMFBR

    International Nuclear Information System (INIS)

    Dzenus, M.; Hundhausen, W.; Jansing, W.

    1980-01-01

    This discourse recounts efforts put into the SNR-2 project; specifically the development of compensation devices. The various prototypes of these compensation devices are described and the state of the development reviewed. Large Na (sodium)-heat transfer systems require a lot of valuable space if the component lay-out does not include compensation devices. So, in order to condense the spatial requirement as much as possible, expansion joints must be integrated into the pipe system. There are two basic types to suit the purpose: axial expansion joints and angular expansion joints. The expansion joints were developed on the basis of specific design criteria whereby differentiation is made between expansion joints of small and large nominal diameter. Expansion joints for installation in the sodium-filled primary piping are equipped with safety bellows in addition to the actual working bellows. Expansion joints must be designed and mounted in a manner to completely withstand seismic forces. The design must exclude any damage to the bellows during intermittent operations, that is, when sodium is drained the bellows' folds must be completely empty; otherwise residual solidified sodium could destroy the bellows when restarting. The expansion joints must be engineered on the basis of the following design data for the secondary system of the SNR project: working pressure: 16 bar; failure mode pressure: 5 events; failure mode: 5 sec., 28.5 bar, 520 deg. C; working temperature: 520 deg. C; temperature transients: 30 deg. C/sec.; service life: 200,000 h; number of load cycles: 10 4 ; material: 1.4948 or 1.4919; layer thickness of folds: 0.5 mm; angular deflection (DN 800): +3 deg. C or; axial expansion absorption (DN 600): ±80 mm; calculation: ASME class. The bellows' development work is not handled within this scope. The bellows are supplied by leading manufacturers, and warrant highest quality. Multiple bellows were selected on the basis of maximum elasticity - a property

  2. The LMFBR safety case

    International Nuclear Information System (INIS)

    Smith, D.

    1990-01-01

    The main objective of this report is to review the present status of the safety case for the liquid metal cooled fast reactor (FBR). A particular emphasis is placed on activities in Europe where the FBR has been progressively developed for many years during which time systems have passed from small experimental plants to the 1200 MWe SPX-1. The FBR has been found to be an easily controlled plant with low impact on the environment and low dose rates to operational personnel. Aspects of reactor design and associated R and D that are required for FBRs to be licensed and the progress made to meet these requirements are described. Fault conditions in the credible range can be dealt with safely, the FBR having several advantageous characteristics which assist safety. Also measures are foreseen to mitigate potential consequences of more severe but improbable accidents. This study sponsored by the Commission of the European Communities was carried out by Colenco Ltd in vlose collaboration with the Safety Working Group (SWG) which is a subgroup of the CEC Fast Reactor Coordinating Committee (FRCC)

  3. Comparative analysis of LMFBR licensing in the United States and other countries - notably France. Executive summary

    International Nuclear Information System (INIS)

    Golay, M.W.; Castillo, M.

    1981-01-01

    The safety-related design aspects and licensing experiences of LMFBR projects in other democratic countries have been studied and contrasted to those in the United States in order to understand the importance of different approaches to safety, and also to understand better the system of the United States. The regulatory systems and LMFBR programs of France and the United States are contrasted in detail, and that of West Germany is also studied. The programs of Japan and the United Kingdom receive considerably less attention, and that of the Soviet Union is ignored

  4. Investigating the Response and Expansion of Plasma Plumes in a Mesosonic Plasma Using the Situational Awareness Sensor Suite for the ISS (SASSI)

    Science.gov (United States)

    Gilchrist, Brian E.; Hoegy, W. R.; Krause, L. Habash; Minow, J. I.; Coffey, V. N.

    2014-01-01

    To study the complex interactions between the space environment surrounding the International Space Station (ISS) and the ISS space vehicle, we are exploring a specialized suite of plasma sensors, manipulated by the Space Station Remote Manipulator System (SSRMS) to probe the near-ISS mesosonic plasma ionosphere moving past the ISS. It is proposed that SASSI consists of the NASA Marshall Space Flight Center's (MSFC's) Thermal Ion Capped Hemispherical Spectrometer (TICHS), Thermal Electron Capped Hemispherical Spectrometer (TECHS), Charge Analyzer Responsive to Local Oscillations (CARLO), the Collimated PhotoElectron Gun (CPEG), and the University of Michigan Advanced Langmuir Probe (ALP). There are multiple expected applications for SASSI. Here, we will discuss the study of fundamental plasma physics questions associated with how an emitted plasma plume (such as from the ISS Plasma Contactor Unit (PCU)) responds and expands in a mesosonic magnetoplasma as well as emit and collect current. The ISS PCU Xe plasma plume drifts through the ionosphere and across the Earth's magnetic field, resulting in complex dynamics. This is of practical and theoretical interest pertaining to contamination concerns (e.g. energetic ion scattering) and the ability to collect and emit current between the spacecraft and the ambient plasma ionosphere. This impacts, for example, predictions of electrodynamic tether current performance using plasma contactors as well as decisions about placing high-energy electric propulsion thrusters on ISS. We will discuss the required measurements and connection to proposed instruments for this study.

  5. My Sassy Girl: a qualitative study of women's aggression in dating relationships in Beijing.

    Science.gov (United States)

    Wang, Xiying; Petula; Ho, Sik Ying

    2007-05-01

    The Korean film My Sassy Girl was very popular among young people in urban China in 2001. After the release of the movie, the new image of the "sassy girl" emerged in different media. This study uses the media image of the sassy girl as a stimulus material in interviews and focus groups to explore how young men and women make sense of women's aggression in dating relationships. This qualitative study is mainly based on two focus groups and in-depth interviews with 20 informants (13 female and 7 male). The study sheds light on how the competing and multilayered discourses in contemporary China regarding gender, love, and sex have left some space for young adults to justify women's aggression in dating relationships. We see how young adults in Beijing situate themselves within this set of social cultural discourses and make use of them to "do" their gender through their aggressive behavior and interactions with each other. New images of men and women are thereby created.

  6. Analytical approach for confirming the achievement of LMFBR reliability goals

    International Nuclear Information System (INIS)

    Ingram, G.E.; Elerath, J.G.; Wood, A.P.

    1981-01-01

    The approach, recommended by GE-ARSD, for confirming the achievement of LMFBR reliability goals relies upon a comprehensive understanding of the physical and operational characteristics of the system and the environments to which the system will be subjected during its operational life. This kind of understanding is required for an approach based on system hardware testing or analyses, as recommended in this report. However, for a system as complex and expensive as the LMFBR, an approach which relies primarily on system hardware testing would be prohibitive both in cost and time to obtain the required system reliability test information. By using an analytical approach, results of tests (reliability and functional) at a low level within the specific system of interest, as well as results from other similar systems can be used to form the data base for confirming the achievement of the system reliability goals. This data, along with information relating to the design characteristics and operating environments of the specific system, will be used in the assessment of the system's reliability

  7. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  8. Review of PRA methodology for LMFBR

    International Nuclear Information System (INIS)

    Yang, J. E.

    1999-02-01

    Probabilistic Risk Assessment (PRA) has been widely used as a tool to evaluate the safety of NPPs (Nuclear Power Plants), which are in the design stage as well as in operation. Recently, PRA becomes one of the licensing requirements for many existing and new NPPs. KALIMER is a Liquid Metal Fast Breeder Reactor (LMFBR) being developed by KAERI. Since the design concept of KALIMER is similar to that of the PRISM plant developed by GE, it would be appropriate to review the PRA methodology of PRISM as the first step of KALIMER PRA. Hence, in this report summarizes the PRA methodology of PRISM plant, and the required works for the PSA of KALIMER based on the reviewed results. The PRA technology of PRISM plant consists of following five major tasks: (1) development of initiating event list, (2) development of system event tree, (3) development of core response event tree, (4) development of containment response event tree, and (5) consequences and risk estimation. The estimated individual and societal risk measures show that the risk from a PRISM module is substantially less than the NRC goal. Each task is compared to the PRA methodology of Light Water Reactor (LWR)/Pressurized Heavy Water Reactor (PHWR). In the report, each task of PRISM PRA methodology is reviewed and compared to the corresponding part of LWR/PHWR PSA performed in Korea. The parts that are not modeled appropriately in PRISM PRA are identified, and the recommendations for KALIMER PRA are stated. (author). 14 refs., 9 tabs., 4 figs

  9. Intelligent type sodium instrumentations for LMFBR

    International Nuclear Information System (INIS)

    Chen Daolong

    1996-07-01

    The constructions and performances of lots of newly developed intelligent type sodium instrumentations are described. The graduation characteristic equations for corresponding transducer using the medium temperature as the parameter are given. These intelligent type sodium instrumentations are possessed of good linearity. The accurate measurement data of sodium process parameters (flowrate, pressure and level) can be obtained by means of their on-line compensation function of the temperature effect. Moreover, these intelligent type sodium instrumentations are possessed of the self-inspection, the electric shutoff protection, the setting of full-scale, the setting of alarm limits (two upper limits and two lower limits alarms), the thermocouple breaking alarm, mutual isolative the 0∼10 V direct-current analogue output and the CENTRONICS standard digital output, and the alarm relay contact output. Theses intelligent type sodium instrumentations are suitable particularly for the instrument, control and protective systems of LMFBR by means of these excellent functions based on microprocessor. The basic errors of the intelligent type sodium flowmeter, immersed sodium flowmeter, sodium manometer and sodium level gauge are +-2%, +-2.3%, +-0.3% and +-1.9% of measuring ranges respectively. (9 figs.)

  10. Tank type LMFBR type reactors

    International Nuclear Information System (INIS)

    Shimizu, Hiroshi

    1985-01-01

    Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

  11. Structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    1983-01-01

    Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start

  12. Structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    Vaze, M.K.K.

    1983-01-01

    The use of elastic analysis for structural design of LMFBR components is discussed. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed Prototype Fast Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is same as that of Rapsodie. Nevertheless, the design had to be checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, ASME Code Section III and the Code Case N-47 are used for high temperature design. The problems faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's shakedown and plastic cycling criteria for ratchet free operation to biaxial stress fields

  13. Status of LMFBR development project in Japan

    International Nuclear Information System (INIS)

    Nagane, G.; Akebi, M.; Matsuno, Y.

    1987-01-01

    Initiation of the LMFBR development project in Japan was decided by the Atomic Energy Commission of Japan in 1966. In 1967, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established to realize the project as a part of its tasks of a wide scope covering all the reseatch and development activities concerning fuel cycle. In the present paper the status of experimental fast reactor (Joyo), which is the first milestone of the LMFBR project, prototype fast reactor (Monju) and R and D activities supporting the project including that for larger LMFBRs in the future is described. (author)

  14. Attenuation of airborne debris from LMFBR accidents

    International Nuclear Information System (INIS)

    Morewitz, H.A.; Johnson, R.P.; Nelson, C.T.; Vaughan, E.U.; Guderjahn, C.A.; Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

    1978-01-01

    Experimental and theoretical studies have been performed to characterize the behavior of airborne particulates (aerosols) expected to be produced by hypothetical core disassembly accidents (HCDA's) in liquid metal fast breeder reactors (LMFBR's). These aerosol studies include work on aerosol transport in a 20-m high, 850-m 3 closed vessel at moderate concentrations; aerosol transport in a small vessel under conditions of high concentration (approximately 1,000 g/m 3 ), high turbulence, and high temperature (approximately 2000 0 C); and aerosol transport through various leak paths. These studies have shown that tittle, if any, airborne debris from LMFBR HCDA's would reach the atmosphere exterior to an intact reactor containment building. (author)

  15. Sodium water reaction R and D for French LMFBR

    International Nuclear Information System (INIS)

    Cambillard, E.; Finck, P.; Lapicore, A.; Simeon, C.

    1985-01-01

    This paper presents the research and development which is underway for the French LMFBR steam generator safety study. The program comprises three major areas: (1) the analysis of realistic leaks, which includes the leak evolution and its consequences; (2) the response time of leak detection systems compared to leak propagation phenomena; and (3) the guillotine rupture (DBA) studies relative to source term evaluation by experimental/calculational approach and mechanical calculations. This program has provided information for the demonstrations of the steam generator safety in respect to a sodium-water reaction

  16. LMFBR steam generator leak detection development in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Magee, P M; Gerrels, E E; Greene, D A [General Electric Company, Sunnyvale, CA (United States); McKee, J [Argonne National Laboratory, Argonne, IL (United States)

    1978-10-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H{sub 2} and O{sub 2}) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  17. Immersed acoustical transducers and their potential uses in LMFBR

    International Nuclear Information System (INIS)

    Argous, J.P.; Brunet, M.; Baron, J.; Lhuillier, C.; Segui, J.L.

    1980-04-01

    Six years satisfactory operation in PHENIX has proved the reliability and effectivness of under-sodium viewing (VISUS) and Acoustic Detection. This fact has been strong incentive to maintain, on the future LMFBR the visus as well as the Acoustic Detection functions. These two functions are performed on SUPER PHENIX, by two sets of distinct systems using the well-known solution. Taking into account of recent improvements in sodium immersible acoustic transducers technology, CEA decided to undertake the development of a multi-functions instrument. This paper gives an outline of this new concept, which should be able to reduce the cost and the complexity of core instrumentation

  18. CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

    1983-01-01

    The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

  19. LMFBR steam generator leak detection development in the United States

    International Nuclear Information System (INIS)

    Magee, P.M.; Gerrels, E.E.; Greene, D.A.; McKee, J.

    1978-01-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H 2 and O 2 ) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  20. Scale modelling in LMFBR safety

    International Nuclear Information System (INIS)

    Cagliostro, D.J.; Florence, A.L.; Abrahamson, G.R.

    1979-01-01

    This paper reviews scale modelling techniques used in studying the structural response of LMFBR vessels to HCDA loads. The geometric, material, and dynamic similarity parameters are presented and identified using the methods of dimensional analysis. Complete similarity of the structural response requires that each similarity parameter be the same in the model as in the prototype. The paper then focuses on the methods, limitations, and problems of duplicating these parameters in scale models and mentions an experimental technique for verifying the scaling. Geometric similarity requires that all linear dimensions of the prototype be reduced in proportion to the ratio of a characteristic dimension of the model to that of the prototype. The overall size of the model depends on the structural detail required, the size of instrumentation, and the costs of machining and assemblying the model. Material similarity requires that the ratio of the density, bulk modulus, and constitutive relations for the structure and fluid be the same in the model as in the prototype. A practical choice of a material for the model is one with the same density and stress-strain relationship as the operating temperature. Ni-200 and water are good simulant materials for the 304 SS vessel and the liquid sodium coolant, respectively. Scaling of the strain rate sensitivity and fracture toughness of materials is very difficult, but may not be required if these effects do not influence the structural response of the reactor components. Dynamic similarity requires that the characteristic pressure of a simulant source equal that of the prototype HCDA for geometrically similar volume changes. The energy source is calibrated in the geometry and environment in which it will be used to assure that heat transfer between high temperature loading sources and the coolant simulant and that non-equilibrium effects in two-phase sources are accounted for. For the geometry and flow conitions of interest, the

  1. Operating conditions of steam generators for LMFBR's

    International Nuclear Information System (INIS)

    Ratzel, W.

    1975-01-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  2. Operating conditions of steam generators for LMFBR's

    Energy Technology Data Exchange (ETDEWEB)

    Ratzel, W

    1975-07-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  3. THE LMFBR, key to the future

    International Nuclear Information System (INIS)

    Chipman, G.L. Jr.

    1982-01-01

    This survey explains the United States prospects for utilizing the LMFBR as a mean of meeting future energy demands. Nuclear option will represent a good financial investment only when breeder will be proved as a cost-effective option. International cooperation and combined programs are very helpful to develop breeder reactor power resource

  4. LMFBR safety experiment facility planning and analysis

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Scott, J.H.

    1976-01-01

    In the past two years considerable effort has been placed on the planning and design of new facilities for the resolution of LMFBR safety issues. The paper reviews the key issues, the experiments needed to resolve them, and the design aspects of proposed new facilities. In addition, it presents a decision theory approach to selecting an optimal combination of modified and new facilities

  5. LMFBR subassembly response to simulated local pressure loadings

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1976-01-01

    The structural response of liquid metal fast breeder reactor (LMFBR) subassemblies to local accidental events is of interest in assessing the safety of such systems. Problems to be resolved include failure propagation modes from pin to pin and from subassembly to subassembly. Factors which must be considered include: (a) the geometry of the structure, (b) uncertainty of the pressure-energy source, (c) uncertainty of materials properties under reactor operating conditions, and (d) the difficulty in performing in-pile or out-of-pile experiments which would simulate the above conditions. The main effort in evaluating the subassembly response has been centered around the development of appropriate analyses based on the finite element technique. Analysis has been extended to include not only the subassembly duct structure itself, but also the fluid environment, both within subassemblies and between them. These models and codes have been devised to cover a wide range of accident loading conditions, and can treat various materials as their properties become known. The effort described here is centered mainly around an experimental effort aimed at verfying, modifying or extending the models used in treating subassembly damage propagation. To verify the finite element codes under development, a series of out-of-pile room temperature experiments has been performed on LMFBR-type subassembly ducts under various loading conditions. (Auth.)

  6. Materials engineering issues, LMFBR steam generators

    International Nuclear Information System (INIS)

    Spalaris, C.N.; Challenger, K.D.; Day, R.A.; Dutina, D.; Ring, P.J.

    1976-01-01

    Selection of 2-1/4 Cr-1 Mo as the reference construction material for LMFBR steam generators assumed a balance between its known intrinsic properties and our ability to accommodate certain of its deficiencies through design allowance. A comprehensive development program was undertaken to define base data needed, confirm assumptions made relative to desired performance, minimize defects by optimization of melting, fabrication and heat treatment processes, and prepare specifications for purchasing reactor components

  7. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, T.; Mukai, K.; Yamamoto, K.

    1980-01-01

    Bellows are employed as useful mechanical elements with their flexibility and imperviousness to liquid and gas in the system in which such chemically active substances as sodium are handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: control rod drive mechanism; intermediate heat exchanger; small valve; mechanical penetration assembly of the containment boundary; outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for FBR use in Japan. (author)

  8. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, Tadao; Mukai, Kazuo; Yamamoto, Ken.

    1979-11-01

    The bellows is employed as a useful mechanical element with its flexibility and imperviousness to liquid and gas in the system in which such chemically active substance as sodium is handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: - control rod drive mechanism, - intermediate heat exchanger, - small valve, - mechanical penetration assembly of the containment boundary, - outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for the FBR use in Japan. (author)

  9. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  10. Strategies in development of advanced fuels for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo

    1976-12-01

    Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

  11. Issues in the selection of the LMFBR steam cycle

    International Nuclear Information System (INIS)

    Buschman, H.W.; McConnell, R.J.

    1983-01-01

    Unlike the light-water reactor, the liquid-metal fast breeder reactor (LMFBR) allows the designer considerable latitude in the selection of the steam cycle. This latitude in selection has been exercised by both foreign and domestic designers, and thus, despite the fact that over 25 LMFBR's have been built or are under construction, a consensus steam cycle has not yet evolved. This paper discusses the LMFBR steam cycles of interest to the LMFBR designer, reviews which of these cycles have been employed to date, discusses steam-cycle selection factors, discusses why a consensus has not evolved, and finally, concludes that the LMFBR steam-cycle selection is primarily one of technical philosophy with several options available

  12. Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

    International Nuclear Information System (INIS)

    Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

    1982-01-01

    One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

  13. Transient behaviour and inherent safety research of LMFBR power plants

    International Nuclear Information System (INIS)

    Zhu Jizhou; Wang Ping; Yu Baoan

    1995-06-01

    Fast Breeder Reactor will be the next generation reactor for nuclear electricity production, the development of FBR will give the profits of efficient utilization of nuclear resources. The fast reactor safety analysis is the foundation and key of FBR research work. Therefore, a block-oriented mathematical model for the primary system of LMFBRs was constructed, and the dynamic simulating results which have been carried out on micro-computer are presented for various transients, i.e. TOP, LOFS, LOHS. The results agree well with the corresponding results of the code NATDEMO and experiment results of EBR-II. Based on previous analysis, various methods are discussed to confirm the inherent safety of LMFBR

  14. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  15. Benchmark calculation programme concerning typical LMFBR structures

    International Nuclear Information System (INIS)

    Donea, J.; Ferrari, G.; Grossetie, J.C.; Terzaghi, A.

    1982-01-01

    This programme, which is part of a comprehensive activity aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, should allow to get confidence in computer codes which are supposed to provide a realistic prediction of the LMFBR component behaviour. The calculations started on static analysis of typical structures made of non linear materials stressed by cyclic loads. The fluid structure interaction analysis is also being considered. Reasons and details of the different benchmark calculations are described, results obtained are commented and future computational exercise indicated

  16. LMFBR flexible pipe joint development program. Annual technical progress report, government fiscal year 1977

    International Nuclear Information System (INIS)

    1978-01-01

    Currently, the ASME Boiler and Pressure Vessel Code does not allow the use of flexible pipe joints (bellows) in Section III, Class 1 reactor primary piping systems. Studies have shown that the primary piping loops of LMFBR's could be simplified by using these joints. This simplification translates directly into shorter primary piping runs and reduced costs for the primary piping system. Further cost savings result through reduced vault sizes and reduced containment building diameter. In addition, the use of flexible joints localizes the motions from thermally-induced piping growth into components which are specifically designed to accommodate this motion. This reduces the stress levels in the piping system and its components. It is thus economically and structurally important that flexible piping joints be available to the LMFBR designer. The overall objective of the Flexible Joint Program is to provide this availability. This will be accomplished through the development of ASME rules which allow the appropriate use of such joints in Section III, Class 1 piping systems and through the development and demonstration of construction methods which satisfy these rules. The rule development includes analytic and testing methodology formulations which will be supported by subscale bellows testing. The construction development and demonstration encompass the design, fabrication, and in-sodium testing of prototypical LMFBR plant-size flexible pipe joints which meet all ASME rule requirements. The satisfactory completion of these developmental goals will result in an approved flexible pipe joint design for the LMFBR. Progress is summarized in the following efforts undertaken during 1977 to accomplish these goals: (1) code case support, (2) engineering and design, (3) material development, (4) testing, and (5) manufacturing development

  17. Monte-Carlo validation of secondary sodium activation in a pool-type LMFBR

    International Nuclear Information System (INIS)

    Plamiotti, G.; Rado, V.; Salvatores, M.

    1980-09-01

    The secondary sodium activation in a pool-type LMFBR is the main parameter to be accurately evaluated in the shield design. In the present work a complete two dimensional description of the system, including core, shielding and sodium up to Heat Exchangers, is coupled to local Heat Exchanger Monte-Carlo calculations. This refined calculation is used to deduce a simplified method to take into account the coupling of radial propagation in the Heat Exchanger and its finite cylindrical structure

  18. Acoustic detection for water/steam leak from a tube of LMFBR steam generator

    International Nuclear Information System (INIS)

    Sonoda, Masataka; Shindo, Yoshihisa

    1989-01-01

    Acoustic leak detector is useful for detecting more quickly intermediate leak than the existing hydrogen detector and is available for identification of leak location on the accident of water/steam leak from a tube of LMFBR steam generator. This paper presents the overview of HALD (High frequency Acoustics Leak Detection) system, which is more sensitive for leak detection and lower cost of equipment for identification of leak location than a low frequency type detector. (author)

  19. Development of an 85,000 gpm (19,303 m3/h) LMFBR primary pump

    International Nuclear Information System (INIS)

    Zerinvary, M.C.; Wagner, E.W.

    1984-01-01

    The development of an 85,000 gpm two-stage primary pump for liquid metal fast breeder reactor (LMFBR) applications is described. The design was supported by air and cavitation model testing of the hyraulics, and development and feature testing of the level control system and the adjustable frequency solid state power supply. Important fabrication and water test items are also discussed, along with some unique assembly tooling requirements

  20. LMFBR accident delineation study: approach and preliminary results

    International Nuclear Information System (INIS)

    Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

    1979-01-01

    Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

  1. CEC activities in the field of LMFBR safety

    International Nuclear Information System (INIS)

    Balz, W.; Finzi, S.; Klersy, R.

    1976-01-01

    The aim of the ECC is to reach a common LMFBR Safety strategy in Europe. To this end the Commission promotes collaboration between the different fast reactor projects in the Community through working groups and collaborative arrangements and contributes with a research activity executed in its Joint Research Centre Ispra. A short description is given of the activity in the working groups and of the Ispra programme on LMFBR Safety. This programme covers: LMFBR thermohydraulics, fuel coolant interactions, dynamic structure loading and response, safety related material properties and whole core accident code development

  2. Single-phase sodium pump model for LMFBR thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.

    1979-01-01

    A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics

  3. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Aubert, M.; Mathieu, B.; Petrequin, P.

    1983-06-01

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  4. Applicability of the Reactor Safety Study (WASH-1400) to LMFBR risk assessments

    International Nuclear Information System (INIS)

    El-Sheikh, K.A.; Feller, K.G.; Fleischer, L.; Greebler, P.; McDonald, A.; Sultan, P.; Temme, M.I.; Fullwood, R.R.

    1976-01-01

    The feasibility of applying the WASH-1400 methods and data to LMFBR risk assessment is evaluated using the following approach for a selected LMFBR: (1) Structuring the LMFBR risk assessment problem in a modular form similar to WASH-1400; (2) Comparing the predictive tools applicable to each module; (3) Comparing the dependencies among the various modules. It is concluded that the WASH-1400 applicability is limited due to LWR-LMFBR differences in operating environments and accident phenomena. WASH-1400 and LMFBR specific methods applicable to LMFBR risk assessments are indicated

  5. LMFBR subassembly response to local pressure loadings: an experimental approach

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1975-01-01

    An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

  6. Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

    1976-01-01

    A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

  7. LMFBR design and its evolution. (2) Core design of LMFBR

    International Nuclear Information System (INIS)

    Uto, Nariaki; Mizuno, Tomoyasu

    2003-01-01

    Sodium-cooled core design studies are performed. MOX fuel core with axial blanket partial elimination subassembly due to safety consideration is studied. This type of core with high internal conversion ratio possesses capability of achieving 26 months of operation cycle length and 100 GWd/t of burnup averaged over core and blanket, which are superior characteristics in view of reducing cost of power generation. Metal fuel core is also studied, and its higher breeding capability reveals a potential of better core performance such as longer operation cycle length for the same level of electricity generation, though core outlet temperature is limited to lower level due to steel cladding-metal fuel compatibility concerns. Another metal fuel core concept using single Pu enrichment and two radial regions with individual fuel pin diameters achieves 550degC of core outlet temperature identical to that of MOX fuel core, keeping operation cycle length comparable with that of MOX fuel core. This series of study results show that sodium-cooled MOX and metal fuel cores have a high flexibility in satisfying various needs including fuel cycle cost and breeding capability, depending on the stage of introducing commercialized fast reactor cycle system. (author)

  8. LMFBR safety and sodium boiling

    Energy Technology Data Exchange (ETDEWEB)

    Hinkle, W.D.; Tschamper, P.M.; Fontana, M.H.; Henry, R.E.; Padilla, A. Jr.

    1978-01-01

    Within the U.S. Fast Breeder Reactor Safety R and D Work Breakdown Structure for Line of Assurance 2, Limit Core Damage, the influence of sodium boiling upon the progression and termination of accidents is being studied in loss of flow, transient overpower, loss of piping integrity, loss of shutdown heat removal system and local fault situations. The pertinent analytical and experimental results of this research to date are surveyed and compared with the requirements for demonstrating the effectiveness of this line of assurance. A discussion of specific technical issues concerned with sodium boiling and the need for future development work is also presented.

  9. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  10. Work plan: transient release from LMFBR fuel

    International Nuclear Information System (INIS)

    Kress, T.S.; Parker, G.W.; Fontana, M.H.

    1975-09-01

    The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

  11. Acoustic leak detection of LMFBR steam generator

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo

    1993-01-01

    The development of a water leak detector with short response time for LMFBR steam generators is required to prevent the failure propagation caused by the sodium-water reaction and to maintain structural safety in steam generators. The development of an acoustic leak detector assuring short response time has attracted. The purpose of this paper is to confirm the basic detection feasibility of the active acoustic leak detector, and to investigate the leak detection method by erasing the background noise by spectrum analysis of the passive acoustic leak detector. From a comparison of the leak detection sensitivity of the active and the passive method, the active method is not influenced remarkably by the background noise, and it has possibility to detect microleakage with short response time. We anticipate a practical application of the active method in the future. (author)

  12. Low cycle fatigue of irradiated LMFBR materials

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1976-01-01

    A review of low cycle fatigue data on irradiated LMFBR materials was conducted and extensive graphical representations of available data are presented. Representative postirradiation tensile properties of annealed 304 and 316 SS are selected and employed in several predictive methods to estimate irradiated material fatigue curves. Experimental fatigue data confirm the use of predictive methods for establishing conservative design curves over the range of service conditions relevant to such CRBRP components as core former, fixed radial shielding, core barrel, lower inlet module and upper internals structures. New experimental data on fatigue curves and creep-fatigue interaction in irradiated 20 percent cold worked (CW) 316 SS and Alloy 718 would support the design of removable radial shielding and upper internals in CRBRP. New experimental information on notched fatigue behavior and cyclic stress-strain curves of all these materials in the irradiated condition could provide significant design data

  13. LMFBR Ultra Long Life Cores

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Doncals, R.A.; Porter, C.A.; Gundy, L.M.

    1986-01-01

    The Ultra Long Life Core is an attractive and innovative design approach with several extremely beneficial attributes. Long Life cores are applicable to the full range of LMR plant sizes resulting in lifetimes up to 30 years. Core life is somewhat limited for smaller plant sizes, however significant benefits of this approach still exist for all plant sizes. The union of long life cores and the complementary inherent safety technology offer a means of utilizing the well-proven oxide fuel in a system with unsurpassed safety capability. A further benefit is that the uranium fuel cycle can be used in long life cores, especially for initial LMR plant deployment, thereby eliminating the need for reprocessing prior to starting LMR plant construction in the U.S. Finally the long life core significantly reduces power costs. With inherent safety capability designed into an LMR and with the ULLC fuel cycle, power costs competitive with light water plants are achievable while offering improved operational flexibility derived through extending refueling intervals

  14. LMFBR plant design features for sodium spill and fire protection

    International Nuclear Information System (INIS)

    Palm, R.E.

    1982-01-01

    Design features have been developed for an LMFBR plant to protect the concrete structures from potential liquid spills and fires and prevent sodium-concrete reactions. The inclusion of these features in the plant design reduces the severity of design basis accident conditions imposed on containment and other critical plant structures. Steel liners are provided in cells containing radioactive sodium systems, and catch pans are located in non-radioactive sodium system cells. The design requirements and descriptions of each of these protective features are presented. The loading conditions, analytical approach and numerical results are also included. Design of concrete cell structures that are subject to high temperature effects from sodium spills is discussed. The structural design considers the influence of high temperature on design properties of concrete and carbon steel materials based on results of a comprehensive test program. The development of these design features and high temperature design considerations for the Clinch River Breeder Reactor Plant (CRBRP) are presented in this paper

  15. 85,000-GPM, single-stage, single-suction LMFBR intermediate centrifugal pump

    International Nuclear Information System (INIS)

    Fair, C.E.; Cook, M.E.; Huber, K.A.; Rohde, R.

    1983-01-01

    The mechanical and hydraulic design features of the 85,000-gpm, single-stage, single-suction pump test article, which is designed to circulate liquid-sodium coolant in the intermediate heat-transport system of a Large-Scale Liquid Metal Fast Breeder Reactor (LS-LMFBR), are described. The design and analytical considerations used to satisfy the pump performance and operability requirements are presented. The validation of pump hydraulic performance using a hydraulic scale-model pump is discussed, as is the featute test for the mechanical-shaft seal system

  16. Comments on US LMFBR steam generator base technology

    International Nuclear Information System (INIS)

    Simmons, W.R.

    1984-01-01

    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

  17. Hydrogen formation and control under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Armstrong, G.R.; Wierman, R.W.

    1976-09-01

    The objective of this study is to experimentally investigate the potential for autoignition and combustion of hydrogen-sodium mixtures which may be produced in LMFBR accidents. The purpose and ultimate usefulness of this work is to provide data that will establish the validity and acceptability of mechanisms inherent to the LMFBR that could either prevent or delay the accumulation of hydrogen gas to less than 4 percent (V) in the Reactor Containment Building (RCB) under accident conditions. The results to date indicate that sodium and sodium-hydrogen mixtures such as may be expected during LMFBR postulated accidents will ignite upon entering an air atmosphere and that the hydrogen present will be essentially all consumed until such time that the oxygen concentration is depleted

  18. Licensing decisions and safety research related to LMFBR accidents

    International Nuclear Information System (INIS)

    Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

    1977-01-01

    The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

  19. Upper shielding body in LMFBR type reactors

    International Nuclear Information System (INIS)

    Shoji, Koichi.

    1986-01-01

    Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

  20. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  1. Coolant mixing in the LMFBR outlet plenum

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-06-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

  2. Cesium vapor cycle for an advanced LMFBR

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

    A review indicates that a cesium vapor topping cycle appears attractive for use in the intermediate fluid circuit of an advanced LMFBR designed for a reactor outlet temperature of 1250 0 F or more and would have the following advantages: (1) it would increase the thermal efficiency by about 5 to 10 points (from approximately 40 percent to approximately 45 to 50 percent) thus reducing the amount of waste heat rejected to the environment by 15 to 30 percent. (2) the higher thermal efficiency should reduce the overall capital cost of the reactor plant in dollars per kilowatt. (3) the cesium can be distilled out of the intermediate fluid circuit to leave it bone-dry, thus greatly reducing the time and cost of maintenance work (particularly for the steam generator). (4) the large volume and low pressure of the cesium vapor region in the cesium condenser-steam generator greatly reduces the magnitude of pressure fluctuations that might occur in the event of a leak in a steam generator tube, and the characteristics inherent in a condenser make it easy to design for rapid concentration of any noncondensibles that may form as a consequence of a steam leak into the cesium region so that a steam leak can be detected easily in the very early stages of its development

  3. 54Mn release from LMFBR cores

    International Nuclear Information System (INIS)

    Polley, M.V.

    1976-10-01

    The inventory of 54 Mn per unit exposed area of stainless steel in LMFBR cores may be calculated using a formula originally derived at HEDL. This treats the simultaneous production by activation and release by corrosion and diffusion of 54 Mn and assumes that the concentration at the steel surface is zero. The inventory per unit exposed area is calculated as a function of temperature and is compared with that calculated simply by assuming stoichiometric corrosion. An effective diffusion coefficient is used in the calculations which include contributions from both lattice and grain boundary diffusion. A general relationship is derived for the effective diffusion coefficient and it is shown how values may be obtained using the Levine-MacCallum and the Fisher theories of grain boundary diffusion. Values of the lattice diffusion coefficient were obtained by analysing data obtained from sodium loop experiments. The effect on the inventory due to the possible formation of a ferrite layers on the exposed surface is discussed and it is also shown how the inventory over several fuel cycles may be calculated. (U.K.)

  4. Measurements of dynamic shape factors of LMFBR aggregate aerosols

    International Nuclear Information System (INIS)

    Allen, M.D.; Moss, O.R.; Briant, J.K.

    1980-01-01

    Dynamic shape factors for branched, chain-like aggregates of LMFBR mixed-oxide fuels have been measured with a LAPS spiral-duct centrifuge. The aerosol was generated by repeatedly pulsing a focused laser beam onto the surface of a typical LMFBR fuel pellet. The measured values of the dynamic shape factor, corrected for slip, vary between kappa = 3.60 at D/sub ae/ = 0.5 μm, and kappa = 2.23 at D/sub ae/ = 1.5 μm

  5. Safety consequences of local initiating events in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.

  6. Safety consequences of local initiating events in an LMFBR

    International Nuclear Information System (INIS)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered

  7. Applications of simulation experiments in LMFBR core materials technology

    International Nuclear Information System (INIS)

    Appleby, W.K.

    1976-01-01

    The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

  8. Status of gamma-ray heating characterization in LMFBR

    International Nuclear Information System (INIS)

    Gold, R.

    1975-11-01

    Efforts to define gamma-ray heating in Liquid Metal Fast Breeder Reactor (LMFBR) environments have been surveyed. Emphasis is placed on both current practice for the Experimental Breeder Reactor-II (EBR-II) and future needs of the Fast Flux Test Facility (FFTF). Experimental and theoretical work are included in this preliminary survey for both high and low power environments. Current ''state-of-the-art'' accuracies and limitations are assessed. On this basis, it is concluded that a broad and sustained effort be initiated to meet requested FFTF goal accuracies. To this end, recommendations are advanced for improving the current status of gamma heating characterization and temperature measurements in LMFBR

  9. Neutronic characteristics simulation of LMFBR of great size

    International Nuclear Information System (INIS)

    Kim, Y.C.

    1987-09-01

    The CONRAD experimental program to be executed on the critical mockup MASURCA in Cadarache and use all the european plutonium stock. The objectives of this program are to reduce the uncertainties on important project parameters such as the reactivity value of control rods, the flux distribution to valid calcul methods and data to use for new LMFBR conception (heterogeneous axial core by example) and to resolve the neutronic control problems for a LMFBR of great size. The present study has permitted to define this program and its physical characteristics [fr

  10. Safety research needs for carbide and nitride fueled LMFBR's. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1975-01-01

    The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

  11. Comparative analysis of quality assurance requirements for selected LMFBR components of classes 1, 2 and 3

    International Nuclear Information System (INIS)

    Kleinert, K.P.

    1992-01-01

    The study analyses and compares German, French, British and Italian practices and procedures applied for various LMFBR projects both related to the quality assurance system and related to the particular type of class of component:Class 1: primary reactor vessel; Class 2: Secondary sodium pump; Class 3: Primary cold trap. Various areas of analysis and comparison were selected to identify the underlying concepts of grading of requirements and measures, to identify the similarities and differences, and to give recommendations for further actions concerning quality assurance requirements 60 refs., 21 tabs., 6 figs

  12. Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

    International Nuclear Information System (INIS)

    Henslee, S.P.

    1987-03-01

    The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs

  13. Social Interpretation Bias in Children and Adolescents with Anxiety Disorders: Psychometric Examination of the Self-report of Ambiguous Social Situations for Youth (SASSY) Scale.

    Science.gov (United States)

    Gonzalez, Araceli; Rozenman, Michelle; Langley, Audra K; Kendall, Philip C; Ginsburg, Golda S; Compton, Scott; Walkup, John T; Birmaher, Boris; Albano, Anne Marie; Piacentini, John

    2017-06-01

    Anxiety disorders are among the most common mental health problems in youth, and faulty interpretation bias has been positively linked to anxiety severity, even within anxiety-disordered youth. Quick, reliable assessment of interpretation bias may be useful in identifying youth with certain types of anxiety or assessing changes on cognitive bias during intervention. This study examined the factor structure, reliability, and validity of the Self-report of Ambiguous Social Situations for Youth (SASSY) scale, a self-report measure developed to assess interpretation bias in youth. Participants (N=488, age 7 to 17) met diagnostic criteria for Social Phobia, Generalized Anxiety Disorder, and/or Separation Anxiety Disorder. An exploratory factor analysis was performed on baseline data from youth participating in a large randomized clinical trial. Exploratory factor analysis yielded two factors (Accusation/Blame, Social Rejection). The SASSY full scale and Social Rejection factor demonstrated adequate internal consistency, convergent validity with social anxiety, and discriminant validity as evidenced by non-significant correlations with measures of non-social anxiety. Further, the SASSY Social Rejection factor accurately distinguished children and adolescents with Social Phobia from those with other anxiety disorders, supporting its criterion validity, and revealed sensitivity to changes with treatment. Given the relevance to youth with social phobia, pre- and post-intervention data were examined for youth social phobia to test sensitivity to treatment effects; results suggested that SASSY scores reduced for treatment responders. Findings suggest the potential utility of the SASSY Social Rejection factor as a quick, reliable, and efficient way of assessing interpretation bias in anxious youth, particularly as related to social concerns, in research and clinical settings.

  14. Feasibility study for adapting ITREC plant to reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Moccia, A.; Rolandi, G.

    1976-05-01

    The report evaluates the feasibility of adapting ITREC plant to the reprocessing LMFBR fuels, with the double purpose of: 1) recovering valuable Pu contained in these fuels and recycling it to the fabrication plant; 2) trying, on a pilot scale, the chemical process technology to be applied in a future industrial plant for reprocessing the fuel elements discharged from fast breeder power reactors

  15. Technical considerations relative to removal of sodium from LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, J S; Asquith, J G

    1975-07-01

    Reviewed in this paper are technical considerations which are of importance in choosing between an alcohol process and a moist nitrogen process for the removal of sodium from LMFBR components. Results observed in laboratory tests and in the cleaning of large scale components (e.g. a 28 MWt Modular Steam Generator Test Unit) are presented and discussed. (author)

  16. German position paper on structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    Angerbauer, A.; Link, F.

    1983-01-01

    During the design period of the German LMFBR, the SNR-300, extensive work had been done in the field of elastic and inelastic analysis. Furthermore, special design rules have been developed. A review of these activities and their state-of-the art is outlined in this paper

  17. Fission-gas bubble modeling for LMFBR accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1977-01-01

    The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

  18. Small leak shutdown, location, and behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Sandusky, D.W.

    1976-01-01

    The paper summarizes an experimental study of small leaks tested under LMFBR steam generator conditions. Defected tubes were exposed to flowing sodium and steam. The observed behavior of the defected tubes is reported along with test results of shutdown methods. Leak location methods were investigated. Methods were identified to open plugged defects for helium leak testing and detect plugged leaks by nondestructive testing

  19. Thermographic mapping of a complex vernacular settlement: the case study of Casalnuovo District within the Sassi of Matera (Italy)

    Science.gov (United States)

    Cardinale, Tiziana; Balestra, Alessandro; Cardinale, Nicola

    2015-04-01

    Never as in the Sassi of Matera (Italy), the parties have a volumetric material identity and a special construction condition for which, first of all, you need to know the whole to which they give life, and then the individual components and their connections. The distribution of these building artifacts in symbiotic connection with the connective calcareous texture that hosts them, resulted in a spontaneously harmonious figurative balance that characterizes the constructive expedients employed and the distributive and morphological solutions. This is the reason why the Sassi, and the overlooking Park of Rupestrian Churches of Matera Murgia have been entered in 1993 in the UNESCO World Heritage List. Our research focuses on a largely abandoned area within the Casalnuovo District, situated in the extreme south of the city, probably the place of the future Demoethnoanthropological Museum. It is known today that the particular shape of the area is made up of settlements mainly in cave; during the time architectures above ground, in a very limited number, filled almost completely the limited building space. Here there are the most archaic types and forms of building and wine cellars within the Sassi, essentially derive from the natural cave only slightly structured by man. The exterior construction complete the elementary cave and it is called lamione. The complexity of a built space like this one determines the need for a non-traditional approach, so you have to combine last generation tools and canonical ones for survey and energy diagnosis within a dialectic between memory, tradition and innovation, in order to identify solutions for an environment friendly recover of a cultural heritage such this one. Since the evaluation of the historical buildings conservation state using destructive techniques should be avoided to prevent the integrity of the cultural heritage, the development of non-destructive and non-contact techniques is very important. For this reason, an

  20. Energy and microclimatic performance of restored hypogeous buildings in south Italy: The ''Sassi'' district of Matera

    Energy Technology Data Exchange (ETDEWEB)

    Cardinale, Nicola [Department of Engineering and Environmental Physics, University of Basilicata, Via Lazazzera, 75100 Matera (Italy); Rospi, Gianluca; Stazi, Alessandro [Department of Architecture, Constructions and Structures, Polytechnic University of Marche, Via Brecce Bianche, 60131 Ancona (Italy)

    2010-01-15

    The site of the ''Sassi of Matera'' classified by the United Nations Educational, Scientific and Cultural Organization (UNESCO) as World Heritage in 1993, is an exceptional example of traditional bioclimatic Mediterranean architecture. Within this immense artistic heritage, we find hypogea habitations, stone buildings, and mixed habitation - half-hypogeum and half-built. In this study, we analyze the energy and microclimatic performance of hypogeous structures in three states: not-restored, immediately after restoration, and a few years after restoration (in normal use). We monitored a surface hypogeum and a deep hypogeum. We performed a dynamic parametric simulation using the software EnergyPlus to quantify the energetic balance of the hypogeous structures during one calendar year. The energetic valuation of the surface hypogea shows that these environments, once restored and in a condition of normal use, give indoor comfort within the limits of comfort thermo-hygrometrics established by the comfort indices of predicted mean vote (PMV) and predicted percentage of dissatisfied (PPD). The huge thermal mass of the walls ensures that the microclimate indoor conditions are regular throughout the seasons, without differences in the daily thermal oscillation. Deep hypogea without an air change system cannot reach thermal-hygrometric comfort values. We determined that these structures have a null thermal balance during mid-season, while in the summer the floor loses heat, thereby cooling the environment. The opposite occurs in winter. We can conclude that these buildings were designed as bioclimatic. In fact they can be used, after restoration, with limited use of technology systems. (author)

  1. Power DRAC for rapid LMFBR deployment and consequent CO2 mitigation

    International Nuclear Information System (INIS)

    Schenewerk, W.E.

    2006-01-01

    A metallic-sodium LMFBR (Liquid Metal Fast Breeder Reactor) can control fuel temperature after a full power SCRAM using natural convection. A 3 percent nominal DRAC (Direct Reactor Auxiliary Cooling) does this without moving parts. DRAC is promoted from tertiary to primary decay heat removal, resulting in what is referred to as a Power DRAC. Power DRAC operates continuously before and after SCRAM, rejecting 3 per cent pile power. Power DRAC operability is validated by having it reject 75 MWt from a 2500 MWt pile at all times. IHX (Intermediate Heat Exchanger) is not required to be operable for primary, secondary, or tertiary core over temperature protection. Original DRAC concept (venturi DRAC) was a 1 per cent nominal tertiary decay heat removal system. Tertiary DRAC patent has expired. Power DRAC rejects 75 MWt through its own secondary sodium heat transfer loop to power a 25 MWe air Brayton cycle. Power DRAC eliminates requiring steam plant operability for decay heat removal. Intermediate sodium heat transfer system and steam plant can be optimized for maximum thermal efficiency. 2.5 GWt pile makes 1.0 GWe net power. Power DRAC maintains pile inlet and outlet temperatures while going from power to post-SCRAM conditions. Steam pressure is maintained post-SCRAM to mitigate SCRAM thermal transient. Not requiring steam plant operability for decay heat removal eases licensing and allows early LMFBR deployment. Each GWe atomic power delays Co2 doubling one week. (author)

  2. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

    Anderson, R.; Hayden, O.

    2002-01-01

    Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

  3. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  4. The Integrated Survey for Excavated Architectures: the Complex of Casalnuovo District Within the World Heritage Site "sassi" (matera, Italy)

    Science.gov (United States)

    Cardinale, T.; Valva, R.; Lucarelli, M.

    2015-02-01

    Never as within the complex of Sassi (Matera, South of Italy), the parties have a volumetric material identity and a special construction condition for which, first of all, you need to know the whole to which they give life, and then the individual components and their connections. In the course of time, in the Lucan city, there were stable and favorable conditions that allowed the development of an architectural language, of juxtaposition of the materials, interpenetration of space and conformation of the volumes, which generated an exceptional urban phenomenon. The distribution of these building artifacts in symbiotic connection with the connective calcareous texture that hosts them , resulted in a spontaneously harmonious figurative balance that characterizes the constructive expedients employed and the distributive and morphological solutions. This is the reason why the Sassi, and the overlooking Park of Rupestrian Churches of Matera Murgia, have been entered in 1993 in the UNESCO World Heritage List. The complexity of a built space, such as this one, determines the need for a non-traditional approach, so you have to combine last generation tools and canonical ones for survey, drawing and representation, within a dialectic between memory and design, tradition and innovation. For this reason, an appropriate cognitive apparatus has been set up for the entire technical process, making use of different non-destructive and non-contact techniques: digital photogrammetry, total station, laser scanner and thermography, in order to obtain a three-dimensional computer model, useful for the diagnosis and the preservation of the integrity of cultural heritage.

  5. A study on reactor core failure thresholds to safety operation of LMFBR

    International Nuclear Information System (INIS)

    Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

    2006-01-01

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  6. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

    International Nuclear Information System (INIS)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1977-01-01

    Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

  7. Water tests for determining post voiding behavior in the LMFBR

    International Nuclear Information System (INIS)

    Hinkle, W.D.

    1976-06-01

    The most serious of the postulated accidents considered in the design of the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) is the Loss of Pipe Integrity (LOPI) accident. Analysis models used to calculate the consequences of this accident assume that once boiling is initiated film dryout occurs in the hot assembly as a result of rapid vapor bubble growth and consequent flow stoppage or reversal. However, this assumption has not been put to any real test. Once boiling is initiated in the hot assembly during an LMFBR LOPI accident, a substantial gravity pressure difference would exist between this assembly and other colder assemblies in the core. This condition would give rise to natural circulation flow boiling accompanied by pressure and flow oscillations. It is possible that such oscillations could prevent or delay dryout and provide substantial post-voiding heat removal. The tests described were conceived with the objective of obtaining basic information and data relating to this possibility

  8. Hydrogen jet recombination under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Wierman, R.W.

    1977-01-01

    Certain conditions may be postulated in LMFBR risk assessments for which the potential of hydrogen release to the reactor containment building needs to be evaluated. The inherent self-ignition characteristics of hydrogen jets entering the air atmosphere of the reactor containment building should be understood for such analyses. If hydrogen jets were to self-ignite (recombine) at the source where they enter the reactor containment building, then undesirable hydrogen accumulation would not occur. Therefore, experiments have been conducted investigating the phenomena associated with the recombination of hydrogen jets under conditions similar to those postulated for LMFBR studies. The data presented define the conditions required for self-ignition of the hydrogen jets

  9. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

    1971-01-01

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  10. Design and economic implications of heterogeneity in an LMFBR core

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1983-01-01

    Much emphasis is currently being placed in LMFBR design on reducing both the capital cost and the fuel cycle cost of an LMFBR to insure its economic competativeness without a rapid increase in the uranium prices. In this study the relationship between two core design options, their neutronic consequences, and their effect on fuel cycle cost are analyzed. The two design options are the selection of pin diameter and the degree of heterogeneity. In the case of a heterogeneous core, with a low sodium void reactivity worth this ratio of fertile internal blanket to driver assemblies is generally about 0.40. However, some advantages of cores with heterogeneity of 0.08 to 0.2 for a fixed pin diameter have been reported

  11. Cover-gas seals: 11-LMFBR seal-test program

    International Nuclear Information System (INIS)

    Steele, O.P. III; Horton, P.H.

    1977-01-01

    The objective of the Cover Gas Seal Material Development Program is to perform the engineering development required to provide reliable seals for LMFBR application. Specific objectives are to verify the performance of commercial solid cross-section and inflatable seals under reactor environments including radiation, to develop advanced materials and configurations capable of achieving significant improvement in radioactive gas containment and seal temperature capabilities, and to optimize seal geometry for maximum reliability and minimal gas permeation

  12. Axial migratin of cesium in LMFBR fuel pins

    International Nuclear Information System (INIS)

    Karnesky, R.A.; Bridges, A.E.; Jost, J.W.

    1981-11-01

    A correlated model for quantitatively predicting the behavior of cesium in LMFBR fuel pins has been developed. This correlation was shown to be in good agreement with experimental data. It has been used to predict the behavior of cesium in the FFTF driver fuel and as the result of this analysis it has been shown that the accumulation of cesium in the insulator pellets at the ends of the fuel column will not be life limiting

  13. Retention of gaseous fission products in reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Burch, W.D.; Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-05-01

    The report is devoted to status of the development programme at the Oak Ridge National Laboratory on methods for retaining iodine-131 and 129, Krypton-85, Tritium and Carbon-14 in reprocessing LMFBR fuels. The Iodox process, Fluorocarbon absorption process and Voloxidation process are described for retention of iodine, Krypton-85 and Tritium, respectively. Flowsheets for the different processes are given and results of experimental runs in small engineering-scale equipment are reported

  14. Future development LMFBR-steam generators SNR2

    International Nuclear Information System (INIS)

    Essebaggers, J.; Pors, J.G.

    1975-01-01

    The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

  15. Impact of LMFBR operating experience on PFBR design

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chetal, S.C.; Chellapandi, P.; Govindarajan, S.; Lee, S.M.; Kameswara Rao, A.S.L.; Prabhakar, R.; Raghupathy, S.; Sodhi, B.S.; Sundaramoorthy, T.R.; Vaidyanathan, G.

    2000-01-01

    PFBR is a 500 MWe, sodium cooled, pool type, fast breeder reactor currently under detailed design. It is essential to reduce the capital cost of PFBR in order to make it competitive with thermal reactors. Operating experience of LMFBRs provides a vital input towards simplification of the design, improving its reliability, enhancing safety and achieving overall cost reduction. This paper includes a summary of LMFBR operating experience and details the design features of PFBR as influenced by operating experience of LMFBRs. (author)

  16. Thermal analysis methods for LMFBR wire wrapped bundles

    International Nuclear Information System (INIS)

    Todreas, N.E.

    1976-11-01

    A note is presented which was written to stimulate an awareness and discussion of the fundamental differences in the formulation of certain existing analysis codes for LMFBR wire wrap bundles. The contention of the note is that for those array types where data exists (one wire per pin, equal start angles), the ENERGY method results for coolant temperature under forced convection conditions provide benchmarks of reliability equal to the results of codes COBRA and TH1-3D

  17. Structural analysis for elevated temperature design of the LMFBR

    International Nuclear Information System (INIS)

    Griffin, D.S.

    1976-02-01

    In the structural design of LMFBR components for elevated temperature service it is necessary to take account of the time-dependent, creep behavior of materials. The accommodation of creep to assure design reliability has required (1) development of new design limits and criteria, (2) development of more detailed representations of material behavior, and (3) application of the most advanced analysis techniques. These developments are summarized and examples are given to illustrate the current state of technology in elevated temperature design

  18. Analytical work on local faults in LMFBR subassembly

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Miyaguchi, K.; Hirata, N.; Kasahara, F.

    1979-01-01

    Analytical codes have been developed for evaluating various severe but highly unlikely events of local faults in the LMFBR subassembly (S/A). These include: (1) local flow blockage, (2) two-phase thermohydraulics under fission gas release, and (3) inter-S/A failure propagation. A simple inter-S/A thermal failure propagation analysis code, FUMES, is described that allows an easy parametric study of propagation potential of fuel fog in a S/A. 7 refs

  19. A new approach to the design of LMFBR liners

    International Nuclear Information System (INIS)

    Polentz, L.M.

    1980-01-01

    An advance in the state-of-the-art of LMFBR liners which permits notable savings in construction costs without any sacrifice of safety is described. The application of the new design concept to the rework of the upper reactor vault liner of the FFTF is discussed. Factors which affect the application of the new design approach to other LMFBRs are delineated and discussed. (author)

  20. Effect of operating temperature on LMFBR core performance

    International Nuclear Information System (INIS)

    Noyes, R.C.; Bergeron, R.J.; di Lauro, G.F.; Kulwich, M.R.; Stuteville, D.W.

    1977-01-01

    The purpose of the study is to provide an engineering evaluation of high and low temperature LMFBR core designs. The study was conducted by C-E supported by HEDL expertise in the areas of materials behavior, fuel performance and fabrication/fuel cycle cost. The evaluation is based primarily on designs and analyses prepared by AI, GE and WARD during Phase I of the PLBR studies

  1. Route survey for LMFBR spent fuel transportation analysis

    International Nuclear Information System (INIS)

    Foley, J.T.

    1977-05-01

    Descriptions are given of surveys that were made along segments of interstate highways to obtain information on objects near the right-of-ways and on highway features that constitute hazards in the event of transportation accidents. Data collected during the surveys are summarized. The work was done in support of the LMFBR Hazards Analysis which was being performed for the Division of Reactor Development and Demonstration of the U.S. Energy Research and Development Administration

  2. Shielding design method for LMFBR validation on the Phenix factor

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

    1983-05-01

    Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

  3. Biological behavior of mixed LMFBR-fuel-sodium aerosols

    International Nuclear Information System (INIS)

    Mahlum, D.D.; Hackett, P.L.; Hess, J.O.; Allen, M.D.

    1979-01-01

    Immediately after exposure of rats to mixed aerosols of sodium-LMFBR fuel, about 80 to 90% of the body burden of 239 Pu is in the gastrointestinal tract; 1.5 to 4% is in the lungs. With fuel-only aerosols, less of the body burden was in the GI tract and more in the lung and the head. Blood and urine values suggest an increased absorption of 239 Pu from sodium-fuel than from fuel-only aerosols

  4. Transport-diffusion comparisons for small core LMFBR disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1977-11-01

    A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for homogeneous cores and to the ZPPR-7 configurations for heterogeneous core. In all TUC cases diffusion theory was shown to be inadequate for the calculation of reactivity changes during core disassembly

  5. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  6. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  7. Structural consideration for hot and cold pipe clamps in LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.; Huang, S.N.; Kappauf, H.; Wagner, S.E.; Wirtz, K.H.

    1983-01-01

    A series of analytical studies are described which evaluate stress levels induced in a 600 mm high temperature, thin-wall sodium pipeline by two alternate clamp designs. The first design consists of a band mounted directly on the pipe and is called the hot clamp. The second design consists of a band mounted using insulation standoffs and is called the cold clamp. Pipe stress levels induced by transient thermal dead weight and seismic loads are discussed. Pipe stress levels and system dynamic spring rates are presented. Procedures utilized to combine clamp induced pipe stress with other short and long term pipe system stresses are detailed. Recommendations for practical application in LMFBR pipe systems are made

  8. Structural considerations for hot and cold pipe clamps in LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.; Huang, S.N.; Wagner, S.E.; Kappauf, H.; Wirtz, K.H.

    1983-01-01

    A series of analytical studies are described which evaluate stress levels induced in a 600 mm high temperature, thin-wall sodium pipeline by two alternate clamp designs. The first design consists of a band mounted directly on the pipe and is called the hot clamp. The second design consists of a band mounted using insulation standoffs and is called the cold clamp. Pipe stress levels induced by transient thermal dead weight and seismic loads are discussed. Pipe stress levels and system dynamic spring rates are presented. Procedures utilized to combine clamp induced pipe stress with other short and long term pipe system stresses are detailed. Recommendations for practical application in LMFBR pipe systems are made

  9. Safety evaluation for the LMFBR plant using probabilistic risk assessment techniques

    International Nuclear Information System (INIS)

    Kani, Y.; Aizawa, K.

    1987-01-01

    This paper presents an application of probabilistic risk assessment techniques to a typical loop-type liquid metal fast breeder reactor (LMFBR) plant in the detailed design stage. A comprehensive systems analysis has been performed to identify event sequences leading to core damage and provide insights into the importance of accident contributors. While traditional event tree/fault tree modeling was used for the analysis, this study involved a thorough investigation of initiating events and of support system faults. The qualification of accident sequences has been conducted by combining the fault trees based on the event trees and obtaining sequence cut sets with the use of the SETS code. This study also attempted to quantify the potential for operator recovery actions in the course of each accident sequence. (author)

  10. Safety criteria for the future LMFBR's in France and main safety issues for the rapide 1500 project

    International Nuclear Information System (INIS)

    Justin, F.; Natta, M.; Orzoni, G.

    1985-04-01

    The main safety criteria for future LMFBR in France and the related issues for the RAPIDE 1500 project are presented and discussed. The evolutions with respect to SUPERPHENIX options and requirements are emphasized, in particular for the concerns of the prevention of core melt accidents, fuel damage limits and related required performances of the protection system, since one main option is not to consider whole core melt accidents in the containment design. One shall also point out the advantages of some mitigating features which were nevertheless added in the containment design, although without any explicit consideration for core melt accidents

  11. The experiment study of the thermal insulation of the roof-slab of the main vessel of a LMFBR

    International Nuclear Information System (INIS)

    Wang Zhifeng; Wang Zhou; Yang Xianyong

    1995-01-01

    The effects of composition of insulation, i.e., reflective multi-plate thermal insulator, protecting the roof-slab of the vessel of the LMFBR on the heat transfer performance has been studied experimentally for CEFR. A economical form of the thermal insulation is suggested for CEFR. In addition, the scheme without reflective thermal insulator which has only a forced convection cooling system has been studied experimentally and a formula to calculate the average Nusselt number of the flow channel, which is valuable for CEFR design, has been raised

  12. LMFBR models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1981-10-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given

  13. Evaluation of integrally finned cladding for LMFBR fuel pins

    International Nuclear Information System (INIS)

    Cantley, D.A.; Sutherland, W.H.

    1975-01-01

    An integral fin design effectively reduces the coolant temperature gradients within an LMFBR subassembly by redistributing coolant flow so as to reduce the maximum cladding temperature and increase the duct wall temperature. The reduced cladding temperatures are offset by strain concentrations resulting from the fin geometry, so there is little net effect on predicted fuel pin performance. The increased duct wall temperatures, however, significantly reduce the duct design lifetime so that the final conclusion is that the integral fin design is inferior to the standard wire wrap design. This result, however, is dependent upon the material correlations used. Advanced alloys with improved irradiation properties could alter this conclusion

  14. Fission and corrosion products behavior in primary circuits of LMFBR's

    International Nuclear Information System (INIS)

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  15. Sodium mists behavior in cover gas space of an LMFBR

    International Nuclear Information System (INIS)

    Himeno, Y.; Takahashi, J.

    1978-03-01

    This paper present the sodium mist behavior in Argon cover gas space of an LMFBR experimentally using a test vessel of 1,400 mm in axial length, 305.5 mm in inner diameter and about 100 l in volume. Experiments are consisted with measurements of the mist concentration and the mist gravitational settling flux between the sodium pool temperature range of 290 0 to 520 0 C. The results are discussed under the monosize assumption of the particles, and the particle sizes and evaporation rate are derived. Transient and steady state mist concentration behavior were also investigated. (author)

  16. LMFBR fuel cycle studies progress report, August 1972, No. 42

    International Nuclear Information System (INIS)

    Unger, W.E.; Blanco, R.E.; Crouse, D.J.; Irvine, A.R.; Watson, C.D.

    1972-10-01

    This report continues a series outlining progress in the development of methods for reprocessing of LMFBR fuels. Development work is reported on problems of irradiated fuel transport to the processing facility, the dissolution of the fuel and the chemical recovery of PuO 2 --UO 2 values, the containment of volatile fission products, product purification, conversion of fuel processing plant product nitrate solutions to solids suitable for shipping and for subsequent fuel fabrication. Pertinent experimental results are presented for the information of those immediately concerned with the field. Detailed description of experimental work and data are included in the topical reports and in the Chemical Technology Division Annual Reports

  17. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  18. Feasibility study on large pool-type LMFBR

    International Nuclear Information System (INIS)

    1984-01-01

    A feasibility study has been conducted from 1981 FY to 1983 FY, in order to evaluate the feasibility of a large pool-type LMFBR under the Japanese seismic design condition and safety design condition, etc. This study was aimed to establish an original reactor structure concept which meets those design conditions especially required in Japan. In the first year, preceding design concepts had been reviewed and several concepts were originated to be suitable to Japan. For typical two of them being selected by preliminary analysis, test programs were planned. In the second year, more than twenty tests with basic models had been conducted under severe conditions, concurrently analytical approaches were promoted. In the last year, larger model tests were conducted and analytical methods have been verified concerning hydrodynamic effects on structure vibration, thermo-hydraulic behaviours in reactor plena and so on. Finally the reactor structure concepts for a large pool-type LMFBR have been acknowledged to be feasible in Japan. (author)

  19. Proposal for computer investigation of LMFBR core meltdown accidents

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

    1974-01-01

    The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

  20. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    Broc, Daniel

    2001-01-01

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  1. LMFBR core flowering response to an impulse load

    International Nuclear Information System (INIS)

    Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

    1993-01-01

    Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

  2. Transient analysis of LMFBR reinforced/prestressed concrete containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Belytschko, T.B.; Bazant, Z.P.

    1979-01-01

    The use of prestressed concrete reactor vessels (PCRVs) for LMFBR containment creates a need for analytical methods for treating the transient response of such structures, for LMFBR containments must be capable of sustaining the dynamic effects which arise in a hypothetical core disruptive accident (HCDA). These analyses require several unique features: a model of concrete which includes tensile cracking, a methodology for representing the prestressing tendons and for simulating the prestressing operation, and an efficient computational tool for treating the transient response. Furthermore, for the sake of convenience, all of these features should be available in a single computer code. For the purpose of treating the transient response, a finite element program with explicit time integration was chosen. The use of explicit time integration has the advantage that it can easily treat the complicated constitutive model which arises from the considerations of concrete cracking and it can handle the slip between reinforcing tendons and the concrete through the use of the well known sliding interface options. However, explicit time integration programs are usually not well suited to the simulation of static processes such as prestressing. Nevertheless, explicit time integration programs can handle static processes through the introduction of damping by what is known as a dynamic relaxation procedure. For this reason, the dynamic relaxation procedure was refined through the introduction of lumped mass, viscous damping. This provision made the prestressing operation of the concrete structures by means of the explicit formulation rather convenient. (orig.)

  3. New approach to the design of core support structures for large LMFBR plants

    International Nuclear Information System (INIS)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions

  4. State of the art review of degradation processes in LMFBR materials. Volume II. Corrosion behavior

    International Nuclear Information System (INIS)

    Dillon, R.D.

    1975-01-01

    Degradation of materials exposed to Na in LMFBR service is reviewed. The degradation processes are discussed in sections on corrosion and mass transfer, erosion, wear and self welding, sodium--water reactions, and external corrosion. (JRD)

  5. Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

    International Nuclear Information System (INIS)

    Moreira, M.L.

    1985-01-01

    A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

  6. LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Dhir, V.K.; Frank, M.; Kastenberg, W.E.; McKone, T.E.

    1979-03-01

    Three aspects of LMFBR safety are discussed. The first concerns the potential reactivity effects of whole core fuel motion prior to pin failure in low ramp rate transient overpower accidents. The second concerns the effects of flow blockages following pin failure on the coolability of a core following an unprotected overpower transient. The third aspect concerns the safety related implications of using thorium based fuels in LMFBR's

  7. Natural circulation in simulated LMFBR fuel assemblies

    International Nuclear Information System (INIS)

    Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

    1985-01-01

    Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation

  8. Argon cover gas purity control on LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Maeda, Hiroshi; Kobayashi, Takayoshi [PNC (Japan); Ishiyama, Satoshi [Toshiba (Japan); Motonaga, Tetsuji [Hitachi (Japan)

    1987-07-01

    Various control methods on chemical impurities and radioactive materials (fission products) in the primary argon gas of LMFBRs' have been studied based on experiences in Joyo and results of research and development. These results are reflected on MONJU design. On-line gas chromatographs are installed both in the Primary and in the Secondary Argon Gas Systems in JOYO. Also, chemical analysis has been done by batch sampling in JOYO. Though the rise of impurity concentration had been measured after periodical fuel exchange operation, impurity concentration has been controlled sufficiently under target control limits. In MONJU detailed design, the Rare Gas Removal and Recovery System which consisted of cryogenic distillation equipment had been eliminated and the capacity of Charcoal Beds in the Primary Argon Gas System has been improved to keep the concentration of radioactive materials sufficient low levels. The necessity to control the impurities in fresh argon gas which is supplied to the Primary Argon Gas System is now considered to keep the concentration of Kr and Xe isotopes in specified level, because their isotopes may make background rise for the Tagging Gas Failed Fuel Detection and Location System. Based on various investigations performed on sodium vapor trapping to obtain its detailed characteristics, design specifications and operating conditions of MONJU's Vapor Traps have been decided. To keep the level of radioactivity in gaseous effluents to the environment as low as reasonably achievable, the following means are now adopted in MONJU: the Primary Argon Gas System is composed of a closed recirculating path, but the exhaust gas discharged has different path after the Charcoal Beds; fresh argon gas is blown down to prevent Primary Argon Gas from releasing to the circumference during opening of the primary argon gas boundary, such as fuel exchange operations. (author)

  9. Maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

    Verriere, P.; Alanche, J.; Minguet, J.L.

    1984-06-01

    After some general remarks on the French fast neutron system, this paper presents the state of the program for the construction of fast reactor in France. Then, the general design of Super Phenix 1 steam generator components is outlined and, the in-service monitoring systems and protective devices with which they are equiped are briefly described. The methods used, in the event of leakage, for leak location, steam generator inspection, steam generator repair and putting the affected loop back into service, are discussed. There are two main lines of research, relating respectively to the means of water leak detection in sodium and the inspection arrangements that will be used either periodically, or following a sodium-water reaction. Finally, after a brief description of the steam generator, this paper describes the four incidents (leaks) that occurred on the Phenix steam generator in the course of 1982 and 1983, and the subsequent repair operations

  10. Study on the seismic response of reactor vessel of pool type LMFBR including fluid-structure interaction

    International Nuclear Information System (INIS)

    Tanimoto, K.; Ito, T.; Fujita, K.; Kurihara, C.; Sawada, Y.; Sakurai, A.

    1988-01-01

    The paper presents the seismic response of reactor vessel of pool type LMFBR with fluid-structure interaction. The reactor vessel has bottom support arrangement, the same core support system as Super-Phenix in France. Due to the bottom support arrangement, the level of core support is lower than that of the side support arrangement. So, in this reactor vessel, the displacement of the core top tends to increase because of the core's rocking. In this study, we investigated the vibration and seismic response characteristics of the reactor vessel. Therefore, the seismic experiments were carried out using one-eighth scale model and the seismic response including FSI and sloshing were investigated. From this study, the effect of liquid on the vibration characteristics and the seismic response characteristics of reactor vessel were clarified and sloshing characteristics were also clarified. It was confirmed that FEM analysis with FSI can reproduce the seismic behavior of the reactor vessel and is applicable to seismic design of the pool type LMFBR with bottom support arrangement. (author). 5 refs, 14 figs, 2 tabs

  11. The frequency-dependent elements in the code SASSI: A bridge between civil engineers and the soil-structure interaction specialists

    International Nuclear Information System (INIS)

    Tyapin, Alexander

    2007-01-01

    After four decades of the intensive studies of the soil-structure interaction (SSI) effects in the field of the NPP seismic analysis there is a certain gap between the SSI specialists and civil engineers. The results obtained using the advanced SSI codes like SASSI are often rather far from the results obtained using general codes (though match the experimental and field data). The reasons for the discrepancies are not clear because none of the parties can recall the results of the 'other party' and investigate the influence of various factors causing the difference step by step. As a result, civil engineers neither feel the SSI effects, nor control them. The author believes that the SSI specialists should do the first step forward (a) recalling 'viscous' damping in the structures versus the 'material' one and (b) convoluting all the SSI wave effects into the format of 'soil springs and dashpots', more or less clear for civil engineers. The tool for both tasks could be a special finite element with frequency-dependent stiffness developed by the author for the code SASSI. This element can represent both soil and structure in the SSI model and help to split various factors influencing seismic response. In the paper the theory and some practical issues concerning the new element are presented

  12. Energetics of LMFBR core disruptive accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.

    1979-01-01

    In general, in the design of fast reactor systems, containment design margins are specified by investigating the response of the containment to core disruptive accidents. The results of these analyses are then translated into criteria which the designers must meet. Currently, uniform and agreed upon criteria are lacking, and in this time while they are being developed, the designer should be aware of the considerations which go into the particular criteria he must work with, and participate in their development. This paper gives an overview of the current state of the art in assessing core disruptive accidents and the design implications of this process. (orig.)

  13. French LMFBR's control rods experience and development

    International Nuclear Information System (INIS)

    Arnaud, G.; Guigon, A.; Verset, L.

    1983-06-01

    Since the last ten years, the French program has been, first of all, directed to the setting up, and then the development of, at once, the Phenix control rods, and next, the Super-Phenix ones. The vented pin design, with porous plug and sodium bonding, which allows the choices of large diameters, has been taken, since the Rapsodie experience was decisive. The absorber material is sintered, 10 B enriched, boron carbide. The can is made of 316 type stainless steel, stabilised, or not, with titanium. The experience gained in Phenix up to now is important, and deals with about six loads of control rods. Results confirm the validity of the design of the absorber pins. Some difficulties has been encountered for the guiding devices, due to the swelling of the steel. They have required design and material improvements. Such difficulties are discarded by a new design of the bearing, for the Super-Phenix control rods. The other parts of these rods, from the Primary Shut-Down System, are strictly derived from Phenix. The design of the rods from the Secondary Shut-Down System is rather different, but it's not the case for the design of the absorber pins: in many a way, they are derived from Phenix pins and from Rapsodie control rods. Both types of rods irradiation tests are in progress in Phenix [fr

  14. Development of a simple estimation tool for LMFBR construction cost

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Kinoshita, Izumi

    1999-01-01

    A simple tool for estimating the construction costs of liquid-metal-cooled fast breeder reactors (LMFBRs), 'Simple Cost' was developed in this study. Simple Cost is based on a new estimation formula that can reduce the amount of design data required to estimate construction costs. Consequently, Simple cost can be used to estimate the construction costs of innovative LMFBR concepts for which detailed design has not been carried out. The results of test calculation show that Simple Cost provides cost estimations equivalent to those obtained with conventional methods within the range of plant power from 325 to 1500 MWe. Sensitivity analyses for typical design parameters were conducted using Simple Cost. The effects of four major parameters - reactor vessel diameter, core outlet temperature, sodium handling area and number of secondary loops - on the construction costs of LMFBRs were evaluated quantitatively. The results show that the reduction of sodium handling area is particularly effective in reducing construction costs. (author)

  15. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

  16. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  17. A technique for computing bowing reactivity feedback in LMFBR's

    International Nuclear Information System (INIS)

    Finck, P.J.

    1987-01-01

    During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

  18. CAT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Betten, P.R.; Tow, D.M.

    1984-04-01

    A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

  19. Validation of turbulence models for LMFBR outlet plenum flows

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-01-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds number (Re) values of 33000 and 70000 in a 1/15 - scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different two-equation turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet flow field, importantly also upon the degree of inlet turbulence, and also upon the turbulent momentum exchange model used in the calculations. In the FFTF geometry, the TEACH-T predictions agree well with the experiments. 7 refs

  20. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800 0 C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800 0 C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800 0 C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (Δanti G 02 congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800 0 C

  1. Fatigue of LMFBR piping due to flow stratification

    Energy Technology Data Exchange (ETDEWEB)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface.

  2. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. (Auth.)

  3. Simple LMFBR axial-flow friction-factor correlation

    International Nuclear Information System (INIS)

    Chan, Y.N.; Todreas, N.E.

    1981-09-01

    Complicated LMFBR axial lead-length averaged friction factor correlations are reduced to an easy, ready-to-use function of bundle Reyonlds number for wire-wrapped bundles. The function together with the power curves to calculate the associated constants are incorporated in a computer pre-processor, EZFRIC. The constants required for the calculation of the subchannels and bundle friction factors are derived and correlated into power curves of geometrical parameters. A computer program, FRIC, which can alternatively be used to accurately calculate these constants is also included. The accuracte values of the constants and the corresponding values predicted by the power curves and percentage error of prediction are tabulated for a wide variety of geometries of interest

  4. Fatigue of LMFBR piping due to flow stratification

    International Nuclear Information System (INIS)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface

  5. Gravitational agglomeration of post-HCDA LMFBR aerosols: nonspherical particles

    International Nuclear Information System (INIS)

    Tuttle, R.F.; Loyalka, S.K.

    1982-12-01

    Aerosol behavior analysis computer programs have shown that temporal aerosol size distributions in nuclear reactor containments are sensitive to shape factors. This research investigates shape factors by a detailed theoretical analysis of hydrodynamic interactions between a nonspherical particle and a spherical particle undergoing gravitational collisions in an LMFBR environment. First, basic definitions and expressions for settling speeds and collisional efficiencies of nonspherical particles are developed. These are then related to corresponding quantities for spherical particles through shape factors. Using volume equivalent diameter as the defining length in the gravitational collision kernel, the aerodynamic shape factor, the density correction factor, and the gravitational collision shape factor, are introduced to describe the collision kernel for collisions between aerosol agglomerates. The Navier-Stokes equation in oblate spheroidal coordinates is solved to model a nonspherical particle and then the dynamic equations for two particle motions are developed. A computer program (NGCEFF) is constructed, and the dynamical equations are solved by Gear's method

  6. Creep strain accumulation in a typical LMFBR piperun

    International Nuclear Information System (INIS)

    Johnstone, T.L.

    1975-01-01

    The analysis described allows the strain concentrations in typical LMFBR two anchor point uniplanar piperuns to be calculated. Account is taken of the effect of pipe elbows in attracting creep strain to themselves as well as possible movements of the thrust line due to strain redistribution. The influence of the initial load conditions is also examined. The stress relaxation analysis is facilitated by making the assumption that a cross-sectional stress distribution determined by the asymptotic fully developed state of creep exists at all times. Use is then made of Hoff(s) analogy between materials with a creep law of the Norton type and those with a corresponding non-linear elastic stress strain law, to determine complementary strain energy rates for straight pipes and bends. Ovalisation of the latter produces an increased strain energy rate which can be simply calculated by comparison with an equal length of straight pipe through employing a creep flexibility factor due to Spence. Deflection rates at any location in the pipework can then be evaluated in terms of the thermal restraint forces at that location by an application of Castigliano's principle. In particular for an anchor point the deflection rates are identically zero and this leads to the generation of 3 simultaneous differential equations determining the relaxation of the anchor reactions. Indicative results are presented for the continuous relaxation at 570 deg C of the thermally induced stress in a planar approximation to a typical LMFBR pipe run chosen to have peak elbow stresses close to the code maximum. The results indicate a ratio, after 10 5 hours, of 3 for creep strain concentration relative to initial peak strain (calculated on the assumption of fully elastic behavior) in the most severely affected elbow, when either austenitic 316 or 321 creep properties are employed

  7. Specialists meeting on LMFBR flow induced vibrations. Summary report

    International Nuclear Information System (INIS)

    1977-12-01

    A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs

  8. Specialists meeting on LMFBR flow induced vibrations. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-01

    A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs.

  9. Study on the phenomena of natural circulation in LMFBR

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Koga, Tomonari

    1993-01-01

    Decay heat removal with natural circulation is to be introduced to the LMFBR operation under loss of the electric power supply. The natural circulation is highly reliable, but the phenomenon is essentially unstable and subtle, which makes fine prediction difficult. The difficulties of experimental prediction are explained by facts that the phenomena are ruled by the delicate balance between the buoyancy force and the low pressure loss and are influenced by the various parameters such as local geometry, heat capacity and so on. Therefore the similarity rule for the natural circulation has not been fully understood. This study has been conducted to establish the simulation method for the natural circulation phenomena and the detailed phenomena have been reviewed. For the natural circulation in an LMFBR plant, there are no readily available reference velocity and temperature. These values are related only with the heating and cooling rate, the characteristic length and physical properties of the testing fluid. Basic equations were transformed by these values, and dimensionless equations were derived and then two dimensionless numbers, the Gr' number and the Bo' number, were identified. In order to examine the similarity rule for natural circulation we performed experiments using the different scale water models, a 1/20th and a 1/6th model. The temperatures and velocities at typical points were measured in the transient condition with various heating rate as a parameter. Measured temperatures and velocities were transformed to dimensionless forms for comparison and the effects of the Bo' number and the Gr' number were examined. As a result, it was clarified that the effect of the Gr' number is negligibly small but the effect of Bo' number still remained in our experimental range. The Bo' number of an actual plant is within the range of this experiment. Accordingly similitude of the Bo' number becomes important in an experiment to simulate an actual plant. (author)

  10. Detailed inelastic analysis of an LMFBR pipeline

    International Nuclear Information System (INIS)

    Hibbitt, H.D.; Leung, E.K.; Ohalla, A.K.

    1982-01-01

    The paper describes detailed inelastic analyses of a large diameter, thin walled pipeline configuration typical of liquid metal cooled reactor primary piping, subject to thermal shock, with intermediate periods of creep hold time. Three such analyses are compared. Two of these analyses are performed with recently developed elements based on a combination of Fourier and polynomial interpolation to describe the deformation of the pipe. One of these two analyses includes continuous deformation of the pipe wall between each elbow and the adjacent straight pipe segments, while the other neglects such ''end effects'' on the elbow deformation. The third analysis is based on a modified axi-symmetric shell element for modeling the elbows (neglecting and effects). The results thus provide an assessment of the relative cost and importance of including consideration of end effects in modeling a realistic piping system, as well as providing a similar comparison between the two basic deforming section pipe models (Fourier/polynomial versus modified axi-symmetric shells)

  11. A review of ANL base technology studies in support of the U.S. LMFBR vibration program

    International Nuclear Information System (INIS)

    Wambsganss, M.W.; Chen, S.S.; Mulcahy, T.M.; Shin, Y.S.

    1977-01-01

    Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

  12. A review of ANL base technology studies in support of the U.S. LMFBR vibration program

    Energy Technology Data Exchange (ETDEWEB)

    Wambsganss, M W; Chen, S S [Components Technology Division, Argonne National Laboratory, Argonne, IL (United States); Mulcahy, T M; Shin, Y S

    1977-12-01

    Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

  13. Absorption process for removing krypton from the off-gas of an LMFBR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Stephenson, M.J.; Dunthorn, D.I.; Reed, W.D.; Pashley, J.H.

    1975-01-01

    The Oak Ridge Gaseous Diffusion Plant selective absorption process for the collection and recovery of krypton and xenon is being further developed to demonstrate, on a pilot scale, a fluorocarbon-based process for removing krypton from the off-gas of an LMFBR fuel reprocessing plant. The new ORGDP selective absorption pilot plant consists of a primary absorption-stripping operation and all peripheral equipment required for feed gas preparation, process solvent recovery, process solvent purification, and krypton product purification. The new plant is designed to achieve krypton decontamination factors in excess of 10 3 with product concentration factors greater than 10 4 while processing a feed gas containing typical quantities of common reprocessing plant off-gas impurities, including oxygen, carbon dioxide, nitrogen oxides, water, xenon, iodine, and methyl iodide. Installation and shakedown of the facility were completed and some short-term tests were conducted early this year. The first operating campaign using a simulated reprocessing plant off-gas feed is now underway. The current program objective is to demonstrate continuous process operability and performance for extended periods of time while processing the simulated ''dirty'' feed. This year's activity will be devoted to routine off-gas processing with little or no deliberate system perturbations. Future work will involve the study of the system behavior under feed perturbations and various plant disturbances. (U.S.)

  14. Review on Japanese activities in the field of maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

    Tsuchiya, T.; Fukuda, T.; Sato, M.; Okabayashi, K.; Takahashi, T.

    2002-01-01

    Summary of Japanese activities on maintenance and repair of LMFBR steam generators (SG) is described in this paper. The concept (adoption of helical coil tube etc.) of MONJU SG was established in conceptional design started from 1968, and research and development (R and D) program was prepared. Parallel with basic studies such as material, welding, sodium water reaction and etc., overall verification tests using mock up SGs were conducted. As the first step, 1 Mw SG with two active helical tubes (and eight dummy tubes) was fabricated and operated, and many maintenance and repair experiences were accumulated through two small water leak troubles. Two 50Mw SGs, 1/5 scale of MONJU SG, were constructed and operated for long time. Post test examinations were carried out for No.1 50 Mw SG and feasibility of this type of SG was confirmed. In regard to maintenance and repair techniques, explosive and welding method for tube plugging and UT and ECT techniques for inspection of tube integrity are under development. Overall verification test for on-site and in-factory maintenance and repair techniques was conducted using No.2 50Mw SG evaporator and applicability of those techniques to real plant was evaluated. Many experiences were accumulated for removal and cleaning of sodium water reaction products after sodium water reaction in the cooling system and pressure relief system, using the Large Sodium Water Reaction Test Facility (SWAT-1 and 3). (author)

  15. Boreside rotating ultrasonic tester for wastage determination of LMFBR-type steam generator tubes

    International Nuclear Information System (INIS)

    Neely, H.H.; Renger, H.L.

    1979-01-01

    Large sodium-water reaction (SWR) leak tests are being run in near-prototypic steam generators at prototypic plant conditions of the Liquid Metal Fast Breeder Reactor (LMFBR). These tests simulate various types of steam tube failure at predetermined locations. A SWR results in a highly energetic-exothermic-caustic reaction which erodes neighboring tubes. A boreside-rotating ultrasonic inspection device was developed to measure wall thickness and inside diameter of the 2/one quarter/Cr-1 Mo, 10.1 mm I.D. steam tubes. Rotation of the UT beam yields a complimentary scan of the full tube in a single pass. The UT system was designed with a 15 MHz transducer in pulse-echo compression-wave mode at a pulse rate of 10,000/second. The UT beam is rotated at 20 r/s on a 1.27 mm pitch. System outputs are diameter, wall thickness, attitude, and axial position. Measurements are processed, then fed to a CRT and computer for later retrieval and plotting

  16. Calculation of Doses Due to Accidentally Released Plutonium From An LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Fish, B.R.

    2001-08-07

    Experimental data and analytical models that should be considered in assessing the transport properties of plutonium aerosols following a hypothetical reactor accident have been examined. Behaviors of released airborne materials within the reactor containment systems, as well as in the atmosphere near the reactor site boundaries, have been semiquantitatively predicted from experimental data and analytical models. The fundamental chemistry of plutonium as it may be applied in biological systems has been used to prepare models related to the intake and metabolism of plutonium dioxide, the fuel material of interest. Attempts have been made to calculate the possible doses from plutonium aerosols for a typical analyzed release in order to evaluate the magnitude of the internal exposure hazards that might exist in the vicinity of the reactor after a hypothetical LMFBR (Liquid-Metal Fast Breeder Reactor) accident. Intake of plutonium (using data for {sup 239}Pu as an example) and its distribution in the body were treated parametrically without regard to the details of transport pathways in the environment. To the extent possible, dose-response data and models have been reviewed, and an assessment of their adequacy has been made so that recommended or preferred practices could be developed.

  17. Operational-safety advantages of LMFBR's: the EBR-II experience and testing program

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lindsay, R.W.; Golden, G.H.

    1982-01-01

    LMFBR's contain many inherent characteristics that simplify control and improve operating safety and reliability. The EBR-II design is such that good advantage was taken of these characteristics, resulting in a vary favorable operating history and allowing for a program of off-normal testing to further demonstrate the safe response of LMFBR's to upsets. The experience already gained, and that expected from the future testing program, will contribute to further development of design and safety criteria for LMFBR's. Inherently safe characteristics are emphasized and include natural convective flow for decay heat removal, minimal need for emergency power and a large negative reactivity feedback coefficient. These characteristics at EBR-II allow for ready application of computer diagnosis and control to demonstrate their effectiveness in response to simulated plant accidents. This latter testing objective is an important part in improvements in the man-machine interface

  18. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  19. Fluid structure interaction in LMFBR cores modelling by an homogenization method

    International Nuclear Information System (INIS)

    Brochard, D.

    1988-01-01

    The upper plenum of the internals of PWR, the steam generator bundle, the nuclear reactor core, may be schematically represented by a beam bundle immersed in a fluid. The dynamical study of such a system needs to take into account fluid structure interaction. A refined model at the scale of the tubes can be used but leads to a very difficult problem to solve even on the largest computers. The homogenization method allows to have an approximation of the fluid structure interaction for the global behaviour of the bundle. It consists of replacing the heterogeneous physical medium (tubes and fluid) by an equivalent homogeneous medium whose characteristics are determined from the resolution of a set of problems on the elementary cell. The aim of this paper is to present the main steps of the determination of this equivalent medium in the case of small displacements (acoustic behaviour of the fluid). Then an application to LMFBR core geometry has been realised, which shows the lowering effect on eigenfrequencies due to the fluid. Some comparisons with test results will be presented. 6 refs, 7 figs, 2 tabs

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

    International Nuclear Information System (INIS)

    Neely, H.H.; Renger, L.

    1977-01-01

    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- 1 / 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 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 1 / 8 -inch long and 0.002-inch deep could readily be detected

  1. Simulation of LMFBR pump transients and comparison to LOF that occurred at EBR-II

    International Nuclear Information System (INIS)

    Koenig, F.F.; Dean, E.M.

    1985-01-01

    In a large LMFBR plant design, a number of pumps in parallel will feed the core. It must be demonstrated that the plant can continue to operate with the loss of one of the primary pumps. It is desirable not to have check valves in the loop from a reliability and economic standpoint. Simulations have been made to determine the consequences of a loss of one pump in a four-loop pool plant in which no plant protection action is taken. This analysis would be used to determine the required power rundown that would accompany pump loss. The two primary centrifugal pumps in EBR-II feed the core and blanket plenums in two parallel flow paths. The loss of one pump will result in decrease core flow and reverse flow through the down pump since no check valves are present in the system. For a large pool plant with four primary pumps, the loss of one pump will also result in reverse flow through the down pump if check valves of flow diodes are not included. The resulting flow transient has been modeled for EBR-II and the large plant using the DNSP program

  2. Assessment of inspectability of LMFBR designs. Final report

    International Nuclear Information System (INIS)

    1981-09-01

    This two-volume report provides a comprehensive review of the inspectability of specific portions of loop- and pool-type LMFBR (1000-MWe) designs selected by EPRI. The designs were developed during the mid to late 1970s by three independent design teams (General Electric Co., Rockwell International, and Westinghouse) under the sponsorship of DOE (formerly ERDA) and EPRI. The requirements for normal, contingency, and post-repair inspections, addressed in this report, were established from Draft 12 of the ASME Boiler and Pressure Vessel Code, Section XI Division 3, issued in September 1979. These requirements, the intrinsic characteristics of the designs, the environmental (radiation, thermal, and atmospheric) aspects, and the available (present and near-term) inspection techniques, formed the basis for assessing the selected portions of the design or (1) accessibility, (2) feasibility, (3) practicality, and (4) costs to perform the above-specified inspections. Changes and additions fly ash has been as a concrete additive; however, extensive pilot scale development is underway to advance ash use in the TVA region in such areas as mineral and magnetite recovery, and mineral wool insulation. Recommended studies include: (1) the feasibility of converting existing wet fly d by the fuels include: residential (which includes residential and commercial), elthodology will be developed and verified in Phase II

  3. Structural analysis for LMFBR applications[Indian position paper

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-05-01

    Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start.

  4. Experience on detection of leakages in LMFBR-steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Smit, C C

    1975-07-01

    One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks.

  5. Hydrodynamic analysis of the LMFBR prompt burst excursion (PBE) experiment

    International Nuclear Information System (INIS)

    Young, M.F.

    1977-01-01

    A series of in-pile experiments has been conducted at Sandia Laboratories to provide information on pressure levels and conversion of thermal energy into mechanical work in LMFBR cores during hypothetical, superprompt-critical excursions. Pressures generated in these experiments are recorded by a pressure transducer located at the top and bottom of a sodium channel surrounding a single, fresh UO 2 fuel pin. Work energy conversion is measured by a linear motion transducer connected to a piston at the top of the sodium column. Since the pressure transducers are located fairly far from the location of pin failure, it becomes necessary to determine the effect of channel geometry and piston motion on the observed pressure data. A two-dimensional, hydrodynamic analysis of pressure pulse propagation in the fuel pin-coolant channel geometry was therefore performed using the CSQII computer code. The initial series of PBE experiments consists of single, fresh UO 2 pins surrounded by a sodium-filled or dry-coolant channel contained in a closed test capsule. The capsule is subjected to a maximum pulse in the Annular Core Pulse Reactor (ACPR) resulting in an energy deposition of from 2350 to 2900 J/g (14 and 20 percent enriched pins). The pulse width at half maximum (PWHM) is about 5 ms

  6. Evaluation of high-pressure containment buildings for LMFBR's

    International Nuclear Information System (INIS)

    Armstrong, G.R.

    1981-01-01

    A study was conducted on the use of High Pressure LMFBR Containment Buildings for 1000 MW(e) LMFBRs. Two principal aspects were investigated: accident consequence mitigation and cost. Two types of hypothetical accidents were analyzed to establish consequence mitigation: melt-through and energetic expulsion. Three Containment Building (CB) design pressures were investigated: 69 kPa (10 psig), 207 kPa (30 psig), and 414 kPa (60 psig). Four types of design structures were analyzed to establish cost: steel, steel with confinement building, reinforced concrete, and prestressed/post-tensioned concrete. Results show that: it is within reason that a high pressure containment for a 1000 MW(e) reactor can be fabricated that will retain its integrity during postulated severe hypothetical accidents, if available measures are taken to reduce or prevent hydrogen production and the cost differential between basic high (414 kPa) and low (69 kPa) pressure containments is $10 x 10 6 or less

  7. Experience on detection of leakages in LMFBR-steam generators

    International Nuclear Information System (INIS)

    Smit, C.C.

    1975-01-01

    One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks

  8. Residual stress effects in LMFBR fracture assessment procedures

    International Nuclear Information System (INIS)

    Hooton, D.G.

    1984-01-01

    Two post-yield fracture mechanics methods, which have been developed into fully detailed failure assessment procedures for ferritic structures, have been reviewed from the point of view of the manner in which as-welded residual stress effects are incorporated, and comparisons then made with finite element and theoretical models of centre-cracked plates containing residual/thermal stresses in the form of crack-driving force curves. Applying the procedures to austenitic structures, comparisons are made in terms of failure assessment curves and it is recommended that the preferred method for the prediction of critical crack sizes in LMFBR austenitic structures containing as-welded residual stresses is the CEGB-R6 procedure based on a flow stress defined at 3% strain in the parent plate. When the prediction of failure loads in such structures is required, it is suggested that the CEGB-R6 procedure be used with residual/thermal stresses factored to give a maximum total stress of flow stress magnitude

  9. Multicell slug flow heat transfer analysis of finite LMFBR bundles

    International Nuclear Information System (INIS)

    Yeung, M.K.; Wolf, L.

    1978-12-01

    An analytical two-dimensional, multi-region, multi-cell technique has been developed for the thermal analysis of LMFBR rod bundles. Local temperature fields of various unit cells were obtained for 7, 19, and 37-rod bundles of different geometries and power distributions. The validity of the technique has been verified by its excellent agreement with the THTB calculational result. By comparing the calculated fully-developed circumferential clad temperature distribution with those of the experimental measurements, an axial correction factor has been derived to account for the entrance effect for practical considerations. Moreover, the knowledge of the local temperature field of the rod bundle leads to the determination of the effective mixing lengths L/sub ij/ for adjacent subchannels of various geometries. It was shown that the implementation of the accurately determined L/sub ij/ into COBRA-IIIC calculations has fairly significant effects on intersubchannel mixing. In addition, a scheme has been proposed to couple the 2-D distributed and lumped parameter calculation by COBRA-IIIC such that the entrance effect can be implanted into the distributed parameter analysis. The technique has demonstrated its applicability for a 7-rod bundle and the results of calculation were compared to those of three-dimensional analyses and experimental measurements

  10. Technical assessment study on pool-type LMFBR

    International Nuclear Information System (INIS)

    1986-01-01

    Technical assessment study on pool-type LMFBR was started in 1984 FY, inheriting the products from the Feasibility study, in order to accomplish cost reduction of reactor structure and enhanced structural reliability. This study consists of four major subjects; aseismic design development, component design optimization, high temperature structural design optimization and thermal hydraulics design optimization. In 1985 FY numbers of large model tests and analytical evaluations have been performed based on the prospects obtained in the first year's study. These tests and analyses have produced a lot of findings in each subject. They are concerning; (1) the effect of various building structures and analysis methods on floor response reduction, and data for evaluation of aseismic design concepts and structural integrity to seismic loading in the aseismic design development study. (2) data for evaluation of size reduction of main components in the reactor vessel, and heat transfer data required for structural integrity evaluation in the component design optimization study. (3) data for verification of inelastic analysis method, and assurance of technical applicability of disimilar weld in the high temperature structural design optimization study. (4) the effect of component size and location on thermal hydraulic characteristics, and data of thermal hydraulic similarity in thermal hydraulic design optimization study. This report summarizes the results obtained in 1985 FY. (author)

  11. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  12. Irradiation effects on low-friction coatings for LMFBR applications

    International Nuclear Information System (INIS)

    Ward, A.L.; Johnson, R.N.; Guthrie, G.L.; Aungst, R.C.

    1975-11-01

    A variety of wear-resistant low-friction materials has been irradiated in the EBR-II in order to assess their reponse to LMFBR environments. Pre- and postirradiation testing and examination efforts have concentrated on candidate materials for application to the wear pads on FTR ducts (fuel, control, and reflector assemblies), and a significant result has been qualification of a proprietary detonation-gun-applied chromium carbide coating which employs a Ni Cr binder. Additional materials such as Inconel-718, Haynes-273, aluminides, and various chromium carbide/binder combinations, and other application processes such as plasma-spray, weld-overlays, diffusion bonding and explosive bonding, have also been studied. The most detailed examinations were conducted on selected chromium carbide coatings and included visual inspection, weight and dimensional measurements, metallography, electron microprobe, epoxy-lift-off, and x-ray diffraction analysis. Chromium carbide coatings applied by the detonation-gun process have demonstrated a marked superiority to those applied by plasma-spray techniques

  13. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  14. Finite element elastic-plastic analysis of LMFBR components

    International Nuclear Information System (INIS)

    Levy, A.; Pifko, A.; Armen, H. Jr.

    1978-01-01

    The present effort involves the development of computationally efficient finite element methods for accurately predicting the isothermal elastic-plastic three-dimensional response of thick and thin shell structures subjected to mechanical and thermal loads. This work will be used as the basis for further development of analytical tools to be used to verify the structural integrity of liquid metal fast breeder reactor (LMFBR) components. The methods presented here have been implemented into the three-dimensional solid element module (HEX) of the Grumman PLANS finite element program. These methods include the use of optimal stress points as well as a variable number of stress points within an element. This allows monitoring the stress history at many points within an element and hence provides an accurate representation of the elastic-plastic boundary using a minimum number of degrees of freedom. Also included is an improved thermal stress analysis capability in which the temperature variation and corresponding thermal strain variation are represented by the same functional form as the displacement variation. Various problems are used to demonstrate these improved capabilities. (Auth.)

  15. KANDY - a numerical model to describe phenomena, which - in a heated and voided fuel element of an LMFBR - may occur

    International Nuclear Information System (INIS)

    Thurnay, K.

    1984-02-01

    Kandy is a model developed to describe the essential destructionphenomena of the fuel elements of an LMFBR. The fuel element is assumed to be a voided one, in which the heat generation is still going on. The main process to be modeled is the melting/bursting/evaporating of parts of the fuel pins and the subsequent dislocation of these materials in the coolant channel. The work presented summarizes the assumptions constituting the model, develops the corresponding equations of motion and describes the procedure, turning these into a system of difference-equations ready for coding. As a final part results of a testcase calculation with the Kandy-code are presentend and interpreted. (orig.) [de

  16. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    International Nuclear Information System (INIS)

    Roake, W.E.

    1977-01-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

  17. Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

    International Nuclear Information System (INIS)

    Minden, C. v.; Schultheiss, G.F.

    1982-01-01

    A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.) [de

  18. Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

    International Nuclear Information System (INIS)

    Shin, Y.S.; Wambsganss, M.W.

    1975-05-01

    This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration

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

    International Nuclear Information System (INIS)

    Kurzeka, W.; Oliva, R.; Welch, F.

    1974-01-01

    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

  20. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

    1977-04-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.

  1. LMFBR safety criteria: cost-benefit considerations under the constraint of an a priori risk criterion

    International Nuclear Information System (INIS)

    Hartung, J.

    1979-01-01

    The role of cost-benefit considerations and a priori risk criteria as determinants of Core Disruptive Accident (CDA)-related safety criteria for large LMFBR's is explored with the aid of quantitative risk and probabilistic analysis methods. A methodology is described which allows a large number of design and siting alternatives to be traded off against each other with the goal of minimizing energy generation costs subject to the constraint of both an a priori risk criterion and a cost-benefit criterion. Application of this methodology to a specific LMFBR design project is described and the results are discussed. 5 refs

  2. Sassis tuumamajandus / John Carey

    Index Scriptorium Estoniae

    Carey, John

    2008-01-01

    USA presidendikandidaadi John McCaini energeetikakava kohaselt tuleks USA-sse ehitada 100 uut tuumaelektrijaama, neist esimesed 45 peaksid valmima aastaks 2030. Tuumaelektrijaamad aitaksid rahuldada USA energiavajadust ning võitleksid ka kliimasoojenemise vastu. Eksperdid on arengukava suhtes kriitilised

  3. Dynamic response of single hexagonal LMFBR core subassembly wrappers

    Energy Technology Data Exchange (ETDEWEB)

    Ash, J. E.; Marciniak, T. J.; (Argonne National Lab., IL (United States))

    1977-07-01

    To analyze the dynamic structural response of the LMFBR core subassembly hexagonal wrappers to postulated local energy releases and the sensitivity of the response to variations in both the pressure loading and the material properties of the stainless steel, a finite-element computer code STRAW has been developed. A series of experiments was performed to study the effects of variations in material properties. The amount of coldworking to which the Type 316 stainless steel is subjected has a strong influence upon the ductility and the elastic yield point. The usual fabrication process produced a nominally 20% coldworking with a yield point of about 680 MPa. By designing a special set of dies for the drawing process, a very low ductility hexcan was produced for which the yield point was raised to 820 MPa. Conversely, the yield point was lowered to 170 MPa by a solution annealing process producing a highly ductile test hexcan. A metallurgical study was conducted to find a representative brittle simulant material for the irradiated end-of-life steel properties. An aging treatment for Type 446 stainless steel was developed which reproduced the expected tensile-flow behavior of the in-pile subassembly. Further study is underway to investigate the fracture properties of the simulant material. The pressure pulses were generated by the controlled expansion of high-pressure detonation poducts from low-density explosives detonated inside a vented steel cannister. The orifice configuration of the cannister and the charge mixture ratio were designed to produce two specified pulse shapes. A charge containing 37,7 g PETN mixed with 35 wt % inert, hollow-glass microballoons developed a pressure pulse peak of 9.5 MPa at 1.0 ms. Increasing the PETN to 41 g resulted in a 14.6 MPa peak pressure, and increasing the explosive concentration to 90 wt % in the mixture increased the burning rate and the pulse risetime, so that the peak occurred at 0.6 ms.

  4. Fuel pin response to an overpower transient in an LMFBR

    International Nuclear Information System (INIS)

    Grosberg, A.J.; Head, J.L.

    1979-01-01

    This paper describes a method by which the ability of a whole-core code accurately to predict the time and location of the first fuel pin failures may be tested. The method involves the use of a relatively simple whole-core code to 'drive' a sophisticated fuel pin code, which is far too complex to be used within a whole-core code but which is potentially capable of modelling reliably the response of an individual fuel pin. The method cannot follow accurately the subsequent course of the transient because the simple whole-core code does not model the reactivity effects of events which may follow pin failure. The codes used were the simple whole-core code FUTURE and the fuel pin behaviour code FRUMP. The paper describes an application of the method to analyse a hypothetical LMFBR accident in which the control rods were assumed to be driven from the core at maximum speed, with all trip circuits failed. Taking 0.5% clad strain as a clad failure criterion, failure was predicted to occur at the top of the active core at about 10s into the transient. A repeat analysis, using an alternative clad yield criterion which is thought to be more realistic, indicated failure at the same position but 24s into the transient. This is after the onset of sodium boiling. Pin failure at the top of the core are likely to cause negative reactivity changes. In this hypothetical accident, pin failures are likely, therefore, to have a moderating effect on the course of the transient. (orig.)

  5. Input parameters to codes which analyze LMFBR wire-wrapped bundles

    International Nuclear Information System (INIS)

    Hawley, J.T.; Chan, Y.N.; Todreas, N.E.

    1980-12-01

    This report provides a current summary of recommended values of key input parameters required by ENERGY code analysis of LMFBR wire wrapped bundles. This data is based on the interpretation of experimental results from the MIT and other available laboratory programs

  6. The state of art of the methods for thermohydraulics design of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1981-09-01

    The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt

  7. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  8. LMFBR safety. 6. Review of current issues and bibliography of literature (1977)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1978-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development. Selected bibliographic information on LMFBRs relative to the development and safety of the breeder reactor is presented for the year 1977. The bibliography consists of approximately 198 abstracts covering research and development, operating experience, and design practices. Keyword, author, and permuted-title indexes are included for completeness

  9. Subassembly faults diagnostic of an LMFBR type reactor by the measurement of temperature noise

    International Nuclear Information System (INIS)

    Kokorev, B.V.; Palkin, I.I.; Turchin, N.M.; Pallagi, D.; Horanyi, S.

    1979-09-01

    The subassembly faults detection possibility by temperature noise analysis of an LMFBR is described. The paper contains the results of diagnostical examinations obtained on electrically heated NaK test rigs. On the basis of these results the measurement of temperature noise RMS value seems to be a practicable method to detect local blockages in an early phase. (author)

  10. LMFBR safety program. Annual technical progress report. Government fiscal year, 1977

    International Nuclear Information System (INIS)

    1977-01-01

    Information is presented concerning the development of the SOMIX-1 computer code for sodium drop burning analysis; experimental analysis of burning sodium drops; aerosol leakage from containment buildings; high-temperature-concentration aerosols; aerosol source term from vaporized fuel; properties of high-temperature fuel mixtures; and development of the COMRADEX computer code for analysis of radiological doses in the environment from LMFBR accidents

  11. Review of pertinent thermal-hydraulic data for LMFBR core natural circulation analyses

    International Nuclear Information System (INIS)

    Bishop, A.A.; Coffield, R.D. Jr.; Markley, R.A.

    1980-01-01

    A literature review and summary of significant data is presented relative to LMFBR core natural convection cooling analysis. First, a brief review of computer codes and respective input data needs is made, significant data areas are then addressed and data for verifying the code calculations are described. Recommendations and conclusions with regard to the data are included

  12. A survey of the French creep-fatigue design rules for LMFBR

    International Nuclear Information System (INIS)

    Tribout, J.; Cordier, G.; Moulin, D.

    1987-01-01

    The paper provides a survey of the creep-fatigue design rules for the LMFBR in France. These rules are the ones currently implemented in French component manufacturing. The background of each item is discussed and the trends for improvements currently investigated are described. The creep-fatigue rules apply to elastic analysis only. (orig.)

  13. A miniature inductive temperature sensor to monitor temperature noise in the coolant of an LMFBR

    International Nuclear Information System (INIS)

    Dean, S.A.; Sandham, C.W.

    1980-01-01

    A description is given of the design and performance of miniature inductive sensors developed to monitor fast temperature fluctuations in the sodium coolant above the core of a LMFBR. These instruments, designed to be installed within existing thermocouple containment thimbles, also provide a steady-state temperature indication for reactor control purposes. (author)

  14. Theoretical study and experimental investigation of mixed and natural circulation in LMFBR core subassemblies

    International Nuclear Information System (INIS)

    Leteinturier, D.; Blanc, D.; Menant, B.; Basque, G.

    1980-02-01

    A presentation is made of theoretical and experimental studies carried out in France on mixed and natural convection in LMFBR wire wrapped bundles. Two codes are described, one for mixed convection THERNAT and the other for natural convection BACCHUS. THe related experimental program FETUNA, with electrically heated bundles in sodium loops, is also presented

  15. The water vapor nitrogen process for removing sodium from LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    Crippen, M D; Funk, C W; Lutton, J M [Hanford Engineering Development Laboratory, Richland (United States)

    1978-08-01

    Application and operation of the Water Vapor-Nitrogen Process for removing sodium from LMFBR components is reviewed. Emphasis is placed on recent efforts to verify the technological bases of the process, to refine the values of process parameters and to ensure the utility of the process for cleaning and requalifying components. (author)

  16. SIMMER-I: an S/sub n/, Implicit, Multifield, Multicomponent, Eulerian, Recriticality code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Bell, C.R.; Bleiweis, P.B.; Boudreau, J.E.; Parker, F.R.; Smith, L.L.

    1976-08-01

    Physical models, numerical methods, and program description are presented for SIMMER-I, a computer program which predicts the neutronic and fluid dynamic behavior of an LMFBR during a hypothetical core disruptive accident

  17. Complementary role of critical integral experiment and power reactor start-up experiments for LMFBR neutronics data and method validation

    International Nuclear Information System (INIS)

    Salvatores, M.

    1986-09-01

    Both critical experiments and power reactor results play at present a complementary role in reducing the uncertainties in Key design parameters for LMFBR, which can be relevant for the economic performances of this type of reactors

  18. The benefits and problems of base seismic isolation for LMFBR reactor plants

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1988-01-01

    The use of seismic isolation as an approach to aseismic design has gained increasing interest as a viable and efficient engineering solution to earthquake ground motion both within and outside of the nuclear field. Seismic isolation design is fundamentally different from conventional design practice. In the conventional approach, seismic loads are resisted by making the structures, equipment, piping, and associated supports strong enough to resist seismic loads and to provide high levels of ductility. The use of seismic isolation approaches the problem by decoupling the structure (and its contents) from the seismic input resulting from ground shaking. Because LMFBR systems operate at virtually atmospheric pressure, vessels, piping, and associated components tend to be quite thin-walled. The problem is that these thin-walled items have little inherent resistance to earthquake effects and are vulnerable to seismic load effects. As a result, earthquake loads have an even greater influence on LMR designs than they already are in LWR plants. The potential benefits of seismic isolation for an LMR plant are considerable, including minimization of high-cost commodities such as stainless steel, large reductions in internal equipment loads, increased margins of safety for beyond-design-basis loads, and enhancement of plant standardization design. There are, of course, a number of issues and concerns in the use of seismic isolation for a nuclear power plant. These issues cover a number of items such as the lack of experience in actual earthquakes, effects of long-period ground motion, effect of vertical loads, traveling waves, and other related concerns. This paper presents an evaluation of the benefits and problems in the use of seismic isolation in LMR plants. 12 refs, 7 figs

  19. Specialists' meeting on maintenance and repair of LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    2002-01-01

    The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topic areas were discussed by participants: National review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; Research and Development work on maintenance and repair; Experience on steam generator maintenance and repair. During the meeting papers were presented by the participants on behalf of their countries and organizations. A final discussion session was held and summaries, general conclusions and recommendations were approved by consensus

  20. LMFBR operational and experimental local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS- and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  1. A critical experimental study of integral physics parameters in simulated LMFBR meltdown cores

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Wade, D.C.; Bucher, R.G.; Smith, D.M.; McKnight, R.D.; Lesage, L.G.

    1978-01-01

    Integral physics parameters of several representative, idealized meltdown LMFBR configurations were measured in mockup critical assemblies on the ZPR-9 reactor at Argonne National Laboratory. The experiments were designed to provide data for the validation of analytical methods used in the neutronics part of LMFBR accident analysis. Large core distortions were introduced in these experiments (involving 18.5% core volume) and the reactivity worths of configuration changes were determined. The neutronics parameters measured in the various configurations showed large changes upon core distortion. Both diffusion theory and transport theory methods were shown to mispredict the experimental configuration eigenvalues. In addition, diffusion theory methods were shown to result in a non-conservative misprediction of the experimental configuration change worths. (author)

  2. LMFBR operational and experimental in-core local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS-and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  3. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle

  4. Acoustics and voiding dynamics during SLSF simulations of LMFBR undercooling transients

    International Nuclear Information System (INIS)

    Anderson, T.T.; Kuzay, T.M.; Marr, W.W.; Miles, K.J.; Pedersen, D.R.; Thompson, D.H.; Wilson, R.E.

    1978-01-01

    The SLSF is the largest U.S. in-reactor test vehicle for steady-state and transient experiments in an environment typical of a LMFBR core. The SLSF experiment program, sponsored by the Department of Energy, contributes to the LMFBR safety assurance program by providing data on key phenomena that occur during postulated reactor accidents. This paper describes completed SLSF experiments, in-core instrumentation used, and methods of data interpretation to determine sodium boiling and voiding dynamics. Boiling inception is shown to be identifiable from several types of in-core instruments. Location of the boiling front and void growth derived from experimental data are compared with analytical predictions. These and other data form the basis to improve understanding of accidents and to validate or guide the development of accident analysis methods

  5. Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-10-01

    The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors.

  6. Report of the IAEA advisory group meeting on LMFBR fuel reprocessing

    International Nuclear Information System (INIS)

    1976-05-01

    A summary of the papers and discussions of the meeting is presented, reviewing the status of development in LMFBR fuel reprocessing and focusing attention on important problem areas. The following topics are discussed: Transport, storage and removal of sodium; decladding and shearing; dissolution; Purex process; fluoride volatility method; off-gas purification; waste disposal. Status reports of national programmes of Belgium, France, Federal Republic of Germany, Italy, Japan, United Kingdom, USSR and USA are included

  7. LMFBR conceptual design study: an overview of environmental and safety concerns

    International Nuclear Information System (INIS)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE

  8. A survey of LMFBR cavitation technology in the U.S.A

    International Nuclear Information System (INIS)

    Cha, Y.S.; Huebotter, P.R.; Hopenfeld, J.

    1976-01-01

    Several experimental programmes of a basic and applied nature were established in the USA in order to develop guidelines to ensure design and operation of LMFBR hydraulic components free from cavitation and/or cavitation damage. As of March 1976, most of these experimental programs are still in progress. Each programme is briefly described. The available interium data are presented. References that are relevant are provided

  9. LMFBR conceptual design study: an overview of environmental and safety concerns

    Energy Technology Data Exchange (ETDEWEB)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE.

  10. Comparison of different LMFBR primary containment codes applied to a Benchmark problem

    International Nuclear Information System (INIS)

    Benuzzi, A.

    1986-01-01

    The Cont Benchmark calculation exercise is a project sponsored by the Containment Loading and Response Group, a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee - CEC. A full-size typical Pool type LMFBR undergoing a postulated Core Disruptive Accident (CDA) has been defined by Belgonucleaire-Brussels under a study contract financed by the CEC and has been submitted to seven containment code calculations. The results of these calculations are presented and discussed in this paper

  11. Comparative study of heterogeneous and homogeneous LMFBR cores in some accident conditions

    International Nuclear Information System (INIS)

    Renard, A.; Evrard, G.

    1978-01-01

    An heterogeneous design and a homogeneous one of a LMFBR core with the same power and similar dimensions are compared from the safety point-of-view. The comparison is performed for several accident conditions, such as Loss-of-Flow and Transient Overpower, with the same failure criteria and model assumptions for both cores. Qualitative trends are deduced from the behaviour of the core designs in the investigated transient conditions. (author)

  12. Seismic criteria studies and analyses. Quarterly progress report No. 3. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    1975-01-03

    Information is presented concerning the extent to which vibratory motions at the subsurface foundation level might differ from motions at the ground surface and the effects of the various subsurface materials on the overall Clinch River Breeder Reactor site response; seismic analyses of LMFBR type reactors to establish analytical procedures for predicting structure stresses and deformations; and aspects of the current technology regarding the representation of energy losses in nuclear power plants as equivalent viscous damping.

  13. Fretting and wear of stainless and ferritic steels in LMFBR steam generators

    International Nuclear Information System (INIS)

    Lewis, M.W.J.; Campbell, C.S.

    1981-01-01

    Steam generators for LMFBR's may be subject to both fretting wear as a result of flow-induced vibrations and to wear from larger amplitude sliding movements from thermal changes. Results of tests simulating the latter are given for stainless and ferritic steels. For the assessment of fretting wear damage, vibration assessments must be combined with data on specific wear rates. Test mechanisms used to study fretting in sodium covering impact, impact-slide and pure rubbing are described and results presented. (author)

  14. Seismic analysis of a large LMFBR with fluid-structure interactions

    International Nuclear Information System (INIS)

    Ma, D.C.

    1985-01-01

    The seismic analysis of a large LMFBR with many internal components and structures is presented. Both vertical and horizontal seismic excitations are considered. The important hydrodynamic phenomena such as fluid-structure interaction, sloshing, fluid coupling and fluid inertia effects are included in the analysis. The results of this study are discussed in detail. Information which is useful to the design of future reactions under seismic conditions is also given. 4 refs., 12 figs

  15. Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    1978-10-01

    The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors

  16. A fundamental study on sodium-water reaction in the double-pool-type LMFBR

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Akimoto, Tokuzo

    1987-01-01

    In order to evaluate the pressure rise by large sodium-water reaction in the Double-Pool LMFBR, basic tests on pressure wave celerity in rectangular tube are carried out. The initial spike pressure in rectangular-shelled steam generator of the Double Pool reactor, strongly depends on pressure wave celerity. In this study, celerity was measured as a function of pressure wave rising time and pulse height, and influence of water around the test section on celerity was investigated. (author)

  17. Damping of the radial impulsive motion of LMFBR core components separated by fluid squeeze films

    International Nuclear Information System (INIS)

    Liebe, R.; Zehlein, H.

    1977-01-01

    The core deformation of a liquid metal cooled fast breeder reactor (LMFBR) due to local pressure propagation from rapid energy releases is a complex three-dimensional fluid-structure-interaction problem. High pressure transients of short duration cause structural deformation of the closely spaced fuel elements, which are surrounded by the flowing coolant. Corresponding relative displacements give rise to squeezing fluid motion in the thin layers between the subassemblies. Therefore significant backpressures are produced and the resulting time and space dependent fluid forces are acting on the structure as additional non-conservative external loads. Realistic LMFBR safety analysis of several clustered fuel elements have to account for such flow induced forces. Several idealized models have been proposed to study some aspects of the complex problem. As part of the core mechanics activities at GfK Karlsruhe this paper describes two fluid flow models (model A, model B), which are shown to be suitable for physically coupled fluid-structure analyses. Important assumptions are discussed in both cases and basic equations are derived for one- and two-dimensional incompressible flow fields. The interface of corresponing computer codes FLUF (model A) and FLOWAX (model B) with structural dynamics programs is outlined. Finally fluid-structure interaction problems relevant to LMFBR design are analyzed; parametric studies indicate a significant cushioning effect, energy dissipation and a strongly nonlinear as well as timedependent damping of the structural response. (Auth.)

  18. Effect of reactor size on the breeding economics of LMFBR blankets

    International Nuclear Information System (INIS)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

  19. Econometric modelling of certain nuclear power systems based on thermal and fast breeder reactors

    International Nuclear Information System (INIS)

    Pavelescu, M.; Pioaru, C.; Ursu, I.

    1988-01-01

    Certain known economic analysis models for a LMFBR fast breeder and CANDU thermal solitary reactors are presented, based on the concepts of discounting and levelization. These models are subsequently utilized as a basis for establishing an original model for the econometric analysis of certain thermal reactor systems or/and fast breeder reactors. Case studies are subsequently conducted with the systems: 1-CANDU, 2-LMFBR, 3-CANDU + LMFBR which enables us to draw certain interesting conclusions for a long range nuclear power policy. (author)

  20. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-09-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

  1. The RCC-MR design code for LMFBR components. A useful basic for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1985-11-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials (Stainless steels), temperature service level (550-600 0 C), loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain

  2. The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1986-01-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

  3. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-02-24

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included.

  4. PNC status report on leak detector development for LMFBR steam generators

    International Nuclear Information System (INIS)

    Kuroha, M.; Sato, M.

    1984-01-01

    Chemical and acoustic type leak detectors have been developed for detecting a small sodium-water reaction in an LMFBR steam generator. This paper presents a summary of the development. (1) Test results on PNC type in-sodium hydrogen meters including a description of the structure, the long-term reliability and the durability, and the improved meter with an orifice, (2) Development of in-cover gas hydrogen meters, (3) Hydrogen detection tests and analyses, (4) Operating experiences of electrochemical in-sodium oxygen meters, and (5) Basic studies on acoustic characteristics of the sodium-water reaction. (author)

  5. Study of thermal-hydraulic characteristics in an LMFBR intermediate plenum

    International Nuclear Information System (INIS)

    Uotani, M.; Naohara, N.; Kinoshita, I.

    1985-01-01

    Experimental studies using water and liquid metal were conducted in order to investigate the thermal-hydraulic characteristics of an LMFBR intermediate plenum. The present study is an attempt to evaluate the effect of natural convection on the temperature field and to validate the prediction method of temperature profile in a thermally stratified cavity. The experimental results indicated that the effect of the natural convection on flow velocity and heat transfer in the cavity is reduced with increasing the modified stratification parameter. The calculation by FEM code and a simple 1-D model are effective to predict the temperature profile in the cavity

  6. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  7. Analysis of hypothetical LMFBR whole-core accidents in the USA

    International Nuclear Information System (INIS)

    Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

    1978-01-01

    The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper

  8. Analytical treatment of large leak pressure behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Hori, Masao; Miyake, Osamu

    1980-07-01

    Simplified analytical methods applicable to the estimation of initial pressure spike in case of a large leak accident in LMFBR steam generators were devised as follows; (i) Estimation of the initial water leak rate by the centered rarefaction wave method, (ii) Estimation of the initial pressure spike by the one-dimensional compressible method with either the columnar bubble growth model or the spherical bubble growth model. These methods were compared with relevant experimental data or other more elaborate analyses and validated to be usable in simple geometry and limited time span cases. Application of these methods to an actual steam generator case was explained and demonstrated. (author)

  9. Bubble behavior in LMFBR core disruptive accidents. Annual report, June 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Kennedy, M.F.; Rao, S.P.; Refling, J.G.

    1976-08-01

    The work reported here is part of the Aerosol Release and Transport program for LMFBR safety assessment for the Reactor Safety Research Division of the U.S. Nuclear Regulatory Commission. Six areas were at various stages of investigation during this reporting period. A study of nonequilibrium mass transfer during fuel expansion and a study of the dynamics of fuel expansion into the sodium pool were completed. Studies are underway on condensation on above-core structures and on generation of aerosols from condensation. Studies were initiated on small-particle generation from hydrodynamic fragmentation, on particle kinematics and on particle-surface interaction

  10. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-03-21

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  11. Whole-core thermal-hydraulic transient code development and verification for LMFBR analysis

    International Nuclear Information System (INIS)

    Spencer, D.R.

    1979-04-01

    Predicted performance during both steady state and transient reactor operation determines the steady state operating limits on LMFBRs. Unnecessary conservatism in performance predictions will not contribute to safety, but will restrict the reactor to more conservative, less economical steady state operation. The most general method for reducing analytical conservatism in LMFBR's without compromising safety is to develop, validate and apply more sophisticated computer models to the limiting performance analyses. The purpose of the on-going Natural Circulation Verification Program (NCVP) is to develop and validate computer codes to analyze natural circulation transients in LMFBRs, and thus, replace unnecessary analytical conservatism with demonstrated calculational capability

  12. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  13. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2

    International Nuclear Information System (INIS)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa; Bando, Masaru.

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author)

  14. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  15. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included

  16. Sodium fire studies in France safety tests and applications on an LMFBR

    International Nuclear Information System (INIS)

    Fruchard, Y.; Colome, J.; Malet, J.C.; Berlin, M.; de Cuy, G.D.; Justin, J.; Duco, J.; Fourest, B.

    1976-01-01

    The risk of sodium fires in an LMFBR requires thorough analysis, and the possible consequences of an accidental fire must be accurately determined. Not only must means of prevention and detection be perfected, but techniques must be developed to limit the damage caused by a fire: extinguishment, aerosol containment, protection of reactor structures. The program currently undertaken by the CEA's Nuclear Safety Department covering these problems is described. The major results obtained as well as their application to the SUPER-PHENIX reactor are included

  17. LMFBR safety. 5. Review of current issues and bibliography of literature (1975--1976)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    The current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA), are discussed. Bibliographic information on worldwide LMFBRs relative to the development and safety of the breeder reactor is presented for the period 1975 through 1976. The bibliography consists of approximately 1618 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Keyword, author, and permuted-title indexes are included for completeness

  18. Seismic isolation structure for pool-type LMFBR - isolation building with vertically isolated floor for NSSS

    International Nuclear Information System (INIS)

    Sakurai, A.; Shiojiri, H.; Aoyagi, S.; Matsuda, T.; Fujimoto, S.; Sasaki, Y.; Hirayama, H.

    1987-01-01

    The NSSS isolation floor vibration characteristics were made clear. Especially, the side support bearing (rubber bearing) is effective for horizontal floor motion restraint and rocking motion control. Seismic isolation effects for responses of the reactor components can be sufficiently expected, using the vertical seismic isolation floor. From the analytical and experimental studies, the following has been concluded: (1) Seismic isolation structure, which is suitable for large pool-type LMFBR, were proposed. (2) Seismic response characteristics of the seismic isolation structure were investigated. It was made clear that the proposed seismic isolation (Combination of the isolated building and the isolated NSSS floor) was effective. (orig./HP)

  19. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  20. Development of concept and neutronic calculation method for large LMFBR core

    International Nuclear Information System (INIS)

    Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.

    1991-01-01

    Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)

  1. Deposition of inhaled LMFBR-fuel-sodium aerosols in beagle dogs

    International Nuclear Information System (INIS)

    Hackett, P.L.; Mahlum, D.D.; Briant, J.K.; Catt, D.L.; Peters, L.R.; Clary, A.J.

    1980-01-01

    Initial alveolar deposition of LMFBR-fuel aerosols in beagle dogs amounted to 30% of the inhaled activity, but only 5% of the total inhaled activity was deposited in dogs exposed to sodium-fuel aerosols. Aerosol deposition in the gastrointestinal tract amounted to 4% of the initial body burden of fuel-aerosol exposed dogs and 24% of the burden of animals receiving sodium-fuel aerosols. Preliminary analytical data for the dog exposures appear to agree with rodent data for deposition and distribution patterns of aerosols of similar sodium: fuel ratios

  2. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-11-22

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  3. LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

    This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

  4. Influence of leakage flow on the behaviour of gas behind a blockage in LMFBR subassembly geometry

    International Nuclear Information System (INIS)

    Fukuzawa, Y.

    1980-07-01

    Observations were made of the behaviour of gas behind a uniform porous 21% corner blockage within a pin-bundle of LMFBR subassembly geometry. The main parameter of the experiment was the leakage flow rate through the blockage. The behaviour of gas is significantly influenced by the leakage flow rate. The measured size and residence time of a gas cavity formed behind the blockage are shown and the mechanisms of the gas cavity dispersion by the leakage flow discussed by using a simple model of the liquid flow distribution behind the blockage. (orig.) [de

  5. LMFBR safety. 5. Review of current issues and bibliography of literature (1975--1976)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-06-08

    The current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA), are discussed. Bibliographic information on worldwide LMFBRs relative to the development and safety of the breeder reactor is presented for the period 1975 through 1976. The bibliography consists of approximately 1618 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Keyword, author, and permuted-title indexes are included for completeness.

  6. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-08-16

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  7. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  8. Techniques for the thermal/hydraulic analysis of LMFBR check valves

    International Nuclear Information System (INIS)

    Cho, S.M.; Kane, R.S.

    1979-01-01

    A thermal/hydraulic analysis of the check valves in liquid sodium service for LMFBR plants is required to provide temperature data for thermal stress analysis of the valves for specified transient conditions. Because of the complex three-dimensional flow pattern within the valve, the heat transfer analysis techniques for less complicated shapes could not be used. This paper discusses the thermal analysis techniques used to assure that the valve stress analysis is conservative. These techniques include a method for evaluating the recirculating flow patterns and for selecting appropriately conservative heat transfer correlations in various regions of the valve

  9. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

  10. Study of structural attachments of a pool type LMFBR vessel through seismic analysis of a simplified three dimensional finite element model

    International Nuclear Information System (INIS)

    Ahmed, H.; Ma, D.

    1979-01-01

    A simplified three dimensional finite element model of a pool type LMFBR in conjunction with the computer program ANSYS is developed and scoping results of seismic analysis are produced. Through this study various structural attachments of a pool type LMFBR like the reactor vessel skirt support, the pump support and reactor shell-support structure interfaces are studied. This study also provides some useful results on equivalent viscous damping approach and some improvements to the treatment of equivalent viscous damping are recommended. This study also sets forth pertinent guidelines for detailed three dimensional finite element seismic analysis of pool type LMFBR

  11. Recent development of a CEC'S elasto-plastic-creep cyclic benchmark programme relevant to LMFBR structural integrity

    International Nuclear Information System (INIS)

    Corsi, F.; Terzaghi, A.

    1984-01-01

    It's presented the programme of elasto-plastic benchmark calculations relevant to LMFBr, which started in 1977 with the support and coordination of the Commission of the European Communities (CEC) and the participation of nuclear engineering and manufacturing companies as well as nuclear research centers of France, Germany, Italy and the United Kingdom. (E.G.) [pt

  12. TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

    International Nuclear Information System (INIS)

    Magnaud, J.P.; Rouzaud, P.

    1985-09-01

    TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

  13. Effects of entrained gas on the acoustic detection of sodium boiling in a simulated LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Leavell, W.H.; Sides, W.H.

    1975-01-01

    The relationship between acoustic intensity of nucleate boiling and void fraction was studied in a simulated LMFBR fuel bundle. Results indicate that as the void fraction increases the detected intensity of nucleate boiling decreased until it was indistinguishable from background noise. (JWR)

  14. Detection of oscillatory components in noise signals and its application to fast detection of sodium boiling in LMFBR's

    International Nuclear Information System (INIS)

    Ehrhardt, J.

    1975-09-01

    In general, the surveillance of technical plants is performed by observating the mean value of measured signals. In this method not all information included in these signals is used. On the other hand - for example in a reactor - disturbances are possible which generate small oscillatory components in the measured signals. In general, these oscillatory components do not influence the mean value of the signals and consequently do not activate the conventional control system; however they can be found by analysis of the signal's noise component. For the detection of these oscillatory signals the observation of the frequency spectra of the noise signals is particularly advantageous because they produce peaks at the oscillation frequencies. In this paper a new detection system for the fast detection of suddenly appearing peaks in the frequency spectra of noise signals is presented. The prototype of a compact detection unit was developed which continuously computes the power spectral density (PSD) of noise signals and simultaneously supervises the PSD for peaks in the relevant frequency range. The detection method is not affected by the frequency dependance of the PSD and is applicable to any noise signal. General criteria were developed to enable the determination of the optimal detection system and its sensitivity. The upper limits of false alarm rate and detection time were taken into account. The detection criteria are applicable to all noise signals with approximately normally distributed amplitudes. Theoretical results were confirmed in a number of experiments; special experimental and theoretical parameter studies were done for the optimal detection of sodium boiling in LMFBR's. Computations based on these results showed that local and integral sodium boiling can be detected in a wide core range of SNR 300 by observing fluctuations of the neutron flux. In this connection it is important to point out that no additional core instrumentation is necessary because the

  15. Model-based temperature noise monitoring methods for LMFBR core anomaly detection

    International Nuclear Information System (INIS)

    Tamaoki, Tetsuo; Sonoda, Yukio; Sato, Masuo; Takahashi, Ryoichi.

    1994-01-01

    Temperature noise, measured by thermocouples mounted at each core fuel subassembly, is considered to be the most useful signal for detecting and locating local cooling anomalies in an LMFBR core. However, the core outlet temperature noise contains background noise due to fluctuations in the operating parameters including reactor power. It is therefore necessary to reduce this background noise for highly sensitive anomaly detection by subtracting predictable components from the measured signal. In the present study, both a physical model and an autoregressive model were applied to noise data measured in the experimental fast reactor JOYO. The results indicate that the autoregressive model has a higher precision than the physical model in background noise prediction. Based on these results, an 'autoregressive model modification method' is proposed, in which a temporary autoregressive model is generated by interpolation or extrapolation of reference models identified under a small number of different operating conditions. The generated autoregressive model has shown sufficient precision over a wide range of reactor power in applications to artificial noise data produced by an LMFBR noise simulator even when the coolant flow rate was changed to keep a constant power-to-flow ratio. (author)

  16. Bulk coolant cavitation in LMFBR containment loading following a whole-core explosion

    International Nuclear Information System (INIS)

    Jones, A.V.

    1977-01-01

    An LMFBR core undergoing an explosion transmits energy to the containment in a series of pressure waves and the containment loading is determined by their cumulative effect. These pressure waves are modified by their interaction with the coolant through which they propagate. It is necessary to model both the induction of bulk cavitation by tension waves and the interaction of pressure waves with cavitated liquid in realistic containment loading calculations. This paper sets out the progress which has been achieved in such modelling and first indications for the effect of bulk coolant cavitation in LMFBR containment loading. Conclusions may be briefly summarised: 1) Bulk cavitation must be included in realistic containment loading calculations. 2) Phenomenological models of cavitated liquid without memory are inappropriate. The best approach is to model bubble dynamics directly, including at least momentum conservation and surface tension. 3) The containment loading resulting from a given explosion is sensitive to the state of preparation of the coolant. The number density of nucleation sites should therfore accompany the results of model tests. (Auth.)

  17. International Atomic Energy Agency specialist meeting on advances in structural analysis for LMFBR applications. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, M A; Roche, R L [eds.

    1983-05-01

    After the first session on review of national positions in the subject field, the meeting was divided into five technical sections as follows: General methods of Structural Analysis for Elevated Temperatures; Inelastic Analysis Methods for Elevated Temperature; Effects of Cyclic loading; Design Codes and Criteria; Instability and Buckling - Piping Analysis in the Creep Range. The conclusions of the Meeting were summarised as follows. In view of the complexity of material behaviour and the variability of properties from cast to cast, continuing work is needed to develop simple constitutive relations which ensure an acceptable level of conservatism for design evaluations. It is recognized that simplified design methods require further development for the assessment of ratchetting and shakedown of high temperature structures. More development work is required in the areas of buckling elastic follow up weld factors and these developments should take account of the imperfections inherent in welded fabrications. There is a need for realistic tests on welded structural features to validate design methods. It is proposed that this subject would be the topic of a future specialists meeting. In several countries, organisations are now preparing Guides and Codes concerning Structural Assessment for LMFBR components. It seems that some of these Codes could be drafted within a few years. In order to make a more realistic assessment of LMFBR structures, defect assessment in elevated temperature range must be considered.

  18. International Atomic Energy Agency specialist meeting on advances in structural analysis for LMFBR applications. Summary report

    International Nuclear Information System (INIS)

    Perez, M.A.; Roche, R.L.

    1983-05-01

    After the first session on review of national positions in the subject field, the meeting was divided into five technical sections as follows: General methods of Structural Analysis for Elevated Temperatures; Inelastic Analysis Methods for Elevated Temperature; Effects of Cyclic loading; Design Codes and Criteria; Instability and Buckling - Piping Analysis in the Creep Range. The conclusions of the Meeting were summarised as follows. In view of the complexity of material behaviour and the variability of properties from cast to cast, continuing work is needed to develop simple constitutive relations which ensure an acceptable level of conservatism for design evaluations. It is recognized that simplified design methods require further development for the assessment of ratchetting and shakedown of high temperature structures. More development work is required in the areas of buckling elastic follow up weld factors and these developments should take account of the imperfections inherent in welded fabrications. There is a need for realistic tests on welded structural features to validate design methods. It is proposed that this subject would be the topic of a future specialists meeting. In several countries, organisations are now preparing Guides and Codes concerning Structural Assessment for LMFBR components. It seems that some of these Codes could be drafted within a few years. In order to make a more realistic assessment of LMFBR structures, defect assessment in elevated temperature range must be considered

  19. Overview of current activities relevant to structural analysis on LMFBR in Japan

    International Nuclear Information System (INIS)

    Ichimiya, Masakazu

    1983-01-01

    This paper presents the structural analysis activities on LMFBR in Japan. The structural analysis activities on LMFBR in Japan have been made mainly toward the validation of the rules of high temperature structural design guide which is to be used for the design of Class 1 components for elevated temperature service of the prototype fast breeder reactor, Monju. Main features of these analyses are as follows. (1) Since the design by elastic analysis is intended in the high temperature structural design guide of Monju, a large progress has been made in the bounding technique for high temperature inelastic behaviors, particularly the elastic follow-up. (2) There has been a progress in the clarification of the creep behavior in order to evaluate creep damage adequately. (3) Analysis techniques and design rules for piping have been developed with considerable emphasis. In addition, buckling analyses were performed considering the thin structures with low internal pressure in Monju components. Further test and analysis were made on ratcheting. (author)

  20. Accommodation of potential hydrogen formation in LMFBR accidents

    International Nuclear Information System (INIS)

    Stepnewski, D.D.; Peak, R.D.; Mahaffey, M.K.

    1981-01-01

    Results of design verification tests for the FFTF reactor cavity liner system are presented which suggest that steel liners would retain their integrity even under certain hypothetical accident conditions, thus avoiding the formation of hydrogen. When liner failures are postulate in hypothetical reactor vessel meltthrough accidents, hydrogen levels can be controlled by an air purging system. The design of a containment purging and effluent scrubbing system is discussed

  1. Microprocessor-based integrated LMFBR core surveillance. Pt. 2

    International Nuclear Information System (INIS)

    Elies, V.

    1985-12-01

    This report is the result of the KfK part of a joint study of KfK and INTERATOM. The aim of this study is to explore the advantages of microprocessors and microelectronics for a more sophisticated core surveillance, which is based on the integration of separate surveillance techniques. After a description of the experimental results gained with the different surveillance techniques so far, it is shown which kinds of correlation can be done using the evaluation results obtained from the single surveillance systems. The main part of this report contains the systems analysis of a microcomputer-based system integrating different surveillance methods. After an analysis of the hardware requirements a hardware structure for the integrated system is proposed. The software structure is then described for the subsystem performing the different surveillance algorithms as well as for the system which does the correlation thus deriving additional information from the single results. (orig.) [de

  2. Quasi-steady state boiling downstream of a central blockage in a 19-rod simulated LMFBR subassembly (FFM bundle 3B)

    International Nuclear Information System (INIS)

    Hanus, N.; Fontana, M.H.; Gnadt, P.A.; MacPherson, R.E.; Smith, C.M.; Wantland, J.L.

    1976-01-01

    Results of sodium boiling tests in a centrally blocked 19-rod simulated LMFBR subassembly are discussed. The tests were part of the experimental series conducted with bundle 3B in the Fuel Failure Mockup (FFM) at ORNL

  3. Review of activities concerning sodium water reaction in LMFBR

    International Nuclear Information System (INIS)

    Rajput, A.K.

    1984-01-01

    This paper presents a review of activities concerning safety engineering programme for steam generators of FBT reactor in India. Leak rate and its effect and leak detection system are briefly discussed

  4. Probabilistic Accident Progression Analysis with application to a LMFBR design

    International Nuclear Information System (INIS)

    Jamali, K.M.

    1982-01-01

    A method for probabilistic analysis of accident sequences in nuclear power plant systems referred to as ''Probabilistic Accident Progression Analysis'' (PAPA) is described. Distinctive features of PAPA include: (1) definition and analysis of initiator-dependent accident sequences on the component level; (2) a new fault-tree simplification technique; (3) a new technique for assessment of the effect of uncertainties in the failure probabilities in the probabilistic ranking of accident sequences; (4) techniques for quantification of dependent failures of similar components, including an iterative technique for high-population components. The methodology is applied to the Shutdown Heat Removal System (SHRS) of the Clinch River Breeder Reactor Plant during its short-term (0 -2 . Major contributors to this probability are the initiators loss of main feedwater system, loss of offsite power, and normal shutdown

  5. Cover gas purification in the German LMFBR-programme

    International Nuclear Information System (INIS)

    Schillings, K.-L.; Wagner, J.; Stade, K. Ch.

    1987-01-01

    A specific problem of sodium-cooled reactor plants is the purity of the noble gas argon which is used to protect the liquid alkali metal sodium in its systems in order to avoid or reduce disagreeable reactions between sodium and gaseous compounds like moisture or air and organic products like oil and grease. But as this contact cannot completely be excluded, we have to recycle such soiled cover gas. Simultaneously this procedure has to correct the release of radioactivity. Therefore the cover gas purification of primary systems of reactor plants contains the removal of the inorganic chemical disposal and of the nuclear waste. (author)

  6. Key asset - inherent safety of LMFBR Pool Plant

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Lancet, R.T.; Mills, J.C.

    1984-04-01

    The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

  7. Air-cleaning devices for vented filtered LMFBR containment

    International Nuclear Information System (INIS)

    Muhlestein, L.D.; Hilliard, R.K.

    1982-07-01

    An effort lasting several years is summarized which evaluated, developed and tested air cleaning devices for potential use in breeder reactor containment venting applications. State-of-technology evaluations were completed for both a hypothetical head release accident and a primary vessel melt-through accident. Commercially available systems or components were tested which included HEPA filters, sand and gravel beds, and aqueous scrubbers. Large-scale demonstration tests were completed and results are presented for two- and three-stage conventional aqueous scrubber systems; and for a newly developed passive, submerged gravel scrubber

  8. Leak detection in LMFBR steam generators during operation

    Energy Technology Data Exchange (ETDEWEB)

    Dumm, K [INTERATOM, Bergisch Gladbach (Germany)

    1978-10-01

    This paper deals with the following four main aspects: requirement on the leak detection of the SNR-300 steam generators; the hydrogen detector of SNR-300; remarks on the disadvantage of impurity detectors; and the first approach to acoustic leak detection systems.

  9. Leak detection in LMFBR steam generators during operation

    International Nuclear Information System (INIS)

    Dumm, K.

    1978-01-01

    This paper deals with the following four main aspects: requirement on the leak detection of the SNR-300 steam generators; the hydrogen detector of SNR-300; remarks on the disadvantage of impurity detectors; and the first approach to acoustic leak detection systems

  10. Flow induced vibration of secondary piping of LMFBR

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.

    1977-01-01

    This paper presents a method for evaluating the characteristics of vibrations caused by internal flow in three-dimensional piping systems conveying high density fluids. The excitation of the circuit is mainly caused by the flow singularities, and it is shown that the problem may be reduced to calculate the response of the circuit to an acoustical pressure discontinuity, localised at each flow singularity. The paper is divided into two main parts: First part is devoted to the theoretical formulation of the coupled acoustical-mechanical problem and to its numerical solution by the french computer code TEDEL. Second part describes an experimental test of the method. The tested piping system consists of a stainless steel tube circuit comprising four 909 bends, conveying water. Vibrations are excited by a half closed gate valve. Satisfactory results are obtained concerning both the frequencies of resonance of the circuit and the level of the vibrations observed

  11. Data acquisition and control for LMFBR component testing

    International Nuclear Information System (INIS)

    Turner, G.E.

    1983-01-01

    Liquid Metal Fast Breeder Reactor components such as pumps, steam generators, and piping components are tested for their ability to withstand expected thermal transients of up to 25 0 F/s (14 0 C/s). The Energy Technology Engineering Center performs these tests in facilities specifically designed for that purpose. Although much of the instrumentation and controls for these test facilities are similar to those used in conventional process plants, the requirement to produce, control, and measure the effects of rapid thermal transients results in some not-so-conventional data acquisition and control system design criteria. This paper describes a typical data-acquisition system used at one of the ETEC test facilities and how the thermal transients are produced and controlled in the pump and steam-generator test facilities

  12. Theoretical approach of the seismic behaviour of a LMFBR core

    International Nuclear Information System (INIS)

    Parent, J.M.

    1983-01-01

    The vibratory behaviour of two one-degree-freedom mechanical systems, having known coupling and damping features, is investigated by transposition to two electrical circuits inductively coupled. This approach is carried out in two steps: the first one supposes that both circuits are identical, while the second one considers the general case where both circuits are unlike. It is shown that one or two resonant frequencies may occur, depending on the coupling and the damping conditions

  13. Method for pre-heating lmfbr type reactors

    International Nuclear Information System (INIS)

    Yokozawa, Atsushi; Kataoka, Hajime.

    1978-01-01

    Purpose: To enable pre-heating for the inside of the reactor container and the inside of the coolant recycling system with no additional facilities. Method: The coolant recycling system is composed of a heat exchanger, a mechanical pump, a check valve, a flow meter or the like and it is connected in series by way of a pipe line to a reactor container. The mechanical pump is used as a gas recycling device upon pre-heating and it is designed so that a blower such as a fan can be replaced for the impeller of the pump. The inside of the reactor container and the inside of the coolant recycling system is at first filled with an inert gas such as for use with cover gas. Then, nuclear fuels are loaded to attain criticality. Simultaneously, the blower is started and the control rods are operated while cooling the nuclear fuel with the inert gas thus to obtain heat required for pre-heating the pipe line or the like from the nuclear fuels. After the completion of the pre-heating, the liquid metal is charged. (Ikeda, J.)

  14. Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

    International Nuclear Information System (INIS)

    Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

    1985-04-01

    The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

  15. An experimental study on sodium-water reaction in the double pool LMFBR, (4)

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo; Uotani, Masaki; Akimoto, Tokuzo

    1989-01-01

    Double Pool type LMFBR set the rectangular cross-sectional steam generator (SGs) inside a secondary vessel. The initial spike pressure rise caused by large sodium-water reaction in SGs might be radiated into a large sodium pool in the secondary vessel. Therefore basic experiments on pressure wave propagation were carried out by generating pressure wave in water by mean of a set of drop hummer and piston. But the experimental apparatus in water was not convenience to simulate the structure near the bottom end of the SGs shell. In this reports, experiments were carried out by generating pulse sound pressure in air, and compared with the results pressure waves in water. (author)

  16. Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.; Gvildys, J.

    1977-01-01

    In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs

  17. Stability of inner baffle-shell of pool type LMFBR - experimental and theoretical studies

    International Nuclear Information System (INIS)

    Lebey, J.; Combescure, A.

    1987-01-01

    I pool type LMFBR, the primary coolant circuit, inside the main vessel, comprises a hot plenum separated from a cold plenum by an inner baffle. For Superphenix 1 reactor, it was judged advisable to built a double-shell baffle, each shell withstanding only one type of loading (primary loading for one shell, secondary loading for the other). Due to the size and intricacy of the structure, this design involves unnegligible supplementary costs and manufacturing difficulties. Thus, an alternative solution has been studied for future plants projects. It consists of a single shell baffle having a shape especially studied to sustain the two types of applied loadings (thermal plus primary loadings). Such a shape was calculated by NOVATOME, and it was decided to check the ability of methods of analysis to predict the ruin of this structure under primary loading. For this purpose, a mock-up has been tested, and the experimental results compared with the calculated ones. (orig./GL)

  18. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  19. Key technological issues in LMFBR high-temperature structural design - the US perspective

    International Nuclear Information System (INIS)

    Corum, J.M.

    1984-01-01

    The purpose of this paper is: (1) to review the key technological issues in LMFBR high-temperature structural design, particularly as they relate to cost reduction; and (2) to provide an overview of activities sponsored by the US Department of Energy to resolve the issues and to establish stable, standardized, and defensible structural design methods and criteria. Specific areas of discussion include: weldments, structural validation tests, simplified design analysis procedures, design procedures for piping, validation of the methodology for notch-like geometries, improved life assessment procedures, thermal striping, extension of the methodology to new materials, and ASME high-temperature Code reform needs. The perceived problems and needs in each area are discussed, and the current status of related US activities is given

  20. Status of the LMFBR thermo- and fluid-dynamic activities at KFK

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hofmann, F.; Rehme, K.

    1979-01-01

    The aim of the thermo- and fluiddynamic analysis is to determine the spatial velocity and temperature distributions in LMFBR-core elements with high accuracy. Knowledge of these data is a necessary prerequisite for determining the mechanical behavior of fuel rods and of structural material. Three cases are distinguished: Nominal geometry and steady state conditions; non-nominal geometry and quasi-steady state conditions; nominal geometry and non-steady state conditions. The present situation for the design calculations of fuel elements is based mainly on undisturbed normal operation. Most of the thermo- and fluiddynamic activities performed under the Fast Breeder Programme at KFK are related to this case. The present status of theoretical and experimental research work briefly presented in this paper, can be subdivided into the following main topics: 1. Physical and mathematical modelling of single phase rod bundle thermo- and fluiddynamics, 2. Experimental investigations on heat transfer and fluid flow in rod bundles

  1. Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

    International Nuclear Information System (INIS)

    Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

    1979-01-01

    This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

  2. Active acoustic leak detection for LMFBR steam generator. Sound attenuation due to bubbles

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Sakuma, Toshio

    1995-01-01

    In the steam generators (SG) of LMFBR, it is necessary to detect the leakage of water from tubes of heat exchangers as soon as it occurs. The active acoustic detection method has drawn general interest owing to its short response time and reduction of the influence of background noise. In this paper, the application of the active acoustic detection method for SG is proposed, and sound attenuation by bubbles is investigated experimentally. Furthermore, using the SG sector model, sound field characteristics and sound attenuation characteristics due to injection of bubbles are studied. It is clarified that the sound attenuation depends upon bubble size as well as void fraction, that the distance attenuation of sound in the SG model containing heat transfer tubes is 6dB for each two-fold increase of distance, and that emitted sound attenuates immediately upon injection of bubbles. (author)

  3. LMFBR post accident heat removal testing needs and conceptual design of a test facility

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

    1977-03-01

    A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

  4. An internal core catcher for a pool L.M.F.B.R. and connected studies

    International Nuclear Information System (INIS)

    Le Rigoleur, C.; Kayser, G.

    1979-01-01

    This paper describes an internal core catcher for a pool LMFBR. Problems related to retention of debris are studied: downward progression of debris from the core to the core catcher, debris bed formation, heat transfer below the core catcher plate and to the main vessel, mechanical resistance. These results are used to estimate the performances of the internal core catcher for a given core melt-down-accident. It is seen that for a uniform thickness layer on the core catcher the retention capabilities are satisfactory. Then the problem of a heap of debris is approached. Dryout is studied. Uncertainties related to the bed characteristics and problems of extended dryout beds are put forward

  5. Material properties requirements for LMFBR structural design: General considerations and data needs

    Energy Technology Data Exchange (ETDEWEB)

    Pugh, C E [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Purdy, C M [U.S. Energy Research and Development Administration (United States)

    1977-07-01

    A statement is given of material properties information needed in connection with the structural design technology for liquid-metal fast breeder reactor (LMFBR) primary circuit components. Implementation of current analysis methods and criteria is considered with an emphasis on data and data correlations for performing elastic-plastic and creep analyses, for establishing allowable stress limits, and for computing creep-fatigue damage. Further development of the technology is discussed in relation to properties information. Emphasis is placed on improved constitutive equations for representing inelastic material behavior, on procedures for treating time-dependent fatigue, and on criteria for creep rupture. The properties are generally discussed without regard to specific alloys, since most categories of information are needed for each major structural material. Some sample experimental results are given for type 304 stainless steel and 2 1/4 Cr-1 Mo steel. (author)

  6. Wire-wrapped rod-bundle heat-transfer analysis for LMFBR

    International Nuclear Information System (INIS)

    Wong, C.N.C.; Todreas, N.E.

    1982-07-01

    Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities

  7. Finite element analysis of irradiation-induced dilation of the fuel subassembly duct in LMFBR

    International Nuclear Information System (INIS)

    Gao Fuhai; Fu Hao; Li Nan; Yang Kongli; Wang Mingzhen

    2013-01-01

    Background: The calculation of irradiation-induced dilation of the fuel subassembly duct in LMFBR is important for fast reactor core design.. Purpose: To investigate how to calculate the dilation by using finite element method (FEM). Methods: First, irradiation-induced creep and swelling material models are introduced. Then, a theoretical solution based on a simplified bending plate model is briefly given. Finally, a stress update scheme for the adopted material models is presented and furthermore embedded into ABAQUS user interface UMAT to conduct finite element analysis. Both solutions are compared and discussed. Results: FEM successfully predicts the duct dilation and its solution agrees well with theoretical one in small deformation. Conclusions: The proposed stress update scheme is effective, The accuracy of the theory solution declines when dilation becomes larger. The maximum stress occurs at the duct corner point, and the location has stress relaxation effect. (authors)

  8. Experimental plans for LMFBR cavity liner sodium spill test LT-1

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Newell, G.A.

    1976-01-01

    Reinforced concrete is an important material of construction in LMFBR cavities and cells. Steel liners are often installed on the concrete surfaces to provide a gastight seal for minimizing air inleakage to inerted cell atmospheres and to protect the concrete from direct contact with sodium in the event of a sodium spill. In making safety assessment analyses, it is of interest to determine the adequacy of the liners to maintain their leaktightness during postulated accidents involving large sodium spills. However, data for basing analytical assessments of cell liners are very meager and an experimental program is underway at HEDL to provide some of the needed information. The HEDL cell liner evaluation program consists of both bench-scale feature tests and large-scale sodium spill demonstration tests. The plans for the first large-scale sodium spill test (LT-1) are the subject of this paper

  9. Role of fuel bubble phenomenology in assessment of LMFBR source term

    International Nuclear Information System (INIS)

    Cho, D.H.; Condiff, D.W.; Chan, S.H.

    1985-01-01

    Phenomenological aspects of a fuel vapor bubble formed in the sodium pool in a hypothetical severe accident are considered. The potential for fuel bubble collapse in the sodium pool is analyzed. It appears that for a wide range of hypothetical LMFBR accidents involving core vaporization, the fuel vapor bubble would likely be quenched and collapse prior to migration to the cover gas region. Such rapid quenching is due mainly to radiative heat transfer from the fuel bubble, coupled with the inherent capability of the sodium pool (large subcooling and high thermal conductivity) to dissipate thermal energy. Major uncertainty in the analysis concerns fuel vapor condensation phenomena at the sodium interface and its effect on the sodium surface radiation absorptivity. This is discussed in detail

  10. Material properties requirements for LMFBR structural design: general considerations and data needs

    International Nuclear Information System (INIS)

    Pugh, C.E.; Purdy, C.M.

    1977-01-01

    A statement is given of material properties information needed in connection with the structural design technology for liquid-metal fast breeder reactor (LMFBR) primary circuit components. Implementation of current analysis methods and criteria is considered with an emphasis on data and data correlations for performing elastic-plastic and creep analyses, for establishing allowable stress limits, and for computing creep-fatigue damage. Further development of the technology is discussed in relation to properties information. Emphasis is placed on improved constitutive equations for representing inelastic material behavior, on procedures for treating time-dependent fatigue, and on criteria for creep rupture. The properties are generally discussed without regard to specific alloys, since most categories of information are needed for each major structural material. Some sample experimental results are given for type 304 stainless steel and 2 1 / 4 Cr-1 Mo steel

  11. TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

    International Nuclear Information System (INIS)

    Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

    1981-01-01

    The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

  12. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  13. Laminar/transition sweeping flow-mixing model for wire-wrapped LMFBR assemblies

    International Nuclear Information System (INIS)

    Burns, K.F.; Rohsenow, W.M.; Todreas, N.E.

    1980-07-01

    Recent interest in analyzing the thermal hydraulic characteristics of LMFBR assemblies operating in the mixed convection regime motivates the extension of the aforementioned turbulent sweeping flow model to low Reynolds number flows. The accuracy to which knowledge of the mixing parameters is required has not been well determined, due to the increased influence of conduction and buoyancy effects with respect to energy transport at low Reynolds numbers. This study represents a best estimate attempt to correlate the existing low Reynolds number sweeping flow data. The laminar/transition model which is presented is expected to be useful in anayzing mixed convection conditions. However, the justification for making additional improvemements is contingent upon two factors. First, the ability of the proposed laminar/transition model to predict additional low Reynolds number sweeping flow data for other geometries needs to be investigated. Secondly, the sensitivity of temperature predictions to uncertainties in the values of the sweeping flow parameters should be quantified

  14. Analytical model for transient fluid mixing in upper outlet plenum of an LMFBR

    International Nuclear Information System (INIS)

    Yang, J.W.; Agrawal, A.K.

    1976-01-01

    A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the outlet plenum of an LMFBR. The maximum penetration of core flow is used as the criterion for dividing the sodium region into two mixing zones. The model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of by-pass flow into the plenum. The results of numerical calculations indicate that effects of flow stratification, chimney height, metal heat capacity and by-pass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas do not play any significant role on sodium temperature

  15. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    International Nuclear Information System (INIS)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified

  16. Materials properties utilization in a cumulative mechanical damage function for LMFBR fuel pin failure analysis

    International Nuclear Information System (INIS)

    Jacobs, D.C.

    1977-01-01

    An overview is presented of one of the fuel-pin analysis techniques used in the CRBRP program, the cumulative mechanical damage function. This technique, as applied to LMFBR's, was developed along with the majority of models used to describe the mechanical properties and environmental behavior of the cladding (i.e., 20 percent cold-worked, 316 stainless steel). As it relates to fuel-pin analyses the Cumulative Mechanical Damage Function (CDF) continually monitors cladding integrity through steady state and transient operation; it is a time dependent function of temperature and stress which reflects the effects of both the prior mechanical history and the variations in mechanical properties caused by exposure to the reactor environment

  17. Research report on design allowable values of structural materials for LMFBR

    International Nuclear Information System (INIS)

    1978-11-01

    The present report is composed of following two main parts. i) review and re-evaluation on test results by FCI Sub-committee studies, performed from 1973 to 1976, ii) review on procedures for determining design allowable values of structural materials for LMFBR components. Re-evaluation works have been made on monotonic tensile properties at elevated temperatures, creep and creep rupture properties, creep-fatigue properties (strain rate and tensile strain hold time effects on strain fatigue properties at elevated temperatures) of Types 316 and 304 stainless steel and 2 1/4Cr-1Mo steel (base and weld metals) produced in Japan. In the first half of the present report, creep-fatigue test results obtained by FCI Sub-committee studies are subjected to re-evaluation by the present P-FCI Sub-committee. Reviews have been made on testing methods on FCI's-creep-fatigue experiments with other test data of the test materials; high temperature monotonic tensile data, creep and creep rupture data, and origin of the test materials. The data of FCI studies are compared with other reference data obtained by several Japanese laboratories. In the latter half of the present report, procedures including ASME's are reviewed for setting design allowable values for LMFBR components on the basis of high temperature strength properties obtained with materials produced in Japan. A creep rupture data of Japanese steels are issued and examined to make proposal for a design allowable stress of S sub(t) through parameter survey. (author)

  18. Structural dynamics in LMFBR containment analysis. A brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper gives a brief survey of the computational methods and codes available for LMFBR containment analysis. The various numerical methods commonly used in the computer codes are compared. It provides the reactor engineers to up-to-date information on the development of structural dynamics in LMFBR containment analysis. It can also be used as a basis for the selection of the numerical method in the future code development. First, the commonly used finite-difference expressions in the Lagrangian codes will be compared. Sample calculations will be used as a basis for discussing and comparing the accuracy of the various finite-difference representations. The distortion of the meshes will also be compared; the techniques used for eliminating the numerical instabilities will be discussed and compared using examples. Next, the numerical methods used in the Eulerian formulation will be compared, first among themselves and then with the Lagrangian formulations. Special emphasis is placed on the effect of mass diffusion of the Eulerian calculation on the propagation of discontinuities. Implicit and explicit numerical integrations will be discussed and results obtained from these two techniques will be compared. Then, the finite-element methods are compared with the finite-difference methods. The advantages and disadvantages of the two methods will be discussed in detail, together with the versatility and ease of application of the method to containment analysis having complex geometries. It will also be shown that the finite-element equations for a constant-pressure fluid element is identical to the finite-difference equations using contour integrations. Finally, conclusions based on this study will be given

  19. Influence of fission product transport on delayed neutron precursors and decay heat sources in LMFBR accidents

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1981-01-01

    A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted

  20. Development of a coupled dynamics code with transport theory capability and application to accelerator driven systems transients

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Ama, T.; Palmiotti, G.; Taiwo, T.A.; Yang, W.S.

    2000-01-01

    The VARIANT-K and DIF3D-K nodal spatial kinetics computer codes have been coupled to the SAS4A and SASSYS-1 liquid metal reactor accident and systems analysis codes. SAS4A and SASSYS-1 have been extended with the addition of heavy liquid metal (Pb and Pb-Bi) thermophysical properties, heat transfer correlations, and fluid dynamics correlations. The coupling methodology and heavy liquid metal modeling additions are described. The new computer code suite has been applied to analysis of neutron source and thermal-hydraulics transients in a model of an accelerator-driven minor actinide burner design proposed in an OECD/NEA/NSC benchmark specification. Modeling assumptions and input data generation procedures are described. Results of transient analyses are reported, with emphasis on comparison of P1 and P3 variational nodal transport theory results with nodal diffusion theory results, and on significance of spatial kinetics effects

  1. Characterization of dynamic loads on the LMFBR rotating shield

    International Nuclear Information System (INIS)

    Morris, E.

    1979-01-01

    The rotating shields structure is a potential weak point of some current designs of primary containment against postulated whole core explosions. The calculation of the effect of transient loads on this structure, resulting from such an explosion, is therefore important in developing a safety case. The transient loads are usually calculated by computer codes such as ASTARTE, SEURBNUK, REXCO or ICECO and the effect of these loads on the structure by a suitable finite element code. Such procedure can be lengthly and costly. The present paper proposed a procedure which allows the consequences of changes in the transient loads, resulting from design changes for example, to be quickly and simply gauged. The load-impulse method of characterizing dynamic response of a structural system is well established. Provided loads with a similar temporal variation are compared, it can be shown that the dynamic response depends on only two features of the load, an average load and a time intregrated load or impulse. The scope of this approach has been extended by Youngdahl who has shown, for structures which deform in a rigid-plastic manner, that complex laoding histories can be equated to a rectangular form of loading, in a precise manner for simple structures and in an approximate manner for more complicated structures. This paper proposes that the failure characteristics of the rotating shields for which extensive plastic deformation is involved, be calculated for rectangular type loadings. The complex transient loadings calculated for various explosions and various changes in the primary vessel design can then be reduced to an equivalent rectangular form and the consequencial response of the shields structure deduced. (orig.)

  2. Aerosol generation and filter behaviour in sodium fires. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Boehm, L; Jordan, S

    1975-11-01

    In the scope of a long-term program (a) aerosol-formation rates during Na fires, (b) the behavior of Na aerosols in a closed system, and (c) the filtration of Na aerosols were investigated. These experiments in the ABRAUS facility should simulate the behaviour of Na aerosols after an accident in the inner and outer containment of the sodium-cooled fast Reactor SNR 300. At the conditions of the inner-containment (0.7% oxygen content in the atmosphere) aerosol-concentrations by Na - 0/sub 2/ reactions of 1 - 10 g/m/sup 3/ are possible. At the conditions of the outer-containment (21% 0/sub 2/-content) aerosol-concentrations at Na fires of 10 - 50 g/m/sup 3/ have been measured. The aerosol-formation rates are proportional to the 0/sub 2/-concentration: the rate at 21% 0/sub 2/-concentration is about 10-times higher than the rate at 0.7% 0/sub 2/. The aerosol formation rate was determined to 20 kgNa/m/sup 2/h at 21% 0/sub 2/. The behaviour of sand-bed-filters was investigated. A sand-bed-filter arrangement was developed which is better than HEPA-standard of fiberglas-filters concerning efficiency (better than 99.99) and load capacity (about 500 g Na/sub 2/0/sub 2//m/sup 2/). Beyond that sand-bed-filters resist high pressure- and temperature-peaks. Liquid Na aerosols are filtered with an efficiency better than 99.9%. A physical model was evaluated to explain pressure increase at the sand-bed-filter during load and penetration of the filter. The calculated values were in good agreement with experimental results.

  3. Monitoring of pipe displacements in French LMFBR SUPERPHENIX

    International Nuclear Information System (INIS)

    Foucher, N.; Debaene, J.P.; Renault, Y.; Blin, B.

    1993-01-01

    In order to check that pipe supports work properly and that the locking of snubbers or the loss of supports do not put a pipe in unacceptable loading conditions, a monitoring of the behaviour of the main pipes of SUPERPHENIX is planned. This monitoring system consists in measuring the displacements at selected points of the pipe by means of measuring rods and checking that these displacements remain inside allowable domains. These allowable domains are defined so that, if the displacements of the pipe are inside all these domains, the plant operator is sure that the stresses verify the allowable limits and then no additional inspection is carried out. In the opposite case, the operator will inspect the pipe in detail in order to determine the consequences and repair if necessary before restarting. Selection of points for monitoring was done with the to minimize the number of measures to be carried out and to use as far as possible the measuring rods that were installed to check that pipe displacements were consistent with what has been obtained in design calculations. However, it appears necessary to ensure that any incident occurring at any point of the pipe can be detected and, if necessary, additional measuring rods may be installed. An incident is said detectable if it induces on at least one measuring rod a deviation with respect to expected displacement not lower than 5 mm. It has been chosen so that small normal changes in measured displacements are not mistaken as incidents. The incidents that are supposed likely to occur are: 1) loss of a support which induces mainly primary stresses, 2) locking of a snubber which induces mainly secondary stresses. Monitoring of pipe displacements is a simple and effective way of checking that no damaging perturbation has occurred on the pipe. Calculations carried out on the DHR loops of SUPERPHENIX show that allowable domains of acceptable size may be obtained using a relatively small number of measuring rods. The method

  4. The radiological significance of transuranium radioisotopes released to the environment during operation of the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Barr, N.F.

    1976-01-01

    Estimates based on current knowledge and conservative assumptions indicate that release of transuranium elements from the Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycle are likely to proaduce population dose commitments small compared to those produced by naturally occurring alpha emitters and globally dispersed transuranium radioisotopes from tests of nuclear weapons in the atmosphere. Potential health consequences of these releases to current and future generations are estimated to be very small compared to risks associated with the production of energy by fossil fuels. The estimates are subject to a number of uncertainties imposed by lack of knowledge. Some of the uncertainties are not likely to be greatly reduced until LMFBR facilities are designed and operated. Others may be significantly reduced prior to facility design and operation. The paper discusses the sensitivity of the estimates to uncertainties and approches to reducing those uncertainties that strongly influence the estimates. (author)

  5. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    International Nuclear Information System (INIS)

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist

  6. Detailed design consideration on wire-spaced LMFBR fuel subassemblies under the effects of uncertainties and non-nominal geometries

    International Nuclear Information System (INIS)

    Hishida, H.

    1979-01-01

    This paper explains some analytical methods for evaluating the effects of deviation in subchannel coolant flow rate from the nominal value due to fuel pin bundle deflection and manufacturing tolerances and of inter-sub-channel coolant mixing and local temperature rise due to a wire-spacer on the hot spot temperature. Numerical results are given in each chapter with respect to a prototype LMFBR core. (author)

  7. Trip report: United States LMFBR Steam Generator Team. IAEA symposium, Bensberg, Germany, October 14--17, 1974

    International Nuclear Information System (INIS)

    1974-01-01

    Information is presented concerning steam generator design characteristics for the AFR reactor, SNR reactor, PHENIX reactor, SUPER PHENIX reactor, MONJU reactor, and BN-350 reactor; steam generator development programs for West Germany, France, Japan, U. K., and the U. S. S. R.; and the fabrication and inspection of steam generator components. Steam generator performance and maintenance requirements for operating LMFBR reactors are reviewed. (U.S.)

  8. Maintenance and repair of LMFBR steam generators: specialists` meeting, O-Arai Engineering Center, Japan, 4-8 June 1984. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The Specialists` Meeting on "Maintenance and Repair of LMFBR Steam Generators" was held in Oarai, Japan, from 4-8 June 1984. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by the Power Reactor and Nuclear Fuel Development Corporation of Japan. The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topical areas were discussed by participants: national review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; research and development work on maintenance and repair; and experience on steam generator maintenance and repair.

  9. A fundamental study on sodium-water reaction in the double pool LMFBR, (3)

    International Nuclear Information System (INIS)

    Uotani, Masaki; Kumagai, Hiromichi; Nishi, Yoshihisa; Yoshida, Kazuo

    1989-01-01

    The double pool LMFBR is an innovative reactor that Central Research Institute of Electric Power Industry proposed for the purpose of reducing the construction cost of FBRs, and it is characterized by immersing steam generators in the annular plenum formed between the primary vessel and the outer secondary vessel. Therefore, it is expected that the pressure behavior at the time of sodium-water reaction due to the breaking of heating tubes is largely different from the case of steam generators of conventional FBRs. In order to ensure the soundness of the primary vessel that containes the reactor core, it is necessary to sufficiently grasp the pressure behavior in the plenum, and this basic experiment and analysis are related to the pressure behavior due to piston motion that arises in the initial period of quasi-steady pressure. About 1/10 scale annular plenum was used, and the generation of reaction product gas was simulated by the release of nitrogen. When gas was released in the plenum, the highest pressure rise occurred in the initial period of release, and thereafter, periodic variation arose. The pressure waveform and the value of pressure rise as the results of the model analysis agreed well with the measured results. (K.I.)

  10. Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

    International Nuclear Information System (INIS)

    Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

    1988-01-01

    This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

  11. Neutronic feasibility of an LMFBR super long-life core (SLLC)

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Aoki, Katsutada; Arie, Kazuo; Tsuboi, Yasushi

    1988-01-01

    The LMFBR Super Long-Life Core (SLLC) concept has evolved over the last few years as one of the targets of innovative approaches for future FBR cost reduction. An idea for SLLC has been developed wherein the core lifetime is extended up to the plant life of about 30 years by applying the radially and axially multi-zoned core concept (the improved homogeneous core concept). The main purpose of the present study is placed on the evaluation of neutronic feasibility of the 1000 MWe class SLLC concept. The core size of the present SLLC, which is approximately 3 to 4 times as large as those of the current 1000 MWe core design, was determined by the limit of the maximum fast neutron fluence level, which was tentatively assumed to be 5-6x10 23 nvt as the target of the future development of advanced cladding materials. Emphasis is placed on the discussion of neutronic performances of cores with oxide fuels rather than metal or carbide fuels. The present study has shown that proper zoning of the different plutonium enrichment fuels at the initial core makes it possible to achieve small enough reactivity loss during 30-year burnup while satisfying mild variation of the subassembly power distributions using a higher fuel volume fraction of about 50%. Effects of important neutronic parameters on the core performances are also discussed. (orig.)

  12. LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

    International Nuclear Information System (INIS)

    Petrykowski, J.C.; Longest, A.W.

    1985-01-01

    The transport of uranium dioxide (UO 2 ) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO 2 pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs

  13. Post-accident fuel relocation and heat removal in the LMFBR

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

    1976-08-01

    Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

  14. Development of computer code models for analysis of subassembly voiding in the LMFBR

    International Nuclear Information System (INIS)

    Hinkle, W.

    1979-12-01

    The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered

  15. Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

    International Nuclear Information System (INIS)

    Langhans, J.

    1991-01-01

    Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs. (orig.)

  16. Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1979-12-01

    The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made.

  17. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  18. Studies needed to prevent the use of expansion bends in LMFBR intermediate heat exchangers

    International Nuclear Information System (INIS)

    Kayser, G.

    1975-01-01

    The LMFBR IHX built in France consist in a vertical tube bundle welded on 2 tube sheets. The secondary sodium flows down a central pipe to the lower collector then up through the tube bundle where it is heated. The solution of the problems raised by the presence of thermal stresses needed thorough studies and led to the following theoretical and experimental developments: 1. A computer code was written for structural analysis. The structure was divided in annular elements that could be studied by means of the elementary theory of shells and plates; and reduced elastic coefficients were given to the tube sheets to account for the presence of drilled holes. 2. An experimental study was undertaken to determine the reduced elastic coefficients of the tube sheets. 3. A computer code was written to study the primary sodium flow around the tube bundle, and experimental studies were made on a mockup, the fluid being water. 4. The results of the previous code were used to determine, by means of a code for thermal analysis, the temperature field in the bundle both in steady state and transient regimes. Up to now, many transients were performed and the Phenix heat exchangers have been operating quite satisfactorily; this seems to prove the design assumptions were correct. (Auth.)

  19. Thermohydraulic and thermal stress aspects of a porous blockage in an LMFBR fuel assembly

    International Nuclear Information System (INIS)

    Kuzay, T.M.; Marr, W.W.; Helenberg, H.W.; Ariman, T.; Wilson, R.E.; Pedersen, D.R.

    1979-01-01

    The current safety scenarios of Liquid Metal Fast Breeder Reactors (LMFBR) under local fault propagation include the study of a hypothetical accident initiated by the formation of an external debris porous blockage in a fuel subassembly. In this preliminary experimental and analytical investigation, a non-heat-generating porous blockage was postulated to cover 18 flow channels of a 37 pin Fast Test Reactor (FTR) type fuel subassembly. The axial extent of the blockage is 50 mm. The blockage material is stainless steel (SS 316) with 30 percent average porosity (percent void volume). The blockage and the pins were modeled with a finite element technique and the thermal field in the blockage was predicted. This thermal field was utilized to do a planar thermal stress analysis of the postulated blockage. To verify the analytical model and also to better understand the thermal-hydraulics of such a porous blockage out-of-pile tests were conducted in a sodium loop. Data from the out-of-pile tests was utilized to calibrate and improve the analytical model

  20. Evaluation of the structural integrity of LMFBR equipment cell liners: results of preliminary investigations

    International Nuclear Information System (INIS)

    McAfee, W.J.; Sartory, W.K.

    1976-01-01

    The behavior of a plane wall segment of a prototype liquid-metal-cooled fast breeder reactor (LMFBR) cell under conditions of a postulated massive sodium spill was studied. Sodium-concrete reaction calculations were performed assuming an initial flaw existed in the liner such that high-temperature sodium could penetrate to the concrete underneath. Based on existing sodium-concrete reaction rate data, bounding values were established for the maximum energy release per unit volume of concrete. The potential effect of this energy release on the deformation of the liner material was determined. The temperature buildup in the liner and the pressure differential across the flaw in the liner were also studied. The transient thermal and structural responses of the steel liner and backup concrete were analyzed in detail using the inelastic computer code ANSYS. The literature on the mechanical, physical, and general behavior properties of concrete at high temperature was reviewed. This review emphasized the structural behavior of concrete and did not cover the sodium-concrete reaction

  1. LMFBR Emergency Deployment Assuming 45 year Time-Delay Excess CO2 Removal

    International Nuclear Information System (INIS)

    Schenewerk, William Ernest

    2008-01-01

    Atmospheric CO 2 is presently increasing 2.25% per year in proportion to 2.25% per year exponential fossil fuel consumption increase. CO 2 removal is modeled as being proportional to 45-year-earlier CO 2 amount above 280 ppmV-C. This is: Exp (-0.0225/year * 45 years) = 0.36 fraction CO 2 removed from anthropological emissions, apparently by seawater. LMFBRs use 15 year doubling time. Deploying 30000 GWe atomic power by year-2080 results in CO 2 doubling year-2065 if World primary energy consumption continues increasing 2.25% per year. CO 2 remains roughly twice pre-industrial until year-2100. Beginning year-2080, CO 2 declines at 2.25% per year. CO 2 will presumably decline back to roughly the year-2000 value by year-2200 if the 45-year-delay sink remains effective. LMFBR and GCFR fleet expands to 30000 GWe by 2080. 1000 GWe LWR fleet consumes 5 Mt HM (Heavy Metal). Breeder first cores require 1 Mt HM. (author)

  2. Measurement of heat and momentum eddy diffusivities in recirculating LMFBR outlet plenum flows

    International Nuclear Information System (INIS)

    Manno, V.P.; Golay, M.W.

    1978-06-01

    An optical technique has been developed for the measurement of the eddy diffusivity of heat in a transparent flowing medium. The method uses a combination of two established measurement tools: a Mach-Zehnder interferometer for the monitoring of turbulently fluctuating temperature and a Laser Doppler Anemometer (LDA) for the measurement of turbulent velocity fluctuations. The technique is applied to the investigation of flow fields characteristic of the LMFBR outlet plenum. The study is accomplished using air as the working fluid in a small scale Plexiglas test section. Lows are introduced into both the 1 / 15 scale FFTF outlet plenum and the 3 / 80 scale CRBR geometry plenum at inlet Reynolds numbers of 22,000. Measurements of the eddy diffusivity of heat and the eddy diffusivity of momentum are performed at a total of 11 measurement stations. Significant differences of the turbulence parameters are found between the two geometries, and the higher chimney structure of the CRBR case is found to be the major cause of the distinction. Spectral intensity studies of the fluctuating electronic analog signals of velocity and temperature are also performed. Error analysis of the overall technique indicates an experimental error of 10% in the determination of the eddy diffusivity of heat and 6% in the evaluation of turbulent momentum viscosity. In general it is seen that the turbulence in the cases observed is not isotropic, and use of isotropic turbulent heat and momentum diffusivities in transport modelling would not be a valid procedure

  3. An appreciation of the events, models and data used for LMFBR radiological source term estimations

    International Nuclear Information System (INIS)

    Keir, D.; Clough, P.N.

    1989-01-01

    In this report, the events, models and data currently available for analysis of accident source terms in liquid metal cooled fast neutron reactors are reviewed. The types of hypothetical accidents considered are the low probability, more extreme types of severe accident, involving significant degradation of the core and which may lead to the release of radionuclides. The base case reactor design considered is a commercial scale sodium pool reactor of the CDFR type. The feasibility of an integrated calculational approach to radionuclide transport and speciation (such as is used for LWR accident analysis) is explored. It is concluded that there is no fundamental obstacle, in terms of scientific data or understanding of the phenomena involved, to such an approach. However this must be regarded as a long-term goal because of the large amount of effort still required to advance development to a stage comparable with LWR studies. Particular aspects of LMFBR severe accident phenomenology which require attention are the behaviour of radionuclides during core disruptive accident bubble formation and evolution, and during the less rapid sequences of core melt under sodium. The basic requirement for improved thermal hydraulic modelling of core, coolant and structural materials, in these and other scenarios, is highlighted as fundamental to the accuracy and realism of source term estimations. The coupling of such modelling to that of radionuclide behaviour is seen as the key to future development in this area

  4. An experimental study of the heterogeneous LMFBR core using FCA assemblies with axial internal blanket

    International Nuclear Information System (INIS)

    Nakano, M.; Iijima, S.; Shirakata, K.; Hirota, J.

    1980-01-01

    To investigate physics properties of the heterogeneous LMFBR core and to examine the reliability of the current data and method for heterogeneous core configuration, an experimental study has been made on FCA VII-3 assemblies which have an internal blanket (IB) at midplane of the cylindrical core. Systematic experiments were carried out on the heterogeneous cores whose IBs were different in composition and thickness. A homogeneous core was also built to compare the results with those obtained on the heterogeneous cores. The sodium-void worth is not sensitive to the composition of IB. The positive void worth in the core of the 40 cm IB is lowered by about 40% compared with that in the homogeneous core. The analysis was made using the JAERI-Fast Set Version II and the diffusion code CITATION. Directional diffusion coefficients were used to take account of the axial streaming. To evaluate transport effects, the S 4 calculation was made. Comparison between the calculated and experimental results reveals the following: ksub(eff) and Pu worth in the core are not well predicted for the heterogeneous core, although they are represented satisfactorily for the homogeneous core. Reaction rates sensitive to the low-energy neutron are underestimated in the IB when they are normalized in the core. Sodium-void worths are fairly well predicted. However, the positive void worth in the heterogeneous core is underestimated, while that in the homogeneous core is overestimated. (author)

  5. Effects of governing parameters on steady-state inter-wrapper flow in an LMFBR

    International Nuclear Information System (INIS)

    Moriya, Shoichi

    2001-01-01

    Hydraulic experiments were performed using a 1/8th scale rectangular model, based on a Japanese demonstration fast breeder reactor design, in order to study fundamental characteristics of interwrapper flows occurring under steady state conditions in an LMFBR. The steady state interwrapper flow of which direction was downward in the center region and upward in the peripheral region of a core barrel was observed because of the radial static pressure gradient in the upper part of the core barrel, produced by a core blockage effect resulting from an above core structure with a perforated skirt. Thermal stratification phenomena were moreover observed in the interwrapper region, created by the hot steady state interwrapper flow from an upper plenum and the cold leakage flow through the separated plate of the core barrel. The thermal interface was generated in higher part of the core barrel when the core blockage effect was smaller and Richardson number and the leakage flow rate ratio were larger. Significant temperature fluctuations occurred in the peripheral region of the core barrel, when the difference between the interface elevations in the center and peripheral regions of the core barrel was enough large. (author)

  6. Qualification testing program plan for SIMMER. A computer code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Basdekas, D.L.; Silberberg, M.; Curtis, R.T.; Kelber, C.N.

    1978-07-01

    The objective of SIMMER qualification testing program is to assure that the mathematical models and input parameters are derived from experimental data, which, on the basis of criteria still to be established, are representative of the phenomena and processes governing the progression of a CDA in an LMFBR. At the present time, the work to meet this objective can be classified into three general task areas as they relate to the use of SIMMER in CDA analysis: (1) The whole-core energetic disassembly accident, or the ''vessel problem'': The objective here is to predict the partition of the total energy release, by a postulated severe power excursion, between the primary containment and the energy absorbed through nondestructive dissipative processes. (2) Single subassembly accident: The objective here is to determine the pertinent phenomena and to develop the capability to assess the significance and likelihood that such accidents might propagate to involvement of larger fraction of the core. (3) The whole-core transition phase accident: The objective here is to advance the understanding of the phenomena and processes involved, so that reliable predictions can be made of the possible consequences of a CDA and the potential for further nuclear excursions through recriticality

  7. Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made

  8. Evaluation of organic coolants for the transportation of LMFBR spent fuel rods

    International Nuclear Information System (INIS)

    Arnold, C. Jr.

    1978-05-01

    The physical and chemical processes that are likely to occur when sodium coated LMFBR spent fuel rods are submerged in various aromatic organic coolants was defined by means of immersion experiments carried out with sodium coated 304 stainless steel coupons. Upon immersion of sodium coated coupons at 220 0 C in hydrocarbon type coolants such as Therminol 88, a mixture of terphenyls, not only was the metallic sodium retained on the coupon, but a carbonaceous coating formed on the surface of the sodium. In contrast, coolants that contained aromatic ether bonds, such as Dowtherm A, reacted with sodium at 220 0 C to form phenolate and other salts, which precipitated from the coolant in the form of a dark sludge. With Dowtherm A, removal of metallic sodium from the coupon was essentially complete in a matter of hours at temperatures of 160--220 0 C. Data on the rate and efficiency of sodium removal upon immersion in Dowtherm A at elevated temperatures were obtained. In addition the kinetics and chemistry of the sodium/Dowtherm A reaction were defined. Because sodium sludges are potentially incompatible with the containing structural materials and the fuel elements, it is recommended that sodium be removed prior to immersion in the coolant via reaction with benzoic acid; this method should be adaptable to the facilities at reactor sites. In aging studies Dowtherm A was found to be thermally stable up to 400 0 C and radiatively stable at ambient conditions. The combined effect of heat and radiation was not defined

  9. Experimental study on fast neutron streaming through grid-plate shield of a LMFBR

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Wakabayashi, Hiroaki; An, Shigehiro; Suzuki, Ikunori.

    1976-01-01

    Neutron streaming through the holes penetrating the grid plate shield of a prototype LMFBR was experimentally examined. The mockups of the grid plate shield were made of iron and aluminum. Experiments were conducted at the vertical column of ''YAYOI'', the fast neutron source reactor of University of Tokyo. A He-3 spectrometer was employed in order to measure the transmitted neutron spectrum, while rhodium and indium threshold foils were for the integral flux above specific energies and their spatial distributions in the form of reaction rates. The streaming factor for usual small bended holes is 1.28+-0.04 as to the integral neutron flux above 0.1 MeV and 1.30+-0.12 as to the reaction rate of indium foil. Use were made of the one and two dimensional neutron transport code ANISN and TWOTRAN for evaluation by computation. The reaction rates calculated by infinite slab model with ANISN code agree well with the experiments when normalized at the source point where neutrons are incident on the grid plate shield. (auth.)

  10. Analysis of transient thermal response in the outlet plenum of an LMFBR

    International Nuclear Information System (INIS)

    Yang, J.W.

    1976-05-01

    A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the upper outlet plenum of an LMFBR. The one-dimensional turbulent jet flow equations were solved to determine the maximum penetration of the core flow. The maximum penetration is used as the criterion for dividing the sodium region into two mixing zones. The lumped-parameter model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of bypass flow into the plenum. Numerical calculations were performed for two cases. The first case corresponds to a normal scram followed by flow coast-down. The second case represents the double-ended pipe rupture at the inlet of cold leg followed by reactor scram. The results indicate that effects of flow stratification, chimney height, metal heat capacity and bypass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas does not play any significant role on sodium temperature. This two-zone mixing model will be a part of the thermohydraulic transient code SSC

  11. Validation study of the COBRA-WC computer program for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Khan, E.U.; Bates, J.M.

    1982-01-01

    The COBRA-WC (Whole Core) computer program has been developed as a benchmark code to predict flow and temperature fields in LMFBR rod bundles. Consequently, an extensive validation study has been conducted to reinforce its credibility. A set of generalized parameters predicts data well for a wide range of geometries and operating conditions which include conventional (current generation LMFBRs) fuel and blanket assembly geometry in the forced, mixed, and natural convection regimes. The data base used for validating COBRA-WC was obtained from out-of-pile and in-pile tests. Most of the data was obtained in fully heated bundles with bundle power skew across flats up to 3:1 (max:min), Reynolds number between 500 and 80,000, and coolant mixed-mean temperature rise (δ anti T) in the range, 78 0 F less than or equal to δ anti T less than or equal to 340 0 F. Within the bundle, 95% of the predicted coolant temperature data points fall within +-25 0 F for 150 0 F less than or equal to δ anti T less than or equal to 340 0 F and within +-17 0 F for 78 0 F less than or equal to δ anti T less than or equal to 150 0 F

  12. Analysis and application of prestressed concrete reactor vessels for LMFBR containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Fistedis, S.H.; Bazant, Z.P.; Belytschko, T.B.

    1978-01-01

    An analytical model of a prestressed concrete reactor vessel (PCRV) for LMFBR and the associated finite element computer code, involving an explicit time integration procedure, is described. The model is axisymmetric and includes simulations of the tensile cracking of concrete, the reinforcement, and a prestressing capability. The tensile cracking of concrete and the steel reinforcement are both modeled as continuously distributed within the finite element. The stresses in the reinforcement and concrete are computed separately and combined to give an overall stress state of the composite material. Attention is given to the fact that cracks do not form instantaneously, but develop gradually. Thus, after crack initiation the normal stress is reduced to zero gradually as a function of time. Residual shear resistance of cracks due to aggregate interlock is also taken into account. Prestressing of the PCRV is modeled by special structural members which represent an averaged prestressing layer equivalent to an axisymmetric shell. The internal prestressing members are superimposed over the reinforced concrete body of the PCRV; they are permitted to stretch and slide in a predetermined path, simulating the actual tendons. The validity of the code is examined by comparison with experimental data. (Auth.)

  13. Studies of flow stratification in the hot plenum of an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Jones, P; Hickmott, S [Central Electricity Generating Board, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire (United Kingdom)

    1983-07-01

    The paper reviews work at Berkeley Nuclear Laboratories on the extent and effects of buoyancy in the hot plenum of an LMFBR. It summarizes the experimental, theoretical and numerical work has has been conducted to aid the understanding of the complex transient flows which occur following a reactor trip. The experimental work has been conducted in small-scale idealised geometries which isolate the essential features of the reactor flows and is not intended to provide detailed design data. An integral theory has been devised to describe the thermal hydraulics of negatively-buoyant jets. The predictions are shown to be in good agreement with the experimental results and emphasize the need to correctly represent the inlet velocity and temperature profiles. Some preliminary calculations with a transient, two-dimensional, finite-element code are compared with the experimental results. These calculations reproduce the overall features of the flows but not the details of the stratified interface. The development of turbulence models for stratified flows is seen as a fruitful area for further research. (author)

  14. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  15. Theory and use of GIRAFFE for analysis of decay characteristics of delayed-neutron precursors in an LMFBR

    International Nuclear Information System (INIS)

    Gross, K.C.

    1980-07-01

    The application of the computer code GIRAFFE (General Isotope Release Analysis For Failed Elements) written in FORTRAN IV is described. GIRAFFE was designed to provide parameter estimates of the nonlinear discrete-measurement models that govern the transport and decay of delayed-neutron precursors in a liquid-metal fast breeder reactor (LMFBR). The code has been organized into a set of small, relatively independent and well-defined modules to facilitate modification and maintenance. The program logic, the numerical techniques, and the methods of solution used by the code are presented, and the functions of the MAIN program and of each subroutine are discussed

  16. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    International Nuclear Information System (INIS)

    Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

    1977-01-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  17. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

    1977-04-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  18. Contribution of the CEC in structural analysis applied to LMFBR problems

    International Nuclear Information System (INIS)

    Larsson, L.H.; Terzaghi, A.

    1983-01-01

    This paper presents both the activity of DG XII in field of Codes and Standards (harmonization) and the research activity carried out at the JRC in Ispra. The first part describes the activity performed in the field of structural analysis by the Fast Reactor Coordinating Committee of the CEC and its Working Group Codes and Standards. This activity, which is aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, has made good progress in most areas. Results from recent inelastic and seismic benchmark calculations are presented as well as future computational exercises and investigations related to piping analysis, defect analysis, material behaviour and life prediction at elevated temperature. In the second part of the paper results of recent research and future plans in the area of structural mechanics at the JRC Ispra are discussed. In the past years, a large effort was devoted to the COVA (code validation) program intended to validate dynamic fluid/structure codes necessary for predicting the response of LMFBR containments. The main conclusions that can be drawn from COVA which finishes this year are presented, and some still open questions related to the prediction of containment response to an HCDA are discussed. The paper then describes the identification technique which is applied for the determination of constitutive equations for the dynamic behaviour of materials. In the field of fracture mechanics JRC has mostly concentrated its efforts on the elastic-plastic fracture toughness properties of irradiated austenitic steels. In the future, also dynamic ductile fracture problems will be investigated, for these a large dynamic test facility with a max. force of 5 MN will be used. The numerical analysis methods associated with these tests are discussed. (author)

  19. Synthesis Report on the understanding of failed LMFBR fuel element performance

    International Nuclear Information System (INIS)

    Plitz, H.; Bagley, K.; Harbourne, B.

    1990-07-01

    In the coarse of LMFBR operation fuel element failures cannot entirely be avoided as experienced during the operation of PFR, PHENIX and KNK II, where 44 failed fuel elements have been registered between 1978 and 1989. In earlier irradiations, post irradiation examinations showed mixed oxide pin diameter increases up to pin pitch distance, urging to stress reactor safety questions on the potential of fuel pin failure propagation within pin bundles. The chemical interaction of sodium with mixed oxide fuel is regarded to be the key for the understanding of failed fuel behavior. Valuable results on the failed fuel pin behavior during operation were obtained from the SILOE sodium loop test. Based on the bulk of experience with the detection of fuel pin failures, with the continued operation and with the handling of failed pins respectively elements, one can state: 1. All fuel pin failures have been detected securely in time and have been located. 2. Small defects are developing slowly. 3. Even large defects at end-of-life pins resulted in limited fuel loss. 4. Clad failures behave benign in main aspects. 5. The chemical interaction of sodium with mixed oxide is an important factor in the behavior of failed fuel pins, especially at high burnup. 6. Despite different pin designs and different operation conditions, on the basis of 44 failed elements in PFR, PHENIX and KNK II no pin-to-pin propagation was observed and fuel release was rather low, often not detectable. 7. In no case hazard conditions affecting reactor safety have been experienced

  20. Safety issues for LMFBR: important features drawn from the assessments of Superphenix

    International Nuclear Information System (INIS)

    Natta, M.

    2002-01-01

    Superphenix, which is built on the site of Creys-Malville, is still the biggest LMFBR plant that has been in operation. It is a pool type reactor, as Phenix and the RNR 1 500 and EFR projects. After the analysis of the preliminary safety (1974-1975), the construction was authorised by decree of the Prime Minister in 1977, the authorization for fuel loading and star-up to 3% was given by the minister of industry in July 1985 and full power was achieved in December 1986. The plant was operated until the end of December 1996, producing the equivalent of 320 EFPD, corresponding to half of the maximum barn-up of the first core. The plant was definitively stopped on the 20. of April 1998 by a decision of the French government. During this period of 25 years of licensing, construction and operation of Superphenix, others discussions and preliminary licensing procedures were started for new projects, mainly the RNR 1500 French project and the EFR European project. The operation of Superphenix was also marked by several incidents, which led to additional licensing procedures and important modifications. This period was also marked by an important work of research and development in the safety field, mostly related to the issues concerning hypothetical core disruptive accidents (HCDA) and sodium fires; further, this period was marked by the Three Mile Island accident in 1979 and the Chernobyl accident in 1986. The purpose of this paper is to present some items which were discussed during this period of 25 years and which should be of interest for future LMFBRs. In this presentation, we shall discuss the key issues concerning the safety criteria and options taken with respect to severe accidents, i.e. core melt accidents, giving details on some specific which are less known since they were assessed only lately for Superphenix, sometimes in connection with the on-going safety researches. (author)

  1. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    1980-01-01

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  2. Radiation protection/shield design: a need for a systems approach

    International Nuclear Information System (INIS)

    Disney, R.K.

    1977-01-01

    Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. The system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection is described, and the program developed to implement this approach is defined. In addition, the principal shielding design problems for LMFBR nuclear reactor systems are discussed in relation to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods are discussed

  3. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-01-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System is presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented. (author)

  4. Tool for generation of seismic floor response spectra for secondary system design

    International Nuclear Information System (INIS)

    Cardoso, Tarcisio F.; Almeida, Andreia A. Diniz de

    2009-01-01

    The spectral analysis is still a valuable method to the seismic structure design, especially when one focalizes the topics of secondary systems in large industrial installations, as nuclear power plants. Two aspects of this situation add their arguments to recommend the use of this kind of analysis: the random character of the excitation and the multiplicity and the variability of the secondary systems. The first aspect can be managed if one assumes the site seismicity represented by a power spectrum density function of the ground acceleration, and then, by the systematic resolution of a first passage problem, to develop a uniformly probable response spectrum. The second one suggests also a probabilistic approach to the response spectrum in order to be representative all over the extensive group of systems with different characteristics, which can be enrolled in a plant. The present paper proposes a computational tool to achieve in-structure floor response spectra for secondary system design, which includes a probabilistic approach and considers coupling effects between primary and inelastic secondary systems. The analysis is performed in the frequency domain, with SASSI2000 system. A set of auxiliary programs are developed to consider three-dimensional models and their responses to a generic base excitation, acting in 3 orthogonal directions. The ground excitation is transferred to a secondary system SDOF model conveniently attached to the primary system. Then, a uniformly probable coupled response spectrum is obtained using a first passage analysis. In this work, the ExeSASSI program is created to manage SASSI2000 several modules and a set of auxiliary programs created to perform the probabilistic analyses. (author)

  5. VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

    International Nuclear Information System (INIS)

    Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

    1980-01-01

    1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

  6. Evaluation of the LMFBR cover gas source term and synthesis of the associated R and D

    International Nuclear Information System (INIS)

    Balard, F.; Carluec, B.

    1996-01-01

    At the end of the seventies and the beginning of the eighties, there appeared a pressing need of experimental results to assess the LMFBR's safety level. Because of the urgency, analytical studies were not systematically undertaken and maximum credible cover gas instantaneous source terms (radionuclides core release fraction) were got directly from crude out-of-pile experiment interpretations. Two types of studies and mock-ups were undertaken depending on the timescale of the phenomena: instantaneous source terms (corresponding to an unlikely energetic core disruptive accident CDA), and delayed ones (tens of minutes to some hours). The experiments performed in this frame are reviewed in this presentation: 1) instantaneous source term: - FAUST experiments: I, Cs, UO2 source terms (FzK, Germany), - FAST experiments : pool depth influence on non volatile source term (USA), - CARAVELLE experiments: nonvolatile source term in SPX1 geometry (CEA, France); 2) delayed source term: - NALA experiments: I, Cs, Sr, UO2 source term (FzK, Germany), - PAVE experiments: I source term (CEA, France), - NACOWA experiments: cover gas aerosols enrichment in I and Cs (FzK, Germany) - other French experiments in COPACABANA and GULLIVER facilities. The volatile fission products release is tightly bound to sodium evaporation and a large part of the fission products is dissolved in the liquid sodium aerosols present in the cover gas. Thus the knowledge of the amount of aerosol release to the cover gas is important for the evaluation of the source term. The maximum credible cover gas instantaneous source terms deduced from the experiments have led to conservative source terms to be taken into account in safety analysis. Nevertheless modelling attempts of the observed (in-pile or out-of-pile) physico-chemical phenomena have been undertaken for extrapolation to the reactor case. The main topics of this theoretical research are as follows: fission products evaporation in the cover gas (Fz

  7. Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

    1980-05-01

    The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

  8. Simulation of heat and mass transfer processes in molten core debris-concrete systems. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Felde, D K

    1979-01-01

    The heat and mass transport phenomena taking place in volumetrically-heated fluids have become of interest in recent years due to their significance in assessments of fast reactor safety and post-accident heat removal (PAHR). Following a hypothetical core disruptive accident (HCDA), the core and reactor internals may melt down. The core debis melting through the reactor vessel and guard vessel may eventually contact the concrete of the reactor cell floor. The interaction of the core debris with the concrete as well as the melting of the debris pool into the concrete will significantly affect efforts to prevent breaching of the containment and the resultant release of radioactive effluents to the environment.

  9. Compilation of references, data sources and analysis methods for LMFBR primary piping system components

    International Nuclear Information System (INIS)

    Reich, M.; Esztergar, E.P.; Ellison, E.G.; Erdogan, F.; Gray, T.G.F.; Wells, C.W.

    1977-03-01

    A survey and review program for application of fracture mechanics methods in elevated temperature design and safety analysis has been initiated in December of 1976. This is the first of a series of reports, the aim of which is to provide a critical review of the theories of fracture and the application of fracture mechanics methods to life prediction, reliability and safety analysis of piping components in nuclear plants undergoing sub-creep and elevated temperature service conditions

  10. Solution of generalized control system equations at steady state

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1987-01-01

    Although a number of reactor systems codes feature generalized control system models, none of the models offer a steady-state solution finder. Indeed, if a transient is to begin from steady-state conditions, the user must provide estimates for the control system initial conditions and run a null transient until the plant converges to steady state. Several such transients may have to be run before values for control system demand signals are found that produce the desired plant steady state. The intent of this paper is (a) to present the control system equations assumed in the SASSYS reactor systems code and to identify the appropriate set of initial conditions, (b) to describe the generalized block diagram approach used to represent these equations, and (c) to describe a solution method and algorithm for computing these initial conditions from the block diagram. The algorithm has been installed in the SASSYS code for use with the code's generalized control system model. The solution finder greatly enhances the effectiveness of the code and the efficiency of the user in running it

  11. Monte-Carlo Modeling of Parameters of a Subcritical Cascade Reactor Based on MSBR and LMFBR Technologies

    CERN Document Server

    Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H

    2001-01-01

    Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.

  12. Frequency interpretation of hold-time experiments on high temperature low-cycle fatigue of steels for LMFBR

    International Nuclear Information System (INIS)

    Udoguchi, T.; Asada, Y.; Ichino, I.

    1975-01-01

    The effect of frequency or hold-time on the low-cycle fatigue strength of AISI 316 stainless steel and SCM 3 Cr--Mo steel for fuel cladding, piping, and other structural members of LMFBR is investigated under high temperature conditions. Push-pull fatigue tests are conducted in air under conditions of fully reversed axial strain-control with a tensile strain hold-time ranging fromm 0 to 120 min for AISI 316, and with a tensile and an equal compressive strain hold-time ranging from 0 to 995 s for SCM 3. In these tests, a decrease of fatigue life is observed as the hold-time is increased. An empirical formula is presented which can predict well the effect of hold-time on high temperature low-cycle fatigue life in terms of frequency. The formula is a little different from those in the literature

  13. A frequency interpretation of hold-time experiments on high temperature low-cycle fatigue of steels for LMFBR

    International Nuclear Information System (INIS)

    Udoguchi, T.; Asada, Y.; Ichino, I.

    1975-01-01

    The effect of frequency or hold-time on the low-cycle fatigue strength of AISI 316 stainless steel and SCM 3 Cr-Mo steel for fuel cladding, piping and other structural members of LMFBR is investigated under high temperature conditions. Push-pull fatigue tests are conducted in air under conditions of fully reversed axial strain-control with a tensile strain hold-time ranging from 0 to 120 min for AISI 316, and with a tensile and an equal compressive strain hold-time ranging from 0 to 995 s for SCM 3. In these tests, a considerable decrease of fatigue life is observed as the hold-time is increased. An empirical formula is presented which can predict well the effect of hold-time on high temperature low-cycle fatigue life in terms of frequency. The formula is a little different from those in the literature. (author)

  14. Monte-Carlo modeling of parameters of a subcritical cascade reactor based on MSBR and LMFBR technologies

    International Nuclear Information System (INIS)

    Bznuni, S.A.; Zhamkochyan, V.M.; Khudaverdyan, A.G.; Barashenkov, V.S.; Sosnin, A.N.; Polanski, A.

    2001-01-01

    Parameters are investigated of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k eff = 0.94 - 0.98), is capable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10 14 cm 12 · s -1 , in the fast booster zone is 5.12 · 10 15 cm 12 · s -1 at k eff = 0.98 and proton beam current I = 2.1 mA. (author)

  15. CRAB-II: a computer program to predict hydraulics and scram dynamics of LMFBR control assemblies and its validation

    International Nuclear Information System (INIS)

    Carelli, M.D.; Baker, L.A.; Willis, J.M.; Engel, F.C.; Nee, D.Y.

    1982-01-01

    This paper presents an analytical method, the computer code CRAB-II, which calculates the hydraulics and scram dynamics of LMFBR control assemblies of the rod bundle type and its validation against prototypic data obtained for the Clinch River Breeder Reactor (CRBR) primary control assemblies. The physical-mathematical model of the code is presented, followed by a description of the testing of prototypic CRBR control assemblies in water and sodium to characterize, respectively, their hydraulic and scram dynamics behavior. Comparison of code predictions against the experimental data are presened in detail; excellent agreement was found. Also reported are experimental data and empirical correlations for the friction factor of the absorber bundle in the entire flow range (laminar to turbulent) which represent an extension of the state-of-the-art, since only fuel and blanket assemblies friction factor correlations were previously reported in the open literature

  16. Development of LIFE4-CN: a combined code for steady-state and transient analyses of advanced LMFBR fuels

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Zawadzki, S.; Billone, M.C.; Nayak, U.P.; Roth, T.

    1979-01-01

    The methodology used to develop the LMFBR carbide/nitride fuels code, LIFE4-CN, is described in detail along with some subtleties encountered in code development. Fuel primary and steady-state thermal creep have been used as an example to illustrate the need for physical modeling and the need to recognize the importance of the materials characteristics. A self-consistent strategy for LIFE4-CN verification against irradiation data has been outlined with emphasis on the establishment of the gross uncertainty bands. These gross uncertainty bands can be used as an objective measure to gauge the overall success of the code predictions. Preliminary code predictions for sample steady-state and transient cases are given

  17. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    International Nuclear Information System (INIS)

    Kirk, W.L.

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters

  18. Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

    International Nuclear Information System (INIS)

    Blomquist, C.A.; Pierce, R.D.; Pedersen, D.R.; Ariman, T.

    1977-01-01

    The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. Thermal analyses were performed with the Argonne-modified version fo the general heat transfer code THTB, based on the instantaneous addition of 3200 0 K molten fuel with a decay heat of 9 W/gm and 1920 0 K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The stress analysis showed that the Inconel vessel would not fail from the pressure loading, it was also shown that brittle fracture of the tungsten liner from thermal gradients is unlikely. Therefore, the melt-down cup meets the structural design requirements. (Auth.)

  19. Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

    1982-01-01

    Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

  20. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, W.L. (comp.)

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

  1. Investigation of the physical and numerical foundations of two-fluid representation of sodium boiling with applications to LMFBR experiments

    International Nuclear Information System (INIS)

    No, H.C.; Kazimi, M.S.

    1983-03-01

    This work involves the development of physical models for the constitutive relations of a two-fluid, three-dimensional sodium boiling code, THERMIT-6S. The code is equipped with a fluid conduction model, a fuel pin model, and a subassembly wall model suitable for stimulating LMFBR transient events. Mathematically rigorous derivations of time-volume averaged conservation equations are used to establish the differential equations of THERMIT-6S. These equations are then discretized in a manner identical to the original THERMIT code. A virtual mass term is incorporated in THERMIT-6S to solve the ill-posed problem. Based on a simplified flow regime, namely cocurrent annular flow, constitutive relations for two-phase flow of sodium are derived. The wall heat transfer coefficient is based on momentum-heat transfer analogy and a logarithmic law for liquid film velocity distribution. A broad literature review is given for two-phase friction factors. It is concluded that entrainment can account for some of the discrepancies in the literature. Mass and energy exchanges are modelled by generalization of the turbulent flux concept. Interfacial drag coefficients are derived for annular flows with entrainment. Code assessment is performed by simulating three experiments for low flow-high power accidents and one experiment for low flow/low power accidents in the LMFBR. While the numerical results for pre-dryout are in good agreement with the data, those for post-dryout reveal the need for improvement of the physical models. The benefits of two-dimensional non-equilibrium representation of sodium boiling are studied

  2. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

  3. Internal fuel motion as an inherent shutdown mechanism for LMFBR accidents: PINEX-3, PINEX-2, and HUT 5-2A experiments

    International Nuclear Information System (INIS)

    Ferrell, P.C.; Porten, D.R.; Martin, F.J.

    1981-01-01

    The PINEX-2 experiment verified the concept of axial internal molten fuel motion within annular fuel, representing an inherent shutdown mechanism for hypothetical transient overpower excursions on the order of 5$/s. The PINEX-3 experiment, simulating a 50 cents/s transient overpower, showed that limitations on the effectiveness of fuel motion may arise from freezing of the fuel and blockage of the internal movement. Analysis of these experiments was performed to assess the physical processes that dominate fuel relocation potential and to apply them to prototypic LMFBR pin conditions. Results indicate that internal fuel motion should be reliable as a shutdown mechanism in LMFBR's for a range of reactivity insertion rates beyond presently available experimental data

  4. Basic analysis and a comparison of the characteristics GCFRs and the LMFBR with the thorium cycle in one-group diffusion theory

    International Nuclear Information System (INIS)

    Sabundjian, G.; Ishiguro, Y.

    1991-09-01

    A preliminary study of neutronics of thorium cycle fast breeder reactor has been done using simplified reactor models and analyses methods with the aim of finding a type of breeder reactor suitable for an efficient utilization of thorium that is abundant in Brazil. Basic methods of cross section processing and reactor calculation are studied and applied to analyse breeding characteristics of GCFRs and LMFBRs. The GCFR is fueled with oxide pins and cooled with helium. The LMFBR is fueled with thin metallic pins to achieve high power densities. Neutronics characteristics are determined as functions of the average power density and the fuel volume fraction. Results show that a high power density and a high fuel volume fraction are desirable to achieve short doubling times, that the GCFR is inferior to the LMFBR in regard to the doubling time and that the LMFBR can achieve reactor doubling times ten years with an average power density of ∼ 600MW/m 3 and fuel volume fraction of 40%. (author)

  5. Heat transfer performance of multi-layer insulation structure under roof-slab of pool-type LMFBR

    International Nuclear Information System (INIS)

    Kinoshita, I.; Yoshida, K.; Uotani, M.; Fukada, T.

    1988-01-01

    At the normal operation of the pool-type LMFBR, the free surface of liquid sodium at about 500 0 C is present below the roof-slab, separated by a space of the argon cover gas. The temperature of the roof-slab has to be maintained low and uniform in the horizontal direction for sufficient strength of the structure. Therefore, thermal insulation structures must be installed on the lower surface of the roof-slab. In addition to the installation of thermal insulator, forced cooling of the roof-slab is required for assured structural integrity of the roof-slab. The capacity of cooling equipment can be reduced by installation of structures with high thermal insulating performance. The objective of this study is to evaluate the thermal insulation characteristics of multi-layer type insulator installed below the roof-slab by analytically and experimentally. The analytical study is intended to evaluate the effect of number, distance and emissivity of layers on the heat transfer performances. This is treated as the one-dimensional heat transfer with natural convection, conduction and thermal radiation. In the experiments, we have evaluated effects of gap distances between adjacent thermal insulators placed below the roof-slab on the thermal insulation performances

  6. Some reliability targets affecting the necessary provisions for in-service inspection and monitoring of LMFBR engineering components

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1980-01-01

    The possible consequences of failure of primary and secondary sodium circuit components are discussed with particular reference to post incident fault diagnosis, remedial procedures and outage durations. The core support structures and steam generator units are identified as particularly important components in terms of economic consequence of their failure. Important safety considerations may also apply. Levels of reliability for core support and steam generator integrity, necessary to meet economic and certain safety criteria, are discussed and quantitative data is given. Possible failure and deterioration mechanisms which could result in unacceptable reductions in reliability are then identified for the core support and steam generator units. Following a consideration of the reliability targets and possible causes of loss of reliability, an appraisal is made of the necessary extent of in-service data to be obtained on component behaviour and condition. In-service inspection and monitoring methods that could be used to obtain this data are described. Consideration is given to UK and overseas inspection experience on LMFBR and other nuclear plant. (author)

  7. Comparison of numerical results with experimental data for single-phase natural convection in an experimental sodium loop. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Ribando, R.J.

    1979-01-01

    A comparison is made between computed results and experimental data for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL) used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility using a variety of initial conditions and testing parameters. Specifically, in this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow. The computer program used to analyze the test, LONAC (LOw flow and NAtural Convection) is an ORNL-developed, fast-running, one-dimensional, single-phase, finite-difference model used for simulating forced and free convection transients in the THORS loop.

  8. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  9. DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, R.D.

    1983-03-01

    A nodal method is developed for the solution of the neutron-diffusion equation in two- and three-dimensional hexagonal geometries. The nodal scheme has been incorporated as an option in the finite-difference diffusion-theory code DIF3D, and is intended for use in the analysis of current LMFBR designs. The nodal equations are derived using higher-order polynomial approximations to the spatial dependence of the flux within the hexagonal-z node. The final equations, which are cast in the form of inhomogeneous response-matrix equations for each energy group, involved spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces of the node. These equations are solved using a conventional fission-source iteration accelerated by coarse-mesh rebalance and asymptotic source extrapolation. This report describes the mathematical development and numerical solution of the nodal equations, as well as the use of the nodal option and details concerning its programming structure. This latter information is intended to supplement the information provided in the separate documentation of the DIF3D code.

  10. Main aspects of the design of a support structure of a LMFBR with particular reference to the explosive accident consequences

    International Nuclear Information System (INIS)

    Giuliano, V.; Lazzeri, L.

    1977-01-01

    The aim of this paper is a review of the main aspects of the design of a support structure of a LMFBR tank, with particular reference to the analysis of the non-linear dynamic behaviour of the structure in the plastic range under the effect of an explosive accident within the tank. The structure is composed by a L-shaped flange, which supports the tank, connected by means of nine square beams to a rigid box-type ring, fixed to the concrete. The plug of the tank is connected to the L-shaped figure by means of a group of SS bars. The non-linear dynamic analysis of the explosive accident has been carried out on a lumped mass model, with elastic-plastic elements which simulate main components of the support structure and tank. The impulsive load connected to the explosive accident has been modelled (on the basis of extensive comparative studies carried out) as two triangular pressure impulses acting on the plug and on the botton of the tank. A large amount of results, which describe displacements, velocities and accelerations of the plug, of the tank, and of the support structure, together with the forces and stresses acting on the main structural components are presented and discussed, with particular reference to the influence of the various parameters involved in the analysis. (Auth.)

  11. Some reliability targets affecting the necessary provisions for in-service inspection and monitoring of LMFBR engineering components

    Energy Technology Data Exchange (ETDEWEB)

    Bolt, P R [Fast Reactor Engineering, Plant Engineering Department, CEGB, Barnwood, Gloucester (United Kingdom)

    1980-11-01

    The possible consequences of failure of primary and secondary sodium circuit components are discussed with particular reference to post incident fault diagnosis, remedial procedures and outage durations. The core support structures and steam generator units are identified as particularly important components in terms of economic consequence of their failure. Important safety considerations may also apply. Levels of reliability for core support and steam generator integrity, necessary to meet economic and certain safety criteria, are discussed and quantitative data is given. Possible failure and deterioration mechanisms which could result in unacceptable reductions in reliability are then identified for the core support and steam generator units. Following a consideration of the reliability targets and possible causes of loss of reliability, an appraisal is made of the necessary extent of in-service data to be obtained on component behaviour and condition. In-service inspection and monitoring methods that could be used to obtain this data are described. Consideration is given to UK and overseas inspection experience on LMFBR and other nuclear plant. (author)

  12. Finite element program ARKAS: verification for IAEA benchmark problem analysis on core-wide mechanical analysis of LMFBR cores

    International Nuclear Information System (INIS)

    Nakagawa, M.; Tsuboi, Y.

    1990-01-01

    ''ARKAS'' code verification, with the problems set in the International Working Group on Fast Reactors (IWGFR) Coordinated Research Programme (CRP) on the inter-comparison between liquid metal cooled fast breeder reactor (LMFBR) Core Mechanics Codes, is discussed. The CRP was co-ordinated by the IWGFR around problems set by Dr. R.G. Anderson (UKAEA) and arose from the IWGFR specialists' meeting on The Predictions and Experience of Core Distortion Behaviour (ref. 2). The problems for the verification (''code against code'') and validation (''code against experiment'') were set and calculated by eleven core mechanics codes from nine countries. All the problems have been completed and were solved with the core structural mechanics code ARKAS. Predictions by ARKAS agreed very well with other solutions for the well-defined verification problems. For the validation problems based on Japanese ex-reactor 2-D thermo-elastic experiments, the agreements between measured and calculated values were fairly good. This paper briefly describes the numerical model of the ARKAS code, and discusses some typical results. (author)

  13. Viljandi Vana kalmistu kuulsused / Kadi Sassi

    Index Scriptorium Estoniae

    Sassi, Kadi

    2005-01-01

    Kunstnik F. B. Dörbeck (1799-1835), kunstnik G. Mooste (1885-1957), kunstnik V. Ormisson (1892-1941), kunstnik P. Kondas (1900-1985), kirjanik Fr. Kuhlbars (1841-1924), kirjanik A. Rennit (1860-1936), lavastaja ja näitleja H. Gleser (1893-1932), tantsija I. Urbel (1900-1983), fotograaf J. Riet (1873-1952), skulptor A. L. Weizenberg (1837-1921), professor Fr. Dryer (1879-1934)

  14. Louvre omadega sassis / Maria-Kristiina Soomre

    Index Scriptorium Estoniae

    Soomre, Maria-Kristiina, 1978-

    2002-01-01

    Prantsuse valitsuskomisjoni muuseumiauditi tulemustest Louvre'is (direktor Henri Loyrette). Muuseumi probleemide põhjuseks on külastajate arvu kahekordistumine peale 15 aastat kestnud ümberehitust. Juhtkond on ette võtnud muuseumi reformimise

  15. Sodium-water wastage and reactions program performed by general electric in support of the US. AEC LMFBR steam generator development

    International Nuclear Information System (INIS)

    Greene, D.A.

    1975-01-01

    This paper constitutes an interim report on the sodium-water reaction programs performed, using the GE-SOWAT, GE-SMALL LEAK BEHAVIOR RIG, and GE-PTTR facilities in support of LMFBR steam generator development and its application to the Clinch River Breeder Reactor Plant. Test data from these rigs are presented, including wastage data as a function of water injection rate, sodium temperature, and orifice geometry. Initial results for self-wastage of defects under prototypical conditions, and from proof-of-principle tests of a protected heat transfer tube concept are also presented. An analytical basis for wastage phenomena is suggested. (author)

  16. Sodium-water wastage and reactions program performed by general electric in support of the US. AEC LMFBR steam generator development

    Energy Technology Data Exchange (ETDEWEB)

    Greene, D A

    1975-07-01

    This paper constitutes an interim report on the sodium-water reaction programs performed, using the GE-SOWAT, GE-SMALL LEAK BEHAVIOR RIG, and GE-PTTR facilities in support of LMFBR steam generator development and its application to the Clinch River Breeder Reactor Plant. Test data from these rigs are presented, including wastage data as a function of water injection rate, sodium temperature, and orifice geometry. Initial results for self-wastage of defects under prototypical conditions, and from proof-of-principle tests of a protected heat transfer tube concept are also presented. An analytical basis for wastage phenomena is suggested. (author)

  17. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  18. On the hazard accumulation of actinide waste in a Pu-fueled LMFBR power economy with and without by-product actinide recycling

    International Nuclear Information System (INIS)

    Anselmi, L.; Caruso, K.; Hage, W.; Schmidt, E.

    1979-01-01

    The actinide waste arisings in terms of hazard potential for ingestion and inhalation are given for a Pu-fueled LMFBR Power Economy as function of decay time. The data were assessed for two simplified fuel cycles, one considering the recycling of by-product actinides and the other their complete discharge to the high-level waste. Two durations of nuclear power and several loss fractions of actinides to the waste were considered. The major contributors in form of chemical elements or isotopes to the actinide waste hazard built up during the nuclear power duration were identified for various decay intervals

  19. Comparative analysis of a hypothetical loss-of-flow accident in an irradiated LMFBR core using different computer models for a common benchmark problem

    International Nuclear Information System (INIS)

    Wider, H.U.; Devos, J.; Nguyen, H.; Goethem, G. Van.; Miles, K.J.; Tentner, A.M.; Pizzica, P.

    1989-01-01

    This report summarizes the results of an international exercise to compare whole-core accident calculations of the initiation phase of an unprotected LOF accident in a large irradiated LMFBR. The results for the accident phase before pin failure are in rather good agreement except for the fuel pin mechanics predictions. There are also some differences in the sodium boiling calculations but the voiding rates which are of key importance are very similar. The post - failure fuel motion and sodium voiding predictions show significant differences. However, the majority of these calculations agree that temporary fuel accumulations occur which increase the power beyond that caused by sodium voiding alone

  20. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  1. A study for optimal transmutation system

    International Nuclear Information System (INIS)

    Park, W.S.; Song, T.Y.; Shin, H.S.; Park, C.K.

    1996-01-01

    Couple of transmutation systems are being under investigation to design the optimal transmutation device. Several basic studies were performed for that objectives: (1) select the radioactive nuclides to be transmuted: (2) investigate the physical characteristics of each nuclide; (3) study the most favorable neutron energy environment for the transmutation. The existing LWR and LMFBR cores were found to be not a satisfiable ones in terms of transmutation rate itself. (author). 5 refs, 2 figs, 3 tabs

  2. Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, C. A. [Argonne National Lab., IL (United States); Ariman, T. [Univ. of Notre Dame, IN (United States); Pierce, R. D.; Pedersen, D. R. [Argonne National Lab., IL (United States)

    1977-07-01

    The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. The cup principal components are: 1. A 38 mm diameter tungsten spike which provides initial fuel quenching and prevents fuel boiling, 2. A 73 mm inside diameter tungsten liner to isolate the support vessel from the molten material high initial temperature, 3. An insulator which is an expedient for extending the experiment time, and 4. An Inconel 625 vessel which provides the structural support to withstand the thermal and pressure stresses. The spike, liner, and insulator are supported by a hemispherical tungsten end cap which fits inside the hemispherical bottom of the support vessel. This vessel is attached to the 316 stainless steel test train with an Inconel 750 wire-formed retaining ring. Thermal analyses were performed with the Argonne-modified version of the general heat transfer code THTB, based on the instantaneous addition of 3200/sup 0/K molten fuel with a decay heat of 9 W/gm and 1920/sup 0/K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The most severe heating occurs when there is no sodium flow outside the cup. For this case the sodium boils (approximately 1200/sup 0/K) and the Inconel vessel and tungsten liner temperatures are approximately 1250/sup 0/K and 2420/sup 0/K, respectively.

  3. Main aspects of the design of a support structure of a LMFBR with particular reference to the explosive accident consequences

    International Nuclear Information System (INIS)

    Giuliano, V.; Lazzeri, L.

    1977-01-01

    The aim of this paper is a review of the main aspects of the design of a support structure of a LMFBR tank, with particular reference to the analysis of the non-linear dynamic behavior of the structure in the plastic range under the effect of an explosive accident within the tank. The structure is composed by a L-shaped flange, which supports the tank, connected by means of nine square beams to a rigid box-type ring, fixed to the concrete. The plug of the tank is connected to the L-shaped flange by means of a group of SS bars. The non-linear dynamic analysis of the explosive accident has been carried out on a lumped mass model, with elastic-plastic elements which simulate main components of the support structure and tank. The impulsive load connected to the explosive accident has been modelled (on the basis of extensive comparative studies carried out) as two triangular pressure impulses has been the object of a parametric evaluation. The dynamic transient on the support structure during and after the explosive accident for each couple of pressure impulses has been analyzed by means of modified version of the NON SAP code running on a CDC 7600 computer. A large amount of results, which describe displacements, velocities and accelerations of the plug, of the tank, and of the support structure, together with the forces and stresses acting on the main structural components are presented and discussed, with particular reference to the influence of the various parameters involved in the analysis

  4. Overview of U.S. LMFBR structural materials mechanical properties program

    International Nuclear Information System (INIS)

    Horak, J.A.; Purdy, C.M.

    This paper presents the objective, scope, and status of the U.S. Department of Energy's Materials and Structures Program to develop a data base on mechanical properties of structural materials for out-of-core structures and components for LMFBRs. Information on the development of a reference data base on materials for the reactor system, reactor enclosure system, primary heat transport system, intermediate heat transport system, and steam generator system is included. In addition, the development of the data and analyses to account for the effects of temperature and stress, as well as water/steam, sodium, and radiation environments, is described. Plans for the development of alternative materials for future out-of-core applications are presented. (author)

  5. International piping benchmarks: use of the simplified code PACE 2. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Boyle, J T; Spence, J [University of Strathclyde (United Kingdom); Blundell, C [Risley Nuclear Power Development Establishment, Central Technical Services, Risley, Warrington (United Kingdom); ed.

    1979-05-15

    This report compares the results obtained using the code PACE 2 with the International Working Group on Fast Reactors (IWGFR) International Piping Benchmark solutions. PACE 2 is designed to analyse systems of pipework using a simplified method which is economical of computer time and hence inexpensive. This low cost is not achieved without some loss of accuracy in the solution, but for most parts of a system this inaccuracy is acceptable and those sections of particular importance may be reanalysed using more precise methods in order to produce a satisfactory analysis of the complete system at reasonable cost.

  6. Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

    International Nuclear Information System (INIS)

    Fox, J.N.; Wei, B.C.

    1978-02-01

    Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

  7. Comparison of LMFBR piping response obtained using response spectrum and time history methods

    International Nuclear Information System (INIS)

    Hulbert, G.M.

    1981-04-01

    The dynamic response to a seismic event is calculated for a piping system using a response spectrum analysis method and two time history analysis methods. The results from the analytical methods are compared to identify causes for the differences between the sets of analytical results. Comparative methods are also presented which help to gain confidence in the accuracy of the analytical methods in predicting piping system structure response during seismic events

  8. A comparison of nuclear power systems for Brazil using plutonium and binary cycles

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Fernandes, J.E.

    1985-01-01

    Nuclear power systems based on plutonium cycle and binary cycle are compared taking into account natural uranium demand and reactor combination. The systems start with PWR type reactors (U5/U8) and change to systems composed exclusively of FBR type reactors or PWR-FBR symbiotic systems. Four loading modes are considered for the PWR and two for the FBR. The FBR is either a LMFBR loaded with PU/U or a LMFBR loaded the binary way. A linear and a non-linear capacity growth and two different criteria for the FBR introduction are considered. The results show that a 100 GWe permanent system can be established in 50 years in all cases, based on 300000 t of natural uranium and in case of delay in the FBR introduction and if a thermal-fast symbiotic system is chosen, a binary cycle could be more advantageous than a plutonium cycle. (F.E.) [pt

  9. U.S. LMFBR steam generators materials considerations and waterside chemistry issues

    Energy Technology Data Exchange (ETDEWEB)

    Spalaris, C N

    1975-07-01

    This report describes the materials and waterside chemistry topics most relevant to the steam generator system for the Clinch River Breeder Reactor Plant. Development programs necessary to support or confirm design and plant operating conditions are summarized, together with selected test results obtained to date. (author)

  10. Design-related inherent safety characteristics in large LMFBR power plants

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Barthold, W.P.; Bowers, C.H.; Ferguson, D.R.; Prohammer, F.G.; van Erp, J.B.

    1976-01-01

    Design-related safety-enhancing features such as (1) extended pump coastdown, (2) increased negative reactivity feedbacks, (3) reduced sodium void reactivity, and (4) self-actuated shutdown systems are evaluated. Primary emphasis is placed on preventing or limiting core damage. Attention is also given to features aimed at mitigation of the energetics potential of hypothetical core-disruptive accidents

  11. U.S. LMFBR steam generators materials considerations and waterside chemistry issues

    International Nuclear Information System (INIS)

    Spalaris, C.N.

    1975-01-01

    This report describes the materials and waterside chemistry topics most relevant to the steam generator system for the Clinch River Breeder Reactor Plant. Development programs necessary to support or confirm design and plant operating conditions are summarized, together with selected test results obtained to date. (author)

  12. Little low-power boiling never hurt anybody

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1985-01-01

    Failures in the shutdown heat removal system of an LMFBR might lead to flow stagnation and coolant boiling in the reactor core. At normal operating power, the onset of sodium boiling will lead to film dryout and melting of the cladding and fuel within a few seconds. On the other hand, both calculations and currently available experimental data indicate that at heat fluxes corresponding to decay heat power levels, boiling leads to improved heat removal; and it limits the temperature rise in the fuel pins. Therefore, when setting safety criteria for decay heat removal systems, there is no reason to preclude sodium boiling per se because of heat removal considerations. As an example that illustrates the beneficial impact of coolant boiling, a case involving temporary loss of feedwater and staggered pump failures in a hypothetical, 1000-MWe loop-type reactor was run in the SASSYS-1 code

  13. Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

    International Nuclear Information System (INIS)

    Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

    1982-01-01

    Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components

  14. Flow induced vibrations of secondary piping of L.M.F.B.R

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.

    1977-01-01

    A method for evaluating the characteristics of vibrations caused by internal flow in three-dimensional piping systems conveying high density fluids, is presented. The excitation of the circuit is mainly caused by the flow singularities, and it is shown that the problem may be reduced to calculate the response of the circuit to an acoustical pressure discontinuity, localized at each flow singularity. The theoretical formulation of the coupled acoustical-mechanical problem and its numerical solution by the french computer code TEDEL, are given. An experimental test of the method is described. The tested piping system consists of a stainless steel tube circuit comprising four 90 0 bends, conveying water. Vibrations are excited by a half closed gate valve. Satisfactory results are obtained concerning both the frequencies of resonance of the circuit and the level of the vibrations observed

  15. Dynamic loadings of sodium-water reactions in LMFBR and fusion power designs

    Energy Technology Data Exchange (ETDEWEB)

    Chan, C. K.

    1977-07-01

    In liquid metal fast breeder reactor and lithium cooled fusion reactor, a sodium loop is being proposed to transfer heat from the primary coolant loop to the steam turbine cycle. Although by careful design and quality assurance programs, the probability for steam generator tube failure can be minimized, failure will still occur. The direct contact of sodium and water would cause a chemical reaction where hydrogen and sodium compounds are produced. This paper presents an evaluation of the potential hazards as a result of such a reaction. An analytical method is developed to investigate the extent of the reaction zone and the propagation of the pressure wave in the sodium system. In the calculation, the chemical reaction is assumed to be instantaneous, governed by the equation 2Na(l)+H/sub 2/O(l)..-->..Na/sub 2/O(l)+H/sub 2/(g)+31.4 K cal/gm. mole. Both the temperature and pressure rise in the reaction zone can be established from the energy balance and the equation of state for the gaseous product. As a consequence of the energy released, the chemical products suddenly expand with a high velocity. The expansion also generates a shock wave in both the water and the sodium systems. Results indicate that the reaction zone can expand in a rate of 1500 ft/sec and a shock wave with initial strength of 2300 atmospheres propagates with a speed of 8000 ft/sec into the sodium system. The propagating characteristics of the shock wave are obtained by solving the basic fluid equations. The shock wave decays rapidly, in the neighborhood of milliseconds, as soon as the reaction zone stops to expand. The decrease in the reaction zone pressure allows more water to react with the sodium and a second pulse is generated.

  16. Dynamic loadings of sodium-water reactions in LMFBR and fusion power designs

    International Nuclear Information System (INIS)

    Chan, C.K.

    1977-01-01

    In liquid metal fast breeder reactor and lithium cooled fusion reactor, a sodium loop is being proposed to transfer heat from the primary coolant loop to the steam turbine cycle. Although by careful design and quality assurance programs, the probability for steam generator tube failure can be minimized, failure will still occur. The direct contact of sodium and water would cause a chemical reaction where hydrogen and sodium compounds are produced. This paper presents an evaluation of the potential hazards as a result of such a reaction. An analytical method is developed to investigate the extent of the reaction zone and the propagation of the pressure wave in the sodium system. In the calculation, the chemical reaction is assumed to be instantaneous, governed by the equation 2Na(l)+H 2 O(l)→Na 2 O(l)+H 2 (g)+31.4 K cal/gm. mole. Both the temperature and pressure rise in the reaction zone can be established from the energy balance and the equation of state for the gaseous product. As a consequence of the energy released, the chemical products suddenly expand with a high velocity. The expansion also generates a shock wave in both the water and the sodium systems. Results indicate that the reaction zone can expand in a rate of 1500 ft/sec and a shock wave with initial strength of 2300 atmospheres propagates with a speed of 8000 ft/sec into the sodium system. The propagating characteristics of the shock wave are obtained by solving the basic fluid equations. The shock wave decays rapidly, in the neighborhood of milliseconds, as soon as the reaction zone stops to expand. The decrease in the reaction zone pressure allows more water to react with the sodium and a second pulse is generated

  17. Fuel-coolant interaction-phenomena under prompt burst conditions. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, H.; Young, M.F.; Reil, K.O.

    1979-01-01

    The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2 (190 MPa peak).

  18. Application of a bistable convection loop to LMFBR [liquid metal fast breeder reactor] emergency core cooling

    International Nuclear Information System (INIS)

    Anand, G.; Christensen, R.N.

    1990-01-01

    The concept of passive safety features for nuclear reactors has been developed in recent years and has gained wide acceptance. A literature survey of current reactors with passive features indicates that these reactors have some passive features but still do not fully meet the design objectives. Consider a current liquid-metal reactor design like PRISM. During normal operation, liquid sodium enters the reactor at ∼395 degree C and exits at ∼550 degree C. In the event of loss of secondary cooling with or without scram, the primary coolant (liquid sodium) initially acts as a heat sink and its temperature increases. For events without scram, the negative reactivity induced by the increase in temperature shuts the reactor down. When the average temperature of the sodium reaches ∼600 to 650 degree C, it overflows from the reactor vessel, activating the auxiliary cooling system. The auxiliary cooling system uses natural circulation of air around the reactor guard vessel. An alternative to the current design incorporates a bistable convection loop (BCL). The incorporation of the BCL concept remarkably improves the safety of the nuclear reactors. Application of the BCL concept to liquid-metal fast breeder reactors is described in this paper

  19. Present status and future plan on LMFBR steam generator safety study

    International Nuclear Information System (INIS)

    Tanabe, Hiromi; Kuroha, Mitsuo

    1985-01-01

    The results of the sodium-water reaction test which has been carried out for the safety evaluation of the water leak phenomena in the steam generators for the prototype FBR were summarized. This is related to the behavior of minute leak, the behavior of wear and damage propagation of neighboring tubes due to small and medium leaking jets, and the pressure/flow phenomena occurring at the time of large leak. Moreover, this is related to the development of analysis codes, the development of water leak-detecting system, and the development of the techniques for treating reaction products remaining in the system at the time of accidents. Also the research and development required hereafter for determining the basic specification of the steam generators for a demonstration FBR and future FBRs and reducing the cost were examined. The water leaks in the steam generators for FBRs have been reported in the Fermi reactor of USA, the PFR of Great Britain, the BN-350 of USSR, the Phenix reactor of France and so on. In Japan, the sodium-water reaction has been well understood, and the facilities for the countremeasures to it have been established. The sodium-water reaction phenomena, the present status of sodium-water reaction research and others are reported. (Kako, I.)

  20. Microelectronics in LMFBR: Activities sponsored by the Commission 1979-1983

    International Nuclear Information System (INIS)

    Nordwall, H.J. de

    1986-01-01

    These studies were designed to draw attention to signal handling techniques which have become applicable to the automatic control of reactors as a result of advances in microelectronics. The underlying concepts being explored are the extent to which more information may be obtained from existing sensors, whether changes in the state of a system may be acted upon automatically and the acceptability of stochastic methods of state estimation. If a given state may arise in a number of different ways, automatic action necessarily involves diagnosis of causes if any but the simplest actions, e.g. scram, is required. Up to now this diagnosis has been made by an operator, though a number of schemes intended to identify the primary cause of an abnormal state or to guide an operator to the most easily accessible safe state are in advanced states of development, e.g. at Halden, Westinghouse and GRS-Garching. The activity decribed is continuing, current emphasis being upon diagnostics using approaches based upon artificial intelligence concepts, system optimisation and quantification of the benefits of networking