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Sample records for hcda

  1. HCDA Simulation

    International Nuclear Information System (INIS)

    Zucchini, A.

    1985-01-01

    CASSIOPEE code for HCDA analysis in LMFBR has been successfully applied to the benchmark problem proposed by the CONT group of the CEC: the results of CASSIOPEE presented here are on the whole in satisfactory agreement with the average of the values obtained by the other main European codes

  2. HCDA bubble experiment, (2)

    International Nuclear Information System (INIS)

    Sakata, Kaoru; Mashiko, Hiroyuki; Oka, Yoshiaki; An, Shigehiro; Isozaki, Tadashi.

    1981-06-01

    An experiment simulating the behavior of the very large steam bubbles generated at the time of an accident of core collapse was carried out with a warm water tank, and the applicability of the theory of very small bubble disappearance known at present was examined. The bubbles generated in HCDA (hypothetical core disruptive accident) are expected to be very large, containing sodium, fuel, FP gas and so on, and play important role in the mechanism of emitting radioactive substances in the safety analysis of LMFBRs. In this experiment, the degree of subcool of the warm water pool, the initial radii of steam bubbles and the blowoff pressure of steam were taken as the parameters. The radius of the steam bubbles generated in the experiment was about 6.5 cm, and the state of disappearance was different above and below the degree of unsaturation of 10 deg C. Comparing the disappearance curve obtained by the experiment with the theory of disappearance of small bubbles, the experimental values were between inertia-controlled disappearance and heat transfer-controlled disappearance, and this result was able to be explained generally with the model taking the pressure change within steam bubbles into account. The rise of bubbles was also observed. (Kako, I.)

  3. Some remaining problems in HCDA analysis

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1981-01-01

    The safety assessment and licensing of liquid-metal fast breeder reactors (LMFBRs) requires an analysis on the capability of the reactor primary system to sustain the consequences of a hypothetical core-disruptive accident (HCDA). Although computational methods and computer programs developed for HCDA analyses can predict reasonably well the response of the primary containment system, and follow up the phenomena of HCDA from the start of excursion to the time of dynamic equilibrium in the system, there remain areas in the HCDA analysis that merit further analytical and experimental studies. These are the analysis of fluid impact on reactor cover, three-dimensional analysis, the treatment of the perforated plates, material properties under high strain rates and under high temperatures, the treatment of multifield flows, and the treatment of prestressed concrete reactor vessels. The purpose of this paper is to discuss the structural mechanics of HCDA analysis in these areas where improvements are needed

  4. Recent results on PEC reactor HCDA containment investigations

    International Nuclear Information System (INIS)

    Cenerini, R.; Palamidessi, A.; Verzelletti, G.

    1979-01-01

    The response of PEC reactor containement structures and of tank supporting arms to HCDA has been investigated by an explosive test on a refined 1:6 scaled mock-up. Experimental strains and pressures are compared with Astarte code calculations. (orig.)

  5. Potential hazard to secondary containment from HCDA-generated missiles and sodium fires

    International Nuclear Information System (INIS)

    Romander, C.M.

    1979-02-01

    The potential hazard of HCDA-generated missiles is analyzed, and the current status of the potential hazards of sodium fires is summarized. Simple analyses are performed to determine lower bounds on the HCDA energetics required to generate missiles that could reach the secondary containment structure of a 1000-MWe LMFBR. The potential missiles considered include the vessel head, components mounted on the head, and conrol rods

  6. Response of the primary piping loop to an HCDA

    International Nuclear Information System (INIS)

    Chang, Y.W.; Moneim, M.T.A.; Wang, C.Y.; Gvildys, J.

    1975-01-01

    The paper describes a method for analyzing the response of the primary piping loop that consists of straight pipes, elbows, and other components connected in series and subject to hypothetical core disruptive accident (HCDA) loads at both ends of the loop. The complete hydrodynamic equations in two-dimensions, that include both the nonlinear convective and viscous dissipation terms are used for the fluid dynamics together with the implicit ICE technique. The external walls of the pipes and components are treated as thin shells in which the analysis accounts for the membrane and bending strength of the wall, elastic-plastic material behavior, as well as large deformation under the effect of transient loading conditions. In the straight pipes, the flow is assumed to be axisymmetric; in the elbow regions, the two dimensions considered are the r and theta directions. The flow in the other components is also assumed to be axisymmetric; the components are modeled as a circular cylinder, in which the radius of the cylinder can be varied to conform with the outside shape of the component and the flow area inside can be changed independently from the outside shape. However, they must remain axially symmetric. The method is applied to a piping loop which consists of six elastic-plastic pipes and five rigid elbows connected in series and subjected to pressure pulses at both ends of the loop

  7. Development of the SCHAMBETA code for scoping analysis of HCDA

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Hahn, D. H

    2000-06-01

    A computer code, SCHAMBETA(Scoping Code for HCDA Analysis using Modified Bethe-Tait Method), is developed to investigate the core disassembly process following a meltdown accident in the framework of a mofified Bethe-Tait method as part of the scoping analysis work to demonstrate the inherent safety of conceptual designs of Korea Advanced Liquid Metal Reactor(KALIMER), A 150 Mwe pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. The methodologies adopted in the code ared particularly useful to perform various parametric studies for better understanding of core disassembly process of liquid metal fast reactors as well as to estimate upper-limit values of the energy release resulting from a power excursion. In the SCHAMBETA code, the core kinetics and hydraulic behavior of the KALIMER is followed over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion, starting at the time that the sodium-voided core reaches the melting temperature of the metallic fuels. For this purpose, the equations of state of pressure-energy density relationship are derived for the saturated-vapor as well as the solid liquid of metallic uranium fuel, and implemenmted into the formulations of the disassembly reactivity. Mathematical formulations are then developed, in the framework of Modified Bethe-Tait method, in a form relevant to utilize the improved equations of state as well as to consider Doppler effects, for scoping analysis of the super-prompt-critical power excursions driven by a specified rate of reactivity insertion.

  8. Roof loading and response following a HCDA in a pool-type reactor

    International Nuclear Information System (INIS)

    Lancefield, M.J.; Leigh, K.M.; Potter, R.; Staniforth, R.

    1979-01-01

    In a pool-type reactor the loading and response of the roof structure to a HCDA is important to safety analysis and design. The U.K. programme of experimental and theoretical work on this topic is described. Good progress in understanding and evaluating the complex processes has been made and this is illustrated by results from experimental and theoretical work. 5 refs

  9. Fluid-structure interaction analysis of a deck structure during a HCDA

    International Nuclear Information System (INIS)

    Kulak, R.F.

    1979-01-01

    Presented is an assessment of the structural integrity of the deck structure of a pool-type LMFBR during a Hypothetical Core Disruptive Accident (HCDA). During this accident the sodium above the core is propelled upward until it impacts against the deck structure. This hydrodynamic loading could produce (1) significant structural damage and (2) sodium leak paths. A finite-element model is used to study the deck dynamics during slug impact. By using the symmetry of the system, a sector model which accounts for the salient features of the system is developed. The main radial I-beam, component support I-beam and bottom annular plate are modeled using triangular plate elements. The concrete fill is modeled using hexahedral continuum elements. Using the above finite-element model the dynamics of the deck during a HCDA are investigated

  10. Confidence level in the calculations of HCDA consequences using large codes

    International Nuclear Information System (INIS)

    Nguyen, D.H.; Wilburn, N.P.

    1979-01-01

    The probabilistic approach to nuclear reactor safety is playing an increasingly significant role. For the liquid-metal fast breeder reactor (LMFBR) in particular, the ultimate application of this approach could be to determine the probability of achieving the goal of a specific line-of-assurance (LOA). Meanwhile a more pressing problem is one of quantifying the uncertainty in a calculated consequence for hypothetical core disruptive accident (HCDA) using large codes. Such uncertainty arises from imperfect modeling of phenomenology and/or from inaccuracy in input data. A method is presented to determine the confidence level in consequences calculated by a large computer code due to the known uncertainties in input invariables. A particular application was made to the initial time of pin failure in a transient overpower HCDA calculated by the code MELT-IIIA in order to demonstrate the method. A probability distribution function (pdf) for the time of failure was first constructed, then the confidence level for predicting this failure parameter within a desired range was determined

  11. Analysis of LMFBR containment response to an HCDA using a multifield Eulerian code

    International Nuclear Information System (INIS)

    Chu, H.Y.; Chang, Y.W.

    1977-01-01

    This paper describes a computer code, MICE (Multifield Implicit Continuous-fluid Eulerian Containment Code), which is being developed at Argonne National Laboratory (ANL) for the analysis of containment response to a hypothetical core distruptive accident (HCDA). The code is applicable to multifield flow problems where material fields are allowed to have penetrations. Reactor structures are treated as axisymmetrical shells and solved by the large-displacement and small-strain theory. Two sample problems have been performed using the MICE code. The first illustrates the relative motions of the material fields after the initiation of a core disassembly accident. Core support structure and core barrel are modelled as rigid obstacles. The second demonstrates the interactions between fluid and structures. Core expansion and reactor wall deformation at several instants are shown by the computer-generated film plots. (Auth.)

  12. Evaluation of Lagrangian, Eulerian, and arbitrary Lagrangian-Eulerian methods for fluid-structure interaction problems in HCDA analysis

    International Nuclear Information System (INIS)

    Chang, Y.W.; Chu, H.Y.; Gvildys, J.; Wang, C.Y.

    1979-01-01

    The analysis of fluid-structure interaction involves the calculation of both fluid transient and structure dynamics. In the structural analysis, Lagrangian meshes have been used exclusively, whereas for the fluid transient, Lagrangian, Eulerian, and arbitrary Lagrangian-Eulerian (quasi-Eulerian) meshes have been used. This paper performs an evaluation on these three types of meshes. The emphasis is placed on the applicability of the method in analyzing fluid-structure interaction problems in HCDA analysis

  13. Analysis of LMFBR containment response to an HCDA using a multifield Eulerian code

    International Nuclear Information System (INIS)

    Chu, H.Y.; Chang, Y.W.

    1977-01-01

    During a hypothetical core disruptive accident (HCDA), a core meltdown may cause the fuel cladding to rupture and the fuel fragments to penetrate into the sodium coolant. The heat in the molten fuel may cause the liquid sodium to boil, changing its phase. The interactions between materials are so complicated that a single-material model with homogenized material properties is not adequate. In order to analyze the above phenomena more realistically, a Multifield Implicit Continuous-Fluid Eulerian containment code (MICE) is being developed at Argonne National Laboratory (ANL) to solve the multifield fluid-flow problems in which the interpenetrations of materials, heat transfer, and phase changes are considered in the analysis. The hydrodynamics of the MICE code is based upon the implicit multifield (IMF) method developed by Harlow and Amsden. A partial donor-cell formulation is used for the calculation of the convective fluxes to minimize the truncation errors, while the Newton-Raphson method is used for the numerical iterations. An implicit treatment of the mass convection together with the equation of state for each material enables the method to be applicable to both compressible and incompressible flows. A partial implicit treatment of the momentum-exchange functions allows the coupling drag forces between two material fields to range from very weak to those strong enough to tie the fields completely. The differential equations and exchange functions used in the MICE code, and the treatment of the fluid and structure interactions as well as the numerical procedure are described. Two sample calculations are given to illustrate the present capability of the MICE code

  14. Extension of HCDA safety analysis to large PCRV containment structures

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Fistedis, S.H.; Bazant, A.P.; Belytschko, T.

    1977-01-01

    The introduction of PCRV as part of LMFBR containment also brings in the need for dynamic analysis. The conventional design would not be adequate for designing PCRVs subject to highly transient loading. The PCRV models described provide a valuable illustration of the design arrived at by conventional means and loaded dynamically by means of the transient computer code. In particular the value of the computer code is realized in the case of the loop-type model. The original pressure-volume source used for the illustrative examples corresponds to a total available energy of 2720 MW-s, if the expansion were continued down to one bar. Thus the energy source is considerably larger than the energy releases utilized in the currently prevailing LMFBR safety evaluations. Although the resultant pressure acting on the walls of the PCRV varies with the particular design, the two sample cases indicate that no apparent difficulty is encountered in designing the PCRV to sustain those higher loads. The computer code is established as a valuable tool to study the PCRV response. It serves as a means to locate critical areas of the particular design. Parameter- or sensitivity studies could thus be made with great efficiency. As an example, one interesting and somewhat unexpected result of the illustrations presented is the occurrence of circumferential cracks through the walls of the PCRV

  15. Transients in a reactor containment after a HCDA

    International Nuclear Information System (INIS)

    Ghosh, A.K.

    1984-01-01

    The consequences of a hypothetical core meltdown accident in a fast reactor is analysed. Shock waves are generated in the surrounding medium after the energy release, which is assumed to be instantaneous and at a point. After discussing the difference in the predicted and experimentally observed peak pressure a semi-empirical approach is taken to arrive at a better estimate. This defines the loading on the containment, which is idealised as a combination of a shell and a plate. To simplify the analysis the shell is assumed to be in plane strain and axial symmetry is assumed for both components. The shell response is evaluated by using elasticity theory. Numerical results are presented for peak overpressure and pressure-time history and dynamic stresses in the containment. (Author) [pt

  16. 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

  17. Deposition of aerosols formed by HCDA due to decay heat transport in inner containment atmospheres

    International Nuclear Information System (INIS)

    Vate, J.F. van de

    1976-01-01

    Coupling of decay heat transfer by aerosol-laden inner containment atmospheres with aerosol deposition from such atmospheres leads to useful and simple models for calculation of the time dependence of the aerosol mass concentration. Special attention is given to thermophoretic deposition (dry case) and condensation followed by gravitational deposition (wet case). Attractive features of the models are: 1) coagulation can be omitted and therefore complicated and doubtful calculations on coagulation are avoided, 2) material and particle size of the aerosol are not important for the aerosol decay rate, 3) the aerosol decay rate is related to the decay heat production which is known function of time, and the relevant part of it must be assessed usually for other purposes as well. (orig.) [de

  18. Debris bed cooling following an HCDA in a fast reactor. Final report

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.

    1983-01-01

    Natural convection within simulated core debris beds has been experimentally and theoretically investigated. The effect of heating method on bed behavior has been found to be important. For directly-heated beds, variations of the downward and upward power fraction and Nusselt numbers with bed loading, power density, particle size, overlying fluid layer height and top surface boundary condition have been determined. Generalized correlations for the upward and downward Nusselt numbers as functions of the internal Rayleigh number have been obtained. Particle tracing techniques have been used to visualize the flow patterns within the bed and overlying fluid layer. The temperature distributions within the bed and overlying fluid layer have also been measured. The experimental data have been compared with COMMIX-lA predictions. Poor agreement has been obtained for both the integral quantities, i.e. downward and upward power fractions and Nusselt numbers, as well as the steady state velocity and temperature distributions. The code does not correctly predict either the magnitude or even the trend of the data

  19. Influence of the representation models of the stress-strain law on the LMFBR structures in an HCDA

    International Nuclear Information System (INIS)

    Daneri, A.; Toselli, G.; Trombetti, T.; Blanchet, Y.; Louvet, J.; Obry, P.

    1981-08-01

    Most of analysis involved in mechanical calculations related to explosive accidents in fast breeder reactors are now aware of the inadequacy of certain rough stress-strain laws to representing the correct behaviour of vessel materials. Indeed stress waves along the vessel walls deform the material at a high strain rate with multiaxial loading or reverse loading. Recently different questions have been under investigation in France in this direction and the present study, performed in the frame of the agreement CNEN-CEA, is an example of the way how two very important factors (strain rate and strain hardening) may be taken into account in the constitutive equations of materials subject to dynamic deformations. Several parametric calculations have been carried out with the hydrodynamic structural codes ASTARTE-3/4 and SIRIUS, which are the Lagrangian validated code now available at the CNEN and CEA-Cadarache computing centres. Analysis was performed by comparing two reference calculations relating to the MARA 01 and COVA IT7 explosive tests with experimental data and with other calculations in which different values of the initial hardening and of the strain rates of the tank shell material were introduced. In general both codes give similar results; improvements of predicted axial and hoop strains and of impulses in water have been reached in certain cases but it is difficult to find a general trend and there is no ideal constitutive model: indeed the strain rate is not constant in time, in place and in direction and some parts of the vessels are uniaxially loaded while others are multiaxially loaded

  20. Influence of the representation models of the stress-strain law on the LMFBR structures in an HCDA

    International Nuclear Information System (INIS)

    Daneri, A.; Toselli, G.; Trombetti, T.; Blanche, Y.; Louvet, J.; Obry, P.

    1982-01-01

    Most of analysis involved in mechanical calculations related to explosive accidents in fast breeder reactor are now aware of the inadequacy of certain roug stress-strain laws to representing the correct behaviour of vessel materials. Indeed stress waves along the vessel walls deform the material at high strain rate with multiaxial loading or reverse loading. Recently different questions have been under investigation in France in this direction and the present study, performed in the frame of the agreement CNEN-CEA, is an example of the way how two very important factors (strain rate and strain hardening) may be taken into account in the consecutive equations of materials subject to dynamic deformations. Several parametric calculations have been carried out with the hydrodynamic structural codes ASTARTE 3/4 and SIRIUS, which are the Lagrangian validated codes now available at the CNEN and CEA Cadarache Computing Centres.Analysis was performed by comparing two reference calculations relating to the MARA 01 and COVA IT7 explosive tests with experimental data and with other calculations in which different values of the initial hardening and of the strain rates of the tank shell material were introduced. In general both codes give similar results; improvements of predicted axial and hoop strain and of impulses in water have been reached in certain cases but it is difficult to find a general trend and there is no ideal constitutive model: indeed the strain rate is not constant in time, in place and in direction and some parts of the vessels are uniaxially loaded while others are multiaxially loaded

  1. 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)

  2. Evaluation of explicit finite-difference techniques for LMFBR safety analysis

    International Nuclear Information System (INIS)

    Bernstein, D.; Golden, R.D.; Gross, M.B.; Hofmann, R.

    1976-01-01

    In the past few years, the use of explicit finite-difference (EFD) and finite-element computer programs for reactor safety calculations has steadily increased. One of the major areas of application has been for the analysis of hypothetical core disruptive accidents in liquid metal fast breeder reactors. Most of these EFD codes were derived to varying degrees from the same roots, but the codes are large and have progressed rapidly, so there may be substantial differences among them in spite of a common ancestry. When this fact is coupled with the complexity of HCDA calculations, it is not possible to assure that independent calculations of an HCDA will produce substantially the same results. Given the extreme importance of nuclear safety, it is essential to be sure that HCDA analyses are correct, and additional code validation is therefore desirable. A comparative evaluation of HCDA computational techniques is being performed under an ERDA-sponsored program called APRICOT (Analysis of PRImary COntainment Transients). The philosophy, calculations, and preliminary results from this program are described in this paper

  3. LMFBR aerosol release and transport program. Quarterly progress report, July--September 1975

    International Nuclear Information System (INIS)

    Fontana, M.H.; Kress, T.S.; Adams, R.E.; Parsly, L.F.; Parker, G.W.

    1976-04-01

    Progress is summarized in the areas of capacitor discharge vaporizer (CDV) development, small-vessel fuel and fuel-simulant aerosol studies, large-vessel aerosol studies, HCDA bubble simulant tests, fuel-simulant response to CDV electrical energy deposition, bubble shape and rise behavior, and structural integrity of equipment cell liners

  4. STATUS, CHALLENGES AND MARKETING OPPORTUNITIES FOR ...

    African Journals Online (AJOL)

    kitonyo

    the Ethiopian rift valley especially in Oromia and north of Lake Ziway regions. Since ... In 2001, the Eastern and Central Africa Bean Research Network ... KEPHIS and Horticultural Crops Development Authority (HCDA), and supermarket ... markets of Westlands, Karen and Ngong and the Nairobi city centre stock a variety of.

  5. Design trade-offs in view of safety considerations

    International Nuclear Information System (INIS)

    Saji, G.; Kishida, K.; Inoue, T.

    1978-01-01

    In view of resolving conflicting demands of cost, safety, flexibility of operation and design margins, safety design of various plant systems is discussed referring to their weight on construction costs. An influence of hypothetical core disruptive accident (HCDA) and loss of piping integrity (LOPI) on plant design and thus on construction materials is discussed, in optimising future commercial FBR plants. (author)

  6. An assessment of the effect of reactor size on hypothetical ore disruptive accidents

    International Nuclear Information System (INIS)

    Buttery, N.E.; Board, S.J.

    1978-01-01

    There is a general tendency towards larger plant sizes, in the interests primarily of economies of scale. In this paper the effect of core size on hypothetical core disruptive accidents (HCDA) is considered, and it is shown that the energy yield increases rapidly with size, primarily due to a tendency towards coherence of voiding in reactors with a large positive void coefficient. Small cores compare favourably in this respect with alternative large designs with low void coefficient cores, because the reduced mass more than compensates for the reduced doppler constant, and they also have a potential advantage in later stages of HCDA (transition phase and after). If energetic HCDA cannot be shown to be unrealistic and if containment of these events is provided as part of the general safety philosophy, then the costs (which may increase disproportionately with yield) of engineering an adequately reliable system needs to be accounted for. Containment costs are only one of many factors which need to be taken into account in optimising the design and so the energy release from a HCDA must take its proper place in the optimisation according to the safety principles and safety case agreed for LMFBRS. (author)

  7. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D.

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well

  8. Heat transfer and fluid flow aspects of fuel--coolant interactions

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

  9. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

  10. Network model of free convection within internally heated porous media

    International Nuclear Information System (INIS)

    Conrad, P.W.

    1977-01-01

    A hypothetical core-disruptive accident (HCDA) in a liquid metal fast breeder reactor (LMFBR) may result in the formation of an internally heated debris bed. Considerable attention has been given to postulated mechanisms by which such beds may be cooled. It is the purpose of the work described to demonstrate a method for computing the heat transfer from such a bed to the overlying sodium pool due to single-phase, free convection

  11. Influence of roof motion in LMFBR containment loading studies

    International Nuclear Information System (INIS)

    Potter, R.; Lancefield, M.J.; Sidoli, J.E.A.; Broadhouse, B.J.; Green, R.S.

    1982-01-01

    Following an HCDA the reactor roof may be threatened by coolant impact. Recent trends in CDFR roof design suggest that roof movement during the impact process may reduce the roof loading as a result of the fluid-structure interaction. The paper describes analytic studies of the phenomena, extensions to the SEURBNUK containment code to the roof flexibility and fluid-structure coupling, and results of experiments which confirm the reduced impulse and provide validation of the mathematical modelling

  12. Proposed heat transfer model for the gas-liquid heat transfer effects observed in the Stanford Research Institute scaled tests

    International Nuclear Information System (INIS)

    Corradini, M.; Sonin, A.A.; Todreas, N.

    1976-12-01

    In 1971-72, the Stanford Research Institute conducted a series of scaled experiments which simulated a sodium-vapor expansion in a hypothetical core disruptive accident (HCDA) for the Fast Flux Test Facility. A non-condensible explosive source was used to model the pressure-volume expansion characteristics of sodium vapor as predicted by computer code calculations. Rigid piston-cylinder experiments ( 1 / 10 and 1 / 30 scale) were undertaken to determine these expansion characteristics. The results showed that the pressure-volume characteristics depend significantly on the presence of water in the cylinder reducing the work output by about 50 percent when a sufficient water depth was present. The study presented proposes that the mechanism of heat transfer between the water and high temperature gas was due to area enhancement by Taylor instabilities at the gas-liquid interface. A simple heat transfer model is proposed which describes this energy transport process and agrees well with the experimental data from both scaled experiments. The consequences of this analysis suggest that an estimate of the heat transfer to the cold slug during a full-scale HCDA due to sodium vapor expansion and the accompanying reduction in mechanical work energy warrants further investigation. The implication of this analysis is that for either sodium or fuel vapor expansion in an HCDA, there is an inherent heat transfer mechanism which significantly reduces the work output of the expanding bubble

  13. Calculation of the deformation limits for failure affected wide plate tensile specimens

    Energy Technology Data Exchange (ETDEWEB)

    Janssen, G T.M.; Meijers, P [TNO-IWECO, Delft (Netherlands); Lorenz, H [INTERATOM, Bergisch Gladbach (Germany)

    1977-07-01

    In 1972 the German reactor safety commission recommended, with respect to the safety concept of the SNR 300 concerning an hypothetical core disruptive accident (HCDA), to design the complete plant against an mechanical energy release of 370 MWs and the reactor vessel against 150 MWs. In 1976 it has been decided to design the reactor vessel system against the defined energy release of 370 MWs. This change greatly increased the extent of design activities with regard to HCDA questions. The integrity proof will be given by deformation analysis of the reactor vessel using continuous mechanical computer codes and by comparison of the maximum deformations to verified design limits. The deformation behavior of the vessel system during the HCDA has been analysed by computer codes which describe the pulse and pressure distribution for all elements by which the geometry has been modelled. Simultaneously the codes calculate the hydrodynamic processes and the time and position dependent stress and strain distributions. The accuracy of these complicated computer codes describing pressure and deformation behavior will be evaluated by an experimental series of explosion tests on vessels.

  14. Hypothetical core disruptive accident analysis of a 2000 MWsub(e) liquid metal cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Struwe, D.

    1977-12-01

    A structural phase diagram for hypothetical core disruptive accidents (HCDA) has been developed based on a variety of analyses for different LMFBR's. The intention was to identify the strategic phases of HCDA's important with regard to safety aspects of the plant. These phases are investigated in detail for a 2,000 MWsub(e) LMFBR (SNR-2,000). Characteristic data of SNR-2,000 are discussed concerning their influence on safety analysis. Reasons for the choice of model parameters for special phenomena as fuel coolant interaction, fuel pin failure mechanisms and sodium voiding are given. The results of calculations with CAPRI-2, HOPE and KADIS are analyzed for possibilities to enter energetic core disassembly with consequences, making power values below 2,000 MWsub(e) necessary. Investigation of these results shows that the expected consequences do not lead to design requirements, restricting the magnitude of the electrical power output of LMFBR's to values below 2,000 MWsub(e). Therefore, consequences of HCDA's are principal not expected to limit the feasibility of conventional core design of this order of magnitude. (orig.) [de

  15. Shock loading of reactor vessel following hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Srinivas, G.; Doshi, J.B.

    1990-01-01

    Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

  16. Calculation of the deformation limits for failure affected wide plate tensile specimens

    International Nuclear Information System (INIS)

    Janssen, G.T.M.; Meijers, P.; Lorenz, H.

    1977-01-01

    In 1972 the German reactor safety commission recommended, with respect to the safety concept of the SNR 300 concerning an hypothetical core disruptive accident (HCDA), to design the complete plant against an mechanical energy release of 370 MWs and the reactor vessel against 150 MWs. In 1976 it has been decided to design the reactor vessel system against the defined energy release of 370 MWs. This change greatly increased the extent of design activities with regard to HCDA questions. The integrity proof will be given by deformation analysis of the reactor vessel using continuous mechanical computer codes and by comparison of the maximum deformations to verified design limits. The deformation behavior of the vessel system during the HCDA has been analysed by computer codes which describe the pulse and pressure distribution for all elements by which the geometry has been modelled. Simultaneously the codes calculate the hydrodynamic processes and the time and position dependent stress and strain distributions. The accuracy of these complicated computer codes describing pressure and deformation behavior will be evaluated by an experimental series of explosion tests on vessels

  17. Effects of recent modeling developments in prompt burst hypothetical core disruptive accident calculations

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Abramson, P.B.

    1978-01-01

    The main objective of the development of multifield, multicomponent thermohydrodynamic computer codes is the detailed study of hypothetical core disruptive accidents (HCDAs) in liquid-metal fast breeder reactors. The main contributions such codes are expected to make are the inclusion of detailed modeling of the relative motion of liquid and vapor (slip), the inclusion of modeling of nonequilibrium/nonsaturation thermodynamics, and the use of more detailed neutronics methods. Scoping studies of the importance of including these phenomena performed with the parametric two-field, two-component coupled neutronic/thermodynamic/hydrodynamic code FX2-TWOPOOL indicate for the prompt burst portion of an HCDA that: (1) Vapor-liquid slip plays a relatively insignificant role in establishing energetics, implying that analyses that do not model vapor-liquid slip may be adequate. Furthermore, if conditions of saturation are assumed to be maintained, calculations that do not permit vapor-liquid slip appear to be conservative. (2) The modeling of conduction-limited fuel vaporization and condensation causes the energetics to be highly sensitive to variations in the droplet size (i.e., in the parametric values) for the sizes of interest in HCDA analysis. Care must therefore be exercised in the inclusion of this phenomenon in energetics calculations. (3) Insignificant differences are observed between the use of space-time kinetics (quasi-static diffusion theory) and point kinetics, indicating again that point kinetics is normally adequate for analysis of the prompt burst portion of an HCDA. (4) No significant differences were found to result from assuming that delayed neutron precursors remain stationary where they are created rather than assuming that they move together with fuel. (5) There is no need for implicit coupling between the neutronics and the hydrodynamics/thermodynamics routines, even outside the prompt burst portion

  18. Estimation of the mechanical effects of a core disruptive accident on a LMFBR

    International Nuclear Information System (INIS)

    Robbe, M.F.; Lepareux, M.; Treille, E.

    2001-01-01

    In case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Reactor, the interaction between fuel and liquid sodium creates a high pressure gas bubble in the core. The violent expansion of this bubble loads the vessel and the internal structures, whose deformation is important. In order to demonstrate the CASTEM-PLEXUS capability to predict the behaviour of real reactors], axisymmetric computations of the MARA series were confronted with the experimental results. The computations performed at the beginning of the years 90 showed a rather good agreement between the experimental and computed results for the MARA 8 and MARA 10 tests even if there were some discrepancies which might be eliminated by increasing the fineness of the mesh. On the contrary, the prediction of the MARS structure displacements and strains was overestimated. This conservatism was supposed to come from the fact that several MARS non axisymmetric structures like core elements, pumps and heat exchangers were not represented in the CASTEM-PLEXUS model. These structures, acting as porous barriers, had a protective effect on the containment by absorbing energy and slowing down the fluid impacting the containment. For these reasons, we developed in CASTEM-PLEXUS a new HCDA constitutive law taking into account the presence of the internal structures (without meshing them) by means of an equivalent porosity method and we simulated the MARS test another time with the new HCDA constitutive law. This paper presents the numerical results relative to the structure behaviour during the accident. The results are described through the evolution of several variables versus time: deformed shape of the structures and the mesh, displacements, stresses and plastic strains. (author)

  19. 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.

  20. 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

  1. 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

  2. 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.

  3. Parameter studies to determine sensitivity of slug impact loads to properties of core surrounding structures

    International Nuclear Information System (INIS)

    Gvildys, J.

    1985-01-01

    A sensitivity study of the HCDA slug impact response of fast reactor primary containment to properties of core surrounding structures was performed. Parameters such as the strength of the radial shield material, mass, void, and compressibility properties of the gas plenum material, mass of core material, and mass and compressibility properties of the coolant were used as variables to determine the magnitude of the slug impact loads. The response of the reactor primary containment and the partition of energy were also given. A study was also performed using water as coolant to study the difference in slug impact loads

  4. Study of the containment system of the planned SNR-2 Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Bunz, H.; Scholle, U.

    1979-01-01

    Research is presented concerning the behaviour of four possible containment concepts for the planned Fast Breeder Reactor SNR-2 under the conditions of a HCDA. Detailed descriptions and calculations are given dealing with the acceptable leakages and the filter loads reached. The transport of the aerosol-type as well as the gaseous activity through the different compartments into the environment is calculated and from this the resulting accident doses. To cover the uncertainties of the accident course a wide range of accident parameters is examined

  5. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper describes the application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third problem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed. (Auth.)

  6. 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

  7. 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

  8. 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

  9. Attenuation of radiological consequences from CDA's by radiation. Progress report, October 1, 1976--September 31, 1977

    International Nuclear Information System (INIS)

    Chan, S.H.

    1977-01-01

    This technical progress report summarizes the research work accomplished during the first six months of the investigation on the significance of radiation heat transfer in attenuating the radiological consequences from LMFBR core disruptive accidents. Considerable progress has been made in modeling and computing the effects of radiative cooling on a rising HCDA bubble buoyant through a sodium pool. Our results reveal that most of the fuel vapor within the bubble can be effectively condensed out by radiating cooling. The finding has a profound implication as it could lead to a substantial reduction in subsequent aerosal releases

  10. Status of ANL out-of-pile investigations of severe accident phenomena for liquid metal reactors

    International Nuclear Information System (INIS)

    Spencer, B.W.; Marchaterre, J.F.; Anderson, R.P.

    1986-01-01

    Research addressing LMFBR whole core accidents has been terminated, and there is now emphasis on quantifying reactivity feedbacks, and in particular enhancing negative feedback, so that advanced LMR designs will provide inherently safe operation. The status of recent HCDA-related laboratory research performed at ANL, up to the time that such activities were no longer needed to support CRBR licensing, is described. Included are descriptions of programs addressing sodium channel voiding, fuel sweepout, fuel dispersal and plugging, boiled-up pool, UO 2 /sodium FCI, and debris coolability. Descriptions of recent investigations involving the metal fuel/sodium system are also included

  11. 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

  12. Failure analysis of carbide fuels under transient overpower (TOP) conditions

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1980-06-01

    The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

  13. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    The application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs is described. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third proem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. The computer code SEURBNUK/EURDYN (Release 1). Input and output specification

    International Nuclear Information System (INIS)

    Broadhouse, B.J.; Yerkess, A.

    1986-05-01

    SEURBNUK/EURODYN is an extension of SEURBNUK-2, a two dimensional, axisymmetric, Eulerian, finite element containment code in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin and thick structures. These codes are designed to model the hydrodynamic development in time of a hypothetical core disruptive accident (HCDA) in a fast breeder reactor. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations, with information on output facilities, and aid to users to avoid some common difficulties. (UK)

  20. Improvement and verification of fast-reactor safety-analysis techniques. Final report

    International Nuclear Information System (INIS)

    Barker, D.H.

    1981-12-01

    The work involved on this project took place between March 1, 1975 and December 31, 1981. The work resulted in two PhD and one Masters Theses. Part I was the Verification and Applicability Studies for the VENUS-II LMFBR Disassembly Code. These tests showed that the VENUS-II code closely predicted the energy release in all three tests chosen for analysis. Part II involved the chemical simulation of pool dispersion in the transition phase of an HCDA. Part III involved the reaction of an internally heated fluid and the vessel walls

  1. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Suk, S. D.

    2002-05-01

    In the present study, the KALIMER safety analysis has been made for the transients considered in the design concept, hypothetical core disruptive accident (HCDA), and containment performance with the establishment of the design basis. Such analyses have not been possible without the computer code improvement, and the experience attained during this research period must have greatly contributed to the achievement of the self reliance in the domestic technology establishment on the safety analysis areas of the conceptual design. The safety analysis codes have been improved to extend their applicable ranges for detailed conceptual design, and a basic computer code system has been established for HCDA analysis. A code-to-code comparison analysis has been performed as a part of code verification attempt, and the leading edge technology of JNC also has been brought for the technology upgrade. In addition, the research and development on the area of the database establishment has been made for the efficient and systematic project implementation of the conceptual design, through performances on the development of a project scheduling management, integration of the individually developed technology, establishment of the product database, and so on, taking into account coupling of the activities conducted in each specific area

  2. Homogenization of the internal structures of a reactor with the cooling fluid

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Bliard, F. [Socotec Industrie, Service AME, 78 - Montigny le Bretonneux (France)

    2001-07-01

    To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)

  3. Behavior of UO2 and FISSIUM in sodium vapor atmosphere at temperatures up to 28000C

    International Nuclear Information System (INIS)

    Feuerstein, H.; Oschinski, J.

    1986-11-01

    In case of a HCDA a rubble bed of fuel debris may form under a sodium pool and reach high temperatures. An experimental technique was developed to study the behavior of fuel and fission products in out-of-pile tests in a sodium vapor atmosphere. Evaporation rates of UO 2 were measured up to 2800 0 C. The evaporation was found to be a complex process, depending on temperature and the 'active' surface. Evaporation restructures the surface of the samples, however no new 'active' surface is formed. UO 2 forms sometimes well shaped crystals and curious erosion products. The efficiency of the used condenser/filter lines was higher than 99.99%. In case of a HCDA all the evaporated substances will condense in the soidum pool. Thermal reduction of the UO 2 reduces the oxygen potential of the system. The final composition at 2500 0 C was found to be UO 1.95 . The only influence of the sodium vapor was found for the diffusion of UO 2 into the thoria of the crucible. Compared with experiments in an atmosphere of pure argon, the diffusion rate was reduced. (orig.) [de

  4. An analytical study of slug impact phenomena

    International Nuclear Information System (INIS)

    Smith, B.L.

    1983-05-01

    Following a HCDA the roof of a fast reactor may be threatened by coolant impact. This article aims to develop at a fundamental level, understanding of the impact process and assess the relevance and magnitude of fluid-structure interaction effects. Reference is made to four 1/30th scale experiments, set up to verify the ideas developed in this work, and to provide quality data for code validation purposes. The impact of a one-dimensional liquid slug on a solid slab is investigated using a simplified form of the Rankine-Hugoniot shock equations derived under the joint assumptions of slight compressibility and small Mach number. Initially the roof slab is considered to be freely supported and of finite thickness. A detailed picture of the shock and expansion wave propagations is built up from the basic equations including the effects of wave reflections at boundaries and wave-wave interactions. Particular attention is paid to the impulse transfer mechanism from the slug as this controls the roof slab acceleration. Bulk fluid cavitation effects are noted. Roof flexural response is then taken into account, together with the effects of the hold-down constraints. It is seen that even very minor structural responses can result in significant mitigation of the impulse loading. Guidelines for the application of the work to HCDA analysis in pool reactor geometries are presented. (Auth.)

  5. Homogenization of the internal structures of a reactor with the cooling fluid

    International Nuclear Information System (INIS)

    Robbe, M.F.; Bliard, F.

    2001-01-01

    To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)

  6. Energy conservation in SIMMER

    International Nuclear Information System (INIS)

    Arnold, L.A.; Knowles, J.B.

    1983-11-01

    The SIMMER code contains models of the many interacting thermo-hydraulic processes that occur during a hypothetical core disruptive accident (HCDA), to provide an overall picture from accident initiation to containment loading. In calculations of roof loadings following the HCDA, errors in computing the overall energy balance were found to be up to ten times the kinetic energy of the sodium slug which creates the loading. On this account, the results were considered to be seriously compromised. This report describes a systematic investigation into the effect, nature and origin of the energy discrepancies. Its main conclusion are that, the errors stem from a systematic rather than a random source, energy errors for individual cells can be two decades larger than the mean value provided by the code, and cellular mass and energy errors are strongly correlated and they can actually increase when the mesh is refined. A likely cause of the conservation errors is identified as the solution of the liquid phase mass and energy equations at effectively different time instants during each timestep. (author)

  7. Simulation of a hypothetical core disruptive accident in the mars test-facility

    International Nuclear Information System (INIS)

    Robbe, M.F.; Lepareux, M.

    2001-01-01

    In France, a large experimental programme MARA/MARS was undertaken in the 80's to estimate the mechanical consequences of an HCDA (Hypothetical Core Disruptive Accident) and to validate the SIRIUS computer code used at that time for the numerical simulations. At the end of the 80's, it was preferred to add a HCDA sodium-bubble-argon tri-component constitutive law to the general ALE fast dynamics finite element CASTEM-PLEXUS code rather than going on developing and using the specialized SIRIUS code. The experimental results of the MARA programme were used in the 90's to validate and qualify the CASTEM-PLEXUS code. A first series of computations of the tests MARA 8, MARA 10 and MARS was realised. The simulations showed a rather good agreement between the experimental and computed results for the MARA 8 and MARA 10 tests - even if there were some discrepancies - but the prediction of the MARS structure displacements and strains was overestimated. This conservatism was supposed to come from the fact that several MARS non axisymmetric structures like core elements, pumps and heat exchangers were not represented in the CASTEM-PLEXUS model. These structures, acting as porous barriers, had a protective effect on the mock-up containment by absorbing energy and slowing down the fluid impacting the containment. For these reasons, we developed in CASTEM-PLEXUS a new HCDA constitutive law taking into account the presence of the internal structures (without meshing them) by means of an equivalent porosity method. In other respects, the process used for dealing with the fluid-structure coupling in CASTEM-PLEXUS was improved. Thus a second series of simulations of the tests MARA8 and MARA10 was realised. A simulation of the test MARS was carried out too with the same simplified representation of the peripheral structures as in order to estimate the improvement provided by the new fluid-structure coupling. This paper presents a third numerical simulation of the MARS test with the

  8. Program improvement and applications; Programmpflege und Anwendungen

    Energy Technology Data Exchange (ETDEWEB)

    Hink, M.; Imke, U.; Pfrang, W.; Porscha, B.; Struwe, D.; Zimmerer, W.; Allan, P.

    1995-08-01

    An account is given about further improvements of the SAS4A-Ref. 94.R0 version of the HCDA code. They concern in particular the DEFORM fuel rod deformation module. For a validation of the new code version, various CABRI experiments have been calculated, especially tests with high burnup fuel rods. Progress was shown to be achieved, but the precise timing and location of the observed fuel failures is still hard to calculate. The work was performed in close cooperation with partners in France, Britain, and Japan. An important application concerns the CAPRA project of a reactor for actinide burning. Its behavior under ULOF conditions was analyzed using the improved SAS4A Ref. 94 R0 code. The core design turned out to tend toward a long-term coolable configuration even more so than the EFR core design would do in an ULOF. (orig.)

  9. Program improvement and applications

    International Nuclear Information System (INIS)

    Hink, M.; Imke, U.; Pfrang, W.; Porscha, B.; Struwe, D.; Zimmerer, W.; Allan, P.

    1995-01-01

    An account is given about further improvements of the SAS4A-Ref. 94.R0 version of the HCDA code. They concern in particular the DEFORM fuel rod deformation module. For a validation of the new code version, various CABRI experiments have been calculated, especially tests with high burnup fuel rods. Progress was shown to be achieved, but the precise timing and location of the observed fuel failures is still hard to calculate. The work was performed in close cooperation with partners in France, Britain, and Japan. An important application concerns the CAPRA project of a reactor for actinide burning. Its behavior under ULOF conditions was analyzed using the improved SAS4A Ref. 94 R0 code. The core design turned out to tend toward a long-term coolable configuration even more so than the EFR core design would do in an ULOF. (orig.)

  10. Fast reactor safety: proceedings of the international topical meeting. Volume 2

    International Nuclear Information System (INIS)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments

  11. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    International Nuclear Information System (INIS)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences

  12. Temperature field in concrete when in contact with hot liquids

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de.

    1981-09-01

    In an HCDA (Hypothetical Core Disruptive Accident) it is postulated that liquid metal coolants and core materials come in contact with the retaining concrete structure. A mathematical model and an associated computer program was previously developed to describle the transient heat and mass transfer in the concrete. Implementations on the original program-USINT- are included to consider the variations of the thermal conductivity as a function of the temperature. Also a subroutine - PLOTTI - is incorporated to the program for the plotting of the results. The new program - USINTG - is used to calculate the temperature and pressure fields and the water released from concrete structures during a sodium leak simulation and with the concrete structures in contact with liquid sodium. No consideration about chemical reactions involving the sodium when in contact with concrete is considered. (Author) [pt

  13. Enhancing the Productivity of High Value Crops and Income Generation with Small-Scale Irrigation Technologies in Kenya. Final Report 2009-2013

    International Nuclear Information System (INIS)

    2014-02-01

    The project was implemented by the Kenya Agricultural Research Institute in collaboration with key irrigation stakeholders including Horticultural Crops Development Authority (HCDA), G North and Son limited, Kenya Irrigation and Drainage Association (KIDA), Jomo Kenyatta University of Agriculture and Technology (JKUAT), Greenbelt Movement and Ministry of Agriculture. The objective was to develop and pilot test appropriate irrigation systems (methods and related water/nutrient management practices) for small-scale farmers for increasing yield, quality of high value crops and farmers income to improved livelihood. The project built on earlier work on low head drip irrigation in Kenya involving KARI led promotion among the peri-urban and rural communities. The Equipment used include Neutron Probe Hydroprobe, Ammonium Sulphate Fertilizers (5% a.e), drip irrigation kits, MoneyMaker irrigation pumps, Pessl imetos weather station, SDEC tensimetre and tensiometers), Venturi injectors, among others.

  14. KALIMER preliminary conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

    2000-08-01

    This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report.

  15. Improvement and verification of fast reactor safety analysis techniques. Progress report, July 1, 1978--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Barker, D.H.; Wheeler, P.; Bybee, D.J.

    1978-01-01

    This report deals with the simulation of a HCDA in a LMFBR using chemical heating. The chemicals used were acetyl chloride and dimethyl sulfoxide in a benzene base. In our last report, it was shown that for high concentrations of the two chemicals, vapor velocities above 8 cm/sec and void fractions above 50% were obtained in a closed system. However, at lower concentrations, such as 4 M, no void fraction was obtained. During this period of work, a closed system with a cooling coil was used to provide a larger heat sink. It was shown that for the 4 M concentration void fractions above 50% was obtained and for the 9 M concentration the void fraction was higher and stayed dispersed much longer than without the cooling coil.

  16. Overview and status of the SIMMER testing program

    International Nuclear Information System (INIS)

    Scott, J.H.

    1979-01-01

    Los Alamos Scientific Laboratory has undertaken an extensive experiment analysis program to test the results of SIMMER Liquid Metal Fast Breeder Reactor (LMFBR) accident calculations. Initially, we will test the postdisassembly work-energy partition problem. The SIMMER-calculated order-of-magnitude reduction of available kinetic energy following a severe hypothetical core-disruptive accident (HCDA) can be attributed to (1) purely fluid-dynamic effects; and (2) rate-controlled effects, such as phase transitions and heat transfer. We have chosen to test separately each class of mitigator. In this paper we review the experiments initially chosen for testing of each class of mitigator and report on the status of the analyses. We enumerate several problems in SIMMER that experiment analysis has disclosed. Finally, needs for future experiments are discussed

  17. Simulation experiments on the radial pool growth in gas-releasing melting system

    International Nuclear Information System (INIS)

    Farhadieh, R.; Purviance, R.; Carlson, N.

    1983-01-01

    Following an HCDA, molten core-debris can contact the concrete foundation of the reactor building resulting in a molten UO 2 /concrete interaction and considerable gas release. The released gas can pressurize the containment building potentially leading to radiological releases. Furthermore, directional growth of the molten core-debris pool can reduce the reactor building structural integrity. To implement design changes that insure structural integrity, an understanding of the thermal-hydraulic and mass-transfer process associated with such a growth is most desirable. Owing to the complex nature of the combined heat, mass, and hydrodynamic processes associated with the two-dimensional problem of gas release and melting, the downward and radial penetration problems have been investigated separately. The present experimental study addresses the question of sideward penetration of the molten core debris into a gas-releasing, meltable, miscible solid

  18. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

  19. Summary of treat experiments on oxide core-disruptive accidents

    International Nuclear Information System (INIS)

    Dickerman, C.E.; Rothman, A.B.; Klickman, A.E.; Spencer, B.W.; DeVolpi, A.

    1979-02-01

    A program of transient in-reactor experiments is being conducted by Argonne National Laboratory in the Transient Reactor Test (TREAT) facility to guide and support analyses of hypothetical core-disruptive accidents (HCDA) in liquid-metal fast breeder reactors (LMFBR). Test results provide data needed to establish the response of LMFBR cores to hypothetical accidents producing fuel failure, coolant boiling, and the movement of coolant, molten fuel, and molten cladding. These data include margins to fuel failure, the modes of failure and movements, and evidence for identification of the mechanisms which determine the failure and movements. A key element in the program is the fast-neutron hodoscope, which detects fuel movement as a function of time during experiments

  20. 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.

  1. Some recriticality studies with SIMMER-II

    International Nuclear Information System (INIS)

    Bohl, W.R.

    1979-01-01

    The SIMMER-II code was applied to the problem of evaluating the potential for recriticality in an LMFBR HCDA. The phenomenology examined was based on the post SAS3D behavior in a CRBR LOF accident. The SIMMER-II results were found to be sensitive to the development of fuel blockages. If blockages were formed close to the core, the core-disruption phase of the behavior resulted from the significant energy required following SAS3D termination to raise the average steel temperature to saturation conditions; also material deposition near heat sinks eliminated condensation surfaces and induced core pressurization and fuel collapse. Excessive ramp rates (greater than $100/s) were judged to be unlikely, but improved knowledge and calculational treatments of the applicable phenomenology would be desirable

  2. Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

  3. Considerations on a PAHR test facility

    International Nuclear Information System (INIS)

    Boenisch, G.; Groetzbach, G.; Heinzel, V.; Kleefeld, K.; Kuechle, M.; Mueller, R.A.; Royl, P.; Schramm, K.; Smidt, D.; Werle, H.

    1976-01-01

    On the basis of a hypothetical core disruptive accident (HCDA) analysis the phenomena of the post accident phase are first identified which require experimental investigations and can only be studied in pile. Then the experimental requests for both debris bed and molten fuel pool studies are specified and grouped into three categories. For two of the categories the requests can be satisfied with loop experiments in thermal reactors. For the third category a 70 cm diameter test bed is needed and here the proposal is to use a flat core fast reactor with the test bed located below the core heated by axial leakage neutrons. Finally a conceptual design for such a reactor is presented where the test bed is loaded into an ex-vessel device and is removable on a carriage to a hot cell building. Maintenance and safety problems are briefly discussed and alternative solutions are mentioned

  4. Thermal-structural response of EBR-II major components under reactor operational transients

    International Nuclear Information System (INIS)

    Chang, L.K.; Lee, M.J.

    1983-01-01

    Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

  5. Development and application of capacitor discharge vaporization technique for fuel aerosol studies

    International Nuclear Information System (INIS)

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

    1976-01-01

    Investigations into the behavior of LMFBR fuel as it responds to the levels of energy deposition projected for prompt-critical excursions have been limited because of the difficulty of experimentally achieving the resultant high thermodynamic states in a laboratory. In order to conduct ''source-term'' assessment and fuel aerosol behavioral experiments, the Aerosol Release and Transport program at ORNL has been developing a Capacitor Discharge Vaporization (CDV) technique for using electrical energy, via the discharge from capacitors, to achieve HCDA-like energy state in UO 2 . The paper reports the details of the technique, the developmental test results that have demonstrated feasibility, and a brief description of the proposed experiments that will use the CDV system for safety related studies

  6. KALIMER preliminary conceptual design report

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

    2000-08-01

    This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report

  7. 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

  8. Structural dynamics in fast reactor accident analysis

    International Nuclear Information System (INIS)

    Fistedis, S.H.

    1975-01-01

    Analyses and codes are under development combining the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage the stresses, strains, and deformations of important primary components, as well as the overall adequacy of primary and secondary containments. An arbitrary partition of the structural components treated evolves into (1) a core mechanics effort; and (2) a primary system and containment program. The primary system and containment program treats the structural response of components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which provide greater accuracy and longer durations for the treatment of HCDA. The codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. Recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of primary piping. Pulses are provided at the vessel-primary piping interfaces of the inlet and outlet nozzles, calculation includes the elbows and pressure drops along the components of the primary piping system. Recent improvements to the primary containment codes include introduction of bending strength in materials, Langrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. Another development involves the combination of a 2-D finite element code for the reactor cover with the hydrodynamic containment code

  9. An assessment of core wide coherency effects in the multichannel modeling of the initiating phase of a severe accident in a sodium fast reactor

    International Nuclear Information System (INIS)

    Guyot, M.; Gubernatis, P.; Suteau, C.; Le Tellier, R.; Lecerf, J.

    2014-01-01

    To consolidate the safety assessment for liquid-metal fast breeder reactors (LMFBRs), hypothetical core disruptive accident (HCDA) sequences have been extensively studied over the past decades. Numerous analyses of the so called initiating phase (or primary phase) of a HCDA have been made with the safety analysis system code SAS4A. The SAS4A accident analysis code requires that subassemblies or groups of subassemblies be represented together as independent channels. For simulating a severe accident sequence, a subassembly-to-channel assignment procedure has to be implemented to produce the consistent SAS4A input decks. Generally, one uses imposed criteria over relevant reactor parameters to determine the subassembly to- channel arrangement. The multiple-assembly-per-channel approach introduces core wide coherency effects, which can affect the reactivity balance and therefore the overall accident development. In this paper, a subassembly-to channel assignment procedure based on the subassembly power-to-flow ratio is presented and implemented to generate the SAS4A input decks over a range of parameter values. The corresponding SAS4A calculations have been performed on a large LMFBR. The purpose of the present series of calculations is to investigate the magnitude of errors encountered in the analysis of the initiating phase related to the subassembly-to-channel arrangement selection, by comparison with a one-subassembly-per-channel reference solution. It appears that a refinement in the channel arrangement substantially reduces core wide coherency effects. Analysis of the calculations also suggests that an accurate representation of the scenario requires the number of channels to be on approximately the same order of magnitude as the total number of subassemblies. Numerical results are examined to provide the reader with quantitative measurements of bias related to subassembly to- channel arrangement. (authors)

  10. Review and Perspective of FCI Related Activities in Japan

    International Nuclear Information System (INIS)

    Kawaguchi, O.

    1976-01-01

    Because of its meager natural resources and high population, Japan needs a strong nuclear program. However, the high population density and the public concern about environmental pollution and safety tend to impose stringent safety criteria for nuclear plants. Safety of the LMFBR, the most promising among several nuclear options, must be established for future commercialization. One of the most crucial issues in LMFBR safety is the molten fuel coolant interaction (FCI), which comes into play in various phases of the hypothetical core disruptive accident (HCDA). The objectives of the present paper are: firstly to summarize the current FCI considerations applied in the design of the Japanese prototype FBR, MONJU; and secondly to review the supporting R and D activities. In conclusion: Although analytical model predictions for FCI work energy per plant electrical output has decreased considerably over the last few years, the current classical approach would yield an inexorable containment design for larger FBRs. The FCI issue should be resolved within the coming few years. Concentrated efforts should be directed toward: (1) Proof both with simulant materials and fuel-sodium systems of the existence of the threshold for vapor explosions. (2) Establishment with out-of-pile fuel-sodium systems of the fragmentation mechanism to evaluate the heat transfer coefficients. (3) A few carefully planned in-pile tests. (4) Development of computer codes at various levels of accident to realistically assess the initial condition, progression and consequence of an FCI event under reactor conditions. (5) Establishment of the world-wide consensus of the HCDA scenario for licensable, economical commercial FBRs. We believe that an extensive data and personnel exchange are readily available in the areas (1) and (2) above, and that the NEA/CSNI Committee should coordinate future experiments on a complementing basis. We believe that promotion of further international cooperation in the

  11. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

    2015-10-15

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  12. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    International Nuclear Information System (INIS)

    Heo, Hyo; Bang, In Cheol; Jerng, Dong Wook

    2015-01-01

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  13. Proceedings of the third specialist meeting on sodium/fuel interaction in fast reactors

    International Nuclear Information System (INIS)

    1976-01-01

    This specialist meeting, sponsored by the OECD-NEA and organized by the Power Reactor and Nuclear Fuel Development Corporation, was attended by 56 delegates from 6 countries and the CEC (Commission of the European Communities). The purpose of the meeting was to bring together and discuss in depth the Fuel-Sodium Interaction, a phenomenon of major importance in the assessment of the Hypothetical Core Disruptive Accident in the Liquid Metal Fast Breeder Reactor. The meeting was essentially a follow-up of an earlier meeting held at Ispra in December 1973. In all, 29 papers were presented, covering the following topics: 1. Current perspective on sodium-fuel interaction in LMFBR safety; 2. Basic experimental and theoretical studies including other materials; 3. In-pile and out-of-pile experimental studies on sodium-fuel interaction; 4. Theoretical models for the interpretation of experiments and for application to reactor situations. The meeting is considered useful in narrowing down the chain of events necessary to get energetic interaction, large work potential, but many points are being clarified on the gap between the basic vapor explosions and the real fuel sodium interactions in the HCDA scenario of LMFBR. Finally another meeting of the same nature as this one has been recommended

  14. Melt Fragmentation Characteristics of Metal Fuel with Melt Injection Mass during Initiating Phase of SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Lee, Min Ho; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of)

    2016-05-15

    The PGSFR has adopted the metal fuel for its inherent safety under severe accident conditions. However, this fuel type is not demonstrated clearly yet under the such severe accident conditions. Additional experiments for examining these issues should be performed to support its licensing activities. Under initiating phase of hypothetic core disruptive accident (HCDA) conditions, the molten metal could be better dispersed and fragmented into the coolant channel than in the case of using oxide fuel. This safety strategy provides negative reactivity driven by a good dispersion of melt. If the coolant channel does not sufficient coolability, the severe recriticality would occur within the core region. Thus, it is important to examine the extent of melt fragmentation. The fragmentation behaviors of melt are closely related to a formation of debris shape. Once the debris shape is formed through the fragmentation process, its coolability is determined by the porosity or thermal conductivity of the melt. There were very limited studies for transient irradiation experiments of the metal fuel. These studies were performed by Transient Reactor Test Facility (TREAT) M series tests in U.S. The TREAT M series tests provided basic information of metal fuel performance under transient conditions. The effect of melt injection mass was evaluated in terms of the fragmentation behaviors of melt. These behaviors seemed to be similar between single-pin and multi-pins failure condition. However, the more melt was agglomerated in case of multi-pins failure.

  15. Evaluation of the confinement option for LMRs

    International Nuclear Information System (INIS)

    Himes, D.A.; Stepnewski, D.D.; Franz, G.R.

    1985-12-01

    The coolant in liquid metal cooled reactors operates at low pressures and therefore contains relatively little stored energy compared to LWR systems. This presents the possibility of using a more conventional building for containment coupled with a confinement system which vents the internal volume of the building through a filter/scrubber. The confinement system would be designed to keep the internal pressure in the containment near atmospheric thereby minimizing unfiltered leakage. The principal benefits of such an arrangement would be lower capital cost and less stringent leaktightness requirements permitting simpler and less disruptive testing. In conclusion, the confinement system assumed here would reduce consequences to the public of an LMR HCDA to acceptable levels. However control room doses are unacceptable due to the noble gas concentration inside the control room. A confinement system is therefore a viable design option for LMR's provided means are included for keeping noble gases out of the control room. Such means are readily available including, for example, selectable remote air intakes, an exhaust stack, or a noble gas filter. Probably the most satisfactory alternative would be a large cryogenic filter on the confinement system exhaust

  16. A review of fast reactor activities in Switzerland - March 1984

    International Nuclear Information System (INIS)

    Wydler, P.

    1984-01-01

    As a result of the noncentralized government in Switzerland there is no clear national policy for the future application of nuclear energy. This is reflected in the lack of a generally agreed nuclear energy research policy in the country. Consequently, activities related to several advanced reactor concepts are funded simultaneously at similar, but relatively low levels. The total expenditure of 5.9 million Swiss Francs (approx. 1 SFr per capita) for fast reactor activities in 1983 must be judged in the light of this situation. The funds have been allocated to an LMFBR safety programme (52%) and a fuel development programme (48%). In the field of LMFBR safety analytical work is performed on hypothetical core disruptive accidents (HCDAs) and on the integrity of components under HCDA loadings with emphasis on the dynamic behaviour of the reactor cover. A considerable effort has recently been devoted to the preparations for the SONACO natural convection experiment. Another relatively new experimental activity, involving small-scale vapour explosions with freon and water, has produced evidence of interesting physical effects which are not in accord with the assumptions of current molten fuel-coolant interaction (MFCI) models. The fuel development programme has continued with the manufacture of spherepac mixed carbide fuel pins for an irradiation experiment in FFTF. However, the time scale of the experiment has suffered a set-back due to an accident in a glove box of the production line

  17. Formation and stability of crust in molten pools. Technical progress report, February 1, 1977--October 1, 1977

    International Nuclear Information System (INIS)

    Bankoff, S.G.; Ganguli, A.

    1977-01-01

    The objective of the research is to study the formation and stability of a freezing-melting process under well-defined hydrodynamic and thermal boundary conditions simulating LMFBR HCDA phenomena. A hot liquid, such as molten UO 2 , comes into contact with a cold solid, such as stainless steel forming a solid crust at the interface, possibly accompanied by the simultaneous formation of a melt layer. The stability and rate of growth of the crust and underlying melt layer are thus of concern. A steady flowpast an initially flat slab of the cold solid is considered. The experimental apparatus consists of a horizontal rectangular test section in which a frozen slab of n-decane 18-in. long and 4-in. wide, supported on a refrigerated copper block, is exposed to a flowing stream of cold water. Temperature profiles in the decane will be measured, as well as the rate of liquid decane collection, and the ice crust and melt layer profiles, if stable. These will be compared with theoretical predictions, following an integral boundary layer method. Current efforts are focussed on improving the mold release properties and other debugging operations

  18. The development of technologies of safety analysis for LMR ('03)

    International Nuclear Information System (INIS)

    Lee, Y. B.; Suk, S. D.; Chang, W. P.; Kwon, Y. M.; Jeong, H. Y.; Ha, K. W.; Heo, S.

    2004-03-01

    The developmental objectives of the project, 'The development of safety analysis techniques in LMR', are the code development for the subchannel blockage analysis, the code development for the system transient analysis, the code development for the HCDA(Hypothetical Core Disruptive Accident) analysis, the preliminary safety analysis for KALIMER-600 equipped with the components of new concepts, and the establishment of data base. The purpose of the analysis for subchannel blockage in the subassembly of LMR is to represent quantitatively that the maximum damage due to the accident is within the safety criteria. The computational program should be developed to simulate the thermal hydraulic phenomena and to verify the safety of LMR for the accident. For the purpose, the hybrid scheme has been implemented into the MATRA-LMR code based on the upwind scheme to analyze the various flow fields occurred in the subchannel blockage accident. The turbulent mixing models using the CFX code were assessed to compute more precisely the heat transfer between subchannels. Through this assessment, empirical correction factors of 1.7 for the heat conduction, 0.006 for the turbulent mixing coefficient were obtained. The distributed resistance model instead of wire forcing function has been developed to represent the more exact flow field due to wire-wrap. Other models, such as heat conductor model and various turbulent mixing model, have been implemented into the MATRA-LMR. The ORNL THORS 19-Pin FFM-5B tests have been assessed to validate above new models using the improved MATRA-LMR. The results using MATRA-LMR were well agreed with the experimental data. The subchannel blockage accidents which assumed to be occurred at the three locations for the conceptual plant of KALIMER-600 have been analysed according to blockage size using the MATRA-LMR code. The results of calculations for the design basis events which 6 subchannels were blocked showed the margins of the 290 7.dog. C up to the

  19. The development of technologies of safety analysis for LMR ('03)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. B.; Suk, S. D.; Chang, W. P.; Kwon, Y. M.; Jeong, H. Y.; Ha, K. W.; Heo, S

    2004-03-01

    The developmental objectives of the project, 'The development of safety analysis techniques in LMR', are the code development for the subchannel blockage analysis, the code development for the system transient analysis, the code development for the HCDA(Hypothetical Core Disruptive Accident) analysis, the preliminary safety analysis for KALIMER-600 equipped with the components of new concepts, and the establishment of data base. The purpose of the analysis for subchannel blockage in the subassembly of LMR is to represent quantitatively that the maximum damage due to the accident is within the safety criteria. The computational program should be developed to simulate the thermal hydraulic phenomena and to verify the safety of LMR for the accident. For the purpose, the hybrid scheme has been implemented into the MATRA-LMR code based on the upwind scheme to analyze the various flow fields occurred in the subchannel blockage accident. The turbulent mixing models using the CFX code were assessed to compute more precisely the heat transfer between subchannels. Through this assessment, empirical correction factors of 1.7 for the heat conduction, 0.006 for the turbulent mixing coefficient were obtained. The distributed resistance model instead of wire forcing function has been developed to represent the more exact flow field due to wire-wrap. Other models, such as heat conductor model and various turbulent mixing model, have been implemented into the MATRA-LMR. The ORNL THORS 19-Pin FFM-5B tests have been assessed to validate above new models using the improved MATRA-LMR. The results using MATRA-LMR were well agreed with the experimental data. The subchannel blockage accidents which assumed to be occurred at the three locations for the conceptual plant of KALIMER-600 have been analysed according to blockage size using the MATRA-LMR code. The results of calculations for the design basis events which 6 subchannels were blocked showed the margins of the 290 7.dog. C

  20. Safety design of SNR-300

    International Nuclear Information System (INIS)

    Traube, K.

    1976-01-01

    The joint German-Belgian-Dutch loop-type 300 MW(e) LMFBR prototype is being constructed at Kalkar on the lower Rhine in Germany. Among the many arguments put forward in defense of SNR-300, that of acquiring licensing exprience has proven to be of major importance to the international breeder scene. The severity of the licensing procedures and of the safety standards imposed are unique in several respects, including timing: generally growing scepticism towards nuclear power increased severity of the licensing practice; organizational features: the procedure and criteria developed for commercial light water reactors have been applied without exemptions. This relates to the commercial-type contract under which SNR-300 is being built for private utilities by a private company; and German nuclear safety standards, known worldwide to be most stringent. The following three important areas are discussed in which SNR-300 decidedly deviates from its forerunners: protection against the hypothetical core disruptive accident (HCDA), protection against external events, and provisions for in-service inspection

  1. Experimental investigation of flow dynamics in the SNR-upper-core structure

    International Nuclear Information System (INIS)

    Meyer, L.

    1985-03-01

    This report describes the results of a simulant-material experimental investigation of flow dynamics in the upper-core (UCS) during a HCDA of a LMFBR. The experiments were designed to verify some of the thermal-hydraulic models in SIMMER-II. Four different liquids were used to simulate the flashing U0 2 ; and numerous parameter variations were made regarding initial pressure, temperature, and configurations of the test apparatus. The experiments showed the large effect of the heat transfer in the UCS and the relatively small effect of friction. The reduction in final kinetic energy by the presence of the UCS is shown as a function of the initial pressure and the temperature difference between core and UCS. Calculations with SIMMER-II for the wide range of experiments produced results for the kinetic energy within a factor of 2 of the experimental results without changing the crucial input parameters. The minimum droplet size during the flashing process and the structure-side heat transfer coefficient were determined to be the crucial and most sensitive parameters. This reflects deficiencies in modeling of both the flashing process and the transient heat conduction in the structure. (orig./HP) [de

  2. Effect of particle stratification on debris-bed dryout

    International Nuclear Information System (INIS)

    Gabor, J.D.; Cassulo, J.C.; Pederson, D.R.

    1982-01-01

    Significant work has been performed on debris-bed dryout on beds of either uniformly sized particles or particles of a wide size range which are well mixed. This work has provided an understanding of the mechanisms of dryout and an empirical basis for containment analysis. However, the debris bed resulting from a HCDA would not consist of uniformly sized particles and for certain scenarios the bed could be stratified rather than well mixed. Tests have been conducted on the effect of particle size distribution on dryout and concluded that not only is the mean particle size an important parameter but also the standard deviation of the distribution and change in porosity. The D6 in-pile test at Sandia with a 114-mm deep stratified bed resulted in a reduced dryout heat flux compared to a uniformly mixed bed. Because of the many questions concerning the dryout behavior of stratified beds of wide size distribution out-of-pile experiments in which metal particles in water pools are inductively heated were initiated at Argonne

  3. Kalaeloa Energy System Redevelopment Options Including Advanced Microgrids.

    Energy Technology Data Exchange (ETDEWEB)

    Hightower, Marion Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Baca, Michael J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); VanderMey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    In June 2016, the Department of Energy's (DOE's) Office of Energy Efficiency and Renewable Energy (EERE) in collaboration with the Renewable Energy Branch for the Hawaii State Energy Office (HSEO), the Hawaii Community Development Authority (HCDA), the United States Navy (Navy), and Sandia National Laboratories (Sandia) established a project to 1) assess the current functionality of the energy infrastructure at the Kalaeloa Community Development District, and 2) evaluate options to use both existing and new distributed and renewable energy generation and storage resources within advanced microgrid frameworks to cost-effectively enhance energy security and reliability for critical stakeholder needs during both short-term and extended electric power outages. This report discusses the results of a stakeholder workshop and associated site visits conducted by Sandia in October 2016 to identify major Kalaeloa stakeholder and tenant energy issues, concerns, and priorities. The report also documents information on the performance and cost benefits of a range of possible energy system improvement options including traditional electric grid upgrade approaches, advanced microgrid upgrades, and combined grid/microgrid improvements. The costs and benefits of the different improvement options are presented, comparing options to see how well they address the energy system reliability, sustainability, and resiliency priorities identified by the Kalaeloa stakeholders.

  4. 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

  5. The State-of-the-Art Report on the Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kim, Yeong Il; Kim, Seong O; Lee, Jae Han; Lee, Yong Bum

    2006-03-01

    The status of the sodium cooled metal fuel core technology and the design methodology were described. The preliminary design concepts of the metal fuel core were established for KALIMER. A systematic study on the development fluid and I and C system has been carried out, and the conceptual design of NSSS of the 150MWe and 600MWe LMRs such as the design of PHTS, IHTS, RHRS, SGS and related technology with BOP is established together with the computational technology. The development of detection system of the sodium-water reaction, core blockage and the conceptual design of the control system of large capacity LMR are being carried. The important technological areas for mechanical structure design of LMR are high temperature thin structure design, seismic isolation design, In-Service- Inspection technology, and the economic design. The highlighted performances for the safety analysis were the developments of the containment analysis code CONTAIN-LMR-K, the safety analysis code SSC-K and the flow blockage analysis code. The safety criteria were set up, the safety analysis for the equilibrium core, the HCDA analysis, and the containment performance analysis were performed. The recent SSC-K 1.3 version turns out to be reliable after the indirect verification throughout qualitative/quantitative assessments

  6. Experimental investigation of the unsteady two-phase flow through perforated plates

    International Nuclear Information System (INIS)

    Tartaglia, G.P.

    1985-07-01

    The coolant flow across the perforated dip-plate during a hypothetical core disruptive accident (HCDA) in a liquid metal fast breeder reactor was simulated in a one-dimensional model. Experiments with a water-air mixture as fluid were run by varying the following parameters: geometry of the dip-plate (perforation ratio, number of the holes), height of the fluid head over the dip-plate, air volumetric fraction, size of the air bubbles, acceleration of the fluid. The pressure drop across the dip-plate, the forces acting on the dip-plate and on the upper plate, acceleration and displacement of the piston, the air volumetric fraction and the size of the air bubbles were measured in a wide range of Strouhal and acceleration numbers. The flow pattern downstream the dip-plate was filmed with a high-speed camera. The following correlations were investigated: resistance coefficients as a function of the acceleration and Strouhal number, time delay of the force on the upper plate as a function of the cavitation number, and forces and impulses acting on the upper plate compared with those acting on the dip-plate. Finally, using high-speed film pictures, the formation of fluid jets downstream the dip-plate was investigated. The following relations were obtained: displacement of the mixture surface and of the jets as a function of the perforation ratio and of the air volumetric fraction, and cavitation volume as a function of the cavitation number. (orig.) [de

  7. Development of a submerged gravel scrubber for containment venting applications: summary

    International Nuclear Information System (INIS)

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

    1981-01-01

    Although hypothetical core disruptive accidents (HCDAs) are not design basis accidents for breeder reactor plants, extensive assessments of HCDA consequences have been made and design features for providing margins beyond the design base have been considered for future fast reactor plants. One feature proposed for increasing the safety margin is a containment vent and/or purge system which would mitigate the challenge to containment integrity resulting from excessive temperature and pressure or excessive hydrogen. A cleanup system would be required for removal of vented aerosols and condensible vapors to mitigate radiological consequences to the environment. A study is in progress at HEDL to select and develop a suitable air cleaning system for use in potential breeder reactor containment venting applications. A concept was conceived whereby the passiveness and high loading capacity of a water pool scrubber was combined with the high efficiency of a sand and gravel bed. It was termed a Submerged Gravel Scrubber (SGS). A schematic drawing of the concept is shown. The SGS consists of a bed of gravel (or other packing) submerged in a pool of water

  8. 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.)

  9. Vapor pressures of oxide reactor fuels above 3000 K: Review and perspective

    International Nuclear Information System (INIS)

    Breitung, W.

    1982-03-01

    Vapor pressures of liquid oxide reactor fuels are among the most important material data required for theoretical analyses of Hypothetical Core Disruptive Accidents in Fast Breeder Reactors. This report is an attempt to completely summarize and critically review the numerous theoretical and experimental results published for the pressure-temperature and pressure-energy relation of unirradiated UO 2 and (U,Pu)O 2 . First - to define the research goal - the precision in the saturation vapor pressure is quantified which is required for the purpose of HCDA calculations. Then the various theoretical and experimental methods used for the determination of p-T and p-U data are reviewed with respect to their principles, results and uncertainties. The achievements of the individual methods are discussed in the light of the research goal and - in view of the widely scattered data - recommendations are made concerning the p-T and p-U relation of UO 2 . Finally, the most important future research areas are identified, including some specific research proposals which aim at reducing the still large uncertainties in fuel vapor pressures down to the desired level. (orig.) [de

  10. Interaction and penetration of heated UO2 with limestone concrete

    International Nuclear Information System (INIS)

    Farhadieh, R.; Pedersen, D.R.; Purviance, R.; Carlson, N.

    1982-01-01

    To safeguard the environment against radiological releases, the major question of concern in PAHR safety assessment, following an HCDA, involves confinement and dilution of the molten core-debris. Significant to the study is the directional growth of the core-debris in the concrete foundation of the reactor building or the concrete below the reactor cavity. The real material experiments were carried out in the test apparatus shown. Casts of CRBRP limestone concrete were prepared in graphite cylinders, each having an internal diameter of 8.9 cm and a depth of 30.5 cm. The 17.8-cm-deep concrete samples were allowed to cure for at least 28 days. Experiments were conducted within two months of curing time. The cavity above concrete was packed with 3 kg of pure UO 2 particles (1 to 3 mm). A uranothermic mixture was placed on the top of UO 2 powder. Heating and possible melting of UO 2 was achieved resistively after the ignition of the thermite. Total experimental time was about 60 minutes, during which time a maximum electrical power input of 1.8 watts/gr was applied to the UO 2 . Three gas samples were taken at temperatures of 100, 600, and 950 0 C, measured in the plane of the No. 2 thermocouple. Selection of three temperatures were to study the amount and the type of gases released from different phases of concrete

  11. 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)

  12. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Lee, Y. B.; Kwon, Y. M.; Suk, S. D.

    2005-03-01

    The MATRA-LMR-FB has been developed internally for the damage prevention as well as the safety assessment during a channel blockage accident and, as a the result, the quality of the code becomes comparable to that developed in the leading countries. For a code-to-code comparison, KAERI could have access to the SASSYS-1 through a bilateral collaboration between KAERI and ANL. The study could bring into the reliability improvements both on the reactivity models in the SSC-K and on the SSC-K prediction capability. It finally leads to the completion of the SSC-K version 1.3 resulting from the qualitative and quantitative code-to-code comparison. The preliminary analysis for a metal fueled LMR could also become possible with the MELT-III and the VENUS-II, which had originally been developed for the HCDA analysis with an oxidized fuel, by developing the relevant models For the development of the safety evaluation technology, the safety limits have been set up, and the analyses of the internal and external channel blockages in an assembly have also been performed. Besides, the more reliable analysis results on the key design concepts could be obtained by way of the methodology improvement resulting from the qualitative and quantitative comparison study. For an efficient and systematic control of the main project, the integration of the developed technologies and the establishment of their data base have been pursued. It has gone through the development of the process control with taking account of interfaces among the sub-projects, the overall coordination of the developed technologies, the data base for the design products, and so on

  13. 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)

  14. Preliminary safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2001-06-01

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  15. Development of Sodium Two Phase Flow Model for Kalimer Core Analysis

    International Nuclear Information System (INIS)

    Chang, W.P.; Hahn, Dohee

    2002-01-01

    An algorithm for sodium boiling is developed in order to extend the applicability of SSC-K, which is a main system analysis code for the KALIMER (Korea Advanced LIquid Metal Reactor) conceptual design. As the capability of the current SSC-K version is limited to simulation of only a single-phase sodium flow, its applicable range should not be enough to assess the fuel integrity under some of HCDA (Hypothetical Core Disruptive Accident) initiating events where sodium boiling is anticipated. The two-phase flow model similar to that used for the light water system is known to be no more effective directly to liquid metal reactors, because the phenomena observed between two reactor coolant systems are definitely different. The developing algorithm is based on a multiple-bubble slug ejection model, which allows a finite number of bubbles in a channel at any time. The present work is a continuous effort following the former study to confirm a qualitative acceptance on the model. Since the model has been applied only to the active fuel region in the former study, a part of its qualification seems to have already been demonstrated. For its application to the whole KALIMER core channel, however, the model needs to be examined the applicability to the fuel regions other than the active fuel. The present study primarily focuses on that point. In a result, although the model may be improved in a sense through the present study over the previous modeling, a clear limitation is also confirmed with the validity of the model. The further development, therefore, is required for this model to achieve its goal by resolving such limitations. (authors)

  16. Development of two-phase Flow Model, 'SOBOIL', for Sodium

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Chang, Won Pyo; Kim, In Chul; Kwon, Young Min; Lee, Yong Bum

    2000-03-01

    The objective of this research is to develop a sodium two-phase flow analysis model, 'SOBOIL', for the assessment of the initial stage of the KALIMER HCDA (Hypotherical Core Disruptive Accident). The 'SOBOIL' is basically similar to the multi-bubble slug ejection model used in SAS2A[1]. When a bubble is formed within the liquid slug, the bubble fills the whole cross section of the coolant channel except for a film left on the cladding or on the structure. Up to nine bubbles, separated by the liquid slugs, are allowed in the channel at any time. Each liquid slug flow rate in the model is performed in 2 steps. In the first step, the preliminary flow rate in the liquid slug is calculated neglecting the effect of changes in the vapor bubble pressures over the time step. The temperature and pressure distributions, and interface velocity at the interface between the liquid slug and vapor bubble are also calculated during this process. The new vapor temperature and pressure are then determined from the balance between the net energy transferred into the vapor and the change of the vapor energy. The liquid flow is finally calculated considering the change of the vapor pressure over a time step and the calculation is repeated until specified elapsed time is met. Continuous effort, therefore, must be made on the examination and improvement for the model to become reliable. To this end, much interest must be concentrated in the relevant international collaborations for access to a reference model or test data for the verification.

  17. Development of two-phase Flow Model, 'SOBOIL', for Sodium

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Chang, Won Pyo; Kim, In Chul; Kwon, Young Min; Lee, Yong Bum

    2000-03-01

    The objective of this research is to develop a sodium two-phase flow analysis model, 'SOBOIL', for the assessment of the initial stage of the KALIMER HCDA (Hypotherical Core Disruptive Accident). The 'SOBOIL' is basically similar to the multi-bubble slug ejection model used in SAS2A[1]. When a bubble is formed within the liquid slug, the bubble fills the whole cross section of the coolant channel except for a film left on the cladding or on the structure. Up to nine bubbles, separated by the liquid slugs, are allowed in the channel at any time. Each liquid slug flow rate in the model is performed in 2 steps. In the first step, the preliminary flow rate in the liquid slug is calculated neglecting the effect of changes in the vapor bubble pressures over the time step. The temperature and pressure distributions, and interface velocity at the interface between the liquid slug and vapor bubble are also calculated during this process. The new vapor temperature and pressure are then determined from the balance between the net energy transferred into the vapor and the change of the vapor energy. The liquid flow is finally calculated considering the change of the vapor pressure over a time step and the calculation is repeated until specified elapsed time is met. Continuous effort, therefore, must be made on the examination and improvement for the model to become reliable. To this end, much interest must be concentrated in the relevant international collaborations for access to a reference model or test data for the verification

  18. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

    2000-07-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

  19. Treatment of sodium spills and leakage detection at loop-type fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Foerster, K; Fortmann, M; Lang, H; Moellerfeld, H [Interatom, Bergisch Gladbach (Germany)

    1979-03-01

    Sodium spills are of great importance in the safety analysis for sodium cooled nuclear plants. Large leakages can lead to a depletion of the heat transfer system and cause the loss of cooling of the reactor. Further the hot sodium may attack structural materials. In areas with air atmosphere large amounts of sodium can burn and cause great damages. Therefore the control of large leakages is an indispensable task in design and construction of sodium cooled reactor systems. Because of the typical arrangement of widespread long pipe systems loop type plants are subject to a gradually greater risk of damage than pool type plants. The sodium catching devices of the SNR-300 are described and their function is illustrated as an example for the treatment of large spills. Since the equipment for the control of large amounts of leaking sodium is very expensive, great efforts are made in order to save costs and to decrease safety problems. It is aimed to minimize the probability of such events to a degree that they no longer are to be considered realistic. The advantageous operating conditions and the favourable material properties support this aim. Under the well known keyword 'leak-before-rupture' criterion this task is pursued. Crack growth measurements are made at structural materials under LMFBR conditions, and leakage detecting systems are being developed. Some test results concerning this task are described. Despite the fact that there are good chances to verify the leak-before-rupture criterion it is assumed that certain hypothetical accidents occur, which are to be considered in the design of the reactor plant. The extremely improbable Bethe-Tait-accident (HCDA) is such an event. It would lead to a super spill, that means to the complete depletion of the reactor tank. For the SNR-300 plant a system is provided that is able to catch this super spill and the core melt. This core catcher must withstand the high temperatures and remove the decay heat. The purpose of this

  20. Treatment of sodium spills and leakage detection at loop-type fast reactors

    International Nuclear Information System (INIS)

    Foerster, K.; Fortmann, M.; Lang, H.; Moellerfeld, H.

    1979-01-01

    Sodium spills are of great importance in the safety analysis for sodium cooled nuclear plants. Large leakages can lead to a depletion of the heat transfer system and cause the loss of cooling of the reactor. Further the hot sodium may attack structural materials. In areas with air atmosphere large amounts of sodium can burn and cause great damages. Therefore the control of large leakages is an indispensable task in design and construction of sodium cooled reactor systems. Because of the typical arrangement of widespread long pipe systems loop type plants are subject to a gradually greater risk of damage than pool type plants. The sodium catching devices of the SNR-300 are described and their function is illustrated as an example for the treatment of large spills. Since the equipment for the control of large amounts of leaking sodium is very expensive, great efforts are made in order to save costs and to decrease safety problems. It is aimed to minimize the probability of such events to a degree that they no longer are to be considered realistic. The advantageous operating conditions and the favourable material properties support this aim. Under the well known keyword 'leak-before-rupture' criterion this task is pursued. Crack growth measurements are made at structural materials under LMFBR conditions, and leakage detecting systems are being developed. Some test results concerning this task are described. Despite the fact that there are good chances to verify the leak-before-rupture criterion it is assumed that certain hypothetical accidents occur, which are to be considered in the design of the reactor plant. The extremely improbable Bethe-Tait-accident (HCDA) is such an event. It would lead to a super spill, that means to the complete depletion of the reactor tank. For the SNR-300 plant a system is provided that is able to catch this super spill and the core melt. This core catcher must withstand the high temperatures and remove the decay heat. The purpose of this

  1. The role of fission product in whole core accidents - research in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Dietrich, L W [Argonne National Laboratory, Division of Reactor Analysis and Safety, Argonne, IL (United States); Jackson, J F [Los Alamos Scientific Laboratory, Q Division - Energy, Los Alamos, NM (United States)

    1977-07-01

    Safety of nuclear reactors has been a central concern of the nuclear energy industry from the very beginning. This concern, and the resultant excellence of design, fabrication, and operation, aided by extensive engineered safety features, has given nuclear energy its superior record of protection of the environment and of the public health and safety. With respect to the fast reactor, it was recognized early in the programme that there exists a theoretical possibility of a core compaction leading to significant energy release. Early analysis of this problem employed a number of conservative assumptions in attempting to bound the energy release. As reactors have grown in size, the suitability of such bounding calculations has diminished, and research into hypothetical accident analysis has emphasized a more mechanistic approach. In the USA, much effort has been directed towards modeling and computer code development aimed at following the course of an accident from its initiation to its ultimate conclusion with a stable, permanently subcritical, coolable core geometry, along with considerations of post-accident heat removal and radiological consequence assessment. Throughout this effort, the potential role of fission products has been recognized and account taken of the effects of fission products in determining accident progression. It is important to recognize that reactor safety is a very diverse topic, requiring consideration of a number of factors. While the major questions of public risk appear to be related to the hypothetical core disruptive accident (HCDA), it is necessary that the probability of having such an accident be extremely low In order that acceptable public risk be demonstrated. Such a demonstration requires sound engineering design and Implementation, with high standards of reliability, inspectability, maintainability, and operation, along with the requisite quality control and assurance. Tile current approach, typified by that taken by the

  2. Progress reports for Gen IV sodium fast reactor activities FY 2007

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Tentner, A. M.

    2007-01-01

    for prevention of progression into severe accident conditions (prevention of core melting) or for mitigation of severe accident consequences (mitigation of the impact of core melting to protect public health and safety). Because design measures for severe accident prevention and mitigation are beyond the normal design basis, established regulatory guidelines and codes do not provide explicit identification of the design performance requirements for severe accident accommodation. The treatment of severe accidents is one of the key issues of R and D plans for the Gen IV systems in general, and for the Sodium Fast Reactor (SFR) in particular. Despite the lack of an unambiguous definition of safety approach applicable for severe accidents, there is an emerging consensus on the need for their consideration for the design. The US SFR program and Argonne National Laboratory (ANL) in particular have actively studied the potential scenarios and consequences of Hypothetical Core Disruptive Accidents (HCDA) for SFRs with oxide fuel during the Fast Flux Test Facility (FFTF) and Clinch River Breeder Reactor Plant (CRBRP) programs in the 70s and 80s. Later, the focus of the US SFR safety R and D activities shifted to the prevention of all HCDAs through passive safety features of the SFRs with metal fuel in the Integral Fast Reactor (IFR) program, and the study of severe accident consequences was de-emphasized. The goal of this paper is to provide an overview of the current SFR safety approach and the role of severe accidents in Japan and France, in preparation for an expected and more active collaboration in this area between the US, Japan, and France

  3. An innovative design approach to a cost effective commercial liquid metal reactor

    International Nuclear Information System (INIS)

    Carelli, M. D.; Garkisch, H. D.; Hundal, R.; Arie, K.; Handa, N.; Ota, H.; Matsuyama, S.; Todreas, N.; Hejzlar, P.; Wells, P.; Louis, T. S.

    2008-01-01

    One of the requirements for the Advanced Recycle Reactor (ARR) in the DOE GNEP program is to be economically competitive with advanced LWRs. Sodium cooled Liquid Metal Reactors have been designed in two basic configurations, pool and loop, and various designs had different features, but the capital cost has consistently been substantially higher than for LWRs in spite of the LMRs higher efficiency due to the higher coolant exit temperature. We have identified two major areas as responsible for the divergence in capital costs. First, LMRs have been designed within the HCDA (Hypothetical Core Disruptive Accident) philosophy. Basically, the plant has to survive a major accident regardless whether this accident has any mechanistically acceptable, albeit very small, probability of occurring. Thus, in spite of their inherent reactivity feedbacks, much more effective than in LWRs, LMRs have been designed with redundant systems and costly design solutions which a probabilistic, rather than a deterministic approach, would eliminate or greatly simplify. The other major consideration is the presence of the intermediate heat transport system to prevent a Na/H 2 O reaction involving primary (radioactive) sodium. No other reactor type has three heat transport systems and in fact the LMR intermediate heat transport system is a substantial contributor to the capital cost. For our ARR rated at 1000 MWt power, we have selected a pool configuration because of its enhanced safety and its inherent capability of yielding a more compact reactor. The intermediate heat transport system is eliminated. Immersed in the pool is a double wall steam generator where water/steam flows through the inner tubes and Helium is in the space between the inner and outer tubes for detection purposes. Models of two different types of double wall steam generators designed independently by Westinghouse and Toshiba had been tested in the 1970-1980's at ETEC with favorable results. As we fully realize the risks