Computation of a core disruptive accident in the MARS mock-up

A hypothetical core disruptive accident in a liquid metal fast breeder reactor (LMFBR) results from the interaction between molten fuel and liquid sodium, which creates a high-pressure bubble of gas in the core. The violent expansion of this bubble loads and deforms the vessel and the internal structures. The MARS experimental test simulates a HCDA in a small-scale mock-up containing all the significant internal components of a fast breeder reactor. The mock-up is filled with water, topped by an argon blanket, and the explosion is generated by an explosive charge.This paper presents a numerical simulation of the test with the EUROPLEXUS code. The top closure is represented by massive structures and the main internal structures are described by shells. The current numerical results are described and compared with the experimental ones, and previous computations with the CASTEM-PLEXUS code.     

Comparison of REXCO code predictions with flexible vessel experiments

This paper deals with the REXCO code predictions of the flexible vessel tests. Comparisons are made with the data from experiments which were designed specifically for the purpose of code verification and/or modification. These experiments were performed with a well-defined and calibrated energy source in cylindrical vessel with precision tolerances and simple well-defined boundary conditions. The experimental data can thus be used as reliable test data for validation of computer codes, as well as of the modeling techniques used in the computer analysis. The inputs to the computer analysis are the vessel and core barrel dimensions and boundary conditions, the stress-strain relationships for the vessel and core barrel materials, the equation of state for the coolant, and the pressure-volume relationship of the energy source. The REXCO-predicted wall deformations, pressure loadings, and integrated impulses at various gauge positions are compared with the experimental data. Results of the comparisons are discussed.     

Fluid-structure interaction analysis of a hypothetical core disruptive accident in LMFBRs

To ensure safety, it is necessary to assess the integrity of a reactor vessel of liquid-metal fast breeder reactor (LMFBR) under HCDA. Several important problems for a fluid-structural interaction analysis of HCDA are discussed in the present paper. Various loading models of hypothetical core disruptive accident (HCDA) are compared and the polytropic processes of idea gas (PPIG) law is recommended. In order to define a limited total energy release, a “5% truncation criterion” is suggested. The relationship of initial pressure of gas bubble and the total energy release is given. To track the moving interfaces and to avoid the severe mesh distortion an arbitrary Lagrangrian–Eulerian (ALE) approach is adopted in the finite element modeling (FEM) analysis. Liquid separation and splash from a free surface are discussed. By using an elasticity solution under locally uniform pressure, two simplified analytical solutions for 3D and axi-symmetric case of the liquid impact pressure on roof slab are derived. An axi-symmetric finite elements code FRHCDA for fluid-structure interaction analysis of hypothetical core disruptive accident in LMFBR is developed. The CONT benchmark problem is calculated. The numerical results agree well with those from published papers.     

Energy and angular flux density spectra of 6 MeV gamma-rays penetrating shielding materials

Double differential flux spectra are presented for 6 MeV gamma photons penetrating aluminium, steel, lead and lead-aluminium shielding slabs. A large diameter disc radiator filled with activated primary cooling water from the Universities Research Reactor provides the source of high energy photons. Experimental pulse height distributions from a NaI spectrometer have been unfolded with the code RADAK. Comparisons are made with theoretical values using the code PIPE. Good agreement is established for most of the energy range, but experimental results show the need for the inclusion of secondary sources of photons in theoretical calculations.     

Comparison of experimental and analytic simulations of reactor structural response to a hypothetical core disruptive accident

A series of scale model tests assessing the ability of the Clinch River Breeder Reactor to withstand the loads resulting from a hypothetical core disruptive accident have been performed. Supporting analytic simulations of these tests using the REXCO-HEP code were also performed. Comparisons of the analytic and experimental results are described in this paper.As a general conclusion, the analytically predicted loads on the coolant boundary and resulting structural deformations were greater than the corresponding experimental loads and deformations.     

Validation of a computational model for the evaluation of fuel-coolant interaction under severe accidental condition in fast breeder reactors

A computational model for fuel-coolant interaction has been validated through calculations for a series of THINA experiments. By the experiments, it was intended to simulate a comparatively massive injection of molten core materials into sodium pool under a core disruptive accidental condition assumed for fast breeder reactors. The calculations by the SIMMER-II code showed that the current models for the fuel-coolant interaction (FCI) had the capability to well reproduce the experimental results. This means that the SIMMER-II FCI model based on the liquid particles size and the heat transfer among liquid components is valid with proper accuracy for the estimation of a FCI even without more detailed mechanistic model for FCI. In particular, the thermal-to-mechanical energy conversion rates were well predicted.     

Energetics of simulated HCDA bubble expansions: Some potential attenuation mechanisms

Experiments conducted to increase our understanding of the dynamics and thermodynamics of expanding bubbles similar to the core disruptive accident (CDA) bubble in liquid metal fast breeder reactors (LMFBR) are described. The experiments were conducted in a transparent model of a typical demonstration-size loop-type LMFBR in which water at room temperature simulated the sodium coolant. Nitrogen gas (1450 psia) and flashing water (1160 psia) qualitatively simulated sodium vapor and molten fuel expansions. Three physical mechanisms that may result in attenuation of the work potential of a hypothetical CDA (HCDA) were revealed by the experiments: (1) the pressure gradient existing between the lower core and the bubble within the pool, (2) the hydrodynamic effects of vessel internal structures, and (3) the nonequilibrium flashing process occurring in the lower core. These three mechanisms combine to result in a coolant axial slug kinetic energy that is only 14% of the work potential of the ideal quasi-static nitrogen expansion and only 5% of the work potential of the ideal quasi-static flashing water expansion.     

Assessment of CDFR primary containment capability under HCDA loading

The paper reviews UK studies of fast reactor containment response under hypothetical core disruptive accident (HCDA) loading, describing the evolution of complementary programmes of model experiments, numerical methods development and code validation. Results are presented from studies of the CDFR primary vessel, roof and core support structure, with particular emphasis on recent experimental work: these examples illustrate the level of detail required in the assessment of containment structures. The status of the work is critically reviewed, drawing attention to problems associated with the extrapolation of data from model experiments to the reactor situation. The likely direction of future work is indicated, focussing on more detailed assessment of particular structural features, the performance of seals and the study of leakage.     

Experimental study on dynamic response of reactor vessel to an HCDA

This paper presents a discussion on the model experiments results for reactor structure dynamic response on FBR hypothetical core disruptive accident (HCDA) and the results of analysis using the dynamic response analysis code under experimental conditions. The purpose of this study is to clarify experimentally the dynamic response by use of scale models, as well as to attempt to confirm and improve the dynamic response analysis code on the basis of experimental data. The experimental results have clarified the inner barrel effects on reactor vessel deformation and its behavior due to a impact load. On the other hand, dynamic analysis was made of the 1/7.5 scale complex model by a dynamic response analysis code “PISCES-2DL”, using the explosive combustion characteristics as inputs. Obtained values were compared with experimental values. Results showed that this method was fairly capable of evaluating radial deformation behavior in lower cylindrical parts of the vessel.     

Analysis of the primary containment response using a hydrodynamic-elastic-plastic computer code

The dynamic response of the primary reactor containment system to a hypothetical core disruptive accident (HCDA) is determined from the basic equations of mass, momentum, and energy, and the equations of state of the medium. These equations are first expressed in material coordinates and then set into finite difference form solved numerically on the computer using a hydrodynamic-elastic-plastic computer code, REXCO-HEP developed at ANL. Propagation of pressure waves, loads imposed on different parts of the reactor components, and the resulting deformations are determined at every time step throughout the sequence of the calculation. As a sample calculation, the code was applied to analyze the response of the FFTF reactor to a 150 MWsec HCDA. The mathematical model is described in detail, particularly in the areas of modeling reactor internals and extending the time range to cover the entire excursion phenomenon. Finally, the results obtained from the computer analysis are discussed in detail.     

Hydrodynamics of post-disassembly fuel expansion

The early expansion of the fuel following disassembly in an LMFBR core disruptive accident is modeled. Spherical expansion in the sodium is assumed. A Lagrangian, finite-difference hydrodynamic code (FEXPAN) describes the motion. Disassembly employs VENUS-II, and a consistent equation of state for fuel was used throughout disassembly and FEXPAN. Time-dependent mechanical work and fuel vaporized without fuel mixing are obtained. FEXPAN is compared with time-independent expansion for the effect of fuel mixing. For example, for a particular accident analysed 75 MW sec mechanical work was calculated for expansion with no mixing versus 10 MW sec with complete core mixing.     

Tests in support of LMFBR

Tests in support of LMFBR projects for potential incidents involving a hypothetical core disruptive accident or sodium-water interactions in steam generators have two possible goals: (1) to evaluate the integrity of the primary containment after the design is frozen or (2) to investigate design options in support of primary containment design. The test planning approach differs depending on the goal. The Fast Flux Test Facility tests had the first goal, the current Clinch River Breeder Reactor tests have the second goal. This paper discusses test planning, sources for simulating loads, modeling, and instrumentation. The main source described involves controlled venting of explosive gases. The source avoids generation of undesirable shock waves and facilitates calibration because the reaction rate of the explosive is independent of the response. It is indicated that the test program requiring the least time and cost involves a mixture of three types of models: small simple models, small complex models, and large complex models. Instrumentation is discussed that is required for validation of loading, response measurements for comparison with calculations, and response measurements on critical members where predictions are not possible.     

Structural response of fast breeder reactors to hypothetical core disruptive accidents

Liquid metal-cooled fast breeder reactors (LMFBRs) so far have been analyzed for the consequences on the plant and the environment for hypothetical core disruptive accidents (HCDAs). To provide the appropriate analytical tools for this effort, analysis and codes are currently under development in several countries. They combine the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage stresses, strains, and deformations of the important components of the system, and the overall adequacy of the primary and secondary containments. The effort is partitioned into the structural analysis of (a) the core components, and (b) the primary system components beyond the core.The core mechanics effort covers the structural response of fuel pins, hexcans, fuel elements, and fuel element clusters to transient pressures and thermal loads. Two- and three-dimensional finite element codes are under development for these core components. The results of these analyses would permit evaluation of the adequacy of the heat removal process to continue following severe core component deformations. Also, these analyses are currently being combined with neutronics, for the core transition phase, to allow for the mass movements for realistic neutronic calculations.The primary system and containment program treats the structural response of the 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 in their combined form provide greater accuracy and longer durations for the treatment of HCDAs. More recently the codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. This step will permit treatment of the instabilities following slug impact, the ultimate reversal of the sodium slug with the rising bubble, the bubble break-up, and the calculations of sodium splillage and radioactive gases, if any, in the secondary containment. The extent of sodium spillage and sodium fires should be known for evaluation of the secondary containment. The mechanics of bubble migration are needed for radiological study of post-accident phenomena. More recently dynamic fracture mechanics considerations are being incorporated to remove arbitrary failure criteria imposed on components such as the core barrel and vessel.Most recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of the primary piping. The pulses are provided at the vessel primary piping interfaces of the inlet and outlet nozzles. The calculation includes the elbows and pressure drops along the components of the primary piping system. Pressures larger than the ones used as input at the inlet and outlet nozzles were observed. As expected, they occur far from the nozzles, in the pipe, where the pulses meet.Recent improvements to the primary containment codes include introduction of bending strength in materials, Lagrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. A further development involves the combination of a 2-D finite element code for the reactor cover with the 2-D finite-difference hydrodynamic code for continuous monitoring of stresses, strains, and deformations in the cover, as well as pressure changes in the hydrodynamic code. Substantial experimental effort is in progress in various countries on the response to energy releases of vessels and internals, piping systems, subassemblies, and subassembly clusters. These experimental results are being utilized for the verification or modification of the analyses and codes under development.     

Numerical interpretation of the 8 experiment simulating a hypothetical core disruptive accident

In the case of a hypothetical core disruptive accident (HCDA) in a liquid metal fast breeder reactor (LMFBR), it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between the molten fuel and the liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel, thus endangering the safety of the nuclear plant. The experimental test 8 simulates the explosive phenomenon in a mock-up included in a flexible vessel with a flexible roof. This paper presents a numerical simulation of the test and a comparison of the computed results with the experimental results and previous numerical ones.     

Modelling of water sump evaporation in a CFD code for nuclear containment studies

During the course of a hypothetical severe accident in a pressurized water reactor (PWR), water can be collected in the sump containment through steam condensation on walls and spray systems activation. This water is generally under evaporation conditions. The objective of this paper is twofold: to present a sump model developed using external user-defined functions for the TONUS-CFD code and to perform a first detailed comparison of the model results with experimental data. The sump model proposed here is based on energy and mass balance and leads to a good agreement between the numerical and the experimental results. Such a model can be rather easily added to any CFD code for which boundary conditions, such as injection temperature and mass flow-rate, can be modified by external user-defined functions, depending on the atmosphere conditions.     

Fast Neutron Transport through Laminated Iron-Water Shield

Reaction rates were measured in a laminated iron-water shield by threshold detectors, from which the neutron spectra were obtained with the aid of the SAND-II code. The error analysis for the unfolding of the spectra proved that the spectra obtained satisfactorily in the energy range of 1–10.5 MeV. One-dimensional calculations were made by the discrete ordinates transport codes ANISN-JR and PALLAS in a spherical geometry. Agreements within a factor of 1.6 for the spectra and 1.31 for the reaction rates were obtained between the measurements and calculations, though rather large discrepancies were found in the spectra at the energy range of 3–7 MeV. All experimental data in absolute value and detailed specifications for source, detector and the experimental geometry are given for a fast neutron transport benchmark calculation.     

Numerical simulations of gas–liquid–particle three-phase flows using a hybrid method

For the analysis of debris behavior in core disruptive accidents of liquid metal fast reactors, a hybrid computational tool was developed using the discrete element method (DEM) for calculation of solid particle dynamics and a multi-fluid model of a reactor safety analysis code, SIMMER-III, to reasonably simulate transient behavior of three-phase flows of gas–liquid–particle mixtures. A coupling numerical algorithm was developed to combine the DEM and fluid-dynamic calculations, which are based on an explicit and a semi-implicit method, respectively. The developed method was validated based on experiments of water–particle dam break and fluidized bed in systems of gas–liquid–particle flows. Reasonable agreements between the simulation results and experimental data demonstrate the validity of the present method for complicated three-phase flows with large amounts of solid particles.     

Decontamination factors and release rates of UO2 particles from boiling pools of sodium

A semi-mechanistic model for calculating solid radionuclide release rates from boiling or bubbling pools of sodium was developed. The influence of particle spacial and size distributions on the decontamination of the releases was analysed and found significant. Decontamination factors are shown as a function of pool depth, bubbling characteristics and particle size distribution. The calculation of a decontamination factor for estimating the source term of large scale hypothetical core disruptive accidents is presented. The decontamination factor for a large scale accident was found to be two orders of magnitude greater than results obtained from small scale experiments conducted with uniform particle distributions.     

Energy Spectra of Heavy Fragments from the Interaction of Protons with Communications Materials

A Monte Carlo computer code is used to calculate the energy spectra of recoil nuclei resulting from interactions of protons with communications materials. Results are presented for 130 MeV and 1 GeV protons incident on O, Si, Ga, As, and Au. The results for 130 MeV protons on Si are compared with previous calculations and measurements.