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.     

Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor

In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant.The MARA 10 experimental test simulates a HCDA in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge.This paper presents a numerical simulation of the test with the EUROPLEXUS code and an analysis of the computed results. In particular, the evolution of the fluid flows and the deformations of the internal and external structures are analysed in detail. Finally, the current computed results are compared with the experimental ones and with previous numerical results computed with the SIRIUS and CASTEM-PLEXUS codes.     

Comparison of various models for the simulation of a Core Disruptive Accident in the MARA 10 mock-up

During a Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor melts down partially and that the interaction between hot molten fuel and relatively cold liquid sodium creates a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant.The MARA 10 experimental test simulates a Core Disruptive Accident in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge.This paper presents some models available within the EUROPLEXUS code to simulate a Core Disruptive Accident and an analysis of the computed results. In particular, results are compared with experimental measurements and previous numerical simulations carried out with the codes SIRIUS and CASTEM-PLEXUS.     

Liquid entrainment by an expanding core disruptive accident bubble—a Kelvin/Helmholtz phenomenon

The final stage of a postulated energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor is believed to involve the expansion of a high-pressure core-material bubble against the overlying pool of sodium. Some of the sodium will be entrained by the CDA bubble which may influence the mechanical energy available for damage to the reactor vessel. The following considerations of liquid surface instability indicate that the Kelvin–Helmholtz (K–H) mechanism is primarily responsible for liquid entrainment by the expanding CDA bubble. First, an instability analysis is presented which shows that the K–H mechanism is faster than the Taylor acceleration mechanism of entrainment at the high fluid velocities expected within the interior of the expanding CDA bubble. Secondly, a new model of liquid entrainment by the CDA bubble is introduced which is based on spherical-core-vortex motion and entrainment via the K–H instability along the bubble surface. The model is in agreement with new experimental results presented here on the reduction of nitrogen-gas-simulant CDA bubble work potential. Finally, a one-dimensional air-over-water parallel flow experiment was undertaken which demonstrates that the K–H instability results in sufficiently rapid and fine liquid atomization to account for observed CDA gas-bubble work reductions. An important byproduct of the theoretical and experimental work is that the liquid entrainment rate is well described by the Ricou–Spalding entrainment law.     

快堆钠回路水锤程序开发与应用

研究开发了快堆钠回路水锤分析专用程序WHA。该程序在一维特征线法（MOC）传统的压力波传播数学模型中补充了钠腔－气腔外边界模型，并采用气泡离散模型模拟低压液柱分离中的蒸汽穴的生成与溃灭。程序用FORTRAN90语言对快堆实验钠回路ESPRESSO中由于阀门的快速开启与关闭引起的压力波传播进行了分析计算。计算结果表明：将钠腔－气腔引入水锤压力波传播的数学模型进行程序计算的结果是合理的。     

Heat transfer from a totally blocked fuel subassembly of a liquid metal fast breeder reactor: II. Numerical simulation

Fast breeder nuclear reactors used for power generation, have fuel subassemblies in the form of rod bundles enclosed inside tall hexagonal cavities. Each subassembly can be considered as a porous medium with internal heat generation. A three-dimensional analysis is carried out here to estimate the heat transfer due to natural convection, in such an anisotropic, partially heat generating porous medium, which corresponds to the typical case of blocked flow in a fuel subassembly inside the reactor core. Using the finite volume technique, the temperatures at various locations inside hexagonal cavity are obtained. The simulations by the three-dimensional code developed are compared with the results of experiments [Suresh, Ch.S.Y., Sateesh, G., Das, Sarit K., Venkateshan, S.P., Rajan, M., 2004. Heat transfer from a totally blocked fuel subassembly of a liquid metalfast breeder reactor. Part 1: Experimental investigation. Nucl. Eng. Design, present issue] conducted using liquid sodium as the heat transfer fluid. Further, the code is used to predict the maximum temperature in typical liquid metal fast breeder reactors to find the power level where the liquid sodium starts boiling. It helps to decide the power level for initiation of monitoring the temperature for the purpose of reactor control.     

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive reactivity effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors.     

Prediction method for thermal stratification in a reactor vessel

The prediction method for thermal stratification phenomena in a fast breeder reactor is described. The focus of attention is placed on the applicability of water test results to predict thermal stratification phenomena in a real plant. The basic feature of thermal stratification was examined in a cylindrical plenum, using water and sodium as test fluids. The similitude relationship between a small-scale test and a real plant is discussed in order to understand the experimental results. The scale-model experiments for LMFBRs (liquid metal-cooled fast breeder reactors) were also performed to see the effects of a reactor configuration and reactor-trip operation condition. Then the magnitudes of the temperature gradient and the ascending speed of stratified interface in the hot plenum of LMFBRs were predicted, based on the results of the water scale-model.     

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.     

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.     

Experimental Study of Nuclear Characteristics of Large Axially Heterogeneous Core Using Fast Critical Assembly

The physics characteristics of large axially heterogeneous liquid-metal fast breeder reactors (LMFBRs), particularly the parameters for use in design and safety assessment, were examined using the JAERI fast critical assembly facility, arranged in Assembly XH-1, a partial mock-up of axially heterogeneous LMFBR. The properties measured were (1) criticality, (2) reaction rates and reaction rate ratios, (3) material sample worths, (4) sodium-void worths and (5) B4C control rod worths.

The results were compared with those of prior experiments with assemblies representing conventional homogeneous core. Confirmation was obtained of the typical nuclear characteristics attributed to axially heterogeneous LMFBRs, including flattening of the axial distribution of power and of the differential worth of control rod, as also lower sodium void worth.

Theoretical analyses paralleling the experiments, using JENDL-2 cross section library and JAERI standard calculation code system for fast reactor neutronics, resulted in some discrepancies, particularly for the internal blanket, in respect of plutonium sample worth, fission rate and fission rate ratio.     

Theoretical and experimental thermal-hydraulic analysis of a new sodium cooled fast breeder reactor core

This paper presents the experimental and theoretical results of the thermal-hydraulic design of a new fast breeder reactor core concept. The main feature of this concept is the omission of fuel element cans.The hydraulic function of these fuel element cans is substituted by a winding flow path through the radial blanket and a ring chamber without tubes.A computer code based on the quasi-continuum-theory and especially adapted to the features of the new core concept is developed for theoretical investigations. The pressure drop of the rod bundles is specified by a resistance tensor.The experimental investigations are realized in a test facility, where sodium is simulated by water. Pressures and velocities are measured.Theoretical and experimental results show good agreement. The aim of flattening of the coolant outlet temperature distribution can be reached with satisfying accuracy.     

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.     

Thermal transient and the temperature profile in a HELICA mock-up simulated by a new finite element homogenous model

This paper deals with a numerical approach for simulating the thermal and mechanical behaviour of pebble beds used as breeder and neutron multiplier in breeding blanket of nuclear fusion reactor. The model of the pebble beds is based on the results of a theoretical and experimental research activity performed by the Authors on ceramic pebble beds (lithium ortosilicate and lithium metatitanate). The results of this activity permitted to determine the effective thermal conductivity of the beds, versus the temperature and the axial pressure and to implement a homogenous model of pebble bed in a FEM code.This paper illustrates an application of the implemented model, considering pebble beds under several cycles of heating and cooling. The examined geometry corresponds to the HELICA mock-up tested by ENEA in the research centre Brasimone. The experimental tests performed on HELICA have been used as a benchmark problem in order to assess the different approaches for simulating pebble beds. In this paper, the simulations performed with two-dimensional models are illustrated. Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported.     

Dynamic behavior of a solid particle bed in a liquid pool: SIMMER-III code verification

Dynamic behavior of solid particle beds in a liquid pool against pressure transients was investigated to model the mobility of core materials in a postulated disrupted core of a liquid metal fast reactor. A series of experiments was performed with a particle bed of different bed heights, comprising different monotype solid particles, where variable initial pressures of the originally pressurized nitrogen gas were adopted as the pressure sources. Computational simulations of the experiments were performed using SIMMER-III, a fast reactor safety analysis code. Comparisons between simulated and experimental results show that the physical model for multiphase flows used in the SIMMER-III code can reasonably represent the transient behaviors of pool multiphase flows with rich solid phases, as observed in the current experiments. This demonstrates the basic validity of the SIMMER-III code on simulating the dynamic behaviors induced by pressure transients in a low-energy disrupted core of a liquid metal fast reactor with rich solid phases.     

Three-Dimensional Analysis Method for Sloshing Behavior of Fast Breeder Reactor and Its Application to Uni- and Multi-Vessel Type FBRs

A three-dimensional analysis method for sloshing behavior of fast breeder reactor (FBRs) is developed. The method treats the coolant in a reactor vessel as a potential flow with moving liquid surfaces. The Laplace equation of a velocity potential is solved by a boundary element method with its boundary condition described by a Bernoulli equation.

The vibration test results of a rectangular water pool are calculated by the method. Then, the method is applied to analysis of sloshing behavior of uni- and multi-vessel type FBRs. The latter consists of vessels for the core, heat exchangers and pumps. These vessels are connected by piping. In the case of the uni-vessel type FBR, heat exchangers and pumps are placed in the reactor vessel. The characteristics of sloshing behavior of both the reactors are presented.     

Inter-subassembly heat transfer of sodium cooled fast reactors: Validation of the NETFLOW code

This paper describes the prediction of temperature at the exit of subassemblies of a sodium cooled fast reactor using the NETFLOW code. Until present time, this plant dynamics calculation code is expected as a tool of nuclear education, and has been validated using data obtained at facilities or reactors cooled with water or sodium. A natural circulation test was conducted in the experimental fast reactor ‘Joyo’ with a 100 MW irradiation core. Also a turbine trip test was conducted in the prototype fast breeder reactor ‘Monju’. These tests were chosen to validate a model to calculate inter-subassembly heat transfer consisting of heat conduction and heat transfer by inter-wrapper flow. Based on the calculation for the natural circulation test in primary and secondary loops of ‘Joyo’, the model to calculate the heat transfer in radial direction of the inter-subassemblies simulated reasonable sodium temperature behaviors at the exit of subassemblies. Good agreement was also obtained in prediction of temperatures at the exit of the ‘Monju’ subassemblies. Through these validations, it was shown that the one-dimensional plant dynamics code NETFLOW could trace temperatures at the exit of the subassemblies of fast reactors with the inter-subassembly heat transfer model.     

3D thermo-fluid MHD simulation of single straight channel flow in LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) blanket for its DEMO fusion reactor. The mock-up blanket (TBM), using this concept, will be tested in ITER for its tritium breeding and high-grade heat extraction efficiency. In this TBM, pressurized helium is used to remove the heat from first wall, top and bottom plates of TBM. The Pb–Li is used to extract heat from the breeder zones. The flow of Pb–Li with average velocity 0.1 m/s inside the channel can be significantly modified due to MHD effects, which arise because of the presence of strong toroidal magnetic field. A numerical approach is established to capture this flow modification at higher Hartmann numbers (≥20,000). As a validation part of the developed code, MHD phenomenon is studied in 2-D square geometry and numerically obtained velocity profile is compared with available Hunt's analytical results. Thermo-fluid MHD analysis using this code, has been carried out for single rectangular duct of LLCB TBM. The heat transfer has been studied by keeping hot breeders at both sides of the flow channel. The results suggest modification in steady state MHD velocity profile as the liquid flows along the flow length. However, the temperature in various zone remains well within the maximum allowable limit.     

Response of liquid metal fast breeder reactor containment to a hypothetical core meltdown accident

This paper presents an analysis of the response of the containment building of a 2500-MWt liquid metal fast breeder reactor to a hypothetical reactor core meltdown. Although not mechanistically justifiable, this type of event is chosen for analysis as a basis for risk assessments. Containment space atmosphere compositions, temperatures, pressures, and structural temperatures are calculated, based on decay energy release and chemical reactions associated with the incident. The CACECO containment analysis code, which was used to make the calculations, is described in detail.Results of the study show that by utilizing the passive heat absorption capability of structures normally present in containment design, reactor plant containment integrity can be maintained for more than a day, even for extreme hypothetical events.