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.     

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.     

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.     

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.     

Analysis of LMFBR primary system response to an HCDA using a eulerian computer code

This paper describes the use of the Eulerian code to predict the response of the fast reactor containment and the primary piping loops to HCDAs, and to analyze the sodium-spillage and the bubble motion problems which cannot be analyzed with a Lagrangian code. The basic equations and numerical techniques used in the Eulerian computer code are described in detail. Four sample problems are given. The first problems are given. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the slug impact on the reactor head and the spillage of sodium through the opening holes. The third problem deals with the dynamics of an HCDA bubble. The fourth problem deals with the response of a piping loop.     

Experimental and numerical methods related to safety factors in the PEC fast nuclear reactor

In this paper the author summarizes the activity of structural analysis related to the safety of the PEC fast nuclear reactor. There are two principal aspects of safety concerning problems of structures: the localized incident and the hypothetical core disruptive accident (HCDA).With regard to the first point, the phenomenon is dependent on the hydrodynnamic and structural behaviour of the fuel elements. With regard to the HCDA, it is necessary that the reactor vessel is able to absorb the explosive energy, whereas the plug must not sustain movements such as to alter the overall seal of the installation.Given the complexity of the phenomena, therefore, it was considered necessary first of all to carry out numerous experimental tests on both full-size and reduced scale models. The experimental tests on the individual hexcan, on the group of seven hexcans and on the vessel were carried out at the EURATOM centre of Ispra, in the context of a collaboration agreement between ENEA and EURATOM.Some of the results of these tests are presented in this paper, together with relevant comparisions with the numerical values.     

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.     

Structural shock tests of prototype FBR “MONJU” scale models

This paper presents some results of experiments which simulate the structural dynamic response of a LMFBR primary coolant boundary to a hypothetical core disruptive accident (HCDA) based on scale models and high explosives. It was noted that high explosives are no longer a good simulant of the HCDA. However, the main purpose of the program, which included this experiment, is not to experimentally predict the dynamic response of the reactor structure at the HCDA, but to validate computer codes, which describe the pressure wave propagation and damage process in the reactor structures, using data obtained from these model experiments. The experiments were undertaken using many 1/15 scale simple models of the reactor vessels and internal structures, as well as 1/15 and 1/7.5 scale complex models of the interim design of prototype LMFBR ‘MONJU’. Simple model experiments involved a series of shock tests using pentolite to investigate the configuration effects of the vessel restraining section, the dipped-plate effect and the core barrel effect, respectively.     

An investigation of containment pressurization by sodium spray fires

An investigation of the sodium spray burning phase of LMFBR hypothetical low probability core disruptive accidents (HCDAs) has been undertaken in order to test the response of various containment designs. The HCDAs are produced by arbitrarily inserting unrealistically large amounts of reactivity in a short period of time. The spray fires result from a HCDA which causes head failure due to high-velocity impact by the sodium pool followed by rotating plug jump, loss of rotating plug seals, control rod ejection, failure of instrument tubes, or breach of in-vessel transfer machine ports. Head failure can in principle lead to the injection of significant amounts of sodium into the reactor containment building by residual pressure of the HCDA gas bubble, which forces the upper plenum sodium through the interstitial spaces (e.g., rotating plug gaps, control rod housings, etc.) in the breached head structure.Calculations were made of the hydraulic behavior of the sodium under various injection scenarios for both pool and loop reactor systems. In the case of plug jump, although massive amounts of sodium can be injected into the containment building, the injection will be primarily in a radial direction, and the major consequences could be a low-intensity pool fire rather than a high-intensity sodium spray fire. However, if the bearing housing on the rotating plug fails, a 20 atm initial HCDA residual bubble pressure has the potential for injecting a sodium stream through the rotating plug gaps which could potentially impact the containment building ceiling. Sodium discharges through broken control rod housings could also impact the ceiling and become widely dispersed.The SOMIX-1 sodium spray fire code was used to calculate the energy releases corresponding to a variety of head failure scenarios corresponding to the cases where a high velocity jet impinges on the ceiling of the containment building. The calculated maximum pressure rise was about 2.1 atm. The analysis showed that containment building pressures do not always increase with increasing sodium injection rates since the oxygen concentration can be reduced to a level where the spray begins to cool rather than heat the gas.     

Analysis of HCDA

Safety regulation requires that the containment of fast reactors be analyzed to a spectrum of hypothetical core-disruptive accidents (HCDAs). To demonstrate the reactor containment can sustain the consequences of HCDAs, studies must be made to understand the reactor core behavior under HCDA conditions, to establish the consequences of HCDAs, to provide features to mitigate the effects of HCDAs, and to determine the response of the reactor primary system to HCDA loads. This paper reviews some approaches and methods used in the experimental and numerical analyses.     

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.     

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.     

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.     

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.     

Two-dimensional numerical simulation of single bubble rising behavior in liquid metal using moving particle semi-implicit method

Gas-lift pump in liquid metal cooling fast reactor (LMFR) is an innovative conceptual design to enhance the natural circulation ability of reactor core. The two phase flow characteristics of gas–liquid metal make significant improvement of the natural circulation capacity and reactor safety. It is important to study bubble flow in liquid metal. In present study, the rising behaviors of a single nitrogen bubble in 5 kinds of common stagnant liquid metals (lead bismuth alloy (LBE), liquid kalium (K), sodium (Na), potassium sodium alloy (Na–K) and lithium lead alloy (Li–Pb)) and in flowing lead bismuth alloy have been numerically simulated using two-dimensional moving particle semi-implicit (MPS) method. The whole bubble rising process in liquid was captured. The bubble shape, rising velocity and aspect ratio during rising process of single nitrogen bubble were studied. The computational results show that, in the stagnant liquid metals, the bubble rising shape can be described by the Grace's diagram, the terminal velocity is not beyond 0.3 m/s, the terminal aspect ratio is between 0.5 and 0.6. In the flowing lead bismuth alloy, as the liquid velocity increases, both the bubble aspect ratio and terminal velocity increase as well. This work is the fundamental research of two phase flow and will be important to the study of the natural circulation capability of Accelerator Driven System (ADS) by using gas-lift pump.     

液态金属内单个气泡上升行为的MPS法数值模拟

液态金属冷却核反应堆采用气泡泵的概念设计来提升堆芯自然循环能力。液态金属内气液两相流动特征将直接影响核反应系统一回路的自然循环能力及堆芯安全。本研究通过采用移动粒子半隐式（MPS）方法,对液态金属中单个上升气泡的气泡动力学行为进行数值模拟。分析了铅铋合金中3种初始直径不同的单个氮气泡在上升过程中的气泡形状和速度的变化趋势;对比了初始直径相同的单个氮气泡在液钾、液钠、铅铋合金、钾钠合金和锂铅合金5种液态金属中的上升行为;同时将模拟得到的气泡形状与Grace经验关系图进行了对比,验证了MPS方法数值模拟结果的正确性。     

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.