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.     

Studies of core disruptive accidents and licensing aspects of fast breeder reactors

This paper discusses the role of the core disruptive accident (CDA) in the safety evaluations and licensing of Liquid Metal Fast Breeder Reactors (LMFBR). Parametric studies of transient overpower (TOP) accidents based on calculations for SNR-300 using the HOPE computer code are presented. Major uncertainties in TOP analysis are identified and discussed with emphasis on the need for reliable fuel failure criteria. A series of calculations illustrating the possible behavior of the U.S. LMFBR demonstration plant following a loss-of-flow (LOF) accident without scram using the SAS-IIIA computer code are described. It is shown that for a beginning of life (BOL) core and end of equilibrium cycle (EOEC) core, the reactivity effects from sodium voiding and clad motion can lead to further sustained reactivity additions from subsequent fuel motion and FCI driven sodium voiding. In these calculations we have used the fuel enthalpy criterion which predicts clad failure around the core midplane. For the EOEC case these effects can add sufficient reactivity to take the system above prompt-critical (LOF driven TOP) and into hydrodynamic disassembly. For the BOL case the sodium void may not be sufficient to bring the system near sustained prompt-critical. However, clad motion appears to be effective in raising the reactivity to prompt-criticality. These results are based on clad failure dynamics modeling in SAS-IIIA. Further work is needed in the area of fuel-clad behavior under severe transients before definitive conclusions can be drawn regarding the applicability of current clad failure models at high clad temperatures (>1000°C). The potential significance of a new concept in CDA analysis called the “transition phase” is briefly mentioned.     

Consequences of degraded core accidents

In risk studies, paramount attention is paid to hypothetical accidents with nuclear reactor core melt down because of possible physical health impacts resulting from this type of accidents. In the Federal Republic of Germany the research project “core melting” was initiated by the Federal Ministry of Research and Technology in 1971 in order to investigate melt down phenomena. One of the main objectives within the research project is the development of experimentally verified models and computed codes to predict the radioactive source term released to the environment during degraded core accidents. It is the purpose of this paper to summarize the present state of the art obtained so far and to show current results of consequence analyses. The analyses, which up to now do not cover all important fission product release paths, are valid for large 1300 MW KWU-type PWRs demonstrating low releases of cesium and iodine to the environment.     

Overview of core disruptive accidents

An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described.     

On certain aspects of pressure transient without scram in a BWR

This paper examines the effects of gamma and neutron heating and pressure wave propagation on the core response during an instantaneous loss of condenser vacuum transient without scram (ATWS) in a BWR. By incorporating the gamma and neutron heating, which contribute about 3% of the total power to the moderator, into the transient thermal-hydraulic analysis, the peak power was found to be 35% lower compared with the case with no direct heating. The incorporation of the two-dimensional radial and axial variations of reactivity feedbacks and moderator density into the transient analysis led to a lower power prediction than the one-dimensional model. The pressure surge was examined by a computer program based on the method of characteristics. The pressure rise calculated by this new code was found to be in good agreement with experimental data, while the results of similar calculations done by computer code RELAP4, which is based on finite differencing of the flow equations, were 50% lower.     

Experimental modelling of hypothetical core disruptive accidents

A common feature to reactor containment programmes is the use of detailed models to furnish data for design and safety assessment purposes. Despite the great strides which have been made in computational methods it is expected that the experimental approach will have a continuing role. It is therefore still pertinent to review the basis of such experiments, to see how they could be improved, and to see how well model experiments describe other processes occurring during an hypothetical core disruptive accident (HCDA).Numerous papers have described experiments on detailed models of a fast reactor scheme, and in all these, the sodium coolant of the reactor is replaced by water in the model for obvious practical reasons, but the scaling consequences of this change seem to have been given little attention. Therefore the object of this paper is to review the fundamentals of the scaling process, and then to discuss in more detail the effects of changing the working fluid in HCDA experiments.It is shown that the usual practice of using a geometrically scaled model, water as the working fluid, and a charge of the same characteristics as expected in the reactor excursion results in an inexact simulation, requiring somewhat uncertain corrections before the data can be used for the reactor case. An alternative possibility which is discussed in this paper would be to model the compressible characteristics of the sodium and the results could then be applied directly to the reactor scale using well defined scaling factors. This proposal, however, does require detailed changes to the experimental model and to the charge, but neither of these is expected to give undue difficulty.Modelling of an HCDA normally refers to modelling of the compressible fluid/structure interaction but in recent years interest has grown in other processes, such as heat and mass transfer. By looking at the appropriate dimensionless numbers in the model and reactor, the possibilities of using scale experiments to investigate certain features can be gauged. It is concluded that with experiments using water as the working fluid many processes associated with heat and mass transfer will not be modelled correctly and therefore special experiments have to be devised. For the same reason, caution should be used in extrapolating to the reactor heat and mass transfer data from experiments designed to reproduce structure deformation and loading.Although the modelling of compressible fluid/structure interactions is without doubt the main interest at the present time, other processes can be modelled without difficulty. In the example given, it is shown that buoyancy effects can be modelled provided an incompressible fluid simulation is sufficient. This simulation requires a low pressure charge such as might be provided by the evaporation of FREON released from a frangible container.     

On the failure modes of alternative containment designs following postulated core meltdown

The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR representative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best estimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. These include the temperature of the non-condensibles escaping the ice condenser into the upper compartment, the performance of the pressure suppression system, the distribution of non-condensibles between compartments, and the degree and rate of combustion of hydrogen generated from metal-water reactions. For the PWR ice condenser case, results indicate that the containment would be breached by (i) steam overpressurization during the blowdown period (time less than 20 sec) if the ice condenser fails to perform its function, (ii) by overpressurization and thermal stress during the core melt period if 25% or more of the core zirconium reacts with water followed by hydrogen burning and, and (iii) by the overpressurization due to non-condensibles before containment floor penetration is completed. For the BWR Mark III case, similar conclusions can be drawn for the loss of vapor suppression, and for the hydrogen burning if the extent of zirconium-water reaction is more than 35% of the core inventory. If the hydrogen burning fails to materialize, the containment can retain its integrity until containment meltthrough provided the melting is confined to the reactor pedestal area. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded.     

Specifics of RBMK core cooling in beyond design basis accidents

The most dangerous beyond design basis accidents for RBMK reactors, leading to the worst consequences, are related to the loss of long-term heat removal from the core. Due to a specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using control rods cooling circuit, depressurisation of reactor cooling system, supply of water into cooling system from low pressure water sources, etc. This paper presents the analysis of such heat removal by employing RELAP5, RELAP5-3D and RELAP/SCDAPSIM codes. The analysis was performed for Ignalina nuclear power plant with RBMK-1500 reactor. The analysis of result shows that the restoration of water supply into control rod channels enables to remove 10-30 MW of the generated heat from the reactor core. This amount of removed heat is comparable with reactor decay heat in long-term period and allows to slowdown the core heat-up process. However, the injection of water to reactor cooling system is considered as main strategy, which should be considered in RBMK-1500 accident management procedure.     

钠冷快增殖堆钠雾火分析计算

在钠冷快增殖堆假想事故中 ,由于管道破裂 ,钠喷出到有氧的房间引起钠雾火 ,导致房间内温度及压力的上升。在NACOM单个液滴燃烧模型的基础上 ,考虑燃烧钠液滴的运动以及由于钠液滴与气体的热平衡关系 ,并忽略由于液滴间的相互作用影响 ,编制程序SPCOM。对钠雾火过程中涉及的液滴运动、液滴燃烧、喷雾燃烧以及质量热量传递问题进行了模拟。计算了钠雾火引起的房间的温度及压力瞬变 ,并与实验进行了比较 ,符合良好     

Application of containment codes to LMFBRs in the united states

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.     

On the analysis of the predisassembly phase of the unprotected overpower transient accident for LMFBRs

A simplified model of the predisassembly phase for unprotected overpower transients in LMFBRs is presented. The model includes fuel movement and sodium expulsion as a result of the molten fuel-coolant thermal interaction in the channels. The SNR-300, a prototype demonstration plant, is analysed and the results compared with an earlier set of calculations using the SAS-VENUS computational models. The previous analysis neglects fuel movement. The results indicate that substantial reductions in the severity of the transient as measured by molten fuel produced, the thermal energy generated and the final temperatures, can be obtained when fuel movement is included.     

Progress in core and cermet fuel modelling to calculate severe accidents

The use of cermet type composite fuels leads to an optimised use of plutonium; a good thermomechanical behaviour due to a low operating temperature thanks to a high thermo-conductivity, that favours high burn-up due to the low fission gas release. However, the increase in the metallic mass, an alloy of zircaloy, in the core, as well as the composite nature of the fuel with two very different melting temperature ( 1873 K for the metal, and 2573 K for the ceramic) lead to a behaviour very different from that of the traditional ceramic fuel in the event of an accident.     

Integrated phenomenological analysis of containment response to severe core damage accidents

Analysis of physical and radiological conditions inside the containment building during a severe (coremelt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g. a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants, which highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite. Results are also presented for applications of CONTAIN to the quantitative estimation of uncertainties in the source term which arise because of uncertainties in containment phenomenology. Taken together, the results described here show that analyses with nonintegrated, separate-effects codes can neglect interactions that are important to the source term and, furthermore, it is impossible to generalize whether the errors in such treatments would be ‘conservative’ or ‘nonconservative’. In many cases, it was not possible to predict in advance whether the phenomenological couplings would prove important or what the nature of their effects would be; the integrated treatment was required in order to obtain even qualitative answers to these questions. It is concluded, therefore, that integrated phenomenological analysis will play an increasingly important role as the technology for severe accident analysis matures.     

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.     

Radiological aspects of fusion reactor safety: Risk constraints in severe accidents

A methodology is proposed for determination of the constraints on severe accidents in lithium cooled fusion reactors, based on the potential hazards associated with such accidents. The method utilizes a probabilistic approach to risk calculation. The most effective mechanism for activation product release is found to be volatilization of structure as a result of lithium fires. Several factors were found to influence the consequence of lithium fires, most notably the reactor structural material type and total volume. It is concluded that the consequences of estimated maximum possible release from a properly designed fusion reactor are substantially less than the maximum light water reactor accident consequences.     

Dynamic structural response of LMFBR head closures to hypothetical core disruptive accidents

Presented here is an investigation of the dynamic structural response of the primary vessel's head closure to a hypothetical core disruptive accident (HCDA). Two head-closure designs were considered: the first represents a loop-type design and the second represents a pool-type design. Using representative configurations of liquid metal fast breeder reactors (LMFBR), independent models were used (1) to derive loading pressure histories and (2) to study the structural response of the head closures. Results for loading pressures, displacement histories, deformed profiles, stress magnitudes and plastically deformed regions are presented.     

Characterization of an energy source for modeling hypothetical core disruptive accidents in nuclear reactors

The experimental methods for and results of determining the expansion characteristics of the detonation products of an energy source that simulates the pressure-volume change relationships for sodium vapor expansions during hypothetical core disruptive accidents in a fast test reactor are presented. Rigid cylinder-piston experiments performed at two scales (ratio 1:3) to determine a pressure-volume relationship as a function of source mass and expansion environment are described. Some of these measurements are compared with code calculations for the source. The results show: (1) that the pressure-volume relationship depends significantly on the presence of water in the cylinder and comparatively little on the timescale of the expansion, the presence of steel balls in the water, or a Mylar sheet on the water surface; and (2) the experiment's scale. A relationship between the measured work energy from the source and the charge mass is presented, and pressure-volume change measurements are compared with previous experimental measurements and with theoretical calculations for a 150 MWsec hypothetical core disruptive accident. The measurements and code calculations of the pressure-volume relationship for the source agree reasonably well.     

Preliminary assessments of carbide fuel pins during mild overpower transients in LMFBRs

Mixed carbide fuel has been proposed for advanced liquid metal fast breeder reactors (LMFBRs). In this paper the general characteristics of carbide fuel are reviewed and a survey of available data for use in safety studies is presented. A preliminary investigation of the unprotected transient overpower accident sequence postulated for a large carbide fueled reactor is also presented. The HOPE computer code is the computational tool used to analyse a conceptual core design fueled with mixed carbides of uranium and plutonium. The details of this design are presented and compared with those of the Clinch River Breeder Reactor Plant. Results of HOPE calculations for both helium and sodium bonded carbide fueled cores are presented and compared with a similar study performed on an oxide core. Conclusions are drawn based on the findings of the preliminary transient overpower studies conducted.     

Dynamical models and numerical simulation of system-wide transients in loop-type LMFBRs

Dynamical models and numerical methods for a digital simulation of protected transients in loop-type LMFBRs resulting in EPRI-CURL code are presented. The model is capable of simulating operational transients, anticipated incidents, and postulated accidents which do not lead to sodium boiling. The dynamical models include: point reactor kinetics, primary, intermediate, and tertiary system heat transfer and coolant flow dynamics governed by forced and natural convection effects; and plant protection and control systems. A numerical method is incorporated which calculated the characteristic times of the 489 state variables modeling the entire system, and compares them with a variable preset integration timestep. A Runge-Kutta algorithm is applied to those state variables with moderate and slow response, and a quasistatic approximation is applied to those with rapid response; i.e., the ‘stiff’ equations. This assures numerical stability and is shown to greatly reduce the computation time requirements without much sacrifice in accuracy. The steady state (quasistatic) equations are further utilized to determine the unperturbed state of the system prior to transient initiation. The system response to a complete loss-of-electric power leading to buoyancy-induced natural circulation is calculated and compared to parallel calculations using DEMO and SSC-L simulation models.