Fast Breeder Reactor Core Research Articles

Application of genetic algorithm for optimization of control rods positioning in a fast breeder reactor core

In this work, an integer coded genetic algorithm is implemented for finding the optimal arrangement of twelve absorber rods (control rods) within the active core of a fast breeder reactor, satisfying and optimizing operational and safety parameters. This is a multi-objective, multi constraint combinatorial optimization problem. The problem is having a large search space due to the huge number of control rod arrangements possible. A chromosome in genetic algorithm, which denotes a particular arrangement of control rods, is represented as a set of 12 integers. Each integer stands for one control rod and denotes the position of the rod in the reactor core. The genetic operations, crossover and mutation are modified and adapted for this particular study to take into consideration the geometric positioning of control rods within the core. The results obtained show that the algorithm is able to converge to optimal, feasible configurations by exploring the search space effectively. The performance of algorithm is further enhanced by applying parallel programming techniques in the evaluation of genetic population. The study attempts to validate the effectiveness of genetic algorithms in handling objectives and constraints related to reactor physics as well as engineering design domains.

NEMHEX solution of benchmark problems on thermal and fast breeder reactor cores with hexagonal lattice

This paper gives the solution of some benchmark problems for reactor core calculations in hexagonal geometry using a computer code NEMHEX (Nodal Expansion Method for HEXagonal geometry). The problems cover a wide range of thermal reactor cores including large heavy water, fast breeder and VVER reactors in both two and three dimensional. The code is based on higher order nodal expansion method (NEM) for the solution of the neutron diffusion equation in 3-D hexagonal geometry. In this method, flux in each mesh/node and energy group is expanded in terms of polynomial up to third order in X–Y plane and fourth order in axial direction (Z). Weighted residual method is used to calculate the higher order expansion coefficients of the polynomials. Both the partially integrated radial and axial leakages are approximated by the quadratic polynomial. Higher order polynomial expansion [with respect to the work by Lawrence (DIF3D nodal neutronics option for two-and three-dimensional diffusion theory calculations in hexagonal geometry, 1983)] of flux as well as leakage in each node improves the accuracy in power distribution of VVER-1000 type reactor (which is having large size lattice) in terms of maximum reduction in error from 8 to 2%. The final inhomogeneous response matrix equation involves spatial moments of nodal flux and surface averaged partial currents across the surfaces of node. These equations are solved using conventional fission source iteration. This code is used for calculation of effective multiplication factor, neutron flux and fission power distribution. The results of the analysis establish the adequacy of NEMHEX for thermal as well as fast reactor core calculations in hexagonal geometry.

Swelling of improved 16Cr–15Ni–2Mo–Mn–Ti–V–B steel under dose rates from 1 × 10−8 to 1.7 × 10−6 dpa/s

Radiation-induced swelling negatively affects the operability of structural units of fast breeder reactor (FBR) cores. Therefore, search for new and improvement of existing steels for reducing swelling is an important task.Since 2003, type 16Cr–15Ni–2Mo–Mn–Ti–V–B steel has shown a significant increase in radiation resistance due to its improved composition and heat treatment. Specialists of the JSC INM studied swelling of 16Cr–15Ni–2Mo–Mn–Ti–V–B steel with improved composition, data on the maximum swelling temperature, average swelling rate within typical coolant temperature ranges, as well as fast reactor dose rate were obtained.The obtained results are based on swelling measurements using hydrostatic weighing and transmission microscopy. Errors in hydrostatic measurements were examined with involvement of metallography data and selection of immersion liquid.It was revealed that the average swelling rate of improved 16Cr–15Ni–2Mo–Mn–Ti–V–B steel at the maximum swelling temperature is within the range of 0.04–0.14%/dpa. Shifting of this temperature from 460 to 520°C with increase of the maximum damaging dose from 60 to 80dpa (1.3× 10−6 and 1.7×10−6dpa/s, respectively), is observed. At doses below 10dpa and temperatures below 400°C the average swelling rate may reach 0.04%/dpa. At temperatures of about 600°C and irradiation doses below 50dpa the swelling rate does not exceed 0.01%/dpa during the whole period of observation.

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

A new approach to nuclear reactor design optimization using genetic algorithms and regression analysis

A module based optimization method using genetic algorithms (GA), and multivariate regression analysis has been developed to optimize a set of parameters in the design of a nuclear reactor. GA simulates natural evolution to perform optimization, and is widely used in recent times by the scientific community. The GA fits a population of random solutions to the optimized solution of a specific problem. In this work, we have developed a genetic algorithm to determine the values for a set of nuclear reactor parameters to design a gas cooled fast breeder reactor core including a basis thermal–hydraulics analysis, and energy transfer. Multivariate regression is implemented using regression splines (RS). Reactor designs are usually complex and a simulation needs a significantly large amount of time to execute, hence the implementation of GA or any other global optimization techniques is not feasible, therefore we present a new method of using RS in conjunction with GA. Due to using RS, we do not necessarily need to run the neutronics simulation for all the inputs generated from the GA module rather, run the simulations for a predefined set of inputs, build a multivariate regression fit to the input and the output parameters, and then use this fit to predict the output parameters for the inputs generated by GA. The reactor parameters are given by the, radius of a fuel pin cell, isotopic enrichment of the fissile material in the fuel, mass flow rate of the coolant, and temperature of the coolant at the core inlet. And, the optimization objectives for the reactor core are, high breeding of U-233 and Pu-239 in desired power peaking limits, desired effective and infinite neutron multiplication factors, high fast fission factor, high thermal efficiency in the conversion from thermal energy to electrical energy using the Brayton cycle, and high fuel burn-up. It is to be noted that we have kept the total mass of the fuel as constant. In this work, we present a module based (modular) approach to perform the optimization wherein, we have defined the following modules: single fuel pin cell, whole core, thermal–hydraulics, and energy conversion. In each of the modules we have defined a specific set of parameters and optimization objectives. The GA system (GAS), and RS together, play the role of optimizing each of the individual modules, and integrating the modules to determine the final nuclear reactor core. However, implementation of GA could lead to a local minimum or a non-unique set of parameters, those meet the specific optimization objectives. The GA code is built using Java, neutronic analysis using MCNP6, thermal–hydraulics calculations using Java, and regression analysis using R.

Analysis on Proliferation Resistance Factor and Fuel Breeding Capability Based on Even Mass Plutonium Isotope Compositions

Abstract Actinide recycling options of spent nuclear fuel are estimated to have many advantages such as reducing nuclear waste, increasing nuclear fuel utilization and improving proliferation resistance aspect of nuclear fuel. Composition of spent nuclear fuel consists of heavy nuclides and fission products including uranium and plutonium as main actinide. In addition, it is also consist of some minor actinides (MA) which can be recycled to reduce the volume and radioactive intensity of long-lived radioactive waste and in the same time to increase level of proliferation resistance. In general, some MAs can be utilized as useful material such as increasing nuclear fuel production and producing some even mass plutonium such as Pu-238, Pu-240 and Pu-242 as a control material for protected plutonium production. In this paper, identification of even mass plutonium isotopes production for the application of fuel breeding capability and proliferation resistant aspect of isotopic plutonium production was investigated. A large fast breeder reactor (FBR) model is adopted by using 800 days operation for one cycling operation for 4 fuel batches systems for optimization and analyses. The analysis of loading MA materials into the FBR core regions shows the Pu-238 production is increasing significantly as well as additional production of other plutonium isotopes. Isotopic Plutonium production of Pu-238 as main product can be estimated from fuel conversion process of MA in the reactor. This fuel conversion phenomena of Pu-238 can be identified as fertile material for fuel breeding capability, that Pu-238 can be converted to Pu-239 by neutron capture which is similar process also for fuel conversion of Pu-240 to Pu-241 as fissile materials. In addition, isotopic production of Pu-239 from converted Pu-238 as additional fuel which can be added with some quantities of Pu-239 production from converted U-238 for extending reactor operation. Production of Pu-238 in the reactor are used for categorizing some level of protected plutonium productions which can increase isotopic barrier level of plutonium in term of decay heat (DH) as well as Pu-240 as additional even mass contributor.

Fuel behavior analysis code FEMAXI-FBR development and validation for core disruptive accident

Japan Nuclear Energy Safety Organization (JNES) is a Technical Safety Organization (TSO) for Japanese Nuclear Regulation Authority to provide technical information on safety reviews including incident and accident analyses of nuclear power plants including Fast Breeder Reactor (FBR). JNES has developed an ASTERIA-FBR code system as an integrated Core Disruptive Accident (CDA) analysis code of the regulatory side for the purpose of evaluating the Unprotected Loss-of-Flow (ULOF) accident of FBR cores. The ASTERIA-FBR consists of three main modules, a thermal-hydraulics calculation module; CONCORD, a fuel behavior calculation module; FEMAXI-FBR, and a neutronic calculation module; dynamic GMVP. The FEMAXI-FBR is based on a LWR fuel behavior simulation code FEMAXI-6 and modified the material properties, the calculation models for steady state and transient operational conditions. The FEMAXI-FBR has been validated by French CABRI slow-TOP (E12) and fast-TOP (BI2) transient test data. Furthermore, the fuel temperature calculation model was validated by B5D Power-To-Melt (PTM) test at the experimental fast reactor JOYO in Japan. Through these validation, the code is found to be capable of evaluating FBR fuel pin behavior from normal to accidental conditions precisely.

ZPPR Benchmarks for Large LMFBR Core Physics from JUPITER Cooperative Program Between United States and Japan

Under the International Reactor Physics Experiment Evaluation Project (IRPhEP) framework, in the cooperative JUPITER program between the United States and Japan, benchmarks are established to study large fast breeder reactor (FBR) core physics utilizing nine Zero Power Plutonium Reactor (ZPPR) critical experimental cores. These benchmarks cover a wide variety of core concepts including homogeneous and heterogeneous configurations, clean and engineering mock-up cores of 600- to 1000-MW(electric)–class sizes, and various core parameters such as criticality, reaction rate, and reactivity. Recently, detailed experimental information from original documents from Argonne National Laboratory has been scrutinized very carefully to establish the benchmark model and to evaluate quantitatively the experimental uncertainty. The benchmarks supply users with heterogeneous cell models and three-dimensional (3-D) core configurations, which are simplified to a degree that preserves the important physical features of the ZPPR cores such as plate heterogeneity, different drawer types, and 3-D core arrangement. Further, the benchmark handbook includes as-built information of the ZPPR cores as a complete set of electronic form; therefore, a user can develop his or her own benchmark model if necessary. The analysis of the benchmark with the deterministic or Monte Carlo method demonstrates its usefulness both for improving analytical methods and for validating nuclear data.

ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120MWe, 500MWe and 1000MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase.

Comparison and sensitivity analysis of the core characteristics of a sodium-cooled fast breeder reactor with 750 MWe output evaluated by JENDL-4.0 and ADJ2000R

Core characteristics of a sodium-cooled fast breeder reactor (FBR) with 750 MWe output using highly decontaminated uranium and plutonium and highly minor-actinide-containing compositions were evaluated using the fast reactor cross-section set generated by the new Japanese nuclear data library JENDL-4.0. The core characteristics were compared with those obtained using the unified cross-section set ADJ2000R in order to investigate the differences between both the results. The effects on the core characteristics caused by the differences in the nuclear data of important reactions and nuclides in the cross-section sets were analyzed by a burnup sensitivity analysis. It was confirmed that adopting JENDL-4.0 to the FBR core design improves the breeding ratio, the burnup reactivity, and the reactivity control balance, because of the differences in the capture cross-sections of U-238 and Pu-239 of both the libraries. The difference in the sodium void reactivity evaluated with both the libraries was less than 1% because the increase caused by the differences in the elastic scattering cross-sections of sodium, the inelastic scattering cross-section, and the μ-average value of U-238 was practically cancelled out by the decrease caused by the differences in the capture cross-sections of Pu-239, the inelastic scattering cross-section of iron, and the capture cross-sections of Am-241.

Study on FBR core concepts to increase proliferation resistance of plutonium in LWR–FBR transition period

The possibility to enhance proliferation resistance of discharged plutonium in fast breeder reactor (FBR) has been investigated in terms of reactor core-design aspects. The provisional target for proliferation resistance measures based on Saito's attractiveness (ATTR) is defined. It is found that a few percent of plutonium loading and/or Am/Cm loading, which come from various types of spent fuel, might satisfy the provisional target with a minimum impact on the core neutronic performances. On Am/Cm loading core, decay heat constraints for fuel-handling aspects are found to be important and should be considered in design. There is not significant change on the current developing scenarios for light water reactor–FBR transition period by applying the measures based on Saito's ATTR. It is found that applying Kimura's proposal, 15% 238Pu content requires about 7% MA loading fraction both in the core and the blanket, and it only can be applied at limited case. The period to consume minor actinides is shorter than to consume plutonium.

SADHANA facility for simulation of natural convection in the SGDHR system of PFBR

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled pool type fast reactor being constructed at Kalpakkam, India. PFBR has all the reactor components immersed in the pool of sodium and the fission heat generated in the core, is removed by the sodium circulating in the pool. During normal operation this fission heat is transferred by primary sodium to secondary sodium, which in turn transfers the heat to water in the steam generator for producing steam. The removal of the decay heat generated in the reactor core after the reactor shutdown is also very important to maintain the structural integrity of reactor core components. PFBR employs two independent systems namely, Operational Grade Decay Heat Removal system (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS) for decay heat removal. SGDHR system is a passive system working on natural convection to ensure the core coolability even under station blackout condition. It is very important to study the thermal hydraulic behavior of Safety Grade Decay Heat Removal system of PFBR to ensure its reliable operation. A scaled down model of the circuit, named SADHANA has been modeled, designed, constructed and commissioned for demonstration and evaluation of these systems. The facility has completed around 2000 h of high temperature operation. The performance of the experimental system is satisfactory and it meets all the design requirements. At 550 °C sodium pool temperature in test vessel the secondary sodium loop generated a sodium flow of 6.7 m3/h. These experiments have revealed the adequacy and capability of SGDHR system to remove the decay heat from the fast breeder reactor core after its shutdown.

Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no doping minor actinide (MA) case. Doping MA in the FBR blanket regions is effective to increase the material barrier level from weapon grade plutonium to MOX fuel or more grade level. Moreover, increasing material barrier of plutonium isotopes can be achieved effectively by increasing Pu-238 production as main contributor for LWR and FBR core regions, and from converted MA in the FBR blanket which affects directly to increase DH and SFN components.

Fast breeder reactor core concept for heterogeneous minor actinide loading

As an alternative method to the homogeneous minor actinide (MA) recycling in fast breeder reactors, a heterogeneous MA loading core concept using a highly concentrated americium (Am)-containing fuel (Am target) is proposed. By the use of an extraction process for Am and curium (Cm) in the reprocessing of the spent fuel, Am (and a small amount of Cm) can be recovered and then concentrated to produce the target. The Am content in the heavy metal is assumed to range from 10 to 20 wt% in accordance with the target development scope. A mixed oxide fuel that contains uranium, plutonium, and neptunium is chosen as the base material of the target, so that the targets can generate a level of power equivalent to that of the driver fuels. It was found that a ring-shaped arrangement of Am targets between the inner and outer core regions exhibits a favorable MA transmutation performance without any significant deterioration in the core neutronic characteristics, including increases of the burnup reactivity and sodium void reactivity worth, and decreases of the breeding ratio and absolute value of the Doppler coefficient, etc., in comparison with those of a reference homogeneous MA loading case. It should be noted that the Am targets in this loading arrangement can contribute to the suppression of the core power distribution change along with burnup. A series of core designs, including core neutronics, thermal hydraulics, and fuel integrity evaluations, was also carried out for a representative Am target loading case. The results indicate that it is possible to design an Am target subassembly that can cope with the issues presented by highly concentrated Am, i.e., the deterioration of thermophysical properties and the accumulation of helium gas inside the target fuel pins. Therefore, the design feasibility of the heterogeneous target loading core has been enhanced.

Correlations among FBR core characteristics for various fuel compositions

In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations.

Recent Application of Nuclear Data to Reactor Core Analysis in JAEA

eld. Sensitivities of ssion product concentrations of LWR burned fuel to nuclear data have been evaluated based on the depletion perturbation theory. Important nuclear data are specied for the accurate prediction of ssion product concentrations. Extended bias factor methods are applied for the uncertainty evaluation of a fast breeder reactor core. From the application, the extemded bias factor methods powerfully eliminate the uncertainty due to cross section errors by utilizing a number of integral data.

Dynamic neutron computer tomography technique forvelocity measurement in liquid metal flow- Fundamental PTV experiment -

The aim of this development is to visualize and measure the velocity distribution inliquid metal flow using the neutron beam with the high-speed imaging technique, computertomography (CT) technique and particle tracking velocimetry (PTV). Final research purpose isto obtain the velocity distribution and flow profile data of liquid metal flow in a heated rodbundle for development of the advanced fast breeder reactor (FBR) core. In this paper,visualization and measurement method using the high intensity and large size neutron beamport of the research reactor JRR-4, design and manufacturing of the experimental apparatus,spring model PTV method for this technique and results of the fundamental PTV experimentwere reported. The test section for the fundamental experiment was a revolving aluminumcolumn with cadmium tracers which simulated the liquid metal flow. As the result, cadmiumtracers buried in the column with the speed of 1.5 revolving per second could be visualized asthe 3D movie under 125Hz and 250Hz sampling conditions, the profile of the tracer could betraced, and fundamental velocity distribution measurement method could be conformed.

Neutronic and nonproliferation characteristics of (PuO 2–UO 2) and (PuO 2–ThO 2) as fast reactor fuels

Computational core physics analysis carried out for a typical fast breeder reactor (FBR) core design is presented through two case studies; one using only (PuO 2–UO 2) MOX fuel and another that replaces ∼41% of (PuO 2–UO 2) MOX with (PuO 2–ThO 2) MOX while conserving the total Pu content. The basic computational framework employed uses the MONTEBURNS2 PERL script to couple the neutronic code, MCNP5 with the depletion/burn-up code, ORIGEN2.2. The parameters computed and compared are: net neutron multiplication factor ( k eff); regionally averaged neutron spectrum; neutron flux; thermal power distribution; breeding ratio; fuel burn-up; fissile material build-up/depletion- 232U build-up; and fuel temperature dependent Doppler Effect, coolant temperature dependent sodium expansion coefficient, and nonproliferation characteristics such as dose rates, spontaneous fission and gamma emissions. The analyses of the case studies indicate that the core physics characteristics, except k eff, are only marginally different in their magnitudes between the two cases, if not equal. The first case study shows that diversion of either 8 radial blanket sub-assemblies (weapon grade Pu) or 1 spent fuel sub-assembly (reactor grade Pu) discharged from an equilibrium core is sufficient to derive a significant quantity (SQ). The second case study shows that a considerable improvement in proliferation resistance can be achieved with the peripheral loading of (PuO 2–ThO 2) MOX pins in all the fuel sub-assemblies of a fast reactor, which should aid in nuclear material safeguards. The comparison of transuranic (TRU) generation for both the cases showed that about 60% reduction in neptunium production has been achieved for the new proposed partially ThO 2–PuO 2 loaded FBR design, whereas other higher TRUs like americium, curium, etc. did not show significant reduction.

Assessment of the FBR Core Disruptive Accident (CDA): The Role and Application of General Behavior Principles (GBPs)

General behavior principles (GBPs) first introduced in 1976 (H. K. Fauske, the Role of Core-Disruptive Accidents in Design and Licensing of LMFBRs, Nuclear Safety, Vol. 17, No. 5, 550–567, 1976) are applied to assess the core disruptive accident (CDA) outcome for a large (3,500MWt) high power density oxide-fueled sodium-cooled fast breeder reactor (FBR). Non-energetic termination in the so-called Initiating Phase is emphasized and is made possible by considering departures from the traditional core design including the Subassembly Inner Duct and Limited Blanket Removal concepts. This early accident scenario termination eliminates potential concerns raised related to recriticality events and also facilitates the potential for in-vessel fuel debris coolability for the large FBR with a compact reactor vessel.

Liquid–liquid interface stability in accelerating and constant-velocity tube flows

This paper presents new experimental data on interface stability when one liquid displaces another in a small-diameter tube. Interface stability theory suggests that the interface is stable when the displacement is directed from a dense and viscous liquid to a less dense and less viscous liquid. This theoretical result has important implications in the evaluation of a fast breeder reactor (FBR) core disruptive accident in that it rules out coolant-boiling-driven compaction of the reactor-core fuel, and has been classified by Fauske (Nucl. Safety, 17 (1976) 550) as one of several general behavior principles (GBPs) which may be used to argue against the occurrence of energetic events that could threaten the reactor vessel. The experimental data agreed well with the GBP for accelerated liquid/liquid systems with liquid density ratios and interfacial tensions similar in magnitude to those of the FBR mixed oxide fuel/sodium coolant pair. The data agreed well with the GBP when the displacement occurred in the laminar regime at constant speed and the viscosity of one of the liquid components was destabilizing. The investigation also uncovered behavior that was not in concert with the GBP, although this behavior appears to be limited to small density-difference, low-interfacial-tension liquid/liquid systems and is probably not relevant to the FBR application.