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.     

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.     

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.     

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.     

Response of gas-cooled fast breeder reactors to depressurization accidents

The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV).The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient reponse following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor.Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.     

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal–fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. Simulations of GEYSER and THEFIS experiments were performed for dispersion and freezing behaviors of molten materials in narrow flow channels. In particular, the latter experiment using melt–solid mixture is also related to fundamental behavior of low energy disruptive core. CABRI-TPA2 experiment was simulated for boiling behavior of molten core pool. Expected mechanism of heat transfer between molten fuel and steel mixture was reproduced by the simulation. Analyses of structural dynamics using elastoplastic mechanics and fracture criteria were performed for SCARABEE BE+3 and CABRI E7 experiments. These two analyses are especially focused on thermal and mechanical failure of steel duct wall and fuel pin, respectively. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.     

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.     

Investigation of material relocation phenomena during severe accidents in fast breeder reactors

The SIMBATH out-of-pile experiments simulate severe accidents in fast breeder reactors. In the tests the nuclear energy released is substituted by the exothermal energy of a thermite reaction. Single pin and small bundle experiments as well as freezing tests are performed. Material ejected from the fuel rod simulators in an early phase is finely dispersed. A portion penetrates the upper breeding zone without freezing. The bulk of molten material ejected afterwards leads to blockages in the colder zones of the bundle. Under these conditions bottled-up situations may occur in the SIMBATH experiments. Residual sodium may become entrapped. The current version of the computer code CALIPSO developed to interpret these experiments is verified by calculation of two single pin experiments. The computations show that the relocation mechanisms in the SIMBATH experiments are mainly controlled by expansion of noncondensible gases originally existing inside the pins. The contribution from fuel vapour pressure or from sodium evaporation due to fuel-coolant-interaction is of less importance during the first 100 ms after fuel pin failure.     

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.     

Study on jet breakup behavior at core disruptive accident for fast breeder reactor

It is important to estimate the cooling possibility of the molten jet in coolant during a core disruptive accident (CDA) of a fast breeder reactor (FBR). In the present study, the molten jet of U-alloy78 simulating the core material is injected into the water simulating the coolant. The visual data of the molten jet breakup behavior is observed by using the high-speed video camera. The front velocity of the molten jet is estimated by using the image processing technique from the visual data. It shows that the front velocity of the molten jet can be divided into three time regions. In the first region, the front velocity of the molten jet increases. In the second region, the front velocity of the molten jet suddenly decreases. In the third region, the front velocity of the molten jet keeps at low and steady. In first region, the column diameter of the molten jet decreases with the passage of time. At the location between first region and second region, the column of the molten jet breaks up and disappears. In the present study, the jet breakup length is defined as the distance from the water surface to the location where the jet column disappears. The results show that the jet breakup length depends on the injection nozzle diameter, but does not depend on the jet penetration velocity. This tendency agrees with the prediction by Epstein's equation. After the experiment, the solidified fragments are collected and the mass median diameter is measured. The mass median diameter is compared with the existing theories. Furthermore, a model to estimate the cooling possibility during a CDA of a FBR is constructed, reflecting the above-mentioned results.     

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

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

Integrated fuel-cycle models for fast breeder reactors

Breeder-reactor fuel-cycle analysis can be divided into four different areas or categories. The first category concerns questions about the spatial variation of the fuel composition for single loading intervals. Questions of the variations in the fuel composition over several cycles represent a second category. Third, there is a need for a determination of the breeding capability of the reactor. The fourth category concerns the investigation of breeding and long-term fuel logistics. Two fuel-cycle models used to answer questions in the third and fourth area are presented.The space- and time-dependent actinide balance, coupled with criticality and fuel-management constraints, is the basis for both the Discontinuous Integrated Fuel-Cycle Model and the Continuous Integrated Fuel-Cycle Model. These models are derived by formally condensing the base equations through spatial integration. Cycle-average isotopic microscopic reaction rate and reactivity-worth coefficients, based on equilibrium behavior, complete the derivation of the discontinuous model. Approximating the discontinuous reload with continuous charge and discharge currents transforms the discontinuous model into the more mathematically elegant continuous model.The results of the continuous model are compared with results obtained from detailed two-dimensional space and multigroup depletion calculations. The continuous model yields nearly the same results as the detailed calculation, and this is with a comparatively insignificant fraction of the computational effort needed for the detailed calculation. Thus, the integrated model presented is an accurate tool for answering questions concerning reactor breeding capability and long-term fuel logistics.     

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.     

Unanticipated accidents in sodium-cooled fast reactors

The definitions and requirements of normative documents for unanticipated accidents at nuclear power plants with fast reactors are analyzed. Definitions are constructed between one another and with a collection of scenarios which can lead to unanticipated accidents, likewise determined by normative documents independently of the probability of these accidents actually happening. It is concluded that the normative approaches to fast-reactor safety must be refined with respect to strengthening the probabilistic criteria as a tool limiting the list of required unanticipated accidents for validating reactor safety. Special attention is devoted to the need to strengthen the motivation of designers to make the maximum possible use of passively triggered safety systems.     

Molten fuel dispersal in hypothetical fast reactor accidents

Some aspects of molten fuel dispersal in hypothetical fast reactor accidents are considered, ranging from the two-phase flow fluid equation forms appropriate to modelling molten fuel dispersal to an analytical self-similar solution as a function of space and time for the dispersal of a two-phase molten fuel/fission gas mixture by fission gas pressures. The analytical solution provides both scaling laws for fuel dispersal velocities as a function of gas content and a solution with which to check code results. A discussion of the COMCYL program for molten fuel dispersal in a sodium volded channel is used to illustrate the types of problems that need to be tackled in a molten fuel dispersal program and typical results obtained from the application of COMCYL to hypothetical loss-of-coolant flow accidents are presented.     

Cellular convection in vertical annuli of fast breeder reactors

In the pool type fast reactors the roof structure is penetrated by a number of pumps and heat exchangers that are cylindrical in shape. Sandwiched between the free surface of sodium and the roof structure, is stagnant argon gas, which can flow in the annular space between the components and roof structure, as a thermosyphon. These thermosyphons not only transport heat from sodium to roof structure, but also result in cellular convection in vertical annuli resulting in circumferential temperature asymmetry of the penetrating components. There is need to know the temperature asymmetry as it can cause tilting of the components. Experiments were carried out in an annulus model to predict the circumferential temperature difference with and without sodium in the test vessel. Three-dimensional analysis was also carried out using PHOENICS CFD code and compared with the experiment. This paper describes the experimental details, the theoretical analysis and their comparison.     

Calculation of a materials relocation experiment simulating a core discruptive accident condition in fast breeder reactors

This paper describes an interpretation of the SIMBATH (Simulationsexperimente in Brennelementattrapen mit Thermit) experiments that use the code. A series of SIMBATH experiments has aimed at simulating fuel pin disintegration and following materials relocation in the test sections of a single pin to 37-pin bundles. In the calculation, three modifications were incorporated into the code. With these modifications, the calculation showed good agreement with the experimental measurements with respect to the void region propagation in sodium flow and the molten materials relocation leading to flow blockage. A set of parametric calculations has clarified the range of applicability of parameters for materials relocation and flow blockage formation. The particle radius r_p in blockage regions and the multiplier for particle viscosity (PARVIS) are recommended to be

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respectively.     

Fuel performance of hollow pellets for fast breeder reactors

Three fuel rods containing hollow mixed oxide (MOX) pellets of uranium and plutonium oxides were fabricated and irradiated at a high linear heat rate (LHR) to burn-up of nearly 30,000 MWd/t in the experimental fast rector, JOYO MK-II. After irradiation, one of the fuel rod pellets was examined by X-ray CT and conventional nondestructive and destructive methods.

Swelling rate was evaluated by both dimensional change and radial density distribution. There were no differences between both types of results and it was concluded that swelling rate can be examined in detail by the X-ray CT technique without dismantling the assembly. In addition, the swelling rate of hollow pellets was nearly the same as values reported for the fuel rods containing solid pellets. LHR was higher in the examined fuel rod containing hollow pellets than in the conventional fuel rod containing solid pellets, but fission gas release rates for both fuel rods were nearly the same.