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.     

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.     

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.     

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.     

A three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor

The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core, including control rods, SIMMER-IV has been developed as a direct extension of SIMMER-III to three dimensions with retaining exactly the same physical models as SIMMER-III. Recently, the parallelization of SIMMER-IV has been achieved, allowing application to reactor calculations within available computational resources. A three-dimensional simulation using SIMMER-IV has drawn more realistic accident scenario including a late stage during the transition phase. Additional static neutronic calculations identified major factors significantly influencing the reactivity change shown in the SIMMER-IV simulation.     

Safety of sodium-cooled fast reactors

Conclusions The accumulated experience in the operation of NPP, including those with fast reactors, shows that during normal operation, with due regard for possible operational difficulties and accidents, they ensure a significantly lower level of risk for personnel and the surrounding population than is present in industrial regions and those prone to natural disasters. Therefore, the dangers connected with the widespread development of nuclear power arise not so much from a real risk as from a risk which in principle can be realized in very improbable accidents. From this point of view sodium-cooled fast reactors have certain advantages. The probability of the maximum accident of the rupture of pipelines in high-pressure reactors must be considerably higher. Here a single event, and one difficult to detect, such as the failure to detect a flaw in manufacture, is enough to initiate the very dangerous first step of an accident. The rupture of equipment in the primary loop of a fast reactor at practically atmospheric pressure is considerably less probable, and the integral assembly is quite safe. All the other chains of development of maximum accidents in a fast reactor require the simultaneous realization of several events in systems and devices which are constantly being monitored (SS and power supply systems, etc.). The above considerations together with such important properties of sodium as the large reserve before the boiling point and the practically inertialess transport of heat from the reactor to structural elements and heat-transfer devices under natural circulation conditions gives one confidence that the level of risk for future industrial NPP with fast reactors will be at least no higher than that for NPP with thermal reactors.Translated from Atomnaya Énergiya, Vol. 43, No. 6, pp. 464–472, December, 1977.     

First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study has attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A principal point of this study was to investigate reactivity effects with fuel relocation under three-dimensional core representation including control rods. The calculation has indicated that the fuel discharge from the core was disturbed by a significant flow resistance at the entrance nozzle in the current design. Additional static neutronic calculations have been performed to compare basic neutronic characteristics between different scale cores. The static neutronic calculations have clarified that the outward fuel compaction within the inner core increased the reactivity in the large-scale core unlike the small-scale core.     

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.     

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.     

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.     

Experimental studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progressing beyond the fuel subassembly scale and that the upward discharge option will be effective in eliminating the energetic potential.     

Inservice inspection of the reactor block of sodium-cooled fast breeder reactors

By using sodium as coolant special boundary conditions result for the inservice inspection (ISI) of fast breeder reactors. For that reason in general it is not successful applying the methods and equipment proved for the 151 of light water reactors.This report presents inspection methods and equipment developed for the ISI of the reactor block of sodium cooled fast breeder reactors. The survey takes into account the state of the art as well as some R&D-work at home and abroad. Entering into particulars the methods and equipment used for leak monitoring, the inspection of the reactor vessel wall, the inspection' of reactor internals above and below the sodium level, monitoring of structure home noise and the measurement of the gap between the reactor vessel and the guard vessel are described.     

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.     

Results of analyzing accidents with core meltdown in fast reactors with sodium as the coolant

Nizhegorod Polytechnical Institute. Translated from Atomnaya Énergiya, Vol. 73, No. 6, pp. 433-438, December, 1992.     

Acoustic resonances as a means of blockage detection in sodium-cooled fast reactors

This paper outlines the basis of a system giving early warning of the presence and build up of a local blockage in sodium-cooled fast reactor sub-assemblies. The method uses acoustic resonances and reports on the analysis of signals recorded during special experiments on the Dounreay Fast Reactor. These tests proved the detectability of the standing waves and showed theoretical predictions of their frequency and shift in the presence of a blockage to be in good agreement with measurements.     

A new cross-section generation model in the FAST code system and its application to gas- and sodium-cooled Generation IV fast reactors

The cross-section generation scheme employed in the 3D spatial kinetics PARCS code included in the FAST code system being used and developed at the Paul Scherrer Institute (PSI) is currently based on region-wise macroscopic cross-sections for reference conditions and their first-order derivatives with respect to the state variables. Since for some transients, feedback effects may likely not in this way be precisely approximated in their interrelations, this standard method was recently complemented for (hex, z) geometry by a more rigorous cross-section representation scheme. The main idea behind the new approach within the FAST code system is that of preparing sets of microscopic cross-sections for a studied design. The full library consisting of such isotopic tables is generated based on using the ECCO cell code of the code system ERANOS developed by the French Atomic Energy Commission (CEA) for a suitable range of temperatures and background cross-sections (σ₀) on a common grid. In the paper, this σ₀-model is described. In addition, within a detailed verification study needed due to its complexity, it is extensively compared to the standard method for both steady-state and transient conditions. Thereby use is made of current Generation IV fast-spectrum concepts. Reactivity and power evolution indicate overall good agreement of the two methods. Such a good consistence in various transient situations for systems characterized by different neutron spectra gives a large degree of confidence in the correct implementation and suitability of the microscopic cross-section methodology. Therefore, besides for the safety analysis of advanced fast-spectrum core concepts in general, it is foreseen, within the FAST code system, to use the σ₀-model for the assessment of uncertainty propagations in transient calculations in conjunction with the available ERANOS isotopic covariance matrices. The new development makes it also in principle possible to use the PARCS code as a reactor kinetics solver for severe accident analysis, allowing through the use of microscopic cross-sections to account for relocation of the core material during the accident.