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.     

The coupling algorithm between fuel pin and coolant channel in the european accident code EAC-2

In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release.This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules will also be briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC.     

Fuel Pin Behavior under Slow-Ramp-type Transient-Overpower Conditions in the CABRI-FAST Experiments

In the CABRI-FAST experimental program, four in-pile tests were performed with slow-power-ramptype transient-overpower conditions (called hereafter as “slow TOP”) to study transient fuel pin behavior under inadvertent control-rod-withdrawal-type events in liquid-metal-cooled fast breeder reactors. The slow TOP test with a preirradiated solid-pellet fuel pin under a power ramp rate of approximately 3%Po/s was realized as a comparatory test against an existing test in the CABRI-2 program where approximately 1%Po/s was adopted with the same type of fuel pin. In spite of the different power ramp rates, the evaluated fuel thermal conditions at the observed failure time are quite similar. Three slow TOP tests with the preirradiated annular fuel resulted in no pin failure showing a high failure threshold. Based on posttest examination data and a theoretical evaluation, it was concluded that intrapin free spaces, such as central hole, macroscopic cracks, and fuel-cladding gap, effectively mitigated the fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in the case of a very large amount of fuel melting. These CABRI-FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database under various fuel and transient conditions.     

Numerical calculations for two-phase flow analyses in pin bundles

Two-phase flow equation systems, in which equations are defined for each phase, are discussed for use in analyzing coolant behaviors in LMFBR pin bundles. These equation systems have not yet consolidated, because of theoretical and experimental difficulties and complexities.One of the problems is the equation systems' stability. This paper shows the stability for the low Reynolds number (O(1)) system, using a one dimensional linear equation system. Based on this fact, a two-phase flow equation system is numerically solved by using the subchannel method for 19- and 37-pin bundles. The calculational examples are LOF and TOP conditions with/without the blockage, and fission gas release.     

Thermal-hydraulic modeling for the analysis of LMFBR disrupted core behavior

This paper presents the analytical models of thermal-hydraulic phenomena of major interest in the analysis of LMFBR hypothetical core disruptive accidents. These models have been incorporated in LEVITATE [1], a code for the analysis of fuel and cladding dynamics under loss of Flow (LOF) conditions. LEVITATE has recently become part of the SAS4A [2] code system, replacing the older, less sophisticated SLUMPY [3] model.The influence of different thermal-hydraulic models on fuel motion is illustrated by a comparison between the results calculated by LEVITATE, the data from the L7 and L6 TREAT experiments [4] and the results calculated by SLUMPY. The results calculated by LEVITATE are in good agreement with the experimentally observed early fuel dispersal.     

一种节块多群精细注量率重构方法

在全堆粗网格节块解的基础上,经节块内精细注量率的重构来获得各燃料棒的功率是目前轻水堆堆芯分析计算中普遍采用的方法,然而,目前的精细注量率重构方法大都只适用于两群,无法进行多群重构。在韩国国立首尔大学提出的多群重构思想基础上,经进一步改进,研制出了多群重构方法,并开展了多个基准问题的检验。     

Simulation of the transient behaviour of LMEBR fuel pins under consideration of special burnup phenomena using the BREDA-II model

A new fuel pin model was developed to describe the influence of specific burnup phenomena on the behaviour of fuel pins under transient overpower conditions in a liquid metal fast breeder reactor (LMFBR). It has been used for transient fuel pin deformation analysis during hypothetical core disruptive accidents (HCDA) and for the purpose of interpreting fuel pin failure tests. The fuel pin model, designated as BREDA-II, is based on the equations of the quasi-static theory of thermal elasticity. The fuel is regarded as elastic and the cladding as elasto-plastic material. The equations for the stress-strain analysis are based on the plane strain approximation. A multiregion fuel pin model allows to simulate long-time and transient burnup phenomena. The long-time effects taken into account are the steady state swelling of fuel, the change in fuel porosity and the production and partial release of fission gases. During a power excursion transient fuel swelling and pressure increase due to transient fission gas behaviour are included in the deformation analysis. Potential fuel pin failure is indicated by the application of various criteria of failure. In subsequent model calculations the behaviour of an irradiated LMFBR fuel pin during an overpower transient corresponding to a reactivity ramp of $5/sec is simulated and interpreted from the point of view of reactor safety.     

Some trends in fast-reactor characteristics on optimization subject to constraints on the functional safety

Conclusions One has determined the general trends in the composition parameters in the treatment of various cases. The basic tendency in improvement in the safety characteristics lies in increasing the size of the core and the fuel-pin pitch. If one incorporates the constraint on the external dimensions or volume, the increase in core radial dimensions as a rule occurs at the expense of reduction in thickness of the side screen. In some cases, the constraints on the safety functionals during LOF WS enable one to improve stability for accidents of TOP WS type and vice versa. There is a considerable temperature margin before the fuel melts in the nominal state, so one can anticipate favorable completion of accident situations involving power increase not controlled by the protection system. However, the large temperature margin in the nominal state before the coolant boils and the sheaths are broken does not guarantee accident-free operation in LOF WS mode. Therefore, when one designs a safe fast reactor of the new generation, it is necessary to solve the optimizations in the presence of constraints on the safety functionals in accidents and it is desirable to have a large temperature margin before the fuel melts in the nominal state. This research was supported by a grant for research in the area of power engineering and electrical engineering. Moscow Physics Research Institute. Translated from Atomnaya Energiya, Vol. 80, No. 6, pp. 429–437, June, 1996.     

The quenching of irradiated fuel pins

This paper presents the results of falling film rewetting experiments on two types of irradiated fuel pin and on a complementary range of tubes and heaters. Preliminary measurements of rewetting rate in bottom flooding are presented and a brief comparison is made between falling film and bottom flooding rewetting, including the effects of surface finish. It is demonstrated that fuel pin rewetting behaviour is broadly consistent with that for tubes and heaters and that there are no large systematic effects of irradiation. Surface finish is shown to have an important influence on the falling film rewetting rate and could have influenced the fuel pin behaviour. The preliminary findings for bottom flooding are that there are smaller differences between various heaters than for falling film, and surface finish does not significantly change rewetting behaviour. It is shown that subcooling local to the quench front is an important parameter in rewetting, and can be used as a basis for correlating data. This is in agreement with recent ideas.     

Homogenized model for fluid-structure interaction of the pressurized water reactor core internals during blowdown

A method for the numerical simulation of the pressurized water reactor core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. In order to investigate the global core motion during the blowdown accident, the core model describes the coupled fluid-rod motion with Homogenization methods. The heterogeneous fluid-rod mixture thus is treated as a special continuum with anisotropic material properties. Furthermore, the core model considers elastical rod forces against bending and axial straining and the direct interaction of neighbouring fuel elements, which is a highly nonlinear process due to the finite gaps. Because this effect is very important, two simulation models have been developed and are compared. All these models have been implemented into the blowdown code FLUX-4. With the new code version FLUX-5 the PWR-blowdown is parametrically investigated.     

Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

In the Shutdown Heat Removal Testing (SHRT) Program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elements, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date.     

Multi-pin modelling of PWR fuel pin ballooning during post-LOCA reflood

Experiments investigating post-LOCA reflood generally indicate little coherence to fuel pin ballooning, and an absence of significant blockages and un-coolable regions of the core. Computational modelling is unable to predict this; the usual ‘representative pin’ models neither take into account the heterogeneity of the core, nor incorporate the dynamic coupling between the changing core geometry, and the flow paths taken by the coolant. In this paper we present a composite model, able to treat distinct pins mechanistically, and able to incorporate their (distinct) swelling behaviour into the thermal–hydraulic model of the reflood, allowing the cooling of a pin to be directly affected by the deformation of itself and its neighbours.     

Pellet-cladding interactions and length changes in SGHWR fuel pins

Fuel pins in the Steam Generating Heavy Water Reactor increase in length during irradiation, due largely to a mechanical interaction between the cladding and the fuel pellets. The length change produced by this process depends critically on the deformation behaviour of the fuel pellet under the interaction stresses. This paper presents a model of the pellet deformation which has been incorporated into a computer programme, SLIM, capable of predicting extensions from this and other causes in SGHWR fuel pins subjected to any power history. SLIM is shown to give good agreement with measured length changes for all variants of fuel rating, burn-up and pin type tested, and predicts that extension of the narrower pin designs now in use should be reduced by a factor of 0.7 compared with similarly rated standard pins.     

U_3Si_2-AI 弥散型燃料元件

U_3Si_2-Al 弥散型板状燃料元件是新发展的高铀密度的研究堆燃料.辐照和全堆芯试用表明:元件外形尺寸稳定;无裂变产物泄漏;直到燃耗达98%时,U_3Si_2颗粒肿胀都是裂变密度的线性函数;U_3Si_2与铝基体和铝包壳在制造和辐照中都是相容的。芯体铀密度最高达4.8g(U)/cm~3的 U_3Si_2-Al 燃料都是适合研究堆使用的低浓铀燃料.     

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the “rim effect.” High burn-up phenomena in WWER-440 UO₂ fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity.In this paper, the radial burn-up and fissile products distributions of WWER-440 UO₂ fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.     

On the failure of fast reactor fuel pins

The prediction of the timing and position of fuel pin failures is an important task in the modelling of fast reactor fuel behaviour. The range of processes that can provoke failure of fast reactor fuel pins in normal operating conditions and during hypothetical accidents is reviewed. Some of the mechanisms of failure are examined in more detail and the effect of hot spots and local stress concentrations is discussed. A review of failure criteria used in fast reactor fuel pin codes is given elsewhere, but the difficulties in applying various types of criteria are examined. Some discussion is also given on probabilistic approaches. Recommendations are given for a future approach to the problem of failure prediction, resolving the dilemma between inadequate empirical criteria and over-complex physically based approaches.     

Development of a 3-D FBR fuel bundle-to-duct interaction analysis code marse

A computer code marse has been developed to analyze structural behaviors of fast breeder reactor (FBR) fuel subassemblies under irradiation conditions, especially of wire spaced fuel pin bundles.The first concept specifically considered is to model on pin oval deformation and dispersion phenomena which were found to occur in highly irradiated fuel pin bundles. Every fuel pin is modeled by three-dimensional finite element method beams with trusses at contact points. Almost all phenomena which would occur under irradiation induced bundle-to-duct interaction (BDI) conditions are included in the marse code; these include bending, expansion, oval deformation and dispersion.The second point is to reduce computing time because the BDI analysis requires much computing time if a conventional solving scheme is applied. This problem is resolved by using only a small stiffness matrix of each pin successively to treat interactions with other pins or ducts as outer forces.The marse code has been applied to BDI analyses and has been confirmed to be applicable to any types of FBR fuel pin bundles with high accuracy and a small computing time.     

Simulation of the thermomechanical effects originated on a fuel pin with a cracked pellet by different power ramp velocities, using a two dimensional finite element model

Thermomechanical effects originated on a fuel pin with a cracked pellet by different power ramp velocities has been studied. To such purpose, the response of a simplified two dimensional finite element model including viscoelastic properties on the pellet and elastoplastic properties on the clad was investigated, under the influence of various ramps with durations ranging from 0 s (instantaneous ramp) to 200 s (slow ramp).The importance of both, the viscoelastic relaxation effects and their interaction with the plastic behaviour of the clad was put in evidence. It has been observed, that the mechanical response of this fuel pin model strongly depends on the ramp's duration.