Werner Maschek

The development of simmer-III, an advanced computer program for lmfr safety analysis, and its application to sodium experiments

Abstract SIMMER-III is a general two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics code coupled with a space-time and energy-dependent neutron transport kinetics model. The philosophy behind the SIMMER-III development was to generate a versatile and flexible tool, applicable for the safety analysis of various reactor types with different neutron spectra and coolants including the new accelerator-driven systems for waste transmutation. Currently, a three-dimensional version is also available, coined SIMMER-IV. The main backbone for analyses, however, is still SIMMER-III. SIMMER-III has proven especially well suited for fast spectrum systems such as the liquid-metal-cooled fast reactor where it is one of the key codes for safety analysis, including its application within licensing procedures. To serve especially the last purpose, the code must be made sufficiently robust and reliable and be tested and validated extensively. A comprehensive and systematic assessment program of the code has been conducted. This paper gives the major achievements of this assessment program. The SIMMER-III code handles by default liquid-metal-cooled fast reactor core materials - fuel, steel, coolant, control rod, and fission gas, in solid, liquid, and vapor states. The total of 27 density and 16 energy components are modeled in three velocity fields and one structure field in order that important fluid motions in a degraded core are simulated adequately. The spatial differencing method is based on Eulerian staggered mesh with a higher-order differencing scheme to mitigate numerical diffusion. An improved analytic equation-of-state model provides good accuracy especially at high temperature and pressure. Multiple flow-regime treatment is available over the entire void fraction range. An interfacial area convection model improves the flexibility of the code by tracing transport and history of interfaces and thereby better represents physical phenomena. A generalized and flexible code framework, along with improved numerical stability and accuracy, allows us to apply it to a variety of simple and complex multiphase flow problems. The code assessment program is an ongoing effort. Two major milestones have been achieved in the past by completing two assessment campaigns, Phase 1 and Phase 2: Phase 1 for fundamental code assessment of individual models and Phase 2 for integral code assessment for key phenomena relevant to liquid-metal-cooled fast reactor safety. Through this systematic code assessment program, comprehensive validation of the physical models has been conducted step-by-step. The assessment program has demonstrated that SIMMER-III is a state-of-the-art code with advanced models sufficiently flexible for simulating transient multiphase phenomena occurring during core disruptive accidents. This paper concentrates on the specifics of the code, mainly reflected in its application to sodium experiments related to the safety of liquid-metal-cooled fast reactors.

Investigations of Sloshing Fluid Motions in Pools Related to Recriticalities in Liquid-Metal Fast Breeder Reactor Core Meltdown Accidents

This paper reports that analyses of unprotected loss-of-flow accidents for medium-size cores of current liquid-metal fast breeder reactors have shown that the accident proceeds into a transition phase where further meltdown is accompanied by recriticalities and secondary excursions. Assuming very pessimistic conditions concerning fuel discharge and blockage formation, a neutronically active whole-core pool of molten m material can form. Neutronic or thermohydraulic disturbances may initiate a special motion pattern in these pools, called centralized sloshing, which can lead to energetic power excursions. If such a whole-core pool is formed, its energetic potential must be adequately assessed. This requires sufficiently correct theoretical tools (codes) and proper consideration of the fluid-dynamic and thermo-hydraulic conditions for these pools. A series of experiments has been performed that serves as a benchmark for the SIMMER-II and the AFDM codes in assessing their adequacy in modeling such sloshing motions. Additional phenomenologically oriented experiments provide deeper insight into general motion patterns of sloshing fluids while taking special notice of asymmetries and obstacles that exist in such pools.

Thermophysical Properties of Lead-Bismuth Eutectic Alloy in Reactor Safety Analyses

A consistent set of thermophysical properties of a lead-bismuth eutectic (LBE) alloy was developed for use in safety analyses of lead-alloy-cooled fast reactor systems. The vapor and liquid thermodynamic states of LBE were modeled up to and above the critical point based on a van-der-Waals type of equation. We assumed that LBE vapor is composed of monatomic lead and bismuth and diatomic, bismuth components, and that liquid LBE is a non-ideal mixture of lead and bismuth. Recommended equations were also presented for the transport properties and surface tension of liquid LBE.

Analysis of Severe Accident Scenarios and Proposals for Safety Improvements for ADS Transmuters with Dedicated Fuel

Abstract So-called dedicated fuels will be utilized to obtain maximum transmutation and incineration rates of minor actinides (MAs) in accelerator-driven systems (ADSs). These fuels are characterized by a high-MA content and the lack of the classical fertile materials such as 238U or 232Th. Dedicated fuels still have to be developed; however, programs are under way for their fabrication, irradiation, and testing. In Europe, mainly the oxide route is investigated and developed. A dedicated core will contain multiple “critical” fuel masses, resulting in a certain recriticality potential under core degradation conditions. The use of dedicated fuels may also lead to strong deterioration of the safety parameters of the reactor core, such as, e.g., the void worth, Doppler or the kinetics quantities, neutron generation time, and βeff. Critical reactors with this kind of fuel might encounter safety problems, especially under severe accident conditions. For ADSs, it is assumed that because of the subcriticality of the system, the poor safety features of such fuels could be coped with. Analyses reveal some safety problems for ADSs with dedicated fuels. Additional inherent and passive safety measures are proposed to achieve the required safety level. A safety strategy along the lines of a defense approach is presented where these measures can be integrated. The ultimate goal of these measures is to eliminate any mechanistic severe accident scenario and the potential for energetics.

Theoretical Modeling of Radial Standing Wave Reactor

This paper is a theoretical study of a radial standing wave, which can be applied in the so-called traveling wave reactor (TWR). Two-dimensional cylindrical core geometry is considered and the fuel is assumed to drift radially, which corresponds to a radial fuel shuffling scheme in practice. A one-group diffusion equation coupled with burn-up equations is set up, where the burn-up solution is obtained numerically. The uranium-plutonium (U-Pu) conversion cycle with pure 238U as fresh fuel is considered under conditions of a typical sodium cooled fast reactor with metallic uranium fuel loaded. The asymptotic problem is solved by a time-stepping iteration scheme and the radial standing wave solution is obtained together with certain eigenvalue keff.The neutron flux, the neutron fluence and the net neutron generation cross section are presented and discussed for the inward fuel drifting motion.

Safety aspects of oxide fuels for transmutation and utilization in accelerator driven systems

Abstract General safety aspects of fuels under development for accelerator driven systems (ADS) are reviewed and discussed. These fuels should allow a maximization of transmutation and incineration rates, which excludes fertile UO 2 as a component or matrix. The accumulated knowledge on data, phenomena and scenarios of fast reactors with (U,Pu)O 2 oxide fuels and sodium cooling serves as background for this review. For future ADS both the reactor system itself, the fuel and the coolant are innovative compared to traditional critical fast reactors. For the fuel, these boundary conditions lead to many open questions, starting from basic thermal physical, thermal mechanical and irradiation data to the behavior under transient conditions. The choice of fuel naturally has a significant impact on whole core behavior and safety too, including the influence on related neutronics parameters, on failure propagation and disruption behavior under accident conditions. Key safety issues are discussed and a first assessment of phenomena and scenarios is given. Areas of research and technology in which further work is required to resolve important safety issues are highlighted.

SAFETY ANALYSES OF THE LEAD-BISMUTH EUTECTIC COOLED ACCELERATOR DRIVEN SYSTEM XT-ADS

Safety analyses for the XT-ADS were performed with the reactor safety code SIMMER-III. Besides a brief description of the numerical model, three typical transients are presented in this paper, namely, the unprotected loss of flow (ULOF), unprotected transient over-current (UTOC), and the unprotected coolant flow blockage accident (UBA). Because of the important phenomenon of mass flow rate undershooting in the ULOF case, an integral equation model was set up for a further theoretical study of ULOF. The model confirms the numerical simulation results for various cases and gives a deeper understanding of this phenomenon. The faster the pump shut down, the larger is the undershooting of the mass flow rate. On the other hand a larger coolant cold leg area leads to a weaker undershooting. The stability analysis shows that the natural convection state is in the region of the damped oscillation for the current XT-ADS design.

Simulation of Molten Metal Penetration and Freezing Behavior in a Seven-Pin Bundle Experiment

Analysis of a hypothetical core disruptive accident is important in the safe design of the future generation of reactors such as liquid-metal-cooled reactors. This study determines the fundamental mechanisms underlying the penetration and freezing behavior of molten metal flowing through a sevenpin channel. We conducted a series of simulant experiments that focused on the fuel-pin-bundle geometry under various thermal conditions of the molten metal and pins, and produced data for the fundamental verification of the safety analysis code SIMMER-III. The liquid penetration length and the solidified frozen metal in the flow channel were investigated in the experiments. Visual information was obtained using a digital video camera. A numerical simulation using the SIMMER-III code was carried out to validate its model and method in terms of the bulk freezing behavior of molten metal. The simulation showed that the code reasonably represents the melt penetration and freezing behavior observed in the experiments.

Optimization of Safety Parameters and Accident Mitigation Measures for Innovative Fast Reactor Concepts

Traditionally the analysis of the evolution of severe core disruptive accidents (CDA) is broken down into different phases. This is mainly done for a better focussing on the key phenomena of the accident phase and also allows the application of specific codes for the analysis. In the current paper we mainly deal with the initiating phase and the transition phase of an accident as the ULOF (unprotected loss of flow). The key phenomenon of the initiating phase is the start of boiling and the development of voiding; key phenomena of the transition phase are the progression of core melting and the occurence of recriticalities by fuel compaction. The first level of optimizing safety is oriented to the initiating phase by reducing the positive void worth in order to avoid that a ULOF accident would enter a severe development. If accident prevention is not achieved the transition phase, characterized by a progressive core degradation leading to the occurrence of recriticalities, can be mitigated by dedicated features that enhance and guarantee a sufficient and timely fuel discharge – e.g. by a controlled material relocation (CMR) - and influence and ‘brake’; the recriticality path. In the paper both phases are analyzed. The results presented are in agreement with the activities performed within the European Collaborative CP-ESFR project.

Numerical Studies of Axial Fuel Shuffling

The concept of traveling wave reactor (TWR) applies the mechanism of self sustainable and propagation nuclear fission traveling waves in fertile media of 238U and 232Th to achieve very high fuel utilization. However, the long wave length of such fission traveling wave puts a limit on the applicability of the TWR concept. The axial fuel shuffling strategy is proposed based on the mechanism of asymptotic nuclear fission traveling wave, and is applied to a sodium-cooled fast reactor (SFR) loading metallic 238U fuel. The multi-group deterministic neutronic code ERANOS with JEFF3.1 data library is used as a basic tool to perform the neutronics and burn-up calculations. The calculations are firstly performed in a 1-D case for parametric understanding, and further extended to a 2-D R-Z case. The shuffling calculations for the 1-D and 2-D SFR model described in this paper brought about some interesting results. The results indicate that keff parabolically varies with the shuffling period, while the burn-up increases linearly. The highest burn-up achieved in 2-D case is 46at%. The power shape distortion in 2-D case is observed, and the power peaking factor is much higher than that in 1-D case, but it decreases with the shuffling period increasing.