Results 1 - 10 of 1422
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[en] The Nuclear Safety Research Center (NSRC) of Japan Atomic Energy Agency - JAEA is conducting computations of criticality characteristics of the fuel debris of the Unit 1-3 reactors of Fukushima Daiichi Nuclear Power Station, and building a database of their results. The database will be useful to evaluate criticality control parameters when samples of fuel debris are taken and analyzed. Further computations will be performed to evaluate uncertainty of neutron multiplication factors due to random distribution of fuel debris composition using the new Monte Carlo solver, Solomon. Mathematical models and voxel geometry have been implemented in the solver to handle the randomness. The computation of new materials such as fuel debris mixed with steel or concrete modeled by the new techniques will be validated by critical experiments using the new Static Experiment Critical Facility (STACY). A risk analysis method supported by computer codes will be provided for estimation of criticality risk of the fuel debris retrieval, which will be useful for study and selection of retrieval work designs to be proposed in a few years.
[en] The fuel debris produced by the accident of the Fukushima Daiichi Nuclear Power Station is probably in a state of mixture of spent fuels with different burnups each other. In such a case, the mixing ratio of spent fuels in fuel debris would affect its criticality. This report shows computation results of criticality characteristics of fuel-debris compositions prepared by mixing nuclide compositions of spent fuels in various patterns based on a fuel loading pattern. The results indicate that fuel debris is potentially subcritical when 1-cycle fuels, whose average burnup is several GWd/t, are included homogeneously in fuel debris because remaining 155Gd and 157Gd in 1-cycle fuels works to reduce neutron multiplication. The results also indicate that 155,157Gd/235U ratio well characterize criticality of fuel debris. (authors)
[en] The paper presents a method to evaluate the potential re-criticality issues in severe accident configurations. This method is developed by Tractebel ENGIE, using respectively MCNP and a dataset of MELCOR calculations. The work has focused on the in-vessel phase and more particularly on a TMI-2-like configuration. A model based on a corium sphere approximation associated to the detailed WIMS10 fuel characterization has allowed us to highlight some degraded configurations presenting re-criticality risks associated to the intact fuel rods that remain in the core. The resulting distributions of the multiplication factor (keff) in function of several parameters (core degradation fraction, boron concentration in water, core burnup, moderator and fuel temperature,...) has allowed us to identify the more sensitive parameters regarding the re-criticality risk. The core degradation fraction is the major parameter, even for a high level of degradation, which highlights the influence of the intact peripheral fuel rods. The boron concentration in the water injected after the accident also plays an important role: even with a relatively high core degradation fraction (about 60%) and a consequent boron concentration (1500 ppm) the re-criticality risk cannot be excluded. These conclusions being well dependent on the assumptions and conservatisms of the study Finally a 'surrogate model' - regression model based on artificial neural network - is presented for the prediction of the multiplication factor for TMI-2-like configurations.
[en] Highlights: • Review of the requirements and recommendations for BEPU methodology. • Summary of the advantages and limitations of the current deterministic bounding method for non-LOCA transient analysis. • Description of a pragmatic, graded approach for application of the BEPU methodology to non-LOCA transient analysis. • Proposal for a demonstration case. - Abstract: Since 1990’s, the use of best estimate plus uncertainty (BEPU) methodology is becoming a common practice for large-break Loss-Of-Coolant Accident (LOCA) analysis. However, the development and application of BEPU methodology requires a higher-level requirement on the verification and validation, and uncertainty quantification (VVUQ) of the used calculational method and computer codes. This may result in a high-cost for BEPU methodology development, and hence prevent the industry to take full benefit from the BEPU applications. This paper proposes a pragmatic, graded approach for application of the BEPU methodology to non-LOCA transient analyses.
[en] Difficulty and importance of determining possibility of occurrence of criticality in the accidents with destruction and even melting of nuclear fuel at NPP differ significantly whether this determination occurs immediately or after a long period of time since accident. Immediately after the accident, when situation is not yet stabilized, there is no sufficient amount of data on the state of fuel and moderator and this requires additional efforts to assembly conservative computational models. Current report presents an analysis of the experience of evaluations of criticality that were performed at the Kurchatov Institute immediately after accidents at the 4. block of the Chernobyl NPP in 1986, in the fuel assembly washing tank at the Paks NPP in 2003 and at the Fukushima NPP in 2011. Based on the experience of the first two accidents, there was developed the specialized complex SAPFIR-2006, designed for evaluation the criticality in severe beyond design basis accidents. The operative criticality evaluations made by this complex during the course of the accident did not significantly differ from the conclusions of subsequent analyzes performed quite a long time after the accident. However, earlier criticality evaluations may influence the management of the accident progression and the determination of optimal strategy of emergency response. After the accident at the Paks NPP, results of computational evaluation of the criticality were taken into account when preserving the subcritical state of tank with destroyed fuel and when removing fragments of fuel assemblies from the tank. In emergency response during Chernobyl accident, such results were not taken into account. In the accident at the Fukushima NPP, there was no opportunity to inform emergency workers with results of computational evaluation of the criticality. (authors)
[en] This series of lectures clarifies the relationship between structural strength in design and actual structural strength, and explains the importance and effectiveness of evaluation of actual structural strength. This 4th session takes up the challenges to new structural problems, and explains the structural problems, failure control technology, and resilience engineering in response to out-of-design criteria events in the nuclear field. (A.O.)
[en] Highlights: • A coupled thermalhydraulic-neutronic flow stability criterion is established for SFRs. • Pressure drop and power evolutions are considered in the criterion. • A linear stability analysis of 2 equations expressing flow and power evolution is done. • Reactivity feedback of enthalpy change on SA power is modelled. • The capability of the criterion to predict ULOF flow behavior is shown. - Abstract: A coupled thermalhydraulic-neutronic extended criterion developed in order to assess the conditions of static instability of the sodium flow in a SFR core during the beginning of an unprotected loss of flow (ULOF) is presented in the first part of this paper. This extended criterion takes into account the evolution of the pressure drop in a sub-assembly (SA) as well as the evolution of its power when the reactor is affected by a flow rate (or power) perturbation by starting from a steady-state. The considered steady states are typical of quasi-steady states reached by a SFR during a ULOF. The temporal evolution of the flow and the temporal evolution of the power can be represented by a system of two differential equations whose linear stability analysis has enabled to define the aforementioned extended criterion. Then, in the second part of the paper, a verification of this extended criterion is performed by simulating the behavior of a GEN IV SFR of 1500 MWth. Four various steady states of the reactor have been investigated with a simulation tool devoted to unprotected loss of flow simulations (MACARENa). First simulations have enabled to calculate the various physical parameters needed to check the extended criterion. Perturbations have been applied to these four states. The extended criterion, proposed in this paper, has predicted flow stabilization for 3 states whereas it has predicted a flow excursion (sudden flow rate decrease leading to SA dry-out) for the last state. In this latter case, the classical Ledinegg criterion, which relies only on pressure drop evolution at constant power, has predicted a stable configuration. Then, the extended criterion predictions have been verified with MACARENa simulations, by slightly and temporary reducing the driving pressure head of the flow rate in the studied SA from each of the four initial states. For stable states, when setting again the pressure head of the flow to the value of the initial state, the initial flow rate is recovered. For the last initial state (predicted as unstable by the extended criterion), after a convergence period carried-out by imposing the SA steady flow rate, the solving of the transient momentum equation have shown that a flow excursion occurs in the MACARENa simulation as expected from the extended criterion. Consequently, the extended criterion proposed in this paper has been verified and could enable to define stable and unstable domain of couple flow rate/power in order to design future cores.
[en] The stabilization of molten core material in the lower head in case of a severe accident by external cooling of the reactor vessel is regarded as an effective severe accident management measure. In the experiments LIVE-L10 and -L11 the late phase melt pool behaviour of the corium is investigated under different cooling conditions – the former under sub-cooled convection, the latter under nucleate boiling conditions. In this work the experiments are calculated with the severe accident analysis code AC2 – ATHLET-CD 3.1A. Objective of the simulations conducted is the analysis and assessment of the code's capability to simulate the most relevant phenomena that occur during the tests. The simulations are performed with two different lower head modules implemented in ATHLET-CD, AIDA (Analysis of the Interaction between Core Debris and the reactor pressure vessel during severe Accidents) and LHEAD (extended Lower Head module). The simulation results, analysed in comparison with the experimental results, show the capability of both modules to reproduce the respective experiments.
[en] The nuclear power plant disaster requires zoning for evacuation and recovery. It can be classified into three types. 1. Zoning for Emergency Evacuation: The current evacuation plan institutionalized after the Fukushima-1, cannot realize its agenda at the next disaster. 2. Zoning for Temporary Sheltering: If the locals require for the strict decontamination, it as to face difficulties in terms of construction work and storage place. 3. Zoning for Protracted Refugee: the protracted term is various including the negative repatriation area. There is a difference between the consciousness of national government and the reality of the damaged area and people. (author)
[en] A new Monte Carlo solver Solomon has been developed for the application to fuel-debris systems. It is designed not only for usual criticality safety analysis but also for criticality calculations of damaged reactor core including fuel debris. This paper describes the current status of Solomon and demonstrates the applications of the randomized Weierstrass function (RWF) model and the RWF model superposed with voxel geometry. (authors)