Relocation Of Debris Research Articles

MAAP code analysis focusing on the fuel debris condition in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 2

Based on the updated knowledge from plant-internal investigations, experiments and computer-model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 2 was analyzed using the MAAP code. In Unit 2, it is considered that the enthalpy of the core materials was relatively low when they relocated to the lower plenum of the pressure vessel, then, cooled by the coolant and solidified there. Although the MAAP code tended to underestimate the degree of core-material oxidation during the relocation period, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. With this study, basic validity of the former prediction of the Unit 2 accident progression behavior was confirmed and detailed boundary condition for the later phase was provided. This boundary condition should be utilized for future studies addressing debris reheating process leading to lower head failure and debris relocation toward the pedestal.

Estimation of debris relocation and structure interaction in the pedestal of Fukushima Daiichi Nuclear Power Plant Unit-3 with Moving Particle Semi-implicit (MPS) method

To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant (NPP) Unit-3 (from hereinafter, Unit-3), the meshless, Lagrangian Moving Particle Semi-implicit (MPS) method has been developed for the evaluation of debris-structure interactions in the pedestal region of Unit-3. The developed MPS method incorporated approaches to enhance numerical accuracy and stability, such as second-order corrective matrix and particle shifting technique, and the calculation cost reduction and efficiency improvement strategies to enable plant-scale simulations. A 1/10 scale numerical model of the pedestal region and structures within of Unit-3 and sensitivity analysis conditions were established for simulation cases. The simulation results showed that the convective vapor cooling from the debris surface in the pedestal region and the debris relocation amount/the time intervals between the relocations play important roles in the structure damage due to thermal attacks from the debris.

Analysis of Fukushima-Daiichi Nuclear Power Plant Unit 3 pressure data and obtained insights on accident progression behavior

The D/W (Drywell) and S/C (Suppression Chamber) pressure data of Fukushima-Daiichi Nuclear Power Plant Unit 3 was analyzed in depth. This analysis provided valuable information related to the accident progression behavior on one hand, and gave a hint for understanding of the debris-to-coolant heat transfer when fuel debris relocated to the pedestal on the other hand. In this unit, the D/W and S/C pressure increased and decreased cyclically with a relationship, which seems to have been dependent on the composition of vapor and non-condensable gases in the S/C cover gas region. Based on this characteristic, the vapor pressure in the S/C cover gas region was evaluated for two pressure decrease cycles during and after the expected debris relocation to the pedestal respectively. This evaluation allowed an understanding that the S/C vapor pressure increased due to the heat transfer from the debris relocated to the pedestal. With this understanding, several possible S/C vapor pressure histories were assumed. Based on this parametric study it was suggested that some part of the S/C water was heated up within about 75 min when the main part of the fuel debris would have been relocating to the pedestal.

Estimation of the fuel debris thermal energy at the time of the major core slumping of Fukushima Daiichi Nuclear Power Plant Unit-3 with MELCOR-2.2

To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant Unit-3, sensitivity analyses have been carried out with MELCOR-2.2 for two scenarios, with/without direct leakage from the Reactor Pressure Vessel (RPV) to the Drywell (D/W). The particular focus of the analyses is the estimated debris thermal energy up to and at the time of the major core slumping event, which may be the key to determine the following debris cooling and failure mechanism of the lower head of the RPV. In the analysis, the debris relocation velocity from the core region to the RPV lower head (VFALL), the number of Safety Relief Valves (SRVs) opening, and the amount of core slumping at the major RPV pressure peak were tuned so that the plant data of RPV and PCV pressure from ca. 9:00 to 12:00 March 13th 2011could be reproduced in each scenario. Best-estimate case conditions are summarized for the with/without direct leakage from RPV to D/W scenarios. As a result, as a consideration common to both scenarios, MELCOR analysis tends to estimate significant core oxidation before Automatic Depressurization System (ADS) actuation, and concomitantly estimates little core oxidation from the time after ADS to the time of the major RPV pressure peak. In addition, although a remarkable difference was found in the amount of hydrogen generated in RPV during the major core slumping between the best-estimate with/without leakage cases, limited difference can be observed for the core oxidation heat generation during the major core slumping. This is because the early Zircaloy oxidation prior to ADS actuation in the current MELCOR modelling resulted that the hydrogen generation source during the major core slumping at 12:00 March 13th was primarily from oxidation of stainless steel, which was not as exothermic as that of Zr oxidation.

Simulation of a Station Blackout Accident for the SMART Using the CINEMA Code

SMART is a small-sized integral type PWR containing major components within a single reactor pressure vessel. Advanced design features implemented into SMART have been proven or qualified through experience, testing, or analysis according to the applicable approved standards. After Fukushima accident, a rising attention is posed on the strategy to cope with a Station Blackout (SBO) accident, which is one of the representative severe accidents related to the nuclear power plants. The SBO is initiated by a loss of all offsite power with a concurrent failure of both emergency diesel generators. With no alternate current power source, most of the active safety systems that perform safety functions are not available. The purpose of SBO analysis in this paper is to show that the integrity of the containment can be maintained during a SBO accident in the SMART (System-integrated Modular Advanced ReacTor). Therefore, the accident sequence during a SBO accident was simulated using the CINEMA-SMART (Code for INtegrated severe accidEnt Management and Analysis-SMART) code to evaluate the transient scenario inside the reactor vessel after an initiating event, core heating and melting by core uncovery, relocation of debris, reactor vessel failure, discharge of molten core, and pressurization of the containment. It is shown that the integrity of the containment can be maintained during a SBO accident in the SMART reactor. It has to be mentioned that the assumptions used in this analysis are extremely conservative that the passive safety systems of PSIS and PRHRS were not credited. In addition, as ANS73 decay heat with 1.2 multiplier was used in this analysis, actual progression of the accident would be much slow and amount of hydrogen generation will be much less.

Development of relocation model for debris particles in core disruptive accident analysis of sodium-cooled fast reactor

During a core disruptive accident in a sodium-cooled fast reactor, the molten fuel or steel is solidified into debris particles that form a debris bed in the lower plenum. When the debris bed starts boiling, the flow of the coolant and vapor together relocates and flattens the debris particles. This process is called debris relocation. The thickness of the debris bed and its porosity have significant influence on the cooling ability of the fuel debris in the lower plenum. Therefore, it is necessary to evaluate the transient changes in the shape and thickness of the debris bed in the relocation behavior for accident analysis. To simulate the relocation behavior, a debris-relocation model based on the COMMEN code was developed in this study. The debris-relocation model is based on the shear strength mechanism, which is widely applied in soil mechanics. Shear strength is a function of the density, position, and shape characteristics of particles. The debris bed is relocated only when the shear stress of the global debris particles is greater than the shear strength. By incorporating the relocation model into the COMMEN code, the relocation transient processes of different particle beds in a nitrogen-injection experiment were simulated, and the results were compared with experimental results. The analysis indicates that the proposed debris-relocation model can effectively predict the occurrence of debris-bed relocation and reasonably simulate its transient process, although future improvement is necessary.

Role of fatigue in damage development of refractories under thermal shock loads of different intensity

Refractories insulation of industrial furnaces often fail under repetitive thermal shock. Degradation of silica refractories under thermal shock loads of different intensity was studied. The load variation was achieved by utilisation of geometrically similar samples of different dimensions. Finite element method modelling predicted loads developing during the test. Resulting damage was determined by the ultrasound velocity and crack patterns. Tests involving up to 150 cycles demonstrated the role of fatigue in enabling sub-critical crack formation and countering the crack arrest. Repetitive cycles reduce crack wake friction and intensify loading due to crack debris re-location. Damage saturation, sigmoidal and near-exponential damage growth was typical for low, intermediate and high loads, respectively. Similar trends of damage accumulation were observed in mechanical displacement controlled cyclic fatigue tests performed in wedge splitting set-up. Strain and strain energy based criteria of thermal shock intensity seem to have complimentary value in predicting the crack formation and growth. Thermal shock damage after the first cycle seems to be an effective parameter to predict overall resistance to the degradation in the sample. Load reduction due to previous crack formation related to the fatigue potential for subsequent crack development can explain the crack size variation typically observed in refractories after multiple thermal shocks. For thermal shock tests, the variation of sample size, instead of the temperature interval, is a suitable alternative for refractories with strongly temperature dependant material properties.

Complex weathering in drylands: Implications of ‘stress’ history for rock debris breakdown and sediment release

Weathering studies have often sought to explain features in terms of a prevailing set of environmental conditions. However, it is clear that in most present-day hot desert regions, the surface rock debris has been exposed to a range of weathering environments and processes. These different weathering conditions can arise in two ways, either from the effects of long-term climate change acting on debris that remains relatively static within the landscape or through the spatial relocation of debris from high to low altitude. Consequently, each fragment of rock may contain a unique weathering-related legacy of damage and alteration — a legacy that may greatly influence its response to present-day weathering activity. Experiments are described in which blocks of limestone, sandstone, granite and basalt are given ‘stress histories’ by subjecting them to varying numbers of heating and freezing cycles as a form of pre-treatment. These imposed stress histories act as proxies for a weathering history. Some blocks were used in a laboratory salt weathering simulation study while others underwent a 2 year field exposure trial at high, mid and low altitude sites in Death Valley, California. Weight loss and ultrasonic pulse velocity measurements suggest that blocks with stress histories deteriorate more rapidly than unstressed samples of the same rock type exposed to the same environmental conditions. Laboratory data also indicate that the result of imposing a known ‘weathering history’ on samples by pre-stressing them is an increase in the amount of fine sediment released during salt weathering over a given period of time in comparison to unstressed samples.

Comparison of Severe Accident Results among SCDAP/RELAP5, MAAP, and MELCOR Codes

This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP.

The modeling of core melting and in-vessel corium relocation in the APRIL code

This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validation are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

BWRSAR calculations of reactor vessel debris pours for peach bottom short-term station blackout

This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented.

A gas-cooled fast reactor core model for the analysis of severe accidents, with applicability to LMFBRS

A new fast-reactor core model has been developed which allows simulation of the behavior of a multizone core during accidental transients (loss of coolant or external reactivity insertions) leading from steady-state operation to fuel-rod failure and then to gross core damage. The model for the rod failure and for the relocation of the fuel debris is based on an interpretation of the meltdown experiments performed in the TREAT reactor. In the presented core model, the thermal response of the fuel rods is determined using a two-dimensional heat-transfer code in r-z geometry. In addition to the standard kinetics feedback relations, the reactivity effects of the gradual removal of the rod cladding are included. Taking into account the addition and removal of the heat of fusion allows the amount of fuel that resolidifies in the lower core part to be estimated. The presence of rod spacer arrangements is included as additional barriers for the debris relocation. The thermal response of the subassembly walls is calculated taking into account the contact with the high-temperature fuel as well as radiative heat transfer.