Yapei Zhang

Comparison of Hydrogen Generation Rate between CORA-13 Test and MELCOR Simulation: Clad Solid-Phase Oxidation Models Using Self-Developed Code MYCOAC

Abstract To better understand the MELCOR oxidation and degradation models, past work compared the MELCOR model to a CORA experiment (CORA Test 13). These MELCOR analyses specifically focused on fuel bundle heatup and clad oxidation when compared to CORA test data. The comparison allowed the authors to investigate differences between hydrogen generation data and simulation results. Several potential reasons were considered for hydrogen generation rate differences, including MELCOR input power, heat transfer modeling, the clad solid-phase oxidation model, and the gaseous steam diffusion model. This work focuses on the possible uncertainty in the clad oxidation models used in MELCOR. First, the MELCOR nodalization approach for the CORA test was reviewed. Then, the temperature history and spatial variation were examined. One main focus was to consider other clad solid-phase oxidation models to compare the MELCOR models. This was accomplished by developing a separate model, MYCOAC, using MELCOR temperature predictions as input. Finally, the mass transfer resistance of steam diffusion to the clad surface was examined and found to be a small effect. While the Baker-Just solid-phase oxidation model showed better agreement with CORA data at low temperatures, the conclusion in this paper is that the oxidation models are not the major source of uncertainty in hydrogen generation rate differences. Future work will focus on heat transfer modeling of the CORA test.

Analysis of the Particulate Debris Bed Quenching During Top and Bottom Flood

The quenching characteristics of particulate debris bed during bottom and top flood is analyzed in this paper. The top flood model is formulated by dividing the quenching process into downward frontal period and upward frontal period, which are controlled by the counter-current flow limitation (CCFL) condition and effects of the incoming coolant subcooling and steam cooling in dry channels during quenching process. The bottom flood model is based on porous media theory under the assumption that the height of the two phase region is negligible and the particulate debris bed is divided into single phase liquid and single phase vapor region. The results calculated by these models were compared with the experimental data. The influences of porosity, initial debris temperature and other parameters on both the top and bottom quenching process were studied in this paper. During the top flood, the quenching velocity increased with the increase of the porosity and the decrease of the initial debris temperature. The porosity and initial debris temperature had a larger influence on quenching velocity compared with other parameters, such as initial coolant temperature and coolant flow rate. During the bottom flood, the quenching velocity also increased with the increase of the porosity and the decrease of the initial debris temperature. However, the coolant flow rate had a large influence on the quenching velocity unlike that during the top flood. Quenching from bottom may be superior to the quenching from top. The results can be expected to be useful to evaluate the quenching process of the particulate debris bed.Copyright

Loss-of-Flow-Accidents (LOFA) Study for 100 MW IPWR

The inherent system safety of the 100 MW integral pressurized water reactor (IPWR) can be improved by placing the core, the efficient once-through steam generators and the main coolant pumps in the reactor pressure vessel, and omitting some large pipes and valves in the primary coolant system which can prevent the occurrence of large break loss of coolant accident and reduce the possibility of core melt accident. The application of the passive safety systems simplifies the structures of IPWR and improves the economy of the reactor. In case of accidents, the primary coolant system establishes natural circulation to take the core decay heat away by passive safety systems using gravity and other natural driving forces, thereby enhancing the safety and reliability of the system IPWR. It’s of great significance to establish reasonable and correctable models, including the primary coolant system model, the second loop model and passive core cooling system model, to study thermal-hydraulic phenomena under steady state, transient state and accident conditions.Based on transient safety analysis program RELAP5/MOD3.4, 100 MW IPWR system was simulated. A series of models of reactor coolant system and passive safety systems were established. The main system models are composed of primary coolant system model, part of second loop model, passive safety injection system model and passive residual heat removal system model. The primary coolant system model includes core, lower plenum, downcomer, region of steam generators, upper plenum, riser, pressurizer, and surge line; the second loop model includes the main feed water line, the steam line, and steam generator tubes; passive safety injection system model includes core makeup tank, accumulator, automatic depressurization system, direct vessel injection line; and passive residual heat removal system model includes passive residual heat removal heat exchanger in containment refueling water storage tank. Based on the established models, the steady state was debugged with the RELAP5 input card.Steady state calculation was performed and the results agree well with designed values, which verifies the validity of the model and the input card. Using the steady state results as initial conditions, transient calculation was performed. Typical accidents (loss of main water accident) were calculated, which were relieved by auxiliary feedwater system (AFWS) and passive residual heat removal system (PRHR SYSTEM). The results, obtained from AFWS and PRHR SYSTEM, were contrasted and process of accident and thermal-hydraulic phenomena were analyzed according to transient calculation results. The transient calculation results showed that the integral PWR system and the passive safety system model can provide a reference for IPWR transient safety analysis.Copyright

Comparison of CORA and MELCOR Core Degradation Simulation and MELCOR Oxidation Model

MELCOR is widely used and sufficiently trusted for severe accident analysis. However, the occurrence of Fukushima has increased the focus on severe accident codes and their use. A MELCOR core degradation calculation was conducted at the University of Wisconsin – Madison. The calculation results were checked by comparing with a past CORA experiment. MELCOR calculation results included the flow rate of argon and steam, the generation rate of hydrogen. Through this work, the performance of MELCOR COR package was reviewed in detail. This paper compares the hydrogen generation rates predicted by MELCOR to the CORA test data. While agreement is reasonable it could be improved. Additionally, the MELCOR zirconium oxidation model was analyzed.Copyright

Analysis of a Loss of Heat Removal Accident in a PWR Spent Fuel Pool

The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents a study on the consequences of loss of heat removal accident in the spent fuel pool of a typical pressurized water reactor using the Modular Accident Analysis Program (MAAP5) code. Analysis of uncompensated loss of water due to the loss of heat removal with initial pool water level of 12.2 m (designated as a reference case) has been performed. The analyses cover a broad spectrum of severe accident in the spent fuel pool. Those consequences such as overheating of uncovered fuel assemblies, oxidation of zirconium and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission product are also analyzed in this paper. Furthermore, as important mitigation measures, the effects of makeup water in SFP on the accident progressions have also been investigated based on the events of spent fuels uncovery. The results showed that spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate higher than evaporation rate defined in the reference case. Although spent fuel assemblies partly exposed due to a mass flow rate of makeup water smaller than the average evaporation rate, continuous steam cooling and radiation heat transfer might maintain the spent fuels coolability as the actual evaporation was balanced by the makeup in a period of time of the order of several days. However, larger makeup rate should be guaranteed to ensure long-term safety of SFP.Copyright

Simulation of Molten Corium Concrete Interaction With the MOCO Code

In the event of a severe accident in a pressurized water reactor, the core of a reactor melts and forms corium, a mixture that includes molten UO2 and ZrO2. If the reactor pressure vessel fails, corium can be relocated in the containment cavity and interact with concrete forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question is what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using a new computer code MOCO. The code considers the heat transfer behavior in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, and incorporates phenomenology that is deemed to be important for analyzing debris cooling behavior. The interaction between thermalhydraulics and physic-chemistry is modeled in MOCO. The main purpose of this paper is to present the modeling used in MOCO and some validation calculations using the data of experiments available in the literature.Copyright

MAAP5 Simulation of the PWR Severe Accident Induced by Pressurizer Safety Valve Stuck-Open Accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) induced by the he pressurizer safety valve stuck-open accident with MAAP5. Three cases corresponding to 1, 2 and 3 pressurizer safety valves stuck open accidents without high pressure injection (HPI) and the auxiliary feed water (AFW) of steam generator (SG) were analyzed. The results showed that the pressure safety valve stuck open with HPI and AFW failure accident would lead to a severe accident. And the severe accident sequence was correlated with the number of the stuck open valve. In the case of 1 valve stuck open, the coolant lost slowly and the primary system depressurized slowly which delayed the activation of the low pressure injection (LPI). The reactor pressure vessel (RPV) failed at this case. On the contrary, in the case of 2 and 3 valves stuck open, the coolant lost faster and the pressure decreased rapidly to the critical pressure of the LPI. Thus, the consequences of the two cases were much slighter. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.Copyright

The Code Development for the Thermo-Hydrodynamic Characteristics Study of Reflood Phase

An thermo-hydraulic analysis model for the reflood phase was established in the paper, based on the flow and heat transfer characteristics of the reflood. A code based on the model was developed for the thermo-hydraulic analysis of reflood. By comparing and analyzing the calculation results with the experimental results, the influences of the system pressure and the subcooling of coolant on reflood phase were studied. The study also provided theoretical basis for safety analysis of fuel element of reflood phase.Copyright

Numerical Research on Oscillation of Two-Phase Flow Under Rolling Motion

Two-phase flow instability and dynamics of a parallel multichannel system has been theoretically studied under periodic excitation induced by rolling motion in the present research. Based on the homogeneous flow model considering the rolling motion, the parallel multichannel model and system control equations are established by using the control volume integrating method. Gear method is used to solve the system control equations. The influences of the inlet, upward sections, and heating power on the flow instability under rolling motion have been analyzed. The marginal stability boundary (MSB) under rolling motion condition is obtained. The unstable regions occur in both low and high equilibrium quality and inlet subcooling regions. The multiplied period phenomenon occurs in the high equilibrium quality region and the chaos phenomenon appears on the right of MSB. The concept of stability space is presented.© 2010 ASME

Research on Two-Phase Flow Instability in Parallel Multi-Channel Under Rolling Motion Condition

Two-phase flow instability of the parallel multi-channel system has been studied under rolling motion condition in this paper. Based on the homogeneous flow model with considering the rolling motion condition, the parallel multi-channel model is established by using the control volume integrating method. Gear method is used to solve the system equations. The influences of the inlet and upward sections and the heating power on the flow instability under rolling motion condition have been analyzed. The marginal stability boundary (MSB) under rolling motion condition is obtained and the unstable regions occur in both low and high equilibrium quality regions. The region with low inlet subcooling is also instable. In high equilibrium quality region, the multiplied period phenomenon is found and the chaotic phenomenon appears at the MSB. The oscillation part of mass flow rate (amplitude) may be averaged into other channels so that the influence of rolling motion is weakened. But the stability of multi-channel system is independent of the channel number and the increase of the channel number could only make the amplitude more uniformity in channels.Copyright