MAAP5 Code Research Articles

Study on spent fuel heatup during spent fuel pool complete loss of coolant accident

Abstract After years of operation, nuclear power plants accumulate a substantial amount of spent fuel stored in spent fuel pools. Prior to the completion of dry storage facilities, this spent fuel remains in the spent fuel pools for several decades. The progression of accidents in spent fuel pools leading to fuel damage typically spans several days or even weeks, which may lead people to overlook the importance of spent fuel pool safety. However, accidents involving spent fuel pools have a characteristic of low probability but high consequences. Their significance has been reemphasized, especially after the Fukushima accident. This study conducts an analysis of the scenario where a spent fuel pool experiences a complete loss of water. A plant equipped with a BWR-4 reactor and a MARK-I containment structure is used as the reference plant. The analysis utilizes the MAAP5 code to assess the impact of different fuel cooling times and ventilation flow rate of secondary containment building on the fuel temperature variation, considering only air circulation for cooling. Based on the analysis results, the plant utilizes the normal ventilation system of the reactor building, which can maintain the fuel temperature below 565 °C with a cooling time of 1 year. Additionally, it was found in the study that adequate ventilation flow rates and sufficient cooling time can ensure that the fuel temperature remains below 565 °C, thereby preserving fuel integrity.

Sensitivity Analysis of Ex-Vessel Corium Coolability Models in MAAP5 Code for the Prediction of Molten Corium–Concrete Interaction after a Severe Accident Scenario

A postulated progressing severe accident scenario has been simulated using MAAP5 code with the focus on ex-vessel cooling of molten corium in the reactor cavity. Various parameters associated with the prediction of molten corium–concrete interaction (MCCI) are identified. Accordingly, a sensitivity analysis is performed to assess the impact of these parameters on the predicted cavity floor erosion depth during this MCCI postulated accident. The sensitivity index of each variable parameter is determined using the Cotter indices method and Sobol′ indices method. At the early stage of the accident, the predicted cavity floor erosion depth is found to be highly sensitive to the downward heat transfer coefficient parameter with Cotter and Sobol′ indices of 94% and 50%, respectively. At the late phase of the accident, however, the cavity floor erosion depth becomes sensitive to melt eruption (Cotter index of 40%), water ingression (Cotter index of 13%), and particulate bed (Cotter index of 15%) parameters alongside the downward heat transfer coefficient (Cotter index of 16%) with the melt eruption parameter becoming dominant. Thus, the sensitivity of the code′s predictions can be minimized by improving the physical models associated with these parameters. Moreover, the sensitivity indices of these parameters can be used by model developers to identify unimportant parameters in a bid to reduce the dimension of the problem with the aim of improving the current predictive capabilities to conduct MCCI-related safety analyses.

Evaluation of temperature and flow area variations through the fuel degradation and relocation of the SFD Test 1–4

We used the MAAP5 code and evaluated temperature and flow area variations through fuel degradation and fuel relocation in the Severe Fuel Damage (SFD) Test 1–4. The measured temperatures in the fuel bundle were predicted by considering the steam diffusion-limited effect within the hydrogen-rich boundary layer near the fuel cladding. After confirming that the temperatures in the fuel bundle could be predicted, we clarified the effect of two different binary phase diagrams between UO2 and Zr on the computed results of relocation and flow area. One was the binary phase diagram based on MATPRO, and the other was the diagram drawn using a thermodynamic database. We found that the height of the relocated U-Zr-O mixture evaluated using the binary phase diagram drawn with the thermodynamic database was slightly lower than that evaluated using the phase diagram based on MATPRO, but the difference was small in the high temperature and diluted steam conditions of the SFD Test 1–4.

Uncertainty studies on hydrogen source term with MAAP5 code

Abstract In the present study, a methodology is developed to quantify the uncertainties of special model parameters of the integral severe accident analysis code MAAP5. Here, the in-vessel hydrogen production during a core melt accident for Lungmen Nuclear Power Station of Taiwan Power Company, an advanced boiling water reactor, is analyzed. Sensitivity studies are performed to identify those parameters with an impact on the output parameter. For this, multiple calculations of MAAP5 are performed with input combinations generated from Latin Hypercube Sampling (LHS). The results are analyzed to determine the 95th percentile with 95% confidence level value of the amount of in-vessel hydrogen production. The calculations show that the default model options for IOXIDE and FGBYPA are recommended. The Pearson Correlation Coefficient (PCC) was used to determine the impact of model parameters on the target output parameters and showed that the three parameters TCLMAX, FCO, FOXBJ are highly influencing the in-vessel hydrogen generation. Suggestions of values of these three parameters are given.

Analysis of AP1000 severe accident induced by SBO using MAAP5

The total loss of AC power is an essential initiating event of PWR which can lead to a severe accident like Fukushima in Japan. For such improbable accidents, three cases of severe accidents of AP1000 induced by the station blackout (SBO) were simulated based on MAAP5 code. This particular analysis was conducted at 100% core power with different approaches. For the case with auto depressurization system (ADS), the primary system pressure appeared to uphold a low level for a longer duration and the external reactor vessel cooling (ERVC) system had sufficient capacity to eliminate the heat from the molten pool with a ratio of ex-vessel heat flux to critical heat flux (CHF) at about 0.49. For the case without ADS, the reactor pressure vessel (RPV) rupture will occur due to the creep effect inducing a sudden shock to the containment, and the peak pressure in the containment is still under the level at which the inelastic deformation may take place. The hydrogen production of the case without ADS in the core is 83% more than that of having ADS, while the hydrogen mole fraction in the upper containment is lower before the RPV failure. Overall, our findings are meaningful and significant for comprehending the detailed process of AP1000 severe accident induced by SBO, which is the basic standard for the development of the severe accident management guidelines (SAMG).

Uncertainty Quantification Using the MAAP5 Code of In-Vessel Hydrogen Generation in a Severe Accident at an Advanced Boiling Water Reactor

MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic safety assessments, verification and validation of mitigation actions specified in severe accident management guidelines, and source term quantification. In this study, the uncertainty of in-vessel hydrogen generation predicted by the MAAP5 code was quantified. The surrogate plant that was analyzed is the Lungmen Nuclear Power Station of the Taiwan Power Company. The plant employs an advanced boiling water reactor. We performed sensitivity studies to identify the important model parameters that affect the target output parameters. A range and distribution were assigned to these parameters on the basis of experimental results and expert judgment. The number of input parameters in the analysis was 27. Multiple MAAP5 calculations were performed with an input combination generated from Latin hypercube sampling. The calculation results were analyzed parametrically and nonparametrically to determine the 95th percentile with the 95% confidence level value of the amount of in-vessel hydrogen generation. The Pearson correlation coefficient was used to determine the effect of the model parameters on the target output parameters. The analysis results provide guidance for code applications. The only parameters that pass the threshold of 0.362 for hydrogen generation in the core are FCO and TCLMAX. For hydrogen generation in the lower plenum, FOXBJ is the only input parameter that passes the threshold.

A method for predicting the structure of corium pool in lower plenum of reactor vessel

The corium in the lower plenum of reactor vessel is expected to stratify into two or three layers. The three-layer corium method adopted by MAAP5 code is based on Salay theory to treat the compositions in oxide and heavy metal layers and assumed that the light metal layer is formed automatically after the corium falls down. In this work, a method is designed for predicting the corium pool structure based on the accident progressions, especially, the formation of the light metal layer is not hypothetical and the condition of forming a three-layer corium pool is given. In addition, the heat flux distribution on the outside wall of the reactor vessel is analyzed with different corium pool structures. It is found that, the structure of the corium pool calculated by this method is unique while it is various calculated by MAAP5 code method, in which the composition of the light metal layer is user input. Although the thickest and thinnest light metal layer can be calculated by MAAP5 code method, the conservative heat flux still cannot be obtained, which confused the validity of IVR strategy.

EFFECTS OF CORE PERIPHERAL COMPONENTS MELTING ON CORIUM POOL CONFIGURATION IN LOWER PLENUM OF ACP1000

The MAAP5 code and CFD method are adopted to estimate the melting behavior of core peripheral components for an Advanced China PWR, called ACP1000 after a large break accident to determine the amount of stainless steel relocating into the corium pool. The results suggest that larger portion of peripheral components melting in CFD method because of considering the radiation of fuel rods on shroud and barrel before core melting and ignoring the cooling of steam on the components in core. Effects of different configurations of corium pool on heat flux imposed on the vessel wall are also investigated in this work. The heat flux in light metal layer for the corium pool that obtained by MAAP5 code could be higher than that of CFD case due to its thinner light metal layer.

Development of a prediction method for SAMG entry time in NPPs using the extended group method of data handling (GMDH) model

The importance of severe accident management is being re-examined because severe accident management failed during the Fukushima nuclear accident. Having an accurate estimation of a severe accident is important for coping properly with unfavorable conditions. In this study, the extended group method of data handling (GMDH) with a fuzzy concept is proposed as a means of predicting the major severe accident events that represent the severe accident management guideline (SAMG) entry time, including the time at which the reactor vessel (RV) water level decreases, the time when the core exit temperature (CET) reaches 450 degrees-centigrade, the time when the hydrogen concentration in the containment is over 4%, and the time when the reactor coolant system (RCS) pressure is over 2.86 MPa under loss of coolant accidents (LOCAs) in pressurized water reactors (PWRs). To train the extended GMDH model, it was necessary to acquire the data needed from a number of numerical simulations due to the lack of actual LOCA data. The data was obtained by carrying out simulations using the MAAP5 code. To optimize the developed model, the optimal input selection processes were performed using the clustering analysis including self-organizing feature map (SOFM), hierarchical and non-hierarchical clustering methods. The prediction accuracy of the three types of initiating accident, small break (SB), medium break (MB), and large break (LB) LOCAs, was high enough to predict the SAMG entry time. When compared with other artificial intelligence (AI) methods, the extended GMDH was found to be superior to the original GMDH and support vector regression (SVR). Therefore, it is expected that the extended GMDH model will help operators to properly mitigate severe accidents in PWRs.

Prediction of severe accident occurrence time using support vector machines

If a transient occurs in a nuclear power plant (NPP), operators will try to protect the NPP by estimating the kind of abnormality and mitigating it based on recommended procedures. Similarly, operators take actions based on severe accident management guidelines when there is the possibility of a severe accident occurrence in an NPP. In any such situation, information about the occurrence time of severe accident-related events can be very important to operators to set up severe accident management strategies. Therefore, support systems that can quickly provide this kind of information will be very useful when operators try to manage severe accidents. In this research, the occurrence times of several events that could happen during a severe accident were predicted using support vector machines with short time variations of plant status variables inputs. For the preliminary step, the break location and size of a loss of coolant accident (LOCA) were identified. Training and testing data sets were obtained using the MAAP5 code. The results show that the proposed algorithm can correctly classify the break location of the LOCA and can estimate the break size of the LOCA very accurately. In addition, the occurrence times of severe accident major events were predicted under various severe accident paths, with reasonable error. With these results, it is expected that it will be possible to apply the proposed algorithm to real NPPs because the algorithm uses only the early phase data after the reactor SCRAM, which can be obtained accurately for accident simulations.

Analysis of PWR RPV lower head SBLOCA scenarios with the failure of high-pressure injection system using MAAP5

Through the use of MAAP5 code we carried out an analysis of the severe accident scenarios initiated by an unusual small-break loss-of-coolant accident (SBLOCA) located at the reactor pressure vessel (RPV) lower head with the failure of high-pressure injection system (HPIS) in a PWR power plant. The cases of 0.4-inch-break LOCA with and without manual depressurization of the reactor coolant system (RCS), and 1.0-inch-break LOCA were studied. The results showed that without manually depressurizing the RCS, reactor core underwent slow heatup and completely melted and eventually failed the RPV lower head at the primary system pressure of 6.87 MPa for 0.4-inch-break LOCA and 2.71 MPa for 1.0-inch-break LOCA. There was a strong interaction between the discharged melt and collected pool water in the reactor cavity at the time of RPV lower head failure by creep. A 0.472 MPa containment peak pressure was caused by hydrogen combustion for 0.4-inch-break LOCA. For 1.0-inch-break LOCA the pressure peaked twice, the first one (0.396 MPa) by intense steam generation and the second one (0.331 MPa) by hydrogen combustion. These pressure spikes were lower than the containment design pressure of 0.52 MPa. It was not likely to stop the cavity from being eroded by molten corium through continuously pouring water down into the cavity. Thus there were constant reactor cavity ablation and hydrogen generation, leading to dangerous erosion depths and elevated hydrogen volume fraction in the presence of containment spray system. On the other hand, timely manually opening the pressurizer (PZR) safety valves (SVs) was an effective mitigation measure to recover the core coolabilty by cold-leg accumulator injection system (AIS) and low-pressure injection system (LPIS). Besides, reasonably manually opening the steam generators (SGs) SVs while keeping the auxiliary feedwater on also helped to depressurize the RCS and prevent the severe accident. Both of the two mitigation measures successfully prevented the core from complete melt, but the latter one is preferable to the former one provided no steam generator tube rupture takes place.

Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

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 the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2m (designated as a reference case) and 10.7m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the top of the spent fuel assemblies by the amount of time to start uncovering the spent fuel assemblies when no mitigation measures were taken. Although spent fuel assemblies were partly exposed due to the mass flow rate of makeup water being 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 the long-term safety of SFP.

The Study of Loss-of-Coolant Accident Parameter Identification for a BWR-4 Plant

In this study the parameter identification tool (PIT) for the loss-of-coolant accident (LOCA) of a BWR-4 plant with Mark-I containment, Chinshan nuclear power plant (NPP) in north Taiwan, is developed. The PIT is composed of the Simplex search algorithm and the MAAP5 code with the integration approach presented by Tsai et al. The postulated LOCA is assumed to occur after station blackout sequence with the break elevation of 17 m and the break area of 2.1×10-2 m2 . Through the developed PIT, the break elevation of 16.93 m and the break area of 2.1×10-2 m2 are identified with the imitated plant data, the RPV pressure and the shroud water level, generated by MAAP5 code. The timing of events and source terms in the LOCA of Chinshan NPP can be predicted with by the MAAP5 code with the identified parameters. It demonstrates that the PIT is helpful and important in severe accident management.

Study on the minimum number of SRVs required for emergency depressurization in EOP of BWR power plants

The minimum number of safety/relief valves (SRVs) required for emergency depressurization (MNSRED) is an important reactor pressure vessel (RPV) parameter described in the appendix C of BWR Owners’ Group, Emergency Procedure Guidelines and Severe Accident Guidelines (BWROG, EPG/SAGs). It is directly related to the emergency depressurization if there are less SRVs available. MNSRED is always coupled with decay heat removal pressure (DHRP) in emergency depressurization. DHRP is used as the depressurization low-pressure end to ensure the decay heat can be removed. The purpose of this paper is to discuss their mathematical models in BWROG EPG/SAGs. After study, the vague concepts existed in these parameters are pointed out and clarified. The variables inappropriately used through their evaluations are revised. With improved modifications, the MNSRED turns out smaller than its original value and DHRP is larger than its previous one. These results affect the heat removal capability and the operation margin regarding emergency depressurization. At the end, the revised DHRP value is verified by MAAP5 code with the same plant status for the Kuosheng power station of Taiwan Power Company.

Analysis of pressurizer level control system using MAAP4 code

The code initialization effort has been troubling code users for decades for system transient and severe accident analyses using codes such as RETRAN, MAAP4, MAAP5 and MELCOR. The purpose of this work is to demonstrate an approach that could be considered a generic method to address the code initialization problem. This was demonstrated by developing a pressurizer level control model and temperature dependent level control logic in MAAP4 without re-compiling with the source code. The method would enhance the simulation capability and accuracy of a severe accident analysis by transient and severe accident analyses codes. The demonstration case used MAAP4 to show that the adopted proportional–integral controller with the temperature dependent level control logic would reduce its code steady state errors to zero. The subsequent transient response would become more realistic. The proposed method provides a convenient and exemplified approach for code initialization which is applicable to the next generation of codes that couple with the balance of plant models. These codes include the MAAP5 code and others future codes that could simulate the whole plant by a single and elaborate plant model with exhausting component and phenomenological models.