Automatic Depressurization System Research Articles

Depressurization strategy of China's advanced pressurized water reactor under typical accidents

Advanced pressurized water reactor was simulated using system analysis program. The depressurization strategy for the automatic depressurization system (ADS) under typical ADS-triggering accidents (2-inch/10-inch cold leg break, inadvertent opening of the ADS, DVI pipeline break) was investigated. The results show that after these accidents, ADS can ensure the operation of the three types of safe injection tanks, with high redundancy and inherent safety. The multi-stage design of ADS aids the sparger in maintaining a stable critical jet state. In 10-inch cold leg break accident, the sparger cannot reached the critical jet state under all of pipeline fault conditions. From a system point of view, adjusting the pipeline design of ADS is necessary to avoid long-term condensation oscillation of the sparger in some accidents. From a design perspective of the local equipment, the nozzle structure of the sparger should attenuate the dynamic load caused by the condensation oscillation and back-attack phenomenon.

Event Sequence Based Fault Tree Analysis to Evaluate Minimal Combination of Event Sequences Leading to the Reactor Core Damage

Probabilistic safety assessment has been widely used to evaluate nuclear power plant safety systems. It couples fault tree (FT) analysis and event tree (ET) analysis. FT analysis is to develop failure scenarios, to quantify the failure probability and to identify critical components. ET analysis is to develop event sequences (ESs) leading to reactor core damage and quantify core damage frequency. Currently, there is no approach in probabilistic safety assessment that can be used to identify minimal combinations of safety system failures leading to reactor core damage. This study proposes an event sequence based fault tree analysis, which integrates FT model with ET model. The ET model is to identify the ESs leading to the reactor core damage. Meanwhile, the FT model is to identify minimal combinations of those ESs found in the ET model. The motivation of this study is on how to identify critical safety systems in nuclear power plants to mitigate disturbances caused by a group of postulated initiating events to avoid core damage. To confirm the feasibility and the applicability of the proposed approach, the performance of the AP1000 passive core cooling system is evaluated. It is found that if in-containment refueling water storage tank, automatic depressurization system - full, accumulator, core make-up tank, and passive residual heat removal work properly, the reactor core will intact. These results confirmed that the proposed approach can be applicable to identify minimal combinations of safety systems to keep the reactor core intact.

Depressurization strategy of automatic depressurization system for advanced pressurized water reactor

China’s advanced pressurized water reactor is simulated using a system analysis code. A strategy for operating the reactor’s automatic depressurization system (ADS) when it is inadvertently opened is investigated. Results show that under four typical ADS pipeline failure conditions, the ADS can effectively release the primary loop pressure and maintain the safety of reactor core. The redundancy of the two sets of independent ADS depressurization pipelines is designed to be 2 × 100%. The step-by-step opening of pipelines ADS-1–ADS-3 can effectively improve the depressurization and cooling rates of the primary loop. Pipeline ADS-2 is effective for accelerating depressurization, and pipeline ADS-3 can be regarded as the redundancy of ADS-2. The step-by-step opening of the pipelines increases the inlet pressure and widens the inlet pressure curve of the sprayer, enabling the sprayer to maintain a long and stable critical jet state.

Multi-Scale study of an innovative safety system for Pressurized water reactors

After the FUKUSHIMA accident, passive systems become an important issue for the new projects of Pressurized Water Reactor (PWR). In response to the demands of its partners, the CEA (French Commission for the Atomic Energy and Alternative Energy) has proposed suitable systems to deal with different accidental sequences. Then, it has dimensioned their components and modelled their integration in the reactor using the CATHARE system code. In particular, the CEA has worked on the ADS system (Automatic Depressurization System) devised to tackle with the well-known LOCA accident (Loss of Coolant Accident), resulting from a breach in the pipes of the primary circuit.Concretely, the ADS system is made up of a tank called IRWST (In-Containment Refueling Water Storage Tank), placed high in the containment, which has the role of achieving a safety passive injection to replace the conventional one using low-pressure pumps, during a LOCA accident. The ADS is in fact already present on the AP1000 reactor, the US PWR designed to be totally passive. The CEA has proposed to improve it by adding a LPWT tank (Low-Pressure Water Tank) pressurisable by steam issued from the pressurizer on the last line of the ADS system. It should allow a better core cooling by the additional water injection in the reactor core, anticipated in time compared to that of the IRWST.The CFD application in reactor safety studies, due to its high-level of accuracy, is now widespread. The CFD calculation is used to predict local fluid temperature in the different regions of a nuclear reactor in nominal conditions and transient situations. Local CFD calculation coupled with a component system code for the whole domain of the reactor is also a current approach (Guelfi et al.).The study presented here exclusively focuses on the LPWT water pressurization phase prior to its injection into the reactor core. Three approaches are exposed:- An analytical approach, in which one considers the pressurization of the volume of gas present in the tank without taking into account the liquid phase;- A system approach with the CATHARE thermal–hydraulic code which allows to describe at the same time the volume of gas and the volume of liquid contained in the tank;- A CFD approach using the NEPTUNE_CFD code to determine the local evolution of the quantities that characterize the system. This approach will notably make it possible to describe the movement of the interface and the velocity profile of the water vapor. It will also help to bring a better understanding of the physical phenomena involved during the LPWT water pressurization phase.At the end, these different approaches show that the increase in pressure by the steam injection is an interesting option to pressurize the water tank. On the other hand, the results of the pressure evolution obtained by the system approach are very close to those obtained analytically, in the absence of liquid phase. However, they are overestimated compared to the CFD approach, which, thanks to a better modeling of the condensation of the injected vapor, leads to a more realistic evaluation.

Experimental Study on Sensitivity of PRHR Pipeline Break Location on ACME Test Facility

To study the safety performance of the advanced passive (AP) nuclear power plant (NPP) under passive system failure condition, the experimental study on the loss of coolant from the passive residual heat removal (PRHR) pipeline break is performed by the advanced core-cooling mechanism experiment (ACME) facility, during which the effect of main test sequences and break location on the key parameters in different phases of the accident is analyzed. As demonstrated by the study results, the ACME PRHR pipeline break test sequences, basically same as those for the small-break loss of coolant accident (SBLOCA) of the cold leg, reproduce the safety characteristics in the natural circulation phase of the passive NPP, blowout phase of the automatic depressurization system (ADS) and safety injection (SI) phase of the in-containment refueling water storage tank (IRWST); in the test at different break locations, the passive core cooling system (PXS) can always ensure the water makeup for core and the core active region remains below the mixed liquid level; and the break locations have notable effect on the ACME LOCA accident sequence, initial depressurization rate of reactor coolant system (RCS), PRHR heat exchanger (HX) flow, blowout flow, core level, IRWST SI flow and other parameters, and have little impact on the SI flow of the core makeup tank (CMT) and accumulator (ACC).

Evaluation of Passive Safety Injection System Performance Under Large Break LOCA for Qinshan PWR

In this work, a brand new passive safety injection system has been designed for the ocean-based Qinshan Phase I nuclear power plant to update and replace the traditional active ones. The passive safety injection system is made up of high pressure, medium pressure, lower pressure safety injection system, and a two-stage automatic depressurization system. To evaluate the safety injection system performance, double-ended cold leg large break LOCA has been analyzed by best-estimated safety analysis RELAP5 code. The main operation and safety parameters such as primary system pressure, safe injection mass flow rates, core water level, and peak cladding temperature have been presented. The results conclude that the safety injection system can act as similar to that of the AP1000, which can assure sufficient core cooling and keep the reactor covered by the cold water under the most severe LBLOCA condition.

Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600

Abstract The AP600 is a Westinghouse Advanced Passive PWR with a two–loop 1 940 MWt. This reactor is equipped with advanced passive safety systems which are designed to operate automatically at desired set-points. On the other hand, the failure or nonavailability to operate of any of the passive safety systems may affect reactor safety. In this study, modeling and nodalization of primary and secondary loops, and all passive reactor cooling systems are conducted and a 10-inch cold leg break LOCA is analyzed using ATHLET 3.1A Code. During loss of coolant accident in which the passive safety system failure or nonavailability are considered, four different scenarios are assumed. Scenario 1 with the availability of all passive systems, scenario 2 is failure of one of the accumulators to activate, scenario 3 is without actuation of the automatic depressurization system (ADS) stages 1–3, and scenario 4 is without actuation of ADS stage 4. Results indicated that the actuation of passive safety systems provide sufficient core cooling and thus could mitigate the accidental consequence of LOCAs. Failure of one accumulator during LOCA causes early actuation of ADS and In-Containment Refueling Water Storage Tank (IRWST). In scenario 3 where the LOCA without ADS stages 1–3 actuations, the depressurization of the primary system is relatively slow and the level of the core coolant drops much earlier than IRWST actuation. In scenario 4 where the accident without ADS stage-4 activation, results in slow depressurization and the level of the core coolant drops earlier than IRWST injection. During the accident process, the core uncovery and fuel heat up did not happen and as a result the safety of AP600 during a 10-in. cold leg MBLOCA was established. The relation between the cladding surface temperature and the primary pressure with the actuation signals of the passive safety systems are compared with that of RELAP5/Mode 3.4 code and a tolerable agreement was obtained.

Large-break LOCA analysis of CSR1000

Large-break loss of coolant accident (LBLOCA) analysis of China Supercritical water-cooled Reactor (CSR1000) has been carried out to evaluate its safety system capability. Analysis results indicate that one significant feature of cold-leg LBLOCA is the reverse-flow which occurs in the core due to the break blowdonwn. The reverse-flow drives the pseudo steam into the core and results in the core heat transfer deterioration which induces a quick increasing of the cladding temperature. After the actuation of the automatic depressurization system (ADS) valves, the core coolant forward-flow recovers and the excessive core heat-up is mitigated. The high feed water tank (HFT) provides enough coolant supply to give the low pressure injection system (LPSI) more response time. After the blowdown phase, the core can be re-flooded by LPSI system. The calculated maximum cladding temperature keeps below the safety criterion.

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.

CFD study on onset of liquid entrainment through ADS-4 branch line in AP1000

The consequences of liquid entrainment through the fourth stage Automatic Depressurization System (ADS-4) of Advanced Passive nuclear power reactor (AP1000) are decreased reactor core coolant inventory and increased resistance to quick depressurization of the reactor primary coolant system. Therefore, the accurate knowledge of the Onset of Liquid Entrainment (OLE) is critical to nuclear reactor safety (NRS) assessment. In this study, three-dimensional computational fluid dynamics (CFD) simulations of OLE were carried out using a two-fluid Eulerian model of ANSYS Fluent CFD code along with the renormalization group (RNG) κ-ε turbulence model for each phase. In OLE phenomena, the interfacial drag has a significant effect but available choices of interfacial drag coefficient in Fluent were not suitable due to the presence of droplet clusters in entrainment. Therefore, a modified interfacial drag coefficient accounting for droplet clusters was used. The transient simulations were performed with this modified CFD model and validated against ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) data. The comparison between CFD calculations and experimental data shows a good agreement. Furthermore, the effects of the liquid mass flow rates and single inlet on OLE were also investigated with the validated CFD model. To improve the understanding of OLE phenomena, the liquid volume fraction and gas velocity field distributions were also studied.

CFD modeling of liquid entrainment through vertical T-junction of fourth stage automatic depressurization system (ADS-4)

In this study, a three-dimensional (3D) transient computational fluid dynamics (CFD) model is presented to investigated liquid entrainment in gas–liquid flow through vertical T-junction of fourth stage automatic depressurization system (ADS-4) in AP1000. A specialized CFD model for the liquid entrainment phenomenon study, which is based on coupled Eulerian-Eulerian VOF (volume of fluid) formulation with suitable interfacial drag and standard κ-ε turbulence model for each phase was proposed. The entrainment rate was calculated and results were validated with ADETEL experimental data, which was built at XJTU-NuTheL. The good agreement between CFD calculations and experimental data demonstrated that the proposed CFD model could reasonably predict entrainment within the studied range. The effect of vertical to horizontal branch diameter ratio and gas mass flow rates on liquid entrainment were also studied. In addition, the liquid volume fraction distributions and velocity field were investigated to develop an understanding of the entrainment process. The relatively high demand for computational resources due to very small timestep size and small grid size to accommodate high flow velocities was found to be a challenge.

Development of severe accident mitigation technology and analysis for SMART

Severe accident mitigation technology was developed and evaluated for a small integral reactor of SMART. The containment pressure and hydrogen behavior were analyzed using MELCOR computer code (Sandia National Laboratory, 2018) in the SBO (Station Black Out) sequence, which was selected from PSA (Probabilistic Safety Assessment) results for the SMART. The severe accident mitigation technology to improve the SMART safety include reactor vessel depressurization using the ADS (Automatic Depressurization System) to prevent DCH (Direct Containment Heating) in case of a reactor vessel failure, a reactor cavity flooding using the CFS (Cavity Flooding System) with the IRWST (In-containment Refueling Water Storage Tank) for the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) to prevent the reactor vessel failure, and a hydrogen control system of PARs (Passive Autocatalytic Recombiners) to remove hazards from hydrogen combustion considering the amount of hydrogen to be generated by 100% fuel cladding oxidation. The MELCOR results showed that the containment pressures during the SBO sequence was below the design pressure, which meant that the containment integrity is maintained during a severe accident in the SMART. The hydrogen mole fraction of the containment was much lower than the hydrogen safety criteria of 10 vol%, which meant that the possibility of the hydrogen burn in the containment was negligible. For this reason, it could be concluded that the hydrogen was controlled by the hydrogen moving path and the hydrogen control system of PARs in the SBO sequence of the SMART.

RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

Qinshan nuclear power plant is the first large Chinese-designed nuclear power station based on pressurized water reactor, and the second generation main stream active safety injection system is adopted for Qinshan nuclear power plant. In this paper, a novel passive safety injection system (PSIS) has been proposed for ocean-based Qinshan Phase One nuclear power plant to replace the original active one. The PSIS contains high-pressure, medium-pressure, and lower-pressure safety injection systems, and a two-stage automatic depressurization system. To evaluate the system performance, small-break LOCA has been investigated using RELAP5. Various break sizes and locations including 2-inch, 10-inch cold leg break, and double-ended direct vessel injection line break were analyzed. Key safety parameters such as safe injection mass flow rates, coolant level of the core, system pressure, and fuel cladding temperature were monitored during the accident process. The results illustrate that the performance of the safety injection system can guarantee the effective core cooling and submerged under different LOCA even with only half of the safety injection system, which can fulfill the single failure criteria. The thermal-hydraulic analysis for the Qinshan passive safety injection system is significant to master the related technologies for Chinese engineer and develop the Chinese-designed third-generation nuclear power plants, and the PSIS can guarantee the reactor submerged under LOCA even plus the station block out accident.

Comparison of Two Different Sized Small-Break LOCAs on the Passive Safety Injection Line Using SMART-ITL Data

Even for small modular reactors (SMRs) with all large pipes removed, a small-break loss-of-coolant accident (SBLOCA) is an important design-basis accident (DBA). Experimental simulation of the SBLOCA scenario is essential before a prototype reactor is realized. The system-integrated modular advanced reactor (SMART) is one of the SMRs developed by the Korea Atomic Energy Research Institute. An integral test loop, SMART-ITL, was also constructed to carry out several types of integral thermal-hydraulic effects tests for the prototype reactor. The SMART-ITL was designed with a preserved height, 1/7th diameter, and 1/49th area, and volume-scaling ratios. Two types of passive safety systems were equipped in the SMART-ITL: a passive safety injection system (PSIS) and a passive residual heat removal system (PRHRS). The PSIS was designed to refill the coolant in the reactor coolant system (RCS) for 72 h after an accident. Under accident conditions the PRHRS prevents overheating and overpressurization of the RCS using two-phase natural circulation. The SBLOCA on the passive safety injection line is a significant DBA that should be validated for differences in break size. In this paper, the effects of two different break sizes, 2 and 7/32 in., were analyzed in order to study the effect of the maximum and minimum mass and energy loss of the RCS. In order to simulate a clear difference between maximum and minimum mass and energy loss of the RCS, heat removal by the PRHRS was performed in the maximum break size (2-in.) accident, and heat removal by the PRHRS was not conducted in the minimum break size (7/32-in.) accident. The difference in mass and energy loss of the RCS will have a significant impact on the operation of the automatic depressurization system. Using the two extreme accident simulations, it was possible to confirm the difference in accident progression caused by the difference in break size and the characteristics of the PSIS.

Analysis of post-LOCA long-term core safety characteristics for the Small Modular Reactor ACP100

The China Small Modular Reactor ACP100 is an integral pressurized water reactor with a passive core cooling system. Analyses have been carried out to evaluate the ability to remove decay heat and to prevent boric acid precipitation during the long term cooling phase post-LOCA. The long-term core cooling characteristics have been analyzed on a postulated DEDVI accident. The results show that the key parameters affecting the core cooling ability are the containment pressure and the steam discharge resistance. Simplified calculations have been performed on the boron concentration variation before the automatic depressurization system (RDP) actuation as well as after the RDP actuation. The results indicate that the climbing speed of the core boron concentration and the maximum value that can be reached depend on the core exit quality. The improved passive core cooling system can prevent the core from boron precipitation and returning to criticality post-LOCA.

Performance analysis of automatic depressurization system in advanced PWR during a typical SBLOCA transient using MIDAC

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

Scaling and integral effect test on transition from depressurization to long-term gravity injection

The passive safety commercial pressurized water reactors (PWRs), AP1000 and CAP1400 are equipped with the automatic depressurization system (ADS). The fourth stage of ADS (ADS-4) has the largest discharging capacity among the all ADS stages. During a small break loss-of-coolant accident (SBLOCA), only when the system is fully depressurized by the ADS-4 to near the containment atmospheric pressure, the in-containment refueling water storage tank (IRWST) can deliver the large coolant inventory for the long-term core cooling by its water head. Before the IRWST injection, the core reaches its minimum inventory. So the transition from the ADS-4 depressurization to the IRWST injection is a very challenging period in a SBLOCA. Therefore, to evaluate the passive system performance by the integral effect test (IET), the transition must be simulated properly. This work conducted the scaling analysis for the ADS-4 depressurization and the IRWST injection for designing the related components of the ACME IET facility. A uniform scaling law based on the existing T-branch liquid entrainment models was proposed. For scaling the onset of the IRWST injection, the requirement for reproducing the synergetic effect in an IET was introduced in order to preserve the important event/process occurrence time with the proper conditions, and the scaling laws for the full pressure IRWST injection were obtained. For investigating the transition thermal hydraulic behaviors and the ADS-4 configuration influences on the core safety, one cold leg break test and three double-ended direct-vessel-injection (DEDVI) line break tests with the different ADS-4 failure modes were selected from the ACME SBLOCA test matrix, and the main test data related to the core safety were analyzed and compared. It was found that the ADS-4 actuation significantly accelerated the system depressurization but worsened the core inventory depletion at the same time. In addition, the break condition and the ADS-4 failure mode could differently affect the system depressurization and the core level. The ACC injection during the ADS-4 depressurization in DEDVI was critical for the core level makeup. The insufficient ADS-4 venting capacity delayed IRWST injection, but also mitigated the core inventory depletion. It suggests the ADS-4 design can be potentially optimized in the passive safety system design.

Experimental study on the condensation of sonic steam in the underwater environment

Abstract Steam jet condensation is of great importance to pressure suppression containment and automatic depressurization system in nuclear power plant. In this paper, the condensation processes of sonic steam jet in a quiescent subcooled pool are recorded and analyzed, more precise understanding are got in direct contact condensation. Experiments are conducted at atmospheric pressure, and the steam is injected into the subcooled water pool through a vertical nozzle with the inner diameter of 10 mm, water temperature in the range of 25–60 °C and mass velocity in the range of 320–1080 kg/m2s. Richardson number is calculated based on the conservation of momentum for single water jet and its values are in the range of 0.16–2.67. There is no thermal stratification observed in the water pool. Four condensation regimes are observed, including condensation oscillation, contraction, expansion-contraction and double expansion-contraction shapes. A condensation regime map is present based on steam mass velocity and water temperature. The dimensionless steam plume length increase with the increase of steam mass velocity and water temperature, and its values are in the range of 1.4–9.0. Condensation heat transfer coefficient decreases with the increase of steam mass velocity and water temperature, and its values are in the range of 1.44–3.65 MW/m2°C. New more accurate semi-empirical correlations for prediction of the dimensionless steam plume length and condensation heat transfer coefficient are proposed respectively. The discrepancy of predicted plume length is within ± 10% for present experimental results and ± 25% for previous researchers. The discrepancy of predicted condensation heat transfer coefficient is with ± 12%.

Safety analysis of a small modular reactor using fully ceramic micro-encapsulated fuel

Fully ceramic micro-encapsulated (FCM) fuel has high thermal conductivity and fission product retention capabilities. It can improve the accident tolerance ability of light water reactors (LWRs). Small modular reactors (SMRs) shows a good inherent safety due to their good design and low power density. Combining FCM fuel and SMR may greatly improve the safety of the reactor. In this paper, a SMR model using FCM fuel is built by referring to the Westinghouse Small Modular Reactor (WSMR). Based on this model, we analyzed some accidents associated with SMR using FCM fuel and compare this information with SMR using UO2 fuel. The results indicate that the steady running temperature of FCM fuel is low and that the temperature increases slowly in accidents. The SMR using FCM fuel has good safety features and delay the melting of the core. Under the conditions of station black-out (SBO) combined with failure of core makeup tanks (CMTs) and ADS automatic depressurization system (ADS), the failure time of the core using FCM fuel is 5.49E4 s, which is 7.10E3 s (1.97 h) longer compared with that of UO2 fuel. When considering reinvestment in CMTs, the time limit of FCM fuel is 1.33E4 s (3.70 h) longer compared with that of UO2 fuel, which offers enough time for the restoration of safety system in the accident state. Because of its low temperature, the hydrogen production rate of FCM fuel is reduced compared with that of UO2. However, given the use of zirconium cladding, the final output of hydrogen is the same as or even more than UO2 fuel, and hydrogen elimination devices must be considered.

Development of MELCOR thermal hydraulic model of AP1000 and its verification for a DECL break

The AP1000 is currently an advanced pressurized water reactor with emergency core cooling systems (ECCS) completely dependent on natural circulation. It is under construction in the USA and has been commissioned in China and is being considered by several countries to use this advanced technology for a power source. With this objective in mind, a representative MELCOR model (for the AP1000) has been developed to better understand the thermal hydraulics involved and to assess the performance of the safety systems for a design basis accident such as large break LOCA (LBLOCA) with availability of the safety systems. The results from these calculations were compared to those obtained using WCOBRA-TRAC, developed by Westinghouse Electric Company, which are reported in the AP1000 European Design Control Document.MELCOR was selected for the design assessment activities specifically because of its capabilities to model severe accident scenarios and features such as containment behaviour. It is one of the system thermal hydraulic codes, which are recommended for severe accident analysis in light water reactors. The aim of the current study is to assess the thermal hydraulic models and correlations in MELCOR against a specific code developed for a design basis accident such as a LBLOCA. This exercise will validate the input model for a LBLOCA and then use the input model for station blackout studies and the impact on the containment.A double ended cold leg (DELC) break in the loop containing the Core Makeup Tanks (CMT) is considered for this study. The calculated break flow, peak cladding temperature, accumulator, and CMT injection flow rates are compared to those reported. These calculated results are found to be satisfactory according to ECCS acceptance criteria for the temperature and hydrogen generation. This study has been further extended to analyse the containment behaviour during the accident. The calculations for the transient were extended to observe the behaviour of the Automatic Depressurization System (ADS) and IRWST injection capabilities which can effectively avoid core uncovery for a longer duration of the time. These systems can only work when the reactor system is depressurized. MELCOR calculated a PCT value (occurs just after the break) as 1242 K while the total hydrogen produced was calculated to be 2.216 kg which is only 0.61% of the total potential hydrogen production from oxidation.