Validation and parametric study of FFRD model in DRACCAR code

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The systems computer code is a key part of the evaluation model for safety analysis of nuclear reactors. The systems code utilizes a set of governing equation that is simplified from the fundamental Navier-Stokes equations and closure models to describe the transport of mass, momentum, and energy of single phase or multiphase fluid throughout the reactor coolant systems. Following the Evaluation Model Development and Assessment Process, an assessment matrix is established where Separate Effects Tests and Integral Effects Tests are selected based on phenomena identification and ranking table. The purpose of the assessment matrix is to validate the systems code against the important phenomena for the safety analysis. The code biases and uncertainties are established and the effect of scale could then be determined. The assessment matrices of major systems codes, RELAP5/MOD3, TRACE Ver.5.0 and WCOBRA/TRAC-TF2, for the reactor safety analysis are reviewed and compared in this study for the Loss of Coolant Accident (LOCA) safety analysis perspectives. The scenarios are divided into small break LOCA and large break LOCA. The phenomena bases of the separate effects tests in those assessment matrices are discussed following its PIRT. The comparison demonstrates the capability of each systems code.

The emergency core cooling system (ECCS) of a nuclear power plant supplies cooling water to the reactor vessel in the case of a loss of coolant accident (LOCA). A LOCA generates debris by the force of coolant impinging upon pipe insulation and entraining a wide variety of particulate matter from the reactor building surfaces that the coolant flows over. During the recirculation phase following a LOCA, if a sufficient quantity of debris accumulates on the sump screens, the ECCS pumps’ suction flow path can be reduced significantly, causing a drop in the available NPSH and, eventually, a loss of pump flow. If the ECCS flow is lost for a sufficiently long time, the core may become uncovered and overheat, causing severe damage to the fuel. Since the Barseback strainer event (July 1992), studies and experiments are being undertaken all ...

The effects of emergency core coolant (ECC) temperature on the performance of the emergency core cooling system (ECCS) during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) were investigated experimentally using the Rig of Safety Assessment (ROSA)-III integral test facility. The ECC temperature had no direct influence on the ECCS core cooling performance, since ECC became nearly saturated before reaching the core irrespective of the initial temperature, however, had indirect effects by changing the vessel pressure response. The ECCS injection timings and flow rates, and the core inlet flooding behavior were affected. The measured peak cladding temperature (PCT) was not affected by the ECC temperature for both large (200%) and small (5%) break tests.

Behavior of Emergency Core Cooling System (ECCS) during the early stage of Loss of Coolant Accident (LOCA) for APR 1400 with Flownex software

The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Separate-Effects Program is part of the overall light-water reactor safety research program of the Nuclear Regulatory Commission (NRC). Other parts of the NRC program cover a wide range of experimental and analytical efforts from laboratory-scale experiments to small-scale experimental nuclear plants. Separate-effects studies fall between laboratory-scale experiments and small-scale nuclear plants and are designed to answer specific questions relevant to the hypothetical loss-of-coolant accident (LOCA). Some of the different separate-effects studies sponsored by NRC are (1) evaluations of pump behavior during transients with two-phase flow, (2) studies of two-phase flow in downcomers, (3) experiments on the reflood characteristics of rod bundles, (4) tests of alternate emergency core-cooling systems (ECCS), (5) evaluations of fuel rod behavior under extreme conditions that might occur during a LOCA, and (6) determination of heat transfer characteristics of rod bundles under transient conditions postulated for a LOCA.

Loss of coolant accident (LOCA) analyses for various configurations of safety injection system (SIS) are performed to optimize the emergency core cooling system (ECCS) performance for the Korean next generation reactor (KNGR). The KNGR is an advanced light water reactor (ALWR) adopting the advanced design feature of a direct vessel injection (DVI) configuration and passive fluidic device in the discharge line of the safety injection tank (SIT). To determine the feasible SIS configuration and the optimum capacities of the SIT and high pressure safety injection pump (HPSIP), licensing design basis and best estimate LOCA analyses are performed for the limiting large break and small break spectrum, respectively. The analyses results show that the four-train DVI injection with the current system design is a more feasible configuration than the other ones considered and the adoption of a fluidic device SIT enhances the ECCS performance for large break LOCA. For small break LOCA, in the case of cold leg break, the DVI4 configuration is better than other configurations and also meets the EPRI ALWR requirement of no core uncovery for up to a 15.24 cm (6 in) diameter small break. However, in the case of DVI line break, slight core uncovery is predicted and also the system behavior is significantly affected by reactor vessel (RV) downcomer modeling. Therefore, the DVI4 configuration is more feasible for KNGR ECCS performance, but further investigations are required to resolve the ECCS bypass issues for large break LOCA and to develop a proper RV downcomer model for analysis of DVI line break in small break LOCA.

The lessons learned from the Three Mile Island Unit 2 (TMI-2) accident and the Fukushima Daiichi accident demonstrate that under accident conditions, loss of coolant may lead to the core being exposed to air. Failure to promptly activate the emergency core cooling system (ECCS) can result in severe core damage due to fuel melting. Simulating the high-temperature quenching behavior of fuel under loss-of-coolant accident (LOCA) conditions is critical for evaluating reactor safety performance. This study employs the mechanistic severe accident analysis code ICARE2 to establish a numerical model for analyzing the high-temperature quenching behavior of fuel assemblies during LOCA scenarios. Model validation was conducted using experimental data from the QUENCH-06 test, demonstrating strong agreement between ICARE2 simulations and experimental results for key parameters such as peak cladding temperature, quench front propagation, and cladding oxidation. This confirms the code’s effectiveness in modeling fuel degradation, core oxidation kinetics, and two-phase thermal-hydraulic phenomena during severe accidents. A systematic parametric sensitivity analysis was performed to evaluate the influence of steam flow rate, coolant flow rate, coolant subcooling, and cladding temperature during reflooding on accident progression. Simulations of fuel quenching behavior under various LOCA conditions revealed that the steam mass flow rate dominantly affects hydrogen production and pre-quench peak cladding temperature. The results highlight that insufficient cooling is the primary driver of accident escalation. Consequently, the ECCS must be activated promptly before cladding overheating occurs to mitigate severe core damage.

Thermal hydraulic analysis of the AHWR—The Indian thorium fuelled innovative nuclear reactor

Simulation tests of BWR non-loca transients using the tbl

The Super System Code of the Korea Atomic Energy Research Institute (SSC-K) has been developed for the transient analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) system. Recently, a detailed three-dimensional (3-D) core thermal-hydraulic model was developed to describe nonuniformities of radial temperature and flow within a subassembly and to decrease the uncertainties in the reactor safety margins during accident situations. The Shutdown Heat Removal Test-17 (SHRT-17) performed in the Experimental Breeder Reactor-II (EBR-II) and the postulated unscrammed events for the KALIMER conceptual design have been analyzed using a code system that has coupled a detailed 3-D core thermal-hydraulic model with SSC-K. The coupled code predicted behaviors for the experimental trends for the protected loss-of-flow SHRT-17. The KALIMER-150 design was adopted for a plant application of the same code system. Three events, unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), and unprotected loss of heat sink (ULOHS) were analyzed, and the simulation results were compared to those obtained using another code system that has coupled the Safety Analysis Section SYStem (SASSYS)-1 code with the same detailed 3-D core thermal-hydraulic model. The results, calculated with SSC-K coupled with the detailed 3-D core thermal-hydraulic model showed good agreement with the calculated results of the SASSYS-1 coupled code system for the UTOP and ULOF; however, some discrepancies were shown in the results for the ULOHS. These were found to have occurred because of a difference of the modeling for the decay heat removal system and primary coolant inventory. Through these analyses, the coupled code system was validated in order to be available for the safety analysis of a liquid-metal reactor (LMR) plant.

With the increased requirement for nuclear power generation as an effective countermeasure against global warming, Mitsubishi has developed the advanced pressurized water reactor (APWR) by adopting a new component of the emergency core cooling system (ECCS), a new instrumentation and control system, and other newfound improvements. The ECCS introduces a new passive component called the Advanced Accumulator which integrates both functions of the conventional accumulator and the low-pressure pump without any moving parts. The Advanced Accumulator uses a new fluidics device that automatically controls flow rates of injected water in case of a loss-of-coolant accident (LOCA). This fluidics device is referred to as a flow damper. In this chapter, the Advanced Accumulator is introduced from the background of its development to its principle, with some experimental results. Furthermore, the features of the flow damper are explained in detail.

In the unlikely event of a loss-of-coolant accident (LOCA) in a pressurized water reactor, break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS.A systematic study was conducted on various types of fibrous and metallic foil debris to determine their transport in water. Test results reported include incipient movement, bulk movement, accumulation on a screen, the ability of debris to jump over 5-cm (2-in.) and 15-cm (6-in.) curbs, and the effects of accelerating flow and turbulence. These data are currently being used in conjunction with computational fluid dynamics modeling to determine the potential for each debris type to reach the suction screen.

Study on the coolant mixing phenomenon in a 45° T junction based on the thermal-mechanical coupling method

A MTR reactor is an open water pool reactor type. The water pool serves as a shield from radioactive radiations. The most serious accident in this type of reactor is the Loss of Coolant Accident (LOCA) due to rupture either of a primary coolant pipe or of any experimental beam tube. In the present work it has been assumed that pool water drains out due to double ended rupture of the tangential irradiation beam tube (TIC) which has a diameter 150 mm. For an operating power level of 22 MW, the equilibrium core would enter into melting conditions if the pool drain time is less than one hour. It was also assumed that Emergency Core Cooling System (chimney water injection system and siphon effect breaker were not working. The reactor interiors have been modeled using the Monte Carlo N-Particle Transport code MCNPX 2.7.0. The source term has been determined using the ORIGEN-2 code. The doses and dose rates calculations in different places of operator (as phantom of Tissue-Equivalent Material) inside of the reactor building were determined by using Monte Carlo N-Particle Transport (MCNPX).The results show that the dose rate in the control room would be 5.24557 SV/h, the dose rate in the reactor hall above the pools on the gate would be7.20137 SV/h and the dose rate in the Emergency control room would be 2.68239 SV/h. Those dose rates are extremely high and would lead to fatal doses in short time.