Loss Of Feedwater Flow Research Articles

Thermal hydraulic analysis of a high-temperature gas reactor steam generator after loss of feed water accidents

The helical coil steam generator is widely employed in the energy conversion process of high-temperature gas-cooled reactors for its high thermal efficiency in producing superheated steam within the helical tubes. These tubes necessitate materials specifically designed to endure high temperatures, prevent ruptures, and ensure durability under various operational conditions. This study aims to identify the maximum temperatures arising from potential transients in the secondary system and to propose immediate safety measures for minimizing associated risks. The RELAP5/MOD3.4 code was used to simulate the behavior of the next generation nuclear plant steam generator's conceptual design using steady-state conditions. Following the validation of the model, the established steady-state conditions were utilized to simulate accident scenarios, including the feed water line break and the loss of feed water flow to the steam generator. The transient scenarios assumptions include closed flap valves and retained hot helium in the shell side. The simulation tasks were conducted both with and without activating the depressurization system, in order to assess its effect on the temperature of tubes on the wall side. The analysis results indicate that tube wall temperatures can reach 720 °C at the steam generator outlet region and 620 °C at the weld region. The immediate activation of the depressurization system with a controlled relief valve can effectively reduce these temperatures by 239 °C in the steam generator outlet and by 133 °C in the weld region. This also lowers the steam generator's secondary side pressure. As a result, an upgrade and enhancement of operational procedures is straightforward.

Accidents and transients analyses of a super fast reactor with single flow pass core

The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST.

RELAP5-3D transient modelling for NGNP integrated plant

The High-Temperature Gas-cooled Reactor (HTGR) is designed with outlet temperatures ranging between 750°C and 800°C. These high outlet temperatures enhance the power production efficiency and facilitate a variety of industrial applications. The objective of this study is to understand the response of the primary system to potential transients in the secondary system. For this analysis, the transient condition originates in the Intermediate Heat Exchanger (IHX) or Steam Generator (SG) of the HTGR-integrated plant. The transients analysed are: a loss of pressure; loss of feedwater flow; inadvertent closure of main steam valve; decrease in returning gas temperature and heat load step change. The results show a large dependence on the negative reactivity added to the fuel as a function of increased temperature. The returning gas temperature decrease transient resulted in the highest fuel temperature (1361°C). Fuel temperature was shown to be less than the 1600°C fuel limit for each case analysed.

Safety analysis of a Super LWR with double tube water rods

A Super LWR with double tube water rods core had been designed for simplifying the upper core structure. Coolant flows upward in the fuel channels. Water inventory in the upper dome of the reactor pressure vessel was available for core cooling for the previous two-path core design, but not for the present one. Safety analysis was conducted to evaluate the safety performance of the present core design. Events of abnormal transients, accidents, and ATWS were included in the analysis. It is clarified that the water inventory in the double tube water rods can work as the coolant source in the event of total loss of feed water flow. The results of safety analysis show that the safety criteria are satisfied for all selected events.

The Development of Lungmen ABWR TRACE Safety Analysis Model

One of the main R & D work in nuclear engineering is the development of computer programs related to nuclear power plant (NPP) safety analysis. The advanced thermal hydraulic code named TRACE for NPP safety analysis is developing by U.S. NRC now. Lungmen NPP is the first ABWR NPP in Taiwan and still under construction. The safety analysis of NPP is very important work. In general, the description and results of the safety analysis for the transients are in the Final Safety Analysis Report (FSAR) of NPP. In this research, the development of the Lungmen NPP TRACE model for the safety analysis is performed. The transient analysis prediction of the TRACE model for Lungmen NPP is also performed. The loss of feedwater flow (LOFW) transient data and MSIV (Main Steamline Isolation Valve) closure direct scram transient data from FSAR are used to compare with the predicted results of the Lungmen NPP TRACE model. The compared results indicate that there are the similar trends of parameters between the TRACE model of Lungmen NPP and FSAR for the analysis of the LOFW transient and MSIV closure direct scram (MSIVCD) transient.

The development and assessment of TRACE model for Lungmen ABWR

Abstract For the nuclear power plant Lungmen (two blocks of Advanced Boiling Water Reactor (ABWR)) in Taiwan a plant model for the thermal hydraulic program TRACE (TRAC/RELAP Advanced Computational Engine) was developed. This model is and will be used for the simulation of normal and anomalous plant behaviour as well as for the simulation of incident scenarios. The presentation of this model is done in three steps: The first step is the development of a TRACE model of Lungmen nuclear power plant (NPP) which includes the vessel, the main steam lines and important control systems (such as the feedwater control system, recirculation flow control system, etc.). Key parameters were identified to refine the model further in the frame of a steady state analysis. The second step is the performance of TRACE transient analyses, such as MSIV closure direct scram (MSIVCD, MSIV = Main Steamline Isolation Valve) and loss of feedwater flow (LOFW). The above transient data of Final Safety Analysis Report (FSAR) are used to verify the Lungmen NPP TRACE model. The trends of their analysis results are roughly similar. It indicates that TRACE model is satisfying for the purpose of Lungmen NPP safety analyses. The third step is the prediction analysis of Lungmen NPP startup tests by using the TRACE model. The prediction analysis results of TRACE comply with the startup tests procedure criteria.

The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using Lungmen startup test data.

Dynamic Comparison of Three- and Four-Equation Reactor Core Models in a Full-Scope Power Plant Training Simulator

A comparative analysis of the dynamic behavior of a boiling water reactor in a full-scope power plant simulator for operator training is presented. Three- and four-equation reactor core models were used to examine three transients following tests described in acceptance test procedures: scram, loss of feedwater flow, and closure of main isolation valves. The three-equation model consists of water and steam mixture momentum, including mass and energy balances. The four-equation model is based on liquid and gas phase mass balances, together with a drift-flux approach for the analysis of phase separation. Analysis of the models showed that the scram transient was slightly different for three- and four-equation models. The drift-flux effects can explain such differences. Regarding the loss-of-feedwater transient, the predicted steam flow after scram is larger for the three-equation model. Finally, for the transient related to the closure of main steam isolation valves, the three-equation model provides slightly different results for the pressure change, which affects reactor level behavior.

Safety analysis of a supercritical pressure, light water cooled and moderated reactor with double tube water rods

The supercritical pressure, light water cooled and moderated reactor (SCLWR) has once-through cooling system. All feedwater which cools the reactor core flows to the turbines. This paper summarizing the safety analysis of the SCLWR with double tube water rods. The plant system is simple but no natural circulation is established at the loss of feedwater flow. The coolant inventory in the reactor pressure vessel is small. The coolant density coefficient is approximately twice as large as that of the BWR. A computer code (SPRAT) was developed to analyze SCLWR behavior against major accidents and transients at supercritical pressure. In loss of flow events such as loss of off-site power, the flow coast down time should be larger than 10 sec. for avoiding the deterioration in heat transfer. In the flow-excess event such as inadvertent start of the auxiliary feedwater pumps, the power increases approximately 25% by coolant density feedback. In the overpressurization transient such as generator load rejection, the power does not increase even if scram fails. This is because flow stagnantion raises coolant temperature and coolant density change at overpressurization is small in supercritical pressure. The reactivity-induced event such as control rod ejection, is not severe because of the small reactivity ingress. In the loss of coolant accident, the double tube water rods delay the reflood of the core. The core is heated up rapidly because of the small heat capacity and tight lattice pitch of the fuel rods. All analyzed accidents and transients satisfied the criteria, and the feasibility of the reactor was confirmed from the safety point of view.

Flow-Induced Accident and Transient Analyses of a Direct-Cycle, Light-Water-Cooled, Fast Breeder Reactor Operating at Supercritical Pressure.

This paper deals with the safety design of a light-water-cooled fast breeder reactor operating at supercritical pressure (SCFBR) hich has the advantages of high thermal efficiency and simple reactor system. A new computer code was developed and five important flow-induced events were analyzed for the 1,245 M We SCFBR with two coolant loops. A core flow rate must be maintained for the safety of once-through type plant. Thus this reactor has two high-pressure auxiliary feedwater, four low pressure coolant injection and automatic depressurization systems. They are activated by the emergency signals of the decrease in coolant flow rate. The feedwater pumps should have fly-wheels for prolonging the coast-down time. It has been found that the peak cladding surface temperature of fuel rods does not exceed the allowable temperature under the accidents of the total loss of feedwater flow and feedwater pump shaft seizure. The minimum deterioration heat flux ratio satisfies the criterion under the transients of the ...

Comparison of RETRAN and RELAP5 Models to Oyster Creek Loss of Feedwater Transient

The Oyster Creek Generating Station is a 1930-MW(thermal) boiling water reactor 2 plant. During the past year, a program to qualify the Oyster Creek RETRAN model against plant data was in effect at GPU Nuclear. As part of this program, a major transient that occurred on May 2, 1979, was chosen for analysis comparison. While operating at 100% power, a spurious high-pressure scram occurred, coupled with a simultaneous trip of the recirculation pumps. Other events resulted in a loss of feedwater flow and the inadvertent closure, by the operator, of the recirculation pump discharge valves, which limited recirculation flow to only five 0.0508-m (2-in.) bypass lines. The operator proceeded to isolate the vessel and use the emergency condensers for decay heat removal until feed flow was restored 45 min later. The plant RETRAN model was benchmarked against this transient for the first 45 min, using 39 volumes, 54 junctions, 25 heat conductors, and a bubble rise model for the separator/upper downcomer regions.The RETRAN results showed good agreement with plant data for downcomer level and dome pressure. The unique coupling between the downcomer and core zone liquid levels during the cyclic operation of the emergency condensers was simulated quite well. The use of the bubble rise model for the separator/upper downcomer, however, resulted in a higher dome pressure given by RETRAN, which is believed to be due to the 100% separation efficiency of the model as compared to the degraded separator efficiencies at off-optimum operating conditions. The fuel zone liquid level was an outstanding issue at the time where a conservative simple calculation showed that the core remained covered during the transient. The RETRAN model confirmed that, but also showed that the fuel zone liquid mass during the transient was more than that at steady state. The good agreement obtained against plant data verifies the adequacy of the RETRAN code and the Oyster Creek model for performing transient and accident analyses. Recently, a RELAP5 model has also provided a benchmark for the same transient, and a good comparison with RETRAN and plant data was obtained.