Emergency Core Cooling System Injection Research Articles

Heat transfer characterization under radial jet and falling film induced rewetting

Abstract Empirical correlations are developed for rewetting velocity and maximum heat transfer coefficient during rewetting phase of single hot vertical Fuel Pin Simulator (FPS) by using radial jet impingement and falling film. Emergency Core Cooling System (ECCS) has been designed for Advance Heavy water Reactor (AHWR) to rewet the hot fuel pin under the loss of coolant accident. Coolant injection takes place from a water rod which is located at the center of the fuel bundle in form of jets to rewet hot surface of fuel pin under loss of coolant accident. This kind of design to reflood the fuel bundle is different than bottom and top spray reflooding practiced in PWR and BWR type of nuclear reactors. There are two different kinds of rewetting found during radial jet induced cooling. The first one is due to radial jet impingement and the second one is due to falling film which is below the jet impingement point. Rewetting velocity has been predicted along the length of fuel pin due to radial jet impingement cooling. Temperature of FPS has been varied from 400°C to 700°C with help of different powers supply, simulating decay heat of reactor. A variation of coolant radial jet mass flow rate is from 0.5 lpm to 1.8 lpm. It is considered during ECCS injection. It has been observed from the experiments that rewetting velocity decreases with increasing the clad surface temperature and increases with increasing the coolant mass flow rate. The rewetting velocity in falling film is found to be nearly 1.8 times higher than rewetting velocity predicted in circumferential direction. Further, it is found that maximum heat transfer coefficient increases with increasing the radial jet coolant mass flow rate. The maximum heat transfer coefficient in case of radial jet impingement is found to be nearly 1.5 times the falling film rewetting. Developed correlation predicts the maximum heat transfer coefficient with experimental data well within the error band of ±10%.

Suggestions for the Proper Treatment of Buoyancy Force Effect on Single Phase Thermal Stratification Phenomenon

In the operating nuclear power plants, piping and components installed in the in-service testing systems connected to the reactor coolant system are potentially exposed to fatigue caused by thermally stratified flow, possibly threatening its integrity. The emergency core cooling system (ECCS) is one of candidates for which the thermal stratification phenomenon can occur. Because temperature difference between the hot coolant in the cold leg and the cold cooling water injected from the ECCS line is 200 K or more, buoyancy due to density difference may have significant influence on thermal-hydraulic characteristics of the mixing flow. Therefore, in this study, the need to consider proper buoyancy models in governing equations, especially turbulence transport equations, for accurate prediction of single phase thermal stratification in the cold leg and downcomer by ECCS injection is numerically studied using ANSYS CFX R17.2.

Suggestion for the Proper Treatment of Buoyancy Force Effect on Single Phase Thermal Stratification Phenomenon

In the operating nuclear power plant, piping and components installed in the in-service testing (IST) systems connected to the reactor coolant system are potentially exposed to fatigue caused by thermal stratification flow, possibly threatening its integrity. The emergency core cooling system (ECCS) is one of candidates for which the thermal stratification phenomenon can occur. During the ECCS operating period, buoyancy due to density difference may have significant influence on thermal-hydraulic characteristics of the mixing flow. Therefore, in this study, the need to consider proper buoyancy models in governing equations, especially turbulent transport equations, for accurate prediction of single phase thermal stratification by ECCS injection was numerically studied using ANSYS CFX R17.2.

Boundary identification between LBLOCA and SBLOCA based on stratification and temperature gradient in two-phase PTS

Temperature gradient on the thick Reactor Pressure Vessel (RPV), caused by sudden overcooling events, especially in the downcomer, would intensify the propagation of structural defects. This situation known as Pressurized Thermal Shock (PTS) could be created in case of Emergency Core Cooling System (ECCS) actuation which leads to injection of cold water into the cold leg of the primary loop in some accidents, e.g. Loss Of Coolant Accident (LOCA). Prediction of Plant response to LOCA and water temperature gradient in the downcomer are performed in thermal-hydraulic section of PTS analysis. Employment of system codes is one of the proposed procedures in literature to obtain plant response and flow condition in the cold leg during LOCA. Also the results of these codes would be used to find the flow regime in the cold leg with some limitations. In this paper simulation of different break sizes in Bushehr Nuclear Power Plant as VVER-1000 reactor is performed by RELAP system code to find the temperature gradient and flow regime in the cold leg according to different criteria. Due to some limitations of system codes, CFX code is employed to evaluate turbulence characteristics at the interface for identification of flow regime. The comparison between results of different LOCA scenarios reveals a sharp reduction of water temperature in downcomer for large breaks which would be used for classification of LOCA. Also the flow regime in the cold leg during ECCS injection changes from stable stratified flow to wavy flow when the break size increases beyond a certain value. Therefore, the difference of temperature gradient in downcomer and flow regime in cold leg will be proposed as a new definition of Small Break LOCA (SBLOCA) and Large Break LOCA (LBLOCA) relevant to PTS analysis.

비상노심냉각계통 주입에 따른 저온관 및 강수관에서 단상 열성층 수치해석 : 부력항 고려 필요성에 관한 연구

비상노심냉각계통 주입에 따른 저온관 및 강수관에서 단상 열성층 수치해석 : 부력항 고려 필요성에 관한 연구

Radial jet induced rewetting study for heated rod

Establishment of rewetting of hot nuclear fuel pins under a pipe break event is an important design aspect of Nuclear Power Plant (NPP). The successful rewetting of heated pins through Emergency Core Cooling System (ECCS) injection stands to be a major requirement as this action mitigates the accident progression. In Advance Heavy Water Reactor (AHWR), ECCS is designed to inject water from a central water rod of the fuel bundle in form of jets to rewet hot surface of fuel pin. This kind of design to reflood the fuel bundle is different than bottom and top spray reflooding practiced in PWR and BWR type of nuclear reactors. The success of the proposed ECCS injection type has been assessed in a separate effect test where the study has been carried out with a simulated single fuel pin. Influence of issuing radial jet on circumferential and axial conduction and associated rewetting pattern has been studied. As ECCS injection may take place at different decay power levels and at different fuel initial temperatures hence experiments are conducted at different power levels (212W, 600W, 780W) and at different initial clad temperatures (400°C, 600°C, 660°C). The injection flow rate has been kept constant at a design flow rate of 1.8lpm for all the experiments. In all the experiments it is observed that a steep circumferential temperature gradient is established at the beginning of injection and the gradient comes down with the time advancement. A maximum gradient of 500°C is observed for a short period of time (∼4s) for the initial temperature of 660°C. The influence of variation of jet velocity arising from height of it’s injection is reflected on circumferential rewetting velocities which are found to vary from 2mm/s to 10mm/s. A correlation developed from experimental data has been proposed to estimate the average fuel temperature at any time instant during rewetting process. The study concludes that the design of multi point injection within the bundle will successfully be able to rewet the simulated fuel pin over the maximum period of time of ∼28.6s.

Large Break Loss-of-Coolant Accident Analysis of VVER-1000 Reactor Using CATHARE Code

In the safety analysis of nuclear power plants, large break loss-of-coolant accidents (LB LOCAs) continue to be one of the major issues. In this study, the latest version of the French best-estimate computer code CATHARE2 V2.5_1 Mod6.1 was used in order to predict the thermal-hydraulic phenomena in the VVER-1000 reactor during LB LOCA. This type of reactor is in operation and under construction in several countries including Russia, Ukraine, Bulgaria, Czech Republic, China, and India.The paper first presents the CATHARE modeling of a VVER-1000 reactor, including the core, the vessel, the primary and secondary circuits with the pressurizer, the main circulation pumps, the horizontal steam generators, and the steam lines. The emergency core cooling system (ECCS) is presented also with hydroaccumulators and high- and low-pressure safety injection pumps. The break, located in the cold leg and close to the reactor, is represented by the RUPTURE module of CATHARE, which modelizes a double-ended guillotine type of break.In the current calculations, the bottom-up and top-down reflooding play a very important role, compared to some western pressurized water reactors, because of the ECCS injection into the downcomer and upper plenum of the VVER-1000. That is why special attention is paid to the validation of the CATHARE reflooding model, based on the REWET-II experiment devoted to VVER. In CATHARE, a special two-dimensional reflooding module has been developed, which takes into account the radial and axial conduction in the vicinity of the quench front. The assessment based on the REWET-II facility shows the ability of the code to predict the progression of two simultaneous quench fronts.Then, the paper presents the results of the LB LOCA analysis for VVER-1000 (reference case and sensitivity study), describing the blowdown, refill, and reflood phases of the accident. Problems related to the countercurrent flow limitation (CCFL) phenomena at the core outlet are also considered. The results of the calculations show reasonable prediction of the basic parameters important to the safety of the plant, such as quench front progression, fuel-cladding temperatures, break flows, pressure behavior, etc. As expected, the temperatures—despite the conservative character of the modeling—remain below the safety criteria. It should be noted that the calculated fuel wall temperatures are very sensitive to the proper selection of the reflooding parameters and CCFL and need skillful choice of the reflooding conditions.

RISK INFORMED DECISION MAKING USING MULTIOBJECTIVE OPTIMIZATION TECHNIQUE

Nuclear Power Plant operations are guided by Limiting conditions of operations (LCO's) laid out in the document referred to as the Technical Specifications (TS). This Technical Specification is a legitimate framework approved by the Regulatory Bodies for the Safe operations of the Nuclear Power plants. In the past, the regulatory bodies used a deterministic approach as the basis for making decisions on safety issues and organizing the activities that they carry out. This was done by applying high level criteria such as the need to provide defence in depth and adequate safety margins. However with the availability of detailed Plant Specific Level-1 Probabilistic Safety Assessment (PSA), these limiting conditions need to be reviewed/revised based on the analysis results. This review of the LCO's is not a trivial exercise if the entire solution space of the variables defining the variables has to be investigated. This paper reviews the case for revision of Surveillance test frequencies of the Emergency Core Cooling System injection valves, using the multiobjective optimization technique.

Analyses of subcooled CCFL tests for evaluation of W̱COBRA/TRAC applicability

The W̱COBRA/TRAC-MOD7A, Rev. 1 code is currently licensed for best estimate large break LOCA analyses of 3 and 4 loop PWRs with emergency core cooling system injection located in the cold legs. As a part of the licensing effort to extend the code applicability to an upper plenum injection plant, the codes ability to predict subcooled flooding on a perforated plate was assessed by analyses of GE counter current flow limit tests and by comparison to the Bankoff flooding correlation (Bankoff, S.G., Tankin, R.S., Yuen, M.C., Hsieh, C., 1981). Counter current flow of air/water and steam/water through a horizontal perforated plate, Int. J. Heat Mass Transfer, 24 (8) 1382). The observed code model bias for subcooled CCFL can be eliminated by applying multiplication factors to the interfacial condensation and the interfacial drag models.

Fracture mechanics analyses including constraint investigations on reactor pressure vessels under pressurized thermal shock loading

In the integrity assessment of reactor pressure vessels the emergency core cooling system injection is one of the main load cases to be analysed. For a more generic assessment, parametric analyses are done concerning the load pattern, the postulated crack geometry and the material properties. With detailed axisymmetric and 3D-finite element (FE) models of reactor pressure vessels, temperature distributions in the structures and the deformations are calculated by elastic-plastic FE analyses for axisymmetric and asymmetric strip-like cooling assumptions. With it, boundary conditions for FE detail models with partly and 360° circumferential cracks are determined. Different material properties for the circumferential weld near the core mid-height, the cladding and the base metal with estimated influence of neutron irradiation are considered. Crack loading based on the J-integral is calculated for different crack geometries, and different parameters describing the constraint on the crack ligament with proposals for a fracture assessment concept are discussed. Furthermore, simplified methods based on analytical formulas to calculate the stress distribution in the wall and the stress intensity factor are applied and compared with FE results.

BWR Loss-of-Coolant Accident Tests at ROSA-III with High Temperature Emergency Core Coolant Injection

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.