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[en] The objective of the present work is to assess the analysis capability of thermo-hydraulic code on the direct contact condensation in the core makeup tank (CMT) of passive high-pressure injection system (PHPIS) in the CARR Passive Reactor (CP-1300). RELAP5/MOD3.1 is chosen to evaluate the code predictability on the direct contact condensation in the CMT. It is found that the predictions of MOD3.1 are in better agreement with the experimental data than those of MOD3.0. From the nodalization study of the test section, the 1-node model shows better agreement with the experimental data than the multi-node models. RELAP5/MOD3.1 identifies the flow regime of the test section as vertical stratification. However, the flow regime observed in the experiment is the subsonic jet with the bubble having the vertical cone shape. To accurately predict the direct contact condensation in the CMT with RELAP5/MOD3.1, it is essential that a new set of the interfacial heat transfer coefficients and a new flow regime map for direct contact condensation in the CMT be developed. (author)
[en] 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
[en] The Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) design was based on ABB-Combustion Engineering's (ABB-CE's) standardized 3800 MWt Nuclear Steam Supply System (System 80). Despite the 25% core power difference between the two plants (3800 MWt for System 80, 2815 MWt for UCN 3 and 4), several design features have not been changed, especially the capacity of Safety Injection System (SIS). Focusing on this fact, the sensitivity study for UCN 3 and 4 SIS capacity on the behavior of the postulated small break Loss of Coolant Accident (LOCA) was performed to optimize the SIS capacity. Firstly, the sensitivity study for Safety Injection Tank (SIT) capacity on the postulated small break LOCA was performed by changing the size of SIT from 100% to 60% (100%, 85%, 75%, 60%, respectively) for the 0.5 ft2 (465 cm2) pump discharge leg break. Secondly, the sensitivity study of High Pressure Safety Injection (HPST) pump capacity on the behavior of small break LOCA was investigated by reducing its discharge flow rate of UCN 3 and 4 to 50% (100%, 85%, 75%, 70%, 60% and 50%, respectively) for the 0.05 ft2 (46.5 cm2) pump discharge leg break. Finally, the spectrum analysis was performed using the acceptable results of these two studies, i.e., the reduced SIT size and HPSI pump capacity to 70% of UCN 3 and 4 to retain sufficient margin for the small break LOCA. The results of this spectrum analysis show that the peak cladding temperature (PCT) and peak local cladding oxidation (PLO) for the limiting break size of 0.05 ft2 (46.5 cm2) pump discharge leg break were 1459 deg. F (793 deg. C) and 0.486% respectively, which are in compliance with the Emergency Core Cooling System (ECCS) acceptance criteria of 2200 deg. F (1204 deg. C) and 17% specified in 10 CFR 50.46. The results of this sensitivity study can be adapted to the optimal design of SIS for the future 2815 MWt Korean Standard Nuclear Power Plant and be contributed to the increase of economical gain
[en] By revising the ECCS licensing rules in 1989, the USNRC has allowed the use of 'best estimate' thermal-hydraulics computer codes (such as RELAP5, TRAC, and TRACE), with the requirement that uncertainty analysis accompany the results. Several methodologies have been developed for the quantification of the uncertainties of such codes. These methodologies are either input-driven or output-driven. They disagree in definition for the uncertainty range, qualification and quantification steps, types of uncertainty sources considered, methods of assignment of uncertainty distribution or range to various parameters, approach to propagation of uncertainty, and the way the dynamic characteristics of TH codes are handled. The IMTHUA methodology, developed by the author, is a hybrid approach where an input-driven 'white box' method is augmented with output correction based on experimental results relevant to code output. This paper offers a comparative assessment of uncertainty analysis methodologies for thermal-hydraulics transient calculations. The methods will be compared based on their approaches for treatment of input, propagation, and code models and correlations, as well as output. Comprehensiveness, approach to data treatment, and interpretation of results are among the criteria for comparison. Several examples are provided to clarify the differences.
[en] In the present study effectiveness of emergency core cooling system (ECCS) has been studied in case loss of coolant accident occurs at Pakistan research reactor (PARR-1). The reactor is a swimming pool type using MTR fuel. It was converted from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel in 1992. It was also upgraded from a steady-state power level of 5-10 MW. Several additional facilities were provided to satisfy the requirements of enhanced power level. For safety consideration, emergency core cooling system (ECCS) was also installed to avoid any possibility of core meltdown. Evaluation of ECCS has been carried out for which standard correlations have been employed to find peak clad temperature profile after loss of coolant accident
[en] In preparation for the possible transition to risk based licensing, it is increasingly important to demonstrate the applicability of Best Estimate codes to more extreme conditions corresponding for example to limited equipment availability. With this idea in mind, we have reviewed the application of TRAC-BF1 to Large and Small Break LOCAs. In this context this work describes the assessment and applications of the Penn State University (PSU) version 2001.2 of TRAC-BF1 with all the PSI updates, to the analysis of hypothetical Large Break (LB) LOCAs in a BWR/4 with postulated limited ECC availability. Since in contrast to a LB-LOCA in a BWR with full ECC availability, the rod surface temperatures reach relatively high values, additional assessment of the code under such conditions is required. Hence, after analyzing bottom flooding separate-effect experiments in two different heater rod bundles and a TLTA LB-LOCA test, we shall present and discuss the results of the analysis of a LB-LOCA with highly restricted Emergency Core Cooling flow in a BWR/4. In this context, we shall also assess different descretization of some terms of the 3D momentum equations, which was found to be important in the analysis of BWR Small Break LOCAs
[en] Highlights: • X-ray analysis was used to investigate fibrous debris beds on sump strainer modules. • Fibrous bed porosity was resolved in up to 12″ of water at 300 kV. • Addition of dense particulate does not significantly affect resolution of fiber pores. - Abstract: During the blow-down stage of a loss of coolant accident (LOCA), fibrous material and associated debris are transported from the location of the pipe break to the sump system as core coolant flows through containment. A series of strainer modules are installed to restrict large debris from entering the sump system, which supplies coolant to the Emergency Core Cooling System (ECCS). The accumulation of fibrous and debris material can result in the formation of semi-porous layers on the strainer module surfaces and possibly result in a reduction of coolant recirculation to the ECCS. The University of New Mexico (UNM) is investigating non-invasive techniques to better visualize the formation of fibrous debris beds on sump strainer modules; real-time visualization of debris bed formation would allow for better assessment of the filtration blockage of sump strainers. A series of exploratory tests were performed at the Center for Non-Destructive Evaluation (CNDE) at Iowa State University (ISU) to determine the viability of using an X-ray system to perform in situ measurements of debris bed formation and particulate filtration on small-scale representative sump pump strainer assemblies.
[en] Highlights: • LOCA is considered as initiating event for thermal-hydraulic analysis of PTS. • Temperature gradient and flow regime in cold leg depend on LOCA characteristics. • Identification of cold legs condition is needed to simulate interfacial phenomena. • The prediction of flow condition in cold leg would be obtained by system codes. • Flow regime and temperature gradient are used for classification of LOCA. - Abstract: 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.
[en] In this work, we shall report on the analysis of a number of postulated small breaks ranging from 1 to 10% of the recirculation line flow area at the Muehleberg (KKM) BWR/4 in Switzerland by using the TRAC-BF1 code. The analysis was performed by assuming both limited as well as full (nominal) emergency core cooling systems (ECCS) availability, and also limited availability of some of the safety relief valves used in the automatic depressurization system. Through these assumptions, we consider the response of the plant to multiple failures and therefore we extend our analysis beyond the normal 'design basis'. Furthermore, in order to provide a measure of the 'uncertainty' of the results, a sensitivity study was performed by varying some plant parameters as well as physical models in the code. To the best of our knowledge, this is the first time that such a systematic analysis of a wide spectrum of postulated small breaks in a commercial BWR has been pursued and the results of this work show that even under the most pessimistic assumptions on the availability of plant safety systems as well as of plant parameters and physical models in the code, the peak clad temperatures never exceed 1000 K
[en] As part of the task of assessing and qualify the frozen version of PennState University TRAC-BF1/v2001.2 as a BWR LOCA code, in this work, we shall study the extent to which changes in physical models in the code can affect the predicted rod surface temperature histories during separate-effect tests and system calculations. In particular, we shall concentrate our attention to the reflooding phase and after first demonstrating by two examples the necessity of using the code with the PSI updates, we shall implement in the code a dispersed flow interfacial shear model which is consistent with the corresponding interfacial heat transfer model, i.e. it utilizes the interfacial area per unit volume for both closure relations and not only for the interfacial heat transfer. For this, we shall implement in the code the dispersed flow interfacial shear model of RELAP5/MOD3; furthermore, we shall assume that the average droplet diameter is a function of the distance from the bottom quench front. Having done this, we shall analyze one FLECHT bottom flooding test with both code versions and show that the differences between measured and predicted PCTs are for both cases approximately the same. Finally, we shall study the effect of the number of nodes (8, 24 and 48) in the core by analyzing a hypothetical LB-LOCA in a BWR with limited ECC availability both with the frozen and the modified versions of the code and we shall show that when the frozen version of the code is used, the predicted peak clad temperatures (PCTs) are 120 K lower if 48 rather than when 8 nodes are used in the active core