Validation of CATHARE V2.5 Thermal-Hydraulic Code Against Full-Scale PERSEO Tests for Decay Heat Removal in LWRs

Author(s):  
Giacomino Bandini ◽  
Paride Meloni ◽  
Massimiliano Polidori ◽  
Calogera Lombardo

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.

Author(s):  
F. Bianchi ◽  
P. Meloni ◽  
R. Ferri ◽  
A. Achilli

PERSEO device was developed in the framework of a domestic research program on innovative safety systems, with the purpose to increase the reliability of passive Decay Heat Removal Systems implementing in-pool heat exchangers. The device was tested at SIET Thermal-hydraulic Research Centre by modifying the existing PANTHERS IC-PCC facility. Two types of tests were performed: integral tests and stability tests. The experimental data acquired in the test campaign allowed a validation of a RELAP5/mod 3.3 beta release and CATHARE2 V1.5a/Mod8.1 full scale model of the PERSEO device. The paper deals with the comparison between the two codes against an integral test considered representative from the point of view of the PERSEO functioning and it highlights capabilities and limits of the codes in simulating such kind of test.


1940 ◽  
Vol 44 (352) ◽  
pp. 338-349
Author(s):  
A. P. West

During the past few years an extensive amount of experimental data on split flaps has been made available to the aircraft industry, through the publications of aeronautical research laboratories, both in this country and abroad. In general, each publication deals with one particular aspect of the problem, and when the effect of wing flaps on the performance of an aircraft is being estimated a certain amount of difficulty may be experienced in deciding which of the many reports available gives results most readily applicable to the case being considered ; and what allowances, if any, should be made for wing taper, flap cut-out, fuselage, etc.In this report the available data has been analysed with a view to answering these questions, and presented in such a form that it may be readily applied to determine the most probable change in the aerodynamic characteristics of a wing that may be expected from the use of this type of flap.From the appendix an estimate of the accuracy of the method can be obtained, as a comparison with full-scale data is given for lift and drag, and for the other flap characteristics the original curves have been reproduced.


Author(s):  
Seong Kuk Cho ◽  
Jekyoung Lee ◽  
Jeong Ik Lee ◽  
Jae Eun Cha

A Sodium-cooled Fast Reactor (SFR) has receiving attention as one of the promising next generation nuclear reactors because it can recycle the spent nuclear fuel produced from the current commercial nuclear reactors and accomplish higher thermal efficiency than the current commercial nuclear reactors. However, after shutdown of the nuclear reactor core, the accumulated fission products of the SFR also decay and release heat via radiation within the reactor. To remove this residual heat, a decay heat removal system (DHRS) with supercritical CO2 (S-CO2) as the working fluid is suggested with a turbocharger system which achieves passive operational capability. However, for designing this system an improved S-CO2 turbine design methodology should be suggested because the existing methodology for designing the S-CO2 Brayton cycle has focused only on the compressor design near the critical point. To develop a S-CO2 turbine design methodology, the non-dimensional number based design and the 1D mean line design method were modified and suggested. The design methodology was implemented into the developed code and the code results were compared with existing turbine experimental data. The data were collected under air and S-CO2 environment. The developed code in this research showed a reasonable agreement with the experimental data. Finally using the design code, the turbocharger design for the suggested DHRS and prediction of the off design performance were carried out. As further works, more effort will be put it to expand the S-CO2 turbine test data for validating the design code and methodology.


Author(s):  
Markus Esch ◽  
Bernd Ju¨rgens ◽  
Antonio Hurtado ◽  
Dietrich Knoche ◽  
Wolfgang Tietsch

In Germany two HTR nuclear power plants had been built and operated, the AVR-15 and the THTR-300. Also various projects for different purposes in a large power range had been developed. The AVR-15, an experimental reactor with a power output of 15 MWel was operated for more than 20 years with excellent results. The THTR-300 was designed as a prototype demonstration plant with 300 MWel and should be the technological basis for the entire future reactor line. The THTR-300 was prematurely shut down and decommissioned because of political reasons. But because of the accompanying comprehensive R&D program and the operation time of about 5 years, the technology was proved and essential operational results were gained. The AVR steam generator was installed above the reactor core. The six THTR heat exchangers were arranged circularly around the reactor core. Both heat exchanger systems have been operated successfully and furthermore acted as a residual heat removal system. The technology knowledge and experience gained on these existing HTR plants is still available at Westinghouse Electric Germany GmbH since Westinghouse is one of the legal successors of the former German HTR companies. As a follow-up project of THTR, the HTR-500 was developed and designed up to the manufacturing stage. For this plant additionally to the 8 steam generators, two residual heat removal heat exchangers were foreseen. These were to be installed in a ring around the reactor core. All these HTRs were designed for the generation of electricity using a steam cycle. Extensive research work has also been done for advanced applications of HTR technology e.g. using a direct cycle within the HHT project or generating process heat within the framework of the PNP project. Because of the critical attitude of the German government to the nuclear power in the past 20 years in Germany there was only a very limited interest in the further development of the HTR technology. As a consequence of the German decision, at the beginning of the 90s, to phase out nuclear power completely, research and funding of further development of HTR reactor design was also cut down. Today’s HTR reactor designs, such as the PBMR in South Africa, use a direct cycle with a gas turbine. This technology is also based on the THTR technology and PBMR is a licensed party. For the HTR-PM in China and the future oil sand projects powered by HTR’s in Canada and Siberia however the use of steam generators is required. Westinghouse and Dresden University cooperate in the field of steam generator technology for HTR reactors. The existing know-how for HTR is based on a huge pool of knowledge gained by the past German HTR projects mentioned above and consists especially of the design methodology, the mechanical layout and material issues for helium heated steam generators. The project team consists of experienced specialists who have worked on HTR projects in the past and of young graduate engineers. Main goal of the project is to analyze the existing know-how and to adjust it to the state of the art. As a first step, the existing design and its methodology is being analyzed and the different points of improvement are identified. The final step of the program is the description of a new methodology which fulfills the severe requirements of the customer and all of the actual licensing conditions. One of the reasons why this project has been launched is that the requirements of life expectancy for HTR components increase and the material limits will be reached, especially at high temperatures. This implies that the design of helix heat exchangers has to allow inservice inspections; this was not a requirement for the previous THTR design. Methodologies for in-service inspections already had been developed, but they are not sufficient for today’s tube lengths and have to be adapted. Another example, based on operating experience, is using reheaters to increase the efficiency is not recommended today. Using supercritical steam conditions to increase the efficiency should be investigated instead. In general, the economic benefit has to be balanced against the additional costs resulting from better material and more complex manufacturing.


Author(s):  
Andrea Bersano ◽  
Mario De Salve ◽  
Cristina Bertani ◽  
Nicolò Falcone ◽  
Bruno Panella

Within the field of research and development of innovative nuclear reactors, in particular Generation IV reactors and Small Modular Reactors (SMR), the design and the improvement of safety systems play a crucial role. Among all the safety systems high attention is dedicated to passive systems that do not need external energy to operate, with a very high reliability also in the case of station blackout, and which are largely used in evolutionary technology reactors. The aim of this work is the experimental and numerical analysis of a passive system that operates in natural circulation in order to study the mechanism and the efficiency of heat removal. The final goal is the development of a methodology that can be used to study this class of systems and to assess the thermal-hydraulic code RELAP5 for these specific applications. Starting from a commercial size system, which is the decay heat removal system of the experimental lead cooled reactor ALFRED, an experimental facility has been designed, built and tested with the aim of studying natural circulation in passive systems for nuclear applications. The facility has been simulated and optimized using the thermal-hydraulic code RELAP5-3D. During the experimental tests, temperatures and pressures are measured and the experimental results are compared with the ones predicted by the code. The results show that the system operates effectively, removing the given thermal power. The code can predict well the experimental results but high attention must be dedicated to the modeling of components where non-condensable gases are present (condenser pool and surrounding ambient). This facility will be also used to validate the scaling laws among systems that operate in natural circulation.


Author(s):  
Tanaka Go ◽  
Sato Takashi ◽  
Komori Yuji ◽  
Matsumoto Keiji

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.


Author(s):  
Ranjan Kumar ◽  
Pavel Kudinov ◽  
Sevostian Bechta ◽  
Florence Curnier ◽  
Michel Marques

Some critical safety systems exhibit the characteristics of hybrid stochastic class whose performance depends on the dynamic interactions of deterministic variables of physical phenomena and probabilistic variables of system failures. However, conventional probabilistic safety assessment (PSA) method involves static event and linked fault tree analysis and does not capture the dynamic interactions of such hybrid stochastic systems. Additionally, the existing dynamic PSA methods do not consider any repair possibility of some failed components during safety assessment. To address these issues, this paper presents a dynamic hybrid reliability assessment scheme for performance studies of repairable nuclear safety systems during a mission time. This scheme combines the features of reliability block diagram (RBD) for system compositions and partial differential equations for system physics using a customized stochastic hybrid automata tool implemented on Python platform. A case study of decay heat removal (DHR) systems has been performed using the introduced scheme. The impacts of failure rates and repair rates on sodium temperature evolution over a mission time have been analyzed. The results provide useful safety insights in mission safety tests of DHR systems. In sum, this work advances the dynamic safety assessment approach for complex system designs including nuclear power plants.


2011 ◽  
Vol 241 (12) ◽  
pp. 4662-4671 ◽  
Author(s):  
Giacomino Bandini ◽  
Paride Meloni ◽  
Massimiliano Polidori ◽  
Calogera Lombardo
Keyword(s):  

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