Fortum Participation in IAEA Benchmark for KKNPP7 RHR Piping Response NCO 2007 Event

Author(s):  
Pentti Varpasuo ◽  
Jukka Ka¨hko¨nen

In this report Fortum participation in a benchmark related to the residual heat removal (RHR) piping system response of the Kashiwazaki-Kariwa Nuclear Power Plant unit 7 (KKNPP7) of Tokyo Electric Power Company (TEPCO) in Japan, during the 16 July 2007 Niigataken-Chuetsu-Oki earthquake (NCO) is described. The goal of this benchmark is to conduct a comparison between different analytical techniques, as used in the usual engineering practice. Equipment behavior during the NCO earthquake constitutes an extensive database. There are no direct acceleration or displacement measurements on equipment. For this reason quantitative simulation comparison is not possible. Qualitative observations of the fuel tank sloshing and tank buckling are available and these can be used in numerical simulation benchmark. The benchmark is divided in three phases and it will be carried out during the years of 2009–2011. The number of participants in the benchmark is 25 organizations and institutions. Detailed goals of the benchmark as a whole can be summarized as follows: 1) Understanding of soil, structures and mechanical equipment response during the Niigataken-Chuetsu-Oki earthquake. 2) Simulation of equipment response for residual heat removal equipment. 3) Simulation of liquid sloshing in the fuel pools in reactor building. 4) Simulation of buckling phenomena of vertical, cylindrical yard tanks. In order to carry out this simulation task a joint model of the reactor building and the residual heat removal-system is developed. The main results of the numerical simulation will be the maximum values of the stresses in the most critical locations of the residual heat removal piping system. These values will be compared to known material properties of the piping such as yield and the tensile failure strain.

Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.


Author(s):  
Haiqi Qin ◽  
Daogang Lu ◽  
Shengfei Wang

Practice has proved that nuclear power technology development and operation of nuclear power is a clean, safe and large-scale provided stable power. AP1000 uses a large number of passive safety technologies. Passive residual heat removal system is an important part, in the long-term cooling stage of nuclear reactor normal operating conditions or accident conditions, to prevent the core meltdown. The research of this paper is to solve the long-term discharge of residual heat of the containment in the accident condition of nuclear power plant. Based on the passive heat removal system of AP1000, combined with the heat transfer characteristics and advantages of heat pipes, the PRHR system is further improved on the basis of the present situation, and a conceptual design of passive containment residual heat removal system is proposed. In order to further verify the feasibility of the conceptual design, we make a simplified simulation of small containment test bench to carry out experimental verification and give a detailed experimental design.


2010 ◽  
Vol 171-172 ◽  
pp. 374-378
Author(s):  
Khan Salah Ud Din ◽  
Min Jun Peng ◽  
Muhammad Zubair

In this paper, a research has been carried out on the normal operational state of IPWR by using the thermal hydraulic system code Relap5/Mod3.4.In this study the conceptual study analysis of the reactor named as Inherent Safe Uranium Zirconium Hydride Nuclear Power Reactor INSURE-100 is considered but is only based on the conceptual study so the current research focuses on the normal operational of the reactor by using the Relap5/Mod3.4 code. For this purpose, two passive safety methods have been included for the safe operation as well as for transient analysis of the reactor. In the first one, Passive Residual Heat Removal System (PRHRS) has been modeled by taking the heat exchanger out of the Reactor Pressure Vessel (RPV) in the water tank so that it can absorb core decay heat in case of transient conditions and in the second one heat exchanger is placed both in inside and outside the RPV so that there can be another way to absorb the core residual heat. Considering these concepts figured out the normal operational state of the reactor by using Relap5/Mod3.4 in comparison with the conceptual design study of the reactor under consideration and the results extracted can be a good agreement for the transient analysis of the reactor.


Author(s):  
Eiji Shirai ◽  
Kazutoshi Eto ◽  
Akira Umemoto ◽  
Toshiaki Yoshii ◽  
Masami Kondo ◽  
...  

Seismic safety is one of the major key issues of nuclear power plant safety in Japan. It is demonstrated that nuclear piping possesses large safety margins in the various piping ultimate test reports. But it is appeared that there still remain some technical uncertainties about the phenomenon when both piping and supports show inelastic behavior in the extremely high seismic excitation level. In order to obtain the influence of the inelastic behavior of the support to the whole piping system response, and the subsequent interaction when both piping and supports show inelastic behavior, the following two tests have been started. • Support element test: Load-displacement characteristics of the support system including U-bolt, support itself and concrete anchorage are obtained by the forced displacement test. • Seismic proving test of piping system: The small bore piping and support system consisted of three dimensional piping, supports, U-bolts, and concrete anchorages will be excited on the table by the extremely higher seismic level. This paper introduces the major results of seismic proving test of piping and support system. The support element test results is presented in the paper of part 2, and the simulation analyses of these tests are summarized in the paper of part 3 [1, 2].


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.


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