Development of Advanced Fire Zone Model Applicable to Fire PRA for Nuclear Power Plant

2020 ◽  
Author(s):  
Junghoon Ji ◽  
Koji Shirai ◽  
Koji Tasaka ◽  
Toshiko Udagawa
Keyword(s):  
Oxygen Concentration ◽  
Nuclear Power ◽  
Power Plants ◽  
Fire Risk ◽  
Zone Model ◽  
Radiant Heat Transfer ◽  
Test Results ◽  
Compartment Fire ◽  
Jet Flame ◽  
Fire Source

Abstract In implementing the fire PRA for nuclear power plants, a highly predictive fire model is required for more realistic fire scenarios and fire risk assessment. The fire simulation zone model BRI2002 developed in Japan has been continuously improved to allow analysis considering the characteristics of a compartment fire. In this study, a heat feedback phenomenon was introduced in BRI2002, in which combustion of a fire source can be accelerated by radiant heat transfer inside the compartment during a compartment fire. Not only the thermal radiation from the flame and smoke layers, but also radiation from the hot ceiling surface and the ceiling jet flame were considered when the flame impinges with the ceiling. In addition, in the zone model, the existing model for predicting the oxygen concentration in a compartment was improved so that the oxygen concentration could be predicted considering the vertical location of a fire source (height from the floor). The prediction results were verified by full-scale compartment fire test results. As a result of the calculation in which the fire source is installed at 2 m above the floor, the prediction results for the burning rate and zone temperature were well consistent with the test results.

Fiber-Optic Transmitter for Nuclear Power Plants

2004 ◽  
Author(s):  
Junichi Higashi ◽  
Shinichi Murakawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fiber Optic ◽  
Optic Fiber ◽  
Test Results ◽  
Maintenance System ◽  
Fiber Technology ◽  
Fabry Perot ◽  
Validation Tests

A promising Fiber-Optic Differential Pressure (DP) Transmitter is under development in Flexible Maintenance System (FMS) Projects that supported by Ministry of Economic, Trade, and Industries of Japan. The object of FMS projects is to improve maintenance works at nuclear power plants with latest technology. The new DP Transmitter uses optic-fiber technology of Extrinsic Fabry-Perot Sensor and Fizeau White-Light Cross-Correlator. Validation tests were performed to evaluate the tolerance of the DP transmitter in Nuclear Power Plant conditions. General requirements of PWR are accuracy (repeatability and linearity) of within +/−0.5%, pressure-proof of maximum 17.16MPa, Irradiation of 100Gy, and temperature range of 10–50 degrees centigrade at normal condition. The test results show the new DP transmitter can be expected as the next generation instrumentation in Nuclear Power Plants.

Fire Risk Analysis for Nuclear Power Plants: Methodological Developments and Applications

Risk Analysis ◽  
1985 ◽  
Vol 5 (1) ◽  
pp. 33-51 ◽  
Author(s):  
Mardyros Kazarians ◽  
Nathan O. Siu ◽  
George Apostolakis
Keyword(s):  
Risk Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fire Risk ◽  
Fire Risk Analysis

Fluid-Structure Dynamics With a Modal Hybrid Method

1992 ◽  
Vol 114 (2) ◽  
pp. 133-138 ◽  
Author(s):  
B. Brenneman ◽  
M. K. Au-Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Hybrid Approach ◽  
Dynamic Responses ◽  
Test Results ◽  
Strongly Coupled ◽  
Large Structures ◽  
Modal Synthesis ◽  
Loss Of Coolant ◽  
Thermal Shields

Large structures in nuclear power plants are often separated by very thin fluid-filled cavities. For example, core support structures, thermal shields, and reactor vessels are usually large concentric cylindrical shells with annuli between them as small as 2 percent of the shell diameter. Such thin cavities cause the structures to be very strongly coupled, and such coupling must be accurately modeled to predict the dynamic responses of new designs to turbulence, pump acoustic loading, loss-of-coolant accidents, and seismic events. This paper summarizes a very versatile and efficient method of solving these problems with small personal computers. Among other things, this method uses component modal synthesis with the hybrid approach, and the solution of the resulting unsymmetric eigenvalue problem for the coupled vibration modes. System responses are then found in terms of “right” and “left” eigenvectors. Comparisons with test results are also presented.

Pressure Tests for Thickness Management of Wall Thinning Tees

2012 ◽  
Author(s):  
Kaina Teshima ◽  
Yoichi Iwamoto ◽  
Kiminobu Hojo ◽  
Tomoyuki Oka ◽  
Kunihiro Kobayashi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Outer Diameter ◽  
Test Results ◽  
Minimum Thickness ◽  
T Joints ◽  
Wall Thinning ◽  
Pressure Tests ◽  
Design Pressure

Although the minimum thickness of pipe wall required (tsr) of T-joints (tees) of class 2, 3 and lower classes of nuclear power plants in Japan is calculated from the design pressure and temperature, there is no rule or standard of wall thinning T-joints for thickness management. This paper describes the pressure tests procedure and six test results with parameters of T-joint geometry such as outer diameter D, thickness T and T/D to establish structural integrity of wall thinning T-joints. Based on the fracture surface observation, a ductile crack initiation of each test mock-ups was confirmed.

scholarly journals Fire Risk Assessment of Transient Ignition Sources in Domestic Nuclear Power Plants

2021 ◽  
Vol 35 (3) ◽  
pp. 59-67
Author(s):  
Jung-Hyun Ryu
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Evaluation Method ◽  
Fire Risk ◽  
Ignition Source ◽  
Fire Risk Assessment

The fire risk of a nuclear power plant is evaluated using fixed and transient ignition sources. In terms of the overall fire risk, the proportion of transient ignition sources is very small. However, because the uncertainty due to the difference between the assumptions and the modeling method is relatively large, it is necessary to establish a methodology to address this. In this study, the new transient ignition source evaluation method presented in NUREG/CR-6850, the ignition source frequency revised in NUREG-2169, and the input parameters for transient fire modeling presented in NUREG-2233 were used to evaluate the fire risk assessment for transient ignition sources. In this new evaluation methodology, the fire ignition frequency is quantitatively evaluated based on the characteristics of the area, and an area-based scenario evaluation method considering the location of the transient ignition source is proposed for the evaluation within the area. As a result of applying the new methodology to the switchgear room of a reference nuclear power plant, an approximately 70% risk reduction was confirmed compared to the existing EPRI TR-105928 method. In the future, if fire risk assessment for transient ignition sources in nuclear power plants is applied using the results of this study, it is expected that areas whose control is important in the event of a fire can be determined, which should help reduce highly rated fire risks.

Ammonia Bottoming Cycle Development at Electricité de France for Nuclear Power Plants

1981 ◽  
Author(s):  
J. Fleury
Keyword(s):  
Power Plant ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Economic Conditions ◽  
Air Cooling ◽  
Test Results ◽  
Power Station ◽  
Major Benefit ◽  
Dry Cooling

An ammonia bottoming cycle is under active development at Electricité de France. To be implemented in a nuclear power plant downstream from the steam cycle, shortened for this application, its purpose is to make it possible to practice air cooling in satisfactory economic conditions. After an analysis of the main parameters of the bottoming cycle (H2O/NH3) (i.e., back pressure and temperature differences in the heat exchangers) its advantages are enumerated: in addition to those the dry cooling concept, the major benefit consists of the fact that the bottoming cycle makes use of low atmospheric temperatures in winter, producing a significant increase in the power output, just when it is most needed in many geographic locations. Emphasis is placed on the experimental work performed on E.D.F. test facilities and the construction of a 20-MWe demonstration bottoming cycle power plant at Gennevilliers power station. A brief account is given of test results and experimental programs.

Prediction of Radiation Embrittlement for Highly Irradiated Reactor Pressure Vessel Steels

2009 ◽  
Author(s):  
Hiroshi Matsuzawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Test Results ◽  
Prediction Formula ◽  
Irradiation Embrittlement ◽  
Nuclear Plants ◽  
Domestic Plant ◽  
Test Reactor ◽  
Reactor Pressure

There are 53 (fifty-three) nuclear power plants (both PWR and BWR type) are now under operating in Japan, and the oldest plant has been operating more than thirty years. These plants will be operated until sixty years for operation periods, and will be verified the integrity for assessment of nuclear plants for every ten years in Japan. Reactor Pressure Vessels (RPVs) are required to evaluate the reduction of fracture toughness and the increase of the reference temperature in the transition region. As the operating period will be longer, the prediction for these material properties will be more important. Recently the domestic prediction formula of embrittlement was revised based on the database of domestic plant surveillance test results for thirty years olds as the JEAC4201-2007 [7]. The adequacy for this prediction formula using for sixty year periods is verified by use of the results of the material test reactors (MTRs), but the effects of the accelerated irradiation on embrittlement has not been clear now. So, JNES started the national project, called as “PRE” project on 2005 in order to investigate how flux influences on the ΔRTNDT. In this project the RPV materials irradiated in the actual PWR plant have been re-irradiated in the OECD/Halden test reactor by several different fluxes up to the high fluence region, and the microstructual change for these materials will be investigated in order to make clear the cause of the irradiation embrittlement. In this paper the overall scheme of this project and the summary of the updated results will be presented.

Seismic Test Results of Air-Operated Valve Actuators for Nuclear Power Plants (Air-Operated Globe Valve (Cylinder Type))

2019 ◽  
Author(s):  
Ryo Kubota ◽  
Yoshitaka Tsutsumi ◽  
Yoshinao Matsubara ◽  
Shigeki Suzuki ◽  
Shin Kumagai
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Shaking Table ◽  
Test Results ◽  
Active Components ◽  
Seismic Testing ◽  
Globe Valve

Abstract It is believed that air-operated globe valves are able to operate during and after earthquakes, leading to maximum accelerations beyond the existing allowable acceleration for nuclear power plants in Japan (6 × 9.8 m/s2). In this work, this assumption is verified using a resonance shaking table for seismic testing at acceleration levels of 20 × 9.8 m/s2 (see Ref. [1]). Results show that the active components used in existing air-operated globe valve designs remain operable at 22 × 9.8 m/s2 (horizontal (X and Y) and vertical (Z) directions).

Selection of the Test Specimens for Seismic Tests of Air-Operated Valve Actuators for Nuclear Power Plants

2019 ◽  
Author(s):  
Nobuo Kojima ◽  
Yoshitaka Tsutsumi ◽  
Yoshinao Matsubara ◽  
Koji Nishino ◽  
Yasuyuki Ito ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Ground Motions ◽  
Test Results ◽  
Earthquake Ground Motions ◽  
Dynamic Function ◽  
Design Earthquake ◽  
Seismic Tests ◽  
Selection Of

Abstract The soundness for the function of air-operated valves in nuclear power plants during earthquake has been investigated via seismic test results and so forth. Since the seismic response acceleration has increased more and more with a recent reassessment of design earthquake ground motions conducted according to the revised Japanese nuclear safety regulation, it is necessary to evaluate the soundness for the function of various valves subject to large acceleration beyond design basis. The air-operated valves currently installed in the nuclear power plants in Japan play the important roles in the sever accident events. In this study, we classified them based on the valve type, manufactures and the previous test results. Furthermore, we proposed the strategy for evaluating the seismic-proof and the seismic test condition for examining the soundness of the dynamic function. Here, the dynamic function is defined as the function required under and after earthquakes.

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