Hasegawa K.,Japan Nuclear Energy Safety Organization |
Meshii T.,University of Fukui |
Scarth D.A.,Kinectrics Inc.
Journal of Pressure Vessel Technology, Transactions of the ASME | Year: 2011
One of the more common modes of degradation in power plant piping has been wall thinning due to erosion-corrosion or flow-accelerated corrosion. Extensive work has been performed to understand flow-accelerated corrosion mechanisms and develop fracture criteria of locally thinned pipes since the tragic events at Surry Unit 2 and Mihama Unit 3. A large number of tests have been performed on carbon steel pipes, elbows, and tees with local wall thinning. In addition, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code provides procedures in Code Case N-597-2 for the evaluation of wall thinning in pipes. This paper provides validation of the evaluation procedures in Code Case N-597-2 by comparing with the field rupture data and pipe burst test data. The allowable wall thinning from the Code Case N-597-2 procedures is shown to maintain adequate margins against rupture. © 2011 American Society of Mechanical Engineers.
Watanabe T.,University of Fukui |
Ishigaki M.,Japan Atomic Energy Agency |
Hirano M.,Japan Nuclear Energy Safety Organization
Annals of Nuclear Energy | Year: 2012
A long-term station blackout accident of a BWR is analysed using TRAC-BF1, and the calculated results are compared with the observed data at the unit 2 reactor of the Fukushima Daiichi nuclear power plant. The RCIC is assumed to be actuated, and the reactor pressure and the liquid level in the core are found to be stable and the dry well pressure increases gradually. The effects of leakage and heat release from the containment are discussed. The reactor pressure increases to the SRV set point and the liquid level and the dry well pressure decrease after the termination of RCIC. The calculated results are in good agreement with the observed data, and the thermal-hydraulic transient of unit 2 reactor is shown to be simulated well. © 2012 Elsevier Ltd. All rights reserved.
Yoshioka K.,Toshiba Corporation |
Ando Y.,Japan Nuclear Energy Safety Organization
Journal of Nuclear Science and Technology | Year: 2010
We have developed a deterministic group constant generation method based on the calculation results of a continuous energy Monte Carlo technique. This method features multigroup scattering matrix generation via a weight-to-flux ratio. We performed both diffusion and transport core calculations with this set of multigroup constants generated by the proposed method, which we then validated by both a comparison with a conventional method and a critical experiment analysis. The developed method is particularly useful for innovative fuel and future core designs as Monte Carlo calculations are applicable to any heavy material and the geometrical heterogeneity thereof. ©Atomic Energy Society of Japan.
Matsu'ura T.,Geological Survey of Japan |
Matsu'ura T.,Japan Nuclear Energy Safety Organization |
Miyagi I.,Geological Survey of Japan |
Furusawa A.,Geological Survey of Japan
Quaternary Research | Year: 2011
We detected late Pleistocene cummingtonite-bearing cryptotephras in loess deposits in NE Japan and correlated them with known tephras elsewhere by using major-element compositions of the cummingtonite. This is the first time cryptotephras have been identified by analysis of a crystal phase rather than glass shards. In central NE Japan, four cummingtonite-bearing tephras, the Ichihasama pumice, the Dokusawa tephra, the Naruko-Nisaka tephra, and the Adachi-Medeshima tephra, are present in late Pleistocene loess deposits. Because the cummingtonite chemistry of each tephra is different and characteristic, it is potentially a powerful tool for detecting and identifying cryptotephras. An unidentified cummingtonite-bearing cryptotephra previously reported to be present in the late Pleistocene loess deposits at Kesennuma (Pacific coast) did not correlate with any of the known cummingtonite-bearing tephras in central NE Japan, but instead with the Numazawa-Kanayama tephra (erupted from the Numazawa caldera, southern NE Japan), although Kesennuma is well beyond the previously reported area of the distribution of the Numazawa-Kanayama tephra. Three new cummingtonite-bearing cryptotephras in the mid and late Pleistocene loess deposits (estimated to be less than 82. ka, 100-200. ka, and ca. 250. ka) on the Isawa upland were also detected. © 2010 University of Washington.
Yamamoto T.,Japan Nuclear Energy Safety Organization
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2013
Based on radioactivity measurement of soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, radioactivity of Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Cs, Ba, La, Pu, Am, and Cm isotopes were compiled as radioactivity ratios to 137Cs. By exponentially fitting or averaging, the radioactivity ratios at the core shutdown were estimated. They were divided by those of the fuel of the core at the shutdown to obtain a deposited radioactivity fractions of the nuclides as relative values to 137Cs, which also correspond to deposition fractions of the elements as relative values to Cs. They were estimated to be orders of 10-4 to 10-3 for Sr, 10 -4 for Nb, 10-2 to 10-1 for Mo, 10-1 for Ag, 10-1 to 100 for Te, 100 for I, 10 -3 for Ba, 10-6 to 10-5 for Pu, 10-6 to 10-5 for Am, and 10-6 for Cm. The observed radioactivity ratios to 137Cs were compared with those obtained by severe accident analysis to assess the validation of the analysis. Copyright © 2013 by ASME.
Ueda Y.,Japan Nuclear Energy Safety Organization
Transactions of the Atomic Energy Society of Japan | Year: 2010
A special committee on "Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs) " was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for probabilistic safety assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to establishing quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g. events of criticality, leakage of molten glass, hydrogen explosion, boiling of radioactive solution and fire (including rapid decomposition of TBP complexes), resulting in release of radioactive materials to the environment. The results of the research were summarized in a series of six reports. This report aims to provide common backgrounds of the events studied in order to promote the understanding of the other five technical reports and shows overviews of abnormal events postulated in a reprocessing plant and their features. ©Atomic Energy Society of Japan.
Yamamoto T.,Japan Nuclear Energy Safety Organization
Journal of Nuclear Science and Technology | Year: 2012
Based on periodically performed radioactivity measurements on soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, activity ratios to 137Cs of fission product and heavy nuclides were obtained for Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Ba, La, Pu, Am, and Cm isotopes. By exponentially fitting or averaging, the activity ratios at the core shutdown were estimated. Using correlations of activity ratios of 134Cs to 137Cs, and 238Pu to the sum of 239Pu and 240Pu against fuel burnup, burnup of the fuel sourcing the deposited activity of the soil was estimated. The activity ratios to 137Cs of each nuclide on the deposited activity were divided by those calculated on the fuel at the shutdown to obtain the deposited activity fraction of each nuclide as a relative value to 137Cs, which also corresponds to the deposited fraction of each element as a relative value to Cs. The obtained deposited fractions relative to Cs are the orders of 10 -4 to 10 -2 for Sr, 10 -5 to 10 -3 for Nb, 10 -2 to 10 -1 for Mo, 1 to 10 for I, 10 -3 to 10 -2for Ba, 10 -2 for La, 10 -6 to 10 -3 for Pu, 10 -6 to 10 -4 for Am, and 10 -7 to 10 -5 for Cm. The deposited fractions for Tc, Ag, and Te were not estimated due to the lack of the calculated inventories in the fuel for the relevant measured radioactive nuclides. © 2012 Atomic Energy Society of Japan. All rights reserved.
Pellegrini M.,Tokyo Institute of Technology |
Endo H.,Japan Nuclear Energy Safety Organization |
Ninokata H.,Tokyo Institute of Technology
Progress in Nuclear Energy | Year: 2011
The correct evaluation of flows at transitional Reynolds number in nuclear reactors is gaining higher importance in relation to the accident analysis for buoyancy-driven flows which dominate the heat decay removal process. In the present paper a comparative study of different turbulence modeling and wall treatment for the evaluation of a fluid flow in transitional Reynolds number, is presented employing computational fluid dynamics (CFD). The relative performance of the models is assessed through benchmarking of fully developed pipe flow at Reynolds number 4900 and of a 90° bend pipe at Reynolds number 5000. Predictions of velocity profiles at different locations are compared to both experimental and accurate numerical simulations. It has been found that the predictions between the models can vary considerably in particular in relation to the different wall treatment employed on the wall. The results show the concerns about the employment of the available turbulence models and wall treatments in low Reynolds number flow regimes and explanation is provided in relation to their formulation. © 2011 Elsevier Ltd. All rights reserved.
Matsu'ura T.,Japan Nuclear Energy Safety Organization
Geomorphology | Year: 2015
Tectonic uplift rates across the Muroto Peninsula, in the southwest Japan forearc (the overriding plate in the southwest Japan oblique subduction zone), were estimated by mapping the elevations of the inner edges of marine terrace surfaces. The uplift rates inferred from marine terraces M1 and M2, which were correlated by tephrochronology with marine isotope stages (MIS) 5e and 5c, respectively, include some vertical offset by local faults but generally decrease northwestward from 1.2-1.6mky-1 on Cape Muroto to 0.3-0.7mky-1 in the Kochi Plain. The vertical deformation of the Muroto Peninsula since MIS 5e and 5c was interpreted as a combination of regional uplift and folding related to the arc-normal offshore Muroto-Misaki fault. A regional uplift rate of 0.46mky-1 was estimated from terraces on the Muroto Peninsula, and the residual deformation of these terraces was attributed to fault-related folding. A mass-balance calculation yielded a shortening rate of 0.71-0.77mky-1 for the Muroto Peninsula, with the Muroto-Misaki fault accounting for 0.60-0.71mky-1, but these rates may be overestimated by as much as 10% given variations of several meters in the elevation difference between the buried shoreline angles and terrace inner edges in the study area. A thrust fault model with flat (5-10° dip) and ramp (60° dip) components is proposed to explain the shortening rate and uplift rate of the Muroto-Misaki fault since MIS 5e. Bedrock deformation also indicates that the northern extension of this fault corresponds to the older Muroto Flexure. © 2015.
Hiroaki D.O.I.,Japan Nuclear Energy Safety Organization
Nihon Kikai Gakkai Ronbunshu, A Hen/Transactions of the Japan Society of Mechanical Engineers, Part A | Year: 2010
A developed crack propagation analysis system using the hnite element method tor components like vessels and pipings in nuclear power plants is described. One of the characteristic features of the system is that its input data include welding residual stresses and the stresses produced by applied forces in the components which have been obtained from previous analyses. In the system, the nodal forces on the crack surface are calculated from these input stresses. The authors extended the Virtual Crack Closure-Integral Method so as to calculate stress intensity factors when nodal forces on crack surface exist and applied the method to the system. In order to set test analysis problems for the system, conditions were found in which an elliptical crack holds the elliptical shape in fatigue crack propagation, and accordingly the crack size changes can be predicted using the theoretical stress intensity factors for an elliptical crack in infinite bodies subjected uniform tensile stresses. Under these conditions, a test analysis was carried out using the system, and the obtained crack size changes were shown to be in good agreement with those obtained from the theoretical stress intensity factors. As an example problem for practical structures, crack propagation due to SCC in a cylinder with a residual stress distribution solved by the system are presented and the results are compared with reliable reference values calculated using stress intensity factor data in a literature.