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Ninokata, Hisashi*; Pellegrini, M.*; Kamide, Hideki; Ricotti, M.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13), Vol.2, p.151 - 166, 2015/04
The paper provides a cursory look at current approaches in numerical modeling and simulation of typical multi-physics phenomena of concern relevant to the sodium-cooled fast reactor design and safety. Emphasis is placed on the methods that are in practice and their verification and validation programs, including for those of fluid-structure thermal interactions due to thermal striping, thermodynamics of sodium-water chemical reactions, multi-component and multi-phase flows in the fuel degradation and core meltdown phases. Several of numerical simulations of these multi-physics phenomena are shown with verification and validation programs that employ not only separate-effect small-scale experiments of clean geometry but also for large-scale integral tests or mock-up experiments. The last part of this paper will be spent on discussions on more quantitative validation basis with identification of errors and/or uncertainties based on the Bayesian rule.
Toyooka, Junichi; Endo, Hiroshi*; Tobita, Yoshiharu; Ninokata, Hisashi*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 12(1), p.50 - 66, 2013/03
In the design of JSFR (Japan Sodium-cooled Fast Reactor), a design measure (FAIDUS: Fuel sub-Assembly with an Inner DUct Structure) is considered to prevent severe re-criticality events even in case of core disruptive accidents by molten fuel ejection out of the core region through the duct equipped within the fuel subassembly. Confirming principle effectiveness of such design measure is important. In this study, systematic heat transfer behavior of the ID1 test, which was conducted in IGR (Impulse Graphite Reactor) in Republic of Kazakhstan, was evaluated applying a heat conduction code TAC2D and a reactor safety analysis code SIMMER-III focusing on the clarification of heat transfer from high-temperature mixture of molten fuel and steel to the duct. As a result, the duct failure by high heat flux from the mixture was identified as one of an important mechanism of early duct failure in FAIDUS. It was also suggested from this study that the high heat flux from the mixture is caused by the direct contact of molten steel without the presence of fuel crust on the duct wall. Based on these findings, it is judged that the mechanism of early duct failure with high heat flux obtained in the ID1 test satisfies the required condition to FAIDUS, i.e., the inner duct of FAIDUS should fail at an early phase of core disruptive accident in advance to wrapper tube failure so that produced molten fuel can escape from the core region, and it supports feasibility of the FAIDUS concept.
Ninokata, Hisashi*; Kamide, Hideki
Nuclear Technology, 181(1), p.11 - 23, 2013/01
Times Cited Count：2 Percentile：17.66(Nuclear Science & Technology)In this paper key issues and highlighted topics in thermal hydraulics are discussed in the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., fully natural circulation decay heat removal and recriticality-free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities are briefly reviewed.
Ninokata, Hisashi*; Kamide, Hideki
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 20 Pages, 2011/09
Key issues in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. Design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., fully natural circulation decay heat removal and recriticality free core, have been investigated in order to achieve higher level of reactor safety. Preliminary evaluations are on-going. Here, progress of design study is introduced.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Kano, Takuma; Merzari, E.*; Ninokata, Hisashi*
Annual Report of the Earth Simulator Center April 2006 - March 2007, p.223 - 228, 2007/09
no abstracts in English
Kriventsev, V.; Ninokata, Hisashi; Yamaguchi, Akira; Ohshima, Hiroyuki
Journal of Fluid Mechanics, 0 Pages, 2003/12
A new model of turbulence is proposed for solving of Reynolds equations for fully-developed flow in a wall-bounded straight channelof arbitrary shape. The main idea of Multi-Scale Viscosity(MSV) model can be expressed in the followingphenomenological rule:A local deformation of axial velocity can generate the turbulence with theintensity that keeps the value of local turbulent Reynoldsnumber below some critical one. Therefore, in MSV, the only empirical parameter is the critical Reynolds number. The turbulent Revnolds numberis defined as Re = K/W, where K is kinetic energy and W is the work of friction/dissipation forces. MSV has been applied to the basic channel flows such as a circulartube, an infinitive plane chanael and an annulus. Calculated velocity profiles are in a good agreementwith expe
Ninokata, Hisashi*; Misawa, Takeharu*; Baglietto, E.*; Aoki, Takayuki*; Sorokin, A. P.*; Maekawa, Isamu*; Ohshima, Hiroyuki; Yamaguchi, Akira
JNC TY9400 2003-010, 170 Pages, 2003/03
JNC-TY9400-2003-010.pdf:5.42MB
A method of large scale direct numerical simulation of turbulent flows in a high burn-up fuel pin bundle is proposed to evaluate wall shear stress and temperature distributions on the pin surfaces as well as detailed coolant velocity and temperature distributions inside subchannels under various thermal hydraulic conditions. This simulation is aimed at providing a tool to confirm margins to thermal hydraulics design limits of the nuclear fuels and at the same time to be used in design-by-analysis approaches. The method will facilitate thermal hydraulic design of high performance LMFR core fuels characterized by high burn-up, ultra long life, high reliable and safe performances, easiness of operation and maintenance, minimization of radio active wastes, without much relying on such empirical approach as hot spot factor and sub-factors, and above all the high cost mock up experiments. A pseudo direct numerical simulation of turbulence (DNS) code is developed, first on the Cartesian coordinates and then on the curvilinear boundary fit coordinates that enables us to reproduce thermal hydraulics phenomena in such a complicated flow channel as subchannels in a nuclear fuel pin assembly. The coordinate transformation is evaluated and demonstrated to yield correct physical quantities by carrying out computations and comparisons with experimental data with respect to the distributions of various physical quantities and turbulence statistics for fluid flow and heat transfers in various kinds of simple flow channel geometry. Then the boundary fitted pseudo DNS for flows inside an infinite pin array configuration is carried out and compared with available detailed experimental data. In parallel similar calculations are carried out using a commercial code STAR-CD to cross-check the DNS performances. As a result, the pseudo DNS showed reasonable comparisons with experiments as well as the STAR-CD results. Importance of the secondary flow influences is emphasized on the momentum a
Ohshima, Hiroyuki; Kriventsev, V.; Yamaguchi, Akira; Ninokata, Hisashi
Journal of Nuclear Science and Technology, 40(9), p.655 - 663, 2003/00
Times Cited Count：9 Percentile：52.55(Nuclear Science & Technology)A new model of turbulence is proposed for the estimation of Reynolds stresses in turbulent fully-developed flow in a wall-bounded straight channel of an arbitrary shape. The main idea of a Multi-Scale Viscosity (MSV) model can be expressed in the following phenomenological rule: A local deformation of axial velocity can generate the turbulence with the intensity that keeps the value of the local turbulent Reynolds number below some critical one. Therefore, in MSV, the only empirical parameter is the critical Reynolds number. MSV has been verified on the pipe flow and applied to simulation of turbulence-driven secondary flow in elementary cell of the infinitive hexagonal rod array. Since MSV can predict turbulent viscosity anisotropy in directions normal and parallel to the wall, it is capable to calculate secondary flows in the cross-section of the rod bundle. Calculations have shown that maximal intensity of secondary flow is about 1% of the mean axial velocity for the low-Re flo
Kriventsev, V.; Ninokata, Hisashi; Yamaguchi, Akira; Ohshima, Hiroyuki
Proceedings of 10th International Conference on Nuclear Engineering (ICONE-10), 0 Pages, 2002/04
A new model of turbulence is proposed for the estimation of Reynolds stresses in turbulent fully-developed flow in a wall-bounded straight channel of an arbitrary shape. The main idea of a Multi-Scale Viscosity(MSV) model can be expressed in the following phenomenological rule:A local deformation of axial velocity can generate the turbulence with the intensity that keeps the value of the localturbulent Reynolds number below some critical one. Therefore,in MSV,the only empirical parameter is the critical Reynolds number. Since MSV can naturally predict turbulent viscosity anisotropy in directions normal and parallel to the wall,it is capable to calculate secondary flows in thecross-section of the rod bundle.
; Ninokata, Hisashi*; Okano, Yasushi; Yamaguchi, Akira
JNC TY9400 2001-010, 175 Pages, 2001/03
JNC-TY9400-2001-010.pdf:3.66MB
None
Ninokata, Hisashi; ; Okano, Yasushi
JNC TY9400 99-008, 31 Pages, 1999/03
no abstracts in English
Ohshima, Hiroyuki; ; Ninokata, Hisashi
Proceedings of 2nd International Topical Meeting on Advanced Reactors Safety, Vol.2, p.1157 - 1164, 1997/00
None
Otaka, Masahiko; Ohshima, Hiroyuki; Ninokata, Hisashi;
PNC TN9410 96-212, 36 Pages, 1996/06
A single phase subchannel analysis code ASFRE-III has been developed at PNC for predicting behavior of coolant and fuel pin temperature distributions in a fast reactor fuel subassembly under various operation and accident conditions such as a local flow blockage event. Salient features of the code are: a distributed resistance model of wire-wrap spacers, a porous blockage model, and an efficient matrix solver suitable for a large vector/parallel computation. In this study, ASFRE-III was applied to the thermal-hydraulic analysis of the two out-of-pile experiments using sodium performed at PNC for the purpose of the code validation. The one was performed around rated flow and heat flux conditions and the other was decay heat removal conditions. The computational results obtained under various flow and heat flux conditions were compared with the experimental data. The predicted coolant temperatures in subassemblies were agreed well with the measured data within 5 
6% in the wide range from low to high Reynolds number regions.
Kamide, Hideki; Ieda, Yoshiaki; Kobayashi, Jun; Ninokata, Hisashi
PNC TN9410 96-076, 72 Pages, 1996/03
A benchmark exercise for multi-dimensional thermohydralinc codes was carried out related to natural convection decay heat removal in liquid metal-cooled fast breeder reactors. A total of twelve computational methods were applied to the benchmark problem which simulated mixed forced and buoyancy driven penetration flow and thermal stratification phenomena. The applicability of turbulence models and higher order schemes of convection terms, was examined, and a combined method incorporating a higher order scheme and a turbulence model was found to be highly effective among the group of finite difference methods. The importance of turbulence models was also recognized for the finite element method. Development of a turbulence model applicable to the mixed convection flow regime was also discussed.
; Ninokata, Hisashi*
Nuclear Technology, (113), p.54 - 72, 1996/01
None
; Ninokata, Hisashi
Nuclear Engineering and Design, 150(1), p.81 - 93, 1994/09
Times Cited Count：11 Percentile：68.60(Nuclear Science & Technology)None
; Ninokata, Hisashi
International Journal for Numerical Methods in Engineering, 37(20), p.3397 - 3415, 1994/00
Times Cited Count：1 Percentile：34.14(Engineering, Multidisciplinary)None
Ninokata, Hisashi*;
Proceedings of 4th International Topical Meeting on Nuclear Thermal Hydraulics, Operating and safety, 0 Pages, 1994/00
None
Ohshima, Hiroyuki; Kamide, Hideki; ; Yamaguchi, Akira; Ninokata, Hisashi
IAEA INTERNATIONAL W/G FOR FAST REACTOR, 0 Pages, 1993/00
None
Ninokata, Hisashi; ;
International Conference on Design and Safety of Advanced Nuclear Power Plants (ANP '92), 3, 29.3-1 Pages, 1992/10
None
