Hisashi Ninokata

Distributed resistance modeling of wire-wrapped rod bundles

Abstract Three-dimensional flow hydrodynamic distributed resistance models for rod bundles were developed. The models specifically account for the presence of the wire-wrap spacer and may be used for any lumped parameter thermohydraulic analysis numerical program. Validation studies of the hydrodynamic resistance models were also performed using a subchannel code ASFRE. The models were tested against subchannel velocity and temperature data taken from bundles of triangular rod array configurations with wire-spacers. Overall the models performed satisfactorily predicting the most important qualitative trends for flows in wire-wrapped rod bundles.

Biglobal linear stability analysis for the flow in eccentric annular channels and a related geometry

Recently, it has been observed that simple geometry characterized by a low level of symmetry present interesting peculiarities in the process of transition from laminar Poiseuille flow to turbulent flow. Examples of this type of geometry are eccentric channels and, more generally, parallel channels containing a narrow gap. In the present work, a global linear stability analysis for the flow in this class of geometry has been performed. The problem is discretized through spectral collocation and the eigenvalue problem has been solved with the Arnoldi-method based algorithms and the QZ algorithm. Since no numerical studies of this type have yet been performed to address the issue of transition in this geometry, the codes have been validated toward results obtained in simplified geometries (e.g., concentric annular channel and square channel). The eigenvalue spectra of the Poiseuille flow in eccentric channels and a U-shaped channel have then been computed and analyzed for a wide range of geometric parameter...

Gallium-cooled liquid metallic-fueled fast reactor

We have examined the effects on core characteristics of using two different types of Pu-based metallic alloy fuels in the gallium-cooled fast reactor core. In the proposed concept, the liquid metal fast nuclear reactor uses metallic fuel in the liquid phase and gallium coolant at high temperature (inlet 1700K, outlet 1900K). The liquid fuel is continuously supplied to the reactor during operation at full reactor power. The reactor power is controlled by rotational control drums with absorber material. The aim was to evaluate reactor core neutronics and safety characteristics demonstrating a feasibility of the reactor system. Although gallium has large absorption cross section in the high neutron energy region, we can design the core with rather good neutronics performances. The large negative reactivity feedback induced by the thermal expansion of liquid metallic fuel ensures the cores inherent safety against the unprotected loss-of-flow transient.

Nodal neutron kinetics model based on nonlinear iteration procedure for LWR analysis

A 3-dimensional neuron kinetics model based on the analytical nodal method and nonlinear iteration procedure is developed for Light Water Reactor (LWR) transient calculations. The solution procedure is decoupled on a local solution of the nodal equations for two-node problems and global iterations of the coarse-mesh finite-difference method. An orthogonality of the basic functions used for the neutron flux expansion results in an efficient algorith of the solution of the nodal equations for the two-node problem. The initial system of 8G nodal equations is reduced to a set of G and 2G equations, where G is a number of neutron energy groups. A fully implicit scheme with an analytical treatment of the delayed neutron precursors equations is used for time integration. An adaptive time-step size control procedure based on the time-step doubling technique is applied. The described numerical methods are implemented into the computer code SKETCH-N. The 3D LWR Langenbuch-Maures Werner (LMW) operational transient and 2D and 3D Boiling Water Reactor (BWR) LRA super-prompt-critical benchmark problems are calculated in order to verify the code. A comparison of the computed results with the solutions obtained by the other nodal computer codes demostrate fidelity and efficiency of the SKETCH-N code.

Evaluation Methods for Corrosion Damage of Components in Cooling Systems of Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics (I) : Major Targets and Development Strategies of the Evaluation Methods

Problems in major components and structural materials in nuclear power plants have often been caused by flow induced vibration and corrosion and their overlapping effects. In order to establish safe and reliable plant operation, future problems for structural materials should be predicted based on combined analyses of flow dynamics and corrosion and they should be mitigated before becoming serious issues for plant operation. Three approaches have been prepared for predicting future problems in structural materials: 1. Computer program packages for predicting future corrosion fatigue on structural materials, 2. Computer program packages for predicting future corrosion damage on structural materials, and 3. Computer program packages for predicting wall thinning caused by flow accelerated corrosion. General features of evaluation methods and their computer packages, technical innovations required for their development, and application plans for the developed approaches for plant operation are introduced in this paper.

Pioneering role of IRIS in the resurgence of Small Modular Reactors

Abstract This paper presents an overview of the first 10 years of the IRIS project, summarizing its main technical achievements and evaluating its impact on the resurgence of small modular reactors (SMRs). SMRs have been recurrently studied in the past, from early days of nuclear power, but have never gained sufficient traction to reach commercialization. This situation persisted also in the 1990s; the focus was on large reactors based on the presumed common wisdom of this being the only way to make the nuclear power plants competitive. IRIS is one of several small reactor concepts that originated in the late 1990s. However, the specific role and significance of IRIS is that it systematically pursued resolving technology gaps, addressing safety, licensing, and deployment issues and performing credible economics analyses, which ultimately made it possible—together with other SMR projects—to cross the “skepticism threshold” and led the making of a convincing case—domestically and internationally—for the role and viability of smaller reactors. Technologically, IRIS is associated with a number of novel design features that it either introduced or pursued more systematically than its predecessors and ultimately brought them to a new technical level. Some of these are discussed in this paper, such as the IRIS Safety-by-Design, security by design, the innovative thermodynamic coupling of its vessel and containment, systematic probabilistic risk assessment-guided design, approach to seismic design, approach to reduce the emergency planning zone to the site boundary, active involvement of academia, and so on. Many individuals and organizations contributed to that work, too many to list individually, and this paper attempts to pay tribute at least to their collective work.

Sabena: Subassembly boiling evolution numerical analysis

Abstract This paper describes the computer code SABENA that has been used in subassembly sodium boiling evolution numerical analysis as a contribution to fast breeder reactor safety analysis. SABENA is a two-fluid model subchannel code system to calculate coolant boiling and two-phase flow in a rod bundle together with external loop characteristics which affects the overall boiling behavior in the bundle section. With the use of relatively simple but reasonable constitutive models, the SABENA code has been applied to and validated against many multi-pin sodium boiling problems. The results have shown excellent agreement with the experiments. The numerical methods and models employed in the code have proven to be robust and efficient in light of the extreme severity of the conditions characterizing low-pressure sodium boiling.

Development of thermohydraulics computer programs for thermal striping phenomena

Thermal striping phenomena characterized by stationary random temperature fluctuations are observed in the region immediately above the core exit of liquid-metal-cooled fast reactors (LMFRs) due to the interactions of cold sodium flowing out of a control rod (C/R) assembly and hot sodium flowing out of adjacent fuel assemblies (F/As). Two thermohydraulics computer programs AQUA and DINUS-3, which are represented by both time- and volume-averaged transport analysis and direct numerical simulation of turbulence, respectively, were developed and validated for the evaluation of thermal striping phenomena. These codes were incorporated with higher order difference schemes to approximate the convection terms in conservation equations and adaptive time-step size control systems based on the fuzzy theory to eliminate numerical instabilities. From validation analyses with fundamental experiments in water and sodium, it was concluded that (a) thermal striping conditions such as spatial distributions of the intensity and the frequency of the fluid temperature fluctuations can be estimated efficiently by a combined approach incorporating the AQUA code and the DINUS-3 code, and (b) the thermal striping phenomena for the in-vessel components of actual liquid-metal-cooled fast reactors can be evaluated by the numerical method without conventional approaches such as large scale model experiments using sodium.

Numerical Study on Observed Decay Ratio of Coupled Neutronic-Thermal Hydraulic Instability in Ringhals Unit 1 under Random Noise Excitation

The core stability in the Ringhals Unit 1 was estimated under the numerical random noise that simulates indefinable two-phase flow noise sources in actual cores. This noise model is expressed as a product of band white amplitude and arbitrary shape functions. In evaluating decay ratios, the conventional free-decaying method based on a clean modal disturbance was replaced with the response analysis method based on a numerical moderator density noise. The stability monitoring procedure was reproduced numerically by giving the spatially random shaped noise disturbance and by linearly varying the moderator density reactivity multiplier. It was confirmed that the observed regional decay ratio based on differential LPRM signals proposed by Hagen sometimes shows a discontinuous jump from the stable to the unstable region as predicted by Pazsit. Nevertheless, the regional decay ratio based on the extracted modal response will show a continuous change under the same condition. It was clarified that this jumping is mainly induced by the local fluctuation of moderator density at the frequency range which overlaps with a predominance range of the fundamental mode. It was demonstrated that this kind of numerical noise analysis is useful in verifying the monitoring algorithm before applying it in actual plants.

IRIS: Proceeding Towards the Preliminary Design

The IRIS (International Reactor Innovative and Secure) project has completed the conceptual design phase and is moving towards completion of the preliminary design, scheduled for the end of 2002. Several other papers presented in this conference provide details on major aspects of the IRIS design. The three most innovative features which uniquely characterize IRIS are, in descending order of impact: 1. Safety-by-design, which takes maximum advantage of the integral configuration to eliminate from consideration some accidents, greatly lessen the consequence of other accident scenarios and decrease their probability of occurring; 2. Optimized maintenance, where the interval between maintenance shutdowns is extended to 48 months; and 3. Long core life, of at least four years without shuffling or partial refueling. Regarding feature 1, design and analyses will be supplemented by an extensive testing campaign to verify and demonstrate the performance of the integral components, individually as well as interactive systems. Test planning is being initiated. Test results will be factored into PRA analyses under an overall risk informed regulation approach, which is planned to be used in the IRIS licensing. Pre-application activities with NRC are also scheduled to start in mid 2002. Regarding feature 2, effort is being focused on advanced online diagnostics for the integral components, first of all the steam generators, which are the most critical component; several techniques are being investigated. Finally, a four year long life core design is well underway and some of the IRIS team members are examining higher enrichment, eight to ten year life cores which could be considered for reloads.Copyright