Tetsuo Sawada

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.

Safety characteristics of the SCNES core

Abstract The core concept of the Self-Consistent Nuclear Energy System (SCNES) and its safety characteristics have been investigated from the view point of the elimination of recriticality. The recriticality potential can be eliminated based on characteristics of self-controllability to prevent the core damage and self-terminability to limit the propagation of core disruption. These two characteristics are simultaneously achieved by the radial heterogeneous two region core with different height. This core consists of leading and driver zones where hybrid metallic fuels with different melting point are installed. The self-controllability can be achieved by decreased coolant density effect due to the above core sodium plenum at the leading zone. The self-terminability is achieved by the Controlled Material Relocation (CMR), which is essentially the preceding downward in-pin fuel relocation selectively generated at the leading zone. U-Pu-1Zr alloy is used to the leading zone fuel due to lower melting point (900°C) than the driver fuel of U-Pu-10Zr(1100°C). Based on the quantitative investigations, it was emphasized that the recriticality potential can be eliminated by the in-pin fuel CMR even for severe unscrammed events such as a total pump stick for the primary coolant system and a total control rods withdrawal.

Inherent and Passive Safety Sodium-Cooled Fast Reactor Core Design with Minor Actinide and Fission Product Incineration

Abstract A self-consistent nuclear energy system (SCNES) can be a promising option as a future nuclear energy source. An SCNES should fulfill (a) efficient energy generation, (b) fuel production or breeding, (c) burning minor actinides with incinerating fission products, and (d) system safety. We focus on the system safety and present a simple evaluation model for the inherent and passive power stabilization capability of intact fast reactor cores under the conditions of an anticipated transient without scram (ATWS), i.e., self-controllability. The simple evaluation model is referred to as the “reactivity correlation model.” The model assesses self-controllability of a core based on the capabilities of reactivity feedbacks to stabilize transient power and maintain temperatures within predefined safety limits. Here the safety limits are “no fuel failure” and “nonboiling of coolant.” The reactivity correlation model was used to survey the self-controllability for metallic-fueled fast reactor cores. The survey was performed by selecting the core volume fractions of fuel, coolant, and structure; the arrangement of material compositions; and core configuration. A variety of reactor cores were examined, ranging from a standard 100-cm height to a flat 40-cm height. The effect of additions of sodium plena and channels, increased/decreased fuel volume fraction (Vf), loading 0 to 10 wt% minor actinides, and installing fission product-burning assemblies was also examined. The core performances were evaluated relative to tolerances against typical ATWSs, i.e., unprotected transient overpower and unprotected loss of flow. An optimum fast reactor core with the self-controllability as well as well-balanced tolerance against ATWSs resulted. The performance of this optimal core was examined for the other three prerequisites of a self-consistent nuclear energy system.

A Recriticality-Free Fast Reactor Core Concept

An outline is given of simple evaluation models for a recriticality in an attempt to construct a fast reactor core that has high potential to terminate an accident and prevent its progression, under postulated core-damage conditions, into further disruption of the degraded core and into possible recriticality leading to an energetic power excursion. The basic idea to prevent recriticality events is to remove a certain amount of fuel material out of the core in order to keep the core subcritical. Based on the simplified models, general guidelines are given that minimize the amount of fuel removal necessary to avoid recriticality events. Multigroup two-dimensional diffusion calculations are also performed to ascertain the tendency obtained by the simple model for the reactivity insertion due to a core collapse. In the sense of controlled material relocation, the fraction of core materials is identified that should be preferentially removed out of the core to eliminate the recriticality potential.

Calculation of a materials relocation experiment simulating a core discruptive accident condition in fast breeder reactors

This paper describes an interpretation of the SIMBATH (Simulationsexperimente in Brennelementattrapen mit Thermit) experiments that use the Simmer-II code. A series of SIMBATH experiments has aimed at simulating fuel pin disintegration and following materials relocation in the test sections of a single pin to 37-pin bundles. In the calculation, three modifications were incorporated into the Simmer-II code. With these modifications, the calculation showed good agreement with the experimental measurements with respect to the void region propagation in sodium flow and the molten materials relocation leading to flow blockage. A set of parametric calculations has clarified the range of applicability of parameters for materials relocation and flow blockage formation. The particle radius rp in blockage regions and the multiplier for particle viscosity (PARVIS) are recommended to be rp ⪆ built12Dh and 0.001 Pa s ⪅ PARVII ⪅ Pa s respectively.

Application of GEMSAFE to ITER CDA and its comparison with FER

A safety analysis for the design of International Thermonuclear Experimental Reactor (ITER) in the Conceptual Design Activity stage was performed by the GEMSAFE methodology, and its results were compared with those of Fusion Experimental Reactor (FER), a Japans facility planned next to JT-60. The objectives of this study are to confirm the applicability of GEMSAFE to ITER and to select design basis events of ITER and identify R&D items with comparison to FER. Function-Based Safety Analyses (FBSA) were carred out to select 19 and 25 design basis events for FER and ITER, respectively. The major reason for the difference is that ITER has a class-2 RI source, e.g., tritium of 7.5 × 105 Ci in mobile form, in the coolant for the first wall and blankets as well as a class-3 RI source, e.g., the immobile tritium of 2.2×107 Ci absorbed in first wall and dust.

On self-controllability and self-terminability of fast reactors

The paper gives an outline of simple evaluation models first for passive and inherent power stabilization of the intact fast reactor cores under the conditions of an anticipated transient without scram, and then for termination of the accident progression into further disruption of the degraded core and into a possible recriticality leading to a power excursion under postulated core damaged conditions. Based on the simplified models, general guidelines are given which enhance the passive and inherent power stabilization capability for intact cores, and those which minimize the amount of fuel removals necessary to avoid recriticality events. Also consistency between the self-controllability and self-terminability requirements is discusse.

Validation of a computational model for the evaluation of fuel-coolant interaction under severe accidental condition in fast breeder reactors

Abstract A computational model for fuel-coolant interaction has been validated through calculations for a series of THINA experiments. By the experiments, it was intended to simulate a comparatively massive injection of molten core materials into sodium pool under a core disruptive accidental condition assumed for fast breeder reactors. The calculations by the SIMMER-II code showed that the current models for the fuel-coolant interaction (FCI) had the capability to well reproduce the experimental results. This means that the SIMMER-II FCI model based on the liquid particles size and the heat transfer among liquid components is valid with proper accuracy for the estimation of a FCI even without more detailed mechanistic model for FCI. In particular, the thermal-to-mechanical energy conversion rates were well predicted.

Safety Analysis of ITER EDA Design by GEMSAFE

General Methodology of Safety Analysis and Evaluation for Fusion Systems (GEMSAFE) was applied to the International Thermonuclear Experimental Reactor (ITER) design in the stage of Engineering Design Activities (EDA) to identify Design Basis Events (DBEs) and the related safety features, which were compared with those of the ITER design in the stage of Conceptual Design Activities (CDA). As a result, 18 DBEs for the EDA design were selected in comparison with 25 DBEs for the CDA design. DBEs related to the fuel area were categorized in higher event category than those of the CDA design due to the increase of the mobile tritium contained in some components. It was necessary to reduce the inventory of the tritium absorbed in the tokamak dust in the EDA design as well as in the CDA design. Some measures were recommended to reduce mobile tritium dissolved in the coolant in the single cooling loop due to the increase of this estimated inventory.

Analysis of an out-of-pile experiment for materials redistribution under core disruptive accident condition of fast breeder reactors

Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration, the molten materials relocation and following materials redistribution that could occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of materials relocation within the identified ranges of the parameters which governed the blockages formation, i.e. the characteristic radius of solid particles jamming and/or sieving out in the flow and the effective particle viscosity. In particular the final materials redistribution calculated by SIMMER-II very well reproduced the experiment. This fact made it possible to interpret theoretically the mechanisms of flow blockages formation and related materials redistribution.