Japan Sodium-cooled Fast Reactor Research Articles

Numerical Simulation and Validation of Debris Bed Self-Leveling Behavior with Mixed-Density Particles Using CFD-DEM Coupling Algorithm

For the Japan Sodium-cooled Fast Reactor (JSFR), should the hypothesized core disruptive accident (CDA) happened, the in-vessel retention (IVR) will be the main target to achieve. In the heat-removal phase of the CDA, the debris bed will be piled up on the debris catcher. The capability of stable cooling and avoiding recriticality on the debris bed will be the main issues for achieving IVR. Previous studies have shown that the homogeneous debris bed can attain stable cooling and eliminate the probability of recriticality. Besides, self-leveling, which is a mechanism redistributing and flattening the debris bed by the natural circulation or vaporization from surrounding coolant, can further suppress the debris bed’s thickness to below the coolable thickness. However, in the real situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles start to redistribute due to self-leveling, the debris bed will form a heterogeneous density distribution. Under this scenario, the capability of coolability and the probability of recriticality could deviate from the previous study. Therefore, it is necessary to obtain a verified coupled model between the computational fluid dynamics (CFD) and the discrete element method (DEM) to track the mixed-density debris particles’ movement under the phenomenon of self-leveling. In this paper, first, the experiments simulating self-leveling on the mixed-density particle bed are performed. Afterward, the random heavy particle movement’s experimental data are extracted and transformed into the statistics form as the benchmark materials. Finally, the CFD-DEM model is validated via a series of sensitivity studies. The verified CFD-DEM can be expected to simulate the self-leveling behavior on the mixed-density debris bed and the real reactor case.

Safety Enhancement Approach Against External Hazard on JSFR Reactor Building

To respond to seismic and other natural hazard events, designers of the Japan Sodium-cooled Fast Reactor (JSFR), an advanced loop-type reactor, are planning to adopt a steel-plate reinforced concrete structure reactor building and an advanced seismic isolation system to strengthen this building. The design changes have been initiated by lessons learned from the Fukushima Daiichi nuclear power plant (Fukushima I NPP) accident. These enhancements to the design are to ensure that the JSFR structure can withstand external hazards and a severe accident. This evaluation and countermeasure study of external hazards and severe accident response by JSFR are based on the JSFR design before the Fukushima I NPP accident (2010 JSFR design). The method to evaluate the influence of external hazards on the JSFR design has applied Japanese codes and standards to show that the 2010 JSFR design of the building can withstand external hazards without additional countermeasures. And, for extreme conditions, few countermeasures are needed; however, the countermeasures considered have limited impact on the JSFR design because they do not significantly change the reactor building design. This paper gives a detailed evaluation of the countermeasures for the external hazards and severe accidents that could impact the JSFR building.

Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

In the design study of advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) called FA with inner duct structure (FAIDUS) is expected to enhance reactor safety during a core-disruptive accident. Evaluating the thermal-hydraulics in FAIDUS under various operating conditions is required for its design. This study is the first step toward confirming the design feasibility of FAIDUS; the thermal-hydraulics in FAIDUS are investigated with an in-house subchannel analysis code called asymmetrical flow in reactor elements (ASFRE), which can be applied to a wire-wrapped fuel pin bundle in conjunction with the distributed resistance model (DRM) and the turbulence-mixing model of the Todreas–Turi correlation model (TTM). Before simulating the thermal-hydraulics in FAIDUS, a few validations of DRM and TTM are conducted by comparing the numerical results of the pressure drop coefficients or temperature distribution obtained using ASFRE with the experimental data obtained using an apparatus with water or sodium for simulated FAs. After these validations, thermal-hydraulic analyses of FAIDUS and a typical FA are conducted for comparison. The numerical results indicate that, at a high flow rate simulating rated operation condition, no significant asymmetric temperature and velocity distribution occur in FAIDUS compared to the distribution in the typical FA. In addition, at a low flow rate simulating natural circulation condition in decay heat removal, the temperature distribution in FAIDUS is similar to that in the typical FA. This is because the local flow acceleration and the flow redistribution due to buoyancy force may occur in FAIDUS and the typical FA.

Water Experiments on Thermal Striping at the Bottom of Upper Internal Structure of Japan Sodium-cooled Fast Reactor

Water Experiments on Thermal Striping at the Bottom of Upper Internal Structure of Japan Sodium-cooled Fast Reactor

Core geometry for recriticality prevention against CDA in sodium-cooled fast reactor

In core disruptive accident (CDA) of sodium-cooled fast reactors (SFRs), fuel and steel can melt to form a molten pool. The objective of the present work is to investigate design measures for core of SFR to prevent recriticality. This paper studied core geometry of convex cross-section shape to increase neutron leakage from core surface in case of core compaction. In order to evaluate effective neutron multiplication factor, 70-group neutron diffusion calculations were performed by changing parameters related to the convex core geometry (i.e. ratio of inner height to inner diameter). Neutron diffusion code and nuclear data library employed in those calculations are CITATION and JENDL-3.3, respectively. By optimizing core configuration, the compaction reactivity of the convex shaped core showed drastic reduction from 2.07 % of reference cylindrical core (Japan Sodium-cooled Fast Reactor) to -3.28 %.

Development of a hybrid ECT sensor for JSFR SG double-wall tubes

ABSTRACTA new design has been adopted for the steam generator (SG) tubes of the Japan Sodium-cooled Fast Reactor (JSFR) using double-wall tubes. This paper estimates and assesses the effectiveness of detecting defects in SG double-wall tubes of the JSFR by using combined high-frequency eddy current testing (ECT) and low-frequency remote field eddy current sensors. We confirm that the proposed hybrid ECT sensor is highly sensitive to small defects, fatigue cracks, and other defects even when located under support plates of tubes. The parameters of the hybrid ECT sensor are designed and optimized to detect small defects using accurate numerical simulations based on the finite element method, using an in-house developed code. The sensitivity and high performance of the hybrid ECT sensor was validated with experimental measurements.

Secondary Sodium Fire Measures in JSFR

The Japan Sodium-cooled Fast Reactor (JSFR) adopts a double boundary for all sodium components. In this paper, design measures are investigated against a secondary sodium fire inside the reactor building, which might be assumed as design extension conditions. Candidate sodium fire measures for the secondary sodium systems compared in terms of safety are the sodium drain, nitrogen injection, pressure release valve, catch pan, and drain system for leaked sodium. Various sodium fires in the steam generator room have been analyzed by the SPHINCS code to evaluate the performance of the candidate sodium fire measures.

Evaluation of influence of the inlet swirling flow on the flow field in a triple elbow system

This study evaluated experimentally the influence of the swirling inflow condition upon the flow field in the cold leg piping, which comprises three 90 ° elbows, of the primary cooling system of Japan sodium-cooled fast reactor. Flow visualization experiment was conducted to reveal the velocity fields of the flow in the piping with swirling flow generator. When comparing to the developed turbulent inflow case, the flow separations in the intrados of the 1st elbow became smaller and the swirling flow downstream of the 2nd elbow became stronger, and there appeared separated regions downstream of the 2nd elbow which was not observed in developed turbulent inflow case. Since the flow separation can be considered as the source of the separated vortices which induce pressure fluctuations, these results imply the possibility that pressure fluctuations downstream of the 1st elbow are reduced and those of the 2nd elbow are enhanced by swirling flows as the inlet condition. The separated region is strongly reduced when a suitable swirling inflow enters the 3rd elbow.

Hazard curve evaluation method development for a forest fire as an external hazard on nuclear power plants

A method to obtain a hazard curve of a forest fire was developed. The method has four steps: a logic tree formulation, a response surface evaluation, a Monte Carlo simulation, and an annual exceedance frequency calculation. The logic tree consists domains of “forest fire breakout and spread conditions”, “weather conditions”, “vegetation conditions”, and “forest fire simulation conditions.” Condition parameters of the logic boxes are static if stable during a forest fire or not sensitive to a forest fire intensity, and non-static parameters are variables whose frequency/probability is given based on existing databases or evaluations. Response surfaces of a reaction intensity and a fireline intensity were prepared by interpolating outputs from a number of forest fire propagation simulations by fire area simulator (FARSITE). The Monte Carlo simulation was performed where one sample represented a set of variable parameters of the logic boxes and a corresponding intensity was evaluated from the response surface. The hazard curve, i.e. an annual exceedance frequency of the intensity, was therefore calculated from the histogram of the Monte Carlo simulation outputs. The new method was applied to evaluate hazard curves of a reaction intensity and a fireline intensity for a typical location around a sodium-cooled fast reactor in Japan.

Severe External Hazard on Hypothetical JSFR in 2010

Severe external hazards on the Japan Sodium-cooled Fast Reactor (JSFR) have been analyzed and evaluated. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. Integrity of the major components has been confirmed covering recent earthquake conditions. In the case of a tsunami, the seawater pumps for the component cooling water system (CCWS) could be damaged by the tsunami, since they are located at sea level. In the JSFR design with full natural convection decay heat removal systems (DHRSs) and an air-cooling emergency gas turbine, safety-grade components are independent of CCWS, and loss of CCWS does not affect reactor cooling. As a conservative case, hypothetical station blackout (SBO) has been evaluated. In the case of SBO, decay heat is removed by natural convection DHRS, but control of the air cooler (AC) damper is lost after the battery power is out. The analysis has revealed that freezing at one of three ACs could happen due to loss of automatic control of AC dampers. However, the time margin to protected loss of heat sink is evaluated to be >10 days. Manual control of the AC damper is also investigated. Transient analyses show that the AC dampers can be controlled manually adopting a simple operation procedure with sufficient operation time. Decay heat cooling in the case of collapse in all air stacks of AC has been evaluated. The result shows that decay heat could be removed maintaining air paths in two of three ACs by accident management. In conclusion, JSFR in the 2010 design version has enough external hazard toughness mainly thanks to passive safety features and a seismic isolation system.

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan sodium-cooled fast reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by the innerduct configurations were confirmed.

Thermal fatigue crack growth tests and analyses of thick wall cylinder made of Mod.9Cr–1Mo steel

In Japan, the basic designing works for a demonstration plant of Japan Sodium cooled Fast Reactor (JSFR) are now conducted. JSFR is an advanced loop type reactor concept. To enhance the safety and the economic competitiveness, JSFR employs modified 9% chromium–1% molybdenum (Mod.9Cr–1Mo) steel as a material for coolant pipes and components, because the steel has both excellent high temperature strength and thermal properties. The steel has been standardized as a nuclear material in Japan Society of Mechanical Engineers (JSME) code in 2012. In JSFR pipes, demonstration of Leak Before Break (LBB) aspect is strongly expected because the safety assessment may be performed on the premise of leak rate where the LBB aspect is assured. Although the authors have already performed a series of thermal fatigue crack growth tests of austenitic stainless steel cylinders (Wakai et al., 2005), crack growth behavior in the structures made of Mod.9Cr–1Mo steel has not been investigated yet. Especially for the welded joints of Mod.9Cr–1Mo steel, “Type-IV” cracking may occur at heat affected zone (HAZ). Therefore, this study performed a series of thermal fatigue crack growth tests of thick wall cylinders made of Mod.9Cr–1Mo steel including welds, to obtain the crack growth data under cyclic thermal transients. The test results were compared to the analytical results obtained from JAEA's simplified methods (Wakai et al., 2005).

Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant and descendant flows in the core channels suppresses the rise in the core temperature instead of the primary coolant loop. However, the test result suggests that the present DHRS has insufficient performance to remove the decay heat in the case of siphon break. Repeated tests have shown that the results of the water test have sufficiently high reproducibility in terms of the flow rate of the primary loop to be used for the verification of numerical codes.

Tagging-Gas Failed Fuel Detection and Location System for Sodium-Cooled Reactors

Failed fuel detection for the Japan Sodium-cooled Fast Reactor (JSFR) has been studied. The present JSFR design adopts a selector-valve (SV) failed fuel detection and location (FFDL) system. In this study, a tagging-gas (Tag) FFDL (Tag-FFDL) system has been investigated as an alternative. Although the identification performance of the Tag-FFDL system has been demonstrated in small and medium-sized reactors, the Tag-FFDL system has not been demonstrated yet in a large reactor like JSFR, which has 1500-MW(electric) power and 562 core fuel subassemblies. Major issues of the JSFR Tag-FFDL system are affected by high-burnup fuel and large cover gas volume. High-burnup fuel leads to a large change of the isotope ratio, which is important for the detection performance of the Tag-FFDL system. Since the cover gas volume in JSFR is larger than that in previous reactors, the tagging-gas concentration in the cover gas is lower than that in previous reactors. Thus, a requirement of the background value is more strict in JSFR. This study investigates whether two issues of the Tag-FFDL system for JSFR would be solvable. Tag gas isotope change in a high-burnup condition has been evaluated regarding transmutation and fission gas release. Taking into account tag gas isotope change due to the high-burnup conditions and large cover gas space, JSFR tagging gas has been designed. The investigation results showed that the JSFR FFDL system can provide an identification capability for 672 subassemblies, which is larger than the number of subassemblies in JSFR combining tagging-gas and burnup estimation. In addition, an allowable background concentration of natural Kr and Xe in the cover gas has been evaluated.

Selector-Valve Failed Fuel Detection and Location System for Japan Sodium-Cooled Fast Reactor

The Japan Sodium-cooled Fast Reactor (JSFR) is studied as an advanced loop-type sodium-cooled reactor. A selector-valve (SV) mechanism is adopted in the design of JSFR for its failed fuel detection and location (FFDL) system. JSFR has only two FFDL units for 562 core fuel subassemblies to reduce construction cost by decreasing the reactor vessel diameter. Consequently, one SV-FFDL unit must handle about 300 subassemblies. Because of the large number of subassemblies per unit, it is predicted that the total duration for measuring all the fuel subassemblies becomes long. In addition, JSFR adopts an upper internal structure (UIS) with a slit above the core. In order to detect the fission products from the subassemblies below the slit, additional sampling nozzles for the FFDL are set in the UIS around the slit. In previous water experiments and numerical simulation, the sampling performance for the subassemblies under the UIS slit has been evaluated to be lower than those under the normal UIS position. In this paper, the outline of the FFDL system is shown, which can be applied to a large number of fuel subassemblies in a compact reactor vessel. The detection capability of the FFDL system was studied to achieve the design conditions. Operation modes and procedures of the FFDL system were also investigated.

Evaluation of recriticality behavior in the material-relocation phase for Japan sodium-cooled fast reactor

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency has selected the advanced loop-type fast reactor, so-called Japan sodium-cooled fast reactor (JSFR). Through the evaluation of event progressions during hypothetical core-disruptive accident (CDA) under the design extension condition, a CDA scenario for JSFR has been evaluated. It has already been demonstrated that in-vessel retention (IVR) against CDA could be achieved by taking adequate design measures under best estimate conditions.The whole sequence of CDA scenario for JSFR was categorized into four phases according to the progress of core-disruption status. In the third phase, so-called material-relocation phase, the accident events would progress in the subcritical state. However, if the uncertainties about the molten state of core remnant and their discharge behavior outward from core are conservatively superposed, the disrupted core may lead up to recriticality.In the present study, the factors leading to recriticality in the material-relocation phase were investigated using the phenomenological diagrams, and the recriticality behaviors were evaluated through parametric analyses using SIMMER-III/IV codes. The results of parametric analyses suggested that a significant mechanical energy leading to the boundary failure of reactor vessel would not be released even assuming recriticality due to the uncertainties about molten state and discharge behavior. Through the present evaluation of the hypothetical recriticality event, the CDA scenario for JSFR could obtain further robustness from the viewpoint of achieving IVR.

Evaluation of External Event Effects on JSFR Fuel Handling System

In response to the accident at the Fukushima Daiichi (1F) nuclear power plant, designers of the 2010 version of the Japan sodium-cooled fast reactor (JSFR) have been studying the robustness of JSFR’s fuel handling system (FHS) against an earthquake and a tsunami. In the earthquake evaluation, the margin of seismic resistance and sloshing in the ex-vessel fuel storage tank (EVST) against an earthquake of the 1F-envelop condition were estimated. In the tsunami evaluation, for the case of loss of emergency power supply, a scenario is studied where fuel subassemblies are led to a stable cooling state, and potentialities for the cooling system are examined. As a result, it is shown that the original design of the JSFR FHS already had the potential to prevent the release of radioactive material. Additionally, some design improvements of the EVST cooling system are introduced.

GS0101-104 高速炉配管のLBB評価に用いる貫通時き裂長さ評価法の提案

This paper describes a through-wall crack length evaluation procedure applicable to Leak Before Break (LBB) assessment of Japan Sodium cooled Fast Reactor (JSFR) pipes made of Mod.9Cr-1Mo steel. In LBB assessment of JSFR pipes, it is required to calculate virtual through-wall crack length, though the crack growth is quite small under design condition. Generally, it is known that the through-wall crack length depends on loading condition, namely the load ratio between tensile and bending and that the length under pure bending load condition is largest. This study proposes a simplified method to evaluate the through-wall crack length both for axial and circumferential cracks as a function of load ratio and fatigue crack growth characteristics. Using the method, through-wall crack length can be predicted as far as we know the loading condition and material properties.

Development of an unstable failure analysis procedure considering change of compliance at a crack part of SFR pipes

This paper describes a failure analysis procedure used in Leak Before Break (LBB) assessment of large sodium pipes in Japan Sodium cooled Fast Reactor (JSFR) made of modified 9Cr–1Mo steel. Since the steel has small thermal expansion rate and excellent high temperature strength, a simplified piping layout can be achieved. However, the fracture toughness of the steel is relatively smaller than that of the austenitic stainless steel used for many former fast reactor (FR) pipes. For this reason, a sophisticated LBB assessment is required for the pipes made of modified 9Cr–1Mo steel. Since the internal pressure of the pipes of JSFR is small, the thermal load caused by thermal expansion is predominant. Supposing a through wall crack in such piping system, the stiffness of the crack part will decrease with crack extension. The rational failure analysis procedure, evaluating the loading conditions at a crack part considering change of compliance of the piping system with crack extension, was proposed. The loading conditions at a crack part are evaluated by a Finite Element Analysis (FEA) in which the stiffness of a cracked pipe is modeled (1), and critical crack size is estimated by the 2-parameter method (2). The stiffness of a crack part is evaluated using the critical crack size and loading condition (3), and the stiffness of the crack part is reflected into the FEA model (4). The critical crack size can be achieved reasonably by repeating from (1) to (4). Using some experimental results, the validity of the analysis procedure was verified. Finally, the critical crack size was calculated for designed JSFR pipes. It was clarified that the proposed procedure was much effective to loosen LBB conditions for JSFR piping system. For the verification of the method, calculated results were compared to some experimental results.

Design features and cost reduction potential of JSFR

To improve the economic competitiveness of the Japan Sodium-cooled Fast Reactor (JSFR), several innovative designs have been introduced, e.g. reduction of number of main cooling loop, shorter pipe arrangement by adopting thermally durable material, in fact high chromium ferrite steel, a compact reactor vessel (RV), integration of a primary pump and an intermediate heat exchanger (IHX). Since they had not been introduced in the past and existing reactors, a new approach for construction cost estimation has been introduced to handle innovative technologies, for example, concerning different kinds of material, fabrication processes of equipment etc. As results of JSFR construction cost estimations based on the new method and the latest conceptual JSFR design, economic goals of Generation IV nuclear energy systems can be achieved by expecting the following cost reduction effects: commodity reduction by adopting innovative design, an economy of scale by power generation increase, learning effect etc. It is well analyzed quantitatively that feasibility of innovative designs is essential for economic competitiveness of JSFR.