Super Fast Reactor Research Articles

Fuzzy Logic-Based Feedwater Controller for the Super Fast Reactor

Abstract Super fast reactor is a Generation IV reactor with water coolant at supercritical pressure. It has advantages of having high efficiency of about 44%, more economical and excellent safety features. Its last core design which adopted one heating in one coolant cycle, however, results a large density change along the fuel channels. It leads to a sensitive steam temperature change against a change of power which might mismatch with the flow change. The sensitivity should be reduced to maintain the structure integrity of the reactor components. A feedwater controller which based on fuzzy logic was used to reduce the sensitivity. Method of fuzzy-based control is applied. Two signals of the temperature and the power changes were taken as the inputs of the fuzzy controller, and change of the feedwater coolant flow was taken as the control output. Fuzzifications used triangular member functions with Tsukamoto fuzzy model for inferencing process. Two control parameters of member function spread and signal values which lead to fuzzy degree of 1 were optimized so that the criteria of outlet coolant temperature and the stability were fulfilled. Test results of the control system showed that the deviation of the water temperature at outlet side could be kept within the criteria when mismatch change of power and water flow happened. The temperature deviation could be supressed by using the fuzzy-based control.

Design Study of Small Modular Reactor Class Super Fast Reactor Core for In-Vessel Retention

Abstract The small modular reactor (SMR) class core design concept of once-through supercritical light water-cooled reactor (SCWR) with fast neutron spectrum (super fast reactor (FR)) is being developed at Waseda University. For the 300 MWel class design, complete core meltdown may need to be considered in case of a severe accident. This study proposes the new in-vessel retention (IVR) concept of the SMR class super FR (super FR-IVR), which can avoid recriticality even if the whole core relocates to the lower plenum of the reactor pressure vessel (RPV). The core characteristics with a given set of design specifications and criteria are evaluated based on fully coupled neutronics and thermal-hydraulics core burnup calculations. The debris criticality is evaluated based on Monte Carlo based method to consider the RPV lower plenum and debris configurations. The relationships between the 300 MWel class core design with the inner vessel diameter of 2.32 m and the IVR design are revealed. By reducing the operation cycle length from 720 days to 360 days and increasing the core inlet temperature from 280 °C to 370 °C, the required IVR submergence level could be reduced from 2.18 m to 1.38 m, assuming that the debris bed (melt pool) is homogeneous. However, when fully stratified debris configurations are assumed, the required disperser height and the corresponding IVR submergence level may increase to about 2.70 m.

Preliminary Core Design Study of Small Supercritical Fast Reactor with Single-Pass Cooling

A supercritical water-cooled reactor (SCWR) adopts a once-through direct cycle, which is compatible with a small modular reactor class (SMR) plant system. The core is cooled by supercritical light water, which does not exhibit phase change, but undergoes large temperature and density changes. A super fast reactor (Super FR) is a fast reactor type concept of SCWR. Unlike other SCWR core concepts, it adopts the single coolant pass flow scheme, in which the coolant passes the core only once from the bottom to the top without any reverse flows or preheating stages. In the meantime, reducing the core size tends to increase the core power peaking and reduce criticality. Therefore, the key issues with the small Super FR core design is reducing the core power peaking and achieving high average core outlet temperature with the single coolant pass scheme. This study aims to highlight the design issues through conceptual core designs of SMR class Super FR. To evaluate the core characteristics, three-dimensional coupled core calculations are carried out. The proposed design with small fuel assemblies, which are equivalent to those of boiling water reactors, attains a high core average outlet temperature of about 500 °C, which is compatible to that of typical large SCWR core design.

Control System of a Super Fast Reactor with Single Flow Pass Core

The one-pass supercritical pressure light water fast reactor has different core structure from that of the two-pass cooling type one. The distance between inlet and outlet channels of the one-pass supercritical pressure light water fast reactor is shorter with less core water inventory due to the removal of the gap channels. Its responses to a perturbation are similar to those of the two pass core. However, change of the outlet temperature regarding to a perturbation of power to flow ratio in the one-pass cooling type fast reactor is larger. It is important to suppress the change as small as possible to keep the integrity of the reactor structure. Early designs of power and pressure control systems of a supercritical pressure light water fast reactor are applicable to control the power and pressure parameters, but it is not for the early outlet temperature control system which considered only the outlet temperature deviation as the feedback. An improved outlet temperature control system which considers both the signals of outlet temperature and power deviations as the control feedbacks can be applied in the one-pass cooling-type fast reactor with adjustment on the control’s parameters. Optimized control’s parameters can suppress the change of the outlet temperature significantly without oscillation. By using the control system, responses of the one-pass supercritical pressure light water fast reactor to the designated perturbations satisfy the criteria.

A novel process maximizing energy conservation potential of biological treatment: Super fast membrane bioreactor

The study evaluated the microbial behavior of a submerged super fast membrane reactor (SFMBR) with settled sewage and a soluble synthetic substrate (peptone mixture) exhibiting similar biodegradation characteristics with sewage. A laboratory scale SFMBR was run at very low sludge retention times of 0.5–2.0d and a hydraulic retention time of 8h. Effluent COD was always lower than the soluble COD in the reactor and remained in the range 14–28mg/L for the peptone mixture and 36–44mg/L for settled sewage. Significant characteristics of SFMBR performance were assessed by particle size distribution analysis; respirometry; modeling defining related process kinetics; and molecular analysis, which revealed changes in microbial community composition under different operating conditions. Model simulation was also performed for raw sewage for revealing COD fractionation in the permeate and biomass in the reactor for the same operating conditions. The results indicated, as predicted, partial removal of the particulate slowly biodegradable COD fraction while soluble biodegradable COD components were almost totally removed. Modeling also highlighted energy conservation and recovery as the major feature of SFMBR, which was assessed to vary between 54% and 77%, a range significantly higher than what can be achieved with different activated sludge alternatives.

CFD analysis of heat transfer in hexagonal subchannels of super fast reactor in upward flow

AbstractA super fast reactor is a fast spectrum, supercritical, water‐cooled reactor. This paper represents CFD analysis of heat transfer in hexagonal subchannels of super fast reactor using FLUENT in ANSYS. The numerical simulation of grid stability was done by considering different mesh sizes and the turbulence model for heat transfer of supercritical water was also carried out and compared with the experimental data. RNG k‐ϵ turbulence model with enhanced wall treatment was considered for simulations. Heat transfer and heat generation rate analysis of the outer surface rod wall is carried out with different subchannels by changing various parameters like boundary conditions and pitch‐to‐diameter ratio. The analyses reveal that the outer surface of the rod wall temperature decreases with increase in pitch‐to‐diameter ratio. Maximum coolant temperature rises in edge subchannels more than corner subchannels. Further analysis is carried out with different mass fluxes. Increases in mass flux has minimal effect on the maximum rod wall surface temperature. Maximum cladding surface temperature for the corner subchannel is less compared to the edge subchannel. Heat generation rate also decreases with increase in pitch‐to‐diameter ratio. This paper also investigates the buoyancy effect on subchannels with varying heat flux as boundary conditions considering constant mass flux.

Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900°C. (2) Maximum cladding stress in circumferential direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be less than 1.0. Unfortunately some criteria could not be satisfied, such as the maximum fuel centerline temperature and maximum cladding stress. Therefore the influence of average linear heat generation rate, Pu enrichment and compressive stress to yield strength ratio are analyzed. Finally the improved fuel rod design of the new core is proposed.

Passive safety system of a super fast reactor

Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

Single pass core design for a Super Fast Reactor

Single pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward coolant flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern significantly simplifies the core concept in terms of upper core structure, coolant flow scheme as well as refueling procedure. In single pass core design, supercritical-pressure water at approximately 25.0MPa enters the core at 280°C and heated up in one through upward flow to the average outlet temperature at 500°C. Great coolant density change in vertical direction will cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO2 also introduces great power increase in blanket assemblies during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of blanket assemblies to mitigate the power change in blanket assembly and to increase the coolant outlet. With exclusive refueling and shuffling scheme as well as partial length control rod design, neutronics & TH coupled calculation shows satisfactory results that can fulfil the requirement of design for both 500°C outlet temperature and negative coolant void reactivity.

Analysis of anticipated transient without scram of a Super Fast Reactor with single flow pass core

Anticipated transient without scram (ATWS) of the supercritical-pressure light water-cooled fast reactor (Super FR) with single flow pass core is analyzed to clarify its safety characteristics. Abnormal reactivity due to loss of feedwater heating and uncontrolled withdrawal of a control rod is mitigated by density and Doppler reactivity feedback. Increase of the coolant density at the upstream of the fuel channel leads the core flow rate to decrease which in turn decreases the power especially due to the density reactivity feedback. Meanwhile, ATWS accidents which are followed by trip of both RCPs are important. Decreased feedwater flow rate leads directly to a decrease of core flow rate in all fuel assemblies. If the ADS is not used, the MCST increases rapidly due to high power density feature of the Super FR. Coolant expansion effect does not appear in the single flow pass core of Super FR while it appears in the second pass of the two-flow pass core leading to mild increase of MCST of the second pass. However, signal of flow low level 3 (6%) is provided for the ADS actuation and the MCST of each fuel channel is effective to be mitigated due to large core coolant flow rate induced by the ADS operation. The power is also decreased by density and Doppler reactivity feedbacks. MCST and core pressure of all the analyzed events satisfy the criteria.

Accidents and transients analyses of a super fast reactor with single flow pass core

The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST.

Improvements of two-pass core design for super fast reactor

Improvements of two-pass core design of Super critical-water cooled fast reactor (Super FR) are proposed. The core height has been shortened; assembly pitch is enlarged to take Control Rod (CR) design into account. Well-coupled fuel loading pattern is proposed, number of control rods location are limited only in second flow pass assemblies due to current flow pattern. Fuel shuffling is also confined within each own flow pass because of the different structure of assemblies. CR withdrawal is considered during the operation, satisfactory radial and axial power distribution can be achieved by implementing appropriate CR withdrawal strategy. Results suggest that all the design target, criteria and limits are satisfied in terms of average coolant outlet temperature, maximum linear heat generation rate (MLHGR), maximum cladding surface temperature (MCST) as well as core shutdown margin, negative coolant void reactivity coefficient and positive coolant density reactivity coefficient in all coolant density range.

Core design for super fast reactor with all upward flow core cooling

A 1000MWe level supercritical-pressure light water-cooled fast reactor (Super FR) is proposed in this paper. The newly improved features include two flow paths which both applies upward flow pattern. Compact lattice of MOX fuel pins are arranged in hexagonal seed assembly while blanket assembly comprised of depleted UO2 fuel pins wrapped with solid moderator ZrH layer for negative void reactivity. This core design includes 162 seed assemblies and 73 blanket assemblies with 1.7cm of solid moderation layer of ZrH, the core active height is 3.6m. The axial coolant density distribution of the core is far from uniform due to the all upward flow pattern, however, by applying axial Pu enrichment zoning and new loading and refueling scheme, flattened power distribution is obtained and all the design criteria as well as limitations have been fulfilled with significant delivery of high power density up to 237kw/L, 500°C average core outlet temperature, both whole core and local negative void reactivity.

LOCA Analysis of Super Fast Reactor

This paper describes loss of coolant accident (LOCA) analyses of the Supercritical-pressure Water-Cooled Fast Reactor (Super Fast Reactor). The features of the Super Fast Reactor are high power density and downward flow cooled fuel channels for the improvement of the economic potential of the Super Fast Reactor with high outlet steam temperature. The LOCA induces large pressure and coolant density change in the core. This change influences the flow distribution among the downward flow parallel channels. It will affect the safety of the Super Fast Reactor. LOCA analysis of Super Fast Reactor is important to understand the safety features of the Super Fast Reactor. Keeping the flow rate in the core is important for the safety of the Super Fast Reactor. In LOCA, it is difficult to maintain an adequate flow rate due to the once-through coolant cycle and the downward flow cooled fuel assemblies. Therefore, the early actuation of the Automatic Depressurization System (ADS) and reduction of the maximum linear heat generation rates of the downward flow seed fuel assemblies and Low-Pressure Core Spray (LPCS) system are necessary for the Super Fast Reactor to cool the core under LOCA. Analysis results show that the Super Fast Reactor can satisfy the safety criteria with these systems.

Numerical Analysis on Thermal-Hydraulics of Supercritical Water Flowing in a Tight-Lattice Fuel Bundle

To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of a supercritical water reactor (Super Fast Reactor), a simplified 19-rod fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D which has been enhanced by Japan Atomic Energy Agency. In ACE-3D, a two-phase flow turbulent model based on the k- model was adopted. In this calculation, a one-twelfth model is adopted as the computational domain and takes advantage of symmetry. As boundary conditions, mass velocity, inlet enthalpy and power per rod are to be the same as the steady state condition of the Super Fast Reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. The effect of grid spacers is taken into account in the analysis. Calculated rod surface temperatures take peak values near the top of the rods. The maximum cladding surface temperature (MCST) is observed at the position facing the narrowest gap on the center rod near the outlet and the value is 901 K (628 °C). It was confirmed that the predicted MCST satisfies the thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650 °C.

CFD analyses in tight-lattice subchannels and seven-rod bundle geometries of a Super Fast Reactor

This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.

ICONE19-43702 Super fast reactor R&D projects in Japan

The Japanese research project of the "Research and Development of Super Fast Reactor" was conducted from December 2005 to March 2010, entrusted by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT). Aiming at a highly economical fast reactor, the plant concept was developed with quantitative characteristics. The databases of the thermal hydraulics and materials including water chemistry were accumulated by experiments. Based on the success of the project, the second phase of Super FR project was initiated in July 2010. The project consists of three subjects; (1) development of the plant concept: (2) thermal-hydraulics: (3) material-coolant interactions: The Super Fast Reactor has the same plant system of the once-through coolant cycle as the Super Light Water Reactor, the thermal reactor. The results of experimental R&D constitute the common database for the development of Super LWR and Super FR. This paper describes the principle of the reactor concept development and the R&D program of the second phase project.

Improvements of Feedwater Controller for the Super Fast Reactor

The main steam temperature of SCWRs sensitively changes with the power-to-flow ratio. In this article, the feedwater controller of the Super FR (fast-spectrum SCWR) is modified from that of the Super LWR (thermal-spectrum SCWR) for suppressing the variation of the main steam temperature. A plant system analysis code SPRAT-F is used. One of three feedback terms is added to the original feedwater controller that took only the deviation of the main steam temperature into consideration. In the feedwater controller (A), the deviation of the power-to-flow ratio is considered. In the feedwater controller (B), the deviation of the power is considered. In the feedwater controller (C), the time derivative of the power is considered. All the modified feedwater controllers keep the variation of the main steam temperature within 2°C, which has been achieved in recent supercritical coal-fired power plants, against typical load change. In order to further confirm the performance of the modified feedwater controllers, five typical perturbations are analyzed. All the feedwater controllers including the original one stably control the Super FR against all the perturbations without a significant oscillation or offset. Among them, the feedwater controller (B) gives a smaller or at least not larger variation of the main steam temperature compared with the original one at all the perturbations, while the feedwater controllers (A) and (C) give a larger variation in particular cases. From these results, it is concluded that the original feedwater controller is successfully improved as the feedwater controller (B).

Analysis of fuel rod behavior under normal operating conditions in Super Fast Reactor

A Super Fast Reactor is a pressure-vessel type, fast spectrum supercritical water-cooled reactor (SCWR) that is presently researched in a Japanese project. A preliminary core has been designed with 1.59E+06 W/m 3 of power density [1]. In order to ensure the fuel rod integrity, the fuel rod behaviors under the normal operating conditions are analyzed using FEMAXI-6 code. Three types of the limiting fuel rods, with the maximum cladding surface temperature (MCST), maximum power peak (MPP) and maximum discharge burnup (MDB), are chosen to cover all the fuel rods in the core. The power histories of these fuel rods are taken from the neutronics calculation results in the core design. The available design range of the fuel rod design parameters, such as the initial gas plenum pressure, gas plenum length, grain size and pellet-cladding gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumstance direction should be less than 100 MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Compressive stress to yield strength ratio should be less than 0.2. (5) Cumulative damage fraction (CDF) on the cladding should be less than 1.0. Finally the improved fuel rod design is proposed.

CFD analysis of heat transfer in subchannels of a Super Fast Reactor

This paper presents CFD analyses of heat transfer in subchannels of a Super Fast Reactor fuel assembly. Analyses are concentrated on the circumferential temperature distribution on the cladding outer surface because the Maximum Cladding Surface Temperature (MCST) has been a crucial design parameter to evaluate fuel cladding integrity of the Super Fast Reactor. Speziale non-linear high Re k–ɛ model, which can reproduce the anisotropic turbulence flow in non-circular flow channels, with two-layer near-wall treatment is adopted. The results show that heat conduction in the cladding should be considered in the CFD analyses. Larger circumferential temperature gradient occurs on the cladding surface in the edge and corner subchannels than that in the ordinary subchannel because of their special geometries causing larger heterogeneity of mass flow rate distribution inside the subchannels. Improved subchannel configurations to reduce the circumferential temperature gradient are proposed. This study will be a good guideline to the future core design improvement.