Advanced CANDU Reactor Research Articles

Modeling and Intelligent Controller Design for Reactor Regulating System of Advanced CANDU Reactor (ACR-700) in LabVIEW

In this research work, an Advanced CANDU Reactor of 700 MWe rating (ACR-700) is attempted for Reactor Regulating System (RRS) modeling and intelligent controller design. ACR-700 is a state-of-the-art advanced generation-III reactor. The reactor regulating system is modeled with special emphasis on internal and external reactivity devices. ACR-700 is designed with Liquid Zone Control (LZC) Model, Adjuster Banks Model, Absorber Banks Model and Shutdown Banks Model. The RRS model is a highly sophisticated model developed based on the principle of spatial nuclear reactor dynamics. The RRS model is a multivariable model. The spatial reactor dynamics is modeled based on five internal feedbacks and three feedbacks. The original controller design of RRS is comprised of a PID control algorithm. The control design is reattempted with an advanced intelligent algorithm in which the ANFIS controller is used as a modern control design tool. The new ANFIS controller is basically a multivariable controller. All the modeling of RRS is implemented in Visual Basic (VB) Software while the controller is configured in LabVIEW. A special toolkit is designed for the interfacing of Visual Basic and LabVIEW known as VBLAB. The optimization of intelligent controller parameters is carried out by Genetic Algorithm (GA). The GA-optimized intelligent controller is configured with the VB RRS model. All the variable trends are visualized in the VB environment. The proposed closed-loop ACR-700 RRS control system is tested for small perturbation Analysis and power ramp-down transient and found with excellent behavior well within the design limits. The performance of the suggested ACR-700 RRS controller is compared with the existing conventional controller. The performance of the suggested control scheme is marked with reduced oscillations and faster as compared to the existing control scheme.

Advanced CANDU reactors axial xenon oscillation controllability validation

Advanced CANDU reactor fuel channel assembly MCNP model with top, bottom, and four lattice sides reflecting boundaries is used to generate a conservative flat axial neutron flux distribution for the 135Xe oscillation controllability analyses. Burnup-dependent neutron Fission Tally and 135Xe axial profiles are calculated using the Monte Carlo burnup script MCOS. The controllability of the 135Xe oscillations is validated by the MCOS-calculated xenon reactivity worth swings with respect to the mean. The validated upper bound of the xenon reactivity swing band of ±3mk is well within the zone controller reactivity bank of ±7mk.

Use of Thorium-Plutonium MOX in the inner pins of CANDU fuel bundles

This research is focused on using Thorium-Plutonium MOX fuel in the inner fuel pins of the CANDU fuel bundles for plutonium incineration and reduction of uranium demand and to reduce coolant void reactivity. The delayed neutron fraction and the power distribution amongst the fuel elements of the fuel bundle have been considered as main safety parameters.The 700 MWe Advanced CANDU Reactor (ACR-700) was selected as a case study. The inner eight UO2 fuel pins of the ACR-700 fuel bundle are replaced by Thorium-Plutonium MOX fuel pins in the proposed design with 3% reactor grade PuO2. This amount represents 23.4 w/o of the fuel in the bundle. The outer two fuel rings (35 pins) enrichment is reduced from 2.1 w/o U-235 to 2 w/o U-235. The simulation using MCNP6 showed that about 27% reduction of uranium demand can be achieved. The proposed fuel bundle eliminate the use of burnable poisons in the central pin that was used for negative coolant void reactivity and more reduction in the coolant void reactivity was achieved (about 3.5 mk less than the reference fuel bundle). The power distribution throughout the fuel bundle is more flat in the proposed fuel bundle. Use of this fuel bundle reduces the delayed neutron fraction from 540 pcm in the reference case to 480 pcm in the proposed case.

Design of a PWR for long cycle and direct recycling of spent fuel

In a previous work, a new design was proposed for the Pressurized Water Reactor (PWR) fuel assembly for direct use of the PWR spent fuel without processing. The proposed assembly has four zircaloy-4 tubes contains a number of 61-element CANDU fuel bundles (8 bundles per tube) stacked end to end. The space between the tubes contains 44 lower enriched UO2 fuel rods and 12 guide tubes. In this paper, this assembly is used to build a single batch loading 36-month PWR and the spent CANDU bundles are recycled in the on power refueling CANDU reactors. The Advanced PWR (APWR) is considered as a reference design. The average enrichment in the core is 4.76%w U-235. IFBA and Gd2O3 as burnable poisons are used for controlling the excess reactivity and to flatten the power distribution.The calculations using MCNPX showed that the PWR will discharge the fuel with average burnup of 31.8MWd/kgU after 1000 effective full power days. Assuming a 95 days plant outage, 36 calendar months can be achieved with a capacity factor of 91.3%. Good power distribution in the core is obtained during the cycle and the required critical boron concentration is less than 1750ppm. Recycling of the discharged CANDU fuel bundles that represents 85% of the fuel in the assembly, in CANDU-6 or in 700MWe Advanced CANDU Reactor (ACR-700), an additional burnup of about 31 or 26MWd/kgU burnup can be achieved, respectively. Averaging the fuel burnup on the all fuel in the PWR assembly, 58 or 54MWd/kgU burnup of the fuel can be achieved, respectively. Comparing this fuel cycle strategy with that of the advanced pressurized water reactors such as AP-1000 and EPR, we find that almost the same burnup can be achieved with longer core cycle: 36 months versus 18 months for these reactors.Recycling these fuel bundles in CANDU-6 reactors requires minimizing the lattice pitch to 22cm. In order to keep the coolant void reactivity of the ACR-700 slightly negative, the calandria tube should be increased from 7.8cm to 8.2cm to decrease the moderator to fuel volume ratio.

Burnup-dependent effect of lattice-level homogenization and group condensation on calculated kinetics parameters for CANDU-type lattices

Modern analysis of nuclear reactor transients uses space-time reactor kinetics methods. In the Canadian nuclear industry, safety analysis calculations use almost exclusively the Improved Quasistatic (IQS) flux factorization method. The IQS method, like all methods based on flux factorization, relies on calculating effective point kinetics parameters, which dominate the time behavior of the flux, using adjoint-weighted integrals. The accuracy of the adjoint representation influences the accuracy of the effective kinetics parameters. Routine full core calculations are not performed using detailed models and transport theory, but rather using a cell-homogenized model and two-group diffusion theory. This work evaluates the effect of homogenization and group condensation at different burnups, for three fuel types: natural-uranium (NU) fuel, low-void reactivity (LVR) fuel and Advanced CANDU Reactor (ACR) fuel. Results show that the use of a two-group lattice-homogenized adjoint consistently overestimates the effective delayed neutron fraction by approximately 5% for all three fuel types and over a wide burnup range. The use of a two-group lattice-homogenized adjoint also introduces errors in the effective neutron generation time, but these are at most 1.3% (and their sign changes with burnup). Errors tend to vary with burnup by approximately 1% (of the individual parameter value). If a 69-group lattice-homogenized adjoint is used, the errors drop to approximately 2% for the effective delayed neutron fractions and 0.5% for the effective neutron generation time, for all three fuel types and over the entire studied burnup range. The variation of errors with burnup is also reduced.

Improved Zr-2.5Nb Pressure Tubes for Reduced Diametral Strain in Advanced CANDU Reactors

Abstract In an Advanced CANDU Reactor (ACR) (ACR is a registered trademark of Atomic Energy of Canada Limited), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure, and flux than those in a CANDU (CANDU is a registered trademark of Atomic Energy of Canada Limited) reactor, and accordingly require a thicker wall (6.5 mm for ACR versus 4.2 mm for CANDU). In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a study to model and forecast the performance of these thicker pressure tubes has been undertaken. One of the main requirements for the pressure tube is to have low diametral creep. Based on previous experience with CANDU reactor pressure tube performance and manufacture, an assessment of the grain structure and texture of the ACR pressure tubes indicates that the in-reactor creep deformation will be improved. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor.

Integration of Nuclear Energy Into Oil Sands Projects

Energy security and greenhouse gas reduction are thought to be two of the most urgent priorities for sustaining and improving the human condition in the near future. Few places pit the two goals so directly in opposition to one another as the Alberta oil sands. Here, Canadian natural gas is burned in massive quantities to extract oil from one of North America’s largest native sources of carbon-intensive heavy oil. However, this conflict need not continue. Nonemitting nuclear energy can replace natural gas as a fuel source in an economical and more environmentally sound way. This would allow for the continued extraction of transportation fuels without greenhouse gas emissions, while freeing up the natural gas supply for hydrogen feedstock and other valuable applications. Bitumen production in Alberta expanded dramatically in the past 5 years as the price of oil rose to record levels. This paper explores the feasibility and economics of using nuclear energy to power future oil sands production and upgrading activities, and puts forth several nuclear energy application scenarios for providing steam and electricity to in situ and surface mining operations. This review includes the Enhanced CANDU 6, the Advanced CANDU Reactor, and the pebble bed modular reactor. Based on reasonable projections of available cost information, steam produced using nuclear energy is expected to be less expensive than steam produced by natural gas at current natural gas prices and at prices above $6.50/MMBtu (CAD). For electricity production, nuclear energy becomes competitive with natural gas plants at gas prices of $10–13/MMBtu (CAD). Costs of constructing nuclear plants in Alberta are affected by higher local labor costs, which this paper took into account in making these estimates. Although a more definitive analysis of construction costs and project economics will be required to confirm these findings, there appears to be sufficient merit in the potential economics to support further study.

Evaluating accuracy of standard homogenization and need for generalized equivalence theory for ACR ®-lattice checkerboard configurations

The Advanced CANDU ® 1 CANDU ® (CANada Deuterium Uranium) is a registered trademark of Atomic Energy of Canada Limited (AECL). 1 Reactor ACR ® 2 ACR ® (Advanced CANDU Reactor) is a registered trademark of Atomic Energy of Canada Limited (AECL). 2 is a new type of CANDU reactor that is slightly-enriched-uranium fuelled, light-water cooled, and heavy-water moderated. The characteristics of the ACR lattice therefore differ somewhat from those of the standard (natural-uranium fuelled and heavy-water cooled) CANDU lattice. Consequently, the validity of the older neutronic-analysis methodology employed for the natural-uranium CANDU lattice and, in particular, that of the standard homogenization method used for that type of lattice, need to be investigated for the ACR lattice. Of particular interest are configurations in which neutronic properties of lattice cells alternate in a checkerboard pattern. Such patterns occur in CANDU reactors because of alternating fuelling directions in neighbouring channels and because neighbouring channels belong to different coolant loops or different passes within the same loop. The work presented in this paper assesses the accuracy of the standard homogenization method when applied to checkerboard ACR-lattice configurations and investigates whether the use of Generalized Equivalence Theory (GET) yields more accurate results. The conclusion is that the most accurate results, aside from using exact (or, at least, leakage-corrected) macroscopic cross sections and discontinuity factors, can be obtained by using exact (or, at least, leakage-corrected) macroscopic cross sections and no discontinuity factors. The expected errors in the latter case, assuming a 6-cm or finer Cartesian mesh, are less than 1% in bundle power and less than 1 mk in k eff .

Materials performance in CANDU reactors: The first 30 years and the prognosis for life extension and new designs

A number of CANDU reactors have now been in-service for more than 30 years, and several are planning life extensions. This paper summarizes the major corrosion degradation operating experience of various out-of-core (i.e., excluding fuel channels and fuel) materials in-service in currently operating CANDU reactors. Also discussed are the decisions that need to be made for life extension of replaceable and non-replaceable components such as feeders and steam generators, and materials choices for new designs, such as the advanced CANDU reactor (ACR) and enhanced CANDU-6. The basis for these choices, including a brief summary of the R&D necessary to support such decisions is provided. Finally we briefly discuss the materials and R&D needs beyond the immediate future, including new concepts to improve plant operability and component reliability.

Modeling reactivity devices for advanced CANDU reactors using the code DRAGON

Full core analysis of typical power reactors generally performed uses few group diffusion theory, it is necessary to generate beforehand, using a lattice code, the required few group cross-sections and diffusion coefficients associated with each region in the core. For the ACR™ (Advanced CANDU Reactor), the problem is more complex because these reactors contain vertical reactivity devices that are located between two horizontal fuel bundles. The usual calculation scheme relies in this case on a 2D fuel cell calculation to generate the few group fuel properties and on a 3D supercell calculation for the analysis of the reactivity devices present in the core. Because of its complexity, the supercell calculations have usually been performed using simplified fuel geometries. The development of new geometry features in DRAGON and the availability of faster computers have made it possible to improve the 2D cell and 3D supercell models by using explicitly 3D assemblies of clusters to simulate the reactivity devices in CANDU reactors, including the ACR. These studies will thus improve the fine reactor core results by generating more accurate and appropriate reactor databases. In this paper, we will review the lattice-cell/supercell calculation procedure using the code DRAGON by introducing a new supercell model. The use of such an explicit 3D geometry implies a very fine spatial mesh discretization that can generate a large number of regions leading to problems that cannot be solved by the collision probability (CP) method. The method of characteristics (MoC) is then the only alternative for such cases. A comparison of results using these two methods will also be presented for 3D models with a coarse mesh discretization.

Improvements of the reactivity devices modeling for the advanced CANDU reactor

In the context of the ACR™ (Advanced CANDU Reactor), 3D transport calculations are required in order to simulate the reactivity devices located perpendicularly to the fuel channels. The computational scheme that is usually used for CANDU-6 and ACR reactors is based on a simplified supercell geometry in which the fuel clusters and devices are replaced by annuli. Recently, an exact modeling of 3D supercell configurations was introduced within the framework of the ACR calculations. However, with such a model, fine meshing requirements lead to problems that are very demanding in terms of computational resources. In this paper, we present improvements introduced in the ACR context to reduce the cost of the 3D supercell calculations. Two avenues of investigations are reported. First, the introduction of an accelerated characteristics method permits to reduce the computational burden of such calculations involving a large number of regions. In addition, contrarily to CANDU-6 supercell configurations, the ACR 3D geometry is prismatic and consequently a special tracking procedure can be used. This approach introduces no approximation and is significantly faster than the general 3D tracking technique. Thanks to these modifications in the computational procedure, 3D supercell calculations with a level of mesh discretization comparable to 2D cell configurations become affordable for industrial applications.

Coolant void reactivity adjustments in advanced CANDU lattices using adjoint sensitivity technique

Coolant void reactivity (CVR) is an important factor in reactor accident analysis. Here we study the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice using the optimization and adjoint sensitivity techniques. The sensitivity coefficients are evaluated using the perturbation theory based on the integral neutron transport equations. The neutron and flux importance transport solutions are obtained by the method of cyclic characteristics (MOCC). Three sets of parameters for CVR-BOC and k eff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR-BOC (CBCVR-BOC). To approximate the EOC sensitivity coefficient, we perform constant-power burnup/depletion calculations using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Our aim is to achieve a desired negative CVR-BOC of −2 mk and k eff-EOC of 0.900 for the first two cases, and a CBCVR-BOC of −2 mk and k eff-EOC of 0.900 for the last case. Sensitivity analyses of CVR and eigenvalue are also included in our study.

Unstructured grid and unstructured grid porous methods for numerical analysis of TACR moderator calandria

In the design of the Thorium-based Advanced CANDU Reactor (TACR), the moderator system is used to remove the moderator heat in normal operations and to remove the residual heat in accidents. The numerical methods to analyze the heat transfer in the moderator calandria are required. In this paper, the moderator calandria is analyzed by using two types of numerical methods, an unstructured grid method and an unstructured grid porous method. Both methods use collocated unstructured grids and a new average filter method to eliminate the non-physical checkerboard pressure. The unstructured grid porous method introduces the porous method to consider the large tube bundles in the calandria. It is concluded that both methods are effective to analyze the thermal performance in the TACR calandria. Four different designs of the calandria are analyzed and are compared.

Development of Phenomena Identification and Ranking Tables for the Advanced CANDU Reactor

The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6® reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by a heavy water moderator, as with all CANDU reactors. However, ACR uses low enriched uranium fuel, compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety.This paper presents the approach used in developing two phenomena identification and ranking tables (PIRTs) for selected ACR-700 events and their results. One of the two selected events is a large loss-of-coolant accident, which is an ACR design basis event, while the other is a severe flow blockage, which is proposed to be classified as a limited core damage event (beyond design basis event).The paper outlines the design characteristics of the ACR-700 reactor that impact the PIRT process and computer code applicability, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the finalized PIRTs.The ACR-1000 reactor design is currently in detailed design at AECL, and it retains similar design features as the ACR-700 design. Although the PIRTs presented in this paper were developed for the ACR-700 design, they are generally applicable to the ACR-1000 design.

Sensitivity and Uncertainty Analysis for Eigenvalue-Difference Responses

Equations for sensitivity coefficients of eigenvalue-difference responses such as reactivity are derived from a unified approach based on both eigenvalue and generalized perturbation theory. The sensitivity coefficients are utilized for uncertainty analysis of reactivity responses, and it is shown that these types of responses have inherently larger relative uncertainties than eigenvalue responses. Monte Carlo calculations are used to apply the methodology to the analysis of the coolant void reactivity in a three-dimensional model of a fuel bundle in an advanced CANDU reactor system. The important data sensitivities are identified, and it is shown that the coolant void reactivity has a large uncertainty due to nuclear data uncertainties.

Electrolytic hydrogen generation using CANDU nuclear reactors

Hydrogen generation on a large scale will be required to power the emerging hydrogen economy without emitting large amounts of carbon dioxide into the atmosphere. All large-scale hydrogen production facilities currently in operation use processes such as steam-methane reforming or partial oxidation of heavy hydrocarbons, which emit large amounts of carbon dioxide. Atomic Energy of Canada Ltd. (AECL) is studying several process options for large-scale production of hydrogen without emitting carbon dioxide. One of these options is to produce hydrogen via steam electrolysis, where the heat and electricity are produced by nuclear reactors. This paper investigates the technical issues and economics of hydrogen production by integrating a steam electrolysis system with the Advanced CANDU Reactor ACR-1000. The overall thermal-to-hydrogen efficiency of the combined ACR-1000 and steam electrolysis process is 33-34% compared to about 27% for conventional water electrolysis. The heat recovered from the steam electrolysis loop is used to provide heat at appropriate locations in the electricity generation cycle.

ICONE15-10680 ACR-1000 SEVERE ACCIDENT PROGRAM

Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor^<TM>-1000 (ACR-1000^<TM>) as an evolutionary advancement of the current CANDU 6^[○!R] reactor. The ACR-1000 is a Generation III+ plant aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features.This paper presents an overview of the ACR-1000 severe accident program. This program has the following objectives: to ensure public safety from severe accidents, to achieve a robust design with respect to severe accidents, and to meet international guidelines and practice with respect to the severe accident design of advanced reactors.

Pressure-tube and calandria-tube deformation following a single-channel blockage event in ACR-700

One postulated accident scenario for the Advanced CANDU Reactor (ACR-700™) is the complete coolant flow blockage of a single pressure-tube (PT). The flow is not monitored within each individual PT, thus during the early stages of this accident the reactor remains at full power and full pressure, resulting in rapid coolant boil-off and fuel overheating. Melting of the Zircaloy (Zry) components of the fuel bundle can occur, with relocation of some molten material to the bottom of the PT, which may cause failure of the PT and/or the calandria-tube (CT). We analyzed several key phenomena occurring after the blockage, including coolant boil-off, cladding heat-up and melting, dripping of molten Zircaloy (Zry) from the fuel pin, thermal interaction between the molten Zry and the PT, localized bulging of the PT, and interaction of the bulged PT with the CT. The main findings of the study are as follows: (1) Most coolant boils off within 3 s of accident initiation. (2) The Zry cladding starts to melt between 7 and 10 s after accident initiation. (3) The very high heat-up rate typical of the flow blockage accident sequence ensures that the molten Zry would drip to the bottom of the PT. (4) After contacting the molten Zry, the PT and CT bulge out radially under the effect of the reactor pressure. (5) PT/CT failure occurs only if the postulated mass of molten Zry in contact with the PT is sufficiently large, i.e., >100 g. The characteristic time scales for this 100-g case are as follows: - PT bulging starts within 3 s of Zry/PT contact; - PT makes contact with the CT in another 3 s; - CT bulging starts in approximately 1 s; - CT failure occurs within another 6 s. Thus, the duration of the PT/CT deformation transient is 13 s, which gives a total duration of the accident (from PT blockage to PT/CT failure) of 20–23 s. The relatively simple models developed in this study and the estimates generated with these models provide a solid physical framework for the key phenomena in the single-channel flow blockage event in ACR-700. As such, they can also assist in the interpretation and verification of future analyses of this event conducted with more sophisticated codes and tools.

CANDU technology for Generation III+ and IV reactors

Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU ® reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second-generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU Reactor™ (ACR™), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor. Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants. This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R&D and engineering development programs to cover all of these elements. The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating margins. The ACR is also the basis for the Generation IV Super Critical Water Reactor, which extends operation to higher temperatures and pressures.

Thermal hydraulic analysis of the thorium-based advanced CANDU Reactor fuel channel

This paper presents a thermal hydraulic analysis of the thorium-based advanced CANDU Reactor (TACR-1300) fuel channel. The paper first uses CATHENA, a system thermal hydraulics code for CANDU reactors, to determine the boundary conditions of the TACR-1300 fuel channel. The boundary conditions are then passed to ASSERT-PV, a subchannel thermal hydraulics code for the CANDU industry, to obtain detailed thermal hydraulic information on the TACR-1300 fuel channel. Through the subchannel analysis, it is found that the onset of dry-out power (ODP) for the TACR fresh fuel channel is much less than that of the ACR-700 because thorium fuel rods as well as uranium rods are not heated and there is bypass-flow effect in the associated subchannels. The paper compares three thorium loading modes, and determines the best one for TACR.