Block-type High Temperature Gas-cooled Reactors Research Articles

High Temperature Test Facility sensitivity and calibration studies to inform OECD-NEA benchmark calculations

The High Temperature Test Facility (HTTF) is a thermal hydraulics test facility for block-type high-temperature gas-cooled reactors (HTGRs) built to provide data to validate HTGR modeling tools. Several experiments were conducted at HTTF in the spring of 2019 that serve as the basis for an upcoming Organization for Economic Cooperation and Development – Nuclear Energy Agency HTGR thermal hydraulics benchmark. We present sensitivity studies of steady state and both the pressurized and depressurized conduction cooldown (PCC and DCC) transients in HTTF in this paper. These studies identified that uncertainties in the thermal conductivity of the ceramic blocks making up the HTTF core are significant drivers of peak temperatures in steady state, PCC, and DCC. The emissivity of the heater rods and core blocks also provide significant control over heater rod temperatures. We also performed calibration studies based on HTTF Experiment PG-26, which was intended to serve as the basis for the DCC benchmark problem. We found that no calibration was able to reproduce PG-26 peak block temperatures within an average of 30%, but by taking the best-estimate PG-26 calibration and applying it to a similar experiment—PG-29—we were able to reproduce the peak block temperatures within an average of 6.2% error. Based on this nearly fivefold improvement in peak block temperature predictions from PG-26 to PG-29 and confounding factors in PG-26 that contributed to previous difficulties modeling the experiment in RELAP5-3D and SAM we conclude that PG-26 data are insufficient for developing a DCC benchmark and recommend using PG-29 as the benchmark experiment instead.

Analysis of air ingress scenario in GEMINI+ plant

The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI+ project. The GEMINI+ reactor is a prismatic block-type High-Temperature Gas-Cooled Reactor (HTGR). Within tasks devoted to safety, an analysis of air ingress scenario was performed by NRG with the thermal–hydraulic system code SPECTRA, while UJV and NCBJ worked with the integral system code MELCOR. Two air ingress scenarios were analyzed:Design Basis Accident (DBA) scenario. The air ingress scenario selected as DBA is a 65 mm break of a Helium purification line on top of the Steam Generator,Large air ingress scenario. The air ingress scenario selected is guillotine break of the coaxial gas duct.The DBA scenario is characterized by a very long period with practically stagnant gas in the primary system after the initial depressurization. However, very small gas flows through the break due to counter flow and diffusion in the break region (gas velocities of 10–2 m/s) are difficult to model with system codes. Therefore the break models in the system code were calibrated by performing CFD simulations. Furthermore, a model of gas mixing by diffusion was introduced in the NRG analysis for the practically stationary gas in the primary system (gas velocities of 10–4 m/s). The large air ingress scenario, with relatively large flow through the break, is easier to model from this standpoint.The NRG SPECTRA results showed that the amount of air that can reach the core is extremely small in the DBA accident (0.5 kg of graphite consumed after 100 h). The maximum depth of oxidation was 2 µm. In the large air ingress scenario the air ingress is significantly larger. At 100 h, approximately 105 kg of graphite was consumed. Even then, the maximum depth of oxidation was very small, approximately 0.2 mm. MELCOR results, performed at NCBJ and UJV, basically confirmed that the amount of air ingress is very small. The numbers were somewhat higher, mainly due to heavy flow oscillations in the break that could not be mitigated in MELCOR calculations.

Comparative study of meta-heuristic algorithms for reactor fuel reloading optimization based on the developed BP-ANN calculation method

In order to obtain optimal core configurations, the problems of fuel reloading optimization must be solved in nuclear engineering field. This problem is a combinational optimization problem with an enormous solution space. In the traditional calculation method, the meta-heuristic algorithms (MHAs) are used to search and update optimization solutions continuously, while the core physics calculation programs are used to calculate the optimization parameters. In this research, we developed a new BP-ANN calculation method to solve the problems of fuel reloading optimization for thorium-based block-type high temperature gas-cooled reactors (HTGRs). In this method, the function of core physics calculation programs are replaced by BP-ANNs in order to reduce calculation time. Besides, five different MHAs, which are genetic algorithm (GA), particle swarm optimization (PSO), Teaching-learning based optimization algorithm (TLBO), harmony search (HS) and Rank-based ant system (RAS), are used under this new method to solve the problems of fuel reloading optimization, respectively. The results shows that PSO is the best MHA with a real feasible solution ratio of 0.30 and the best fitness function value of 0.455904. Compared to the traditional calculation method, the calculation time of BP-ANN calculation method is reduced to 0.7% of the former.

Optimum coating layer thicknesses of a TRISO having an 800-μm UO2 kernel under normal operation conditions of a 10-MWth block-type HTGR

The study evaluates the optimal thicknesses of the coating layers of an 800-μm UO2 tri-structural isotropic coated fuel particle (TRISO) for a 10-MWth block-type micro modular high temperature gas-cooled reactor (HTGR) using a response surface methodology (RSM) and an HTGR fuel performance analysis code, COPA. The objective function to be minimized in the optimal design of a TRISO consists of the product of the packing fraction of TRISO particles in a compact, the fractional releases of major metallic fission products from a TRISO and the failure probability of the silicon carbide (SiC) layers. A constraint on the sum of the coating layer thicknesses was made using a maximum packing fraction. Two fuel performance analyses such as a failure fraction analysis and a fission product release analysis are applied to the optimal designs of a TRISO that the RSM produced. An 800-μm UO2 TRISO having buffer, inner high-density pyrocarbon (IPyC), SiC, outer high-density pyrocarbon (OPyC), each with a thickness of 110, 40, 57, 35 μm, or 120, 35, 60, 27 μm is suitable for the operation of a 10-MWth block-type micro modular HTGR within 7600 days when both the SiC and through-coatings failure fractions do not occur and the fractional release of silver, cesium, strontium and krypton are below 1 × 10−9. The packing fraction of the optimal TRISOs in a compact is 40%.

DEVELOPMENT OF PIN-LEVEL NEUTRONICS/THERMAL-FLUID ANALYSIS COUPLED CODE SYSTEM FOR A BLOCK-TYPE HTGR CORE

Recently, the coupling between computer codes that simulate different physical phenomena has attracted for more accurate analysis. In the case of high-temperature gas-cooled reactor (HTGR), the coupling between neutronics and thermal-fluid analysis is necessary because of large change of temperature in the reactor core. Korea Atomic Energy Research Institute (KAERI) has developed the coupled code system between a reactor physics analysis code CAPP and a thermal-fluid system safety analysis code GAMMA+ for a block-type HTGR. The CAPP/GAMMA+ coupled code system provides more accurate block-wise distribution data than CAPP or GAMMA+ stand-alone analysis. However, the block-wise distribution data has the limitation in order to predict safety parameters such as the maximum temperature of the nuclear fuel. It is necessary to calculate refined distribution, for example, pin-level (fuel compact level) distribution. In this study, we tried to solve this problem by coupling CAPP and a high-fidelity thermal-fluid analysis code CORONA. CORONA can perform a high-fidelity thermal-fluid analysis of Computational Fluid Dynamics (CFD) level by dividing a block-type HTGR core into small lattices. On the other hand, CAPP can provide a pin power distribution. It is expected that the refined, more accurate distribution data for a block-type HTGR can be obtained by coupling these two codes. This paper presents the development of coupled code system between CAPP and CORONA, and then it is tested on a simple HTGR column problem with encouraging results.

Comparison and analysis of different selection strategies of genetic algorithms for fuel reloading optimization of Thorium-based HTGRs

The nuclear fuel cycle cost can be effectively reduced through fuel reloading optimization. Genetic algorithm (GA) is a classic optimization algorithm that is widely applied in fuel reloading optimization. In the GA, selection is a key operator. However, few studies have compared and analyzed different selection strategies. In this study, 1/6 core of thorium-based block-type high temperature gas-cooled reactor was considered as an example, and ten different selection strategies were compared and analyzed. Five of these strategies were the roulette wheel and proportionate selection, tournament selection, uniform sorting, exponential sorting, and deterministic selection, whereas the other five were the aforementioned selection strategies combined with the truncation selection strategy. These ten different selection strategies were evaluated for single-objective and multi-objective problems. In single-objective optimization problems, the effective neutron multiplication factor was selected as the only optimization objective, whereas in multi-objective optimization problems, the effective neutron multiplication factor and power peak factor were considered as optimization objectives. The results indicated that exponential sorting was the best selection strategy for single-objective optimization problems, whereas hybrid truncation exponential sorting was the best selection strategy for multi-objective optimization problems.

Block-type high temperature gas cooled reactor reloading pattern optimization using genetic algorithm

The reactor reloading pattern optimization is a typical combinatorial optimization problem with a huge search space. It is very hard for traditional optimization algorithm to find the global optimal solution in such huge search space. However, for combinatorial optimization problem, the genetic algorithm (GA) provides a very effective solution by its excellent adaptive ability and optimization ability. This paper is focused on the reloading pattern optimization by using GA in a block-type high temperature gas cooled reactor(HTGR) and corresponding programs were written to realize this goal. To improve the calculation accuracy of core physics, the transport calculation with 26 groups is adopted in the core calculation, which will also be time-consuming. To make up for this shortcoming, the parallel optimization of GA is carried out. Finally, a refueling optimization benchmark in a small HTGR is constructed to test the optimization ability of GA. The results show that GA has a good optimization ability and computational stability for reloading pattern optimization in block-type HTGRs.

Burn-Up Dependency of Control Rod Position at Zero-Power Criticality in the High-Temperature Engineering Test Reactor

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR), which was constructed in Japan. The operating data of HTTR with burn-up to about 370 EFPD (effective full-power days), which are very important for the development of HTGRs, have been collected in both zero-power and powered operations. In the aspects of code validation, the detailed prediction of temperature distribution in the core makes it difficult to validate the calculation code because of difficulty in measuring the core temperature directly in powered operation of the HTTR. In this study, the measured data of the control rod position, while keeping the temperature distribution in the core uniform at criticality in zero-power operation at the beginning of each operation cycle were compared with the calculated results by core physics design code of the HTTR. The measured data of the control rod position were modified based on the core temperature correlation. At the beginning of burn-up, the trends of burn-up characteristics are slightly different between experimental and calculation data. However, the calculated result shows less than 50 mm of small difference (total length of control rod is 4060 mm) to the measured one, which indicates that the calculated results appropriately reproduced burn-up characteristics, such as a decrease in uranium-235, accumulation in plutonium, and decrease in burnable absorber.

Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core temperature of the HTTR is too high. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross-section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

Development of fuel temperature calculation code for HTGRs

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code. In addition, the effect of cooling forms on fuel is investigated by using FTCC.

Burnup performance of small-sized long-life CANDLE high temperature gas-cooled reactors with U–Th–Pa fuel

The innovative CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy has been successfully applied to both fast and thermal reactors. In particular for thermal reactor applications, CANDLE block-type high temperature gas-cooled reactors (HTGRs) with either U–Pu or U–Th fuel cycle had been proposed and investigated for their simple and safe reactor operation, and the ease of designing a long life reactor. In this work, we investigated the use of Pa-231 mixed homogeneously in the (Th-232/U-233)O2 fuel kernel of the TRISO particles, that is the U–Th–Pa fuel, to obtain better CANDLE burnup performance. Some consideration on how to supply the required U-233 and Pa-231 was also thoroughly discussed. U–Th–Pa fuel decreases the required amount of natural Gd burnable poison in the fresh fuel for establishing a valid CANDLE HTGR design since Pa-231 has a large thermal absorption cross section. Besides the role as a burnable poison nuclide, Pa-231 also serves as a fertile nuclide during the CANDLE burning since Pa-231 is finally transmuted to a fissile U-233 nuclide. A promising analysis result shows that for U-233 and Pa-231 concentrations of 15 and 7.50w/o, respectively, the discharged burnup can be increased from 138 (U-Th fuel) to 149GWd/t. This extends the core life time about 16%, i.e. from 35 (U-Th fuel) to 41years with CANDLE active core height of 800cm and reactor thermal power of 30MWth.

Study on Operation Scenario of Tritium Production for a Fusion Reactor Using a High Temperature Gas-Cooled Reactor

To start up a fusion reactor, it is necessary to provide a sufficient amount of tritium from an external device. The fusion DEMO reactor is planned to start up in the 2030s. Herein, methods for supplying the reactor with tritium are discussed. For the initial startup of the fusion reactor, use of a high temperature gas-cooled reactor (HTGR) as a tritium production device has been proposed. So far, the analyses have been focused only on the operation in which fuel is exchanged at stated periods (batch) using the block type HTGR. In this paper, to improve the performance of tritium production, properties of the HTGR are studied from the viewpoint of continuous operation for several conditions. In continuous operation, for example, in the pebble bed type HTGR, it is possible to design an operation that has no time loss for refueling. The pebble bed modular reactor (PBMR) and the gas turbine high temperature reactor of 300 MWe nominal capacity (GTHTR300) are assumed as the calculation and comparison targets, and simulation is made using the continuous-energy Monte Carlo transport code MVP-BURN. It is shown that the continuous operation using the pebble bed type HTGR has almost the same tritium productivity compared with the batch operation using the block type HGTR. The issues for pebble bed type HTGR as a tritium production device are discussed.

Thermal-hydraulic design and transient evaluation of a small long-life HTR

Small long-life high temperature gas-cooled reactors (HTRs) may provide electricity or heat for remote areas or industrial users in developed and/or developing countries. Moreover, small HTRs have advantages over large nuclear reactors of demonstrated inherent safety, transportability, modular construction, and flexible site selection. This paper presents the thermal-hydraulic evaluations of the U-Battery, which is a small, long-life and block-type HTR using the DALTON/THERMIX code system. The thermal-hydraulic characteristics of a cylindrical design and an annular design of the U-Battery were evaluated for loss of forced-cooling (LOFC) incidents including depressurized LOFC (DLOFC) and pressurized LOFC (PLOFC) incidents. The calculations show that the stronger natural circulation during the PLOFC makes the reactor core cool faster than during the DLOFC, flattens the radial solid temperature distribution, and transfers more heat from the hot regions (bottom and center of the reactor core) to cold regions (top and periphery of the reactor core). Although the natural circulation in the reactor core is so weak that it is neglected during the DLOFC, the decay heat is removed passively by conduction without any violation of the temperature limits for the 20MWth U-Battery. The comparisons of the cylindrical and annular reactor core configurations show that the latter is a better design with a lower maximum core temperature during the LOFC because of the central graphite reflector. Moreover, it is possible to adopt the current reactor configuration when the thermal power of the U-Battery increases from 20MWth to 40MWth.

Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

DYN3D is a nodal diffusion code for 3D steady-state and transient analysis of Light Water Reactor (LWR) cores with hexagonal or square fuel element geometry. In addition to the neutron kinetics, it comprises of a thermal-hydraulics model for flow in parallel coolant channels. Macroscopic cross section data libraries generated with variation of burn-up, reactor poisons concentrations and thermal-hydraulic feedback parameters are linked to the code. Two-group and multi-groups versions of the code are available.Currently, at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the DYN3D code is being extended and adopted for the application to block-type High Temperature Gas-Cooled Reactors (HTGRs). In this paper, we give an overview of the latest developments of DYN3D concerning block-type HTGR.The simplified P3 (SP3) transport approximation is implemented into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method previously implemented into DYN3D for square fuel element geometry of LWR is being extended for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes to be able to perform a systematic mesh refinement.One of the main challenges in cross section generation for the HTGR core calculations is the treatment of the so-called “double heterogeneity”. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. The main steps of the RPT method are described. The use of the method for the cross section generation of a simplified HTGR core including its verification is presented.A 3D heat conduction module coupled with a channel-type coolant flow model is implemented to take the temperature reactivity feedback to neutronics physically correctly into account. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks.

Development of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

Abstract The Light Water Reactor (LWR) dynamics code DYN3D is extended and adopted for the application to block-type High Temperature Gas-Cooled Reactor (HTGR). A procedure for the cross section generation for the HTGR core calculations was developed. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. A full core analysis of the reference simplified HTGR core is performed with DYN3D using macroscopic nodal cross sections provided by HELIOS. The SP3 transport approximation is integrated into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method was developed for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes. A 3D heat conduction module coupled with a channel-type coolant flow model is implemented into the code. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks.

Experimental validation of effectiveness of rod-type burnable poisons on reactivity control in HTTR

In block-type high temperature gas-cooled reactors (HTGRs), insertion depth of control rods (CRs) into a core should be retained shallow to keep fuel temperature below 1495 °C through a burnup period, and hence excess reactivity should be reduced through a different method. Loading burnable poisons (BPs) into the core is considered as a method to resolve this problem as in case of light water reactors (LWRs). Effectiveness of BPs on reactivity control in LWRs has been validated by experimental data, however, this has not been done yet for HTGRs, because there was not enough burnup characteristics data for HTGRs required for the validation. The High Temperature Engineering Test Reactor (HTTR) is a block-type HTGRs and it adopts rod-type BPs to control reactivity. The HTTR has been operated up to middle burnup, and thereby the experimental data was expected to show effect of the BPs on the reactivity control. Hence, in order to validate effectiveness of rod-type BPs on reactivity control in the HTTR, we investigated on the HTTR results whether the BPs have functioned as designed. As a result, the CRs insertion depth has been retained shallow within allowable range, and then effectiveness of rod-type BPs on reactivity control in the HTTR was validated.

高温ガス炉ガスタービン発電プラントにおける設計パラメータの最適化に関する研究

A modular high-temperature gas-cooled reactor (HTGR) coupled with a closed cycle helium gas turbine is expected as one of the promising power generation plants due to inherent safety and economics. Effects of main design parameters, particularly gas pressure, on cycle thermal efficiency and power generation cost were examined for the block-type HTGR power plant. A previous design of 600 MWt reactor with inlet and outlet temperatures of 460 and 850°C and a gas pressure of 6 MPa was assumed as a reference design. A computer analysis program combining a core thermal design and a gas turbine cycle was developed. From the results of parametric analyses regarding both the pressure and the fuel channel diameter based on this program, the reactor pressure vessel (RPV) cooling flow ratio of 1% was clarified to be sufficient. With increasing pressure, pressure drop decreased, and consequently, a higher cycle thermal efficiency and a lower power generation cost were achieved. Namely, it was clarified that an optimal pressure exists at around 8 MPa, which is different from the existing generally employed pressure of 6 or 7 MPa. In addition, the effects of fuel cost escalation were examined.

Small high temperature gas-cooled reactors with innovative nuclear burning

Since the innovative concept of CANDLE ( Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety, easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles.

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel ( 233U, 232Th)O 2 are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.

06/02127 Optimization of the power ascension path for a boiling water reactor using genetic algorithms: Lee, C. -S. and Lin, C. Annals of Nuclear Energy, 2006, 33, (1), 53–61.

The innovative CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy has been successfully applied to both fast and thermal reactors. In particular for thermal reactor applications, CANDLE block-type high temperature gas-cooled reactors (HTGRs) with either U–Pu or U–Th fuel cycle had been proposed and investigated for their simple and safe reactor operation, and the ease of designing a long life reactor. In this work, we investigated the use of Pa-231 mixed homogeneously in the (Th-232/U-233)O2 fuel kernel of the TRISO particles, that is the U–Th–Pa fuel, to obtain better CANDLE burnup performance. Some consideration on how to supply the required U-233 and Pa-231 was also thoroughly discussed. U–Th–Pa fuel decreases the required amount of natural Gd burnable poison in the fresh fuel for establishing a valid CANDLE HTGR design since Pa-231 has a large thermal absorption cross section. Besides the role as a burnable poison nuclide, Pa-231 also serves as a fertile nuclide during the CANDLE burning since Pa-231 is finally transmuted to a fissile U-233 nuclide. A promising analysis result shows that for U-233 and Pa-231 concentrations of 15 and 7.50 w/o, respectively, the discharged burnup can be increased from 138 (U-Th fuel) to 149 GWd/t. This extends the core life time about 16%, i.e. from 35 (U-Th fuel) to 41 years with CANDLE active core height of 800 cm and reactor thermal power of 30 MWth.