High Temperature Engineering Test Reactor Research Articles

Study on Natural Convection Heat Transfer of High Temperature Gas in a Vertical Annular Space of a Double Coaxial Cylinder

Water cooling panels are adopted as a vessel cooling system of a High Temperature-engineering Test Reactor (HTTR) to cool the reactor core indirectly by natural convection and thermal radiation. In order to investigate heat transfer characteristics of high temperature gas in a vertical annular space between the reactor pressure vessel and the cooling panels of the HTTR, we carried out experiments and numerical analyses on natural convection heat transfer coupled with surface-to-surface thermal radiation in a vertical annular space of a double coaxial cylinder. In the present experiments, Rayleigh number based on the height of the vertical space was set to be 2.0 × 107 < Ra < 5.4 × 107 for helium and 1.2 × 109< Ra < 3.5 × 109 for nitrogen. The numerical results were in good agreement with the experimental ones regarding the temperature on the heating and cooling walls, and so on. As a result of the experiments and the numerical analyses, a heat transfer coefficient of natural convection coupled with surface-to-surface thermal radiation was obtained as functions of Rayleigh number, radius ratio, temperature and emissivities on the heating and cooling walls.

Fuel and Fission Gas Behavior during Rise-to-Power Test of the High Temperature Engineering Test Reactor (HTTR)

The rise-to-power tests of the High Temperature Engineering Test Reactor (HTTR) have been carried out successfully by the Japan Atomic Energy Research Institute (JAERI). For the safe operation of HTTR, the continuous and reliable measurement of the coolant activity is required to evaluate the fuel performance during normal operating conditions. In order to measure the primary coolant radioactivity (PCR), the PCR instrumentation of the safety protection system, the fuel failure detection (FFD) system and the primary coolant sampling system have been installed in the primary circuit. The PCR was less than 103 MBq/m3, and measured Kr and Xe isotopes were less than 0.1 MBq/m3 during the rise-to-power tests. The measured fractional releases are constant at 2x10-9 up to 60% of the reactor power, and then increase to 7x10-9 at full power operation. The prediction shows good agreement with the measured value. These results showed that the release mechanism varied from recoil to diffusion of the generated fission gas from the contaminated uranium in the fuel compact matrix.

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams : Takeda, T. et al. Annals of Nuclear Energy, 2003, 30, (7), 811–830

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams : Takeda, T. et al. Annals of Nuclear Energy, 2003, 30, (7), 811–830

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).

Investigation of Irradiation Behavior of SiC-Coated Fuel Particle at Extended Burnup

In current high-temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. The maximum burnup of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is limited to 3.6%FIMA (% fission per initial metallic atom) to certify its integrity during the operation. In order to investigate fuel behavior under extended burnup condition, irradiation tests were performed. The irradiation was carried out as HRB-22 and 91F-1A capsule irradiation tests. The fuel for the irradiation tests was called extended burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the HTTR. In order to keep fuel integrity up to over 5%FIMA, the thickness of buffer and SiC layers of fuel particle were increased. The fuel compacts were irradiated in the HRB-22 and the 91F-1A capsules at the High Flux Isotope Reactor of Oak Ridge National Laboratory and at the Japan Materials Testing Reactor of the Japan Atomic Energy Research Institute, respectively. The comparison of measured and calculated release rate-to-birth rate ratios showed that there were additional failures in both irradiation tests. A pressure vessel failure model analysis showed that no tensile stresses acted on the SiC layers even at the end of irradiation and no pressure vessel failure occurred in the intact particles even in a particle with thin buffer layer with failed OPyC layer. The presumed failure mechanisms are additional through-coatings failure of as-fabricated SiC-failed particles or an excessive increase of internal pressure by the accelerated irradiation. Further study is needed to clarify the failure mechanism.

Assessment of irradiation temperature stability of the first irradiation test rig in the HTTR

The High Temperature Engineering Test Reactor (HTTR) can provide very large irradiation spaces at high temperatures for various irradiation tests. The first irradiation test rig for the HTTR, the I–I type irradiation equipment, was developed for an in-pile creep test on a stainless steel with large standard size specimens. The equipment uses the ambient high temperature of the core for the irradiation temperature control. The target irradiation temperatures are 550 and 600 °C with the target temperature deviation of ±3 °C. In this study, the specimen temperature stability at the irradiation test was assessed by both analytical and experimental approaches. The irradiation temperature changes at transient conditions were analyzed by a finite element method (FEM) code and the temperature controllability of the equipment was examined by a mockup test. The controllability was evaluated with the measured temperature transient data at the core graphite components in the Rise-to-Power tests of the HTTR. The result indicates that the temperature control method of the I–I type irradiation equipment is effective to keep the irradiation temperature stable in the irradiation test.

GTHTR300 design and development

In Japan Atomic Energy Research Institute (JAERI), successful development and operations of the High Temperature Engineering Test Reactor (HTTR) are coordinated by programs to study applied systems of the promising reactor technology. One such program being carried out from 2001 to 2007 is the design and development for the Gas Turbine High Temperature Reactor of 300 MWe nominal capacity (GTHTR300), aiming at demonstration of a prototype plant in 2010s in Japan. GTHTR300 features an original design of fuel cycle based on improved HTTR fuel element and of greatly simplified plant system in pursuit of economical performance with minimal development requirements. The fuel cycle characterizes on high burnup, low power peaking factor, and extended refueling interval. The plant system incorporates salient design simplification including conventional steel reactor pressure vessel, non-intercooled power conversion cycle, horizontal single-shaft turbomachine, and distributed modular maintenance of the overall plant. Planned research and development supporting detailed design efforts include performance tests of subscale components and control system. This paper describes the GTHTR300 reference design and related R&D activities in JAERI.

Indirect air cooling techniques for control rod drives in the high temperature engineering test reactor

The high temperature engineering test reactor (HTTR) is the first high-temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 °C and thermal power of 30 MW. Sixteen pairs of control rods are employed for controlling the reactivity change of the HTTR. Each standpipe for a pair of the control rods, which is placed on the top head dome of the reactor pressure vessel, contains one control rod drive mechanism. The control rod drive mechanism may malfunction because of reduction of the electrical insulation of the electromagnetic clutch when the temperature exceeds 180 °C. Because 31 standpipes stand close together in the standpipe room, 16 standpipes for the control rods, which are located at the center, should be cooled effectively. Therefore, the control rod drives are cooled indirectly by forced air circulation through a pair of ring-ducts with proper air outlet nozzles and inlets. Based on analytical results, a pair of the ring-ducts was installed as one of structures in the standpipe room. Evaluation results through the rise-to-power test of the HTTR showed that temperatures of the electromagnetic clutch and the ambient helium gas inside the control rod standpipe should be below the limits of 180 and 75 °C, respectively, at full power operation and at the scram from the operation.

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 °C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 °C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open.

405.高温ガス炉燃料の再処理技術

A High Temperature Gas-Cooled Reactor (HTGR) is particularly attractive due to capability of producing high temperature helium gas and its inherent safety characteristic. Research and development of high temperature gas turbine plant and high temperature heat utilizing technology are now undergoing. The High Temperature Engineering Test Reactor (HTTR) is a research facility constructed by the Japan Atomic Energy Research Institute (JAERI). All rise-topower tests have been successfully carried out and the performance of the HTTR has been evaluated. Now, preparation for the operation with outlet coolant temperature of 950°C and safety demonstration tests are undergoing.This paper describes reprocessing technology of HTGR fuels. Coated fuel particles, consisted of a microsphere of low enriched UO2 with TRISO particles, are used as the HTGR fuels. In order to reprocess HTGR fuels, a head-end process is needed and JAERI had confirmed jet-grind method as basic technologies of the head-end process. Since Purex method can be used after the head-end process, a reprocessing system for the HTGR fuels could be established. Also the preliminary study on the methodology for disposing graphite blocks in a HTGR was carried out, and its evaluation results were briefly presented.

404. 高温ガス炉ガスタービン発電システム(GTHTR300)発電系設備の点検方法・手順

Maintenance methods and procedures for the power conversion system of the Gas Turbine High Temperature Reactor 300 (GTHTR300) were examined. Because of being installed in the power conversion vessel which is primary pressure boundary and exposed to primary coolant contaminated with fission products, unique methods and procedures which are different from those for existing power plants and High Temperature Engineering Test Reactor (HTTR) are needed for especially the gas turbine and the compressor. The gas turbine-compressor assembly is pulled out from the power conversion vessel by a special trolley after generator vessel section is separated. Then, the gas turbine-compressor assembly is lifted and carried to a maintenance facility by a building overhead crane. During the removal process, preventive measures to reduce the exposure are taken by limiting working time and bagging the assembly to keep fission products from scattering. Before casings are open, the gas turbine-compressor assembly shall be in storage for a certain period to reduce the dose rate. In order to achieve a high plant availability, a spare gas turbine-compressor assembly is installed as an alternative to the assembly cooled in the storage. It was confirmed that, by applying those proposed methods and procedures, it is possible to keep the plant availability more than 90%. Present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.

Fuel and Fission Gas Behavior during Rise-to-Power Test of the High Temperature Engineering Test Reactor (HTTR)

Fuel and Fission Gas Behavior during Rise-to-Power Test of the High Temperature Engineering Test Reactor (HTTR)

Safety Shutdown of the High Temperature Engineering Test Reactor during Loss of Off-site Electric Power Simulation Test

The high temperature engineering test reactor (HTTR) is a graphite-moderated and helium-gas-cooled reactor, which is the first high temperature gas-cooled reactor in Japan. The HTTR achieved its first full power of 30 MW at rated operation on December 7 in 2001. In the rise-to-power test of the HTTR, simulation test of anticipated operational occurrence with scram was carried out by manual shutdown of off-site electric power from 30 MW operation. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, mass flow rates of helium and water decreased to the scram points. Sixteen pairs of control rods were inserted at two-steps into the core by gravity within the design criterion of 12 s. In 51 s after the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. In 40 min after the startup of the auxiliary cooling system, one of two auxiliary helium circulators stopped for reducing thermal stresses of core graphite components such as fuel blocks. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. Blackout sequences of the HTTR dynamic components were in accordance with the design. As a result of the loss of off-site electric power simulation test, it was confirmed that the HTTR shuts down safely after the scram.

Inspection techniques for primary pressurized water cooler tubes in the high temperature engineering test reactor

A primary pressurized water cooler (PPWC) with 136 inverse-U-tubes is installed in the primary cooling system of the high temperature engineering test reactor (HTTR). The HTTR is the first high temperature gas-cooled reactor in Japan with an outlet gas temperature of 950 °C and thermal power of 30 MW. The heat transfer tubes form the reactor pressure boundary of the primary coolant. Inspection techniques for the tubes should be established to carry out the in-service inspection efficiently. An automatic inspection system for the tubes uses probes for eddy current testing and ultrasonic testing. Defect detecting characteristics of the inspection probes and the application of the automatic inspection system to nondestructive test of the tubes were examined by a mockup test utilizing artificially degraded tubes. The automatic inspection system could smoothly insert and withdraw the probe at its moving velocity in the fixed positions of the defected tube. Nondestructive test of the tubes using the automatic inspection system was conducted during reactor shutdown period of the HTTR after test operation of about 55% of the full power. Through the nondestructive test, it was concluded that there was no defect on the outer surface of the heat transfer tubes of the PPWC inspected.

373. 高温ガス炉ガスタービン発電システム(GTHTR300)用高燃焼度燃料の成立性評価,(I)

Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300) based on the experiences of development of the High Temperature Engineering Test Reactor (HTTR) of JAERI. Feasibility of the GTHTR300 fuel which is irradiated under high burnup condition of approximately 140GWd/t at the maximum was investigated. It was certified that failure of coated fuel particles during irradiation will not occur and the integrity of coated fuel particles will be kept. Cost evaluation for fuel production and fuel cycle was also studied in order to clarify the economical feasibility which is important item for the development of the GTHTR300. The fuel cost per electricity production is estimated to be less than 1.1Yen/kWh. The feasibility of the fuel cost is certified.The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.

1937 高温工学試験研究炉の定格熱出力 30MW 到達

1937 高温工学試験研究炉の定格熱出力 30MW 到達

Safety Shutdown of the High Temperature Engineering Test Reactor during Loss of Off-site Electric Power Simulation Test.

Safety Shutdown of the High Temperature Engineering Test Reactor during Loss of Off-site Electric Power Simulation Test.

372. HTTR(高温工学試験研究炉)の出力上昇試験

372. HTTR(高温工学試験研究炉)の出力上昇試験

Elastic–plastic analysis of nuclear structures with millions of DOFs using the hierarchical domain decomposition method

The hierarchical domain decomposition method (HDDM) proposed by Comp. Sys. Eng. 4 (1993) 495 is applied to the large scale elastic–plastic finite element (FE) analysis of nuclear structures. The HDDM is a method to implement the finite element method (FEM) on various kinds of parallel environments. The substructure-based iterative methods can effectively be used with the HDDM to solve the large scale linear algebraic equations derived from the implicit FEM. In this paper, some key techniques to parallelize the static elastic–plastic FE analysis by the HDDM are described. As illustrative examples, a support structure of the high temperature engineering test reactor (HTTR), a pressure vessel, and an internal pump of a pressure vessel are analyzed. The structure of HTTR and the pressure vessel are modeled by hexahedral solid elements whose total degrees of freedom (DOFs) are about 1.3 millions (M) and 3 M, respectively. The internal pump is modeled by quadratic tetrahedral elements whose total DOFs are about 2 M. The elastic–plastic analysis of a simple cube with 10 M DOFs is also carried out. Both the conjugate gradient method for solving the linear equations and the Newton–Raphson method for solving nonlinear problems successfully converge.

Modeling of fuel performance and metallic fission product release behavior during HTTR normal operating conditions

The computer codes PANAMA and FRESCO developed at the Research Center Jülich have been used for the prediction of fuel performance and fission product release behavior during the normal operation of the Japanese High-Temperature Engineering Test Reactor, HTTR. Basis for the calculations was the so-called ‘Standard HTTR Operation Plan’ with a nominal operation time of 660 efpd including a 110 efpd period with enhanced fuel temperatures. Fuel performance model calculations with the PANAMA code have shown that for the temperature distribution given, only a small additional failure fraction is expected. The diffusive release of metallic fission products from the fuel occurs mainly from the central core layers with the maximum temperatures whereas there is little contribution from the upper layer. Silver most easily escapes the fuel. The release data for strontium and cesium also reveal a significant fraction to originate from still intact particles. The comparison with the calculations obtained with the JAERI models has shown a good agreement for the release from the coated particles.