Kazuhiro Sawa

Fabrication of the First-Loading Fuel of the High Temperature Engineering Test Reactor

The High Temperature Engineering Test Reactor (HTTR). which is the first high temperature gas-cooled reactor (HTGR) in Japan, attained its first criticality in November 1998. The fabrication of the...

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

Abstract The computer codes PANAMA and FRESCO developed at the Research Center Julich 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.

An Investigation of Irradiation Performance of High Burnup HTGR Fuel

In order to investigate fuel behavior under high burnup irradiation condition of high temperature gas-cooled reactor (HTGR), an irradiation test was performed. An irradiation was carried out as a part of a cooperative effort between the US DOE and the Japan Atomic Energy Research Institute. The fuel for the irradiation test was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR). In order to keep fuel integrity up to high burnup over 5%FIMA (% fission per initial metallic atom), thickness of buffer and SiC layers of fuel particle were increased. This report describes the fuel behavior under high burnup condition in the irradiation test.

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).

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.

Prediction of Fuel Performance and Fission Gas Release Behavior during Normal Operation of the High Temperature Engineering Test Reactor by JAERI and FZJ Modeling Approach

In high temperature gas-cooled reactors (HTGRs), coated fuel particles are employed as fuel to permit high outlet coolant temperature. The essential feature of HTGRs is the role of the coated fuel particles acting as tiny containment and the principal barrier against radionuclide release under any operational and accident condition. For a safe operation of the High Temperature Engineering Test Reactor (HTTR), the continuous and reliable measurement of the coolant activity is required to allow an evaluation of the fuel performance and the radiological assessment of the plant during normal operation conditions. Since fission gases do not plateout on the inner surface of the primary cooling system, their concentrations in the primary coolant reflect the core-average through-coatings failure fraction and the fuel matrix contamination fraction. The main purpose of this report was to give a prediction of the fuel performance and fission product behavior in the HTTR under normal operating conditions by applying the calculation models as used at Japan Atomic Energy Research Institute and Research Center Juelich (FZJ), Germany, and to compare the results and methodologies. In the prediction of core average failure fraction, JAERI model gave much earlier failure and about two times larger than FZJ model. In the fission gas release prediction, FZJ model showed later increase of fractional release than JAERI model and rapidly increased towards the end of the operation.

Integrity Confirmation Tests and Post-irradiation Test Plan of the HTTR First-Loading Fuel

Since the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is the first mass-production High Temperature Gas-cooled Reactor (HTGR) fuel in Japan, their quality should be carefully inspected. For the quality control related to the fabrication process, Japan Atomic Energy Research Institute (JAERI) carried out the tests to certify the fuel integrity during operation. The tests comprise (1) as-fabricated SiC failure fraction measurement, (2) high-temperature heatup test of irradiated fuel and (3) accelerated irradiation test. For (1), the SiC failure fraction was measured independently in JAERI in addition to the measurement in the fabrication process. The measured failure fractions agreed within 95% confidence limit. In order to confirm the integrity of the SiC layer with respect to the 1,600°C criterion, the high-temperature heatup test of irradiated fuel compact was carried out. The result showed that no failed particle was present in the fuel compact after heating. The diffusion coefficient of metallic fission products in SiC layer was also examined in a series of post-irradiation heating tests. The measured diffusion coefficient of 137Cs showed a good holding ability as those obtained for research and development fuel specimen. The measured fission gas release rate in accelerated irradiation test showed no additional failure up to 60 GWd/t which was about two times higher than 33 GWd/t of the maximum burnup in the HTTR core. Through the tests, integrity of as-fabricated first-loading fuel of the HTTR was finally confirmed. The future post-irradiation test plan, which will be carried out to confirm the fuel irradiation performance and to obtain the data on its irradiation characteristics in the core, is also described.

Safety characteristics of the high temperature engineering test reactor

Abstract Various safety evaluations had been performed to confirm the validity of the design of High Temperature engineering Test Reactor (HTTR) facility taking into account the inherent safety features and characteristics in the design of the HTTR. It is shown that the reactor facility is so designed that (1) the integrity of fuel and reactor coolant pressure boundary is not damaged against the trouble of equipments, etc., during operation, (2) the influence of accidents including the rupture of reactor coolant pressure boundary, the reactivity initiated accident, etc. is not spread and (3) the release of radioactive materials under accidents is well mitigated.

Method and results of safety evaluation of the high-temperature engineering test reactor

A modular high-temperature gas-cooled reactor (MHTGR) is expected to be one of the best energy sources in the near future because it can supply high-temperature heat and have high thermal efficiency and sufficient safety features. The safety evaluation of the future MHTGR should be performed based on the experience obtained from the safety evaluation of the High-Temperature Engineering Test Reactor (HTTR). The safety evaluation of the HTTR was performed considering the specific safety design features of the HTGR and is applicable to the future MHTGR. Before the detailed safety evaluation of the future MHTGR, the safety evaluation method and results of the HTTR should be reviewed, and newly established acceptance criteria and methods for selecting evaluation events must be clarified. This paper describes in detail the method and results of the safety evaluation of the HTTR.

An Investigation of the Effects of Water Content on the Shielding Performance of the Primary Upper Shield in the High Temperature Engineering Test Reactor (HTTR)

High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor in Japan. The plant layout and radiation shielding are designed so that the plant can be operated without any employee receiving a high radiation dose rate. The primary upper shield of HTTR is composed of concrete (grout) and carbon steel. The function of the primary upper shield is to attenuate neutrons and gamma rays generated in the core to satisfy dose rate criterion for the operating floor. Since the HTGR uses high temperature helium as coolant, temperatures of shielding materials could be higher than that of conventional reactors. According to the analytical simulation, the maximum temperature inside of the primary upper shield during full-power operation was estimated to be about 85 °C. Since water content in the primary upper shield concrete depends on its temperature, shielding performance of the concrete under HTTR operational conditions need to be confirmed. Thus the water content of the concrete was experimentally investigated by out-of-pile heat-up tests to 175 °C. According to these tests, a water release model was developed. The out-of-pile heat-up tests showed that the water content in the concrete was larger than 78 kg/m3 up to 110 °C in a 95 % confidence limit.