Daisuke Tochio

Thermal Performance of Intermediate Heat Exchanger during High-Temperature Continuous Operation in HTTR

To use the High-Temperature Gas-cooled Reactor (HTGR) system as a commercial reactor similar to the future HTGR ‘GTHTR300C,’ supplying stable high-temperature heat from the reactor to the heat utilization system such as the hydrogen production system over a prolonged period should be demonstrated. The 50-day high-temperature continuous operation with full power, which is the first long-term operation with a reactor outlet coolant temperature over 900°C, was achieved in March 2010 in the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA). In order to supply stable high-temperature heat from the reactor to the heat utilization system using the HTTR, the heat exchange performance and structure integrity of the intermediate heat exchanger (IHX) should be confirmed. In this paper, the heat exchange performance and structure temperature of the IHX were evaluated with the high-temperature continuous operation data and compared with the designed one or estimated one. As a result, it was confirmed that the IHX can exchange stable high-temperature heat from the primary coolant to the secondary helium, and the IHX structure temperature is under the allowable working temperature in the operation. Moreover, it was confirmed that the heat exchange performance is almost the same as the designed one and that the structure temperature estimated with the analysis code is almost the same as the measured one.

Experiments and validation analyses of HTTR on loss of forced cooling under 30% reactor power

In a safety demonstration test involving the loss of both reactor reactivity control and core cooling, the high-temperature engineering test reactor (HTTR) demonstrates spontaneous stabilization of the reactor power. The test and analytical results of tripping one or two out of three gas circulators without reactor scram have already been reported. Moreover, the pre-analytical result of tripping all three gas circulators without reactor scram has been presented. On the other hand, the test and analytical results of tripping all three gas circulators without reactor scram are shown in this paper. About experiments, at an initial reactor power of 30% (9 MW), when all three gas circulators were tripped without reactor scram to reduce the coolant flow rate to zero, the fuel temperature did not show a large increase because the large heat capacity of the graphite core could absorb heat from the fuel in a short period. Moreover, the decay heat could be transferred through the graphite core and the reactor pressure vessel (RPV), emitted by thermal radiation from its outer surface and removed to the active vessel cooling system; therefore, the core at 9 MW was never exposed to the danger of a core melt, and the reactor power was stabilized spontaneously. About analyses, the reactivity performance is important for predicting the converging level of reactor power that affects the fuel temperature during a loss of forced cooling (LOFC) without reactor scram. With regard to thermal hydraulics, the performances of graphite heat conduction in the reactor core and thermal radiation from the RPV surface to the reactor cavity cooling system are crucial for predicting the temperature behavior of the fuel and RPV in the LOFC condition. It was confirmed that reactor kinetics coupled with heat transfer could be applied to reactor safety and accident analysis based on the comparison between the experiments and the analyses.

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, (I) thermal mixing behavior of helium gas in HTTR

The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.

Establishment of control technology of the HTTR and future test plan

The high-temperature engineering test reactor (HTTR) has carried out several demonstration tests to obtain basic data for the design of a future high-temperature gas-cooled reactor (HTGR). In addition, for commercial HTGRs, it also should be demonstrated that the HTGR system can supply stable heat to heat utilization system for the long term. To demonstrate the capability of HTGRs of stable heat supply, the operational experience of the HTTR would be useful. The control characteristics of the HTTR are evaluated by the result of the long-term high-temperature operation performed in 2010. As the result, it is confirmed that control technology of the HTTR is enough to achieve the control response with reasonable damped characteristics and the stable normal operation. In addition, a future possibility of the thermal-load fluctuation and lost test is examined. The data would contribute design of heat utilization system because heat utilization system is required to be a non-safety system. For the future HTGR system design, the requirement, method and result of determination of control constants are discussed. The behavior of control system is described with the characteristics of the HTGR during the past test using HTTR. In addition, this paper proposes a future test plan for demonstration of the reactor stability and control technology against thermal load fluctuation for future HTGR with a heat utilization system design.

Experience and Future Plan of Test Operation Using HTTR

The High Temperature Engineering Test Reactor (HTTR) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR is a graphite-moderated and helium gas-cooled reactor with thermal power of 30MW and the maximum reactor outlet coolant temperature of 950°C. Main objectives of the HTTR are to establish and develop HTGR technology and to demonstrate process heat application. The HTTR has conducted two test operations which are safety demonstration test and continuous operation. The safety demonstration tests focus on the verification of the inherent safety features of the HTGR that is the negative reactivity feedback effect of the core brings the reactor power safely to a safe and stable level without a reactor scram and the temperature transient of the reactor is slow in case of anticipated operational occurrences (AOOs). The safety demonstration tests include reactivity insertion test, coolant flow reduction test and loss of forced cooling (LOFC) test. Reactivity insertion test and coolant flow reduction test have been conducted since 2002. These tests demonstrated the inherent safety features of the HTGR in case of reactivity insertion and coolant flow reduction, and provided valuable data for code validation. LOFC test will start in the middle of 2010. LOFC is one of the important accident scenarios in the safety assessment of the HTGR. This test result will show extreme safety features of the HTGR and further improve the safety design approach of the HTGR. Obtained data can be useful to validate plant safety analysis codes. The continuous operation has been conducted to obtain plant data to establish HTGR technology and to demonstrate capability of the HTTR to supply stable heat to heat utilization system for long-term. Two operations of 30-day continuous operation in rated operation mode (in which designed reactor outlet coolant temperature of 850°C) and 50-days continuous operation in high temperature test operation mode (in which designed reactor outlet coolant temperature of 950°C) have been conducted so far. The 30-day continuous operation was achieved in 2007 and a good fuel performance to retain fission products within the coated fuel particle was clarified. The HTTR conducts 50-days continuous operation in 2010 to add useful operation data at high temperature to improve technical basis of HTGR and to realize high temperature heat application of HTGR.Copyright

Helium leak and chemical impurities control technology in HTTR

Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly. The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system. This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown.

Evaluation of maximum fuel temperature in HTTR

In order to ensure the thermal integrity of fuel in the high temperature engineering test reactor (HTTR), it is necessary that the maximum fuel temperature in the normal operation is to be lower than a thermal design limit of 1495°C. In the core thermal and hydraulic design of the HTTR, the maximum fuel temperature was estimated to be 1492°C, which satisfied the thermal design limit. However, the estimated temperature was derived by using hot spot factors with a large safety margin for the consideration of uncertainties in the design stage without the HTTR practical operating data, and thus there is no doubt that the estimated temperature includes excessive conservativeness. In order to obtain the maximum fuel temperature with appropriate conservativeness, the maximum fuel temperature has been re-evaluated on the basis of the HTTR operating data. In this paper, the random factors of the hot spot factors are revised by using the HTTR first fuel fabrication data, and the new maximum fuel temperature is estimated. As a result, the estimated maximum fuel temperature can be reduced to 1424°C. The reduction of the maximum fuel temperature leads to a larger thermal margin in nuclear and fuel designs.

A Safety Evaluation of HTTR Core Components Against 2011 Tohoku Earthquake

On March 11th, 2011, the 2011 Tohoku Earthquake which is one of the largest earthquakes in japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.Copyright

Preliminary Study on Thermal-Load Fluctuation Test Using HTTR

JAEA has designed and developed a commercial very high temperature reactor hydrogen cogeneration system named GTHTR300C. JAEA has investigated that a hydrogen production facility is treated as not a reactor grade facility but a general industrial plant grade facility. The reactor plant system should be designed absorbing the disturbance under AOO and becoming to steady state in the case that thermal-load of the facility is fluctuated or lost. Also, high-accuracy plant dynamics code should be developed to design the GTHTR300C.In order to demonstrate continuing the reactor operation in if the AOO is occurred in the hydrogen production facility and to obtain the validation data for high-accuracy plant dynamics code, the thermal-load fluctuation tests were planned using the HTTR in JAEA. In this study, it is concluded that the thermal-load fluctuation tests can be carried out without reactor scram and without modification of the HTTR facility and that the test data for validation of the code can be obtained sufficiently.Copyright

Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.Copyright