Yosuke Shimazaki

Investigation on structural integrity of graphite component during high temperature 950 °C continuous operation of HTTR

Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950 °C continuous operation, a high temperature continuous operation with reactor outlet temperature of 950 °C for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950 °C continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000 °C. In addition, the programs of surveillance test and ISI using TV camera were introduced.

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

Improvement of neutron startup source handling work by developing new transportation container for high-temperature engineering test reactor (HTTR)

ABSTRACT The high-temperature engineering test reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of 252Cf with 3.7 GBq, installed in NS holder and subsequently in a control rod guide block (CR block). The NSs are exchanged at the interval of approximately seven years. The NS holders are transported from the dealers hot cell to the HTTR using a transportation container. The loading work of NS holders to the CR blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Technical issues, which are the reduction and prevention of radiation exposure of workers and the exclusion of falling of NS holder, were extracted from the experiences in the past two exchange works of NSs to develop a safety handling procedure. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container. As a result, the NS handling work using the new transportation container was safely accomplished by developing the new transportation container which can reduce the risks of radiation exposure dose of workers and exclude the falling of NS holder.

Development of the maintenance technologies for the future high-temperature gas cooled reactor (HTGR) using operating experiences acquired in high-temperature engineering test reactor (HTTR)

This paper describes the lessons learned of the maintenance technologies, which have been and will be developed by using the high-temperature engineering test reactor (HTTR), which should be expected to apply to the future high-temperature gas cooled reactors (HTGRs). For example, the periodical maintenance for the future HTGRs is planned to carry out in each two years. The duration of periodical maintenance is planned about 60 days to satisfy the availability of operation up to 90%, in which decay heat removal, refueling and maintenances of equipment and components are carried out. As the key issue is to make a practical application of the technologies to the future HTGRs, shortening the duration of the periodical maintenance is expected to be important by excluding the maintenance related to the reactor equipment from the critical path by shifting the time-based maintenance, which is defined to be carried out in certain time interval, to the condition-based maintenance, which is defined to be carried out when a parameter exceeds its criteria, by using the experience and data acquired in HTTR, which is expected to make the maintenance interval significantly longer than that of the time-based maintenance as the operation experiences are accumulated. The maintenance technologies for the reactor system, which have been developed by using HTTR, are categorized as the following. (1) Establishment of the maintenance technologies specific to the HTGRs. (2) Development of the maintenance technologies for the future HTGRs. (3) Efficient maintenance works for the general equipment.

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

Dust Generation and Transport Behavior in the Primary Circuit of HTTR

Dust is to be limited in the primary circuit of the high temperature gas-cooled reactors (HTGRs) with regard to the reactor operational stability and structural integrity of the heat exchanger because the dust in the coolant adheres to the heat transfer pipe surface, and it is lowered with the performance of the heat exchanger. Furthermore, the dust, including the fission products (FPs) which adhered to the piping, must be reduced in order to be discharged in the depressurization accident with coolant helium.In High Temperature Engineering Test Reactor (HTTR), the rise of filter differential pressure of primary helium gas circulator (HGC) showed that the dust was arising in the coolant. The increase of differential pressure at the filter of the HGC for the primary pressurized water cooler (PPWC) in HTTR was observed from the beginning of the HGC operation. The differential pressure rose with the increase in operating time of the HGC, and the filter had to be exchanged the filter approximately in about every 8400 hours.Therefore, the source of the dust should be investigated using The Scanning Electron Microscope (SEM) and The X-ray Fluorescence (XRF) analyzer. SEM was used to observe the appearance of the captured dust in detail and to identify its shape and its size. This result clearly indicated that the dusts mainly composed of carbon. XRF was used to estimate the sources of the carbonaceous dust particles. It was clarified that the source of dust is the slide member of the compressor which is not of a primary cooling system.© 2012 ASME