PRESENT STATUS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR (HTTR)

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1 PRESENT STATUS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR (HTTR) Shusaku Shiozawa * Department of HTTR Project Japan Atomic Energy Research Institute (JAERI) Japan Abstract It is essentially important to make efforts to obtain more stable but cheaper energy supply by extended use of nuclear energy. The High Temperature Gas-cooled Reactor (HTGR) is expected to be one of the best reactors as future energy source since it can provide high temperature helium gas for electricity generation and hydrogen production. Also, its outstanding safety features should be inevitable considering the movement of the recent social requirement for nuclear safety. Hence, efforts are to be continuously devoted to establish and upgrade HTGR technologies in the world. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950 C at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and was successfully finished with the achievement of the first criticality in 10 November 1998 by the Japan Atomic Energy Research Institute (JAERI). The rise-to-power test is now underway, attaining the full power of 30 MW at reactor outlet temperature of 850 C early 2001 year. The construction of the HTTR started in March 1991 and ended in After functional test operations in 1996 and 1997, the HTTR achieved first criticality on November 10, 1998, and low power physics tests were finished in January 1999, producing technical data useful for the development of future HTGRs. Then, the plant was shut down to modify a standpipe cooling system to reduce upper concrete biological shield temperature. In parallel, the rise-to-power test program of the HTTR has been defined by considering experience of previous Japanese test reactors, German AVR and US Fort St. Vrain with help of experts inside and outside JAERI. The first phase of the rise-to-power test was started with the plan of a series of the tests in September Unfortunately, the reactor was shut down by minor circulator trouble of the helium circulating system in October 1999, and thorough safety inspection for every system was conducted again since then. So far, most of the inspections were finished, and no major problems were found. The rise-to-power test will restart from April in 2000, and full * This paper is co-authored by: O. Baba, M. Ohkubo, S. Fujikawa, T. Iyoku, K. Kunitomi, and T. Kojima, Department of HTTR Project, Japan Atomic Energy Research Institute. 111

2 power operation will be attained in early After achievement of full power operation, High temperature test operation will be carried out to achieve the high reactor outlet coolant temperature of 950 C. In addition, safety demonstration tests, nuclear heat application tests and various kinds of irradiation tests are being planned. Introduction A High Temperature Gas-cooled Reactor (HTGR), which is a graphite moderated, helium cooled reactor, can supply high temperature heat as high as 1000 C which has a potential of obtaining high thermal efficiency as well as high heat-utilizing efficiency. It also has excellent features such as high inherent safety, easy operation and high fuel burnup. From the viewpoint of the global environmental protection and diversification of energy usage, non-electrical application of nuclear energy, hydrogen production for example, is very important. Therefore, in order to establish and upgrade the technological basis for HTGRs and also to use as a tool of basic researches for high temperature and neutron irradiation the Japan Atomic Energy Research Institute (JAERI) has been constructing a 30MWt High Temperature engineering Test Reactor (HTTR) at the Oarai Research Establishment. This report describes the present status of the HTTR. Present Status of HTTR Project Outline of HTTR Design The HTTR plant is composed of a reactor building, a machinery building and so on. The reactor building is 48 m by 50 m with two floors above the ground and three below (Fig. 1). Major components such as a reactor pressure vessel (RPV), primary cooling system components etc. are all inside a steel-made containment vessel. Air cooling towers for a final heat sink of the cooling systems are located on the roof of the reactor building. The reactor core is designed to generate 30 MW of thermal power and consists of array of hexagonal graphite fuel assembly so-called pin-in-block type fuel, control rod and graphite reflector. The core is supported by a graphite support structure and is tightened by a core restraint mechanism (Fig. 2). The block type fuel is adopted in the HTTR, instead of the pebble-bed type fuel (Fig.3). The biggest reason is that the HTTR is a test reactor for not only for HTGR technology development, but also a tool to accommodate irradiation spaces for innovative basic research in the field of high temperature engineering, which can be attained easily in the block type fuel rather than the pebble-bed one. It was also pointed out that the domestic own technology of fuel fabrication should be developed in selecting the pin-in-block type fuel and the technology of pebble-bed fuel fabrication might be imported from Germany as necessary. The design criteria specific to the HTGR fuel was examined and established in connection with the HTTR construction. The criteria is featured by the fact that the 112

3 fuel failure can not be defined and controlled for a single individual fuel particle, but whole core or group of the fuel assemblies, because the number of the fuel particles is as many as 14,000 particles in a fuel compact and amounts to totally 10 9 in the whole HTTR core. Therefore, the design criterion defines that the failure fraction of the coating layers of the fuel particles in as-fabricated fuel shall be allowed within a limit. For the same reason, a possible degradation of the coating layer during normal operation shall be allowed in some extent. This is quite different from the idea for the LWR fuels, in which not a single fuel rod failure is allowed in the fabrication and the fuel failure during operation is allowed up to the prescribed limit. In the HTTR design, the fuel failure during normal operation is defined on the basis of the current technology to occur when the wall-through failure fraction of the coating layers reaches 0.2%. Below this level, the reactor operation can be continued as an expected and allowable fission product release from the fuel to the primary circuit. This idea is supported by the HTGR fuel characteristics that no failure propagation is quite unlikely to happen and the fission product release from a single particle is negligibly small. During normal operation, radioactive level in the primary circuit is monitored to shutdown the reactor as necessary in prior to possible significant fuel failure propagation. Fig.1 Bird s eye view of HTTR building. The reactor cooling system is composed of a single main cooling system (MCS), auxiliary cooling system (ACS) and vessel cooling system (VCS) as schematically shown in Fig. 4. The MCS is operated in normal operation to remove heat from the core and send it to the environment via intermediate heat exchanger (IHX) of 10 MW and pressurized water cooler (PWC) of 20 MW in parallel or via only the PWC of

4 MW in case no IHX is in use. The coolant gas of helium comes in the reactor core at 395 C and flows downwards through the core, heated up to 850 C in normal operation and 950 C in high temperature test operation. The ACS is designed as a safety engineered feature to operate upon a reactor scram and cool down the core and the metallic core support components. On the other hand, the VCS works as a concrete cooling device used for biological shield in normal operation and acts as a passive cooling system upon a postulated accident of loss of forced convection cooling like pipe rupture of primary cooling circuit. The decay heat and residual heat are removed by the heat transfer from the RPV to the cooling panel of the VCS. No active cooling action on demand is necessary upon the accident. The major specification of the HTTR is summarized in Table 1. Fig.2 Table 1. Core structures of HTTR. Major Specification of HTTR Thermal power Outlet coolant temperature Inlet coolant temperature Fuel Fuel element type Direction of coolant flow Pressure vessel Number of cooling loop Heat removal Primary coolant pressure Containment type Plant lifetime 30MW 850 C/950 C 395 C Low enriched UO 2 Prismatic block Downward Steel 1 IHX and PWC (parallel loaded) 4MPa Steel containment 20 years 114

5 Fig.3 HTTR Fuel. Fig.4 Cooling system of HTTR. Safety Aspect of the HTTR Technology development activities for advanced HTGRs with passive safety features are underway in several countries. Advanced HTGR designs currently being developed are predicted to achieve a high degree of safety through reliance on passive safety features. Such design features should permit the technical demonstration of adequate public protection with significantly reduced emergency planning requirements. For advanced HTGRs this predicted high degree of safety derives from 115

6 (1) The ability of the ceramic coated fuel particles to retain the fission product under normal and accidental conditions, (2) The safe neutron physics behavior of the core, (3) The chemical stability of the core and (4) The ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. The passively safe design or inherent safe concept is taken in the HTTR as is usual in other current commercial HTGR designs. This design concept is widely understood to require that the core withstands a loss of forced circulation core cooling without sever fuel damage that would preclude subsequent operation and release of fission products in excess of regulatory guidelines without reliance on engineered safeguards including control rod insertion on demand. The reactivity control system and the core heat removal function were evaluated in the HTTR design to meet this requirement. Taking the passively safe design approach to the reactivity control system requires that sufficient negative reactivity can be inserted to ensure long-term hot or cold shutdown under any condition. Furthermore, movable metal clad control rods must not be degraded during any core heatup event. For both pressurized and depressurized losses of cooling, the fuel temperature is shown to remain well below the range of temperatures where coated fuel particles degradation occurs. Some of these safety aspects are to be demonstrated in the HTTR as will be shown below as Safety Demonstration Test, although this was confirmed by a safety analysis to be well reliable. Status of HTTR Construction Construction History in Brief In February 1989, JAERI applied for the permission of the HTTR construction to the Science and Technology Agency of Japan (STA) for the safety review by the Government. The safety review by Japanese Nuclear Safety Commission followed and was finished in November In parallel, the Graphite Structural Design Code and the High Temperature Structural Design Code were developed by JAERI. The Inspection Criteria for the fuel and the graphite components were also established by JAERI. These Codes and the Criteria were endorsed by the STA in December The construction of the HTTR was initiated in March 1991 and the excavation of ground was completed in May 1992, followed by the construction of the reactor building. The assembly and installation of the containment vessel were completed with success of its pressure-proof and leakage tests in November The reactor pressure vessel and the intermediate heat exchanger were installed in the containment vessel, followed by the installation of the cooling systems, reactor internals and others except fuel with completion in March The pressure-proof test and leakage test of the reactor cooling system was carried out with nitrogen gas at 6.0MPa. No leakage of the nitrogen gas was confirmed to 116

7 guarantee a design limit of helium gas leakage rate to the atmosphere of 0.28 percent of the inventory per day. Essential to the HTGR technology are to minimize the leakage of helium gas out of the cooling boundary and prompt detection of the possible leakage. Meanwhile, not a few leakages were observed in the very beginning of a functional test before fuel loading and adequate countermeasures were taken against this leakage. Also, a prompt and reliable detection of the possible leakage is under investigation by means of acoustic measurement. Here, it should be noted that a certain level of the helium coolant leakage is allowed in the HTTR design, therefore HTGR design, since a limited small leakage does not cause a serious problem from a safety point of view. Furthermore, absolutely no leakage is technically possible. Only problem, if any, is a radioactive material release coupled by the helium leakage, since a small content of fission products exist in the primary circuit. Thus, the controlling factor to determine the allowable limit of the helium leakage is the level of radioactivity around the circumference of the primary circuit. Establishing the technology of the helium closure and its prompt monitoring is one of the key technologies to be exploited through the HTTR operation. In parallel to the construction of the components mentioned above, the fuel fabrication started in June 1995 and completed in November The failure fraction of the coating layers was 100 times lower than the design limit of 0.2%. Fuel loading into the reactor core was started in May A special fuel loading machine is prepared which can replace one column by one column of fuel assemblies through a connection passage to a standpipe. The fuel assemblies are lifted up one by one from the core and moved into a rotating rack in the fuel loading machine and then fresh fuel assemblies are lowered into the core. Experience from Functional Tests before Fuel Loading Functional tests before fuel loading has been conducted to validate the design of the HTTR and confirm its performance. Several improvements for the system have been accomplished in terms of, for example, securing safety margin and easy operation. At the same time, some findings including unexpected troubles were revealed out through the tests, which were resolved or are to be resolved in the very near future. Among the troubles, the biggest was unexpected temperature increase of atmosphere in standpipes to accommodate a control rod drive mechanism and consequently temperature increase at a localized part of radiation shield concrete penetrated by the standpipes. This became evident during non-nuclear heatup test up to about 200 C by supplying heat with gas circulators. A posttest thermal analysis suggested the temperature increase to have occurred due to excessive helium bypass leakage upwards within the standpipe. The temperature of the atmosphere in the standpipes was successfully decreased by adding pressure balance holes in a shroud containing the control rod drive mechanism and by providing an additional seal to limit the bypass leakage. However, the temperature of the concrete shield was still higher than expected by calculation. According to an extrapolation to 400 C of reactor inlet temperature at 117

8 the rated power, the concrete temperature was evaluated to exceed 100 C, whereas the design expectation is less than about 90 C. It was suspected that the unexpected temperature increase was due to underestimation of heat input to the concrete via the standpipes penetrating the radiation shield concrete. Adequate countermeasures were taken of thermal insulator installed additionally on the circumference of the standpipes. Moreover, the standpipes were additionally equipped with copper tubes to enhance the conduction cooling to the VCS. Such countermeasures as these resulted in significant decease of the concrete shield temperature. Beside the unexpected temperature increase mentioned above, some troubles were experienced in helium leakage, control systems, gas circulators, purification system etc. Those were not serious either and thought to be usual troubles that appears due to the first of kind nuclear reactor to be built in Japan. Fuel Loading, First Criticality test and the Rise-to-power Tests Fuels were loaded into the core since July 1998 and the first criticality was attained on November 10, Fuels were column-wise loaded from the outer fuel column to the inner one and 19 of the total 30 fuel columns made the reactor criticality. The first criticality was attained almost in a shape of annular core, installing the outer ring of fuel columns and one inner fuel column and leaving the other inner ones filled with dummy fuels. The predicted minimum number of fuel columns attained the first criticality was 16±1. The discrepancy between the predicted and the measured values was due to impurity and air contents in the dummy fuel block used for the calculations. The post calculation with revised graphite block composition showed that the number of critical fuel column was 19. The other inner fuel columns were filled in to give the full size core by December In the course of fuel loading, low power physics tests were conducted for the 21, 24 and 27 fuel column loaded core. These tests provided useful data for designing future annular cores of advanced HTGR. The rise-to-power tests was begun on September 28,1999. Unfortunately, the reactor was shut down by the cause of a malfunction of the gas circulator control system on October. Immediately, safety check and countermeasures to prevent from a malfunction were conducted by November After that, the annual inspection such as overhauling pumps, leakage testing primary coolant pipe, checking and calibrating the instrumentation and control panel and so on will be carried out by April The raise to power tests will restart in April 2000 and will be expected to attain the full power by the first half Future Plan The HTTR Project is an overall national project for HTGR technologies and its application. This section introduces the future plan to be done within a framework of the HTTR Project. Establishment of HTGR Technologies 118

9 Accumulation of HTGR Operation Experience The highest prioritized objective of the HTTR Project is to establish HTGR technologies, which will be almost attained upon the completion of the HTTR construction and its commissioning within one and half years from now. The rest will be obtained through an operational experience of the HTTR. A long-term rated power operation at 850 C of reactor outlet temperature is scheduled after the commissioning test, followed by high temperature test operation at 950 C. During the long-term rated power operation, basic data concerning HTTR performance is to be accumulated on the focus of fission product release in the primary circuit and function of purification system, helium gas leakage from the primary circuit and identification of the leakage place if significant, general performance of high temperature components such as the gas circulators, intermediate heat exchanger, pressurized water cooler and vessel cooling system. Data specific to nuclear technology will be also collected through the reactor operation concerning operability and controllability of the reactor, amount of radioactive waste, as well as human engineering data for operators. The data obtained here will be expected to be useful for the future HTGR development. Especially, as will be stated below, the operational experience obtained here is to be utilized to improve the economy of commercial HTGRs, by, for example, simplifying safety-related equipment, decreasing the management effort for the rector operation and maintenance, increasing the availability of the reactor operation, minimizing waste management etc. Upgrade of HTGR Technologies Evaluation of Reactor Performance Based on the HTTR operational and test data, the HTGR reactor performance is to be evaluated and analytical computer codes will be developed for predicting realistic rector performance under steady state and operational transient conditions. The evaluation is focused on (1) Core physics in relation with thermal response and control system, (2) Thermal analysis for fuel, reactor internals and high temperature components, (3) Fuel performance on fission product release and degradation of the coating layers to contain the fission products, (4) Structural integrity of reactor internals and high temperature components, (5) Decay heat and residual heat removal characteristics etc. The fruits from here are expected to be utilized for the future commercial HTGR design underway in South Africa, United States, Germany, the Netherlands and so on, as well as Japanese designed advanced HTGRs. Among those listed above, the most important are fuel performance and decay heat removal characteristics, since those are key factors to finally assure the reactor 119

10 safety. For the fuel performance, peak fuel temperatures and its distribution in the core will be calculated and compared with the data to accurately predict the fuel temperature. Also, the fission product release from the fuel in the primary circuit is measured to well predict the failure fraction of the coating layers and fission product release. For decay heat removal characteristics, the cooling ability of the VCS is precisely evaluated in relation with the establishment of the passive cooling system. Safety Demonstration Test It is well known that the HTGRs has inherent safety features, characterized by no risk of reactor core meltdown even in case of no forced cooling systems functioned with reactor shutdown failed. This is generally confirmed by analyses. The reactor power rapidly decreases after a postulated depressurized accident, which is featured by a sudden pipe rupture of the primary circuit. The decrease of the power is due to strong negative feedback by the Doppler effect. It is also shown that the peak temperatures of fuel and reactor pressure vessel do not exceed the limit of 1600 C and 550 C, respectively. A similar phenomenon is calculated in other commercial HTGRs designed in the USA, Germany, South Africa and so on. However, it is of great importance and the best way for the wide public acceptance to demonstrate such an inherent safety of the HTGRs using an actual HTGR. It is, therefore, planned in the HTTR to conduct a safety demonstration test under simulated accidental conditions. In the most extreme test to be planned in the final stage, all the forced cooling systems including the vessel cooling system are stopped under a rated power and the position of the control rods kept as it is. It is hoped that the public can really understand inherent safety features of the HTGRs from this test and come to show an interest in introducing HTGRs in vicinity of their living district Development of Key Components for Advanced HTGRs From a technical point of view, most essential to the development of the future advanced HTGRs is to develop high quality of fuel, since it is the most important parameter to determine the plant capability on rector power and safety. For example, high quality fuel makes high power density of the core possible, consequently increasing the reactor power in one unit and improving the economy of the HTGRs. As well, high quality of fuel to hold the capability of fission product release retention at elevated temperatures can assure higher degree of reactor safety against possible core heat up accidents. Furthermore, the high quality of fuel can significantly decrease the amount of the radioactive material release, resulting in the easy-maintenance and low cost for the waste management. In the HTTR, an effort to develop such high quality of fuel as zirconium-carbide coated fuel will be taken through irradiation test using the HTTR together with related out-pile-experiments. Beside the fuel development, advanced graphite materials for control rod clad material and core support component are explored to be developed through the HTTR experiment. With respect to the high temperature component, advanced intermediate 120

11 heat exchanger to enhance the heat transfer and minimized the size of the component is under planning in the HTTR Project. A spin-off effect is expected in the development of the advanced graphite and the intermediate heat exchanger. Concluding Remark The HTTR is a high temperature gas cooled test reactor, which has various aims and operational modes. The construction of the HTTR has progressed rather smoothly and the achievement of the first criticality was successfully finished in 10 November The rise-to- power tests will restart in April 2000 and will be expected to attain the full power by the first half The various tests by the HTTR will make a great contribution to confirm salient characteristics of HTGRs including reliable supply of high temperature heat as high as 950 C and high inherent safety and the application of high temperature heat from HTGRs to various fields will also contribute to relax global environmental problems. Furthermore the HTTR has a unique and superior capability for carrying out high temperature irradiation tests not only for development of advanced HTGRs but also for basic researches. The HTTR is highly expected to contribute so much to promoting the international cooperation in these fields. 121