Future High Temperature Gas-cooled Reactor Research Articles

Fractographic studies and tensile strength of MPG-6 graphite based on uncalcined coke, neutron-irradiated at a high temperature

This paper presents the results of studies of the mechanical properties and fracture characteristics of MPG-6 graphite in the unirradiated state and neutron-irradiated at high temperature. MPG-6 is considered as a material for future high-temperature gas-cooled reactors (HTGR). In particular, MPG-6 is considered as one of the candidate materials for prismatic fuel blocks of the Russian experimental hydrogen-producing nuclear power plant with the VTGR high-temperature reactor currently under design.It is shown that MPG-6 graphite, based on uncalcined petroleum coke, which causes a small difference in the size of crystallites in areas with different preferred orientations of crystallites, has higher radiation stability of mechanical properties than GR-280 graphite based on calcined coke.

Development of the Prediction Technology of Cable Disconnection of In-Core Neutron Detector for the Future High-Temperature Gas-Cooled Reactors

Maintenance technologies for the reactor system have been developed by using the High Temperature engineering Test Reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for future high-temperature gas-cooled reactors (HTGRs) by shifting from time-based maintenance to condition-based maintenance. The technical issue of the maintenance of the in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because it is difficult to observe directly inside the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside the reactor using the time-domain reflectometry (TDR). The disconnection position, which was specified by the electrical method, was identified by nondestructive and destructive inspection. The accumulated data are expected to be contributed for advanced maintenance of future HTGRs.

Principle Design of Graphite Components for HTTR and R&D on Nuclear Graphite for HTGR in JAEA

Nuclear energy is one of the most promising energy sources to satisfy energy security, environmental protection, and efficient supply. The High Temperature Gas-cooled Reactor (HTGR) has attractive inherent safety features and it can be used as many kinds of heat applications such as hydrogen production, electricity generation, process heat supply, district heating and desalination. Many countries, especially developing countries, show their interests in HTGR. Graphite materials are used for the core components of the HTGR. IG-110 graphite, fine-grained isotropic graphite, with high strength and high oxidation resistance is used in the High temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) and High Temperature Reactor-Pebble-bed Modules (HTR-PM) in China. IG-110 graphite is a major candidate for the core graphite components of the Very High Temperature Reactor (VHTR) which is one of HTGRs and one of the most promising candidates as the Generation-IV nuclear reactor systems. This paper describes design of core components of HTTR and R&D on nuclear graphite for HTGR. JAEA established the graphite structural design code and inspection standard of graphite to construct the HTTR. JAEA developed an in-service inspection method and a draft graphite structural design code for future HTGR on the basis of the HTTR technologies. Moreover, JAEA are now developing the design data base of IG-110 graphite and IG-430 graphite including irradiation data for HTGR.

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.

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.

Operation and maintenance experience from the HTTR database

The Japan Atomic Energy Agency has been establishing a database of operation and maintenance experience for the High Temperature Engineering Test Reactor. The objective of this database is to share information from operation and maintenance experience and make use of the knowledge gained in the design, construction, and operation of future High Temperature Gas-cooled Reactors (HTGRs). Between 1997 and 2012, more than 1000 events have been registered in this database system.This paper describes trends in operation and maintenance events recorded in this database, including experience gained from the Great East Japan Earthquake. The paper also identifies the following significant items that are expected to be useful in the design of future HTGRs: (1) performance degradation of helium gas compressors, (2) malfunction of the reserved shutdown system in the reactivity control system, (3) problems with emergency gas turbine generators, and (4) consequences of the Great East Japan Earthquake.

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.

Structure of different grades of nuclear graphite

Owing to its low neutron absorption cross-section, large scattering cross section and thermal and chemical stability, graphite is a key component of operational nuclear reactors where it is used as a moderator, reflector and as major structural component for 90% of current UK nuclear plants. It is also of interest for use in developing the future high temperature gas-cooled reactors. The properties of the nuclear graphite are influenced by its structural characteristics, which change as a function of neutron irradiation, temperature and oxidation. The principal structural changes during neutron irradiation that affect the integrity and dimensions of nuclear graphite components, thereby affecting service lifetime, are that the a-axis contracts and the c-axis expands in the crystallites. Characterization of virgin graphite structure and of the damage evolution after irradiation of nuclear graphite has an important role to play in the understanding and development of materials used in current and future nuclear reactors, respectively.

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.

Numerical simulation of three-dimensional thermal-hydraulic behavior for HTTR (High Temperature Engineering Test Reactor)

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are now in progress in order to verify the inherent safety features and to improve safety designing and analysis technologies for future HTGR (high temperature gas-cooled reactor). Coolant flow reduction test is one of the safety demonstration tests for the purpose of demonstration of inherent HTGR safety features in the case that coolant flow is reduced by tripping of helium gas circulators. If reactor core element temperature and core internal structure temperature during abnormal events are estimated by numerical simulation with high-accuracy, developed numerical simulation method can be applied to future HTGR designing efficiently. In the present research, three-dimensional in- and ex-vessel thermal-hydraulic calculations for the HTTR are performed with a commercially available thermal-hydraulic analysis code “STAR-CD ®” with finite volume method. The calculations are performed for normal operation and coolant flow reduction tests of the HTTR. Then calculated temperatures are compared with measured ones obtained in normal operation and coolant flow reduction test. As the result, calculated temperatures are good agreement with measured ones in normal operation and coolant flow reduction test.

Safety design

JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R&D needs for establishing the safety philosophy for the future HTGRs are reported.

Development of a Simulation Model and Safety Evaluation for a Depressurization Accident Without Reactor Scram in an Advanced HTGR

It is important to use analyses to prove outstanding inherent reactor safety during a severe accident in order to convince the public and licensing authority of the safety advantage of the high-temperature gas-cooled reactor (HTGR). In this study, the simulation of a depressurization accident without reactor scram (DAWS) was performed for a future HTGR with 450-MW thermal output, introducing the annular core of pin-in-block-type fuel, which was originally designed in Japan. The DAWS has the possibility of becoming one of the severe accidents postulated in the HTGR. To perform an accurate simulation, a new analytical model for reactor dynamics and indirect decay heat removal from the surface of the reactor pressure vessel (RPV) during the DAWS was developed. The features of the new simulation model are as follows:1. A single-channel model is coupled with a two-dimensional reactor thermal model in the new simulation model. The reactor kinetics with a single-channel model during the DAWS is simulated taking into account heat removal from the reactor calculated in the R-Z reactor thermal model, including the RPV and indirect vessel cooling system. No conventional calculation codes with a single channel have a heat removal model from an RPV or were able to simulate precisely the transient during DAWS.2. A xenon buildup and decay model for the reactivity calculation is made in addition to one point-kinetics approximation to simulate a recriticality and a power oscillation following the initiation of the DAWS.3. A transient simulation can be performed for two kinds of core models of pin-in-block- and multihole-type fuels.The accurate evaluation of xenon density and core temperature is of prime importance in the simulation of the DAWS. From the simulation result with a proper safety margin, it was confirmed that the safety performance of passive decay heat removal with cooling indirectly from the surface of the RPV is outstanding for the DAWS, and a severe-accident-free HTGR can be designed. The newly developed code is applicable to the detailed safety evaluation necessary to future HTGR design.

Advanced Vessel Cooling System Concept for High-Temperature Gas-Cooled Reactors

The high-temperature gas-cooled reactor (HTGR) program will be attractive to a broad range of owner / operators and meet public acceptance if the future HTGRs would be completely free from accidents, which cause a significant release of radioactivity into the environment. An advanced vessel cooling system concept, in which there is no heat loss in normal operation and the decay heat is removed by the natural circulation of air in an accident, is proposed for the High-Temperature Engineering Test Reactor to meet this requirement. The depressurization accident, one of the severest accidents of the HTGR, is selected and the analysis shows no significant core heatup. Applicability to the future HTGR is also investigated.

高温ガス炉による高効率発電，（Ｉ） 高温ヘリウムガスタービンによる発電

The development of High-Temperature Gas-cooled Reactors (HTGRs) has a long history, and originated from the Dragon Project in 1959. After the achievement of the Dragon Reactor's criticality in 1964, the Fort-Saint Vrain Reactor in the U.S. and the THTR-300 in Germany were constructed and successfully operated.In Japan the High-Temperature engineering Test Reactor (HTTR) with a thermal output of 30MW is now under construction. Its main objectives are: fundamental or basic studies for high-temperature engineering, development for future HTGR's technology, and research on heat application to industrial purposes.In addition to this widely envisaged use of high-temperature heat, HTGRs with excellentsafety and easy operation features have a possibility of obtaining higher thermal efficiency, because of its high outlet temperature. In this sense, an HTGR system with a high-temperature helium-gas turbine is most promising. The maximum thermal efficiency of this system under the same operational condition of HTTR is expected to be about 45% or more.

Planning and Performance of Dynamic Experiments with the AVR Power Plant

In collaboration with Kernforschungsanlage Julich, Federal Republic of Germany, and other companies, dynamic experiments have been carried out with the Arbeitsgemeinschaft Versuchsreaktor (AVR) to test advanced dynamic computer models with the goal of using low-enriched uranium (LEU) fuel in future high-temperature gas-cooled reactors. Since LEU fuel has been used for the AVR since 1982, both experimental and theoretical behavior has been studied during the changeover from highly enriched uranium to LEU. The experiments comprise fast power transients that are initiated by either a fast control rod movement or a fast change of coolant flow. The neutron flux and other important parameters are registered in suitable time expansion.