Present status of energy in Japan and HTTR project

An amount of primary energy supply in Japan is increasing year by year. Much energy such as oil, coal and natural gas is imported so that the self-sufficiency ratio in Japan is only 20% even if including nuclear energy. An amount of energy consumption is also increasing especially in commercial and resident sector and transport sector. As a result, a large amount of greenhouse gas was emitted into the environment. Nuclear energy plays the important role in energy supply in Japan.Japan Atomic Energy Research Institute (JAERI) has been carried out research and development of a hydrogen production system using a high temperature gas cooled reactor (HTGR). The HTTR project aims at the establishment of the HTGR hydrogen production system. Reactor technology of the HTGR, hydrogen production technology with thermochemical water splitting process and system integration technology between the HTGR and a hydrogen production plant are developed in the HTTR project.     

A pilot test plan of the thermochemical water-splitting iodine–sulfur process

Research and development (R&D) of hydrogen production systems using high-temperature gas-cooled reactors (HTGR) are being conducted by the Japan Atomic Research Institute (JAERI). To develop the systems, superior hydrogen production methods are essential. The thermochemical hydrogen production cycle, the IS (iodine–sulfur) process, is a prospective candidate, in which heat supplied by HTGR can be consumed for the thermal driving load. With this attractive feature, JAERI will conduct pilot-scale tests, aiming to establish technical bases for practical plant designs using HTGR. The hydrogen will be produced at a maximum rate of 30 m³/h, continuously using high-temperature helium gas supplied by a helium gas loop, with an electric heater of about 400 kW. The plant will employ an advanced hydroiodic acid-processing device for efficient hydrogen production, and the usefulness of the device was confirmed from mass and heat balance analysis. Through design works and the hydrogen production tests, valuable data for construction and operation will be acquired to evaluate detailed process performance for practical systems. After completing the pilot-scale tests, JAERI will move onto the next R&D step, which will be demonstrations of the IS process to which heat is supplied from a high-temperature engineering test reactor (HTTR).     

Reactor technology development under the HTTR project

JAERI is developing HTR technology, hydrogen production technology, and system integration technology under the HTTR Project. The HTTR is the Japanese first HTR with a 30-MW thermal power. The first criticality of the HTTR was achieved in 1998, and the full-power operation at an outlet coolant temperature of 850°C was attained in 2001. The outlet coolant temperature was reached to 950°C in 2004. A seven-year program on the gas turbine HTR was launched in 2001. The program consists of the design of a GTHTR300 plant and R&D on a closed-cycle helium gas turbine system for the GTHTR300. It is designed to have a 600-MW thermal power at an outlet coolant temperature of 850°C and a 275-MW electric power. The objectives of the program are to establish a feasible plant design and to demonstrate key technologies for the helium gas turbine. The GTHTR300 design will demonstrate competitive economy and high degree of safety. It will also provide technology basis of VHTRs for power generation, hydrogen production, and cogeneration.     

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 high temperature structural materials for the HTGR

In accordance with the HTGR program in Japan, a series of R&D for high temperature structural materials in particular with respect to the HTTR design code has been performed in JAERI for more than 20 years. This paper introduces R&D results of the pressure retaining low alloy steel 2 1/4Cr-1Mo and the high temperature structural alloys Hastelloy XR and Ni-Cr-W superalloy for the design code together with some fruits of recent studies.     

Present status of the High Temperature Engineering Test Reactor (HTTR)

In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis.The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR.The installation permit of the HTTR was issued by the Government in November 1990.     

Achievement of Reactor-Outlet Coolant Temperature of 950°C in HTTR

A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30 MW and reactor-outlet coolant temperature of 950°C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950°C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950°C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.     

17O NMR Study on Dissolution Reaction of UO2 in Nitric Acid Mechanism of Electron Transfer

Japan Atomic Energy Agency (JAEA) started design studies of the thermochemical water-splitting iodine-sulphur (IS) process to be coupled with the HTTR to demonstrate hydrogen production from a very high-temperature reactor (VHTR) system. It is important from an economic point of view that a non-nuclear-grade, rather than nuclear-grade, IS process plant be built based on conventional chemical plant construction standards. In order to construct the IS process as a conventional chemical plant, some critical safety issues must been studied and clarified prior to the application for safety case review from the government. JAEA has launched R&;D for a non-nuclear-grade IS process to be coupled with the HTTR, which is the Japan's first VHTR capable of supplying 900°C secondary helium for process heat application. In this paper, we describe the development scenario for a non-nuclear grade hydrogen production system. Utilizing the HTTR-IS system as a reference system, the R&;D map is proposed for the VHTR-IS hydrogen production system.     

Partitioning and transmutation studies at JAERI both under OMEGA program and High-Intensity Proton Accelerator Project

At the Japan Atomic Energy Research Institute (JAERI), active and comprehensive studies on partitioning and transmutation (P&T) of long-lived nuclear waste from the reprocessing processes of spent fuel has been carried out under the OMEGA program. Studies at JAERI include a design study of dedicated transmutation systems both of an MA burner fast reactor (ABR) and an accelerator-driven subcritical system (ADS), and the development of a high intensity proton accelerator as well as the development of partitioning process, nitride fuel fabrication/dry separation process technologies and nuclear data studies.

During the course of studies, JAERI developed the concept of the double-strata fuel cycle, in which a dedicated system is used for transmutation. Comparing the various transmutation systems, such as thermal neutron spectrum or fast neutron spectrum systems, power reactors or dedicated systems, from the viewpoints of reactor physics, nuclear fuel cycle and socio-technical issues, it was concluded that the ADS is the best option for transmutation of minor actinide(MA). JAERI, therefore, decided to concentrate its R&D efforts on the development of ADS and related technologies.

One of the goals of R&D is to provide a basis for designing demonstration facilities of ADS, aqueous partitioning process and nitride fuel fabrication and dry separation technologies. As the initial step toward this purpose, the construction of an ADS experimental facility is planned under the High-Intensity Proton Accelerator Project which JAERI and the High Energy Accelerator Research Organization (KEK) are jointly proposing since 1998.

The paper discusses the some of the results of P&T studies and the outline of the High-Intensity Proton Accelerator Project under which ADS experimental facility will be constructed.     

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10⁻⁸ year⁻¹ which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10⁻⁸ year⁻¹, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.     

Study on control characteristics for HTTR hydrogen production system with mock-up test facility: System controllability test for fluctuation of chemical reaction

The Japan Atomic Energy Research Institute has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High-Temperature Engineering Test Reactor (HTTR). Prior to the coupling of a hydrogen production plant with the HTTR, simulation tests with a mock-up test facility of the HTTR hydrogen production system (HTTR-H2) is underway. The test facility is a 1/30-scale of the HTTR-H2 and simulates key components downstream from an intermediate heat exchanger of the HTTR. The main objective of the simulation tests is the establishment and demonstration of control technology, focusing on the mitigation of a thermal disturbance to the reactor by a steam generator (SG) and on the controllability of the pressure difference between the helium and process gases at the reaction tube in a steam reformer (SR). It was confirmed that the fluctuation of the outlet helium gas temperature at the SG and the pressure difference in the SR can be controlled within the allowable range for the HTTR-H2 in the case of the system controllability test for the fluctuation of chemical reaction. In addition, a dynamic simulation code for the HTTR-H2 was verified with the obtained test data.     

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.     

Research activities for accelerator-driven transmutation system at JAERI

Japan Atomic Energy Research Institute (JAERI) performs Research and Development (R&D) for accelerator-driven systems (ADS) for transmutation of long-lived nuclides. To study the basic characteristics of ADS, Transmutation Experimental Facility is proposed within the framework of the J-PARC project. The facility consists of two buildings, Transmutation Physics Experimental Facility to research the neutronics and the controllability of ADS and ADS Target Test Facility for material irradiation and partial mockup of beam window. A comprehensive R&D program for future ADS plant is also underway in three technical fields, 1) accelerator, 2) lead-bismuth target/coolant and 3) subcritical core.     

Investigation of absorption characteristics for thermal-load fluctuation using HTTR

A commercial very high-temperature gas-cooled reactor (VHTR) hydrogen cogeneration system named gas turbine high temperature reactor 300-cogeneration (GTHTR300C) is designed and developed in Japan Atomic Energy Agency (JAEA). Moreover, it has been planned that hydrogen production system and gas turbine system is connect to high-temperature engineering test reactor (HTTR). The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system like as the hydrogen production facility. Then, it is necessary to confirm that the thermal-load fluctuation could be absorbed by the reactor system so as to continue the stable and safety operation. The thermal-load fluctuation absorption tests using the HTTR were planned to clarify the absorption characteristics of the HTGR system. However, it is difficult to clarify the phenomenon due to many kinds of fluctuation in nuclear thermal power in the reactor core. Moreover, the actual data regarding how the delay of the temperature response is effective for the reactor system had been gained quantitatively.

The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the intermediate heat exchanger (IHX). The absorption characteristics of thermal-load fluctuation can be revealed with sufficient temperature fluctuation. So the tests were conducted with the primary coolant temperature 120 °C which is the start-up temperature of the HTTR. As a result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than the expected one and that the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from their results that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover, the calculation with the safety evaluation code based on RELAP5/MOD3 was performed. It was confirmed that the calculated temperature for the reactor is almost same to the measured one with the new analysis model. On the other hand, it was confirmed that the calculated temperature for the IHX decreased faster than the measured one due to smaller absorption capacity in the calculation model than that in actual one. It can be considered that the calculation for the IHX produces the conservative result. It was summarized that the safety evaluation code can represent the thermal-load fluctuation absorption behavior conservatively.     

Safety design consideration for HTGR coupling with hydrogen production plant

Safety requirements and design considerations are examined for a nuclear hydrogen production system that consists of High Temperature Gas-cooled Reactor (HTGR) and a hydrogen production plant by thermochemical water splitting iodine–sulfur process (IS process). Requirements in order to construct hydrogen production plants under conventional chemical plant regulation are identified in order to take into account a fundamental difference in safety philosophy between the nuclear facility and chemical plant and meet requests from the potential users of nuclear heat. In addition, safety requirements for the collocation of the nuclear facility and hydrogen production plant utilizing IS process (IS plant) are investigated. Furthermore, design considerations to comply with the requirements are suggested and the technical feasibility of the design considerations is evaluated. The evaluation results for a reference plant showed that safe distance determined by the chemical plant regulation against combustible gas and hazardous chemical leakages comply with the plant layout design. Furthermore, the results demonstrated the feasibility of IS plant construction under non-nuclear regulation by showing that the tritium concentration in IS plant can be maintained below the regulation limit and reactor normal operation can be achieved during abnormal conditions in the IS plant. These results clarified that design considerations suggested for coupling the IS plant to HTGR are reasonably practicable. The proposed criteria can be used not only for coupling hydrogen production plants but also for other chemical plants such as steam reforming plants, etc.     

R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC

Research and development(R&D) activities on partitioning and transmutation of trans-uranium nuclides (TRU) and LLFP and future R&D program in JNC were summarized. Feasibility design studies have been conducting to investigate the characteristics of a fast reactor core with TRU and LLFP transmutation. It was reconfirmed that the fast reactor has a strong potential for transmuting TRU and LLFP, effectively. R&D for establishing partitioning process of TRU apart from the high-level radioactive wastes have been carried out. By several counter-current runs of the TRUEX process using highly active raffinates, a process flow sheet capable of selective partitioning of actinides and fission products was newly developed. JNC has settled a new R&D program concerning partitioning and transmutation of long-lived radioactive waste based on recommendation of check & review for OMEGA program performed by the Ad Hoc Committee under the Atomic Energy Commission of Japan (AEC). The R&D program is composed of the design studies and development of element technologies (fabrication, irradiation) in the “Feasibility Studies” on commercialized fast reactor system and the basic studies with experiments (nuclear data, reactor physics, fuel property, etc.) to establish database and analytical tools for the TRU- and LLFP- containing fuel and core design.     

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H₂O, H₂SO₄ and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.     

Study on mo steel for high temperature gas-cooled reactor material characterization of forged low Si mo steel

The Japan Atomic Energy Research Institute (JAERI) has carried out a series of research and development work related to the high temperature gas-cooled reactor (HTGR) and, accordingly the high temperature engineering test reactor (HTTR) will be constructed in the near future. As the reactor pressure vessel (RPV) material, Mo steel will be used. Material characterization tests have been carried out to evaluate the applicability of the Mo steel for the RPV and to prepare for the licensing. The present paper summarizes the fracture toughness behavior including K_Id and K_Ia, irradiation embrittlement susceptibility and degradation of steel due to the long term aging at high temperature of the forged low Mo steel. These tests reveal good fracture toughness which well meets the requirements of the ASME Code, low neutron irradiation embrittlement susceptibility, little embrittlement by long term aging and so on. The present test results demonstrate good applicability of forged low Mo steel to the RPV of 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.     

Irradiation behavior of HTGR coated particle fuel at abnormally high temperature

Irradiation behavior of high temperature gas-cooled reactor (HTGR) coated particles under temperature transient conditions was investigated in accordance with a design-base accident scenario for HTTR, a 30 MWth HTGR under construction at JAERI. One of the scenarios predicts that the fuel temperature of the block-type fuel elements rises to abnormally high temperature by blocking a coolant channel with some foreign substance. For simulating this scenario the fuel compacts incorporating the coated particles were irradiated at normal temperature in three capsules, followed by temperature transient up to a maximum of approximately 2000°C. The post-irradiation examinations, including surface inspection, metrology, ceramography and a measurement of coated particle failure were applied to the fuel compacts to investigate the thermal-transient effect on the fuel integrity. Integrity of the fuel compact was also assessed by an estimation of tangential stress introduced into the compact by the temperature transient.