Tetsuo Nishihara

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.

Development Scenario of the Iodine-Sulphur Hydrogen Production Process to be Coupled with VHTR System as a Conventional Chemical Plant

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 Japans 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.

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

The tritium concentration in the hydrogen product in Japans future very high temperature gas-cooled reactor (VHTR) system coupled with a thermochemical water-splitting iodine-sulfur (IS) process (VHTRIS system), named GTHTR300C, was estimated by numerical analysis. The tritium concentration in the hydrogen product significantly depended on undetermined parameters, i.e., the permeabilities of a SO3 decomposer and a H2SO4vaporizer made of SiC. Thus, the estimated tritium concentration in the hydrogen product for the conservative analytical condition ranged from 3.4 × 10−3 Bq/cm3 at STP (38 Bq/g-H2) to 0.18 Bq/cm3 at STP (2,000 Bq/g-H2). By considering the tritium retained by core graphite and the reduction in permeation rate by an oxide film on the heat transfer tube of the IHX and the HI decomposer, the tritium concentration in the hydrogen product decreased to the range from 3.3 × 10−5 Bq/cm3 at STP (0.36 Bq/g-H2) to 5.6 × 10−3 Bq/cm3 at STP (63 Bq/g-H2), which were smaller than those for the conservative analytical condition by factors of about 3.2 × 10−2 and 9.6 × 10−3, respectively. The effectof the helium flow rate in the helium purification system on the tritium concentration in the hydrogen product was also evaluated.

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

Uranium-based TRU multi-recycling with thermal neutron HTGR to reduce environmental burden and threat of nuclear proliferation

ABSTRACT To reduce environmental burden and threat of nuclear proliferation, multi-recycling fuel cycle with high temperature gas-cooled reactor has been investigated. Those problems are solved by incinerating trans-uranium (TRU) nuclides, which is composed of plutonium and minor actinoid, and there is concept to realize TRU incineration by multi-recycling with fast breeder reactor. In this study, multi-recycling is realized even with a thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium and natural uranium are enriched and mixed with recovered TRU to fabricate fresh fuels. The fuel cycle was designed for a gas turbine high temperature reactor (GTHTR300). Reprocessing is assumed as existing reprocessing with four-group partitioning technology. As a result, the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for high-level waste can be reduced by 99.7% compared with the standard case. Surplus plutonium is not generated by this cycle. Moreover, incineration of TRU from light water reactor cycle can be performed in this cycle.

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.