Shusaku Shiozawa

GTHTR300 design and development

In Japan Atomic Energy Research Institute (JAERI), successful development and operations of the High Temperature Engineering Test Reactor (HTTR) are coordinated by programs to study applied systems of the promising reactor technology. One such program being carried out from 2001 to 2007 is the design and development for the Gas Turbine High Temperature Reactor of 300 MWe nominal capacity (GTHTR300), aiming at demonstration of a prototype plant in 2010s in Japan. GTHTR300 features an original design of fuel cycle based on improved HTTR fuel element and of greatly simplified plant system in pursuit of economical performance with minimal development requirements. The fuel cycle characterizes on high burnup, low power peaking factor, and extended refueling interval. The plant system incorporates salient design simplification including conventional steel reactor pressure vessel, non-intercooled power conversion cycle, horizontal single-shaft turbomachine, and distributed modular maintenance of the overall plant. Planned research and development supporting detailed design efforts include performance tests of subscale components and control system. This paper describes the GTHTR300 reference design and related R&D activities in JAERI.

Research and development of HTTR coated particle fuel

The coated particle fuel has been developed within a framework of the HTTR (High Temperature engineering Test Reactor) Development Program at the Japan Atomic Energy Research Institute. The HTTR fuel is a prismatic block type containing TRISO-coated U02 particles. Research and development on the fuel has been progressed in three categories; a work for fuel production technology, a proof test of fuel performance and a safety-related research. In the present report the concept and outline of the fuel in the HTTR design are firstly described, and then fuel fabrication technology including recently developed methods for improving fuel quality is followed. Tests for proving fuel performance have been carried out extensively on the reference fuel of the HTTR design by irradiation in an in-pile gas loop and capsules, and typical results are presented in this report. Concerning the safety-related research, fuel failure and 137Cs release at abnormally high temperature are described.

Fabrication of the First-Loading Fuel of the High Temperature Engineering Test Reactor

The High Temperature Engineering Test Reactor (HTTR). which is the first high temperature gas-cooled reactor (HTGR) in Japan, attained its first criticality in November 1998. The fabrication of the...

A study of fuel behavior under reactivity initiated accident conditions — review

Results obtained in the 400 tests performed to simulate reactivity initiated accidents since 1975 in the Japanese Nuclear Safety Research Reactor, are described. Tests included the effects of cooling environment, defective fuel elements, fuel design parameters, the behaviour of fuel elements for various reactor types, all done for a wide range of energy deposition. Four types of basic fuel failure mechanisms have been established, and are discussed in detail: cladding melt failure, UO2 melt failure, high temperature burst failure and low temperature burst failure. Future test plans up to 1990 are out-lined and features requiring particular attention are pointed out.

Graphite core structures and their structural design criteria in the HTTR

Abstract The Japan Atomic Energy Research Institute (JAERI) is now proceeding with the construction design of the High Temperature Engineering Test Reactor (HTTR) to achieve first criticality in FY 1995. The reactor internal structures of the HTTR are made up of mainly graphite components. The characteristics of graphite are quite different in stress-strain behavior from metals, since the ductility of graphite is significantly less than metals. Therefore, the design codes provided for metal components cannot be applied directly to graphite components. The design criteria for the graphite core and core support components in the HTTR have been developed by JAERI and will be authorized by the Japanese licensing authorities. The design criteria were developed by partially modifying ASME Sec.III, Div.2, Subsection CE Code (draft) in the items of bi-axes failure theory, buckling limit and oxidation effects on the basis of test data. This paper describes the graphite core structures of the HTTR and the design criteria developed by JAERI and details the limits different from the ASME CE Code. A brief explanation is also made in this paper for quality control specified in the design criteria.

Performance Test Results of Mock-up Test Facility of HTTR Hydrogen Production System

For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m3/h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004.

Seismic response of the high-temperature engineering test reactor core bottom structure

This paper discusses the High-Temperature Engineering Test Reactor (HTTR) a 30-MW (thermal) helium gas-cooled rector that uses a prismatic block. The core bottom structure (CBS) of the HTTR consists of an arrangement of graphite components, and it supports the core elements within the reactor vessel. vibration tests are performed with two scale models to clarify the seismic response of the CBS. The vibration characteristics of the CBS are clarified quantitatively, and the structural integrity of the graphite components is confirmed.

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 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.

Integrity assessment of the high temperature engineering test reactor (HTTR) control rod at very high temperatures

The high temperature engineering test reactor (HTTR) is the first high temperature gas-cooled reactor (HTGR) in Japan with a reactor outlet coolant temperature of 950°C at high temperature test operation. The HTTR contains 16 pairs of control rods for which Alloy 800H is chosen of the metallic parts. Because the maximum temperature of the control rods reaches about 900°C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Under the guideline, temperature and stress analysis was conducted, and it is confirmed that the target life of the control rods of 5 years can be achieved.

Design and safety consideration in the High-Temperature engineering Test Reactor (HTTR)

The High Temperature engineering Test Reactor (HTTR) is a test reactor in Japan with thermal output of 30MW and reactor outlet coolant temperature of 950 °C at a high temperature test operation. This report describes features of the HTTR, emphases being laid on the safety design.