Kazuhiko Kunitomi

JAEA'S VHTR FOR HYDROGEN AND ELECTRICITY COGENERATION : GTHTR300C

Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2030. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of . The intermediate heat exchanger (IHX) and the gas turbine are arranged in series in the primary circuit. The IHX transfers the heat of 170MW to the secondary system used for hydrogen production. The balance of the reactor thermal power is used for electricity generation. The GTHTR300C is designed based on the existing technologies of the High Temperature Engineering Test Reactor (HTTR) and helium turbine power conversion and on the technologies whose development have been well under way for IS hydrogen production process so as to minimize cost and risk of deployment. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the GTHTR300, IHX, etc. Also, the advantage of the GTHTR300C is presented.

Thermal-Hydraulic Characteristics of Coolant in the Core Bottom Structure of the High-Temperature Engineering Test Reactor

This paper discusses the high-temperature engineering test reactor (HTTR), a 30-MW (thermal) helium gas-cooled reactor being constructed by the Japan Atomic Energy Research Establishment. A thermal mixing study of the coolant in the core bottom structure (CBS) of the HTTR is conducted to clarify the thermal-hydraulic characteristics of the coolant and estimate the influence of a hot streak on the intermediate heat exchanger (IHX) and a pressurized water cooler (PWC) down-stream from the core. An experiment is carried out using an in-core structure test section (a full-scale simulation model of the (CBS) of the helium engineering demonstration loop (HENDEL), and a numerical analysis is made using a three-dimensional time-dependent flow and heat transfer code including a k-{epsilon} model of turbulence. It is confirmed that the coolant is mixed sufficiently in the CBS and the outlet gas duct of the HTTR, and the hot streak had little effect on the IHX and the PWC.

Concept of an inherently-safe high temperature gas-cooled reactor

As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of...

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.

Proposal of a plutonium burner system based on HTGR with high proliferation resistance

An innovative plutonium burner concept based on high temperature gas cooled reactor (HTGR) technology, “Clean Burn”, is proposed by Japan Atomic Energy Agency (JAEA). That is expected to be as an effective and safe method to consume surplus plutonium accumulated in Japan. A similar concept proposed by General Atomics (GA), Deep Burn, cannot be introduced to Japan because of its adopting highly enriched plutonium, which shall infringe on a Japanese nuclear nonproliferation policy according to Japan–US reprocessing negotiation. The Clean Burn concept can avoid this problem by employing an inert matrix fuel (IMF) and a tightly coupled fuel reprocessing and fabrication plants. Both features make it impossible to extract plutonium alone out of the fabrication process and its outcomes. As a result, the Clean Burn can use surplus plutonium as a fuel without mixing it with uranium matrix. Thus, surplus plutonium alone will be incinerated effectively, while generation of plutonium from the uranium matrix is avoided. High neutronic performance, i.e., achievement of burn-up of about 500 GWd/t and consumption ratio of plutonium-239 reaching to about 95%, is also assessed. Furthermore, reactivity defect caused by the inert matrix is found to be negligible. It is concluded that the Clean Burn concept is a useful option to incinerate plutonium with high proliferation resistance.

Evaluation of high temperature gas reactor for demanding cogeneration load follow

Modular nuclear reactor systems are being developed around the world for new missions among which is cogeneration for industries and remote areas. Like existing fossil energy counterpart in these markets, a nuclear plant would need to demonstrate the feasibility of load follow including (1) the reliability to generate power and heat simultaneously and alone and (2) the flexibility to vary cogeneration rates concurrent to demand changes. This article reports the results of JAEAs evaluation on the high temperature gas reactor (HTGR) to perform these duties. The evaluation results in a plant design based on the materials and design codes developed with JAEAs operating test reactor and from additional equipment validation programs. The 600 MWt-HTGR plant generates electricity efficiently by gas turbine and 900°C heat by a topping heater. The heater couples via a heat transport loop to industrial facility that consumes the high temperature heat to yield heat product such as hydrogen fuel, steel, or chemical. Original control methods are proposed to automate transition between the load duties. Equipment challenges are addressed for severe operation conditions. Performance limits of cogeneration load following are quantified from the plant system simulation to a range of bounding events including a loss of either load and a rapid peaking of electricity.

A STUDY OF AIR INGRESS AND ITS PREVENTION IN HTGR

Abstract A rupture of the primary piping in the helium-cooled and graphite-moderated high-temperature gas-cooled reactor (HTGR) represents a design-basis event that should not result in significant safety consequences. In such a loss-of-coolant event, the reactor would be shut down inherently, and the decay heat would be removed passively with the ultimate reactor temperature rise being less than the design limit. Still, an important concern for reactor safety continues to be graphite oxidation damage to the fuel and core should a major air ingress take place through the breached primary pressure boundary. Two major cases of air ingress are studied. The first case results from the rupture of a control rod or refuel access standpipe atop the reactor pressure vessel (RPV). To rule out the possibility of such a standpipe rupture, a design change is proposed in the vessel top structure. The feasibility of the modified vessel local structure is evaluated. The second case of air ingress results from the rupture of one or more main coolant pipes on the lower body of the RPV. Experiment and analysis are performed to understand the multiphased air ingress phenomena in the depressurized reactor. Accordingly, a new passive mechanism of sustained counter air diffusion is proposed and shown to be effective in preventing major air ingress through natural circulation in the reactor. The results of the present study are expected to enhance the HTGR safety and economics.

Conceptual design of a 50-MW severe-accident-free HTR and the related test program of the HTTR

The severe-accident-free high-temperature gas-cooled reactor (SFHTR) is a prototype design for a next generation reactor. It is suitable for widespread deployment by virtue of its inherent safety features and very long refueling interval. Furthermore, its inherent safety features can be demonstrated by full-scale tests. Many of these features may be demonstrated in the High-Temperature Engineering Test Reactor (HTTR). The SFHTR is designed to have the probability of a severe accident at least two orders lower than existing systems. The fuel will not exceed its failure temperature even in the event of complete loss of coolant or complete withdrawal of two control rods. A unique configuration of burnable poisons allows a fuel cycle of 16 yr and a burnup exceeding 120 GWd/t. This feature promises very high availability and good economics. We have designed two SFHTR systems. The larger one, called the MSFHTR, has a 450- to 600-MW thermal capacity and is intended for the production of hydrogen and electricity. The smaller SFHTR (SSFHTR) is intended for remote areas, off the electrical grid, for simultaneous production of electricity and desalinated water. The SSFHTR can produce 23.5 MW(electric) plus 40 t/h of water with a net efficiency of 47%. The HTTR is capable of conducting full-scale simulation testing of key SFHTR design features in order to confirm and extend the designs and as a first step in convincing the public and the licensing authorities of the validity of demonstrable inherent safety. Design features of a 50-MW SFHTR focusing on the safety concept, safety evaluation, and core design are described. In addition, an HTTR-based test-and-development program for the SFHTR is presented.

A SMALL MODULAR REACTOR DESIGN FOR MULTIPLE ENERGY APPLICATIONS: HTR50S

HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEAs 950°C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750°C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900°C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.