Search

The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107 kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9 MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments.

The Tokamak Fusion Test Reactor (TFTR) at Princeton began operation in December 1982. While it has operated succesfully with protium and deuterium achieving record energy confinement time and record ion temperatures. TFTR's ultimate goal is to achieve the scientific break even point of Q = 1 for which deuterium/tritium injection is needed. This paper will discuss the design parameters resulting from using tritium as a fuel, the design and operating philosophies employed, the additional systems and equipment required, the effect on the heating, ventilating and air conditioning systems, the tritium monitoring system and the personnel training.

Introducing tritium into the plasmas of the controlled fusion program will require stringent control and processing of effluents from the vacuum system. The reduction in the amount of gas required for discharge cleaning of the vacuum vessel during D-T operations is discussed. A steady-state glow discharge at pressures of 5*10/sup -3/ torr (D/sub 2/) for periods of up to 24 h are currently performed as part of the procedure to prepare the TFTR (Tokamak Fusion Test Reactor) vacuum system for high-power pulsing. The present vacuum system, with a speed of 5000 L/s, would generate about 5 mol. of tritium-contaminated effluent per hour. Purification and real-time recycling of the discharge cleaning gas using a Stirling cycle cryopump (10 K to 20 K) in series with a turbopump would reduce the discharge cleaning gas requirements to a single gas filling. The usual charcoal coating on the low-temperature stage of the cryopump will be removed to prevent hydrogen isotope sorption pumping. Impurities such as CO, H/sub 2/O, CH/sub 4/, etc., will be cryopumped on the 80 K and 10-20 K cryo-surfaces. At the termination of glow discharge, the cleaning of a single charge (250 torr-L) of tritium contaminated gas will be transferred to a tritium cleanup system, possibly for recycling. The cryopump is then regenerated and the tritium-contaminated impurities are transferred to the tritium clean up system. >

The original design of the Tokamak Fusion Test Reactor (TFTR) and tritium handling are described. Modifications to these designs inspired by operational experience gained from running TFTR since 1982 with deuterium-deuterium (D-D)-fueled experiments are also presented. The presently installed gas handling systems work well for the D-D phase of operation, but they do not have the level of sophistication needed for tritium injection, tritium-contaminated-gas exhaustion, and gas detritiation for deuterium-tritium (D-T) operations. The pumping and gas-handling systems were designed prior to the operation of TFTR to handle the tritiated process streams anticipated during the DT phase. Since then, it has been found that certain changes and refinements to this design would make tritium handling simpler and safer. Most (90%) of the process gas (deuterium fuel) is pumped by the cryopanels in the neutral beams during operations. This nearly pure elemental hydrogenic gas is radiologically 25000 times more safe to handle than the corresponding amount of tritium oxide. It has been decided to transfer the hydrogenic species from the neutral beam cryopanels into special containers for off-site shipment as elemental tritium rather than as oxide. This reduces the exposure of oil-lubricated pumps from large contact with tritium-laden gas and reduces the risk of exposure to tritium oxide. Conditioning of the torus will be processed by impurity cryopumps, designed to pump only the nonhydrogenic species (impurities) and reduce the total amount of tritium used during conditioning. >

The program and facilities necessary for the safe handling of tritium at the Tokamak Fusion Test Reactor (TFTR) have undergone substantial evolutionary changes as the period of deuterium-tritium operation approaches. Details of these changes (which pertain to plasma-exhaust handling, water swamping, molecular sieve regeneration, premixed-gas composition, system venting, and tritium supply to neutral beams) are presented. Tritium and process flow are diagrammed, and equipment arrangements are given. >

The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium in deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility.

The TFTR tritium stack and area monitoring systems have been developed to provide the required level of reliability in a cost-effective manner consistent with the mission of the tritium handling system on TFTR. Personnel protection, environmental responsibility, and tritium-containing-system integrity have been the considerations in system design. A variety of monitors has been used. Real-time area and stack monitors are tied to HVAC control systems, and passive samplers provide a continuous effluent monitoring without AC power. It is concluded that the challenge of TFTR's radiological and electrical environment has been met. >

The Tokamak Fusion Test Reactor (TFTR) vacuum system is being upgraded to trap SF/sub 6/ gas which may leak from the neutral beam injectors during the planned deuterium-tritium experiments. This is necessary since SF/sub 6/ can, in 25% of a gram-mole, completely inactivate the platinum catalyst in the tritium cleanup system (TCS). Any SF/sub 6/ leaking into the injectors will initially be held on the liquid helium or liquid nitrogen cryopanels. During regens of the liquid helium panel, the deuterium and tritium will pass to the TCS, with the SF/sub 6/ being retained on the cryopanels. In the final phase of the accident recovery scenario, the cryopanels will be warmed to room temperature and the SF/sub 6/ collected by a liquid-nitrogen cooled cold trap. Design of the cold trap and measurements of the vapor pressure of SF/sub 6/ at 77 K as a function of surface coverage are presented. The SF/sub 6/ vapor pressure at 77 K, with only a few monolayers absorbed, was found to be >

The Tokamak Fusion Test Reactor (TFTR) at Princeton began operation in December 1982. Since then it has operated successfully with protium and deuterium achieving energy confinement time at peak electron density of 10/sup 19/ m/sup -3/s, with ion temperatures of 20 keV. This paper describes the systems and preparations required for D-T operation, i.e. introducing and operating the tokamak with tritium in order to achieve the scientific break even point of Q=1. These systems include the tritium storage and delivery system, the tritium injection systems, the tritium clean-up systems, and the plasma exhaust and collection systems. It is expected that TFTR will have these systems fully operational, with trained personnel.

The Tokamak Fusion Test Reactor (TFTR) is scheduled to begin DT operation in 1990 with the on-site tritium inventory limited to 5 grams. The physics and chemistry of the in-vessel tritium inventory will impact safety concerns, and also the entire operating schedule of the tokamak. We have investigated plasma-material interaction processes that will affect this first tritium-fueled tokamak. Tritium inventory estimates for TFTR are derived from: (1) laboratory simulation, (2) in-situ plasma measurements, (3) post-run surface analysis, and (4) modeling. This paper presents the results of these investigations, the derivation of a tritium inventory estimate and its uncertainties, and a discussion of its impact. A particular discharge-by-discharge operating schedule has been developed and evaluated. The major source of in-vessel tritium inventory will be codeposition of tritium and eroded carbon onto surfaces. We find that the on-site limit may be approached unless specific inventory reduction techniques are invoked, e.g., discharge cleaning.

Ontario Hydro is constructing a Tritium Removal Facility at its Darlington Nuclear Generating Station. The facility uses the Vapour Phase Catalytic Exchange and Cryogenic Distillation process to remove tritium from heavy water. This paper describes the background and design of Tritium Removal Facility.