Tritium-processing operations at the oak ridge national laboratory with emphasis on safe-handling practices

Nuclear Instruments and Methods in Physics Research A282 (1989) 329-340 North-Holland, Amsterdam

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Section X. Tritium targets and neutron production TRITIUM-PROCESSING OPERATIONS AT THE OAK RIDGE NATIONAL LABORATORY WITH EMPHASIS ON SAFE-HANDLING PRACTICES E.H. KOBISK * *, D.W. RAMEY, W .S. AARON, J.A. TOMPKINS, K.W. HAFF, J.R. DEVORE and H.L. ADAIR Chemical Technology Division, Oak Ridge National Laboratory P.O. Box 2008, Oak Ridge, Tennessee 37831, USA

1. Introduction The ever-increasing world inventory of tritium mandates improvement of handling procedures to prevent accidental releases and/or personnel exposures. Tritium releases to the environment can arise from several sources: 1) tritium production ; 2) nuclear weapons testing; 3) natural production by the interaction of cosmic radiation with nitrogen nuclei ; 4) reactor operations and nuclear fuel reprocessing; and 5) disposal of consumer products containing tritium . The natural production of tritium by cosmic rays presently accounts for most of the world's inventory of tritium (2.59-5.18 EBq in 1985); however, fusion power reactor programs will probably increase tritium inventories by about 3.7 EBq [1]. Uncertain long-term effects to processing personnel or the general population resulting from exposure to tritium in the environment necessitate improvement in tritium production and/or processing procedures to minimize tritium releases. Tremendous progress has been made in the handling of tritium since it was first produced by Luis Alvarez in 1936. At the Oak Ridge National Laboratory (ORNL), considerable improvement in tritium-handling procedures has resulted from developments in the Isotope Program . Specific ORNL programs involving tritium will be described with emphasis on the safe-handling practices as developed and used in these programs. 2. ORNL Isotope Program tritium operations 2 .1. Target fabrication

Since the early 1960s, the Isotope Research Material Laboratory (IRML) at ORNL has been preparing tri' Research sponsored by the Division of Basic Energy Sciences and Division of Nuclear Sciences, U.S. Department of Energy, under contract DE-AC05-84OR21400 with the Martin Marietta Energy Systems, Inc . " Retired . 0168-9002/89/$03 .50 © Elsevier Science Publishers B.V. (North-Holland Physics Publishing Division)

tium targets that are used to produce intense beams of 14.5-MeV neutrons by the (d, t) reaction. These intense beams of neutrons are used in programs involving cancer research, materials evaluation, materials identification, and nuclear physics research. Tritium-containing targets from 1 to 50 cm diameter having up to 220 TBq of tritium have been prepared for these applications . The fabrication of tritium targets in the IRML facility usually requires the use of large quantities of tritium. Safe handling of such large quantities of tritium is approached in a slightly different manner than the handling of most other radioactive materials . For the most part, radioactive materials are handled in glove boxes or hot cells which are connected to an off-gas system that produces a negative pressure inside the box or cell with respect to the ambient pressure of the laboratory in which the box or cell is located. This practice is still followed for very small quantities of tritium (_< 3.7 TBq). However, for larger quantities of tritium, this practice usually results in a considerable amount of tritium exposure to operating personnel because of tritium diffusing into the laboratory through plastic bags or rubber gloves which normally cover glove box ports. The present IRML facility that has been in operation for approximately 20 years was designed for handling quantities of tritium (37 PBq) with minimum personnel exposure. The metal prefabricated building (9.1 m X 6.1 m) was designed specifically for housing tritium and deuterium target fabrication equipment ; an interior view of the facility shows a 7.3-m-long stainless steel hood (fig. 1). The purpose of the hood is to house and ventilate the tritium-handling systems by air sweep. Six Plexiglas doors permit easy access to the work area, and a continuous upward air sweep is maintained in the hood. Continuous air supply to the building is provided by a forced air system in conjunction with electrical heaters and cooling equipment to maintain a tolerable operating temperature for personnel . In the event of failure of this forced air supply, air is brought from outside the building through filters at a rate governed by the hood exhausts system . Minimum airflow velocity across the X. TRITIUM TARGETS

E.H. Kobisk et al. / Tritiumprocessing operations at ORNL

330

HVAC TRANSFORMER

WATER FOUNTAIN LAVATORY BALANCE TABLE TRITIUM MONITOR/" ALARM

URANIUM EVAPORATO FILTER'

I

O

0

I

E EC TRON ^ BEAM POWER UP PL

MOBILE LARGE-TARGET LOADING STATION

CONTROL CONSOLE

50-cm TRITIUMSYSTEM CONTROL CONSOLE

TITANIUM EVAPORATOR CONTROLCONSOLE

10-INCH TRITIUM SYSTEM CONTROL CONSOLE

TITANIUM EVAPORATOR

VACUUM PUMP CABINETS

WATER HEATER

PRESS

DESK

STACK TRITIUM MONITOR

0

TRITIUM MONITORIALARM

HOT SINK

STACKFAN MONITORALARM

PTO STACK

Fig. 1 . Interior schematic of IRML tritium facility.

face of the hood has been maintained and measured at > 45 m/min with one hood door open ; with all doors closed, the flow across the vents at the base of the hood exceeds 30 m/min . The hood exhausts to a stack that is 4 m high . The effective stack height with an air sweep velocity out the stack of 19.4 m/s is 16.5 m. Operating personnel who worked inside the hood are required to work in airsuits, and those who work on the tritium sorption system (but from in front of an open door) are required to wear fresh-air masks and protective clothing . The main hazard of tritium to personnel is by body tissue absorption through skin contact or inhalation . Tritiated water (HTO or T20) is more rapidly absorbed by body tissue and, as a result, is approximately 10 ° times more of a biological hazard [2] than molecular (gaseous) tritium. Exterior surfaces exposed to tritium gas will exchange with the surface layer moisture to form HTO, and these contaminated surfaces must be routinely handled during all tritium work. As a result, the potential routine personnel exposure in the tritium facility is much greater through the handling of tritium-contaminated surfaces than through the accidental breathing of tritium gas. The tritium-processing equipment was designed in a long, narrow configuration to permit access for nearly all operations and maintenance from the hood doors. On the rare occasions when personnel actually have to

enter the hood area, airsuits are used to minimize exposure to tritium contamination in the hood. Most operating and maintenance procedures are conducted within arm's length of the hood doors allowing personnel to remain outside the hood area. In this case, personnel wear an extra layer of protective clothing, latex gloves, and a fresh-air-supplied full face mask . For simple operations such as manipulation of valves located directly in front of the hood doors, gloves are the only protective clothing that is necessary, since personnel stay completely outside the hood area and the highvelocity airflow into the hoods ensures that any tritium contamination is flowing away from personnel . The air in the personnel operating area of the IRML tritium facility is continuously monitored by two tritium monitors which alarm at 0 .26 MBq/m3 . One monitor is located in the general lab area while the other is located above the hood enclosure and is equipped with an intake hose that is positioned over whichever hood door is opened for access to the operating equipment. When the hood doors are closed, this monitor samples the air immediately above the hood door. Since tritium is lighter than air, this position provides the earliest detection capabilities. At ORNL, tritium air monitors are calibrated by Health Physics personnel on a quarterly basis . In addition to protecting personnel, minimization of tritium to the environment is an important considera-

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E.H. Kobisk et al. / Tritiumprocessing operations at ORNL

tion in the operation of the IRML tritium facility . For many years, a tritium stack monitor consisting of a catalytic furnace and a silica-gel sample collection chamber was operated by building personnel to provide incremental tritium release data. In recent years, the operation of the stack sampling equipment has been the responsibility of the ORNL Department of Environmental Management. In the past year, a dual-ionization, real-time tritium monitor giving a readout of WCi/m3 (3 .7 X 10 6 Bq) and an anemometer were installed in the stack. A new catalytic furnace and collection system for cumulative tritium release sampling were also installed . Efforts to calibrate this system and correlate sample measurements to total tritium emission are in progress, but they have been complicated by the very low levels of release from the IRML tritium facility . The fact that operations result in periodic small release (bursts) rather than a steady release rate further complicate efforts to interpret emissions data. Equipment used in the fabrication of tritium targets is shown in figs. 2 through 4. A base metal layer into which tritium gas is sorbed is prepared by evaporation-condensation in an oil diffusion-pumped vacuum system (fig . 2). The system is pumped with a 2000-1/s liquid-nitrogen trapped oil diffusion pump . An interior view of the system is shown in fig . 3 . The base metal layer is usually titanium metal evaporated using an electron-beam heat source and condensed onto a rotating substrate that is heated to 675-700 K. The vacuum level during the base metal layer evaporation is approximately 2 .1 X 10 -4 Pa . The tritium sorption chamber (having a volume of 69 .25 1) is pumped with two 270-1/s ion pumps that can maintain a vacuum level of _< 1 .3 X 10 -6 Pa during the substrate heating cycle and before the tritium is expanded into the chamber. Since each target will contain approximately 0 .22 PBq of tritium, a large tritium inventory must be maintained . The tritium is contained in large uranium-containing storage traps that have a total capacity of approximately 24.6 PBq. Approximately 3 .33 PBq of tritium are needed in the sorption chamber to maintain the necessary tritium pressure of 400 mm . Operating procedures require sampling of the system vacuum manifold with a quadrupole mass analyzer prior to any operations that could result in a substantial release from the systems (fig . 4) . If analysis indicates a high level of 3 H in the system, that level is reduced by further absorption of tritium onto the uranium traps. It has been routinely found that even when large amounts of tritium are being moved throughout the system, the residual gas is primarily 3 He and tritium in the form of methane . When the mechanical roughing pump is used to evacuate the system, this residual gas is exhausted to the stack. Typically, no more than 1 .1-1 .3 TBq are released in this procedure ; and when smaller quantities of tritium are used in the particular preparation, such as

Fig. 2 . Oil diffusion vacuum system used for preparing base metal layers for tritium targets.

Fig. 3 . Interior of oil diffusion vacuum system. X . TRITIUM TARGETS

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E.H. Kobisk et al. / Tritium processing operations at ORNL NCV

go

CHAMBER VENT VALVE

CHAMBER

ION PUMPS

ROUGHING VALVE

ION PUMP

ROUGHING PUMP

ION PUMP

Fig. 4. A schematic of tritium sorption system.

the tritium-trick helium doping process described below, emissions are even lower. More detailed information on tritium targets prepared in the IRML facility is contained in another publication [3] . 3. Tritium trick helium doping The IRML tritium target facility described above is also used for implanting 3 He in alloy specimens . Although this technique involves the handling of relatively small quantities of tritium, usually 5 37 TBq, it provides a convenient means of studying the mechanism of helium embrittlement without irradiation and provides a rapid screening method in the development of embrit-

tlement-resistant alloys. Experience obtained in processing several vanadium alloys implanted with various levels of 3 He should be useful to those who wish to use this technique . One aspect of controlled thermonuclear reactors (CTRs) that will require extensive research is the change in the mechanical properties of first-wall structural materials caused by the high helium concentrations expected to be generated in these materials exposed to the high flux of 14.5-MeV neutrons. Techniques for simulating and exploring the effects of helium are limited by the fact that helium is practically insoluble in metals. Elaborate methods of implantation, such as alpha bombardment, are confined to thin, impractical test specimens . An alternative helium-charging method using

EH Kobisk et al. / Tritium processing operations at ORNL tritium to deposit 3 He by radioactive decay has been shown to be feasible [4] . This so-called tritium trick has the capability of implanting 3 He in engineering-size specimens at a relatively high rate . Materials to be implanted with helium are allowed to sorb tritium at an elevated temperature . The specimens are then cooled to room temperature . While at room temperature, a calculable fraction of the sorbed tritium (- 0 .5% per month) will decay to 3 He. When the desired level of helium is achieved by the elapse of a predetermined decay time, the remaining tritium is removed by heating in a vacuum environment leaving the relatively immobile 3 He deposited in the metal lattice . Several vanadium alloys have been subjected to an improved version of the tritium trick . The procedure described by Remark et al . [5] was modified to include wrapping the specimens in tantalum foil to minimize oxygen contamination . Instead of permitting tritium decay at the low temperatures, sample temperature was maintained at 675 K to prevent vanadium tritide formation and to produce a 3 He bubble distribution similar to that produced during elevated temperature irradiation. Vanadium tritide formation is not desirable because the phase change associated with this reaction produces material stresses not typically experienced in CTR applications . The system used to charge the vanadium alloys with tritium is schematically presented in fig . 4 . This all-metal system can be used to charge alloys with tritium at pressures up to 0 .1 MPa . The rate of sorption is dependent on temperature, pressure, tritium solubility (specific to each material), and specimen surface condition . Tritium sorption can be attenuated by a poorly activated condition of the specimen surface. Thus, close attenuation was given to good prevacuum cleaning of the specimens and the tantalum wrapper. After ultrasonically cleaning the specimens and tantalum wrapper in methanol . The specimen bundles are triple wrapped in the - 50-[m-thick foil . Samples are then placed in the tritium-trick furnace and outgassed for approximately 16 h at 673 K under - 1 WPa pressure. The specimens consisted of transmission electron microscope (TEM) disks, 3 mm diameter X 0 .3 mm thick, and small sheet-tensile specimens (gauge section of 7 .6 mm long X 1 .5 mm wide X 0 .84 mm thick) . The tantalum wrapper, if properly cleaned, does not restrict the flow of tritium to and from the specimens. Heat treatment of the specimens and wrap under a good vacuum at elevated temperatures facilitates a more rapid tritium sorption . After vacuum heat treatment, the sample chamber is isolated from the pumping system ; and tritium is admitted to the chamber from the heated (- 700 K) tritium storage trap . After a nominal equilibrium pressure of 53 kPa is achieved (usually within 16-24 h), the specimen chamber is sealed ; and the samples are held at 675 K for a prescribed tritium decay time . The bulk of the remaining tritium is then removed

333

from the specimens by opening the valve between the tritium-trick chamber and the tritium storage trap; chamber temperature is increased to 975 K (later batches were outgassed at 875 K) . Typically room-temperature uranium-trap sorption action achieves a vacuum of 106 Pa in approximately 4 h . Further tritium removal is achieved by first pumping with a mechanical pump to 0.4 Pa and then with an ion pump to an ultimate pressure of 3 l,Pa . This ultimate pressure is achieved in 13-19 h from the initiation of tritium removal. After cooling the furnace to room temperature, the specimens are removed, unwrapped, and cleaned for 15 s in an ultrasonic bath containing 1 part concentrated HN03 , 1 part (48`x) HF, and 1 part H 2 0 by volume. Specimens were normally smear tested at 10 kBq before cleaning and less than 4 Bq after cleaning. Analysis for residual tritium is performed by electrolytically dissolving a TEM specimen in a solution of 7 parts methanol and 1 part sulfuric acid (by volume) . A few drops of HF (48%) solution are added to approximately 60 ml of electrolyte prior to each run to help prevent an oxide film from forming on the specimen . A schematic drawing of the system is shown in fig . 5 . With an applied do potential of 7 V, dissolution rates from 0.5 to 2 mg/min are achieved . The evolved HT gas is purged from the electrolytic cell with air at a rate that provides a 20E% excess of oxygen for the complete combustion of all hydrogen isotopes. The gas mixture is passed through a catalytic furnace which contains nickel sputter-coated with platinum, thus converting the HT to HTO. The tritiated water is then collected in a series of scrubbers which contain water ; both the electrolyte and HTO scrubber solutions are analyzed by scintillation counting . The data are used to determine the total tritium content of the specimen . The tritium-trick technique is used to prepare vanadium alloys containing from 7 to 300 atomic parts per million (appm) 3 He [6] .

CATALYTIC

FURNACE

ITadC I

211T + 02 -+ 2HTO

EIISSOLVER

HTO

SCRUBBER

Fig. 5 . Schematic of system used for obtaining sample for residual tritium analysis. X . TRITIUM TARGETS

334

E.H. Kobisk et al. / Tritium-processing operations at ORNL

Table 1 Residual tritium levels for various vanadium alloys Batch number

Alloy

Outgas

0 a) 1 4 b) 7 8 8 8 9

V-15% Cr-5% Ti V-15% Cr-5% Ti V-15% Cr-5% Ti V-15% Cr-5% Ti V-15% Cr-5% Ti Vanstar-7 Vanstar-7 V-3% Ti-1% Si

975 975 975 875 875 875 875 875

a) b)

Temperature (K)

Time (h) 7 19 25 16 13 13 13 20

Ultimate vacuum (P Pa)

Residual 3 H (MBq/g)

12 .0 1.5 5.3 3.1 4.0 4.0 4.0 2.9

17 .02 3.96 53 .28 19 .60 20 .02 48 .58 49.36 31 .38

Alloy scrap. Sample was only partially dissolved.

Residual tritium levels must be known because of potential contamination problems during specimen testing. It is also desirable to know the amount of 3He that will be generated by the residual tritium over a long period . For example, a modest 1000 appm of residual tritium (21 GBq/g) would generate approximately 55 appm 3 He in one year . Table 1 gives a summary of the residual tritium levels for three different alloys. They ranged from 17 to 53 MBq which translates to an appm range of 0.88 to 2.29, respectively . The one exception of this concentration range is Batch 1 (3 .9 MBq/g) . This particular sample was from a section of a tensile specimen after it had been tested to failure at 875 K in high vacuum for approximately 30 min. All other specimens analyzed were from samples that had undergone no further heat treatment other than the tritium removal step of the tritium-trick process. The important conclusion derived from these data is that the level of tritium is low enough to allow physical testing with minimum precautions against tritium contamination of the environment . 4. Tritium sales program Tritium in the form of TZ is provided by the ORNL Isotopes Program for a wide range of research, development, and production applications. Major commercial applications include tritium lights, electronic tubes, and products using a tritium-based luminous compound such as in clocks and compasses. Gaseous tritium is provided to ORNL by the Savannah River Laboratory (SRL) where the tritium is produced by irradiating aluminum-lithium targets in a heavy water reactor (HWR). The tritium received from SRL is sorbed on uranium traps in a second ORNL tritium facility and subsequently used for filling specific orders. The amount of

tritium processed in this facility each year is approximately 74 PBq. Tritium processing for the ORNL sales program involves a chemical purification process for tritium gas received from the SRL. As the tritium decays, the concentration of the 3He daughter in the storage containers increases to the point that it is difficult to process by the customers. For the most part, the ORNL purification process involves removal of the 3 He . A flow diagram of the tritium system is shown in fig. 6. The gas is received in a cylinder at up to 93 .3 kPa (700 mm Hg) pressure. The cylinder is shown in fig. 7. In the initial stage, the system is in the automatic mode of operation, and the gas is processed through the main uranium trap which removes the tritium from the gas stream by the formation of uranium tritide. The helium is allowed to pass through the trap and discharge into the Impure Tritium Storage Tank (IST). At some point, when the trap becomes loaded, the tritium breaks through the main trap. This occurs when the absorption efficiency is reduced due to the reduction in pressure in the feed cylinder . When the gas flow drops due to pressure equalization between the feed cylinder and the IST, the oilless vacuum pump is started which reduces the feed cylinder pressure further. This is continued until the feed cylinder is empty. At this point, the system is switched to manual, and tritium is withdrawn from the IST. It is circulated over the main uranium trap until it has reacted. This is continued until the entire quantity of tritium in the IST has reacted with the uranium. The tritium is stored in the main uranium trap until it is needed for shipment . The uranium trap is then heated to decompose the tritide and load a gas cylinder . There are three types of containers which are normally loaded using this system : namely, 1-ml to 10-ml glass ampules, 10-ml to 50-1 gas cylinders, and uranium

E.H. Kobisk et al. / Tritiumprocessing operations at ORNL

TO EXHAUST

335

FE FT FI N8 NY NS NV PT PI PV PIS PA RE RIS RA RR AT TC TE TIC TAN TI 00 OV

FLOW ELEMENT FLOW TRANSDUCER FLOW INDICATOR HAND SELECTOR SWITCH SOLONOID VALVE HAN SNITCH AIR OPERATED VALVE PRESSURE TRANSMITTER MESSURE INDICATOR MESSURE SIGNAL CONDITIONER MESSURE INDICATING SWITCH MESSURE ALARM RADIATION ELEMENT (IONIZATION CNNASERI RADIATION INDICATING SWITCH RADIATION ALARM RADIATION RECORDER RADIATION SIGNAL CONDITIONER TEMPERATURE CONTROLLER TEMPERATURE ELEMENTTTNERMOCOUPLEI OVER TEMPERATURE CONTROLLER HIGH TEMPERATURE ALARM TEMPERATURE INDICATOR INTEGRATORTOTALIZEN MULTIPLIER

Fig . 6 . Flow diagram of ORNL tritium system used for loading tritium gas for sales program .

trap shipping containers. The required quantity of tritium loaded into a shipping cylinder is determined by the pressure reading and the ideal gas law relation, PV = nRT; the temperature is taken to be the measured room temperature. The uranium trap shipping containers are limited to 1 .1 PBq and are loaded from two expansion tanks in the same manner as the gas cylinders . Loading of shipping containers is done in the manual mode of operation. 4.1 . Process containment and equipment 4.1 .1 . Containment The tritium is located inside a stainless steel hood which is itself located in the tritium process room. The hood is connected to the Isotope Area Cell Ventilation System and is maintained under a nominal 99 .3 kPa (-0.5 in . w.g .) pressure. The hood is the primary containment area for the gaseous tritium . The building, which is also maintained at a negative pressure with respect to barometric (- 99 .43 kPa, 746 nun Hg) serves as the secondary containment area.

4.1 .2 . Equipment There are four major components to the tritium process : the uranium traps, in-line ionization chambers, vacuum systems, and process instrumentation . Each is described below. 4.1.2.1 . Uranium traps. A typical uranium trap is made of welded 304 stainless steel and contains 0 .4 kg of depleted uranium metal turnings . Each trap has a 0.45kW resistance heater which provides heat to decompose the UH 3 and has a 0.64-cm stainless steel tube wrapped around it for cooling with compressed air. Each trap heater is controlled with an SCR controller, which is backed up by an overtemperature controller that has a cutoff set point at 725 K (measured by a separate thermocouple) . These temperature limitations are to prevent stainless steel-uranium alloying damage which would occur should the traps be heated to greater than 775 K . Protection against overpressure in the trap system is provided by a rupture disk (0 .207 ± 0 .014 MPa, 30 psi f 2 psi) in a line leading to a 2-1 tank . The X . TRITIUM TARGETS

EH. Kobisk et al. / Tritium processing operations at ORNL

336 30.48c. j*

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US DOE SR AIKEN, S.C. RADIOACTIVE MATERIALS USA/6678/BL (DOE-SR) TYPE B GROSS WT 260 LB MODEL N0. LP-50 SN

DETAIL A

NOTE " SERIAL NUMBER AS SPECIFIED ON PURCHASE ORDER

by Hoke gas cylinder connectors or special rubber compression fitting connectors . 4 .1 .2.4. Process instrumentation. Process instrumentation, as designated in fig. 6, can be grouped into four areas : pressure, temperature, flow, and radiation . Radiation instrumentation is described in section 4 .1 .2 .2 . All pressure instrumentation consists of capacitance-type digital pressure sensors . Chromel-alumel thermocouples make up the temperature instrumentation. Flow is measured by laminar meters, each of which are connected to a transducer that measures a pressure difference across the device . One flowmeter is read directly, and the other flowmeter inputs, together with the output from an ionization chamber, to a multiplier which is then integrated to obtain the total quantity of tritium discharged from the system . 4.2 . Waste management and radiation protection

FLAT-TYPE ARRANGEMENT FIRE AND IMPACT SHIELD FOR LP-50 GAS CYLINDER

Fig . 7 . Savannah River Laboratory tritium cylinder (outside schematic).

secondary containment for the uranium traps is piped through a valve to the vacuum system . 4.1 .2.2. In-lin e ionization chamber. The system is instrumented to detect the release of tritium gas into the cell ventilation system. An in-line ionization chamber between the main trap and guard trap senses process activity. This instrument provides a signal for analytical purposes . An additional ionization chamber, located on the vacuum pump discharge to cell ventilation, alarms to alert personnel to pump discharge to cell ventilation, alarms to alert personnel to activity being discharged, and if in the automatic mode, operator activity being discharged, and if in the automatic mode, operator action would be required to close valve HV 20 (fig . 6) . 4.1 .2.3. Vacuum system . The vacuum system consists of one oil-sealed vacuum pump and an oilless bellows pump . All process valves consist of bellows valves with a bellows to body seal . Most joints in the 9 .5 mm stainless steel tubing are Heli-arc welded. The oil vacuum pump discharges into cell ventilation. Connections to the Savannah River Operations (SRO) feed cylinder are made by a compression fitting (Cajon VCR) . Connections to shipping cylinders and ampules are made

4.2 .1 . Waste management Tritium process waste is handled as both gaseous and solid . The gaseous waste is discharged into the cell ventillation system and results in approximately 13 TBq per month being discharged . The solid waste consists of gloves, wipes, shoe covers, and a certain amount of contaminated equipment . This is bagged and treated as solid waste in accordance with Division and Laboratory Solid Waste Disposal procedures. 4.2.2. Radiation protection Tritium radiation consists of a relatively soft beta (E,n,,, = 2 .98 fJ, 18 .6 keV) . Therefore, there is no penetrating radiation hazard . Since the elemental tritium ( 3 H 2 ) has such an affinity for nearly everything, there is significant potential for contamination. The tritium beta is not detectable with conventional survey meters, so contamination is monitored with swipe/smear samples which are counted on a liquid scintillation detector. Contamination levels are maintained at less than 16 .6 Bq/100 cm2 (1000 d/min/100 cm2 ) at all times . All personnel wear airline masks, gloves, and shoe covers during all operations in the tritium room. There is a constant air monitor located adjacent to the tritium hood as shown in fig . 8, which will alarm at a concentration of about 37 MBq/m3 of tritium, which is one-half of MPC40 for elemental tritium gas . Risk is substantially increased when handling tritium in the oxide form (HTO) compared to that of handling the element form (T2) . This is because tritium oxide is much more readily assimilated into the body's water than is elemental tritium . Consequently, the MPC value for tritium oxide is a factor of 400 less than elemental tritium. Therefore, a body fluid analysis program is maintained and used by the facility operators .

EH Kobisk et al. / Tritium processing operations at ORNL

33 7

overtemperature controller fails to turn off the electrical power.

KRYPTON PERMANENT STORAGE TANKS

5. Radioluminescent lights

BUILDING 3033 LOCATION OF INSTRUMENTATION

7C

O71 TS

TRITIUM CONSTANTAIR MONITOR CONTROLS TRITIUM IONIZATI0NCHAMBER TRITIUM SAMPLER

Fig. 8 . Facility instrumentation for ORNL tritium system used for loading tritium for sales program.

The tritium process system neither uses nor produces tritium oxides. The only mechanisms for tritium oxide formation are (1) oxidation of elemental tritium contamination and (2) hydrogen exchange with normal water which is the primary source of personnel exposure in the tritium operation . Tritium gas is either loaded or unloaded from closed gas cylinders, uranium traps, or sealed glass ampules . These are cleaned with disposable paper wipes prior to removal from the hood into the tritium room where they are decontaminated further. The cylinders are cleaned to less than 16.6 Bq/100 cm2 (1000 d/min/100 cmZ ) prior to their removal from the building. 4.3 . Tritium safety system Each uranium trap in the system is equipped with a rupture disc rated at 207 kPa (30 psi). This rupture disc is in a line which leads to a dump tank which has been pressure tested at 1 .03 MPa (164.7 psia), has been sized at 501, and can contain the system inventory of 4 .4 PBq at a pressure of 105 kPa absolute (15 .2 psia) . The maximum test pressure of the uranium traps is 791 kPa absolute (114 .7 psia) . A rupture disc/dump tank system and the dual heater controls eliminate the possibility of a rupture of the uranium trap due to a temperature excursion if the

Radioluminescent (RL) lighting systems are selfpowered alternatives to conventional electrically powered lighting systems for remote, tactical, and emergency airfield and helipad lighting and marking applications. They are ideally suited for uses where electrically powered lights are not acceptable or impracticable for reasons of mobility, ruggedness, dependence upon external power, etc . However, they are not suitable for applications where high intensity is required . In the basic RL light, a phosphor compound is coated on the inner surface of an impermeable transparent vessel . A radioactive solid may be mixed with the phosphor in the coating, or the vessel is filled with a radioactive gas and sealed . The radioactive source decays and emits energy in the form of decay products (alpha or beta particles or neutrons) or electromagnetic radiation (gamma- or X-rays or photons) . Energy from the decay products excites the molecules in the phosphor to a higher energy state . This energy state is unstable ; the energy is released as light as the molecules return to their normal ground state. Once in the ground state, the phosphor is available to be excited again, repeating the cycle. The continuous repetition of this process produces the light emitted by a RL lighting device . The most common RL lighting technology uses tritium gas as the radioactive source and glass tubing as the vessel material . Tritium is relatively benign and emits only a low-energy beta particle that is easily shielded, producing no detectable radiation outside its enclosure . Glass tubing is easy to work, is inexpensive, and resists the diffusion of tritium gas . Because tritium is an isotope of hydrogen, it shares many of the characteristics of hydrogen, including its chemical reactivity and fast diffusion rate . Tritium is relatively plentiful and reasonably priced . It is, therefore, almost an ideal RL light energy source. The program for the development and fabrication of RL lights was begun at ORNL in 1979 . Since 1981 research and development efforts have been directed at improving light output through improved phosphors, better light-tube geometry, and increased tritium content. During this time, light tubes have been assembled in several different configurations that have produced light outputs ranging from 0 .15 to 0.38 cd (the latter being one from the set purchased from a commercial vendor) . The commercial set had a median light output value of 0 .355 cd with a standard deviation of 0 .011 cd. The process of producing a light box, which contains 1-7 light tubes, involves coating the inner surface of X . TRITIUM TARGETS

E.H. Kobisk et al / Tritiumprocessing operations at ORNL

33 8

SEVEN TUBE LIGHT PANEL

Fig . 9. RL light tubes. Pyrex glass tubes with copper-doped zinc-sulfide phosphor. The tubes, shown in fig . 9, are then loaded with a small quantity (- 2 TBq) of tritium using the system shown in fig. 6. The loading operations are conducted by personnel dressed in contamination zone clothing and fresh air masks. The hood that contains the tritium system and the laboratory are exhausted to the cell ventilation system which maintains a negative pressure of >_ 75 Pa (- 0.3, in. w.g.) on the laboratory and a nominal 124 Pa (- 0.5 in w.g.) on the hood. Tritium monitors are used to detect airborne tritium, and routine tritium smears are taken to determine if tritium contamination is present in the facility. Operating personnel routinely submit urine samples to verify that current operating procedures are adequate to maintain personnel tritium exposure levels to values that are as low as reasonably achievable (ALARA) . The ORNL RL light development efforts culminated in this technology being transferred to the private sector . This was accomplished by conducting several technology transfer conferences, the first of which was held at ORNL in March of 1982. 6. Tritium recovery

During the early stages of the RL light program at the ORNL, it became evident that a method of recovering tritium from these gas-filled light sources was needed . The R&D effort of the RL light program has produced numerous outdated prototype lights as well as quantities of lights which have failed during various stages of quality control testing. To date, over 2.2 PBq of tritium exist in an unusable form in these rejected light sources at ORNL . To prevent a large inventory of

these tubes from accumulating and creating a potential safety problem, a tritium-recovery system has been designed, constructed, and (at present) tested with hydrogen. A development effort in the RL light program aimed at tritium recovery was initiated to meet the immediate on-site needs at ORNL with an ultimate goal of transferring the technology to private industry in the RL light manufacturing business. Tritium recovery from gas-powered RL lights and lights generated from future R&D efforts to develop new metal tritide-powered RL lights will require a safe and effective technology. This technology not only has application as a method of recovering a valuable commodity from RL light production rejects, but may also be applied to old RL lights no longer having usable light intensity. The useful life of a tritium gas-filled light source is from 8 to 10 years. With a tritium half-life of 12.3 years, this 8- to 10-year life span would yield 50-60% of their original charge of tritium available for recovery and reuse. ORNL has produced gas-powered RL lights containing over 18.5 PBq of tritium . When these lights are no longer usable, this system can be used to recover a potential 11 PBq of tritium for reuse. The present RL light source design is a glass tube coated internally with phosphor and filled with tritium gas. Recovery from light sources of this design requires mechanically breaking the glass tube and purifying and storing the evolved tritium gas . Options for purifying and storing the evolved gas are numerous. The method of purification chosen for this process was diffusion through a thin palladium-silver alloy membrane. Gas permeation through palladium-silver will act as an absolute filter to all gases except hydrogen isotopes . This purification technique is used in industry today in several commercially available hydrogen purifiers ; the technique has also been applied to the separation of hydrogen isotopes in two R&D efforts at ORNL [7]. Tritium storage is accomplished using uranium traps. The basic processing flow scheme of the recovery system is shown in fig. 10. Glass tubes are placed in a hermetically sealed chamber, that is evacuated and backfilled with argon, and then are mechanically broken. Tritium separation from the argon carrier gas and the 3He decay product is affected by a palladium-silver membrane heated to 620 K. A nominal 93-kPa (700 Ton) forepressure is maintained by an oilless, metal-bellows vacuum pump . A near-zero backpressure is maintained by a 0.925-PBq capacity tritium trap filled with activated uranium. Circulation of the argon-tritium mixture is maintained until the tritium partial pressure is less than 133 Pa (1 Torr), as determined from residual gas analyses . The remainder of impure tritium will then be scavenged by a second uranium trap. Impure gases and vapors removed from the tritium do not accumulate on the separator membrane surface but are continuously depleted via a gaseous bleed stream having a flow

EH. Kobisk et al. / Tritiumprocessing operations at ORNL

339

cuss

Fig. 10. Process flow schematic for tritium recovery system.

rate of > 509 of the feed flow rate. This ensures that the surface of the membrane is continually flushed with fresh feed gas and that impure gases are swept away . After the residual tritium is scavenged, the manifold is backfilled with argon, and the broken glass tubes are removed to a metal waste receptacle . When the waste receptacle is filled, it is sealed and placed in a 55-gal stainless steel drum containing a nonhardening asphalt. Additional asphalt is added to fill all void space around ô

10 -1

U

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the primary waste can. The drum is then transferred to ORNL's solid waste storage area . To date, functional testing of the manifold has been completed. Figs. 11 and 12 represent typical test results with argon-hydrogen mixtures . The test with 4% hydrogen illustrates the maximum practical separation achievable in a reasonable period of time with the palladium-silver membrane separation technique. This test was performed with the downstream side of the membrane being continuously evacuated with a mechanical pump, i.e ., zero backpressure . It appears that a reasonably achievable minimum hydrogen concentra-

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0.8

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Fig. 11 . Test results from tritium recovery system using argon-4%1 hydrogen mixtures.

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Fig. 12 . Test results from tritium recovery system using argon-50%1 hydrogen mixtures. X . TRITIUM TARGETS

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E.H. Kobisk et al / Tritium processing operations at ORNL

tion is 6 x 10 -3 mol fraction. The test with 50% hydrogen illustrates the effect of backpressure on tritium permeation . The downstream side of the membrane was initially evacuated to 1 .3 Pa. The permeating hydrogen was then allowed to accumulate and raise the backpressure to a level equivalent to the hydrogen partial pressure (H Z forepressure) on the upstream side of the membrane . During normal operations, the pumping action downstream of the tritium separation membrane will be supplied by the sorption of the uranium trap .

7. Summary The uncertain long-term effects resulting from the exposure of processing personnel (or the general population) to radiation from tritium contamination would seem to dictate that tritium production and/or atmospheric processing procedures be improved to minimize tritium releases . Procedures used in several tritium-handling operations at ORNL have been described with

emphasis on procedures that are being used or will be implemented to accomplish this goal .

References

[21

[41 [51 [61

[81

I .R. Brearley, The Hazard to Man of Accidental Releases of Tritium, SRDR 331 (1985). E.A. Pinson and W.H . Lagham, J . Appl . Physiol . 10 (1957) 108 . H .L . Adair, E .H. Kobisk, and B .L. Byrum, Nucl . Instr. and Meth . 200 (1982) 99. R.G . Hickman, Proc. 1 Topl . Mtg. on Technology of Controlled Nuclear Fusion 2, ed . G .R . Hopkins (1974) p . 535 . J .F . Remark, Nucl . Technol. 29 (1976) 369. D .W . Ramey and D .N. Braski, Nucl. Instr. and Meth . B10/11 (1985) 976 . D .W . Ramey, Thesis, The University of Tennessee, Knoxville, Tennessee, USA (1977). D .W . Ramey and M . Petek, Sep. Sci. Technol. 15 (1980) 405 .