WO1999063546A2 - Device for storage of gaseous radioisotopes - Google Patents

Abstract

A system for disposing of gaseous waste containing at least one radioisotope comprising: (a) a cyclotron (60) for producing the radioisotope; (b) a radiochemistry laboratory (66) for producing the gaseous waste including the at least one radioisotope; (c) a gas waste disposal unit (10) connected to the radiochemistry laboratory (66) for receiving the gaseous waste from the radiochemistry laboratory (66) the gas waste disposal unit (10) comprising: (i) a storage tank (12) for storing the gaseous waste, the storage tank (12) featuring: an outlet valve (22) and an inlet valve (24); (ii) a compressor (26) connected to the inlet valve (24) for pressurizing the gaseous waste in the storage tank (12); and (iii) an EPC (Electrical Pressure Controller)(28) for controlling the compressor (26) and for determining the pressure of the gaseous waste, such that the compressor (26) is substantially intermittently activated by the EPC for maintaining the pressure of the gaseous waste in the storage tank (12).

Description

DEVICE FOR STORAGE OF GASEOUS RADIOISOTOPES

FIELD AND BACKGROUND OF THE INVENTION

The present invention relates to a device for safely collecting and

containing radioisotopes in gaseous form until they can be safely released into

the atmosphere, and in particular for storing short-lived radioisotopes until the

radioisotopes have sufficiently decayed to permit safe release.

The storage and disposal of radioactive wastes from various industrial

processes has proven problematic. Long-term storage and disposal solutions

are required for radioactive waste containing long-lived radioisotopes, such as

waste from the processing of uranium for nuclear fuel. These radioisotopes

have half-lives of hundreds of years, significantly prolonging the process of

decay, during which these wastes must be safely contained. Thus, the

containers for storing these wastes must be able to contain the wastes for such a

prolonged period of time, in particular as single, sealed, separate units.

Examples of such containers are disclosed in U.S. Patent No. 4,673,814,

which discloses a container featuring a hollow vessel capped with a tightly

sealed, gas impermeable cover. The tightly sealed unit can be safely

transported and stored as a single, separate unit. Similarly, U.S. Patent Nos.

4,783,309; 4,818,878; 5,042,679; and 4,437,578 all disclose containers which

are intended to be transported and stored as a single, separate unit. Also similarly, U.S. Patent No. 4,626,414 discloses an apparatus which fills

individual containers for storing and transporting radioactive waste as separate,

sealed, single units.

By contrast, radioisotopes used in medical applications such as PET

(positron emission tomography) often have very short half-lives ([F-18] tι/2 =

110 min., [C-l 1] ti/2 = 20 min., [N-13] ti/2 = 10 min., [0-15] ti/2 = 2 min.),

typically on the order of minutes. These radioisotopes must be produced on site

using a medical particle accelerator with a satellite radiochemistry laboratory.

The production of short lived PET radiopharmaceuticals involves in most cases

handling of radioactive gases and radiochemical multistep reactions. These

various steps take place in different locations of the site and require reliable

processes of transport and trapping of radioactive liquids and gases.

Furthermore, most of these radiochemical conversions are not exclusive and

volatile radioactive by-products are generated during the process.

Storage of the radioactive waste in devices suitable for long-lived

radioisotopes would be prohibitively expensive and inefficient, in addition to

being unnecessary since these radioisotopes have often sufficiently decayed to

permit release of the waste into the general atmosphere after only a few hours

to a few days. Thus, a device which can collect and store the short-lived

radioisotopes for a short period of time is required, until the radioactive waste

will decay to a level so that the waste can be disposed of in a manner suitable

for non-radioactive waste. Currently, these short-lived radioisotopes such as the radioactive gases

produced as a by-product of certain chemical and radiochemical processes are

either directly released to the general atmosphere or are first stored in highly

pressurized containers before being released. However, these currently

available devices all suffer from significant drawbacks.

For example, U.S. Patent No. 3,871,842 and U.S. Patent No. 4,038,060

both disclose an exhaust gas cleaning system for removing radioactive material

from gases. The radioactive material is removed with an adsorbent material,

which is contained in a tank. However, such adsorbent material requires

elaborate procedures for insertion and removal, as detailed in both U.S. Patent

No. 3,871,842 and U.S. Patent No. 4,038,060, and must itself be treated as a

radioactive waste. Furthermore, the processes involved in the production of

PET radiopharmaceuticals have radioactive gaseous by-products which are

highly difficult to trap, such as nitrogen-13 (N-13) and C^π0₂. Thus, these

disclosed background art systems also have a number of significant

disadvantages, particularly with regard to the inclusion of an adsorbent

material.

U.S. Patent No. 5,368,633 discloses a system for compressing and

storing radioactive gas until the radioisotopes have decayed to a safe level,

which does not require adsorbent material. The disclosed system maintains

pressure with two gas circuits between an original storage container and a

decay tank. The first circuit is a recirculation line through a pressure control

valve back to the original storage container, and is used to maintain pressure within the system. The second circuit goes to the decay tank. A gas pump is

used to suck gas from the original storage container into the decay tank.

However, the disclosed system maintains pressure through a complicated two-

circuit system, with periodic injections of carrier gas if necessary. Such a

system is unnecessarily complicated, requiring too many components in order

to maintain pressure throughout the system. Thus, the disclosed two-circuit

system is overly and unnecessarily complex, and may therefore be expensive

and prone to mechanical failure.

There is therefore a need for, and it would be useful to have, a device for

safe collection and short-term storage of radioactive gases, which would

contain these gases until the radioisotopes had sufficiently decayed to permit

safe release of the gases into the general atmosphere, and which is able to

maintain pressure at a constant level in the radiochemistry system connected to

the device without highly complex, expensive machinery.

SUMMARY OF THE INVENTION

According to the present invention, there is provided a gas waste

disposal unit for disposing of gaseous waste including at least one radioisotopic

material, the gas waste disposal unit comprising: (a) a storage tank for storing

the gaseous waste, the storage tank featuring: (i) an outlet valve for removing

the gaseous waste from the storage tank; and (ii) an inlet valve for enabling the

gaseous waste to enter the storage tank; (b) a compressor connected to the inlet

valve for pressurizing the gaseous waste in the storage tank, such that the radioactive gaseous waste has a pressure; and (c) an EPC (Electrical Pressure

Controller) for controlling the compressor and for determining the pressure of

the gaseous waste, such that the compressor is substantially intermittently

activated by the EPC for maintaining the pressure of the gaseous waste in the

storage tank.

Preferably, the storage tank further comprises: (iii) a substantially rigid

outer wall; and (iv) a flexible inner membrane, the flexible inner membrane

being contained within the outer wall and the flexible inner membrane being

connected to the inlet valve for receiving the gaseous waste. More preferably,

a general atmosphere is accessible through the outlet valve when the outlet

valve is open, such that the gaseous waste is removed from the storage tank by

escaping to the general atmosphere through the outlet valve when the outlet

valve is open. Most preferably, a pressure is present between the outer wall

and the flexible inner membrane, such that the pressure of the gaseous waste in

the flexible inner membrane is substantially similar to the pressure between the

outer wall and the flexible inner membrane. Also most preferably, the pressure

between the outer wall and the flexible inner membrane is in a range from

about 3 bars to about 9 bars.

Preferably, the gas waste disposal unit further comprises: (d) a detector

for detecting an amount of radioactivity of the at least one radioisotopic

material in the storage tank, such that the outlet valve is opened substantially

only when the amount of radioactivity reaches a predetermined level. According to preferred embodiments of the present invention, the gas

waste disposal unit further comprises a trap for trapping a substance in the

gaseous waste substantially before the gaseous waste is pressurized by the

compressor. Preferably, the trap is a liquid trap and the substance is moisture

in a form of liquid droplets. Alternatively and preferably, the trap is a charcoal

trap. Also alternatively and preferably, the trap is a soda-lime trap.

According to other preferred embodiments of the present invention, the

at least one radioisotopic material includes a radioisotope with a short-life.

Preferably, the radioisotope with the short-life is selected from the group

consisting of [F- 18] (fluorine-18), [C-11] (carbon-11), [N-13] (nitrogen-13),

and [0-15] (oxygen- 15).

According to another embodiment of the present invention, there is

provided a method for disposing of gaseous waste, the gaseous waste

containing a radioisotopic material with a radioisotope having a short-life, the

method comprising the steps of: (a) providing a storage tank with an inlet valve

and an outlet valve; (b) providing a compressor connected to the inlet valve of

the storage tank; (c) providing an EPC; (d) placing the gaseous waste in the

storage tank through the inlet valve; (e) pressurizing the gaseous waste in the

storage tank with the compressor, such that the gaseous waste has a pressure;

(f) controlling the compressor with the EPC such that the EPC determines

activity of the compressor and such that the EPC determines the pressure of the

gaseous waste; (g) storing the gaseous waste in the storage tank until a predetermined limit has been achieved; and (h) releasing the gaseous waste

from the storage tank through the outlet valve.

Preferably, the predetermined limit is a period of time. Alternatively

and preferably, the method further comprises the step of monitoring an activity

of the radioisotope, and the predetermined limit is a preset level of

radioactivity, such that when the activity of the radioisotope is substantially

below the level of the radioactivity, the gaseous waste is released through the

outlet valve. More preferably, the step of releasing the gaseous waste through

the outlet valve includes the step of releasing the gaseous waste from the outlet

valve to a general atmosphere.

According to other preferred embodiments of the present invention,

there is provided a system for disposing of gaseous waste, the gaseous waste

including at least one radioisotopic material containing at least one

radioisotope, the system for disposing of gaseous waste comprising: (a) a

cyclotron for producing the at least one radioisotope; (b) a radiochemistry

laboratory for producing the at least one radioisotopic material with the at least

one radioisotope and for producing the gaseous waste including the at least one

radioisotope; (c) a gas waste disposal unit for being connected to the

radiochemistry laboratory and for receiving the gaseous waste from the

radiochemistry laboratory, the gas waste disposal unit comprising: (i) a storage

tank for storing the gaseous waste, the storage tank featuring: (1) an outlet

valve for removing the gaseous waste from the storage tank; and (2) an inlet

valve for enabling the gaseous waste to enter the storage tank; (ii) a compressor connected to the inlet valve for pressurizing the gaseous waste in the storage

tank, such that the radioactive gaseous waste has a pressure; and (iii) an EPC

(Electrical Pressure Controller) for controlling the compressor and for

determining the pressure within the system, such that the compressor is

substantially intermittently activated by the EPC for maintaining the pressure

within the system.

Hereinafter, the term "short half-life" refers to a half-life of less than

about 2 years, preferably less than about 6 months, more preferably less than

about two weeks, and most preferably less than about 24 hours. Examples of

radioisotopes with short half-lives include, but are not limited to, [F-18] tl/2 ⁼

1 10 min., [C-11] tl/2 = 20 min., [N-13] tl/2 = 10 min., and [0-15]

ti/2 = 2 min. Hereinafter, the term "general atmosphere" refers to the

atmosphere not specifically contained within a particular device or container,

such as the air surrounding a building, for example.

BRIEF DESCRIPTION OF THE DRAWINGS

The foregoing and other objects, aspects and advantages will be better

understood from the following detailed description of a preferred embodiment

of the invention with reference to the drawings, wherein:

FIG. 1 is a schematic diagram of an illustrative example of a gas waste

disposal unit according to the present invention;

FIG. 2 is a schematic diagram of an illustrative example of a gas waste

disposal system according to the present invention; and FIGS. 3A-3C illustrate radioactivity produced and stored with the

present invention during three different complete production runs.

DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS

The present invention is of a device for short-term storage of radioactive

material, particularly radioactive gases containing short-lived radioisotopes.

After the amount of radioactivity has decayed to a safe level, the material can

be disposed of according to any suitable method, for example by releasing the

gases into the general atmosphere. The device of the present invention has the

advantage of being relatively simple to construct, operate and maintain, yet is

sufficiently durable for safe operation over a long period of time.

The principles and operation of a device for short-term storage of

radioactive material according to the present invention may be better

understood with reference to the drawings and the accompanying description, it

being understood that these drawings are given for illustrative purposes only

and are not meant to be limiting.

Referring now to the drawings, Figure 1 shows a schematic diagram of

an illustrative embodiment of a gas waste disposal unit for radioactive or other

hazardous gaseous waste according to the present invention. A gas waste

disposal unit 10 features a storage tank 12. Storage tank 12 preferably includes

a substantially rigid outer wall 14 constructed of a substance such as metal, and

a flexible inner membrane 16 within outer wall 14. Flexible inner membrane

16 is preferably made from a polymeric substance such as plastic, more preferably a plastic which is relatively chemically inert. One advantage of this

preferred construction of storage tank 12 is that if flexible inner membrane 16

should tear or otherwise permit the escape of gaseous waste, outer wall 14

would be able to prevent any escape of gaseous waste outside of storage tank

12. Such a type of storage tank 12 can be obtained from Oran Ltd. (Jerusalem,

Israel) for example.

In addition, preferably there is compressed air in the space between

outer wall 14 and flexible inner membrane 16, and this space is substantially

sealed for example at the time of manufacture, to prevent escape of the

compressed air. Preferably, the compressed air has a predetermined pressure of

about 3 bars when flexible inner membrane 16 is substantially empty. As the

gaseous waste enters flexible inner membrane 16, the volume of flexible inner

membrane 16 expands, for example from a substantially completely collapsed

state to a substantially completely expanded state. The expansion of this

volume causes the pressure of the compressed air between outer wall 14 and

flexible inner membrane 16 to increase, preferably to a maximum pressure of

about 9 bars.

Storage tank 12 should be able to hold radioactive or other hazardous

gaseous waste for at least about 24 hours, more preferably for at least about one

week, and most preferably for at least about one month, before the gaseous

waste is removed, for example by being released to the atmosphere. Most

preferably, the gaseous waste includes at least one radioisotopic material

having a radioisotope with a short half-life, including but not limited to [F-18], [C- l 1], [N- 13], and [0- 15], so that the waste can be held in storage tank 12

until the radioisotopic material has sufficiently decayed to permit safe release

of the gaseous waste to the atmosphere.

According to preferred embodiments of the present invention, storage

tank 12 is more preferably substantially cylindrical in shape, and is most

preferably about 145 cm in height and 60 cm in diameter, for a final volume of

about 300 liter, although of course other geometrical configurations and

dimensions are possible and would be obvious to one of ordinary skill in the

art. More preferably, the pressure of the radioactive gas within storage tank 12

is in a range of from about 1 bar to about 11 bar, and most preferably in a range

of from about 3 bar to about 9 bar. Also most preferably, the initial air pressure

within storage tank 12 should be about 3 bar. Optionally and preferably,

pressure within storage tank 12 is measured by a tank pressure gauge 18.

Storage tank 12 features an inlet valve 20, which is preferably a one-way

valve, and an outlet valve 22. When outlet valve 22 is open, gases are able to

leave flexible inner membrane 16 for direct release to the general atmosphere

or to another storage container (not shown). Gases enter inlet valve 20 from a

tank inlet tube 24. Preferably, tank inlet tube 24 is made from a polymeric

material, more preferably a semi-rigid or rigid material. Most preferably, the

polymeric material is substantially chemically inert. An example of a suitable

polymeric material is P VC (polyvinyl chloride) plastic, although of course

other suitable polymeric materials would be well known to one of ordinary skill

in the art. Tank inlet tube 24 is connected to a compressor 26. Compressor 26 is

preferably a 0.5 HP (horsepower) compressor. More preferably, compressor 26

develops an adjustable pressure of up to 12 bars at the junction of compressor

26 and tank inlet tube 24. More preferably, compressor 26 can create a vacuum

of 0.1 bar and can work continuously at least up to two hours (available from

OMA (Italy), for example).

Compressor 26 is preferably controlled by an Electrical Pressure

Controller (EPC) 28. EPC 28 could be a type RT-121 compressor (Danfoss,

Denmark), for example. Preferably, EPC 28 is capable of maintaining pressure

of gas in a range of from about 0.09 to about 1 bar. EPC 28 is in gaseous

communication with a compressor inlet tube 30. The term "gaseous

communication" is intended to connote that a sample of gas is able to escape

compressor inlet tube 30 through an EPC tube 32, such that EPC 28 is able to

measure the pressure of the gas within compressor inlet tube 30. EPC 28 is able

to control compressor 26 according to the measured pressure of gas within

compressor inlet tube 30, for example by causing compressor 26 to become

activated or inactivated through an electrical connection to compressor 26.

Such alternating activation and inactivation could be performed by alternately

enabling or blocking a flow of power from a power supply (not shown) of

compressor 26. In addition, a pressure gauge 34 may also be attached to

compressor inlet tube 30 for determining a pressure of the gas within

compressor inlet tube 30. Compressor inlet tube 30 and EPC tube 32 are both preferably made

from a polymeric material, more preferably a semi-rigid or rigid material. Most

preferably, the polymeric material is substantially chemically inert. An

example of a suitable polymeric material is PVC (polyvinyl chloride) plastic,

although of course other suitable polymeric materials would be well known to

one of ordinary skill in the art.

Compressor inlet tube 30 is preferably connected to at least one trap, of

which a plurality are shown and are generally designated as "traps 36". Traps

36 are able to remove or "trap" substances from the gaseous waste travelling

through gas waste disposal unit 10 which might prove detrimental to the

operation of compressor 26. Traps 36 are interconnected by a plurality of trap

tubes 38, such that the last trap 36 is then connected to compressor inlet tube

30.

Preferably, plurality of traps 36 include a liquid trap 40 for trapping any

moisture within the gaseous waste, for example in the form of liquid droplets.

In addition, preferably plurality of traps 36 also include a soda-lime trap 42 and

a charcoal trap 44.

Preferably, gaseous waste enters at least one, and preferably a plurality

of, traps 36 from a one way valve 46. The gaseous waste enters one way valve

46 from a manifold outlet tube 48. Preferably, manifold outlet tube 48 features

a manifold pressure gauge 50 for determining the pressure of the material

leaving a manifold 52. Manifold 52 is connected to the chemistry waste valves

radiochemistry lab (not shown, see Figure 2). Preferably, trap tubes 38 and manifold outlet tube 48 are made from a polymeric material, more preferably a

semi-rigid or rigid material. Most preferably, the polymeric material is

substantially chemically inert. An example of a suitable polymeric material is

PVC (polyvinyl chloride) plastic, although of course other suitable polymeric

materials would be well known to one of ordinary skill in the art.

According to a particularly preferred embodiment, gas waste disposal

unit 10 is preferably part of a gas waste system (shown in more detail in Figure

2 below), which includes a detector of radioactivity 54 for detecting a level of

radioactivity in storage tank 12. Detector of radioactivity 54 is preferably

connected to a monitor 56, which may be placed at a distance from storage tank

12, such that personnel are able to monitor the level of radioactivity within

storage tank 12 substantially without exposure to radioactivity from storage

tank 12. Preferably, once the level of radioactivity has reached a predetermined

level, for example after being stored to allow decay of the radioisotopes, the

gaseous waste in storage tank 12 is removed by opening outlet valve 22. For

example, if the predetermined level is the amount of radioactivity considered to

be sufficiently low for safe release to the atmosphere, then the gaseous waste

could be allowed to escape to the general atmosphere through outlet valve 22.

The amount of radioactivity which is considered to be safe for release to the

atmosphere may be determined by suitable national and international regulatory

bodies, such as the International Atomic Energy Commission.

The preferred operation of gas waste disposal unit 10 is as follows. Gas

waste disposal unit 10 is preferably connected to a PLC (programmable logic controller, not shown, see Figure 2 below) which automatically activates gas

waste disposal unit 10 whenever a radiochemical process is started or

completed. Alternatively and preferably, gas waste disposal unit 10 is

manually activated and deactivated. In either case, such control of the

activation and deactivation of gas waste disposal unit 10 is preferably

performed by activating or deactivating EPC 28. In addition, preferably inlet

valve 20 of storage tank 12 is either opened upon activation of gas waste device

10 or else is left opened substantially continuously, so that gas can enter

storage tank 12. Also preferably, one way valve 46 is either opened upon

activation of gas waste disposal unit 10 or else is left opened substantially

continuously, so that gas can leave the radiochemistry facility (not shown)

through manifold 52.

Upon activation, EPC 28 controls the pressure of the gas by activating

and deactivating compressor 26 in order to maintain the gas pressure within a

certain desired range. Preferably, the gas pressure lies within a range of from

about 0.75 bars to about 0.9 bars during operation of gas waste disposal unit

10. Such a relatively narrow range of pressures is particularly preferred in

order to maintain a narrow range of pressure in the radiochemistry facility (not

shown, see Figure 2 below), since the chemistry waste valves of the

radiochemistry facility are opened to gas waste disposal unit 10 through

manifold 52. Furthermore, by controlling the activity of compressor 26 with

EPC 28, this range of pressures can be maintained while avoiding continuous

use of compressor 26. Without the requirement for such continuous use, compressor 26 may be relatively inexpensive, since a relatively less robust type

of compressor can be employed.

Figure 2 depicts an illustrative schematic preferred embodiment of an

exemplary gas waste system of the present invention, featuring the gas waste

disposal unit of Figure 1. A gas waste storage system 58 features three major

components: a cyclotron 60, at least one on-line radiation monitoring channel

62, and gas waste disposal unit 10. Gas waste disposal unit 10 is substantially

as shown in Figure 1 above. Cyclotron 60 could be a 18 MeV negative ion (H^~)

cyclotron (Model 18/9, Ion Beam Application (IB A), Belgium), for example.

Preferably, cyclotron 60 is connected to a radiochemistry facility 66 for

producing short-lived radioisotopically-labelled materials.

Each radiation monitoring channel 62 is preferably composed of at least

one detector 68. Detector 68 could be a GM-tube or a high sensitivity 2"x2"

Nal (Tl) scintillation detector, for example. The GM-tube could be a high

sensitivity GM-42 detector, based on a ZP-1201 Geiger tube, Centronic, UK.

Detector 68 is preferably capable of detecting radioactivity of various types,

more preferably including alpha, beta and gamma radiation.

Preferably, there are a plurality of radiation monitoring channels 62,

more preferably at gas waste disposal unit 10, and at a chimney 67 of

radiochemistry facility 66.

The choice of a suitable type of detector 68, whether ionization

chamber, scintillator, Geiger or semi-conductor, depends upon both the required sensitivity, and on the detector life time. This choice could easily be

made by one of ordinary skill in the art.

Example 1 Testing of Gas Waste System

The exemplary gas waste system of the present invention as shown in

Figure 2, including the gas waste disposal unit of Figure 1, was tested as

follows. A brief description of the experiment is described below. For more

details, see the attached Appendix I.

Materials and Methods

Data was collected over a year during the production of each batch of

[F- 18]FDG (fluorodeoxyglucose, an exemplary PET radiopharmaceutical) (50-

800 mCi), [C-l l]Deprenyl (4-50 mCi), [0-15] Water (50-250 mCi), [N-

13] Ammonia (50-300 mCi) and [F-18]F-DOPA (20-30 mCi).

The preferred embodiment of the gas waste system of Figure 2, which

was tested, included two detectors, one high sensitivity GM-42 GM-tube

detector and one high sensitivity 2"x2" Nal (Tl) scintillation detector. The

GM-tube detector was used for monitoring the radiation level in the gas waste

tank itself The scintillation detector was located in the ventilation system, at

the end of the hot cell (radiochemistry facility) chimney.

Calculations based on detector surrounding geometry and properties

were made in order to convert the pulse rates obtained by the detector at the top

of the chimney into activity concentration levels (see Appendix I for details). Results

The radiation levels measured at the gaseous waste decay tank and at the

top of the hot cells chimney during production of commonly used PET

radiopharmaceuticals are summarized in Table 1. High radiation levels were

measured at the gas waste decay tank during labeling processes with [C-

1 1 ]C02 and [0- 15]θ2- These radiation levels increased to values of more than

150 mR/hr when [0- 15]water was produced continuously or more than 300

mR/hr when 1.5 Ci of [C- l 1]C02 was produced, and increased also during a

failure in the production of [C-l l]deprenyl. In contrast to the high levels

observed in the gas waste tank while working with [C-l 1]C02 or [0-15]θ2,

the count rate measured at the top of the chimney was very low. The activity

concentration released during [C-11] production was 3.1μCi/m³ (see Appendix

I)-

En vironmental Radiation safety

Since all the chemistry waste valves were connected to the gas waste

system, the activity concentration levels of carbon- 11 at the top of the chimney

were lower than the Drived Air Concentration (DAC) recommendation

(0.27 mCi/m³) of the International Atomic Energy Agency (IAEA) [9]. The gas

waste system prevents the release in the atmosphere of radioactive by-products

generated during the production of radiopharmaceuticals and of volatile

compounds of failed radiochemistry processes. The design of the gas waste

storage tank increases the safety in the site and minimizes the space needed for such a system in comparison with traditional solutions such as large size

ballons.

Production Optimization

The radiation levels measured in the gas waste tank during production of

[C- l l]deprenyl [3] were used in order to optimize the reaction process (Figure

4). As shown in the graph of the first run (Figure 3 A) with a flow of 500

ml/min of [C- 1 1 ]C02 coming from the target, 98 mR/hr were measured in the

waste tank during the trapping step. During the next steps-the reduction with

LiAlH4 and the iodination with HI - large quantities of radioactive gas reached

the gas waste tank and the radiation level exceeded 240 mR/hr. Increasing the

flow of [C-l 1]C02 at the second run (Figure 3B) in order to decrease the

trapping time resulted in a higher loss of activity (during this first step) to the

gas waste storage tank. However, increasing the volume of L1AIH4 in THF,

decreasing the reaction temperature and Argon flow and other modifications

(such as changing the quantity of starting material, increasing the reaction time

of the third step and decreasing the temperature of the third step) which were

performed during the second and third steps of the syntheses, resulted in lower

quantities of radioactivity loss to the gas waste. The results of the third run with

optimal conditions is shown in Figure 3C. The maximum level was observed

during the trapping step and did not exceeded 98 mR/hr. The activity loss was

due to a known side nuclear reaction that produced [N-13]nitrogen which

cannot be trapped [ 12]. Therefore, as demonstrated by this experiment, the gas waste system of

the present invention provides an effective solution for the control of various

aspects of production and radiation safety in a cyclotron-radiochemistry

facility. In addition the gas waste disposal unit enables the radiochemistry

facility to meet the radiation safety recommendations published by the IAEA

(International Atomic Energy Association). The design of the gas waste tank

increases the safety in the site, minimizes the space needed for a gas waste

system, especially in comparison with currently available storage devices such

as large size balloons, and provides a longer period of time for decay.

It will be appreciated that the above descriptions are intended only to

serve as examples, and that many other embodiments are possible within the

spirit and the scope of the present invention.

Table 1 : Radiation Levels Measured During Production

^ppesica I Radiation Levels in Cyclotron-Radiochemistry Facility measured by A

Novel Comprehensive Computerized Monitoring System.

E. Mishani¹, N. Lifshits', A. Osavistky , J. Kaufman², N. Ankry², N. Tal², R. Chisin¹ ' Department of Medical Biophysics and Nuclear Medicine, Hadassah Hebrew University Hospital POB 12000 , Jerusalem

91 120, Israel. ² Rotem Industries POB 9046, Beer-Sheva 84190, Israel.

Abstract

Radiation levels in a cyclotron-radiochemistry facility were measured during the production of commonly used PET radiopharmaceuticals by a comprehensive computerized monitoring system. The system consists of three major components : on-line radiation monitoring channels, an area control unit, and a gas waste management unit. During production the radiation levels were measured in the cyclotron vault, inside automatic chemistry production and research shielded cells, in the radiochemistry room, in the gas waste decay tank, in the chimney filters, and at the top of the cells chimney. Each detector was calibrated in a known radiation field, and a special detector dead time correction was performed in order to achieve detected signal-to-radiation linearity for the Geiger tubes located in the radiochemistry production and research cells. During production of C-11 and O-15 PET radiopharmaceuticals, high radiation levels were measured in the gas waste decay tank (240 and 80 mR hr respectively). In contrast, the radiation levels at the chimney filters and at the top of the cells chimney did not exceed the International Atomic Energy Agency (IAEA) Drive Air Concentration (DAC) recommended for C-11 or O-15. During production of FDG, high radiation levels were measured at the chimney filters, however the radiation level at the top of the chimney (3.7 μCi/m³) did not exceed the F- 18 DAC recommendation (27μCi/m³). Low radiation levels of approximately 0.5-1 mR hr were measured in the radiochemistry room during production of PET radiopharmaceuticals. In the cyclotron vault, 2 minutes after bombardment the radiation levels at 2 meters from the cyclotron decreased to 1-2 mR hr. The addition of a gas waste decay system to computerized monitoring channels located near each strategic point of the site allows loi a comprehensive sui vey ol the radiochemical piocesscs

Key words : Radiation-level, cyclotron, PET, FDG, radioactive gas-decay.

Eyal Mishani : Department of Medical Biophysics and Nuclear Medicine, Hadassah Hebrew

University Hospital POB 12000 , Jerusalem 91120, Israel. email : [email protected]

Fax # : 972 2 6421203

Introduction.

Positron Emission Tomography (PET) is becoming a major tool for both biochemical research and noninvasive diagnostic purposes [1]. Since the majority of radioisotopes used in PET have a short half life ([F-18] t_1/2 = 1 10 min., [C-11] t_1/2 = 20 min., [N-13] t_l/2 = 10 min., [O-15] t_1/2 = 2 min.), these isotopes must be produced on site using a medical particle accelerator with a satellite radiochemistry laboratory. The production of short lived PET radiopharmaceuticals involves in most cases handling of radioactive gases [2] and radiochemical multistep reactions [3]. These various steps take place in different locations of the site and require reliable processes of transport and trapping of radioactive liquids and gases. Furthermore most of these radiochemical conversions are not exclusive and radioactive byproducts are generated during the process. Therefore, in this type of integrated cyclotron-radiochemistry facility a reliable gas waste decay system and a continuous monitoring of radiation levels in real time is mandatory for both radiation safety [4] and production purposes. We report here the radiation levels measured during the production of commonly used radiopharmaceuticals by a comprehensive computerized monitoring system. This provides a factual basis for establishing radiation guidelines and enables to monitor the different aspects of production.

Materials and Method

Radioisotopes were generated with a 18 MeV negative ion (H^~) cyclotron (Model 18/9, Ion Beam

Application (IB A), Belgium). [O-15]Water, [F-18]FDG were produced with IBA automated chemistry units. [C-l l]deprenyl was produced with a Nuclear Interface automated chemistry unit (Nuclear Interface, Munster, Germany). Data was collected over a year during the production of each batch of [F- 18]FDG (50-800 mCi), [C-l l]Deprenyl (4-50 mCi), [O-15]Water (50-250 mCi), and [N-13]Ammonia (50-300 mCi).

The monitoring system consists of three major components : on-line radiation monitoring channels, an area control unit, and a gas waste management unit. Each of the radiation monitoring channels is composed of a detector and a Data Processing Unit (DPU). The DPU's are connected to one control PC with an analysis software. The area control unit includes field sensors (pressure, humidity, doors position, etc.), a Programmable Logic Controller (PLC) and a Man Machine Interface (MMI) (scheme 1 ). Monitoring Channels

Detectors

The system includes sixteen detectors, twelve are GM-tubes and four are high sensitivity 2"x2" Nal (Tl) scintillation detectors. Out of theses twelve GM-tubes three are high sensitivity GM-42 detectors (based on a ZP-1201 Geiger tubes, Centronic, UK) and are used for area monitoring in the radiochemistry lab and the basement area, and for monitoring the radiation levels in the gas waste tank (scheme 2). Nine other Geiger detectors (GM-41, based on ZP-1313 Geiger tubes, Centronic, UK) are located inside automatic chemistry production and research cells and in the cyclotron vault (Scheme 2). The scintillation detectors (PM-11) are located in the ventilation system (one at the chimney filters, two at the end of each chimney), and one is used to monitor the radiation levels in the liquid disposal tank (scheme 2). Each type of detector has its own electronics circuit for providing the power and amplifying the signal. It also has its own identity frequency recorded on the DPU, allowing for simple and automatic system recognition of each type of detector. This frequency is produced by an internal oscillator and transferred to the DPU on one of the detectors leads. In addition, because of statistical variance between the detectors, each detector is calibrated by adjusting the frequency of the oscillator in a known radiation field in order to achieve accurate signal. This calibration factor is saved in the detector as a specific frequency for each detector, and is transferred to the DPU. In order to achieve detected signal-to-radiation linearity for the Geiger tubes, a special detector dead time correction was included in the DPU software. The minimum detectable level (MDL) of the scintillation detectors was improved by the use of a built-in hardware energy window of 511 KeV. Calculations based on detector surrounding geometry and properties were made in order to convert the pulse rates obtained by the detector at the top of the chimney into activity concentration levels. Taking in consideration the positron range [5] before it annihilated (for C-11 and F-18 the positron range (X^) is up to 300 cm, and 170 cm respectively), we assumed that the positron anhilation take place mainly on the duct aluminum surface (for C- 1 1 and F- 18 the positron range (X_A1) is up to 0.145 cm, and 0.084 cm respectively). This assumption allows to define the duct as a uniform surface source of activity that emits two 511KeV photons caused by the annihilation. Equation 1 determines the conversion factor (F) between the detector pulse rate (in cps) and the activity concentration ( LiCi/m'). This factor is obtained from the probability of a photon generated on the surface of the duct to scintillate in the detector.

Eq 1:

F - conversion factor

V - duct volume

A ' - duct surface

I - Intensity

S- Symmetry

C_A - Specific area activity [photon/m²].

L - Attenuation factor caused by 3 mm thick aluminum

A - Detector surface [m³]. η- Nal(Tl) detector efficiency for 511 KeV photons. d - The distance between the detector and the duct [m].

X- Duct height [m].

Y- Duct width [m].

Z- Duct length [m].

The solution of equation 1 with the following estimation and dimension gives F = 3.24 m³.

L - 93% , η - 40%

S = 4 , A = 0.0025[m³] , a = 0.12[m], X= 0.3[m] , Y = 0.3[m] , Z = 0.5[m].

In order to confirm the estimation made on the detector efficiency an additional calculation was made with the MCNP software [6]. This software uses the Monte Carlo method to calculate the conversion factor [7]. The conversion factor calculated by using this software was 4.17 m\

DPU

Each DPU consist of a microcontroller 80L32 (Intel, IL, USA), operation file programmed in EPROM-27C256 (Texas Inst., USA), and a display unit. The microcontroller opens and closes the communication channels, commands the transmission, and the display unit driver. The DPU performs the following functions: It provides the radiation levels monitored by each detector, alerts by audible and visual alarm in case of radiation levels exceeding a predetermined threshold or in case of detector failure, and communicates with the control station. The DPU also calculates the calibration factor as the ratio of the detector specific frequency to the original detector type frequency, and the correct radiation field by multiplying the detected rate of pulses by the calibration factor (eq. 2). The radiation levels are transferred from the DPU via a RS 485 communication network to the PC for on-line display and documentation. eq. 2: Displayed radiation field = count rate of pulsesxdeteclor specific frequency/original detector type frequency.

Gas waste system

The gas waste system consists of an expansion tank, a compressor, an Electrical Pressure Controller (EPC), a liquid trap, a one way valve, and manifold and pressure gauges (scheme 3). The compressor is a 0.5 HP and can develop an adjustable pressure up to 12 bars at the exit. It can create a vacuum of 0.1 bar and can work continuously up to two hours. The EPC ( type RT-121, Danfoss, Denmark) has a range of 0.09-1 bar. The gas waste system is connected to the PLC which switches the system on and off whenever a radiochemical process is started or completed. The pressure in the waste system is maintained by the EPC between 0.9 to 0.75 bars during operation in order to maintain a narrow range of pressure in the chemistry units when its valves are opened to the waste system, and in order to avoid continuous use of the compressor.

Results.

The radiation levels measured by the system during production of commonly used PET radiopharmaceuticals are summarized in table 1. High radiation levels were measured at the gas waste decay tank during labeling processes with [C-l l]CO₂ and [O-15]O₂. These radiation levels increased to values of more than 150 mRhr when [O-15]water was produced continuously or more than 300 mR/hr when 1.5 Ci of [C-l l]CO₂ was produced, and increased also during a failure in the production of [C- l l]deprenyl. In contrast to the high levels observed in the gas waste tank while working with [C- 1 1 JCO, or [O- 1 JO,, the count rate measured in the hot cells chimney filters and at the top of the chimney were very low. Based on equation 1 the activity concentration released during [C- 1 1] production was 3. 1μCi/m\ During FDG production, high count rates were observed in the hot cells chimney filters. When 300 mCi of FDG was produced the count rate reached a maximum of 3470 cps. This maximum was doubled when 800 mCi of FDG was produced. In contrast, the count rate at the top of the chimney during FDG production did not exceeded 12 cps, which, based on equation 1, translates in an activity concentration of 3.7μCi/m³. During production of [F-18] fluoride, under optimal operation of the cyclotron with full target and constant beam current on target, a constant radiation level was observed in the cyclotron vault (scheme 4). The high radiation levels in the vault decreased to 2-3 mR/hr two minutes after transferring the activity to the radiochemistry lab.

Discussion

Detector configuration

The choice of the adequate detector (ionization chamber, scintillator, Geiger or semi-conductor) for a specific task depends on the MDL needed (determined by radiation levels and background level), and on the detector life time. In the cyclotron vault, the transmission of gamma rays is higher than the transmission of neutrons [8], therefore only one GM-tube was located in the cyclotron at a distance of 2 meters from the cyclotron. For operational purposes, i.e. to determine the efficiency of radioisotope production and radiochemical conversions (such as [O-15]O₂ to [O-15]Water or [C-l l]CO₂ to [C- l l]deprenyl), to monitor transfer and activity trapping, and to evaluate the radiochemical yield of each step during a multi-step radiosynthesis, GM tubes were placed in each of the strategic parts of the lab, and in the gas waste decay tank (expansion tank) (scheme 2). On the other hand for regulatory and health safety purposes, sensitive scintillation probes were located in the ventilation deck and near the discharged liquid tank. The probes in the ventilation deck were placed inside the chimney near the filters, and outside the top of the chimney, in order to avoid absorption of radioactive particle on the detector surface. Theoretical calculations and software simulation based on detector surrounding geometry and properties were made in order to convert the count rates obtained by the detector at the top of the chimney into activity concentration levels. The conversion factor that was obtained by software simulation

(4.17 m³) was in the same range as the conversion factor obtained by theoretical calculations (3.24 m³). Environmental Radiation safety

Since all the chemistry waste valves were connected to the gas waste system, the activity concentration levels of carbon- 1 1 at the top of the chimney were lower than the Drived Air Concentration (DAC) recommendation (0.27 mCi/m³) of the International Atomic Energy Agency (IAEA) [9]. The gas waste system prevents the release in the atmosphere of radioactive by-products generated during radiopharmaceuticals production and of volatile compounds of failed radiochemistry processes. The design of the gas waste tank increases the safety in the site and minimizes the space needed for such a system in comparison with traditional solutions such as large size ballons. During FDG production, since several evaporation steps [10] were performed with open vials the activity concentration levels observed in the hot cells chimney filters were higher by a 4-5 order of magnitude than the DAC recommendation for F-18 (0.027 mCi/m³) (table 1). At the top of the chimney, the radiation levels were lower than the DAC recommendation for fluorine-18, underlining the importance of the filters located in the chimney.

Production Optimization

Radiation levels measured inside the vault were related to cyclotron operation. During production of [C-l l]CO₂ with a beam current of 20 μA on target, a constant radiation level of 147 mR hr was observed in the vault at a distance of 2 meters from the cyclotron (scheme 5). When the beam current was decreased to 13 μA the radiation level decreased to 88 mR/hr. During production of [F-18]FDG, under normal operation of the cyclotron with full target and constant beam current on target, a constant radiation level was observed in the cyclotron vault (scheme 4). This value changed when bombardment was performed on an empty target. The radiation level observed in the vault was also used for fine tuning of the beam. Upon adjustment of the striper position, although the beam on target or the striper/target beam ratio was not changed, a 30-40% increase in the radiation level was observed in the vault. The adjustment of the striper did not change any measurable parameter of the cyclotron operation, however it centered the beam on the target window and increased the formation of [F-18]fluoride and neutron (^l8O(p,n)^,8F). As a result, the radiation level in the vault increased and the yield of [F-18] also increased by 20-30%. This shows that the monitoring channel in the cyclotron vault can be used as a sensitive probe o\^' the target condition and the cyclotion opeiation. In addition, in the case of FDG production at the end of bombardment after transferring the activity to the second station (the water recovery unit), the radiation level in the vault decreased to 1 -2 mR after 2 minutes. Unsuccessful or non-completed transfer will cause higher radiation levels and will be detected by the monitoring channel in the vault.

The monitoring channel in the water recovery unit (scheme 6), enables, after calibration, quantification of the activity at the begining of the synthesis and the calculation of the yield of [F- 18]fluoride production. In addition, the yield of the FDG production process can be easily calculated. The follow up of radiation levels in the FDG hot cell is shown in scheme 7. Although we used only one monitoring channel it was sufficient, as shown in this graph, to identify transfers of the activity from the first reactor of the chemical unit to the second and to the purification columns. In addition loss of activity as volatile byproducts during evaporation steps was also observed by this monitoring channel and by the monitoring channel located at the shielded cell chimney filter.

In radiochemical processes involving a radioactive gas, the determination of the activity flow plays a key role in the optimization of chemical conversions. In the case of [O-15]water production, the conversion of [O-15]O₂ to [O- 15] water over the palladium catalyst is not exclusive and the [O- 15] oxygen which does not convert to water, is trapped in the gas waste system [11]. The radiation levels in the gas waste tank and in the [O- 15] water chemistry unit (due to [O- 15] water trapped in the product vial) were measured by two monitoring channels (scheme 8). This allows for the calculation of the quantity of the remaining [O-15]Oxygen and the radiochemical yield. In addition, the reaction parameters such as quantity and quality of the catalyst, temperature, flow rate of gas, were reflected by the ratio of radiation levels in the gas waste to the [O- 15] water unit and optimization could easily be performed by adjusting these parameters.

The radiation levels measured by the system in the gas waste decay tank during production of [C-l l]deprenyl [3] were similarly used in order to optimize the reaction process (scheme 9). As shown in the graph of the first run (scheme 9 A) with a flow of 500 ml/min of [C-l l]CO₂ coming from the target, 98 mR/hr were measured in the waste tank during the trapping step. During the next steps-the reduction with LiAlH₄ step and the iodination with HI - large quantities of radioactive gas reached the decay tank and the radiation level exceeded 240 mR/hr. Increasing the How of [C- l 1 ]CO, at the second run (scheme 9B) in order to decrease the trapping time resulted in a higher loss of activity (during this first step) to the decay tank. However, increasing the volume of LiAlH₄ in THF, decreasing the reaction temperature and Argon flow and other modifications (such as changing the quantity of starting material, increasing the reaction time of the third step and decreasing the temperature of the third step) which were performed during the second and third steps of the syntheses, resulted in lower quantities of radioactivity loss to the gas waste. The results of the third run with optimal conditions is shown in scheme 9C: The maximum level was observed during the trapping step and did not exceeded 98 mR/hr. The activity loss was due to a known side nuclear reaction that produced [N-13]nitrogen which can not be trapped [12].

Conclusions

This new monitoring system provides an effective solution for the control of various aspects of production and radiation safety in a cyclotron-radiochemistry facility. The combination of a gas waste decay system and computerized monitoring channels located near each strategic point of the site allows for a comprehensive evaluation of radiochemical processes. Since signal to radiation level linearity was achieved for Geiger tubes and each monitoring channel calibrated, the results obtained can be used to quantify the yield of each step during the various radiosyntheses. In addition the gas waste unit permits to meet the radiation safety recommendations published by the IAEA. The design of the gas waste tank increases the safety in the site, minimizes the space needed for a gas waste system (in comparison with traditional solution such as large size balloons) and provides longer time for decay.

References

1. McCarthy, T. J.; Schwarz, S. W.; Welch, M. J. J. of Chemical Education, 1994, 71 (10), 830- 836.

2. McCarthy, T.J., Bonasera, T. A., Welch, M. J.; Rozen S. J. Chem. Soc, Chem Commun. 1993, 561.

3. MacGregor, R. R.; Fowler, J. S.; Wolf, A. P. J. of Label. Compd. Radiopharm. 1988, 25 (1), 1-9.

4. International Atomic Energy Agency, Vienna, 1988; Radiological safety aspects of the operation of proton accelerators. Technical Report Series NO. 283. 5. Turner. J. R.; Atoms, Radiation, and Radiation Protection, pp 94-95; Pergamon Piess Inc. 1986.

6. A General Monte Carlo N-Particle Transport Code software, Loss Alamos National Labs. Nov. 1993.

7. Arthur, B. C; Shultis, J. K.; Faw, R. E.; Principels of Radiation Shielding, pp 348-358; Prentice Hall, Inc., New Jersey 1984.

8. Mukherjee B.; Parcell, S. Appl. Radiat. hot. 1997, 48 (4), 453-457.

9. International Atomic Energy Agency, Vienna, 1982; Basic safety standards for Radiation Protection, safety series No. 9.

10. Hamacher K.; Coenen H. H.; Stocklin G. J. Nucl. Med. 1986, 27, 235-238.

1 1. Meyer G. J.;et al. Eur. J. Nucl. Med., 1984, 9, 220.

12. Qaim, S. M.; Clark, J.C; Crouzl, C; Guillaume, M.; Helmeke,H. J.; Nebeling, B.; Pike, V. W.; Stocklin, G.- PET radionuclide production; In Radiopharmaceuticals for Positron Emission Tomography, Stocklin, G. and Pike, V. W.; Kluwer Academic Publishers 1993.

Scheme 1 : Monitoring system layout.

Waste gas System Area control

Monitoring Channels

Cyclotron PLC

PC

Scheme 3 : Gas waste decay unit.

from chemistry waste valves

Table 1 .- Radiation levels measured by the system during production.

* At a distance of 2 meters from the cyclotron.

** The detector was located inside the chimney.

*** The detector was located outside the top of the chimney.

Scheme 5 : Radiation levels in the vault during [C-11] production.

During irradiation the beam current was reduced from 20 to 13 μAm.

Scheme 4 Radiation levels in the vault during [F-18Jfluorιde production

Cyclotron was operated at stable beam current of 1 lμAm

Scheme 6 Radiation in the watei

shielded cell

Total starting activity 17Cι

Scheme 7 Radiation levels in the ladiochenu sii \ i ell dm mi; FDG production

Total starting activity of 1 3 Ci Total FDG product 350mCι

Scheme 8 : Radiation levels during [015 Jwater production.

A : radiation level in the gas waste tank, B : radiation level in the chemistry cell. [0-15] was produced in batches providing approximately 250mCi of final product for each batch.

Scheme 9 : Radiation levels in the gas waste tank during [C-l 1 [deprenyl production.

A : loss of activity during the second and third steps; 9B : loss of activity during the trapping step, and improvment in the third and second step; 9C : "optimal" n.

Claims

WHAT IS CLAIMED:

1. A gas waste disposal unit for disposing of gaseous waste

including at least one radioisotopic material, the gas waste disposal unit

comprising:

(a) a storage tank for storing the gaseous waste, said storage tank

featuring:

(i) an outlet valve for removing the gaseous waste from said

storage tank; and

(ii) an inlet valve for enabling the gaseous waste to enter said

storage tank;

(b) a compressor connected to said inlet valve for pressurizing the

gaseous waste in said storage tank, such that the radioactive

gaseous waste has a pressure; and

(c) an EPC (Electrical Pressure Controller) for controlling said

compressor and for determining said pressure of the gaseous

waste, such that said compressor is substantially intermittently

activated by said EPC for maintaining said pressure of the

gaseous waste in said storage tank.

2. The gas waste disposal unit of claim 1, wherein said storage tank

further comprises:

(iii) a substantially rigid outer wall; and (iv) a flexible inner membrane, said flexible inner membrane being

contained within said outer wall and said flexible inner

membrane being connected to said inlet valve for receiving the

gaseous waste.

3. The gas waste disposal unit of claim 2, wherein a general

atmosphere is accessible through said outlet valve when said outlet valve is

open, such that the gaseous waste is removed from said storage tank by

escaping to said general atmosphere through said outlet valve when said outlet

valve is open.

4. The gas waste disposal unit of claim 3, further comprising:

(d) a detector for detecting an amount of radioactivity of the at least

one radioisotopic material in said storage tank, such that said

outlet valve is opened substantially only when said amount of

radioactivity reaches a predetermined level.

5. The gas waste disposal unit of claim 4, further comprising a trap

for trapping a substance in the gaseous waste substantially before the gaseous

waste is pressurized by said compressor.

6. The gas waste disposal unit of claim 5, wherein said trap is a

liquid trap and said substance is moisture in a form of liquid droplets.

7. The gas waste disposal unit of claim 5, wherein said trap is a

charcoal trap.

8. The gas waste disposal unit of claim 5, wherein said trap is a

soda-lime trap.

9. The gas waste disposal unit of claim 3, wherein the at least one

radioisotopic material includes a radioisotope with a short-life.

10. The gas waste disposal unit of claim 9, wherein said radioisotope

with said short-life is selected from the group consisting of [F-18], [C-11], [N-

13], and [0-15].

11. A method for disposing of gaseous waste, the gaseous waste

containing a radioisotopic material with a radioisotope having a short-life, the

method comprising the steps of:

(a) providing a storage tank with an inlet valve and an outlet valve;

(b) providing a compressor connected to said inlet valve of said

storage tank;

(c) providing an EPC;

(d) placing the gaseous waste in said storage tank through said inlet

valve; (e) pressurizing the gaseous waste in said storage tank with said

compressor, such that the gaseous waste has a pressure;

(f) controlling said compressor with said EPC such that said EPC

determines activity of said compressor and such that said EPC

determines said pressure of the gaseous waste;

(g) storing the gaseous waste in said storage tank until a

predetermined limit has been achieved; and

(h) releasing the gaseous waste from said storage tank through said

outlet valve.

12. The method of claim 11, wherein said predetermined limit is a

period of time.

13. The method of claim 11, further comprising the step of

monitoring an activity of the radioisotope, and said predetermined limit is a

preset level of radioactivity, such that when said activity of the radioisotope is

substantially below said level of said radioactivity, the gaseous waste is

released through said outlet valve.

14. The method of claim 13, wherein the step of releasing the

gaseous waste through said outlet valve includes the step of releasing the

gaseous waste from said outlet valve to a general atmosphere.

15. The method of claim 14, wherein the radioisotope with a short-

half life is selected from the group consisting of [F- 18], [C- 1 1], [N-13], and [O-

15].

16. A system for disposing of gaseous waste, the gaseous waste

including at least one radioisotopic material containing at least one

radioisotope, the system for disposing of gaseous waste comprising:

(a) a cyclotron for producing the at least one radioisotope;

(b) a radiochemistry laboratory for producing the at least one

radioisotopic material with the at least one radioisotope and for

producing the gaseous waste including the at least one

radioisotope;

(c) a gas waste disposal unit for being connected to said

radiochemistry laboratory and for receiving the gaseous waste

from said radiochemistry laboratory, said gas waste disposal unit

comprising:

(i) a storage tank for storing the gaseous waste, said storage

tank featuring:

(1) an outlet valve for removing the gaseous waste from

said storage tank; and

(2) an inlet valve for enabling the gaseous waste to

enter said storage tank; (ii) a compressor connected to said inlet valve for pressurizing

the gaseous waste in said storage tank, such that the

radioactive gaseous waste has a pressure; and

(iii) an EPC (Electrical Pressure Controller) for controlling

said compressor and for determining said pressure of the

gaseous waste, such that said compressor is substantially

intermittently activated by said EPC for maintaining said

pressure of the gaseous waste in said storage tank.

17. The system for disposing of gaseous waste of claim 16, wherein

said storage tank further comprises:

(3) a substantially rigid outer wall; and

(4) a flexible inner membrane, said flexible inner membrane being

contained within said outer wall and said flexible inner

membrane being connected to said inlet valve for receiving the

gaseous waste.

18. The system for disposing of gaseous waste of claim 17, wherein a

general atmosphere is accessible through said outlet valve when said outlet

valve is open, such that the gaseous waste is removed from said storage tank by

escaping to said general atmosphere through said outlet valve when said outlet

valve is open.

19. The system for disposing of gaseous waste of claim 18, wherein

said gas waste disposal unit further comprises:

(iv) a detector for detecting an amount of radioactivity of the at least

one radioisotopic material in said storage tank, such that said

outlet valve is opened substantially only when said amount of

radioactivity reaches a predetermined level.

^'20. The system for disposing of gaseous waste of claim 19, further

comprising a trap for trapping a substance in the gaseous waste substantially

before the gaseous waste is pressurized by said compressor.

21. The system for disposing of gaseous waste of claim 20, wherein

said gas waste disposal unit further comprises a trap for happing a substance in

the gaseous waste substantially before the gaseous waste is pressurized by said

compressor.

22. The system for disposing of gaseous waste of claim 21, wherein

said trap is a liquid trap and said substance is moisture in a form of liquid

droplets.

23. The system for disposing of gaseous waste of claim 21, wherein

said hap is a charcoal trap.

24. The system for disposing of gaseous waste of claim 21, wherein

said trap is a soda-lime trap.

25. The system for disposing of gaseous waste of claim 21, wherein

the at least one radioisotopic material includes a radioisotope with a short-life.

26. The system for disposing of gaseous waste of claim 26, wherein

said radioisotope with said short-life is selected from the group consisting of

[F-18], [C-1 1], [N-13], and [0-15].