SAFETY SERIES /A -A - Nucleus Safety Standards/Safety... · 2012. 11. 1. · The Agency’s Statute...

/ A - ASAFETY SERIES-A _____ L *-

No. 19

The Management of Radioactive Wastes Produced by

Radioisotope Users

Technical Addendum

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1966

This publication is not longer valid Please see http://www-ns.iaea.org/standards/

THE M ANAGEM ENT OF RADIOISOTOPE W ASTES PRODUCED BY

RADIOISOTOPE USERS

TECHNICAL ADDENDUM


The following States are Members o f the International Atom ic Energy Agency:

AFGHANISTAN FEDERAL REPUBLIC OF NICARAGUAALBANIA GERMANY NIGERIAALGERIA GABON NORWAYARGENTINA GHANA PAKISTANAUSTRALIA GREECE PANAMAAUSTRIA GUATEMALA PARAGUAYBELGIUM HAITI PERUBOLIVIA HOLY SEE PHILIPPINESBRAZIL HONDURAS POLANDBULGARIA HUNGARY PORTUGALBURMA ICELAND ROMANIABYELORUSSIAN SOVIET . INDIA SAUDI ARABIA

SOCIALIST REPUBLIC INDONESIA SENEGALCAMBODIA IRAN SOUTH AFRICACAMEROON IRAQ . , SPAINCANADA ISRAEL SUDANCEYLON ITALY SWEDENCHILE IVORY COAST SWITZERLANDCHINA JAMAICA SYRIACOLOMBIA JAPAN ’ ' THAILANDCONGO. DEMOCRATIC KENYA TUNISIA

REPUBLIC OF REPUBLIC OF KOREA TURKEYCOSTA RICA KUWAIT UKRAINIAN SOVIET SOCIALISTCUBA LEBANON REPUBLICCYPRUS LIBERIA UNION OF SOVIET SOCIALISTCZECHOSLOVAK SOCIALIST LIBYA REPUBLICS

REPUBLIC LUXEMBOURG UNITED ARAB REPUBLICDENMARK MADAGASCAR UNITED KINGDOM OF GREATDOMINICAN REPUBLIC MALI BRITAIN AND NORTHERNECUADOR MEXICO IRELANDEL SALVADOR MONACO UNITED STATES OF AMERICA .ETHIOPIA MOROCCO URUGUAYFINLAND NETHERLANDS VENEZUELAFRANCE NEW ZEALAND VIET-NAM

YUGOSLAVIA

The Agency’ s Statute was approved on 23 October 1956 by the Conference on the Statute o f the IAEA held at United Nations Headquarters, New York: It entered into force on 29 July 1957. The Headquarters o f the Agency are situated in Vienna. Its principal objective is "to accelerate and enlarge the contribution o f atomic energy to peace, health and prosperity throughout the world".

© IAEA, 1966

Permission to reproduce or translate the information contained in this publication may be obtained by writing to the International Atom ic Energy Agency, KSrntner Ring 11, Vienna I, Austria.

Printed by the IAEA in Austria

May 1966


SAFETY SERIES No. 19

THE MANAGEMENT OF RADIOISOTOPE WASTES PRODUCED BY

RADIOISOTOPE USERS

TECHNICAL ADDENDUM

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA,, 196,6


International A tom ic E nergy Agency.The management of radioactive wastes p ro

duced by radioisotope u sers . Technical Addendum. Vienna, the Agency, 1966.

81p. (IAEA, safety ser ies no. 19)

621 .039 .7

This Addendum is a lso published in F rench , Russian and Spanish

THE MANAGEMENT OF RADIOISOTOPE WASTES PRODUCED BY RADIOISOTOPE USERS. TECHNICAL ADDENDUM

IAEA, VIENNA, 1966 ST I/P U B /119


FOREWORD

The International A tom ic E nergy A gency published in 1965, as part of its Safety Standards, a Code of P ractice entitled: "The Management of Radioactive Wastes Produced by Radioisotope U sers" (Safety Series No. 12, ST I/P U B /87), based on the work of two international panels convened by the A gen cy . This Addendum contains detailed technical inform ation on p ro ce sse s and procedu res that are outlined in m ore general term s in the Code of P ra ctice .

As in the Code o f P ra ctice itse lf, the in form ation given in this Addendum is p a rticu la r ly relevant to the p rob lem o f handling the re la tively sm all quantities of waste a ris in g frosri the use o f ra d io isotopes in laboratories, hospitals and industry when no special facilit ie s fo r ra d io a c t iv e w aste trea tm en t a re a v a ila b le on the s it e .

The Addendum is d ire cted tow ard prov id in g the ra d io iso top e user with the type and amount of inform ation required fo r him to be able to (a) a sse ss the a lternatives available to him in term s o f his particu lar needs and restra in ts , (b) develop a p re lim in a ry design of an optimum waste-handling system , and (c) get help and guidance in his search for m ore detailed inform ation.

The Addendum has been prepared by M r. W illiam H. Regan, J r ., whose se rv ice s w ere provided to the A gency by the Governm ent of the United States of A m erica , with the assistance of the Secretariat. The A gency b e liev es that this Addendum w ill prov ide in form ation of considerable value and publishes it fo r whatever use M em ber States and others m ay wish to make o f it. H ow ever, it should not be regarded as representing the Agency's offic ia l judgement o f policy on the matter.


CONTENTS

1. TYPES OF WASTE ASSOCIATED WITH SOME USES OF ANUMBER OF RADIOISOTOPES ................................ ................... 1

2. W ASTE-CO LLECTIO N CONTAINERS AND SYSTE M S_____ . 4

2 .1 . Segregation ............................................................................. 42 .1 .1 . P hilosophy o f segregation ................................ 42 .1 .2 . Methods ..................................................................... 5

2 .2 . L iqu id -w aste c o l le c t io n ...................................................... 82 .2 .1 . C o n ta in e rs ................................................................ 82 .2 .2 . Marking and r e c o r d s ........................................... 10

2 .3 . Solid-w aste c o l l e c t io n ........................................................ 102 .3 .1 . C ollection -con ta in er design and m ateria ls 142 .3 .2 . Marking and r e c o r d s ........................................... 182 .3 .3 . C ollection and transportation on -site ......... 19

3. DIRECT DISPOSAL OF RADIOACTIVE WASTESTO SEWERS............................................................................... 203 .1 . D ischarge p r a c t i c e s ............................................................. 213 .2 . Dilution techniques and com p u tation s............................ 27

3 .2 .1 . General c o n s id e ra tio n s ...................................... 283 .2 .2 . C alculations fo r d isposal p r a c t i c e ............... 303 .2 .3 . C onstant-drip d ischarge bottle . . . . . . . . . . . 323 .2 .4 . R ecom m endations fo r d i s p o s a l ...................... 34

3 .3 . Behaviour of radionuclides in sew age-treatm entplants .......................................................................................... 35

4. LIQUID-WASTE TREATM ENT TECHNIQUES SUITABLEFOR USERS OF RADIOISOTOPES.............................................. 37

4 .1 . B atch -ch em ica l treatm ent ............................................... 374 .1 .1 . L im e-sod a ash trea tm en t.................................. 404. 1 .2 . Aluminium and fe rr ic hydroxide coagulation . 414 .1 .3 . Phosphate coa gu la tion ......................................... 434 .1 .4 . F errocyanide precipitation ............................. 444 .1 .5 . Strontium phosphate precip itation ............... 474 .1 .6 . M assive ch em ica l t r e a tm e n t ........................... 474 .1 .7 . Treatm ent o f sludges ........................................ 49


4 .2 . Ion exchange using organ ic resin s ................... 504 .2 .1 , Treatm ent by ion exch an ge ................................. 504 .2 .2 . Cation exchanger fo r p rocess in g general

laboratory wastes ................................................ 524 .3 . E v a p o ra t io n ............................................................................. 59

4 .3 .1 . L ow -tem perature e v a p o ra to r ........................... 594 .3 .2 . W iped -film e v a p o r a t o r ...................... ............... 62

4 .4 . C onclusions and recom m endations ....................... ; . . . 64

5. SOLID-WASTE TREATM EN T AND DISPOSAL BY INDIVIDUAL USERS OF RADIOISOTOPES ................................... 65

5 .1 . In c in e r a t io n ............... .............................................................. 655 .1 .1 . In trod u ction .............................................................. 655 .1 .2 . A p p lic a b il it y ............................................................ 665 .1 .3 . Simple approach to calculating safe lim its

fo r incineration ..................................................... 665 .2 . C o m p r e s s io n .................................................................. 695 .3 . Solid-w aste buria l ..................................................... 70

5 .3 .1 . Packaging ................................................................. 705 .3 .2 . Site se lection fo r ground disposal ................. 72

6. AIR-BORNE WASTE MANGEMENT ......................................... 74

6 .1 . F um e-hood design recom m endations ............................. 746 .2 . Gaseous and a eroso l waste con trol system s suitable

fo r use by sm all laboratories ......................................... 766 .2 .1 . H igh -e fficien cy filte rs ....................................... 766 .2 . 2. M ed iu m -efficien cy p re -filte rs ....................... 786 .2 .3 . A ctivated carbon adsorbents ............................ 78

REFERENCES 79


1. TYPES OF WASTE ASSOCIATED WITH SOME USES OF A NUMBER OF RADIOISOTOPES

Isotope H alf-life Main use Amounts Type o f waste

A c227 22 yr A ctin ium -227 / beryllium neutron sources

1-100 me None unless source is broken or lost

Sb124 60 d ( a ) An tim ony-124 / beryllium neutron sources

102 - 103 c None unless source is broken or lost

(b ) As tracer m icrocurie o f m illicurie amounts

Small hazard

Ba140 12. 8 d Tracer in steel industry

10-50 me Remains as insoluble in slag

Br80andBr82

1. 5 yr

36 h J

f (a ) Diagnostic ■j use [ (b ) Industrial

tracer; e. g. study o f retention in tanks

~ 10 jj,c

~ 100 me

Excreted in urine

Liquids and solids

Cs137 30 yr Sealed sources used in therapy and radiography

1 me to 1000 c

None unless source is broken or lost

C 14 5760 yr (a ) Over 200 la be l-'I led compounds.

(b ) Tracer in b io log ica l work

~ 10 fic Various-depending on type o f experiment

C r51 27. 8 d C lin ical purposes 10 ficper patient

Excreted slowly

1


Isotope H alf-life Main use

C oou 5. 27 yr (a ) W ide range o fsealed sources for radiographic, m ed ica l or general use

(b ) Industrial irradiation. Sealed sources up to m egacurie le ve l

Auls“ 2.7 d Therapy:(a ) C o llo id a l

(b ) "Grains sources sealed in platinum

12.26 yr Various industrialuses: b io log ica l tracers, etc.

8 .04 d Diagnosisand treatment

2.26 h Diagnosis andtreatment, e. g . :(a ) hyperthyroidism

(b ) thyroid carcinom a

74. 4 d Industrialradiography

F e " 45 d Diagnosis andresearch purposes

Amounts

up to 50 c

Up to m egacuries

C o llo id a l m etal - up to 150 m e per dose

Grains -m illicu rieamounts

Very varying

1-50 (Jc

5 to 10 me

100 to 150 me

Usually 1 to 10 cper source

~ 10 jic per patient

Type of waste

Very little use as an industrial tracer; alm ost en tirely used as a sealed source

Largely retained in patient

Often le ft in patient but cou ld constitute solid waste

Large proportion gets into liquid effluent

Approxim ately 75% excreted in urine

As for I131 but short h a lf-l ife reduces hazard

Sealed sources

Some excreted

2


Isotope H a lf-life Main use Amounts Type o f waste

H g197andHg203

65 h

47 d

Rising in im portance in diagnoses

~ 10 lie per patient

Som e excreted

p 32 14.2 d (a ) Diagnosis Up to 300 |ic

(b ) Treatment of b lood disorders

5 to 10 me per dose

Largely excreted in urine

( c ) Agricultural research work including field trials

Up to 100 c Liquid and solid waste depending on type o f experim ent

Ra226 1620 yr (a ) C lin ica l M illicurieamounts

Valuable expensive closed sources - should be little if any waste

(b ) Industrial - as a fo il-typ e source and in certain types o f e le ctro n ic valves and switches

M icrocurie amounts in sources: m icrocurie amounts in valves and switches

M ainly solid waste

( c ) Manufacture o f luminous com pounds

Up to about 50 m c /k g

M ainly solid waste

S c 46 84 d Silt tracer in rivers, etc.

Curies per operation

Various, depending on type o f experim ent

Na22 2. 58 yr Diagnostic tracer for humans

M icrocurieamounts

Largely excreted

3


Isotope H a lf-life Main use Amounts Type o f waste

Na24 15 h (a ) Tracer in industry, e. g. detecting leaks in new water mains.

(b ) Human diagnostic tracer

►

M icrocurie tom illicurieamounts

Liquids and solids; When used for diagnosis largely excreted

Sr 89 51 d Large scale agricultural tracer work

100 to 1000 me per experim ent

Liquids and solids

S r 90 28 yr Sealed sources for thickness gauges gauges, etc.

1 m e to 1 c Sealed sources - there should be little waste problem

S 35 87.2 d Agricultural experim ents (insecticides and fungicides)

Up to curie amounts

Depending on type o f experim ent

2. W A STE -C O LLE C TIO N CONTAINERS AND SYSTEMS

2 .1 . SEGREGATION

2.1.1. Philosophy of segregation

The operations o f waste management can be greatly sim plified by the segregation o f the w astes into c la s s e s of m a teria l in such a way that a ll constituents o f any one batch can be dealt with in the sam e way. This ideal situation can seldom be attained in p ractice ,

4


because segregation can in itself be quite a complicated operation which entails significant costs, and may introduce hazards that will outweigh the convenience and enhancement of safety intended to result from the segregation. As in other aspects of waste management, benefits must be weighed thoughtfully against penalties.

Wastes can be divided into classes as shown in Table I. Additional classes could be added, but the 30 shown in the Table are those most commonly considered. The particular class into which a given object will fall will depend upon many circumstances - for example a "high activity" waste from a laboratory concerned with measurement of environmental radioactivity might well be regarded as "possibly active" in an irradiated-fuel treatment plant. An object regarded as "recoverable" in one establishment might be classed as garbage in another, and definitions of "bulky", "disposable-onsite" and "physically dangerous" will clearly vary from place to place.

In classifying wastes thought must be given not only to the wastes themselves, but to the qualifications and number of the people that handle them, and to their equipment and procedures. Wastes that might be a serious problem to an unskilled labourer with a handcart would present no difficulty to a trained crew with a shielded pick-up truck. If wastes are collected frequently, the problems differ from those arising when there are long periods of accumulation.

When wastes are taken from a radioactive area, they should be presumed to be active unless shown to be otherwise. This is particularly true in hot laboratories, where paper tissue and even writing paper may become significantly contaminated. The decision as to what level of contamination should be considered significant for purposes of classification can only be made locally.

2.1.2. Methods

(a) SolidsSegregation should begin at the source. Procedures must be

simple, or they will not be used. In tracer laboratories two waste containers for individual workers may be used - one marked "active" and the other "inactive". In laboratories handling larger quantities of. activity, all waste should be placed in "active" containers. When incineration is being used, each laboratory should have two corresponding bins for active (and, where applicable, inactive) glass and metal objects.

5


TABLE I

CLASSIFICATION OF WASTES FOR SEGREGATION

(1) High activity Low activity Possibly active

(2) Long half-life Short half- life

(3) Solid Liquid Gaseous

(4) Combustible Non- combustible

(5) Acid Neutral Alkaline

(6) Aqueous solutions Organic liquids

(7) Corrosive Non- corrosive

(8) Bulky (compressible) Bulky (not compressible) Not bulky

(9) Physically dangerous (sharp, explosive, fragile, etc.) Not dangerous

(10) Putrefiable Not putrefiable

(11) Recoverable Non- recoverable

(12) Disposable on-site Not disposable on-site

(13) Disposable to garbage or sewer Not disposable to garbage or sewer

The containers should be emptied regularly, and a limit should be set to the maximum permissible external radiation field emitted from a container. The contents should be taken to disposal racks or special areas outside the building, marked "active" or "inactive", the two sites being physically separated to ensure that mistakes in collection are unlikely. Collection of active wastes should be carried out by a different vehicle from the one used for "inactive" solids.

The above technique is applicable to the operation of a laboratory, but the same principles apply to any installation, with appropriate modifications according to the nature and size of the establishment and the kind of radionuclides being used.

(b) Liquids

A decision must be made as to the level of radioactive content below which a liquid will - for purposes of control - be regarded as "non-radioactive". This will depend on the nature of the sewer system, local regulations, and the kind of work being done. Design of facilities must be such that the probability of "active" liquid being

6


put into an "inactive" drain is small, but it must be realized that such accidents will occasionally happen. Experience shows that radionuclides do in fact get into the sanitary sewage system of "active" establishments from time to time even when this is strictly forbidden.

Organic liquids should be segregated from aqueous solutions, especially if ground disposal is used. The possibility of the occurrence of violent reactions or explosions must be prevented - for example nitric acid and alcohol in the same vessel can cause extensive spread of contamination as a result of a reaction.

When the disposal method in use depends for its success upon adsorption or ion exchange, the effect of acids, alkalis, complexing and wetting agents, detergents, e tc., on the system must be considered. For example, in a case where a large volume of plain water with a high total content of radioactive material is being handled, .together with a smaller volume of a "chemical" waste containing a comparatively small total content of activity, it would be unwise to mix the two streams before treatment or disposal.

When disposals are made in cheap but impermeable containers such as glass or polyethylene bottles, it is often advantageous to keep the material in the container rather than to pour it out before disposal.

Many industrial operations have demonstrated the advantages, primarily in efficiency of operation, of providing separate drainage systems for storm water, sanitary sewage, various process and waste streams, etc. In nuclear installations the use of separate drainage systems reduces the volume of waste material requiring special handling and treatment and, at the same time, reduces occupational exposure to radioactivity. Segregation of wastes may range from the collection of radioactive wastes in specially provided containers to disposal through specially constructed sinks and drain lines leading to holding tanks for monitoring and processing. Sometimes a combination of the two systems is used.

One of the advantages of collecting waste materials in special containers is that the volume of material handled at any one time is greatly reduced. However, because these containers must be accurately labelled as to content, this system requires the full co operation of laboratory staffs. This type of system is especially applicable to limited laboratory operations involving tracer levels of radioactivity and small volumes of total solution. Special containers are also useful in segregating long-lived or more hazardous radionuclides that require specialized handling, treatment (decontamination), or disposal.

7


The use of special sinks and separate drainage systems leading to holdup tanks for monitoring is common in many large facilities. Such systems permit segregation of low-volume high-activity wastes from the large-volume low-activity wastes. The form er require tank-storage facilities or treatment, whereas the latter in many cases can be released directly into the environment after monitoring or dilution with non-radioactive waste streams. In some cases even the large-volume low-activity wastes may require treatment (chemical precipitation, filtration, ion exchange, etc.) before discharge to the environment.

2.2. LIQUID-WASTE COLLECTION

2.2.1. Containers

For small volumes of liquid waste which are unsuitable for disposal into the sanitary system or the active drain system because of activity level, half-life, or chemical reactivity, containers similar to the one shown in Fig. 1 are recommended. This consists of an 8-litre polyethylene bottle mounted within a 20-litre paint can, the cover of which has been modified as shown to accept the bottle. Glass bottles may also be used where organic solvents are present, which would attack the plastic, but the general use of glass is not recommended. The use of a drip ring placed around the neck of the bottle is recommended to reduce the possibility of contamination of bottle and can surfaces. If certain wastes generated within a laboratory require segregation from the bulk of the liquid wastes co llected, (for example, organics, hydrochloric acid where subsequent stainless-steel processing equipment may be involved) the can cover should be painted a distinctive colour and have printed on it the type of waste for which it is intended.

Where the size of the laboratory and the volume of waste generated does not warrant a separate active drain system connected to retention tanks, a small stainless-steel sink mounted in a hood and draining into a container such as described above has proven useful, particularly for initial decontamination of heavily contaminated laboratory ware. Because of the limited capacity of the receiver, running water should not be supplied to the sink, and reagent and wash water should be applied by means of plastic wash bottles.

The use of a simple liquid level alarm is recommended for the above application, because of the lack of visibility with, regard to the container contents. Such an alarm may be constructed as follows:


FIG. 1. Liquid-waste collection container

Two stainless-steel rods, bent into a hook-shape so that they will hang over the lip of the bottle, and extending about 10 cm into it, are connected in series with the coil of a low-voltage sensitive relay and an appropriate power supply. When the waste reaches the level of the electrodes, the circuit is completed and the relay closes. The relay contacts complete the circuit between a power source and a bell or buzzer, thereby providing a warning signal. If larger collection tanks are used, commercially available liquid-level indicators should be utilized.

If the volume of low-activity waste is too large to be conveniently handled by means of small containers, or if it is desired to monitor the total effluent from a laboratory before discharge, a retention tank system may be employed. The volume of these tanks will be dictated by the waste-generation rate, time required for analysis, and the rate at which the tank may be emptied, either to the sanitary drain system if permissible, or to treatment facilities if required.

9


In some circumstances, where only short-half-life radionuclides are involved, decay time may be a factor in sizing. Two tanks should always be provided. When one tank is full, flow is diverted to the other, while the contents of the first are sampled and monitored for radioactivity level. It is also advisable to interconnect the tanks so that any .overflow from one automatically spills over into the second. Materials of construction may be carbon or stainless steel, depending on the chemical characteristics of the waste. The use of mild steel tanks lined with glass., hard rubber, plastic or chemically resistant coatings has been shown to be very practical. Stainless steel, polyethylene, glass tubing and ceramics are examples of materials commonly used for the piping system. For low-level wastes where corrosive substances are absent the use of mild steel pipes coated with bitumen has.proved to be satisfactory.

2. 2. 2. Markings and records

In addition to the conventional trefoil radiation warning symbol and a "radioactive" label as shown in Fig. 1, a tag such as the one shown in Fig. 2 should be attached to the filled waste container, particularly if the waste is to be held for decay storage or if it is to be processed and disposed of by someone other than the individual generating the waste. Each container should be sampled and analysed for activity level and pH. Information provided on the tag should include the isotope or isotopes present in the waste, approximate quantity, external radiation levels, date, and information concerning chemical characteristics,5 as well as additional remarks which may be of value to others involved in treating or disposing of the wastes.

Records should be kept of quantities of radioisotopes disposed, in order to permit maintenance of an accurate inventory of isotopes on hand. Examples of radioisotope inventory control forms are shown in Figs. 3 and 4. •

2.3. SOLID WASTE COLLECTION

At any site where radioactivity is handled, a remarkable variety of solid material becomes contaminated to a jgreater or lesser degree. These articles range from ordinary paper, rubber gloves and laboratory glassware tc» large pieces of equipment a:nd even entire buildings. The usual practice for handling solid wastes consists of

10


FIG. 2. Label and tags for marking radioactive waste container

accumulating all hazardous wastes in suitable containers for storage, shipment, further treatment, or ultimate disposal. In handling these wastes, provision must be made to safeguard personnel from radioactive hazards and to prevent the spread of contamination. P rotective clothing is usually required, masks are worn when inhalation hazards exist, and radiation surveys are made before and during handling. Segregation of the wastes into combustible or noncombustible and compressible or non-compressible types as well as by activity level and half-life may be practiced.

Collection practices for low-level solid wastes usually consist of distributing suitable containers throughout the work areas to receive discarded contaminated material. These containers are plainly marked with brightly coloured paint and .radiation symbols to distinguish them from ordinary uncontaminated trash cans. They may range from small cardboard cartons to 208-litre (55 gal) steeldrums.

11


RADIOISOTOPE INVENTORY

S eria l N o; _________________________________________ Isotope ______________O rdered by: ________________________________________ Compound ___________Date ordered: _______________________________ :______ No. m illicu ries re c 'd

Date re ce iv ed :____________________________Location of stored isotope _______________________________________________________________________

Room s used in: ___________________________________________________________ ________________________Location of sinks used fo r disposal: _____________________________________________________________________________________

PORTIONS REMOVED FINAL DISPOSAL OF PORTION REMOVED

Date No. of m icrocu ries Purpose

Decay Sewage Dry waste M iscellaneous disposal

/uCi Date MCi Date M Ci Date Method AiCi Date

FIG. 3. Radioisotope inventory control form


MONTHLY ISOTOPE DISPOSITION REPORT

DEPT.' __________ Radiology______________________________ AUTHORIZED USER ______________ R. Haas_____________________

This rpport covers the month o f June 1961. Please complete this form for all radioactive material you received this month or have on hand from previous months and return to the Department o f Radiology. A report is not necessary if there were no isotopes under your control during the month.

| | No changes since last report.

ISOTOPES REMAINING FROM ORDERS RECEIVED THIS MONTH PREVIOUS REPORTS

ISOTOPE AMOUNTPURCHASE

ORDER NUMBER

1 1131 1 mCi 37143

2 P32 . 10 mCi 41178

3 C14 50 ^Ci 51790

4

DISPOSAL

Given to patients

Into sewage

system 1

Waste disposal service 2

Lost by decay

Other disposal’AMOUNT REMAINING

UNDER YOUR CONTROL (Carry to next report)

TOTALAMT.

NO. AMT.BIOLAMT.

DRYAMT.

AMT. AMT. METHOD AMT. LOCATED IN ROOM(S) NO.

1 0. 56 mCi 30 0. 1 mCi 1 125 fiCi 1 215 MCi 345

2 150 fiCi 2 7. 5 mCi 2 2 .4 mCi

3 10 fjCi 30 jiCi 3 10 pCi CO 2 3

4 • 4 4

5 5 5

6 1.1 mCi 3 mCi 6 6 1. 6 mCi 335

7 0. 045 mCi 4 7 30 pCi 7

8 20 MCi 2 mCi 8 8 0. 23 mCi 376

9 9 9

10 10 10

COMMENTS: Isotopes under "Amount Remaining e t c ." are reported the following month under "Isotopes Remaining etc. Separate orders may be grouped together by isotope when reporting. In all cases amounts received plus that carried from the previous month must equal that disposed plus that remaining under your control.

June 31, 1961

1 Sewage disposal is permitted only if within the limits (MCi/day) set in the Federal Register (Title 10, Part 20.303) and included in a circular provided by the Radiation Safety Service.

2 Incineration o f radioactive material in any form or amount is not permitted.

FIG. 4. Radioisotope inventory control report


FIG. 5. Solid-waste collection can with removable fibre drum insert

2. 3. 1. Coliection-container design and materials

Refuse cans with foot-operated lids are particularly suitable for radioisotope laboratories. These may be lined with removable fibre-board cartons, plastic or heavy paper bags. Figure 5 illustrates a stainless-steel sliding-cover can with its removable 28- litre (one cubic foot) fibre drum insert. This unit is less likely to spread contamination than the container shown in Fig. 6 if the contents are dusty, although the latter type of can is less expensive.

Figures 7 and 8 show a larger drum and cardboard carton which are used for objects that will not fit in the smaller containers, or in cases where a large volume of waste is being produced in a very short time. For small quantities of waste produced in glove or dry boxes, the cartons shown in Fig. 9 have proved useful.

When filled, the tops of the boxes or lids of the cartons and drums should be sealed with wide masking tape, which has been found to be the most satisfactory type of sealant for this purpose.


FIG. 6. "Flip-open" solid-waste can with paper-bag liner

For non-combustible or non~compressible waste such as broken glassware and sharp metal objects which require a container stronger than paper, large metal paint pails such as shown in Fig. 10 may be used. Smaller metal paint cans shown in Fig. 11 may be used for the same purpose or to contain liquid waste which has been solidified in vermiculite, plaster of paris, cement, etc. For the disposal of small animal carcasses which contain relatively large amounts of activity, the 4 litre (one gallon) can has been found suitable. The carcass is placed in a wide-mouthed one-litre jar filled with formaldehyde. The jar is capped, placed in the can, and the remaining space filled with vermiculite or other absorbent material. After inserting the lid on the can, the package may be handled as solid waste.

15


FIG. 7. Large fibre drum of Q. 08 m3 capacity

Large heavy-gauge plastic bags supported inside of 120-litre (32 gal) refuse cans are widely used for solid-waste collection. The use of plastic bags for trash containers has several advantages over the use of cardboard boxes under some circumstances. The waste material, if wet, may seep through the boxes and contaminate floors, whereas the plastic bags will provide much better containment of moist material. If packages are stored in locations exposed to the weather, it has been found that the plastic bags contain the material much better than the boxes.

Figure 12 shows a standard 208-litre (55 gal) barrel which is commonly used for packaging solidified concentrates such as eva-

1 6


FIG. 8. 0.1-m3 cardboard carton

porator bottoms and chemical treatment sludges. They may also be used for packaging higher level solids where shielding is required, as shown in Fig. 13. These drums are prepared for the wastes by pouring concrete around a cylindrical paper or fibre form centred in the drum. The result is a hollow concrete cylinder. Varying wall thicknesses are obtained by using different sizes of form s. After wastes are inserted, the remainder of the drum is filled with concrete, thereby forming a cap which completes the shielding and seals the drum. Experience has shown the need for reinforcing bars to ensure a strong joint between the main body and the cap.

1 7


FIG. 9. Small cartons of about 1-, 2 -, and {-litre capacity for use in glove boxes and other confined areas

2. 3. 2. Marking and records

In general, the comments presented in section 2.2.2. also apply to solid wastes. Labels such as that shown in Fig. 2 should be affixed to the package before it is released for further handling.

Because of the non-homogeneity of solid wastes, it is very difficult to obtain a representative sample for analysis. Samples can sometimes be taken and the activity determined; however, in most cases volume reduction of the sample is necessary, and this can lead to the possible loss of volatile radionuclides associated with the waste. Because of this inherent difficulty, waste containers are generally only monitored externally to ascertain the need for special handling or shielding. In view of these considerations, records should be kept of the type and estimated quantity of activity placed in the container wherever possible.

18


FIG. 10. Large metal paint pails with about 40- and 20-litre capacity, which may be used for collecting non-compressible or non-combustible waste

2. 3. 3. Collection and transportation on-site

Several installations utilize large steel bins, which are designed for rapid attachment and unloading using special trucks, for the collection of filled waste containers and for their transport on-site (Fig. 14). Others use i~t pick-up trucks, 2-§-t stake-body trucks, or other vehicles.

Each building should have two separate and distinctively marked collecting points - one for "inactive" and the other for "active" solids. Packages placed at either station should be monitored, because occasionally radioactive material will find its way into an "inactive" can. It is usually convenient for laboratory staff to carry active solid waste to the collecting point because it is unwise to admit the crew of the collecting vehicle into a building, especially if low background counting is done there. Most building contamination occurs through the tracking of radioactive dust on the feet, and the laboratory staff are likely to be more sensitive to such risks in their own area than are the crew of the active waste truck.

19


FIG. 11: Small cans of about 1- and 4-litre capacity

The vehicle used for collecting radioactive waste will sooner or later become contaminated. It must be an absolute rule that no package placed at the "active" collecting point be contaminated on the surface with removable radioactive material. This may require that some objects be wrapped in polyethylene film or paper, but this must be done not only to protect people handling the waste but also to prevent loose contamination from being detached during transfer. However, with all precautions accidents do happen. It is therefore advisable to have the vehicle lined with some easily disposable, cheap material such as plywood, which can be discarded when it becomes contaminated. Decontamination of an unprotected vehicle is expensive, difficult and time-consuming, and during decontamination the vehicle is out of use.

3. DIRECT DISPOSAL OF RADIOACTIVE WASTES TO SEWERSFor small quantities of soluble radioactive wastes containing

nuclides of short half-lives, the most convenient and generally the

2 0


FIG. 12. Standard 55-gal (208-litre) barrel used to contain solidified waste concentrates

most practicable procedure is disposal into the sanitary sewerage system. This provides a period of delay for decay before the radioactive materials can reach water or food supplies, and provides some degree, of dispersion and dilution.

3.1. DISCHARGE PRACTICES

Several countries have establishe'd guides or limits for the discharge of radioactive wastes to sewer systems. For example, United States regulations require that (1) the material be readily soluble or dispersible in water; and (2) the quantity of any licensed or other radioactive material released into the system in any one

21


FIG. 13. Shielded container made from standard barrel

day does not exceed either (a) the quantity which, if diluted by the average daily quantity of sewage released into the sewer by the installation, will result in an average concentration equal to the 40-h occupational MPC's, or (b) ten times the quantity of such material specified in Table II, whichever is the larger amount. The regulations further require that the quantity of any radioactive material released in any one month, if diluted by the average monthly quantity of water released by the installation will not result in an average concentration exceeding the 40-h occupational MPC's and that the gross quantity of radioactive material released into the sewer system by the installation does not exceed 1 C i/yr. The following examples illustrate the calculations which are made in conducting disposal to sewers under the above regulations:

Isotope to be disposed of: 131I 40-h MPCW: 6 X 10'5 ,uCi/ml

Table II value: 10/^Ci

22


FIG. 14. Transportable steel bin for collecting solid wastes

Case A - average daily sewage flow from installation = 15 000 litres Under 2 (a); 6 X 10'5 /uCi/ml * 10? m l/l X 1. 5 X 104 1/d =

900 MCi/dUnder 2 (b); 10X 10 /tic/d = 100 nCi/d

Therefore, installation could dispose of up to 900 juCi of 131I/d.Case B - average daily sewagfe flow = 1000 litres

Under 2 (a); 6X10-5 X lO&X 103 = 60 /uCi/d- Under 2 (b); 10X 10 mCi/d = 100 /uCi/dUnder these circumstances, installation could operate under

2' (b) and dispose of up to 100 nC i/d .However, under the further restriction that the monthly d is

charge not-exceed-MPCw when diluted with the average monthly flow of sewage, the installation would be limited to 1800 /uCi/month rather

2 3


TABLE II

SPECIFIED RADIOACTIVE MATERIAL

Material (fiCi) Material (fiCi)

105 Ag 1 103 Pd + 103Rh 50m Ag 10 109 Pd 10I6As,71As 10 147 Pm 10198Au 10 210 po 0.1199Au 10 143 Pr 10u°Ba +U0La 1 239 Pu 1^ e 50 226 Ra 0.214C 50 86 Rb 10^Ca 10 186 Re 10109Cd + 109Ag 10 105 Rh 10144Ce + 144Pr 1 106 Ru + 106Rh 13<C1 1 35 s 5060Co 1 124Sb 151Cr 50 46 Sc 1137Cs + 137Ba 1 153 sm 10«Cu 50 u3Sn 10154Eu 1 89 Sr 118p 50

>*+CO 0.1“ Fe 50 182 T a 1059Fe 1 96 Tc 112Ca 10 » Tc 171Ge 50 127Te 10

sH(HTO or’ HjO) 250 i » i e 113 l j 10 Th (natural) 50114In 1 204T1 50192Ir 10 Tritium. See H3 25042k 10 U (natural) 50140La 10 233 u 152 Mn 1 234y _ 235 y 5056Mn 50 48 v 199Mo 10 185w 1022 Na 10 90 y 1MNa 10 91y 195 Nb 10 ffiZn 1059 Ni 1 Unidentified radioactive materials or63Ni 32 p

110

any o f the above in unknown mixtures 0.1

2 4


than the 3000 ^Ci which would be discharged if the installation d is posed of the m axim um allowed under 2(b) every day.

W here a com bination o f isotopes is involved in known amounts, the lim it fo r the com bination should be derived by determ ining, fo r each isotope, the ratio between the quantity present in the com bination and the lim it oth erw ise estab lished fo r the s p e c if ic iso top e alone, ex p ressed as a fra ction . The sum o f such fra ction s fo r a ll o f the is o to p e s in the com bin ation m ay not e x ce e d " l " ( i . e . , "u n ity " ) .

The United States Regulations sp ecifica lly exempt excreta from individuals undergoing m edical diagnosis o r therapy with radioactive m aterials from these regulations.

The Soviet Union's health and safety regulations governing work with rad ioactive m ateria ls and sou rces o f ionizing radiation estab lishes d ischarge lim its for disposal to sew ers. The follow ing paragraphs reproduced from these regulations apply:

Paragraph 98

"L iquid and solid wastes from installations shall be considered radioactive i f their activity (in C i/l itr e and C i/k g ) is m ore than 100 tim es the m axim um p erm issib le concentration in open water (C i/lit re in the case o f m a teria ls with a h a lf- life o f up to 60 d, o r m ore than ten tim es the maximum perm issib le concentration in open water in the case of m ateria ls with a h a lf-life of over 60 d ".

Paragraph 138

"W aste w ater from installation show ers and laundries and from the washing of flo o rs and walls, etc. may be discharged into the norm al sewage system provided that its activity, without p relim in ary dilution , does not exceed the le v e ls indicated in paragraph 98 and provided that a tenfold dilution by non -rad ioactive waste water is ensured in the in sta lla tion 's co lle ctin g tank. W aste w ater re lea sed d irectly into bodies o f water must not contain m ore activity than the maximum p erm issib le concentrations for open water. "

Paragraph 163

"Solid and liquid radioactive waste products containing sh ortlived isotop es with a half life no g rea ter than 15 days are kept fo r a tim e that w ill guarantee a d e cre a se in the activ ity to the values indicated in paragraph 98, after which solid radioactive waste p ro

2 5


ducts are rem oved with the usual rubbish, and liquid waste products are disposed of in the sew erage (see paragraph 138) after being o ffic ia lly reg istered with the appropriate r e p o r t ."

Regulations in the United Kingdom provide for the establishment o f d isch arge lim its on a c a s e -b y -c a s e basis by the appropriate authorities because o f w idely d ifferin g lo ca l conditions.

In a government publication entitled "The Control of Radioactive W astes", C-MND 884, broad prin cip les concern ing waste m anagement are stated and som e general guidance given. The follow ing is a section of this document dealing with discharge of radioactive waste to sew ers . , .

"in the ca se o f d isch a rg es to sew ers the fo llow in g c o n s id e r ations a r ise : .

(i) the contam ination of the drains, which might present a hazard during repa ir p ro ce s s e s ;

(ii) the contam ination o f the sew age it s e lf which cou ld endanger m en w orking in the sew er;

(iii) the contam ination o f the purified sew age effluents which may- a ffect their ultim ate d isch arge ; *

(iv) the build-up o f'ra d ion u clid es on filte r beds .( v ) the possib le use o f sewage sludge.i" E ach radionuclide d ischarged m ay w ell behave d ifferen tly and

contribute to one or m ore of the above potential hazards-. In con sequence, the fixing o f general • lim its is not. easy, but a level of 10‘4 mCi/ml in the sewage flow from the user establishments would norm ally be p erm iss ib le . E xceptions would have to be made in the case of the m ore hazardous isotopes or, for example, where a large fa cto ry or hospital drains to a sm a ll’ v illage sew age w ork s.

As to the use of sewage sludge as; a fertilizer, there is evidence that those radionuclides, such as radio strontium , which are taken up rea d ily by plants, are not adsorbed on the sludge to any great extent. -’ The above suggested lev e l of 10'4 juCi/m l in the d isch arge a s sum es a hundredfold dilution in the m ain sew er".

The International C om m ission on R adio log ica l P rotection , in their publication Num ber 5, states: "V alid estim ates o f quantities o f rad ioactive w astes from other sou rces , including hospital la b o ratories, which may appropriately be released into sewerage systems- should be based on lo ca l fa ctors . Under conditions rep resen tative of m ost sm all la b ora to r ie s and of hosp itals using d iagnostic quantities of radionuclides, perm issib le discharges w ill correspond to concentrations (averaged.over, say, 1 month) in the range of 10-4

2 6


to 10-5 /uC i/m l in the effluent from the establishm ent. Such leve ls w ill usually be sufficient to allow disposal of the wastes which arise . Even with m ore active wastes, re lease to the sew erage system may be accep tab le , fo r exam ple, i f the system does not d isch a rge to a potential sou rce of drinking w ater. On the other hand, som e uses* o f w ater m ay req u ire even lo w e r con cen tra tion s o f ra d ion u clid es than does use fo r drinking w ater. In such -cases, the concentration of radioactivity in the receiv ing water should1 also be considered and related to the uses o f the w ater; U ses fo r agricu lture and industry should be exam ined". The ICRP further states: "D irect discharge into the sanitary sew erage system is 'particu larly sa tis fa ctory for the d isposa l of excreta fro m ‘patients given radioactive m ateria ls •'in m edical diagnosis and therapy". ' ■ ■■ ■ •

"T his recom m endation is made because of their putrescent nature and the suitability of sew erage system s for their d isposal. The relation sh ips betw een quantities of rad ioactiv ity r e le a s e s , ra d io active h a lf- liv e s , and quantities o f sew age handled by the system w ill usually be such as to require no specia l precautions other than! those which may be n ecessary to protect plum bers and sewage w orkers near points of d ischarge. " •- . . . . • •- ;

To re s tr ic t the p oss ib ility b f general sink 'contam ination each laboratory or group o f room s should designate1 one sink as a "hot"- sink. Only this sink should be :us.ed for first* cleaning 'of contam in-' ated glassw are (initial washings -should be co llected in carboys) or fo r d isposing o f liquid w astes: It should be labe lled w ith-cautiontape or-tags both on top and on the drain . Traps and' p ipes should be m onitored before d isassem bly for repairs to avoid radiation ex posure of maintenance personnel by radioactive m aterials that) have' been precipitated, adsorbed, or plated on exposed1 surfaces.- •

3 .2 . DILUTION TECHNIQUES AND COMPUTATIONS- - !

1 The United States National Com m ittee on R adiation-Protection, in their'Handbdok 49, has1 published recom m endations'for waste d is posal of 32P and 131I for m edical u sers . These recom m endations in large m easure address them selves to the procedures which iriay be em ployed to protect sewage plant w orkers in cases where large therapeutic doses of 32P o r 131I are em ployed . The fo llow ing is va sum m ary of the d iscussions and recom m endations presented in this Handbook.' ‘ ................... ' ■

2 7


3. 2 . 1 . General considerations

It is not rea lis tic to in sist on dilution of radioactive w astes in sew age to the lev e l established as p e rm iss ib le in drinking w ater. Ingestion o f this fluid w ill o c cu r only as the resu lt o f an accident, and the hazard should be considered from this point of view . In a c tual practice , high concentrations in pipelines w ill occu r over short p eriod s ; the duration o f such high transient concentrations w ill ha about 30 s . It m ay be assum ed that in the case o f accid en ta l im m ersion in sew age, not m ore than 0. 25 litre s would be sw allow ed in 30 s. It would be reasonable to assum e further that this w ill not happen often and that therefore the ingestion of a perm issib le tracer dose of the isotope could be tolerated in the accident. The p erm issible tracer dose of either 32P or i31I is approximately 100 mCi; this would be contained in 0. 25 litres if the concentration were 0. 4/uCi/m l. To allow fo r an additional m argin o f safety the value fo r m axim um short period contamination of 32P or 131I in sewage used in the fo llow ing calculations is 0. 1 /LtCi/ml (0. 1 m C i/lit re ).

In the same sewage plant an external radiation hazard to plant personnel might be thought to exist in case of accidental im m ersion ina concentration of 0. 1 m C i/litre . This concentration would be possib le , but im probable, because of the dilution of the high transient con cen tration s in th'e p ipelin es by the tim e they reach ed the plant (unless the institution has its own treatm ent plant). Even in a con centration as high as 0. 1 m C i/litre , however, the radiation received on the su rface o f the body by such im m ers ion would be re la tiv e ly low . F or an im m ersion o f one hour, the calculated dose on the surface of the body is 0 .2 rem for 32P and 0.1 rem for 1311. If the sludge from institutional o r com m unity sew age treatm ent plants is to be used as a fe rt ilizer , the hazard to the general population must be con sid ered . It m ay be assum ed that the concentration of 131I in the sludge cake w ill not exceed that of the sewage as rece ived , and that this concentration w ill be reduced depending upon the tim e in volved in d igestion , conditioning, filtra tion and storage o f sludge. At this level the danger from exposure to the fe r t ilize r is obviously le s s than that to the sew age plant w o rk e r . F rom the above co n siderations, it appears that the lim iting factor in the determ ination o f the quantity of rad ioactive isotopes that may be d ischarged daily to a sewage treatment plant w ill be the rate of water flow at the plant.

The sim plest way to d ispose of radioactive wastes encountered in connection with the adm inistration of the m ateria l to patients is,

2 8


o f cou rse , to allow the patient to use the toilet without res tr ic tion . This w ill be ca lled to ile t d isp osa l. An alternative is to pour the radioactive m aterial into the sink. In this case it may be preferable f ir s t to put it into a 4 -l it r e (one gallon) bottle, f i l l this to the top, using tap water if n ecessary , and pour this into the sink. This w ill be called batch-bottle d isposal. The batch disposal (toilet or bottle) o f a single sam ple takes from three to 30 s; by the tim e it a rr iv es at the sew age plant it m ay be con s id ered as diluted with the p r o portional part of the 24~h sewage flow . In an institutional or municip a l system having an averaged d ry -w ea th er flow o f one m illion gallons (3800 m3) a day at the treatm ent plant, the flow in four secon ds is about 100 l i t r e s . This w ill dilute 10 m C i to a con cen tration of 0 .1 m C i/lit r e , which has been shown above to be without p racticab le hazard to plant p erson nel. W hile it would be expected that this concentration would be further reduced in treatm ent tanks, it m ight again be in crea sed in sludge con cen tration . F o r lack o f accu ra te data on these points, it is fe lt w ise at p resen t to set the lim it o f 10 m C i fo r a sin g le batch d isch a rg e , p er m illio n ga llon s (3800 m 3) o f flow a day. In any case where the daily d isch a rge of 32P and 1311 exceeds 10 m C i/d per m illion gallons of sewage flow the d isp osa l should be m ade in sm all batches at intervals,, o r through a constant head o r i f ic e o r s im ila r m eans to m aintain a re la t iv e ly uniform d isch arge ov er a period o f six daylight hours a day or lon ger. One such dev ice is the constant drip d ischarge bottle d escribed b e low , and shown in F ig . 15. With the insta lla tion and op era tion o f such a uniform d isch arge d ev ice , 100 m C i o f these iso top es m ay be d ischarged in any s ix -h ou r daylight period into a system having a on e -m illio n -g a llo n average d ry -w eath er flow . The p e rm iss ib le d ischarges for other sewage flows w ill be proportional to the above, e .g . , 50 mCi for 0 .5 m illion gallons (1960 m3) daily.

The above lim its are subject to rev is ion in any com m unity on the basis of actual radioactive m easurem ents in the sludges. Where quantities o f rad ioactive iso top es o f the o rd e r suggested by these lim its are discharged, radioactive m easurem ents o f the sludge should be made and the lim its revised , if n ecessary . Responsible offic ia ls at institutions using considerable quantities (100 mCi or m ore a week) o f rad ioactive iso top es , should in form and co -o p e ra te with m unicipal and regional health authorities, in order that proper arrangem ents fo r m onitoring m ay be made when there is any question r e garding safety o r hazard.

29


■FIG. 15.. 4 -litre:bottle set-up for constant-pressure drip discharge.

3. 2. 2. Calculations for disposal practice ' '

On the basis o f the above discussed general considerations, fo r m ulas have b een d eve lop ed '‘for com puting p erm issib le d ischarge of 32p o r 1311 by variou s m ethods,, in system s with d ifferen t average w ater flow s. C alcu lations based on these form ulas fo llow :

' "'.V' N ™ = M / q • • f •_ ( I V

wheire Nm = maximum contamination'Occurring iri a Water column flowing through the sewer (/LtCi/litre),

M = activity, in m icrocu ries , introduced in a single disposal event,

q - water used during a single d isposa l event; fo r toilet flushing, q .=12 to' 3 2 litre s.

(a) T ra ce r and therapeutic dose's u p ’to1 1 m C i. L es ’s than 25% is usually excreted , during the fir s t day, arid this o cd u rs in not

le s s than in four evacuations. M 'is , therefore, usually not m ore than 6% of the administered dose, or 60 /uCi.

3 0


The maximum contamination that can occu r i f only flushing water is considered is calculated from E q .( l ) :

Nm = 2X lO-3 to 5X 10"3 juCi/ml

i . e . , below p e rm iss ib le le v e l o f 100 / jC i /l i t r e (0 .1 )LtCi/ml).(b) Therapeutic doses o f 32P . The la rgest single dose used at

present is 7 m C i. This resu lts in M equals 420 juCi (6% of adm inistered dose). Again considering toilet-flushing water alone1we get from equation (1 ):

Nm =0.013 to 0. 035 AiCi/ml

i . e . , a lso below p erm issib le level.(c) 131I in treatm ent of hyperthyroidism . A single dose w ill ra re ly

exceed 10 m C i. In ca ses o f hyperthyroid ism requ iring such a high dose , the fir s t 24~h ex cre tion w ill be not m ore than 30%, and it may be assum ed that not m ore than half w ill be evacuated at one tim e: M equals 1. 5 m C i. M axim um contam ination that w ill occu r, if only flushing water is considered calculated from equation (1 ):

Nm = 0. 047 to 0. 125 £iCi/m l

i . e . , still essentia lly at p erm issib le level.(d) Treatment of thyroid cancer. A single dose o f 100 m Ci is ra re

ly exceeded . H ow ever, when uptake by m etastases is low and thyroidectom y has been perform ed, up to 90% may be excreted within the firs t 24 h. Again M is equal to half o f this value: ,M equals 45 m C i. C onsideration of flushing water alone w ill lead to e x ce s s iv e va lu es o f Nm . Taking into account the average w ater use o f 550 litre s p er person per day, varia tion s in flow during day and night, and using 100 fxC i/l i t r e as the lim itin g con cen tration , (N m ), T able III w as p rep a red .F o r the d isp o sa l o f la r g e r quantities the con sta n t-d r ip bottle

d escr ib ed below m ay be used . The la rgest amount that can be ex creted by one patient per day w ill seldom exceed 100 m Ci. It is perm iss ib le to d isch arge this amount by the constant drip bottle, p ro vided the dry-w eather flow to the sewage treatm ent plant is one m illion gallons per day or m ore .

31


TABLE III

PERMISSIBLE ACTIVITIES IN MILLICURIES FOR SINGLE DISPOSAL EVENTS

Number o f beds Number o f people aToilet disposal

Batch bottle

disposal day*3Day Night

(mCi) (mCi) (mCi)

25 50 1 to 4 1 to 3 1

50 100 2 to 4 1 to 4 2

100 200 2 to 5 2 to 4 4

200 400 2 to 6 2 to 5 8

300 600 2 to 8 2 to 6 12

■ 500 1000 4 to 11 2 to 8 20

1000 2000 6 to 20 4 to 13 40

a It is assumed that the hospital population is equal to twice the number of beds, b Batch discharge o f a full 4-litre bottle (add tap water if necessary) emptied into a sink, not

into a toilet.

3. 2. 3. Constant-drip discharge bottle

A sim ple d ev ice fo r d ischarg ing 4 litre s (one gallon) o f liquid w aste at a constant rate is illu strated in F ig . 15. It co n s is ts o f a 4 -litre jug and a tw o-hole stopper with two glass tubes. One g lass tube (a ir inlet tube) rea ch es to about 6 cm above the bottom o f the bottle . The second glass, tube (outflow tube) reaches to the bottom . A rubber tubing is attached to this outflow tube, and a capillary glass tube is attached to the other end of the rubber tubing. The capillary tube is attached to the bottle (with w a ter-p roo f tape or adhesive p la ster) so that the top o r if ic e o f the ca p illa ry is 5 cm below the low er end o f the a ir - in le t g lass tube.

When the bottle is filled and the stopper with the tubings in stalled, it may be set in a sink and the flow started by pumping air into the open end o f the a ir - in le t tube. This m ay be conven iently done by attaching a piece of rubber tubing to this open end and using

3 2


C A P ILLA R Y DIAM ETER (mm)

FIG. 16. Length of capillary tube as a function of its diameter for an emptying time of 6 h for a 4-litre bottle with a water-head of 5 cm

the inflating rubber bulb with the re lease valve of a b lood -p ressu re m onom eter. A fter the liquid begins to flow from the ca p illa ry the flow w ill be maintained by syphon action . The p re ssu re is d e te r mined by the leve l d ifference between the low er end of the a ir-in let tube and the ca p illa ry o r if ic e (this lev e l d iffe ren ce is m ade equal to 5 cm ). The p ressu re w ill rem ain constant until the liquid lev e l inside the bottle drops below the end of the a ir -in le t tube; then the p ressu re w ill gradually drop until the lev e l sinks below the end of the outflow tube.

F low -ra te is determ ined essentia lly by this p ressu re and by the length and inner diam eter of the capillary tubing. Suitable capillary tubes with an inner d iam eter betw een 4 and I 4 -m m are gen era lly available from laboratory equipment d ea lers . F igure 16 is an em p ir ic a l plot indicating the requ ired lengths o f ca p illa ry tube o f various inner d iam eters, fo r a flow rate at which the bottle w ill be em ptied in 6 h.

The actual emptying time with the set-up described will generally be within about 30% of 6 h, because of a num ber of fa ctors which it is difficult to control (for instance: change of v iscos ity with tem perature, variations o f the bore o f the same capillary, e t c . ) . This uncertainty in emptying tim e, however, is satisfactory for practical pu rposes.

3 3


3. 2. 4. Recommendations for disposal

A fter the daily or weekly w aste-d isposal level for an institution has been determ ined, the method o f d isposa l must be decided from considerations of safety both to sanitation w orkers and sewage-plant personnel. It has been pointed out that for the first, a transient concentration o f 0. 1 m C i/l it r e should be p e rm iss ib le , while fo r the second, a single-batch d ischarge of 10 m Ci, or a 6~h constant d is ch arge o f 10 m C i, per m illion ga llons (3800 m 3 ) o f w ater flow is sa tis fa c to ry .

The values tabulated in Table III for institutions of various sizes are valid provid ing the dry -w eath er flow to the sew age-treatm ent plant equals or exceeds a m illion gallons a day for each 10 mCi d is charged. If the flow to the sew age-treatm ent plant is not great enough to perm it the use of Table III, a constant-drip bottle should be used, on the basis o f a 100 m Ci d isch arged in this m anner p er m illion gallon per day w ater flow to the sew age-treatm ent plant. Thus, fo r le ss than 10 m Ci the lim iting fa ctor is transient con cen tration in pipeline, as determined from the Table; for larger amounts the lim itin g fa cto r m ay be concentration at the sew age plant to be determ ined by total daily flow to this plant.

3. 2 . 4 . 1 . Sm all quantity d isp osa l

In d iagnostic and therapeutic uses of 32P, in d iagnostic use of 1311, and in treatm ent o f h yp erth yro id ism with 131I, patients m ay use the to ile t without any in stru ction s o r r e s tr ic t io n s .

3. 2. 4. 2. C arcin om a treatm ent with 131I

A . H ospitals

It is not possib le to form ulate sim ple instructions for the levels excreted in the treatm ent o f carcin om a; m illicu r ie s excreted by a pa rticu la r patient m ust be calcu lated on the basis of dose and uptake, and method o f d isp osa l must be decided from Table III.

When the quantities fo r d isposal exceed the p erm issib le values either from Table III o r from the general sew age-plant lim itations, the follow ing methods can be used:

(1) Storage fo r decay to p erm issib le activity;(2) D istribution o f activ ity in a num ber o f on e -g a llon bottles

each containing perm issib le activities; filling bottles with tap water;

3 4


su ccess ive emptying of these bottles in the sink at proper intervals, (batch bottle d ischarge);

(3) S ix-hour d isposal by constant-drip bottle.In deciding on the use of one of the above m ethods, o r a batch -

bottle d isp osa l in accord an ce with Table III, radiation exposu re to laboratory personnel must be con sidered . This is a m atter fo r the institu tional rad iation sa fety o f f ic e r and outside the s cop e o f this section .

B . A partm ent houses and sm a ll hom es

On the basis of Table III it would appear that toilet d isposal would ra re ly be p erm iss ib le fo r patients treated without h osp ita lization. However, it must be considered that in these cases, for one hom e, only a single patient is involved; the probab ility o f sev era l persons in the same building being treated for cancer with radioactive isotopes at the sam e tim e is n eglig ib le . High contam ination in the lo ca l sew age system w ill thus o ccu r seldom , w ill be o f b r ie f duration, and w ill be prom ptly rem oved by further flow o f sew age and by radioactive d ecay . This con sideration p erm its re com m en dation of sim ple toilet d isposal for patients who are not hospitalized.

3. 2. 4. 3. Simultaneous d isposal of 32P and 1311

When 32P and 131I are used sim ultaneously in an institution, d is posa l ru les can be based on the sum of the m illicu r ie s o f both is o to p e s .

3 . 3 . BEHAVIOUR OF RADIONUCLIDES IN SEW AGE- TRE ATM E N T PLANTS .

B io lo g ica l treatm ent p r o c e s s e s are ra th er in e ffic ien t for .th e rem oval of radioactive m aterials from water. There have been some d irect determ inations of the behaviour of radioisotopes during prim ary sedim entation . V ery little rad ioiod ine settled out from crude sewage to which had been added urine from patients who'had received diagnostic d oses . L ess than 5% absorption on p rim a ry sludge was found for radiosodium , radiocobalt, radiophosphorus and radioiodine at concentrations in the range o f 0 .5 to 50 juCi/litre.

3 5


The most important characteristics determining the behaviour of a radioisotope in biological purification are its chemical nature and the concentration of its inactive isotopes in the sewage. The active isotopes are normally present in such small concentration that their presence does not affect the total concentration of isotopes of that element in the sewage. Consequently, the degree of removal of a radioisotope, expressed as a percentage of the initial concentration of the radioisotope, is independent of that concentration.

The behaviour of several radioisotopes in sewage can be at once predicted without reference to published experimental work. Tritium is the radioisotope of hydrogen and will normally be discharged to the sewers as water or, less often, as organic compounds containing tritium. The latter will usually be oxidized to water on the filters. Ultimately the tritium will be flushed out with the effluent or the sludge and will not remain at the disposal plant. The ionizing radiation from tritium is so soft, and the permissible level in drinking water is so high that much larger amounts than are commonly used would be necessary to produce a hazardous level.

Radiosodium and radiopotassium will behave like salt and pass through the disposal plant. Experimental work has indicated that radiobromine will behave in a similar fashion. Radiocarbon will be oxidized and released either as gas or in the effluent and sludge and no concentration at the plant can occur. Radiosulphur will mostly be in the effluent. Radiocalcium and radiostrontium also will pass through.

Most experimental work on the biological behaviour of radioisotopes has concerned radioiodine and radiophosphorus. Their behaviour depends largely on the concentration in sewage. For radiophosphorus, removals as high as 80 to 90% occurred on trickling filters at levels of 1 ppm in the sewage, but these fell to 20% at levels of 6 ppm. With the high phosphorus content of sewage now common from the use of synthetic detergents, the latter figure is likely to be more representative of what will be achieved in practice.

Reports of the behaviour of radioiodine on trickling filters are contradictory. Results range from only 5% removed to 75-90% removal. The difference is perhaps attributable to different levels of inactive iodine in the sewages used by the investigators. Studies have shown that the proportion removed by the filter depends on this factor, as with radiophosphorus.

Another radioisotope whose behaviour has been investigated on trickling filters is radiocobalt. Removal of 80% was observed. It

3 6


is probable that the behaviour Of radiocobalt, like that of radiophosphorus and radioiodine, is dependent on the concentration of inactive cobalt in the same chemical form, and it is not clear how representative of the general behaviour are the results quoted above.

The removal of radioisotopes by activated sludge is similar to that by trickling filters. Here again the removal of radioiodine ranges from 90% to only 10-15% removal, again probably due to the concentration of inactive iodine in the sewage. There is little removal of radiocalcium, radiosodium and radiobromine. 85% removal of radiophosphorus and radiocobalt has been reported.

Thus, there is experimental evidence for the behaviour in biological treatment of most of the radioisotopes at present used in research. There is enough evidence to assess the situation in a system of sewerage and sewage disposal with sufficient accuracy. It can be stated confidently that the amount taken up on filters or in activated sludge is not sufficient to be hazardous to workers. Only if relatively large amounts of a long-lived radioisotope like cobalt-60 were discharged to the sewers would hazardous levels build up. As stated in section 2. 1, where quantities of radioisotopes approaching the permissible limit are discharged to the sewer, sludge measurements should be made to assess the validity of these limits for the particular conditions which exist.

4. LIQUID-WASTE TREATMENT TECHNIQUES SUITABLE FORUSERS OF RADIOISOTOPES

4.1. BATCH-CHEMICAL TREATMENT

Most of the chemical methods involved in radioactive-waste treatment are adaptations of standard water-treatment practice, and have been extensively used for processing large volumes of lightly contaminated wastes using equipment designed for continuous operation.

However, it is feasible to utilize batch chemical treatment where the volumes involved are small, and where the nature of the wastes points to chemical treatment as the best method.

It is suitable for application where the required reduction of activity in the waste is small, since decontamination factors (DF)*

^ _ . . r ___ con c. o f radioactivity in feedDecontamination factor = DF = ------------- ------------------------L-----------------con c. o f radioactivity in effluent

3 7


of more than 10 are seldom attained. If it is possible to select and optimize the process for a single radioactive species, however, much better results may be achieved. Advantages of the process are low cost, the ability to handle a wide range of solid content in the feed, and in some cases the production of a waste sludge volume which is relatively independent of feed solids content. Suitable handling, storage and disposal facilities, of course, have to be provided for the resultant radioactive sludge.

Chemical coagulation involves the de-stabilization, aggregation and binding together of colloids. These colloids form chemical floes that adsorb, entrap, or otherwise bring together suspended matter. Commonly used c.oagulants are alum and iron salts, which are precipitated as aluminium and iron hydroxides.

In chemical precipitation, chemicals are added to produce an insoluble, somewhat heavy precipitate, which removes radioactivity as it settles out of solution. The radioactivity may be removed by direct precipitation, by adsorption on the resultant floes, and by entrainment in the settling precipitates. Precipitation and flocculation may occur simultaneously, as in the case of calcium phosphate, ferric ion, or lime-soda treatment, or may involve the formation of a suspension which requires the subsequent introduction of a flocculating agent such as ferric hydroxide. Details of these and other processes are discussed below.

There are three steps in the process of coagulation and flocculation:

(1) Addition of coagulating chemical to the liquid waste. To ensure that the chemicals are distributed uniformly and promptly throughout the liquid rapid agitation or mixing must occur. . This is extremely important; otherwise the coagulant would slowly disperse and the initial chemical reactions would be localized at the point of coagulent introduction. This may produce reactions other than those intended.

(2) Coagulation, i .e . complex chemical and physico-chemical reactions and changes occur leading to the formation of finely divided precipitating solids. '

(3) By means of gentle agitation of the liquid, flocculation takes place, i .e . , the finely divided particles contact and adhere to one another and form progressively larger floe . This insoluble flocculent precipitate will then settle out carrying with it the colloidal materials present in the liquid.

3 8

I


W ell-formed floe settles at a rate of about 3 m/h (10 ft/h) although it is generally advisable to assume a rate of 0. 75 m /h (2.5 ft/h) to allow for periods of poor flocculation.

In the development of a method of treatment, representative quantities of the waste material to be treated are collected and are tested in the laboratory. Most frequently jar tests of one kind or another are run to determine specific chemicals needed, optimum quantities required, and conditions necessary to obtain the desired degree of removal. The apparatus for performing jar tests is shown in Fig. 17.

FIG. 17. Apparatus for conducting jar tests

For a series of tests, a given quantity of the waste, generally 500 ml or one litre, is placed in each of several suitable beakers. Varying amounts of the necessary chemical or chemicals are added to the different beakers, and rapid mixing is provided for periods up to one minute to ensure complete contact between the wastes and the chemicals. Then the stirrer speed is reduced to permit maximum precipitate formation. This slow stirring may be continued for approximately 30 mlji after which the precipitate is allowed to settle. Samples of the waste before treatment are compared with aliquots of the supernatant after treatment to determine the conditions under which maximum removal of activity occurs. These findings then serve as the basis for chemical additions under full- scale conditions.

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With regard to equipment for batch-chemical treatment, the main requirements consist of a collection tank, where wastes are accumulated for treatment, and a treatment tank with steeply sloping conical bottom section-on which a stirring device is mounted. The chemicals are added to the waste in the tank and the mixture is stirred rapidly for 1 to 10 min, then stirred slowly, for a longer period (10 to 60 min). After the stirring has been completed and good flocculation achieved, the precipitate is allowed to settle to the bottom of the tank (this may take several hours) after which the sludge is drawn, off.

4.1.1. Lim esoda ash treatment

Studies on the removal of radioactive contamination by standard water-treatment procedures have shown that an appreciable removal of polyvalent cations can be achieved by the lime-soda ash-softening process. Lime (calcium hydroxide) and soda ash (sodium carbonate) removals reported for a variety of radionuclides are summarized in Table IV. These data show that reasonable amounts of chemical will provide a 90% or better removal of soluble barium, lanthanum, strontium, cadmium, scandium, yttrium and zirconium-niobium, but that much larger quantities of chemical were ineffective for the removal of caesium-barium and tungsten. It will also be noted that lime alone was effective for the removal of 95Zr~95Nb. Lime and soda-ash softening has also been tested for the removal of iodine, but the results have been negative.

Treatment with lime and soda ash finds its greatest use in the removal of strontium. In this process, hydroxides and basic carbonates of the heavy metals are precipitated, while strontium carbonate is precipitated together with calcium carbonate in mixed crystals, the efficiency of removal of strontium being directly proportional to the degree of softening. Therefore, for the effective removal of strontium, it is imperative that the calcium hardness be reduced to a very low value. Such a requirement suggests a considerable excess of soda ash in the water during treatment and an initial reduction of calcium hardness to a low value. Repeated additions and precipitations of small quantities of calcium could then be used to reduce the radioactive strontium to a very low amount.

Studies have shown that an excess dosage of lime and soda ash is in fact instrumental in removing greater quantities of strontium. Thus the amount of material added is theoretically sufficient tosatis-

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TABLE IV

APPROXIMATE MINIMUM COMBINED DOSAGE OF LIME AND SODA ASH TO GIVE STATED REMOVAL*

Chemical dosage (grains/gal)** for stated removal

Nuclide50%

Lime Soda75%

Lime Soda90%

Lime Soda95%

Lime Soda99%

Lime Soda

i4oBa - ‘«La 2 2 4 2 6 4

»9Sr 4 3 5 5 7 9 20 20

115Cd 2 2 3 3 4 4 4 4 6 4

46 Sc 3 3 3 3 5 5

91Y 2 2 4 4 6 6 12 6

95 Zr - 95Nb 2 0 5 0 12 0 17 0 22 0

i37Cs - 137 mga 48 48

185 W 48 48

* Minimum combined dosage is defined such that, o f the variable dosages studied, the number o f grains per gallon o f lime plus the grains per gallon o f soda ash is a minimum.

* * 1 grain/gal = 17.1 m g/litre.

fy the stoichiometric requirements of the raw wastes and also to provide a pre-established excess of lime and soda ash in the system. The amount of lime feed is based upon the consumption of this chemical by carbon dioxide, calcium hardness, and magnesium hardness in the waste, and the excess to be introduced. Sufficient soda ash is added to combine with the non-carbonate hardness and the excess lime. The effect of excess dosages in the range of 20 to 300 mg/1 of lime and soda ash are shown in Table V.

The use of an illite clay in the amounts of 30 mg/1 in conjunction with the lime-soda ash process has been shown to increase caesium removals from about 16% to approximately 55%.

4. 1 . 2 . Aluminium and ferric hydroxide coagulation

Another standard method of water treatment consists in adding filter alum (aluminium sulphate) or ferric salts and raising the pH

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TABLE V

RESULTS OF LIME-SODA ASH TREATMENT FOR REMOVAL OF STRONTIUM

TreatmentRemoval o f activity

(% )'

Stoichiometric amounts 75.0

20 ppm excess lime- soda ash 77.0

50 ppm excess lime-soda ash 80.1





of the solution until aluminium or ferric hydroxide is precipitated; in some cases a mixed floe is formed. Increasing alkalinity causes the precipitation of heavy metals as hydroxides, while the bulky floe acts as an efficient scavenger by absorbing or co-precipitating with the other hydroxides. In some cases it may not be necessary to add carrier aluminium or iron, but in most low-level wastes the pre- cipitable salt content is low, and floe-forming chemicals are necessary. Dosages in the range of 10 to 200 mg/1 are currently used. Such a process will in general remove efficiently all cations other than those of the alkali metals and alkaline earths. It will have little effect on anions, since the hydroxide floe is negatively charged.

It has been shown that in the treatment of mixed fission product solutions alum-sodium carbonate floes are superior to alum-sodium hydroxide floes, since in the former case strontium is also removed as carbonate; iron floes are slightly superior to aluminium floes, and may be operated at much higher pH with a consequent improvement in removal. Iodine removal could be improved by introducing into the floe suitable absorbants or specific precipitants such as activated carbon, copper sulphate, or silver nitrate; these produce up to fivefold increase in removal on alum floes. The results of work on the removal by hydroxide floes are summarized in Table VI.

Hydroxide floes are highly efficient in removing alpha emitters from low-activity wastes. Plutonium is removed readily on alum

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TABLE VI

REMOVAL OF ACTIVITY ON HYDROXIDE FLOCS

Element FloeRemoval

m

Sr Alum-NaOH 0.

Alum-N^COj 6

FeCl-N a Co3 2 32

Y Alum-NaOH 35Alum-Na2C 03 96

FeC l-N aC O3 2 3

9

Ce Alum-NaOH 89Alum-Na2C 03 94

FeCl3-Na2C03 95-6

I Alum-NaOH 15Alum-Na2COs 25

or iron floe, preferably the latter at pH 10; colloidal plutonium hydroxides are efficiently removed at pH 10.3. Lime, sodiumhydr- oxide, or ammonium hydroxide all give good removals, but the best floe is obtained by using lime for neutralization. In the presence of citrates and other complexing agents, which may be present in waste from alpha laboratories, decontamination centres, e tc ., a higher pH must be used to obtain both good removal and a satisfactory floe, the two factors running in parallel. At pH 12.0 up to 400 ppm citrate may be tolerated; higher alkalinity is also necessary in the presence of complexing agents for removal on phosphate floes.

4.1.3. Phosphate coagulation

Since the heavy metal phosphates are less soluble than the hydroxides a better removal would be expected on alkaline phosphate floes; since strontium phosphate is not soluble a considerably greater over-all removal of this nuclide should be expected. If the waste water is a hard one, the addition of phosphates followed by alkali to

43


pH 10 will precipitate basic calcium phosphate, which forms a good floe and acts as a carrier for the precipitated activity; in the case of soft or de-mineralized water it is necessary to add calcium as well as the other reagents.

Studies of the removal of strontium upon a calcium phosphate floe have shown that the degree of removal depends upon the pH and upon the ratio of sodium phosphate to calcium hydroxide. As these two parameters are increased the efficiency increases, steeply at first but later reaching a steady value; optimum conditions are given by a pH of 11 .0 -11 .5 and a ratio of 2. 2 by weight of sodium phosphate to calcium, hydroxide, corresponding to 46% excess sodium phosphate. The excess phosphate may be removed by adding ferric ions to give a mixed floe of basic calcium and ferric phosphates which possesses better settling properties than a calcium phosphate floe alone; a cheaper alternative is to use ferrous salts, but although rem ovals are good the quality of the floe is not always as good.

The attainment of a pH greater than 11 is difficult using lime alone, and the large amounts of calcium thus introduced would adversely effect strontium removals unless large amounts of phosphate were used; it is customary instead to use calcium chloride or calcium hydroxide and sodium hydroxide, trisodium phosphate and ferric ion being added after adjusting the pH.

One laboratory has found that for their wastes optimum conditions involve a concentration of 50 ppm of calcium, 80 ppm of phosphate, and 40 ppm of Fe+3 at a pH of 11. 5. As shown in Table VII, most of the polyvalent elements are removed with up to 99% efficiency, and removals in excess of 90% may be achieved for mixed fission products. Little removal of the anionic activity is obtained but strontium removal is good. Removal of ruthenium is variable, depending upon the chemical state of the element. As in the case of hydroxide floes, removal of alpha emitters, most of which are heavy metals, is very efficient, figures of greater than 99% being obtained.

4.1.4. Ferrocyanide precipitation

None of the methods outlined above are very effective for re moving caesium from waste solutions, the maximum removals reported being in the range of up to 30%. Investigations have shown that precipitation of certain metal ferrocyanides is quite effective for scavenging caesium and other isotopes under a variety of conditions. These techniques involve the reaction of a metal salt, com-

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TABLE VII

REMOVAL OF ACTIVITY ON PHOSPHATE FLOCS

Nuclide ConditionsRemoval

(%)Remarks

89 Sr Calcium phosphate floe, excess Na3P04

97.8

91 y Calcium phosphate floe, excess KH.PO.Z 4

99.9

144 ce Calcium phosphate floe, excess Na3P04

99.9 Jar tests with settling

124 Sb Calcium phosphate floe, excess Na„PO,3 4

67.4

185W (anionic) Calcium phosphate floe, excess KH PC)2 4

10.7

137Cs Calcium phosphate- ferric Cs 0V

90Sr %2 Mixed

phosphate floe, pH 11.5; 50 ppm C a2+, 40 ppm

Sr- 92.6 Ce- 95.6 Flow system in

i44Ce a-a solution 0106Ru g

FeHI, 80 ppm PO^' Ru- 47.2* Zr- >99

sludge blanket precipitator

Overall removal 93- 94+95Zr 96-97+

* Ru removal is dependent upon the chemical state o f the element.+ After settling.+ Centrifuged.

monly the sulphate, with potassium ferrocyanide at the proper pH to form the precipitate.

Ferrous ferrocyanide will scavenge caesium, ruthenium and other nuclides at pH's in the range of 4 to 9. Beyond a pH of 9, ferric hydroxide begins to form while at low pH Prussian blue (ferric ferrocyanide) is produced. The addition of a reducing agent such as sodium bisulphate tends to speed the reaction and keep the iron in the ferrous form . Removals of 98% of the activity from waste liquids have been achieved using 400 to 600 ppm of ferrous ferrocyanide, as shown in Table VIII.

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TABLE VIII

RADIOCHEMICAL ANALYSES OF WASTE TREATED WITH FERROUS FERROCYANIDE AS A FUNCTION OF TIME OF STANDING IN CONTACT WITH THE PRECIPITATE AT DIFFERENT pH VALUES

Original untreated waste

pH 8.3 4.0 5.8 8.9 8.9

Time the precipitate stood in contact with solution (days) 8 15 5 8

Ruthenium (counts/min ml) 6075 356 1300 823 327

Antimony (counts/min ml) 1620 105 310 548 404

Zirconium (counts/min ml) 1070 55 150 61 38

Rare earths (counts/min ml) 10860 35 320 810 151

Tellurium (counts/min ml) 1620 31 65 84 35

Caesium (counts/min ml) 1080 3 8 0 1

Per cent solids 13.4

Supernatant after treatment

Copper, iron, cobalt and particularly nickel ferrocyanide have been found very effective for removing caesium from waste solutions containing high concentrations of sodium ions, DF's of 103 having been obtained in the presence of sodium concentrations as high as 5 N. These metal ferrocyanides are also capable of yielding caesium DF's of several hundred to a thousand when precipitated in wastes at acidities ranging from 2. 5 M to pH 6. Above pH 6, ferric ferrocyanide loses its effectiveness for caesium, as does the copper form at a pH above 8. Cobalt and nickel ferrocyanides are effective up to pH 10.

Studies on the effect of concentration of nickel sulphate and potassium ferrocyanide have shown that equal concentrations of these reagents in the range of 0. 002 to 0. 01 M give essentially the same caesium decontamination. The order of addition of these reagents is unimportant when treating acid wastes, but in alkaline solutions the ferrocyanide should be present when nickel sulphate is added to prevent the formation of nickel hydroxide.

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Recent work has shown that by combining phosphate treatment with a copper ferrocyanide precipitation, improved decontamination of radiocaesium is possible without seriously reducing the removal of other radioactive isotopes. The procedure is as follows: The pH is raised to 10. 0, using spdium hydroxide solution, after adding the following chemical doses:; 15 mg/1 Cu‘2+, 40 mg/1 Fe (CN6)4-, 16 mg/1 Fe2+, 80 m g/’l PO^' , and 50 mg/1 Ca2+. Calcium is only added if less than 50 m g/l is present in the raw effluent and the PC>43- dose must be sufficient to ensure no Ca2+ residual in the treated effluent. Under these conditions radiocaesium removal in the range of 98% is attained.

4.1.5. Strontium phosphate precipitation

Addition of stable strontium nitrate and subsequent precipitation of strontium phosphate has proved to be an attractive method of removing radiostrontium. A comparison of the effectiveness of strontium and calcium phosphate for this purpose is shown in Table IX. In each case precipitation was carried out at pH 9. 5 in the presence of 0. 005 M potassium ferrocyanide and nickel sulphate, which were added to' remove caesium. The calcium of strontium nitrates were added after neutralization. At equal concentrations, strontium phosphate is a more efficient precipitant than calcium phosphate. . In fact, better decontamination of radiostrontium is obtained with 0. 004 M strontium nitrate than with the highest concentration of calciurr} nitrate as shown in Table IX. The lower concentration of strontium nitrate required is advantageous for keeping sludge volumes small. Results show thaf strontium nitrate is most effective when added before neutraliza.tion. Decontamination factors of around 10 to 100 have been obtained.by adding strontium nitrate to a concentration of 0. 004 M to the acidic waste and then neutralizing to pH 9. 7. '

The .precipitation of radiostrontium is enhanced by high pH. Strontium phosphate, although less, sensitive to pH control than calcium phosphate, nevertheless rapidly loses its efficiency for radiostrontium removal below pH 8.5. '.

4.1.6. .Massive chemical treatment.

. Where.a variety of radionuclides are, or may be, in the waste, and relatively small volumes of liquid are to be treated, a '-univer-

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TABLE IX

COMPARISON OF CALCIUM AND STRONTIUM PHOSPHATE PRECIPITATION OF- RADIOSTRONTIUM

Procedure: Acidic waste made 0.005 M in potassium ferrocyanide.Solution adjusted to pH 9.5 and made 0.005 M in nickel sulphate.Calcium or strontium nitrate added to indicated concentration.Slurry stirred one hour at 25°C.

Concentration of Ca(N03)2 or Sr(N03)2

M

Final radiostrontium concentration in supernatant liquid, pCi/ml*

Ca(N03)2 added Sr(N03)2 added

0.004 1.18 0.32

0.008 0.71 0.10

0.015 0.86 0.17

0.030 0.73 0.12

0.035' 0.60 0.16

* Original radiostrontium concentration = ~ 2 jiCi/m l.

sal" precipitation process may be used which will simultaneously remove a large number of nuclides of various groups.

Recent studies have resulted in the development of dosage formulas which will, in a single precipitation, remove strontium, ruthenium, caesium, iodine, mercury, cobalt, iron, cerium, lanthanum, chromium, phosphorus and rare earths from liquid wastes.

Two mixtures of precipitants, shown in Table X, have been found effective; the formulas differ in that procedure number 2 utilizes Bi+3 instead of the Fe+3 used in procedure number 1. Procedure number 2 is slightly more expensive on account of the bismuth salt, but the precipitate is deposited more rapidly and forms a smaller volume of sludge which is important in practice.

Both formulas are equally effective with regard to decontamination, the DF's ranging between 20 and 500 depending on the concentration of chemical and radiochemical constituents in the waste.

Experience has shown that the quantities given in Table X are suitable for low-salt-content liquids. Where wastes are encountered which have high salt and organic or mineral acid contents, or which

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TABLE X

TWO PRECIPITANT MIXTURES

Precipitant formula 1 Precipitant formula 2

Add to 1 m3 o f liquid waste:1 litre carrier solution containing 0.5% each of Sr, Ru, Ce, and Zr chloride

(g) Antifoam agent (g) Antifoam agent

35 w 10 KI

150 K4Fe(CN)6 -3H20 50 NaHS03 • 7H20

250 NiS04 • 6H20 30 AgN03

50 NaHSO -1 H O5 2150 K4Fe(CN)6.3 H 20

10 KI 250 NiS04 • 6H,0

30 AgNOj 250 Bi(N03 ) 3

250 FeCl3 50 Ca(OH)2 tech.

250 Ca(OH)2 tech.

NaOH to pH 10.5

NaOH to pH 10.5

contain detergents, considerably larger quantities of precipitation chemicals must be used.

4.1. 7. Treatment of sludges

All the processes discussed above produce sludges as a result of the reactions taking place during chemical treatment. Slud.ge volumes vary according to the process used, phosphate and lim e- soda methods producing the largest quantities, particularly where a hard water is being treated. The volume of the wet sludge will also depend on the dosage of chemicals used, and the nature of the precipitated solids.

The water content of these sludges is usually high, solids contents of less than 5% resulting from normal settling techniques. Further treatment of the sludge depends upon the method of disposal employed. If only short or medium-lived nuclides are involved, the sludge may be stored in tanks to permit decay. Extended storage

4 9


will also result in some thickening of the sludge and consequent volume reduction. In all cases it is desirable to reduce sludge volumes to a minimum and various de-watering processes may be used, such as filtration, centrifugation, or a freeze-thaw step followed by filtration. By these means the solids content of the sludge can be increased from an initial 1 to 3% to a final content in the range of 30 to 50%. The freezing and thawing.treatment gives rise to a granular material which settles and filters well. In this process the sludge must be frozen slowly and completely and, while the rate of thawing is unimportant, filtration should take place as soon as possible after thawing.

Vermiculate or other absorbent materials should be added to dry out de-watered sludge before seating it in metal drums for storage or disposal. If dilute sludges are to be packaged for disposal without further treatment, they can be solidified by mixing with cement or a blend of cement and vermiculite, again using steel drums as the final disposal container.

4.2 . ION EXCHANGE USING ORGANIC RESINS ,

4.2.1. Treatment by ion exchange

One of the simplest methods of decontaminating a waste solution is to mix it in batches with an ion-exchange resin; the contaminants are concentrated on the exchanger. Removal of over 99% of most long-lived fission products have been achieved in this manner by using 2700 ppm of a mixed anion and cation exchange resin. However, more efficient use of the exchanger may be achieved by passing the waste down a column packed with the resin. Since most of the radioactive species, are cationic, decontamination of the wastes resembles .water softening by a cation exchanger on the hydrogen- or sodiumrcycle. On passing a waste of constant composition down a column of cation exchanger the order in which the inactive cationic species appear, in the effluent frona the column, is determined, by the relative affinity of the exchanger for all the cationic sp.ecies, .including that originally, present on the resin. On parsing.hard water through a. column of resin in a sodium form, calcium and magnesium are removed from solution and the water is thus softened. When the resin is saturated with the total hardness cations, these appear in the effluent causing the breakthrough of hardness.

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The use of ion-exchange resins may be of considerable value to industrial laboratories and hospitals for the removal of radioactivity from aqueous wastes having a total solids content of less than 2500 ppm, preferably below 1000 ppm. The principles governing the removal of radioactive ions by ion exchange are not different from those of similar chemical elements. However, in a waste, particularly one composited from many laboratory operations, the radioactive elements are in tracer quantities with respect to inactive ions. -

The wastes which are amenable to treatment by ion exchange are of two varieties - those from general laboratory operations and those from laboratories where only one tracer is used. In the first group the wastes may contain any or all types of radioactivity together with acids, bases and salts and the composition both chemically and radiochemically will be unknown. To analyse for all constituents would be not only difficult but too time-consuming to be economically feasible. Under these circumstances an analysis is made for pH, approximate total solids, and for total alpha and beta-gamma activity. An approximate total solids content may be obtained quickly by simply weighing a planchet before adding a known volume of the liquid for counting and then reweighing the planchet after the sample has been evaporated. The difference in weight in milligrams divided by the volume of the sample in' millilitres gives the milligrams per millilitre. This result multiplied by 1000 gives the parts per million. This result may be an error as much as 10-20%, but it is sufficient for an estimate'of the amount of total solids present. If the total solids content is less than 2500 ppm, ion exchange may be used. However, for most efficient use total solids should be less than 1000 ppm and for general laboratory wastes this is usually the case-.

Although many cation exchange resins are available on the market, those of the polystyrene divinylbenzene sulphonic acid type (Dowex 50, Nalcite HCR) have been found most satisfactory for good removal of radioactivity.

The flow-rate recommended by the manufacturers for high- capacity cation resins of the divinylbenzene sulphonic acid type is 270 litre/min m3 of resin. However, it has been shown that for the removal of gross fission products from radioactive waste solutions there was very little difference in the removal of radioactivity when using a flow-rate ranging from 270 to, 1350 1/min m3.

For the best over-all removal of gross fission products pH 2. 5 is recommended. A comparison of the removals obtained from total

5 1


beta-gamma radioactivity at various pH's is shown in Table XI. These data show a high through-put capacity and a greater removal at pH 2.5 for the ion-exchange column alone. Although a better over-all removal of activity occurred at pH 9. 5, half this removal was obtained by filtration before hardness breakthrough and over twice as much by filtration after hardness breakthrough.

The reasons for these results are apparent when one considers the chemistry of the fission products. With the exception of z irconium and niobium all the fission products should be ionized at pH2.5. The poor results particularly at pH 5. 5 may be due to the formation of radiocolloids of some of the other fission products. At pH 9. 5-there will be formation of the hydroxides of the rare earths and of radiostrontium.

The regeneration of resins presents some difficulty because radioactive contaminants, particularly those present as colloids, are not always removed as easily as are the inactive species. Furthermore, another liquid waste is produced as a result of regeneration which must be treated and disposed. Accordingly, disposal of the resin following exhaustion is recommended for sm all-scale application.

The results experienced in the treatment of laboratory wastes over a period of several months by ion exchange on a divinylbenzene sulphonic acid type resin is given in Table XII. The wastes were filtered before passage through the ion-exchange column. These wastes may have contained various ions; complexing agents such as oxalic and citric acids, and minute amounts of organic solvents, as well as some soap and detergents.

4. 2. 2. Cation exchanger for processing general laboratory wastes

A. SizeThe size of a cation exchanger will depend on the total hardness

of the waste, the species and level of radioactivity to be removed, the permissible level for discharge, and the volume.

(a) For illustration, suppose the total hardness of the waste averaged 170 ppm as CaC03 (the maximum was 1700 ppm). The activity to be removed was mixed fission products averaging 1.6X 10"4 juCi/ml (maximum 4X 10-4 /uCi/ml), and the tolerance level for discharge was 4X 10“5 (LiCi/ml. The waste was to be diluted with non-radioactive wastes before final discharge, and the volume was 4000 litres/d .

5 2


TABLE XI

EFFECT OF pH AND FILTRATION ON REMOVAL OF RADIOACTIVITY BY CATION EXCHANGE

FeedpH

Filterelem ent

A Volum e throughput

Bed volumes

Average 0- y a ctiv ity , (cp m /m l)

Integrated decontam ination factors

Feed Filtrate Finaleffluent

Filter Ionexchange

O ver-a ll

1 .8 none 800 1260 _ 137 _ 9 .2 Ha . . .3400 978 . . . 197 . . . 5 .0 . . .

2 .5 none 850 1180 85 . . . 14.1 Ha . . .2020 942 . . . 179 . . . 5 .2 . . .

2 .5 yes 815 1180 . . . 83 . . . . . . 1 4 .2 Ha1960 950 . . . 169 . . . 5 .6

4 .0 none 655 1280 . . . 122 . . . 1 0 .5 . . .2900 1088 . . . 291 . . . 3 .7 . . .

5 .5 none 680 1200 . . . 426 . . . 2 .8 . . .

8 .02050 1160 . . . 514 . . . 5 .0 . . .

8 .0 none 815 1300 . . . 262 . . . 5 .0 . . .

8 .0 yes 815 1390 . . . 233 . . . . . . 6.0 Ha1120 1170 . . . 408 . . . . . . 2 .91320 830 493 256 1 .7 1 .9 3 .3

9 .5 yes 790 920 308 49 3 .0 6 .3 18.6 Ha175 4990 569 274 8.7 2 .1 18.2670 880 289 268 3.0 1 .1 3.3

Feed: Tap water (85 ppm total hardness as CaCC^) adjusted to pH shown and 1 to 2 - yr-old m ixed-fission products added.Colum n: 1 .6 cm x 60 cm Pyrex glass tube, containing 50 m l N alcite HCR-Na form.Flow-rate: 270 1/m in m sa H = Hardness breakthrough.


Cn►fc.TABLE XII

ANALYSIS OF RETENTION TANK WASTES AND CATION EXCHANGE EFFLUENT

Filtered Feed-Analysis

Batch volum e ( litres) PH

Hardness as CaCOs (ppm )

Totalsolids

(ppm )b

8 activity (/jC i/m l)

Effluent 6 activity (fiC i/m l)

Removal(%)

DF

3500 2 .3 0 2420 ' 2 .4 X 1 0 "4 5 .9 x 1 0 '6 98 41

3800 , 2 .5 22 1660 1 .6 x 10"4 4 .0 x 1 0 '5 75 4 .0. 5000 2 .7 26 ; 2120 9 .9 x 10’ 5 1 .5 x 1 0 '5 85 6 .7

3800 3 .2 111 1100 1 .4 x 1 0 '4 6 .3 x 10"5 54 2 .23000 4 .4 a 21 880 1 1 .6 x 10"4 8 .8 x 10*6 94 17.85000 3. l a 102 1000 8 .2 -* 1 0 '5 2 .1 x 1 0 '5 74 3 .55800 7 .4 a 102 , 860 2 .4 x l O '4 3 .6 x TO"5 85 6 .64550 • 2 .3 144 1000 . - 2 .1 x 10" 4 7 .9 x 1 0 '5 62 2 .64250 1 .9 111 3680 2 .5 x 1 0 -4 2 .1 x lO '4 17 1 .23500 2 .1 150 1610 1 .2 x 1 0 -4 3 .8 x 1 0 -5 69 3 .34500 2 .3 141 1070 9 .8 x 10" 5 7 .8 x lO-5 21 1 .35400 2 .8 131 - 1125 9 .3 x 1 0 "5 4 .3 x 10" 5 . 54 2 .24100 1 .9 * 150c . 2090 . 1 .1 x 10‘ 4 2 .6 x 10‘ 5 77 4 .34100 1 .9 s 150c ; 1860 9 .4 x 1 0 "5 2 .1 x 10" 5 78 4 .66500 3 .2 a 148 ' 1020 2 .4 x 1 0 -4 4 .3 x 10“ 5 82 5 .44350

71150

3 .6 195 1240 3 .6 x 10" 4 4 .8 x 1 0 '5 87 7 .7

Average 3 .0 106 1550 1 .7 x 1 0 '4 4 .8 x 10" 5 72 ' 3 .6

Filter: Alsop, 30 cm diam . paper No. 40.Resin Colum n: 30 cm diam . containing 0.1 m 3 N alcite HCR-Na form .Flow rate = 270 1 /m in m 3apH adjusted to ~ 2 . 5 before' filtration.bpianchet method ± lO^o.E stim ated hardness.


The average decontamination factor for the above will be:

1.6X 10-4 juCi/ml 4X10-5 MCi/ml

The maximum total hardness as CaCC>3 is 1. 7 g /litre , and if the capacity of the cation resin used is 68 700 g of CaCC>3 per m3, this will be equivalent to:

68 700 g /m 3 / 3- —y g/lit re" ~ litres/m .

Since the volume to be treated per day was- 4000 litres:

4000 litres - . 340 000 litres/m 3 • m •

However, on the basis that a decontamination factor of 4 is sufficient and the average hardness is 170 ppm as CaCC>3, a cation exchanger could be used for about the equivalent of 4X 107 litres /m 3 before the resin must be replaced.

If a 0. l~m3 bed were used, 4X 106 litres could be processed or the exchanger could be used for 4X 1 0 6 /4X 103 = 1000 d or 3 yr before needing to be discarded.

There are two things to be particularly noted in the above calculations: capacity for hardness breakthrough was based on maximum hardness in waste whereas capacity for activity removal was based on the average figure. In the above example a 30-litre exchanger could be installed as it would be sufficient for a year's use- before having to be discarded.

(b) Suppose, however, that all the above data for the wastes were the same but it was required to remove strontium-90 to below its maximum permissible level, namely 8 X 10_7juCi/ml.

The maximum activity will not have to be taken into consideration (in the above example, 4 X 10-4 /uCi/ml), but rather the strontium-90 content of the feed must be considered. Let us also assume, based on previous analyses of these wastes for strontium-90, that it is safe to consider that 10% of this will be strontium-90.

5 5


Therefore, the level of activity to be considered as feed is 4X 10"5 juCi/ml and a decontamination factor of only 50 would be required.

4X 10~5 juCi/ml8 X 10"7 /uCi/ml

From Table XIII it is seen that at pH 2. 5 it is possible to remove strontium-90 from tap water containing 85 ppm hardness as CaC0 3

with a decontamination factor of 1 X 105 if not more than 615 bed volumes of water are treated. But in the example given above the maximum hardness was 1700 ppm or 20 times greater than 85 ppm. However, a decontamination factor of only 50 will be required. Referring to the same Table, it is noted that about 1400 bed volumes of 85-ppm hardness water may be processed before the DF is reduced to 50. Therefore, if the water in the example is 20 times harder than the tap water in the data of Table XIII it might only be possible to process 70 bed volumes of waste before reaching the maximum permissible level for strontium-90. If the volume to be treated per day was 4000 litres it would require about 60 litres of resin. Therefore, a cation exchanger containing 60 litres of resin will be required for each day's run and, under these conditions, ion exchange would be practical only if the resin were regenerated.

B. Equipment and materials of construction

Since the wastes being processed will be on the acid side, pH2.5, a plastic or rubber-lined vessel and either stainless-steel pipe and valves, plastic lined steel, or plastic pipe will be required.

The vessel may be any standard ion-exchange unit. A glass, saran, or stainless-steel screen, instead of gravel and sand, is preferable for holding up the resin particles, since the gravel and sand become radioactive and must be disposed of as solid waste.

C. Process design

In the example given above, where regeneration is not contemplated, the only equipment required is a conductivity meter, a flow meter, ion-exchange unit and holding tank. The latter is used as it is necessary to get a sample of the composite effluent to determine the activity before discharge. A diagram of such a system is shown in Figure 18.

5 6


TABLE X m

R E M O V A L OF S T R O N T IU M -90 B Y C A T IO N EXCH A NG E

V o lu m e throughput

90Sr from M ix ed fission products In effluents

( c p m /m l )D econ ta m in a tion Factor**

90Sr fr o m MSr tracer In effluents ( c p m /m l)

D econ ta m in a tionfactor*5

(b e d v o lu m es) pH l . J pH 2 .5 pH 5 .5 pH 1 .8 pH 2 .5 pH 5 .5 pH 2 .5 pH 7 .0 pH 2 .5 pH 7 .0

0 16 15 16 ----- ----- --- 1 .4 X 1 0 5 1 .3 X 105 --- ---40 0 .2 7 0 .0 4 0 .0 5 59 375 320 0 .8 7 4 . 3a 1 .6 x 1 0 6 3 . 0 X 1 0 4

200 0 .0 5 0 .0 0 .0 5 320 co 32 0 0 .0 7 6 .1 2 .0 X 106 2 .1 X 10*400 0 .0 8 0 .0 0 .1 9 200 <o 84 0 .2 4 0 .5 3 5 .8 X 106 2 .5 X 1 0 s615 0 .2 7 0 .3 0 0 .2 2 59 50 73 1 .3 a 2 .4 1 .1 x 105 5 .4 X 1 0 *820 0 .4 3 0 .4 5 0 .0 7 37 33 229 20 42 7 .0 X 10* 3 .1 X 10s

1000 0 .1 1 0 .9 7 0 .3 1 145 15 52 145 49 966 2 .7 X 10*1320 1 .8 a 0 .5 5 0 .5 6 8 .9 27 29 665 300 210 4331630 1 .3 2 . 0 a 1 .4 a 1 2 .3 7 .5 1 1 .4 3 .9 X 10s 3 .3 x 10S 36 391920 3 .5 1 .4 1 .3 4 .6 1 0 .7 1 2 .3 1 .1 x 104 8 .6 x 10* 1 2 .7 15

F eed : T ap w atef con ta in in g 85 ppm tota l hardness as C a C O j. pH and a c t iv ity a d ded as show n.C o lu m n : 1 . 6 -c r o .-d ia m . Pyrex glass con ta in in g 50 m l N a lc ite H C R -N a fo rm .F low rate: 270 1 /m in m 3a M axim u m p erm issib le le v e l e x c e e d e d at this p o in t, s in ce 0 .7 c p m /m l = 8 x 10” 7 j iC i /m l for instrum ent used.

b . . . . a c t iv ity in feed 'D econ ta m in a tion fa ctor -------- r r r — :------t t •a c t iv ity in e f f .

Ui-n]


AIR VENT

BACKWASH AND SPENT REGENERANTS TO SEPARATE HOLDING TANK

HARDNESS-FREE WATER FOR RINSING AND BACK- WASHING

FILTERED FEED

FIG. 18. Cation exchanger - schem atic

1. Pump to supply active feed

2. Pump to supply nitric acid

3. Pump to backwash colum n4. Flow indicator for side stream

5. Flow m eter for nitric acid and rinse water

6. Flow m eter for active feed

7. Conductivity m eter

8. Conductivity cells

9. Sam pling point for effluent

10. Tank 150 cm high by 35 cm I. D. ,

containing 0.1 m 3 cation resin.

Steel shell with Koroseal lining

D. Operation

Any ion-exchange resin must be first conditioned by exhausting and regenerating it at least twice. This is usually done by simply passing hard water through the bed until it is' exhausted and then regenerating it with acid if the hydrogen form of the resin is desired.

The gross beta activity and approximate total solids of the feed are determined. If the latter is above 3000 ppm it is not practical to process by ion exchange. If the total solids are below 3000 ppm then the pH of the feed is determined and it is adjusted to pH 2. 5 with acid or alkali. If permissible discharge level is based on strontium-90, the amount of this radioactive element in the feed must be determined.

5 8


The feed is then filtered and pumped to the ion-exchange column at the rate of 270 to 1350 litres/min m3 . Two conductivity cells are installed, one in the effluent line and the other in a line-bleeding feed just before it has passed completely through the resin; this is in' order to get the hardness breakthrough point at the instant of initial breakthrough of calcium and magnesium. An effluent sampling point is required in order to check for activity from time to time.

4 .3. EVAPORATION

While evaporation is the most effective means decontaminating liquid radioactive wastes, the use of this process has in the past been confined to large treatment centres on installations because of the high capital, and operating expenses involved.

Recently, however, several packaged units have become commercially available which are intended for treatment of relatively small batches of low“level liquid wastes, and which therefore might be of interest to small installations which generate waste requiring DF's of from 104 to 106 before discharge. Two of these units are described below.

4.3.1. Low-temperature evaporator

The- first system, a flowsheet for which is shown in Fig. 19, concentrates radioactive liquid waste by using a low-tempera.ture high-vacuum evaporation process. When the desired concentration is achieved, residual wastes are then transferred to storage or to a solidification system for ultimate disposal.

On the average, the radioactivity of the input wastes has been reduced from approximately 10-2 to 10-4 pC i/cm 3 to the order of 10-8 to 10-9 juCi/cm3 in the distillate.

The principal features of the process are:(1) Low temperature operation(2) Efficient multiple system of separators(3) Automatic purity sensing system(4) Automatic self-cleaning during shutdown.During operation, the hot-waste batch tank (1500-litre capacity

on the 190-litres/h unit) is filled with raw waste from the facility waste-collection system. Once the raw waste is in the batch tank

5 9


w P U M PLEGEND

© T EM P E R A T U R E IN D IC A TO R

m E D U C TO R GLOBE V A LV E © F LOW R A TE IN D IC A TO R

s P U R IT Y S E N S 1 N 6 E L E M E N T B A TE V A L V E (N O R M A L L Y C LO SED ] © LIQ U ID LE V E L IN D IC A TO R

§ LIQ U ID LE V E L C O N T R O L L E R & FEEDER #C*3 GLOB E V A L V E (N O R M A L L Y C LO S E D ) D E N S IT Y IN D IC A TO R

T H R E E -W A Y SO LE N O ID VA LV E o LO C AL M O U N T E D IN S TR U M E N T © C O N D U C T IV IT Y C O N TR O LLE R

X G A TE V A LV E © P A N E L M O U N T E D IN S TR U M E N T © V A C U U M IN D IC A TO R

FIG. 19. Low-tem perature waste evaporation system

and the equipment started, the system operates unattended until the cycle is completed. The batch-tank pump circulates the radioactive liquid waste continuously through the evaporator section of the concentrator. The waste in the evaporator section is boiled at a low temperature and at a high vacuum until it reaches the desired concentration. The concentrator operates at approximately 38-44°C

6 0


boiling temperature and at a vacuum of approximately 690 mm of mercury. The rate of evaporation and distillate production is controlled by throttling the heating and cooling water valves.

A level controller on the evaporator maintains a constant level of radioactive liquid waste. A portion of this waste is extracted continuously by an eductor and returned to the batch tank, resulting in an uniformly concentrated batch of raw waste. When the desired solids concentration is achieved, the concentrated waste is transferred to sludge storage, a drumming station, settling tank or other means of disposal by the batch-tank pump. The amount of concentration varies with the type of waste being concentrated. This process has achieved a total solids content as high as 53% in the waste. The density of the material being concentrated is continuously monitored and indicated on the control panel. The distillate cycle begins by the vapour in the concentrator being condensed on the condenser tube bundle to form the distillate. This distillate is withdrawn from the concentrator by an eductor in the distillate circulating line. The distillate pump circulates distilled water from the distillate tank through the eductor in a closed loop in the distillate circulating line. This eductor also creates and helps maintain the vacuum in the concentrator.

Before vapours reach the condenser section, they pass through specially designed separators to remove droplets from the vapour, thus increasing the decontamination efficiency of the unit by preventing the carry-over of contaminated particulates.

If the purity of the distillate is not as high as desired, a conductivity cell (purity sensing element) in the line to the distillate tank operates a solenoid valve which returns the distillate to the waste cycle for reprocessing. As soon as the concentrator operation is again stabilized and the desired purity is achieved, the valve automatically re-directs the flow to the distillate tank. The distillate pump automatically discharges the distillate as it is generated. At the discretion of the operator, the distillate can be passed through a mixed-bed ion-exchange column which acts as a polishing unit to purify the water further, if desired. The final distillate is suitable for reactor-pool make-up or other purposes requiring highly purified water. Since the water from the concentrator has already been highly purified, the de-mineralizer resins will last through thousands of litres of distillate. The ion-exchange resins are visible, and can be readily removed and replaced.

6 1


The process has been designed to be safe in operation. Since steam is not used, a steam accident cannot occur. If leakage occurs, it will be into the evaporator since it is at a lower pressure than atmospheric. A power failure will automatically shut down the process, regardless of heat input conditions. Since hot water is used as the heating medium, loss of power will cause the vacuum to be lost and boiling in the evaporator will cease since it will now be at atmospheric pressure.

A unique feature of the process is the automatic self-cleaning action of the concentrator using a portion of the distillate from the distillate tank when the unit is shut down. The self-cleaning occurs when the distillate pump is de-energized, thus breaking the vacuum in the concentrator. When the vacuum is broken, a portion of the distillate tank water is forced back through the distillate line to the concentrator. This distilled water spills over into the concentrator, rinsing the separators in addition to re-submerging and cleaning the tube bundle. This water is then withdrawn to the batch tank by the batch-tank pump.

The system is available in various capacities, including 190 litres/h , 380 litres/h, 1150 litres/h and larger sizes to meet special requirements. It is shipped completely assembled on a skid, and factory tested. Complete instrumentation is provided with the unit. It is only necessary to connect utility services to make the unit operational. Dimensions of the 190-litres/h unit are 2. 85 m long, 1 .2m wide and 2 .7m high.

Power requirements for the 190-litres/h system are a maximum of 3 kVA to operate pumps and instrumentation. Approximately 2.8 litres/m in of air are required. Temperatures and flow of hot water can be varied to suit most optimum local conditions, ranging between 65-80°C and 40 to 120 litres/min for the 190-litres/h unit. Cold water is used as the condensing medium.

4.3.2. Wiped-film evaporator

The second system, a flowsheet for which is shown in Fig. 20, is based upon the use of a steam-heated wiped-film evaporator. This type of unit is particularly advantageous when treating wastes which have a tendency to foam.

The operation of the system at distillation rates averaging 150 litres/h yielded DF's of the order of 107 when treating waste having an initial activity level of ~ 1 0 -1 ^Ci/m l.

6 2


'/^adeaactcae JtifyciicL ‘TOatfe

‘p&ciiity

FEEDTANK-

FIG. 20. Wiped-film evaporator system

In operation/, raw waste is pumped to the feed tank at a rate controlled by the liquid level in the residue tank. In this way evaporation to dryness on flooding of the system is prevented. - '

Solution flows by gravity from the feed tank to the evaporator, where it runs down the inside of the steam-heated walls. Residue drains from the evaporator to the residue tank, where it is recycled through an'overflow pipe in the feed tank and mixes with fresh dilute feed.

Specific gravity of the concentrate is measured by bubbling air slowly into the overflow pipe, thereby measuring the hydrostatic head in this fixed liquid leg. When the specific gravity reaches a predetermined level, concentrate flow from the residue tank is diverted in whole or in part to a sludge receiving drum. This removal of concentrate or sludge is done without interrupting the evaporation process.

Water vapour produced in the evaporator passes through two internal separators, an external demister, and is condensed. Distillate from the condenser flows to one of two 1200-litre distillate re

6 3


ceiver tanks. When one is full, flow is automatically diverted to the other. The distillate is then pumped through one or both ion- exchange columns. De-mineralized water is collected in either of the two 1200-litre de-mineralizer tanks, where it is monitored and, if within specifications, discharged to the sewer.

Utility requirements include 180 kg/h or steam at 5 atm pressure and 150 litres/m inof cooling water. The facility is mounted ona steel base and has over-all dimensions of 4.2 m long by 3 m wide, with a height of 3.2 m.

4.4. CONCLUSIONS AND RECOMMENDATIONS

It should-be emphasized that the data reported in the previous sections, while serving as a general guide to the characteristics of a given process, cannot be considered as necessarily representative of what will be obtained in practice. Many factors such as local water quality, wast-e-water composition, radionuclide concentration and small quantities of detergents will have a strong influence on the decontamination factor actually achieved.

In many instances, careful attention to good segregation practice will eliminate the need for liquid-waste treatment entirely, since small volumes of more concentrated liquid wastes which do accumulate can best be handled by solidification without prior treatment.

Where wastes have a low total solids content, ion-exchange treatment is the most suitable method. It has the advantages of simple equipment requirements, high volume reduction factors, and the activity concentrated in a form which is easily handled for packaging and disposal. The main disadvantage is cost of the resins, but unless rather large volumes of waste are involved, the costs should not be excessive.

Where the solids content of the waste is too high to permit the use of ion-exchange methods, and where large decontamination factors are not required, chemical treatment may be useful. However, despite the low costs associated with the treatment process itself, volume reduction factors are often low, particularly where sludge treatment is not practical. Consequently, considerable costs and effort may have to be expended in the solidification and disposal of the sludges resulting from chemical treatment.

The choice of a particular process will be governed by the nature of the installation and operations producing the waste. For many

6 4


of the chemical treatment processes discussed above, a considerable amount of laboratory work is required to determine optimum dosages of chemicals. This may be warranted in cases where the waste is produced on a regular basis and has a fairly uniform chemical and radiochemical composition, but where the volumes are too low to justify the installation of a chemical treatment plant. On the other hand, if it is necessary to treat only an occasional batch of waste which is unpredictable in its characteristics, use of the massive chemical treatment process described in section 4 .1 .6 will eliminate the need for much of the laboratory work and will probably be more economic in the final analysis, despite the cost of chemicals involved.

Evaporation is the method of choice where high DF's are re quired, or where the waste is otherwise not suited for processing by alternate techniques. Capital costs compared to the other methods are high, but the units described in section 3 will treat waste for a few dollars per cubic metre including amortization if their processing capacity-is fully utilized.

5. SOLID-WASTE TREATMENT AND DISPOSAL BY INDIVIDUAL USERS OF RADIOISOTOPES

5.1. INCINERATION

5.1.1. Introduction

Incineration, as a unit operation, will substantially reduce combustible waste volumes and as such represents a useful waste treatment or handling process. Where radioactive contamination is involved, other factors must be considered. The radioactivity, per se, remains unaltered. Depending upon the physical and chemical characteristics of the isotopes involved, the radioactivity may be transferred from its original combustible carrier and pass up the stack as a contaminated gaseous or particulate effluent; may adhere to or "plate out" on interior surfaces of the incinerator or associated piping; it may be retained in the ash; or it may be distributed in various proportions in the stack effluent, on the incinerator surfaces and in the ash.

From the foregoing, it is apparent that where incineration is utilized as a waste-handling method, there are implicit consider

6 5


ations of treatment or handling of stack effluents and handling and disposal of radioactively-contaminated ash material. Where conventional incinerators are involved, there generally will be little or no provision for-treatment of stack effluents. In many cases, the stack itself may consist essentially of a roof outlet, and the diffusion and transport of effluent from the outlet will be dependent upon its relationship to the building and surrounding buildings. Also, in the case of many incinerators, the ash-handling facilities may not be very highly developed and, in fact, may constitute the major source of contamination to personnel. Ash-handling facilities and operations can be modified and improved. These factors, therefore, must be carefully considered in assessing the ability of specific installations to handle adequately different quantities of radioactive materials.

5.1.2. Applicability

Small domestic or institutional incinerators have been found to be very useful treating combustible low-level radioactive solids, particularly putrifiable wastes such as animal carcasses, cage cleanings and physiological and pathological remains. However, since no special air-cleaning or ash-handling facilities are usually associated with such units, activity content of the wastes must be restricted to levels which will not result in exposure of the general population to concentrations which exceed MPCair .

The use of incineration for the treatment of wastes containing relatively large quantities of radioisotopes requires specially designed or modified units utilizing elaborate air-cleaning and ash- handling systems; because of the cost and complexity involved, the consideration of such systems by individual users is not recommended. By segregating small volumes of highly contaminated material from larger volumes of slightly contaminated combustibles, activity levels in many, if not most, cases will fall within the range where use of institutional incinerators may be permitted.

5.1. 3. Simple approach to calculating safe limits for incineration

There are two limitations on the safe levels of activity which may be disposed of by means of institutional incinerators; concentration in flue gases and concentration in ashes.

66


5.1. 3.1. Concentration of activity in flue gases

Assuming a relatively uniform distribution of radionuclides in the waste material;

Let t = minutes of burning per dayA = rate of discharge of flue gas in cm3/min y =/uCi of radioactive material which may be disposed of per

day.Then the average concentration in unfiltered flue gases, assuming complete liberation

iuCi/cm3

Assuming a dilution of 100 between stack mouth and point of breathing (a conservative assumption), and using 1/10 of the 168-h ICRP occupational MPCa as the limiting factor, then

y / t A ^ X M P C a

y = 10 MPCatA (1)

For example, if the incinerator is operated 300 min/d, the flue gas discharge = 2X 107 cm3 /min, and it is desired to determine the quantity of 131I (168-h occupational MPCa = 3X 10_9/juCi/cm3) which may be disposed of, then

y ( 131I) = 10 X 3X 10‘ 9 X 300 X 2X 107 i Ci

= 180 juCi

Complete liberation in the flue gases need only be assumed for those isotopes known to be volatile, such as 3H, 14C, 35 S, 74 As, 82 Br, 131I and 203Hg. In calculating flue-gas concentrations for other nonvolatile isotopes it may be assumed that only 10% of the activity is released with the flue gases, and equation (1) becomes

y = 100 MPCa tA (la)

67


5.1. 3. 2. Concentration of activity in ash

When wastes contaminated with non-volatile radioisotopes are incinerated, most of the activity will remain with the ash, and may constitute an inhalation hazard during ash removal and subsequent handling.

The limits on quantities of activity which may be incinerated due to ash contamination may be computed as follows:

Ash residue = 10% by weight of waste incinerated.Let C = Maximum concentration of dust which may be suspended

in air = 3 X 10“7g/ cm 3 .W= Weight in grams of waste (contaminated and non

contaminated) charged to incinerator, y = fj,Ci of radioactive materials which may be charged to

incinerator in W grams of waste.then

y /0 . 1 W = specific activity of ash

Again using 1/10 of the 168-h ICRP occupational MPCa as the limiting factor,

y /0 . 1 W = 0. 1 MPCa fC

y = W MPCa/ 100C

y = W M PQ/3X lO"5 (2)

For example, if 100 kg of waste are incinerated before ash removal, the amounts of 32P (MPCa =2X 10"8/iCi/cm3 ) which could be contained in the waste before special ash handling precautions would be required is

vf3 2 p w lO-5--X 2 X _10-8‘ 3X10-5

= 67 /uCi

In case of volatile isotopes, it may be assumed that only 10% of the activity remains in the ash, and for these isotopes, equation (2) becomes:

y = WMPCa/3X 10-6 (2a)

6 8


Where a combination of isotopes is involved in known amounts, the limit for the combination should be derived by determining, for each isotope, the ratio between the quantity present in the com bination and the limit otherwise calculated for the specific isotope alone using the above equations. This ratio should be expressed as a fraction, and the sum of such fractions for all the isotopes in the combination may not exceed 1 (unity).

Where activity levels exceed those calculated using the above equations, special ash-handling procedures should be used. Respirators should be worn by those involved in ash handling, the ashes should be moistened to inhibit spread of dust, and the ashes should be packaged as low-level solid waste rather than disposed of via normal non-active refuse-disposal procedures. Figure 21 illustrates the removal and packaging of ashes from a small domestic incinerator.

FIG. 21. Rem oval and packaging o f ashes from a sm all dom estic incinerator used for incinerating radioactive

waste

5.2. COMPRESSION

For other than putrifiable wastes, compression or baling re presents a useful method of solid-waste volume reduction, particularly if incineration facilities are not already available or if ac

6 9


tivity levels in the waste frequently exceed those which may be safely incinerated in institutional or domestic type units.

Although it may appear that incineration provides a considerably larger volume reduction than compression, this apparent advantage may not in fact be realized when one considers that the majority of low-level solid wastes which are not combustible are compressible. In general, compression involves lower capital investment and operating costs than incineration, and is a considerably simpler operation. However, it would not be worthwhile to install compression equipment unless the facility generated at least 100 m3 of solid waste per year. These factors should be carefully evaluated in terms of the installation's particular waste characteristics when considering the acquisition of new solid-waste treatment facilities.

Many installations utilize conventional baling presses, suitably enclosed in a ventilated housing to reduce the likelihood of spread of contaminated dusts. Such units are not suitable for the compression of waste contaminated with alpha emitters or other highly toxic materials.

Recently a unit designed especially for compacting contaminated solid waste has become commercially available. The device, shown in Fig. 22, involves a hydraulic cylinder that compacts the solids in a standard 55-gal (208 litres) drum; with a lid in place when the drum is full, it then serves as a shipping, storage or burjal container. Solid waste may be loaded into the drum while it is in place within the compaction device. To prevent the spread of radioactive dust the compaction is accomplished within an enclosure that may, if desired, be fitted with an exhaust fan and air filters. The enclosure is a 1 i 2-m-diam. cylinder 1. 65 m high; over-all height is 2. 5 m. The compacting force is 1300 kg. The device comes equipped with an integral hydraulic power unit and associated controls.

5.3. SOLID WASTE BURIAL

5. 3. 1. Packaging

Solid wastes that are to be buried should be packaged in a manner such that the contents will not be exposed or released during the handling, transportation, temporary storage and burial operatipns. Most of the containers discussed in section 2.3.11 will

70


FIG. 22. D evice for com p actin g contam inated solids and rubbish

meet these criteria, with the exception of cardboard boxes, which are likely to lose their integrity when exposed to wet weather. Semiliquid wastes such as evaporator concentrates or chemical treatment sludges should always be solidified with cement or mixed with vermiculite or other absorbent materials and packaged in metal containers before transportation and burial. Plastic bags which have been sealed with tape are satisfactory containers for low-level solids where sharp objects such as broken glassware or metal wastes are not involved. Plastic bags may also be usefully employed as liners for the containers discussed in section 2 .3 .1 . as an additional barrier between the wastes and ground water.

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5. 3. 2. Site se lec tion fo r ground disposal

In selecting a site for ground disposal of radioactive wastes the objective must be to find a place from which wastes moving through the soil will not enter the public domain in amounts that could cause unacceptable radiation exposure of the population.

If the site is sufficiently far from human habitation this object will be attained. The underground movement of wastes is caused by the movement of ground water, therefore the absence of ground water is desirable, but this can seldom be attained. The entry of waste into ground water introduces a dilution factor as well as ensuring that the movement will probably be slow. Slow movement ensures time for radioactive decay. However, distance from human habitation entails high cost of transport, especially if wastes are sufficiently active to require shielding or special vehicles.

Ground waters move slowly when the head is small, i .e . movement in flat country is slower than in hilly country. Clay soils are less permeable than sands, but it must be remembered that soil horizons are never uniform and water will move more quickly in one stratum - e .g . a gravel or sharp sand layer - than in another, e .g . a silt or loam. This demands a rather detailed knowledge of the local soil stratification. The tendency of clay soils to become waterlogged and form surface swamps and streams must be remembered.

Consideration must be given to the nature of the bedrock. Limestone frequently contains channels through which water runs at rates comparable with those in a stream, and granite is frequently fis sured. The geological situation must therefore be assessed with these factors in mind.

Many radionuclides are strongly adsorbed into soils, particularly silts and clays. This imposes a very important delay upon waste movement. Even when the cation exchange capacity of soils is low - e .g . sands - significant adsorption will usually occur. Some radionuclides - e .g . isotopes of Sr, Ru, Cs, Co, S, move more rapidly than others through soils, but the rate of movement will depend on the properties of the local soil. For example, caesium moves more rapidly than strontium through calcareous soil, but the situation is reversed in acid sand. Thus a knowledge of local soil chemistry is helpful in site selections, but again it must be remembered that soil structure is not uniform.

An ideal disposal site should be dry to great depths or be drained by slow-moving water flowing through soil with a high cation ex-

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change capacity, and remaining below ground for a long time. On emerging into surface water, the ground water would be greatly diluted by flowing into a large river or tidal water which would not be used to any significant extent by man.

These site characteristics are usually unobtainable in practice, so the choice must be made of the best available compromise. The amount of radioactive material premitted to enter the soil will depend on the degree to which the requirements can be met. If drainage from a disposal area can be sampled - e .g . if it emerges at one point - this will simplify monitoring, and if it could be segregated and treated at this point the requirements might be somewhat less strict. A considerable relaxation of requirements would result from a decision that wastes having more than a set content of radioactive material per package would be placed in water-tight containers such as high-quality concrete-lined facilities.

It would appear that more use could be made of municipal dumps for the disposal of Small quantities of low-level packaged wastes than has been the case in the past. Such dumps are carefully sited to prevent contamination of surface and other waters with nonradioactive pollutants. Usually the site, when exhausted, is covered and used as a park, playground, or for agriculture, for which purpose only the topsoil is disturbed. Even if it is developed in other ways a period of several years will elapse.

It is common practice for local authorities to carry out some form of salvaging, sorting, incineration or other procedure on solid waste. This would be undesirable in the case of radioactive material, and such waste, in consequence, should be.taken directly to the point of burial and buried beneath at least four feet of other rubbish.

Although the amount of material sent to municipal dumps would have to be the subject of discussion with the local authorities concerned, it would not seem unreasonable to work to the following limits:

(a) The material should be encased in a suitable container. Normally a 20-gauge drum with a well-fitting lid would suffice.

(b) The estimated amount per container should be less than 100 iuCi of long-lived radionuclides and less than 1 mCi of nuclides with a half-life less than one year.

(c) The gamma dose-rate at any point on the surface of the completed container should be less than 20 mr/h.

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6. AIR-BORNE WASTE MANAGEMENT

6.1. FUME-HOOD DESIGN RECOMMENDATIONS

Ordinary chemical fume hoods are often used for handling small amounts of radioactive material. Many modifications have been made in their design, the major refinement being to ensure a constant air velocity at the opening, regardless of the position of the door or front. Figure 23 shows some ways of achieving this.

EXHAUST EXHAUST

MOTOR OPERATEO f DAMPER AND CONTROL 7 " ' PROPORTIONS

DAM PER-LINK OPERATEO

EXHAUST

•fHr

BYPASS-

~REMOVABLE FILTERS OR BLANKS

FIG. 23. Methods o f controlling face v e lo c ity in ra dioactive-m aterial hoods, (a) Controlled face velocities;

(b ) Proportional by-pass link-operated ; (c ) Proportional by-pass fa ce opening. Entry loss = 0.25 VP (velocity

pressure) plus dirty filter resistance. D uct v e lo c ity = 3500 ft /m in . Filters: ( i) Prefilter; ( i i) A fter-filter ;

(A m erican C on feren ce o f G overnm ental Industrial H ygienists, "Industrial Ventilation M anual".)

The basic need is to provide sufficient air velocity into the hood to prevent the outward escape of contaminants. This may result from abnormal heat loads within the hoods, cross drafts in rooms, and rapid movements (walking in front of hoods).

The following general rules should be followed in the design of hoods for radioactive materials:

(1) Operations in which radioactive materials are handled should be enclosed as much as possible to prevent contaminating large air volumes.

(2) High velocities and cross drafts should be avoided because they may increase contamination and dust loading considerably.

74


(3) The volume of air withdrawn from the hood must be greater than the volume of contaminated gases, fumes or dust created in the hood.

(4). If possible, the operations requiring large amounts of wet digestion and volatilized acid or solvent treatment should be confined to one group of hoods and the handling of dry material in others.

(5) When levels of radioactivity are such that filtration is required, radioactive aerosols should be removed by filters placed as close to the hood as practical to prevent the unnecessary contamination of equipment and duct work.

(6) Accessibility for decontamination of the hood and the duct system must be made as easy as possible and quite fre quently stainless steel is used for the metal parts of the hood for this reason.

(7) The fan should be located so that duct work within the building is under a negative pressure.

To provide uniform distribution of flow at the face of the hood, several methods have been used. A perforated wall located at the back panel provides better distribution than a panel with slot entry at the top and bottom. A filter installed in the hood panel before the plenum is satisfactory in most instances and it can be easily changed when it becomes loaded or dirty. It is desirable to check the velocity through the hood opening at regular intervals to see whether the filter has become clogged.

The laboratory will normally be ventilated through the fume hood and to maintain adequate air changes when the hoods are not in use a by-pass system is necessary. The hoods may be down-draft or up-draft with the proportional by-pass overhead or at floor level.

An alternative approach to having a hood with a constant air velocity is to design the system to give an average face velocity of about 200 ft/min (60 m/min) with the door or sash open in the working position. This is usually 1. 5 ft (45 cm) high.

As this velocity may lead to large air-heating requirements some allowance is normally made in the design for the fact that all the fume hoods will not be used simultaneously, but such diversity factors should be used with caution.

The average face velocity of 130 ft/min (40 m/min) is optimal for most fume cupboards. As absolute minimum the velocity should not fall below 100 ft/min (30 m/min). As it was already.mentioned, for some special purposes higher velocities may be necessary. How

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ever, above velocities of approximately 150 ft/min (45 m/min) with dry (i.e . powder) operations, there is a danger that considerable and unnecessary quantities of radioactive material will be exhausted from the fume hood spreading contamination through the exhaust duct system, clogging the filters, thereby necessitating frequent filter changes and also increasing the contamination and radiation hazards. Also, there is a danger with pyrophoric powders of creating a fire hazard. With other types of operations done in fume cupboards it has been found that, with linear velocities in excess of 200 ft/min (60 m/min), pieces of paper, e .g . tissues, will be sucked into the exhaust system thereby blocking filters and tending to cause a fire hazard.

6 .2 . GASEOUS AND AEROSOL WASTE CONTROL SYSTEMS SUITABLE FOR USE BY SMALL LABORATORIES

For- most uses of radioisotopes in the quantities usually associated with small laboratories, the purification of exhaust air streams is unnecessary, provided that the exhaust outlet is suitably placed. In cases where direct discharge would create an unacceptable hazard, systems such as those described below may be used to remove radioactive aerosols and gases from the effluent air stream. However, it must be emphasized that the design and installation of anything more than the simplest types of exhaust and air-cleaning equipment should be undertaken only by specialists in these fields.

6.2.1. High-efficiency filters

The extremely low permissible level of air activity in radioactive work has required a filter capable of giving high removal efficiencies, often better than 99. 95% for submicron particles. These filters must sometimes remain intact at high temperatures, or at high humidities during normal operation or during an accident.

Most of the materials used in high-efficiency filtration are of a fibrous nature, and the filters consist of an assembly in depth of fibres of suitable diameter. The packing density of the fibres may vary from a loosely packed pad to a compressed paper, but the dimensions of the fibres and of the pores between them are, in general, greater than those of the particles being filtered. The filter does not act simply as a sieve, but aerosols are arrested by direct inter

7 6


ception, inertial impaction, diffusion, electrostatic attraction and gravitational effects.

Many materials are available for forming fibrous media suitable for filtration: some materials, e .g . asbestos/wool mixture, are formed into pads which are made up in filter cases of flat or cylindrical construction. While such filters can have very high efficiencies (up to 99. 9995%) and satisfactory pressure drops (~0.4 in (10 mm) water gauge for face velocities up to 25 ft/min (7.5 m/min)) they are rather bulky and less used in modern installations.

More attractive are materials which can be formed into a lap that can be pleated around spacers such that a large surface can be presented to the air flow in a small volume.

Materials which can be used in this way include:

(a) Glass paper ~ made from very fine glass fibre (1 - 2 /nm diam.)

The fire-resistant nature of this filter has made it particularly attractive for radioactive installations. A wide range of sizes are available, from about 3 ft3/min to 1000 ft3/min. The dimension of a filter unit with a capacity of 1000 ft3/min need only be 2 ft X 2 ft X 1ft (60 cmX 60 cmX 30 cm). Larger installations contain banks of unit s .- A typical face velocity is 3 - 5 ft/min (1-1.5 m/min) with a pressure drop of about 1 in (25 mm) water gauge. The glass paper is pleated around spacers of corrugated aluminium and sealed with fireproof cement into a mild steel frame. When inserted into wall panels or canisters, the leading edge of the filter insert is sealed against a face in the filter assembly by means of a glass paper gasket. The whole assembly is designed to remain intact and not contribute to combustion on heating to 500°C.

(b) Paper - made from a mixture of asbestos and cellu lose

These filters are very similar in design and performance to the glass, paper type. The spacers are often of corrugated kraft paper and the frame may be wooden. They are hyroscopic, not fire re sistant and are gradually being displaced by the glass-fibre paper.

(c) Cotton-asbestos mixture

These filters were widely used for radioactive work a few years ago and provided very high efficiencies, approximately 99. 99% with

7 7


submicron dusts. They are not fire resistant and tend to weaken at high humidities.

6. 2. 2. Medium-efficiency prefilters

Where the dust loading is high it may be necessary to precede the high-efficiency filter by a coarser pre-filter to prolong its life. This pre-filter is often of glass or synthetic fibre with diameter of 20 - 30 /urn.

After a period of operation the dust loading on the filter and consequently the pressure drop increases. At this stage it becomes necessary to replace the filter. This involves special techniques to protect the operators, to prevent contaminati6n spreading outside the filter installation and to confine the activity within the filter element which is being removed. Many fibre and paper disposable filters are available but in the case of radioactive filtration plant a pleated pad of glass fibres several inches thick has been most commonly used. This coarse glass-fibre pad also serves as an excellent spark arrester giving protection to the main filters. Efficiencies > 90% are obtained for particles down to 0. 5 jum. The pressure drop at the beginning of its life is usually 0. 5 in. (12 mm) w .g. Replacement of pre-filters is often an economic matter as their cost is usually much less than that of the main filters. The useful life of a prefilter may vary from a few weeks to perhaps a year depending on the dust load in the air. As a general rule the pre-filter would be changed when the pressure drop increases by a factor 3 or 4.

6. 2. 3. Activated carbon adsorbents

These are efficient and cheap adsorbents which will remove practically any radioactive vapour or gas from air and other gases. They are especially suitable for trapping the vapours of both organic and inorganic labelled compounds (during synthesis) and for removing radioactive iodine and phosphorus and their compounds from air. Even a comparatively thin layer of activated carbon (20-30 cm) will remove 99. 9% of the iodine from air at a filtration rate of up to 1 m/min. The pressure drop of such a layer at the indicated filtration rate is not more than 20 mm water gauge. Usually m icro- porous carbon is used (to get more effective capillary condensation),with granule sizes of 6-14 mesh and good mechanical stability, since dust formation may lower the ignition temperature. Vegetable car

7 8


bons are the best, for instance those obtained from birch, from stones of amygdalaceous plants, from coconut shells and the like. The volume density of such carbons reaches 0.55 g /cm 3.

When a sorbent is used to remove relatively short-lived contaminants such as 1311 and 32P, it can be automatically regenerated owing to the radioactive decay of the adsorbed material; this is not true, however, in the case of labelled compounds, in which the number of. molecules containing the radioisotope is small compared with the number of non-radioactive molecules. .

Activated carbon is the best adsorbent for radioactive inert gases as well. Its disadvantage lies in the fact that it is comparatively easily inflammable. Although the ignition point of carbon in a flow of air normally lies about 200°C, it is safer to limit the temperature of the adsorber to 100-150° C when gases containing oxygen are to be filtered. This makes, carbon practically useless for decontaminating gases that contain strong oxidizing agents (e .g . appreciable concentrations of nitric oxides).

R E F E R E N C E S

Note:

Chapter 2: Waste C ollection Containers and Systems Main References: [3, 10, 15, ,18, 24, 27, 32]

Chapter 3: Direct Disposal o f Radioactive Wastes to SewersMain References: [7, 13, 14, 16, 21, 22],

Chapter 4: Liquid Waste Treatment Techniques Suitable for Users o f RadioisotopesMainReferenc.es: [1, 2, 5, 6, 8, 9. 11, 17, 19, 24, 25, 27, 28, 29, 30, 31, 32]

Chapter 5: Solid Waste Treatment and Disposal by Individual Users o f RadioisotopesMain References: [3, 4, 7, 10, 20, 22, 23, 24, 26, 27, 32]

Chapter 6: A ir-borne Waste Management Main References: [12].

[1] AMPHLETT, C .B ., SAMMON, D .C ., "Survey o f Treatments Considered for Low-Activity Wastes", in Atom ic Energy Waste, Its Nature, Uses and Disposal (Ed. E. Glueckauf), Interscience Publishers, New York, and Butterworth and C o ., London (1961).

[2] ANON., "Ray-Di-Pak Process for Radioactive Liquid Waste Concentration", AMF Atomics, York, Pennsylvania (1965).

[3] BURNS, R. H ., "Radioactive Waste Control at the United Kingdom Atom ic Energy Research Establishment, Harwell” , Disposal o f Radioactive Wastes, _1, IAEA, Vienna (1960).

[4] BURNS, R. H ., "T he Disposal o f Radioactive Solid Wastes" in Radioactive Wastes, Their Treatment and Disposal (Ed. T. C . C ollins) E. and F .N . Spon Lim ited, London (1960).

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[5] BURNS, R. H ., GLUECKAUF, E ., "Development of a Self-Contained Scheme for Low-Activity Wastes", Proc. 2nd UN Int. Conf. PUAE 18 (1958) 150.

[6] CHRISTENSON, C .W ., et a l . , "Removal o f Plutonium from Laboratory Wastes", Ind. Eng. Chem. 43 (1951) 1509-15.

[7] Code of Federal Regulations, Title 10, Part 2o, "Standards for Protection Against Radiation".US Government Printing Office, Washington, DC (1962).

[8] COWSER, K. E. et a l . , "L im e-Soda Treatment o f Low Level Wastes", Proc. 2nd UN Int.Conf. PUAE 18 (1958) 161.

[9] DEJONGHE, P. et a l . , "Treatment of Radioactive Effluents at the Mol Laboratories", Proc. 2nd UN Int. Conf. PUAE 18 (1958) 68.

[10] ENDERS, J .W ., "Radioactive Trash Disposal at Los Alamos", J. Amer. industr. Hygiene Assn., 21:2 (1960).

[11] GOFF, D .L ., BLOORE, E.W ., THIEME, A ., "A Semi-fixed Radioactive Liquid Waste Treatment Facility", NDL-TR-61, US Army Nuclear Defense Laboratory, Edgewood Arsenal, Maryland (in press).

[12] INTERNATIONAL ATOMIC ENERGY AGENCY, "Techniques for Controlling Air Pollution From the Operation o f Nuclear Facilities", IAEA, Vienna, (in press).

[13] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, "Report of Committee II on Permissible Dose for Internal Radiation", Publication 2, Pergamon Press (1959).

[14] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION. "Report of Committee V on the Handling and Disposal o f Radioactive Materials in Hospitals and M edical Research Establishments", Publication 5, Pergamon Press (1965).

[15] IOSEPH, A .B ., "Radioactive Waste Disposal Practices in the A tom ic Energy Industry, A Survey o f the Costs", Rept. NYO 7830 (1955).

[16] KENNEY, A. W ., "The Behaviour o f Radioisotopes in Sewage Treatment and the Disposal o f Radioactive Wastes to Sewers", J. Brit. nucl. Energy Conf. 2 (1957).

[17] LAUDERDALE, R. A. I r . , "Studies on the Removal o f Radioisotopes from Liquid Wastes by Coagulation", Rept. ORNL 932 (1951).

[18] MAWSON, C .A ., "Management of Radioactive Wastes", D. Van Nostrand C o ., In c ., Princeton, N.J. (1965).

[19] MEAD, F .C ., Jr., "Some Observations on the Decontamination o f Fission Product Wastes" in Sanitary Engineering Conference Proceedings, USAEC Report WASH 275 (1954).

[20] MORGAN, J. M ., Jr. et al., "Land Burial o f Solid Packaged Low Hazard Potential Radioactive Wastes in the United States", Proc. Conf. on Ground Disposal o f Radioactive Wastes, Chalk River, Canada, Sept. 1961, Rept. TID-7628.

[21] NATIONAL COMMITTEE ON RADIATION PROTECTION, "Recommendations for Waste Disposal of Phosphorus-32 and Iodine-131 for Medical Users", National Bureau of Standards Handbook 49. Govt. Printing Office, Wash. D.C. (1951).

[22] PCHELINTSEVA, G .M ., e d ., "Sanitary Regulations for Work with Radioactive Substances and Sources of Ionizing Radiation", State Publishing House of Literature in the Field of Atomic Science and Technology, Moscow (1960).

[23] PECKHAM, A.E. and BELTER, W .G ., "Considerations for Selection and Operation of Radioactive Waste Burial Sites", Proc. Conf. on Ground Disposal o f Radioactive Wastes, Chalk River, Canada, Sept. 1961, Rept. TID 7628.

[24] RODGER, W. A . , "Radioactive Waste Disposal", Rept. ANL-6233 (1960).

8 0


[25] SCHULTZ, W .W ., McKENZIE, T. R ., "The Removal of Cesium and Strontium From Radioactive Waste Solutions" in Sanitary Engineering Aspects of the Atomic Energy Industry, USAEC

Report TID-7517 (1956).[26] SILVERMAN, L. B. and DICKEY, R. K ., "Reduction of Combustible, Low-Level Contaminated

Wastes by Incineration", UCLA-368 (1956).[27] STRAUB, C .P ., "Low-Level Radioactive Wastes, Their Handling, Treafment and Disposal",

U.S. Government Printing Office, Washington, D .C . (1964).[28] STRAUB, C. P ., et al. "Method for Decontamination of Low Level Radioactive Liquid Wastes",

Proc. 1st UN Int. Conf. PUAE 9 (1956) 24.[29] SWOPE, H .G ., "Treatment of Radioactive Wastes" in Ion Exchange Technology, (Ed. Nachod,

F.C . and Schubert, J.) Academic Press, Inc., New York (1956).[30] SWOPE, H .G ., ANDERSON, E., "Cation Exchange Removal o f Radioactivity from Wastes",

Industr. Engng Chem. 47 (1955) 78-83.[31] TALSKY, J ., "Decontamination o f Radioactive Effluent by Chem ical Precipitation", Waste

Management Research Abstracts No. 1, IAEA, Vienna (1965).[32] UNITED STATES CONGRESS, Joint Committee on Atomic Energy, Special Subcommittee on

Radiation, Industrial Radioactive Waste Disposal, "Hearings"', Govt. Printing O ffice , Washington, D .C . (1959).

81


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