radiation protection PART 2

2.2 radiation protection 1

1.      Radiation Protection

2.      MAXIMUM PERMISSIBLE DOSES

3.      Radiation Exposure

INTRODUCTION

The history of radiation exposure control is  very briefly outlined  in   this   lesson,   and  the main maximum permissible doses   are  discussed.

2.2.1 THE HISTORY OF RADIATION EXPOSURE CONTROL

Man has always  been  subjected  to  radiation,   both   from outside and  inside  his  body. This   RADIATION BACKGROUND,   as we call it,   arises  both   from  cosmic  rays   from space  and  from naturally occurring radioactive materials.     For example,  we know that radioactive  carbon  and potassium atoms   are   found in the body itself   (carbon-14   and potassium-40). These  and other radioactive  substances   occurring  in  the  earth's   surface (uranium,  radium,   thorium)   all  contribute  to  the  radiation background.

The radiation dose we  receive  from  this  background depends on where v/e   live,   what we  eat,   the  structural materials of our homes   and  so  on,   but  a  figure  of  100   to 200 mrem per year  is   a  reasonable  average  in Canada.     These radiations provide  a   lower  limit of exposure which  is unavoidable,   and which  is   not  normally  considered  significant.

However,   following  the  introduction  of man-made  sources of radiation,   the harmful effects  of  an overdose were  soon discovered as,   for example,   in  the  experience  of the early X-ray experimenters.  It is  generally believed  that  the harmful effects  of nuclear radiations   are due  to  their ion­izing effects  on the   tissues   in  the body. Although  the biological consequences   of  this  will be  considered in  a later section,   it is   sufficient here  to  realize  that ionizing radiation does  result  in  damage  to  the body.

For large doses of radiation,  the damage may range  from a surface burn to some  sort of effect deep  in  the body. This kind of radiation damage is called SOMATIC damage.     Radiation may also affect the reproductive organs in such a way as to produce malformations in future generations; this is called GENETIC damage.

Now it should be pointed out that there are no biological effects that are uniquely caused by radiation - all we can say is that radiation increases the probability or rate at which these effects occur.  Present evidence seems to indicate that although the body may recover from the somatic effects of small doses of radiation, it is probable that damage to the genetic organs may not be repaired.

During the 1920's it was well known that the amount of radiation received by an individual had to be limited to prevent injury and various national organizations began to study the problem and issue recommendations for the control of radiation exposure.  In 1928 an international commission (then called The International X-ray and Radium Protection Commission) was formed to consider this problem and to make recommendations with regard to radiation protection.

In 1950 the Commission was reorganized.  The name was changed to "The International Commission on Radiological Protection" - universally known as the ICRP,  The Commission is composed of a chairman and not more than 12 members who are chosen on the basis of their recognized expertise in radiation protection and related fields, without regard to nationality.  The ICRP is widely recognized today as the chief authority on radiation protection.  Its policy with regard to its recommendations is stated by the Commission as follows:

"The policy adopted by the ICRP in preparing its recom­mendations is to deal with the basic principles of radiation protection, and to leave to the various national protection committees the right and the responsibility of introducing the detailed technical regulations, recommendations, or codes of practice best suited to the needs of the individual countries."

2.2.2 PERMISSIBLE DOSE

In order to be able to enjoy the tremendous advantages of the proper use of nuclear energy and other sources of radia­tion, we obviously cannot restrict ourselves to no radiation exposure whatsoever - particularly since we cannot avoid back­ground radiation.  While there is no conclusive evidence that exposures above background are completely harmless, evidence in the past 60 years does indicate that an individual limited to PERMISSIBLE DOSES of radiation should not suffer any ill effects from radiation,  The ICRP's definition of the Per­missible Dose for an individual is:

"The PERMISSIBLE DOSE for an individual is that dose, accumulated over a long period of time or resulting from a single exposure, which, in the light of present knowledge, carries a negligible probability of severe somatic or gen­etic injuries; furthermore_, it is such a dose that any ef­fects that ensue more frequently are limited to those of a minor nature that would not be considered unacceptable by the exposed individual and by competent medical authorities."

2.2.3 CRITICAL ORGANS AND TISSUES

All body tissues are sensitive to ionizing radiation, but some are more RADIOSENSITIVE than others.  Furthermore, certain organs have functions which are essential to the well-being of the entire body.  When one considers the radiosensitivity of organs with respect to their specific function, some tissues and organs assume a greater import­ance, and they are said to be CRITICAL ORGANS.

In the case of more or less uniform irradiation of the whole body, the blood-forming organs (red bone marrow) and the gonads are the critical organs.

In the case of irradiation more or less limited to parts of the body, the critical tissue or organ is that part of the body most likely to be permanently damaged either because of its inherent radiosensitivity, or because of a combination of radiosensitivity and localized high dose.

2.2.4 DOSE RATE

Dose rate is the time rate at which a dose is received, i.e. dose per unit time.

Dose rates are expressed as rads or rems (or mrad, mrem) per unit of time (year, week, hour, minute).

2.2.5 DOSE LIMITS

THE ICRP

Even before the 1920's it became well known that the radiation dose received by an individual had to be limited to prevent injury. Various organizations began to study the problem and issue recommendations for the control of radiation exposure. In 1928, an international commission ( then called the International X-Ray and Radium Protection Committee) was formed to make recommendations with regard to radiation protection.

This Committee was reorganized in 1950. The name was changed to the International Commission on Radiological Protection - universally abbreviated to the ICRP. The ICRP is widely recognized today as the chief authority in protection from the harmful effects of ionizing radiation and has responsibility for presenting recommendations on all aspects of this subject. These recommendations usually are adopted without significant change by most countries and are incorporated into their laws.

ICRP 26

The ICRP published an important document in 1977. It is ICRP Publication 26, known as ICRP 26, and it describes the ICRP system of dose limitation Given below are some of the concepts that are described in detail in ICRP 26.

2.2.6 THE OBJECTIVES OF RADIATION PROTECTION

The primary objective of radiation protection is to protect individuals, their off-spring and mankind as a whole, while still allowing necessary and beneficial activities involving radiation exposure.

The biological effects of radiation are classified as somatic and hereditary. ICRP 26 treats somatic effect as STOCHASTIC or NON-STOCHASTIC. Stochastic means " arising from chance; involving probability ". It is worth quoting from ICRP 26:

Stochastic ' effects are those for which the probability of an effect occurring, rather than its severity, is regarded as a function of dose, without threshold.

'Non - stochastic ' effects are those for which the severity of the effect varies with the dose, and for which a threshold may therefore occur.

For example, cancer is a somatic effect that is stochastic. In other words, the probability of contracting cancer increases with the dose, but once you get it, the severity of the disease is the same no matter how big the dose was that caused it. We assume that the relationship is linear, in other words, twice the dose means twice the chance of getting cancer. Hereditary effects are also stochastic effects. No threshold is assumed for stochastic effects.

In contrast to this, cataract of the lens of the eye is a non-stochastic effect with a threshold value of 7.5 Sv (750 Rein). As long as the dose is below 7.5 Sv (750 Rein), radiation induced cataracts cannot form.,

To illustrate, given below are a couple of practical examples of non-stochastic and stochastic effects. Sunburn has a threshold; above this threshold exposure, the degree of sunburn becomes more and more severe with increasing exposure to the sun, and below the threshold no harm is done. Compare this with winning a million rupees in a lottery; this is pure chance - the probability depends on the exposure ( the number of tickets you buy), but the magnitude of the effect doesn't change. You either win a million or you don't.

According to ICRP-26, (he aim of radiation protection should be to prevent detrimental non-stochastic effects and to limit the probability of stochastic effects to levels believed to be acceptable.

This is a most important objective. The non-stochastic effects can be prevented by setting annual dose limited low enough so that no threshold dose would ever be reached during a person's lifetime. The stochastic effects are limited by applying annual dose limits which, if not exceeded, would ensure that the level of risk from radiation work is no greater than the risk for other occupations which are recognized as having high standards of safety.

The main features of the ICRP recommendations >         known as  SYSTEM  OF ^t)OSE LIMITATION, are the following :

a)      No practice shall be adopted unless its introduction produces a positive net benefit.

This eliminates the " frivolous " use of radiation. For example, in the 1950's, many shoe stores would X-ray feet to see whether the new shoes fitted. This is no longer permitted.

b)      All exposure shall be kept as low as reasonably achievable, economic and social factors being taken into account. This statement ( known as the ALARA, (As Low As Reasonably Achievable ) principle ) implies that a value judgement must be made on the economic and social cost of say, one man-mSv. The important point is that all unnecessary radiation exposure should be avoided.

c)      The dose equivalent to individuals shall not exceed the limits recommended for appropriate circumstances by the Commission.

Like other national practices, Pakistan Atomic Energy Commission has also been adopting the recommendations of the ICRP.

the dose equivalent limits

hi any organ or tissue, the total dose due to occupational exposure consists of the dose contributed by external sources (i.e., those outside the body) during working hours and those contributed by internal sources (i.e., those inside the body) taken into the body during working hours. The limits apply to this dose received on the job - they do not apply to medical exposure or exposure to radiation background.

Furthermore, the limits presented here apply to Radiation Workers.

RADIATION WORKERS are people who may be routinely exposed to radiation as a result of their occupation.

At mentioned before, the dose equivalent limits are intended to prevent non-stochastic effects and to limit the occurrence of stochastic effects to an acceptable level.

2.2.7 SUMMARY OF CURRENT ANNUAL DOSE EQUIVALENT UMITS FOR RADIATION WORKERS

Organ or tissue	Dose quantity	Dose limits in mSv
Whole body	Effective dose equivalent	50 (5 rem)
Partial body	Committed effective dose equivalent or effective dose equivalent from partial body exposure	50 (5 rem)
Individual organs and tissues except the lens of the eye or the skin	Dose equivalent or committed dose equivalent.	500 (50 rem)
The lens of the eye	Dose equivalent or committed dose equivalent	150 (15 rem)
Skin averaged over any area of 100cm2	Dose equivalent or committed dose equivalent	500 (50 remO
Hands, face, amis, feet and ankles.	Dose equivalent	500 (50 rem)

* Averaging   over   an contamination; a smaller radiation beams area   of   100   cm²   applies   to   doses   from   radioactive area should be used for averaging in case exposure is to radiation beams.

3.1 RADIATION PROTECTION 2

1.      Radiation Protection

2.      Radiation Exposure

3.      PROTECTION AGAINST EXTERNAL EXPOSURE TIME, DECAY AND DISTANCE

INTRODUCTION

The fundamental aim of Radiation Protection is to re­duce exposures to the lowest practical level.

Protection against radiation is concerned with two separate situations:

EXTERNAL RADIATION, which arises from a source outside the body;

INTERNAL RADIATION, which arises from a source inside the body.

                                                                 FIG 3.1

The general biological effects of ionizing radiation from external and internal sources are not very different from one another.  However, as we progress it will become evident that the precautions taken against the one hazard are of little use in protecting against the other, and that the corresponding methods of estimating dose are different. Therefore external and internal radiation are treated separ­ately, and in this lesson we shall consider three methods of controlling external exposure.

3.2 TIME; DECAY; DISTANCE; SHIELDING

Radiation exposure can be decreased in any one of four ways:

1)      By decreasing the length of TIME spent near a source.

2)      By allowing the source to DECAY for some time before approaching it.

3)      By increasing the DISTANCE between yourself and the source.

4)      By absorbing the radiation in SHIELDING material placed between yourself and the source.

In this lesson we shall discuss the first three methods; shielding will be left for the next lesson.

3.2.1 TIME

Radiation exposure can be controlled very simply by limiting the length of time a person spends in the radiation area.  For example, if the radiation level in an area is 5 mR/h, then in 1 hour a worker receives an exposure of 5 mR, in 2 hours 10 mR, and in 6 hours 30 mR.  The dose he receives is therefore 5 mrem, 10 mrem and 30 mrem.

If you wish to limit the dose received by a person to a certain value, and you know the radiation dose rate, you may calculate the maximum length of time to be spent in the area by using the formula:

Time Limit =

Dose Rate

THE UNITS OF TIME MUST BE THE SAME FOR THE DOSE RATE AND THE TIME LIMIT.                                                                                                                                                                                                                                                                                                      

3.2.2 DECAY

A second way of reducing exposure when working near a radioactive source is to allow the source to decay to some extent before starting work.  The radiation dose rate will be decreased by a factor of 2 for every half-life delay. This is a good approach when work has to be done in a radia­tion area where the dose rate decreases very rapidly with time, say a half-life of the order of minutes or even hours. If there is no need for the work to be done immediately, then waiting a day or so would reduce the dose rate quite appreci­ably.

3.2.3DISTANCE

Increasing the distance between a person and the source results in a marked reduction in the radiation exposure. The diagram on the opposite page shows how this comes about.

S is a point gamma source which emits photons isotropically.  (This is an elegant way of saying "uniformly or equally in all directions¹¹.)  If we stand at A, we can count 7 gamma photons striking us every second, whereas if we move out to B, only 3 photons strike us per second.  Thus, as we move away from the source, the gamma ray intensity decreases. This is entirely due to the spreading out of the emitted photons.

FIG 3.2.3 (a)

The INVERSE SQUARE LAW describes the decrease in the radiation intensity with distance:

The intensity of the radiation varies inversely as the square of the distance from the source; that is, doubling the distance drops the exposure rate to 1/4, tripling the distance drops it to 1/9, and so on.

The decrease in intensity predicted by the inverse square law is shown in the diagram below.

INTENSITY DECREASES AS DISTANCE INCREASES

                                                               FIG 3.2.3 (b)

4.1 RADIATION   PROTECTION 3

1.      Radiation  Protection

3.  Radiation Exposure

4.  PROTECTION AGAINST EXTERNAL EXPOSURE SHIELDING

INTRODUCTION

In some cases the only practical way of reducing radia­tion exposures to an acceptable level is to install shield­ing between the source and yourself.  Radiation shielding is a very complex subject, and therefore only a few ele­mentary ideas are discussed.  Some aspects of reactor shielding are described.

4.2 ALPHA PARTICLE SHIELDING

You may recall from lesson R.P.T. 1.2.2 that alpha par­ticles have a relatively small penetrating power - even in air the most energetic alpha particles don't have a range of more than 10 cm; the dead layer of the skin will stop them completely.  Because of this, alpha sources outside the body do not present an external hazard and shielding against alpha particles is therefore quite unnecessary.

However, alpha particles with their QF of 10 are a very serious internal hazard and great care must be exercised to ensure that alpha sources are not taken into the body.

4.3 BETA PARTICLE SHIELDING

Normally, at a nuclear power station, radioactive mate­rials are enclosed in systems which completely shield the operators from the beta particles.  You will recall from lesson R.P.T. 1.2.2, page 7, that not very much shielding is required to do this.

However, when radioactive materials are released into the plant (e.g., Ar^4l) , or when systems are opened for main­tenance (e.g., removal and maintenance of a primary pump), the shielding is removed from around the beta sources.  Then an external hazard can exist, because beta particles have a considerable range in air depending on their energies. 

The most restrictive critical organs for beta radiation are the skin (30 rem/year) and the lens of the eye (15 rem per year).  Serious damage may be caused to both if the source is strong enough.  The hazard to the eye lens which is already shielded by the cornea may be further reduced by the routine wearing of safety glasses.

Since beta particles are easily absorbed, no one should receive an appreciable external dose from beta ratiation if proper techniques are applied. Maximum beta energies, E^x' vary widely but average about 1 MeV for mixtures of old fis­sion products. The percentage of incident 1 MeV beta radia­tion absorbed in some common materials is given in the table below.

ABSORPTION OF BETA RADIATION:  E-max = 1 MeV. Table 4.3

Type of Material	Thickness (inches)	Percent Absorption
Surgeons gloves		30
Cotton gloves		30
Neoprene gloves		50
Double neoprene gloves		70
Light coveralls		20
Plastic hood (PVC)	0.008	30
Safety glasses (lens)	0.14	90
Army respirator (lens)	0.14	90
Air	36	80
Plywood	0.25	100
Asbestos	0.125	90
Paper	0.125	90

Asbestos or heavy paper is  useful for draping over contaminated equipment to reduce beta dose rate.  About 1/8 inch of either material reduces the dose rate by a factor of 10 for 1 MeV beta particles.

4.4 GAMMA RAY SHIELDING

We already know that gamma rays will penetrate to great depths in materials and that no amount of shielding will stop all of the radiation.  The effectiveness of gamma ray shielding is frequently described in terms of the half-value layer (HVL), which is the thickness of absorber re­quired to rediice the gamma radiation to half its former intensity.  But this you know already from lesson R.P.T. 1.2.2.

The first HVL reduces the radiation to one-half.  The second HVL reduces the radiation by one-half again, i.e. ½ × ½ = ¼  of the original level.  The radiation levels after successive HVL's are:

Radiation after 1 HVL                                  =   1/2     of the original

Radiation after 2 HVL's   =   1/2 × 1/2 =   1/4     of the original

Radiation after 3 HVL's   =      (1/2)³   =   1/8     of the original

Radiation after 4 HVL's   =      (1/2)⁴   =   1/16   of the original

Radiation after 5 HVL's   =      (1/2)⁵   =   1/32   of the original

The above is strictly applicable only to a narrow beam of radiation .  A source emits radiation in all directions, and then the reduction in the intensity of the radiation with shielding is less than would be expected.  However, with small sources the error is not too great and the method is useful for estimating the shielding required for such sources .

The most effective gamma shields are those which have both a high density and a high atomic number, such as uranium, tungsten, gold, lead, etc.  Generally speaking, these heavy materials tend to be expensive - even lead isn't cheap - and therefore less costly, medium-weight materials such as iron and concrete are often used.

Concrete is a good structural material, but lead is not. Large lead shields require some sort of a supporting frame. On the other hand, lead shields will be thinner than shields made of less dense materials, and therefore lead is often used where space is very limited.

Water may be used where it is necessary to see the source 1 of radiation or to work through the shield with long-handled tools.  Used nuclear fuel (called spent fuel) is usually stored in deep bays filled with water.  The water also absorbs the heat being generated by the fuel elements.  If the water requires cooling it can be pumped through a heat exchanger.

On page 5 the KVL's of various materials are shown for a range of gamma energies.  The HVL^’s are not constant for a given material, because the relative probabilities of the three gamma absorption processes vary with the gamma energy (see lesson R.P.T. 1.2.3, pages 4 to 6).  In the range of energies of interest to us, it is generally true to say that the HVL will increase with energy.

For gamma radiations with energies in the range where Compton scattering is the predominant absorption process, the is generally about the same, regardless of the material used.  For instance, the HVL of iron for 1 MeV gamma radiation is about 0.65" as com­pared to 1.3" for heavy concrete (220 Ib/cu ft).  Since the iron is just about twice as dense as heavy concrete, the total mass required for a shield will be roughly the same for both materials.

4.5 NEUTRON SHIELDING

Fast neutrons must be slowed down before they are read­ily captured.  Fast neutrons may be slowed down by two inter­actions :

1)      Inelastic scattering of neutrons with heavy elements (especially iron).  This interaction predominates for neutron energies greater than 1 MeV.

2)      Elastic scattering with light nuclei such as hydrogen.

The resulting slower neutrons cire captured by nuclei in the shielding in an (n,γ) reaction.  Therefore gamma radia­tion will be produced as a result of the capture process and additional shielding must be cidded to absorb the gamma rays produced.

Water, paraffin, masonite and polyethylene contain a high proportion of hydrogen and are therefore effective in slowing down neutrons.  Ten inches of water or paraffin will reduce the fast neutron dose rate by more than a factor of 10.  Concrete retains some water permanently, and is there­fore very useful as a neutron absorber.  For example, the HVL of ordinary concrete (150 Ib/cu ft) for 10 MeV neutrons is only about 3 inches.

                                                               

         

4.6 REACTOR SHIELDING

To protect personnel from neutron and gamma radiation from the core, it is necessary to completely surround the reactor with a thick shield.  This is usually called the BIOLOGICAL SHIELD or PRIMARY SHIELD, because its main purpose is to protect people rather than equipment.

The radiation intensity on the outside of the shield must be such that exposures to personnel are well below the maximum permissible limits.  Concrete, which is suitable for neutron and gamma ray shielding, is commonly used because of its low cost and good structural qualities.  Most reactors require 7 or more feet of concrete to reduce the field at the outside of the shield to 1 mrem/h or less.

If space is too limited to permit the use of sufficient ordinary concrete, then special heavy concrete can be used. Such concrete has incorporated into it steel punchings, iron ore or other mineral ores to increase its density.  A shield fabricated from heavy concrete will then not need to be as thick as one fabricated from ordinary concrete.

The inner face of the shield will heat up because it absorbs a great deal of energy.  If the shield requires cooling, water pipes are embedded in the concrete near the inner surface where most of die heat is generated.

Part of the NPD shielding is shown in the diagram on page 7 to illustrate some of the problems encountered in reactor shielding.  The reactor itself is housed in the reac­tor vault, whose roof, floor and walls constitute the primary shield.  There are three types of area near the reactor zone:

1)      Areas which may never be entered are said to be .inacces­sible at all times The reactor vault and the dump pipe room are the only rooms which are never accessible.

2)      Areas which are accessible at all times; that is, they may be entered even during reactor operation.  These areas must be heavily shielded from the reactor, since the radiation levels from the core are much higher while the reactor is operating.  In NPD, 7 feet of concrete with a density of 220 Ib per cubic foot are sufficient to reduce the radiation level to 1 mrem/h or less during reactor operation.

Areas which are inaccessible during reactor operation.

The reactor, at full power, is a source of gamma rays with energies up to 10 MeV and neutrons with energies up to 15 MeV.  When the reactor is shut down,

FIG 4.6

on the other hand, only gamma rays - emitted by fission products and activation products - with energies up to about 2.5 MeV need be considered.  Therefore, for areas only accessible during reactor shutdown substantially thinner shielding is quite adequate.  Such areas usually contain radiation sources of their own during reactor operation  (e.g., short-lived activation products) and wouldn't then be accessible anyway, even if they were fully shielded.  Here the use of thinner shields will save both space and cost.

An example of an area which is inaccessible during re­actor operation is the boiler room.  It is accessible only during reactor shutdown.  The boiler room in NPD requires only 4% feet of 220 Ib/ft³ concrete shielding to reduce the radiation levels from the reactor to 7 mrem/h one minute after it is shut down.  Since the initial fission product decay is very rapid , these levels of course soon drop below 7 mrem/h.  (The fields actually measured in the boiler room are much higher, because of the residual contribution from activation products in the systems located there.)

The clue as to why the boiler room is inaccessible; during reactor  operation  is that this room contains the primary heat transport and moderator equipment. The heavy water in both systems passes through the reactor core and will become radio­active by neutron activation during reactor operation.  Most of the gamma activity is clue to nitrogen-16, which is pro­duced by the (n,p) reaction in oxygen- 16:

₀n¹ + ₈O¹⁶ → ₇N¹⁶ + ₁P¹

Nitrogen-16 emits Jilgh energy gamma rays (6-7 MeV) and has a half-life of 7.4 seconds.  Therefore the items of equipment which contain this heavy  water are high radiation sources during reactor operation, and cause a gamma radiation field of several R/h.  The boiler room itself is shielded to protect adjacent areas from this radiation field.

After the reactor is shut down, no more N¹⁶ is formed and the N^l6 already there decays very rapidly, so that the boiler room may soon be entered under normal conditions.

5.1 RADIATION PROTECTION 4

1.  Radiation Protection

3.  Radiation Exposure

5.1 RADIATION PROTECTION 4

1.  Radiation Protection 3.  Radiation Exposure 5.  INTERNAL RADIATION

INTRODUCTION

Radionuclides will be potentially much more harmful when they have been taken into the body to become INTERNAL SOURCES than they would have been outside the body.  Therefore, one of the primary and also one of the most difficult objectives of radiation protection is to minimize or prevent the intake of radioactive materials into the body.  In this lesson we will consider some of the problems experienced with internal ra­diation.

5.2 INTERNAL RADIATION HAZARDS

Among the first people to suffer injury and death from the effects of internal sources were the radium dial painters They were employed in a factory during the 1920^fs applying radium-containing paint to instrument dials.  Many of these workers later suffered from anemia and bone injury.  Investi­gations showed that they used to point their brushes between their lips after dipping them in the radioactive paint.  In this way relatively large quantities of radium entered their gastrointestinal tracts, some was absorbed into their bodies, the majority of this being deposited in their bones.  After death it was shown that some of their skeletons contained many micrograms of radium.

When dealing with external exposure the protection pro­blem is relatively straight forward; usually the radiation level can easily be measured with an instrument, and if de­sired the exposure can be terminated at any time.  However, the hazard presented by internal sources is much more dif­ficult to cope with for the following reasons:

1)      Internal sources irradiate the body tissues 24 hours a day, seven days a week, until they have been eliminated from the body by excretion and decay.

2)      Many radionuclides remain in the body for years, and in many cases it is difficult if not impossible to increase their rate of elimination from the body.

3)      It is difficult to determine the quantity and the distribution of many of the radionuclides when they are contained in the body, and therefore the dose rate frequently cannot be accurately assessed.

4)      The decaying rcidioactive atoms emit particles within the tissue.  Thus alpha and low energy beta particles which are not sufficiently penetrating to be an external hazard are a very serious inter­nal hazard.  They dissipate all of their energy in a very small volume of tissue.  Gamma emitters taken internally are of course a hazard too_, but the damage is not so concentrated.  In any case, only a fraction of the gamma photon energies are absorbed before the photons leave the body.

5)      Some radionuclides are concentrated chemically in certain organs of the body, thus increasing the probability of injury to these organs.

5.3 ENTRY OF RADIONUCLIDES INTO THE BODY

Radioactive materials may occur in many physical or chemical forms just as other materials do.  They may appear as solids, powders, dusts, liquids, gases, vapours or solu­tions.  Internal contamination can result from the careless handling of such radioactive material.  It may enter the body in four different ways:

1)      Inhalation (breathing it),

2)      Ingestion (eating it), Absorption through the unbroken skin,

3)      Absorption through wounds.

In the nuclear power industry inhalation is generally considered to be the most likely route of entry of toxic materials into the body.  Ingestion is the second most likely. Inhaled material cleared from the lungs often enters the gastrointestinal tract and then a secondary ingestion type of exposure occurs.

The amount of a radionuclide taken up by the body depends on the magnitude of the intake, solubility (which in turn depends on the particular chemical form of the radionuclide), and whether the intake is via inhalation, ingestion or ab­sorption.

Inhalation - A portion of insoluble dust which is inhaled is retained for some time in the respiratory tract and irradiates the lung tissues.  The rest is exhaled or swallowed. In the case of soluble substances which are inhaled_, a high percentage of that taken into the lungs passes rapidly from the lungs into the blood stream.

Ingestion  - Although ingested insoluble material is mostly eliminated from the body in the faeces, the main hazard is the dose it delivers to the gastrointestinal tract during its passage. Ingested soluble material is largely absorbed into the blood stream.

Absorption - Radioactive materials absorbed through the

skin or through a wound also enter the blood stream. The latter can obviously be a route of rapid entry.

To summarize the above, soluble materials which are inhaled and ingested go largely to the blood stream, while insoluble materials irradiate the respiratory tract or the gastrointestinal tract while they are present.

5.4 DISTRIBUTION OF RADIONUCLIDES IN THE BODY

A large percentage of radionuclides which have entered the body are eliminated in the first few days.  However, a portion will be absorbed in various organs depending on the type of radionuclide.  The body deals with elements and compounds on a chemical basis.  For example, normal inactive iodine (I¹²⁷) concentrates in the thyroid gland.  If radio-afetive I^l31 i_s present as well, the body cannot differentiate between the two isotopes which are chemically identical, and the active I¹³¹ also concentrates in the thyroid.

Some elements are so closely related chemically that the body cannot always effectively differentiate between two different elements.  For example, chemists say that calcium, strontium, barium and radium are in the same group (all have 2 electrons in their outer orbit).  Calcium present in the body is largely deposited in the bone, and any radioisotopes of strontium, barium and radium which enter the body will therefore also collect to a considerable extent in the bone. Such radioisotopes (called BONE-SEEKERS) are excreted at a very slow rate once they have been deposited in the bone. If their radioactive half-life is long, they may therefore irradiate the sensitive bone marrow, as well as the bone, for many years.

5.6 elimination of radionuclides from the body

In addition to the tendency for a particular element to be taken up by a particular organ or tissue, the main consid­eration in determining the hazard of a given radioisotope inside the body is the total radiation dose delivered to the critical organ while it is in the body.  The most important factors determining this dose are

1)      The mass of radioactive material deposited,

2)      The nature and energy of the radiations emitted,

3)      The length of time it is effective in the body.

This time depends on two factors: one is the ordinary radio­active half-life and the other is the

5.7 BIOLOGICAL HALF-LIFE.

The BIOLOGICAL HALF-LIFE is the time taken for the amount of a particular element in the body to decrease to half its initial value due to elimination by natural (biological) processes alone.

The biological half-life depends on the rate at which the body normally uses a particular compound of an element.  Combina­tion of the radioactive half-life and the biological half-life gives rise to the EFFECTIVE HALF-LIFE.

The EFFECTIVE HALF-LIFE is the time taken for the amount of a specified radionuclide in the body to decrease to half its initial value as a result of both radioactive decay and natural elimination.

The effective half-life, T_e, is given by the equation

Where T_r and T_b are the radioactive and biological half lives respectively.

Serious internal hazards are presented by those radio-nuclides which have long effective half-lives, such as Ra²²⁶ and Pu²³⁹ with45 and 200 years respectively.  Once deposited in the bones they remain there essentially unchanged in amount, during the lifetime of the individual.  The continued action of the emitted alpha particles, which deposit their energy in a limited region, over a period of years can cause significant injury.  This was the case with the radium dial painters .

A similar situation exists for the beta-emitting bone-seeker Sr^O, one of the more common fission products.  It has '^radioactive half-life of 28 years and a biological half-life estimated to be about 50 years.  Its effective half-life, T_e, is then given by

This long effective half-life of 18 years, coupled with the fact that it is a bone-seeker, makes Sr⁹⁰  one of the most hazardous of fission products.

5.8THE PHYSICAL FORM OF SOME COMMON INTERNAL CONTAMINANTS

Dusts:

Many of the fission products and activation products which, are encountered as air contaminants in a reactor are present as dust.  The radionuclides are often collected on small dust particles of other materials, and these particles floating about in the air act as carriers for the radioactiv­ity, which may then be inhaled.  If the dust is collected u0n a filter paper, so is the radioactivity.

Gases:

The radioisotopes of the rare gases are of course present as gases.  Some of these are argon-41, xenon-133, xenon-135 and krypton-88.  The rare gases are chemically inert, which means they will not combine with other elements. Fortunately the body has no use for these elements, and therefore they do not tend to become concentrated in the body.  The rare gases are considered to be an external radia­tion hazard (i.e. exposure to a cloud of radioactive gas) rather than an internal one.  However, this is not so for any radioactive daughters they may produce.  Rubidium-88, formed by the beta decay of krypton-88, is collected on dust particles and can present an internal hazard.

Vapour:

The radioiodines are usually present in the form of vapours, although a part of the iodine may be absorbed on dust.  These radionuclides are readily absorbed by the body and are concentrated in one specific organ, namely the thyroid.  The rare gases and the iodines are the radioele-ments which escape most readily from defective fuel elements.

Tritium vapour is an internal radiation hazard of con­siderable importance in heavy water reactors.  For this reason the next but one lesson is devoted entirely to tri­tium and therefore we need say no more about it here.