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Internal SFUND Box Number 35 OF 283 

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P.2 

/<79-/ef->^c"S -2. 

TPRD/B/0340/R83 

Central Electricity Generating Board 

Technology Planning and Research Division 

Berkeley Nuclear Laboratories 

THE Q SYSTEM FOR THE CALCULATION OF A. 

AND A2 VALUES WITHIN THE IAEA REGULATIONS 

FOR THE SAFE TRANSPORT OF RADIOACTIVE 

MATERIALS 

By H.F. Macdonald and E.P. Goldfinch 

September 1983 
P.3 

The Q System for the Calculation of A. and A Values 

Within the IAEA Regulations for the Safe Transport 

of Radioactive Materials 

- by -

H.F. Macdonald and E.P. Goldfinch 

Contents 

Approved 

Classification and Distribution 

Approved 

1A>" 

Head of 

Plant Chemistry Branch 

Date: "tH" 

Direc tor \sJ4{^^ 

Date: . ; ••' / - / s •* 

To be published 

* 

Health and Safety Department, Courtenay House, London 
P.4 

V. 

1. INTRODUCTION 

In 1978 the IAEA entered into a Research Agreement (No. 2291/CF) with 

the CEGB entitled "Release Limits for Irradiated Fuel Transport Flasks". 

Subsequently, SAGSTRAM (The IAEA Standing Advisory Group on the Safe 

Transport of Radioactive Materials) strongly urged the Agency to establish 

an international coordination of research, development and testing programmes 

and to disseminate the resulting information to member states. In September 

1979 the IAEA Committee for Contractual Scientific Services recommended the 

initiation of a Coordinated Research Programme and the Agency approved 

incorporation of the CEGB Research Agreement into that programme. 

An essential part of the work associated with the CEGB Research 

Agreement has entailed a re-examination of the dosimetric models used in 

the derivation of the Type A package contents limits within the IAEA 

Regulations for the Safe Transport of Radioactive Materials; these limits 

are also used in specifying Type B package activity leakage limits, LSA 

(low specific activity) and exemption limits, etc. Cl, 2]. The re-examination 

of the earlier models in turn resulted in the development of the Q system, 

an improved method for the evaluation of the contents limits for special 

form (non-dispersible"i and non-special form (dispersible) materials, the A 

and A„ values respectively. 

The results of this work have been reported extensively elsewhere 

[3-8] and have been considered by the IAEA Panels reviewing the Transport 

Regulations and by SAGSTRAM. In March 1982 the IAEA Advisory Group for the 

Comprehensive Review of the Regulations adopted the Q system for the 

calculation of A. and A„ values in the third draft revision of the 1984 

Edition of the Transport Regulations, subject to review in the light of 

comments received from member states at a Special Working Group (SWG) 

meeting convened in London in September 1982. 

The revised methods of determining A. and A- values reported here 

result from the discussion of the Q system at the London SWG meeting, the 

participants at which are listed in Appendix A. The earlier, somewhat 

pragmatic, dosimetric models incorporated in the 1973 IAEA Transport 

Regulations have been extended and improved and now have a firmer physical 

basis which is more defensible in public debate. In addition, they 

incorporate the latest metabolic data and concepts recommended in ICRP 

Publications 26 and 30 [9-12]. This report is intended to provide a 

- l 
P.5 

comprehensive account of the basis of the Q system and the detailed 

considerations'underlying the methods now used in the derivation of A., and 

A- values within the IAEA Transport Regulations. 

2. BACKGROUND 

The various limits for the control of releases of radioactivity from 

transport packages prescribed in the IAEA Transport Regulations are based 

upon the activity contents limits for Type A packages. Type A packages 

are intended to provide economical transport for large numbers of smallactivity consignments, while at the same time ensuring a high level of 

safety. The contents limits are set so as to ensure that the radiological 

consequences of severe damage to a Type A package are not unacceptable and 

design approval by the Competent Authority is not required, except in 

respect of criticality considerations. 

Activities in excess of the Type A package limits are covered in the 

IAEA Regulations by the requirements for Type B packages, which do require 

Competent Authority approval. The design requirements for Type B packages 

are such as to reduce to a very low level the probability of significant 

activity release from such packages as a result of a very severe accident. 

Originally, radionuclides were classified into seven groups for 

transport purposes [13, 14], each group having its Type A package contents 

limits for special form material and for material in all other forms. 

Special form material was defined as that which was non-dispersible when 

subject to specified tests. In the 1973 Regulations the group classification 

system was developed into the A.-/A„ system [1, 2], in which each nuclide 

has a Type A package contents limit, A_ curies, vhen transported in special 

form and a limit, A- curies, when in non-special form. 

The dosimetric bases of the A.-/A9 system relied upon a number of 

somewhat pragmatic assumptions. A whole body dose limit of 3 rem (30 mSv) 

was assumed in the derivation of A, although in calculating A. values the 

exposure was limited to 3R at a distance of 3 metres in a period of 3 hours 

(see para. 4.12 of reference 2). Also, an intake of 10 x A-., leading to 

half the maximum permissible annual intake for a radiation worker, was 

assumed in the derivation of A-, as a result of a "median" accident. The 

median accident was defined arbitrarily as one which leads to complete loss of 

-3 shielding and to a release of 10 of the package contents in such a manner 

-3 

that 10 of this released material was subsequently taken in by a bystander. 

The Q system described here includes consideration of a broader range of 
P.6 

specific exposure pathways compared with the earlier A./A- system. Many of 

the assumptions made are similar to those stated, or implied, in the 1973 

Regulations, but in situations involving the intake of radioactive material 

use is made of new data and concepts recently recommended by the ICRP [9-12]. 

In particular, subjective assumptions are made regarding the extent of package 

damage and release of contents, as discussed later, without reference to a 

median accident. 

3. THE BASIS OF THE Q SYSTEM 

Under the Q system a series of exposure routes are considered, each 

of which might lead to radiation exposure, either external or internal, 

to persons in the vicinity of a Type A package involved in a severe transport 

accident. The dosimetric routes are illustrated schematically in Figure 1 

and lead to five contents limit values Q., CL,, Q_, Q_ and Q_, for external 

A 0 L U _. 

photon dose, external beta dose, inhalation dose, skin and ingestion dose 

due to contamination transfer and submersion dose respectively. Contents 

limits for special form alpha and neutron emitters and tritium are considered 

separately. In addition, consideration is given to the physical form of 

package contents where physical limitations may be more restrictive than 

radiological considerations. 

Type A package contents limits are determined for individual radionuclides, as in the 1973 Regulations. The A. value for special form 

materials is the lesser of the two values Q, and Q_, while the A„ value for 

A H i. 

non-special form materials is the least of A. and the remaining Q values. 

Specific assumptions concerning the exposure pathways used in the derivation 

of individual Q values are discussed below, but all are based upon the 

following fundamental assumptions: 

a) The effective or committed effective dose equivalent 

to a person exposed in the vicinity of a transport 

package following an accident should not exceed the 

annual dose limit for radiation workers, namely 

50 mSv (5 rem). 

b) The dose or committed dose equivalent received by 

individual organs, including the skin, of a person 

involved in the accident should not exceed 0.5 Sv 

(50 rem), or in the special case of the lens of the 

eye 0.15 Sv (15 rem). 
P.7 

c) A peeson is unlikely to remain at 1 metre from the 

damaged package for more than 30 minutes. 

Here the annual dose or intake limits for radiation workers replace the 

earlier quarterly limits used in the 1973 Regulations, the latter limits no 

longer being included in the recommendations of the ICRP [9]. This approach 

was considered by the London SWG meeting as being acceptable on a once in a 

lifetime basis for members of the public inadvertently exposed near the 

scene of a severe accident involving a Type A package. Compliance with the 

assumptions (a) and (b) above is automatically ensured by the use of the 

lowest of the stochastic or non-stochastic annual limit on intake (ALI) or 

the derived air concentration (DAC) values recommended by the ICRP [10-12], 

as appropriate. 

Further, the exposure period of 30 minutes at a distance of 1 metre 

is a cautious judgement of the incidental exposure of persons initially 

present at the scene of an accident, it being assumed that subsequent 

recovery operations take place under health physics supervision and control. 

This is considered to be more realistic than the earlier assumption of 

exposure for 3 hours at a distance of 3 metres. Coupled with the dose 

limits cited above it leads to a limiting.dose rate from the damaged package 

for whole body photon irradiation of 0.1 Sv.h at 1 metre. 

4. DOSIMETRIC MODELS AND ASSUMPTIONS 

In this section the dosimetric models and assumptions underlying the 

derivation of five principle Q values are described in detail. The 

specific radiation pathways considered are outlined and the considerations 

affecting the methods of derivation used are discussed. 

Q - External Dose Due to Photons 

A _____ " ' 

The Q value for a radionuclide is determined by consideration of the 

external radiation dose due to gamma or x-rays to the whole body of a 

person exposed near to a damaged Type A package following an accident. The 

shielding of the package is assumed to be completely lost in the accident 

and the consequent dose rate at a distance of 1 metre from it is limited 

to 0.1 Sv.h . Assuming further that the damaged package may be treated 

effectively as a point source, the free air exposure rate at 1 metre can be 

expressed as D = 1.4 x 10 x C x E^ R.h , where C is the source activity 

in Bq, and E is the mean photon energy release per disintegration expressed 

••'n MP V f TS,—L6_l_ Thp ranw nf var-iar-fnn in the mrniprical constant in 
P.8 

I/. 

this expression for photon energies in the range from 50 keV to 5 MeV is 

± 30%, while over the same energy range the effective dose equivalent per 

'""*"• -2- 1 

unit exposure in free air is approximately 0.67 x 10 Sv.R (see Figure 10 

12 

of reference 17). This leads to an expression for Q of 1.07 x 10 /E Bq, 

A y 

which in deriving the results presented here has been rounded to give: 

^A

 = yfevT' TBq' 

The values of E^ employed here were evaluated taking into account the detailed 

decay schemes of individual radionuclides [18], with no.lower energy limit cutoff applied. This approach is not strictly valid for E^ values below 50 keV 

since the exposure per unit energy flux increases rapidly, while the effective 

dose equivalent per unit exposure decreases, with decreasing gamma energy. 

These two effects compensate to some degree so that the above expression remains 

valid within the stated limits down to about 20 keV, although for veak gamma or 

soft x-ray emitters of around 10 keV energy it may overestimate Q, values by 

about an order of magnitude. This effect is only important where E is dominated 

by photon emissions in this range and has been ignored in the tabulation of Q 

values presented here. 

Q„ - External Dose Due to Beta Emitters 

The Q_ value is determined by consideration of the beta dose to the 

B 

skin of a person exposed following an accident involving a Type A package 

containing special form material. The shielding of the transport package 

is again assumed to be completely lost in the accident, but the concept of 

a residual shielding factor for beta emitters (associated with materials 

such as the beta window protector, package debris, etc.) included in the 

1973 Regulations is retained [2]. These assumed a very conservative 

shielding factor of 3 for beta emitters of maximum energy _: 2 MeV and 

within the Q system this practice is extended to include a range of shielding 

factors dependent on beta energy based on an absorber of ^ 150 mg.cm 

thickness as follows: 

E_ (MeV) 3max 

<Q.5 

0.5-<0.7 

0.7-<1.0 

1.0-<1.5 

1.5-<2.0 

>2.0 

Shielding Factor 

>100 

100 

20 

6 

3 

2 
P.9 

These factors are derived frcm the standard expression for the attenuation 

of beta-rays aV a function of energy [19], and the approximation involved 

can be seen from Figure 2 in which the above values are compared with the 

continuous curve obtained from' the full analytical expression. In addition, 

the factors used here differ from those originally used in the Q system [4] 

in that following the discussions at the SWG meeting no attempt is made 

to represent any self-shielding effects associated with the beta source 

itself. 

Based on the dose limits cited earlier the beta dose equivalent rate 

to the skin at 1 metre frcm the damaged package is limited to 1 Sv.h ; 

this is closely equivalent to an absorbed beta dose rate in air of 1 Gy.h 

The variation of beta dose rate with maximum beta-ray energy at 1 metre 

from a 37 MBq (1 mCi) source is shown in Figure 3 [20];.also shown as a solid 

line is the simplified representation of these data which in conjunction 

vith the shielding factors and dose rate limit cited above lead to the full 

set of Q_ values as follows: 

a 

E

8max

 (MeV ) 

<0.5 

0.5-<0.7 

0.7-<1.0 

1.0-<1.5 

1.5-<2.0 

>2.0 

QB CTBq) 

>370 

21 

4.2 

0.63 

0.32 

0.21 

For radionuclides having more than one beta-ray end-point energy, the weighted 

maximum beta energy is used in deriving Q_ values, taking into account their 

detailed decay schemes [18]. Further, it should be noted that although the 

dose limit for the lens of the eye is lower than that for the skin (0.15 Sv 

c f . 0.5 Sv), consideration of the depth doses in tissue for beta emitters 

-2 

and in particular the absorption at the 300 mg.cm depth of the sensitive 

cells of the lens epithelium indicates that the dose to the skin is always 

limiting for maximum beta energies up to ^ 4 MeV [9, 20, 22]. Specific 

consideration of the dose to the lens of the eye is thus unnecessary. 

Finally, mention should be made of the treatment of positron annihilation 

radiation and conversion electrons in the determination of Q values. The 

latter are treated as beta particles of maximum energy equal to the conversion 

electron energy or an equivalent weighted value where more than one decay 
P.10 

been included in the evaluation of the beta dose to the skin since it 

contributes only an additional few percent to the local dose to the basal 

layer. However, the 0.51 MeV gamma rays are included in the photon energy 

per disintegration used in the derivation of Q as discussed above. 

Q - Internal Dose via Inhalation 

\_i 

The Qr value for a radionuclide transported in a non-special form is 

determined by consideration of the inhalation dose to a person exposed to 

the activity released from a damaged Type A package following an accident. 

Compliance with the limiting doses cited earlier is ensured by restricting the 

intake of activity under accident conditions to the annual limits on intake 

(ALI) recommended by the ICRP [10-12]. The concept of the "median" accident 

used in the 1973 Regulations is no longer employed since its definition 

involved a circular argument, namely that a median accident was one leading 

-3 .. . 

to a release of 10 of the package contents coupled with a dosimetric 

-3 

model vhich assumed that such an accident released 10 of the package 

-3 . contents and that 10 of this release was incorporated into a person. 

Under the Q system a range of accident scenarios is-considered, 

including that originally proposed for the derivation of Qr, encompassing 

accidents occurring both indoors and out-of-doors and including the possible 

effects of fires. These are discussed in detail in Appendix B. Following 

—6 

discussion at the London SWG meeting an intake factor of 10 vas agreed, 

as in the 1973 Regulations. However, within the Q system this value 

-3 -2 

represents a combination of releases typically in the range up to 10 -10 

of the package contents as a respirable aerosol, combined with an uptake 

-4- 3 .. . 

factor of up to 10 -10 of the released material. Together vith the limiting 

doses cited earlier, this leads to an expression for the contents limit 

based on inhalation of the form: 

ALI (Bq) _,_, 

L IO*5 

This expression does not explicitly take account of the chemical form in 

vhich material is transported, but in applying it use of the most restrictive 

of the ALI values recommended by the ICRP ensures that the most restrictive 

situation is considered. Where the likely degree of solubility (class D, 

W or Y) or the chemical form (organic or inorganic) of a particular radionuclide are known, the degree of pessimism introduced by this procedure can 

be quantified by reference to the tabulated ALI values [10-12]. 
P.11 

Q - Skin Contamination and Ingestion Doses 

The Q value for beta emitters is determined by consideration of the 

beta dose to the skin of a person contaminated with non-special form 

material as a consequence of handling a damaged Type A package. This is a 

dosimetric route which was not considered previously. The model proposed 

within the Q system assumes that 1% of the package contents are spread 

2 . . 

uniformly over an area of 1 m ; handling of the debris is assumed to result 

in contamination of the hands to 10% of this level [21]. It is further 

assumed that the exposed person is not wearing gloves but would recognise 

the possibility of contamination or wash the hands within a period of 5 hours. 

Taken individually these assumptions are somewhat arbitrary, but the 

SWG meeting concluded that taken as a whole they represent a reasonable 

basis for estimating the level of skin contamination which might arise under 

-3 -2 accident conditions. This is 10 xQ m , with a dose rate limit for the 

skin of 0.1 Sv.h based on a 5 hour exposure period. Noting the equivalence 

of grays and seiverts for beta irradiation, these parameters coupled with 

the beta dose rate data shown in Figure 4 result in the following values of 

Q_ as a function of weighted maximum beta energy. 

E. (MeV) Smax 

<0.1 

0.1-<0.15 

0.15-<0.2 

0.2-<0. 5 

2:0.5 

QD (TBq) 

>46 

9.3 

2.3 

0.9 3 

0.46 

-2 

The data of Figure 4 are the beta dose rate at a depth of 5-10 mg.cm in 

-2 -2 tissue for an extended plane source of 0.37 TBq.m (1 mCi.cm ) [22]; this 

depth is that recommended by the ICRP for the estimation of the depth of 

the sensitive basal layer of the skin [23], while the solid line shown in 

Figure 4 is the simplified representation of these data for a representative 

-2 

basal layer depth of 7 mg.cm 

The models used in deriving the Q values here may also be employed to 

estimate the possible uptake of activity via ingestion. Assuming that a 

-3 2 2 

person may ingest all the contamination from 10 m (10 cm ) of skin over 

—6 

a 24 hour period [21] the resultant intake is 10 xQ , compared with that 

- 8 
P.12 

vl-. 

via inhalation of 10~ xQ derived earlier. Noting that the dose per unit 

intake via inhalation is generally of the same order or greater than that 

via ingestion [10-12], the inhalation pathway vill normally be limiting for 

internal contamination due to beta emitters under the Q system. Where this 

does not apply, almost without exception Q <<Q_, and explicit consideration 

of the ingestion pathvay is unnecessary. 

0 - Submersion Dose Due to Gaseous Isotopes 

^E 

The Q_. value for gaseous isotopes which do not become incorporated 

into the body is determined by consideration of the submersion dose following 

their release in an accident when transported as non-special form materials 

in either a compressed or uncompressed state. A rapid 100% release of the 

package contents into a store room or cargo handling bay of dimensions 

3 .. . 

3x10x10 m with 4 air changes per hour is assumed. This leads to an initial 

airborne concentration of Q /300 m , which falls exponentially with a decay 

-1 * 

constant of 4 h due to ventilation over the subsequent 30 minute exposure 

-3 -3 period to give a mean concentration level of 1.44x10 xQ_ m . Over the same 

period the concentration leading to the dose limits cited earlier is 

-3 

4000xDAC Bq.m , vhere DAC is the derived air concentration recommended by 

the ICRP for 40 hours per veek and 50 veeks per year occupational exposure in 

3 

a 500 m room [10-12], hence: 

Q_. = 2.8xlO~6x DAC(Bq.m~3) TBq. 

The use of DACs in the derivation of Q here automatically ensures that the 

most restrictive limit, i.e. effective dose equivalent, skin or eye dose, 

is applied. Also, the derivation given here applies to noble gases vhich 

are not incorporated into the body and vhose daughter products are either 

a stable nuclide or another noble gas isotope. In the few cases vhere this 

condition is not fulfilled dosimetric routes other than submersion in a 

radioactive cloud must be considered, as discussed belov. 

5. SPECIAL CONSIDERATIONS 

The dosimetric models described in the previous section apply to the 

vast majority of radionuclides of interest and may be used to determine 

their Q values and associated A, and A_ values. Hovever, in a limited 

number of cases the models are inappropriate or require modification. The 

special considerations vhich apply in such circumstances are discussed 

further in this section. 

- 9 -
P.13 

Alpha and Neutron Emitters 

For alpha emitters it is not in general appropriate to calculate Q. or 

A 

Q values for special form material due to their relatively weak gamma and 

B 

beta emissions. In the 1973 Regulations an arbitrary upper limit for special 

3 -

form alpha sources of 10 x A- was introduced [IJ. There is no dosimetric 

justification for this procedure and in recognition of this, coupled with 

good record in the transport of special form materials and the reduction 

in many Q values for alpha emitters resulting from the use of the latest 

ICRP recommendations, the London SWG meeting recommended a tenfold increase 

3 . . 

in the arbitrary factor of 10 above. Thus an additional Q value, 

4 

Q = 10 x Q„, is defined for special form alpha emitters and listed in 

F L 

the column headed Q. where appropriate in the tabulation of Q values. 

A 

Overall the special form alpha limits remain unchanged or are increased 

slightly compared with those in the 1973 Regulations since changes in ICRP 

data reduce many of the Q values for alpha emitters by factors of up to 10. 

Finally, in respect of the ingestion of alpha emitters arguments analogous 

to those used for beta emitters in the discussion on Q apply and the 

inhalation rather than the ingestion pathway is always more restrictive; 

hence the latter is not explicitly considered. 

In the case of neutron emitters it was originally suggested under the 

Q system that there were no known situations with (ct,n) or (y,n) sources or 

252 the spontaneous neutron emitter Cf for which neutron dose would contribute 

significantly to the external or internal radiation pathways considered 

earlier [4], However, it was noted at the SWG meeting that neutron dose 

252 

cannot be neglected in the case of Cf sources. Data given in ICRP 

Publication 21 for neutron and gamma emissions indicates a dose rate of 

3 -1 252 

2.54 x 10 rem.h at 1 metre from a 1 g Cf source [24], Combined with 

the dose rate limit of 10 rem.h at this distance cited earlier, this leads 

252 . . 

to a Q value ror Cf of 0.095 TBq, which is more restrictive than the 

Q value of 10 TBq obtained on the basis of the revised expression for 

special form alpha emitters. The neutron component dominates the external 

252 dose due to a Cf source and similar considerations apply to the two other 

potential spontaneous fission sources Cm and Cf. The Q values for 

these radionuclides were evaluated assuming the same dose rate conversion 

252 factor per unit activity as for the Cf source quoted above, with allowance 

for their respective neutron emission rates relative to that of this source. 

- 10 -
P.14 

Bremsstrahlung Radiation 

The A./A_ values tabulated in the 1973 Regulations are subject to an 

upper cut-off limit of 1000 Ci in order to protect against possible effects 

of Bremsstrahlung radiation [2j. Within the Q system this cut-off is 

retained at 40 TBq. However, following discussion at the SWG meeting it is 

now recognised as an arbitrary cut-off and is not specifically associated 

with Bremsstrahlung radiation or any other dosimetric consideration. 

Parent Nuclides of Short-Lived Daughters 

The special case of short-lived daughters which may come into equilibrium with a longer lived parent during transit is treated under the Q 

system as in the 1973 Regulations [2]. Thus, based on an assumed maximum 

journey time of 50 days short-lived daughters of half-life less than 10 days 

are assumed to be in equilibrium with a longer lived parent. In such cases 

Q values, and hence A./A„ values, for both the parent and daughter nuclides 

are calculated and the most limiting values are assigned to the parent. 

It was further agreed at the London SWG meeting that where this procedure 

had been applied this would be indicated in all future tabulations of A 

and A„ values. Finally, in cases where daughter nuclides have half-lives 

either greater than 10 days or than that of the parent, the parent and such 

daughters should be treated as mixtures of different radionuclides and their 

Type A package contents limits determined as discussed below. The tabulation 

of A./A_ values considered below has been expanded beyond that given in the 

1973 Regulations to include all radioisotopes of-half-life _: 10 days. 

Tritium and its Compounds 

In the case of both inorganic and organic forms of tritium the dosimetric 

models discussed earlier result in Q values in excess of the arbitrary 

40 TBq cut-off limit for A_/A_ values adopted in the Q system. However, 

due to the special biological behaviour of tritium and' its compounds 

additional concentration limits are necessary for this radionuclide. A 

range of possible accident scenarios involving the release of tritium with 

consequent wetting of the skin of the hands of exposed persons, and 

inhalation of saturated vapour in a confined space or out-of-doors are 

considered in Appendix C; these lead to concentration limits of 50 Ci.1 , 

5 Ci 1 and 20 Ci 1 respectively. Following discussion at the SWG 

.... . -3 

meeting a concentration limit for tritium and its compounds of 1 PBq.m 

(27 Ci.l ) was determined as being appropriate to cover such cases, to be 

applied in addition to the A./A, value limits of 40 TBq. 

- 11 -
P.15 

Low Specific'Activity Materials 

The 19 73 IAEA Transport Regulations recognised a category of materials 

whose specific activities are so low that it is inconceivable that an 

intake could give rise to a significant radiation hazard, namely low specific 

activity (LSA) materials. These were defined in terms of a model which 

assumes that it is most unlikely that a person would remain in a dusty 

atmosphere long enough to inhale more than 10 mg of material [2]. This 

model is retained within the Q system and leads to an LSA limit of 

-4 -1 10 x Q g ; thus the Q values for those radionuclides whose specific 

activity is below this level are listed as "Unlimited". In the cases where 

this criterion is satisfied the radioactivity associated with 10 mg of the 

nuclide is less than the appropriate ALI value recommended by the ICRP. 

Natural uranium and thorium, depleted uranium and other materials such 

238 232 235 

as U, Th, U, etc. satisfy the above LSA criterion, as does uranium 

235 235 

enriched to < 5% in U. However, for U enrichments in the range from 

5 to 20% the permitted intakes of material estimated on the basis of the 

model outlined above range from 10 to 2.7 mg. These levels of intake compare 

with a maximum permitted daily intake of 2.5 mg of uranium recommended by 

the ICRP on the basis of chemical toxicity considerations [25]. Such 

considerations will be more restrictive than the radiological limits 

235 embodied within the Q system for U enrichments up to ^ 20%, although 

following the practice in the 1973 Regulations they are not taken into 

account in tabulating the Q values and associated A./A values for uranium 

isotopes. 

A further consideration relevant to LSA materials in the context of 

the skin contamination model used in the derivation of Q is the mass of 

material which might be retained on the skin for any significant period of 

time. The consensus view of the London SWG meeting was that typically 

-2 1-10 mg.cm of dirt present on the hands would be readily discernable and 

would be removed promptly by wiping or washing, irrespective of the possible 

presence of radioactivity. It was agreed that the upper extreme of this 

range was appropriate as a cut-off for the mass of material retained on 

the skin and in combination with the skin contamination model for Q discussed 

-5- 1 earlier, this results in an LSA limit of 10 x Q g . On this basis Q 

values for radionuclides for which this criterion applies are also listed 

as "Unlimited" in the tabulation of Q values. 
P.16 

Radon and it£ Daughter Products 

As noted earlier, the derivation of Q applies to noble gases which 

_. 

are not incorporated into the body and whose daughter products are either 

a stable nuclide or another noble gas. In a few cases this condition is 

not fulfilled and dosimetric routes other than external exposure due to 

submersion in a radioactive cloud must be considered [26]. The only case 

of practical importance within the context of the IAEA Regulations is that 

229 

of Rn where the lung dose associated with the inhalation of the shortlived radon daughter products has received special consideration by the 

ICRP [27]. 

222 

In deriving the Q values for Rn here full account is taken of the 

daughter radiations, rather than applying the most restrictive of the 

shorter lived daughter product Q values as discussed earlier. The inclusion 

of the daughters in this manner results in a small reduction in the Q and 

222 . Q values compared with those obtained on the basis of the Rn radiations 

o 

alone. The corresponding Q value is 3.6 TBq; however, allowing for a 100% 

-3- 2 release of radon, rather than the 10 - 10 aerosol release fraction 

incorporated in the Q model, this reduces to a Q value in the range 

-3 -2 Z22 

3.6 x 10 to 3.6 x 10 TBq. Further, treating Rn plus its daughters 

-3 

as a noble gas results in a Q value of 4.2 x 10 TBq, towards the lower 

F 

end of the range of Q_ values, and this is the Type A package non-special 

222 form limit cited for Rn in the tabulation of Q values. 

6. TABULATION OF Q VALUES 

A full listing of Q values determined on the basis of the models 

described in the previous sections is given in Table 1. Also included are 

the corresponding Type A package A. and A- contents limit values for special 

form and non-special form materials respectively. The Q values shown in 

Table 1 have been rounded to 2 significant figures and the A. and A_ values 

to one significant figure; in the latter case the arbitrary 40 TBq cut-off 

has also been applied. 

A comparison of A./A values determined on the basis of the Q system 

with those in the 1973 Regulations is presented in Figure 5 for a representative range of radionuclides. In general, the Q system values lie 

within a factor of 2-3 of the earlier values. Where major variations 

occur these are generally associated with reductions in the Type A package 

contents limits for beta emitters determined by Q_ and particularly Q_. 

a u 

- 13 
P.17 

In addition," relaxations in the contents limits occur for some special form 

alpha emitters due to the revised definition of Q and in certain instances 

due to changes in dosimetric data recommended by the ICRP, as in the case 

f

 90

ot Src . 

In drafting the 1973 Regulations the IAEA considered the retention of 

a group classification system, rather than listing individual A, and A,, 

values for each radionuclide (see para. 4.06 of reference 2). Inspection 

of the A, and A? values given in Table 1 indicates that a grouping on the 

basis of three groups per decade, say in the ratio 1:2:5, would not involve 

undue loss of accuracy in rounding. However, the London SWG meeting felt 

.on balance that the individual listing used in the 1973 Regulations had not 

proved inconvenient and should be retained for the present. 

Consideration of Physical Properties 

A further factor considered by the SWG meeting was the need to apply 

additional limits for materials whose physical properties might render 

invalid the assumptions used in deriving the Q values discussed above. 

Such considerations are relevant to materials which may become volatile 

at the elevated temperatures which could occur in a fire, or which may be 

transported as very finely divided powders, and especially for the model 

used to evaluate the Q values. However, on balance it was felt that only 

-6 in the most extreme circumstances would the assumed intake factor of 10 

be exceeded and that special modification of the Q„ model was unnecessary 

for these materials. 

Multiple Exposure Pathways 

Following the 1973 Regulations the application of the Q system as 

described here treats the derivation of each Q value, and hence each 

potential exposure pathway, separately. In general, this will result in 

compliance with the dosimetric criteria defined earlier provided that the 

doses incurred by persons exposed near a damaged package are dominated by 

one pathway. However, if two or more Q values closely approach each other 

this will not necessarily be the case. For example, in the case of a 

radionuclide transported as a special form material for which Q = Q , 

the effective dose equivalent and skin dose to an exposed person could both 

approach 50 mSv and 0.5 Sv respectively on the basis of the Q system models. 

Examination of Table 1 shows that this consideration applies only to a 

relatively small number of radionuclides and for this reason the independent 

treatment of exposure pathways is retained within the Q system. 
P.18 

SUMMARY 

This report contains a comprehensive description of the Q system, an 

improved method for evaluating the Type A package contents limit A and A. 

values within the IAEA Regulations for the Safe Transport of Radioactive 

Materials. The work reported here was performed in collaboration with the 

Board's Health and Safety Department under the terms of Research Agreement 

No. 2291/CF, which forms part of the IAEA Coordinated Research Programme 

on the Safe Transport of Radioactive Materials. The Q system was adopted 

by the IAEA in March 1982 for the calculation of A., and A_ values to be 

published in the 1984 revised edition of the Transport Regulations. This 

present report was prepared at the request of the Agency for publication 

at the same time as the revised regulations later next year. For this 

reason the format used is somewhat different from that of standard TPRD 

reports, following instead that used by the Agency in its publications. 
P.19 

CONTENTS 

1. INTRODUCTION 

2. BACKGROUND 

3. THE BASIS OF THE Q SYSTEM 

4. DOSIMETRIC MODELS AND ASSUMPTIONS 

Q - External Dose Due to Photons 

^A 

Q - External Dose Due to Beta Emitters 

Q - Internal Dose via Inhalation 

Li 

Q - Skin Contamination and Ingestion Doses 

Q - Submersion Dose Due to Gaseous Isotopes 

E 

5. SPECIAL CONSIDERATIONS 

Alpha and Neutron Emitters 

Bremsstrahlung Radiation 

Parent Nuclides of Short-Lived Daughters 

Tritium and its Compounds 

Low Specific Activity Materials 

Radon and its Daughter Products 

6. TABULATION OF Q VALUES 

Consideration of Physical Properties 

Multiple Exposure Pathways 

Mixtures of Radionuclides 

7. CONCLUSIONS 

REFERENCES 

APPENDIX A - Participants in the IAEA Special Working Group Meeting held 

in London, 31 August - 2 September 1982 

APPENDIX B - Intake Factors for Type A Package Transport Accidents 

APPENDIX C - Concentration Limits for Consignments of Tritium Transported 

in Type A Packages 
P.20 

Mixtures_of Radionuclides 

Finally, it is necessary to consider the package contents limits for 

mixtures of radionuclides, including the special case of mixed fission 

products. For mixtures whose identities and activities are known it is 

necessary to show that: 

I B(i)/A (i) H 1 or I 3(i)/A2(i) S 1, 

i i 

where B(i) is the activity of radionuclide i in the mixture, and A_(i) and 

A_(i) are the A and A values for that radionuclide. In the case 

of mixed fission products the composition of the mixture will in general be 

dependent upon the irradiation history and cooling period of the fissile 

material involved. In addition, the fission products will normally be accompanied by alpha emitting transuranic isotopes. In principle the inequalities 

given above may be applied in such cases but where the detailed composition 

of the mixture is unknown the lowest possible A. and A- values may be used. 

-5 

On this basis A. =0.1 TBq and A-. = 2x10 TBq for unknown mixtures, or if 

the presence of alpha emitters can be discounted A. = 0-.2 TBq and 

A- = 0.02 TBq. • 

7. CONCLUSIONSv 

The Q system described here represents a development and extension of 

the A./A-, system used in the 1973 IAEA Transport Regulations for the determination of Type A package contents and other limits. It incorporates the 

latest recommendations of the ICRP and by explicitly identifying the dosimetric considerations underlying the derivation of these limits provides a 

firmer and more defensible basis for the Regulations. Overall, the adoption 

of the Q system introduces only relatively minor variations in the A. and A_ 

values compared with those in the 1973 Regulations. Where major variations 

occur, say by greater than a factor of 2-3, these are generally associated 

with reductions in the Type A package contents limits for beta emitters 

determined by Q_ and particularly Q_. This latter Q value refers to a 

dosimetric route not previously considered in the Regulations, namely beta 

contamination of the skin of exposed persons. Isolated relaxations in the 

contents limits occur due to changes in dosimetric data introduced in ICRP 

90 Publication 30, as in the case of Sr, and more generally in the case of 

special form alpha emitters due to the revised definition of 0_. Finally, 

the Q system can also be applied to the specification of permitted leakage 
P.21 

limits for Type 3 packages, exemption and LSA (low specific activity) limits, 

etc., in a similar manner to the use of A_/A_ values in the 1973 IAEA Transport 

Regulations. 

ACKNOWLEDGEMENTS 

The authors wish to express their appreciation to the members of the 

London SWG meeting for their contributions to the refinement of the original 

Q system concepts. In particular to Dr. Keith Eckerman of the Health and 

Safety Division, ORNL, who undertook the verification of the Q values under 

the sponsorship of the US Department of Transportation, and to Mr. Ken Shaw 

of the NRPB who provided ALI values for radionuclides not included in ICRP 

Publication 30. Finally, it is a pleasure to acknowledge the continuing 

interest and encouragement received from Mr. Gerald Swindell, formally of 

the IAEA Division of Nuclear Safety and Environmental Protection, during the 

development of the Q system. 

REFERENCES 

1. IAEA, Regulations for the Safe Transport of Radioactive Materials, 

1973 Revised Edition (as Amended), Safety Series No. 6, IAEA, Vienna, 

1979. 

2. IAEA, Advisory Material for the Application of the IAEA Transport 

Regulations, 2nd Ed., Safety Series No. 37, IAEA, Vienna, 1982. 

3. H.F. Macdonald and E.P. Goldfinch, "An Alternative Approach to the 

A./A- System for Determining Package Contents Limits and Permitted 

Releases of Radioactivity from Transport Packages", pp 48-56 in 

Proceedings of the Sixth International Symposium on Packaging and 

Transportation of Radioactive Materials, Berlin (West), Nov. 10-14, 1980. 

4. H.F. Macdonald and E.P. Goldfinch, Radiat. Prot. Dosim., 1:29 (1981). 

5. H.F. Macdonald and E.P. Goldfinch, Radiat. Prot. Dosim., 1:199 (1982). 

6. E.P. Goldfinch and H.F. Macdonald, Radiat. Prot. Dosim., 2:75 (1982). 

7. E.P. Goldfinch and H.F. Macdonald, "A Review of some Radiological 

Aspects of the IAEA Regulations for the Safe Transport of Radioactive 

Materials", pp 205-11 in Proceedings of the Third International 

Symposium of the SRP, Radiological Protection - Advances in Theory and 

Practice, Inverness, Scotland, June 6-11, 1982. 
8. E.P- Goldfinch and H.F. Macdonald, "IAEA Regulations for the Safe 

Transport of Radioactive Materials: Revised A. and A- Values", pp 

in Proceedings of the Seventh International Symposium on Packaging and 

Transportation of Radioactive Materials, New Orleans, La., May 15-20, 

1983. 

9. ICRP, Publication 26, Recommendations of the International Commission 

on Radiological Protection, Pergamon Press, Oxford, 1977. 

10. ICRP, Publication 30, Part 1, Limits for Intakes of Radionuclides by 

Workers, Pergamon Press, Oxford, 1980. 

11. ICRP, Publication 30, Part 2, Limits for Intakes of Radionuclides by 

Workers, Pergamon Press, Oxford, 1981. 

12. ICRP, Publication 30, Part 3, Limits for Intakes of Radionuclides by 

Workers, Pergamon Press, Oxford, 1982. 

13. IAEA, Regulations for the Safe Transport of Radioactive Materials, 

Safety Series No. 6, IAEA, Vienna, 1961. 

14. IAEA, Regulations for the Safe Transport of Radioactive Materials, 

1967 Edition, Safety Series No. 6, IAEA, Vienna, 1967. 

15. R.G. Jaeger (Ed.), Engineering Compendium on Radiation Shielding, 

Vol. I Shielding Fundamentals and Methods, Springer-Verlag, 

Berlin-Heidelberg, 1968. 

16. G.W.C. Kaye and T.H. Laby, Tables of Physical and Chemical Constants 

and Some Mathematical Functions, 12th Ed., Longmans, London, 1959. 

17. R. Kramer and G. Drexler, Radiat. Prot. Dosim., 3:13 (.1982). 

18. CM . Lederer and V.S. Shirley (Eds.), Table of Isotopes, 7th Ed., 

John Wiley and Sons Inc., New York, 1979. See also ICRP, 

Publication 3S, Task Croup on Dose Calculations - Energy and 

Intensity Data for Emissions Accompanying Radionuclide Transformations, 

to be published, 1984. 

19. J.J. Fitzgerald, G.L. 3rownell and F.J. Mahoney, Mathematical Theory 

of Radiation Dosimetry, Gordon and Breach Science Publishers Inc., 

Nev York, 1967. 

20. W.G. Cross, Tables of Beta Dose Distributions, AECL Report 2793, 1967. 

21. H.J. Dunster, Maximum Permissible Levels of Skin Contamination, 

UKAEA Report AHS3(RP)R78, 1967. 
P.23 

--22. M.R. Bailey, BETA: A Computer Program for Calculating 3eta-dose Rates 

from Point and Plane Sources, CEGB Report RD/B/N2763, 1973. 

23. ICRP, Publication 23, Task Group Report on Reference Man, Pergamon 

Press, Oxford, 1975. 

24. ICRP, Publication 21, Data for Protection against Ionising Radiation 

from External Sources: Supplement to ICRP Publication 15, Pergamon 

Press, Oxford, 1973. 

25. ICRP, Publication 6, Recommendations of the International Commission 

on Radiological Protection (as Amended ,1959 and Revised 1962) , 

Pergamon Press, Oxford, 1964. 

26. A. Fairbairn, F. Morley and W. Kolb, "The Classification of Radionuclides for Transport Purposes", pp 44-66 in The Safe Transport 

of Radioactive Materials (R. Gibson Ed.), Pergamon Press, Oxford, 1966. 

27. ICRP, Publication 32, Limits for Inhalation of Radon Daughters by 

Workers, Pergamon Press, Oxford, 1981. 

28. D.H. Lohmahn, "Transport of Radioactive Materials: A Review of Damage 

to Packages from the Radiochemical Centre during Transport", pp 818-25 

in Proceedings of the Sixth International Symposium on Packaging and 

Transportation of Radioactive Materials, Berlin (West), Nov. 10-14, 

1980. 

29. A. Hadjiantonion, J. Armiriotis and A. Zannos, "The Performance of 

Type A Packaging under Air Crash and Fire Accident Conditions", 

pp 826-32 in Proceedings of the Sixth International Symposium on 

Packaging and Transportation of Radioactive Materials, Berlin (West), 

Nov. 10-14, 1980. 

30. C.3.G. Taylor, "Radioisotope Packages in Crush and Fire", pp 1347-54 

in Proceedings of the Sixth International Symposium on Packaging and 

Transportation of Radioactive Materials, Berlin (West), Nov. 10-14, 

1980. 

31. K. Stewart, "Principal Characteristics of Radioactive Contaminants 

which may Appear in the Atmosphere", pp 321-60 in Progress in Nuclear 

Energy, Series XII Health Physics, Volume II, Pergamon Press, Oxford, 

1969. 
P.24 

32. G. Wfehnej:, "The Importance of Reportable Events in the Public Acceptance", pp in Proceedings of the Seventh International 

Symposium on Packaging and Transportation of Radioactive Materials, 

New Orleans, La., May 15-20, 1983. 

33. G.E. Swindell, Private communication Feb. 16, 1983. 

34. P.M. 3ryant, "Methods of Estimation of the Dispersion of Windborne 

Material and Data to Assist in their Application, UXAEA Report 

AHSB(RP)R42, 1964. 

35. D.J. Gorman, "Consequences of a Transportation Accident Involving a 

Large Quantity of Tritiated Heavy Water", Ontario Hydro Report 

HPD-76.7, 1976. 

36. G.J. Dicke, "Assessment of Radiation Risk in the Case of Accidents with 

Truck Transportation of Tritiated Water in Drums", Ontario Hydro 

Report RMEP-IR-790-R, 1980. 

37. G.J. Dicke, "A Review of the Q System as Applied to the Safe Transport 

of Tritiated Water", Ontario Hydro Report RMEP-IR-03457-1, 1982. 

- 19 -
APPENDIX A 

Participants in the IAEA Special Working Group Meeting 

held in London, 31 August - 2 September 1982 

P.25 

Chairman: 

Rawl, R. 

Working Group Members: 

BELGIUM 

Baekelandt, L. 

CANADA 

Jack, G.C. 

Dicke, G.J. 

DENMARK 

Ulbak, K. 

FRANCE 

Hamard, J. 

Grenier, M. 

Ringot, C. 

GERMANY, FEDERAL REPUBLIC 

Kolb, W.A. 

ITALY 

Piermattei, Mrs. S. 

Faloci, C. 

JAPAN 

Kuwashima, K. 

Akaishi, J. 

Office of Hazardous Materials Regulation 

US Department of Transport, Washington DC. 

Ministry of Public Health, Brussels 

Atomic Energy Control Board, Ottawa. 

Radioactivity Management and Environmental 

Protection, Ontario Hydro, Toronto. 

National Institute of Radiation Hygiene, Bronshoj 

Commissariat a l'Energie Atomique, 

CEN Fontenay-aux-Roses/Seine. 

Commissariat a l'Energie Atomique, 

CEN Fontenay-aux-Roses/Seine. 

Commissariat a l'Energie Atomique, 

CEN Fontenay-aux-Roses/Seine. 

OF 

Physikalisch-Technische Bundesandstalt, 

Braunschweig. 

Comitato Nazionale per l'Energia Nucleare, Rome. 

Comitato Nazionale per l'Energia Nucleare, Rome 

Nuclear Transport Services Co. Ltd. Tokyo 

Tokai Research Establishment, JAERI, Ibaragi-ken. 

- 20 -
P.26 

SWEDEN -

Pettersson, B. 

Svahn, Ms-. B. 

UNITED KINGDOM 

Blackman, D.J. 

Goldfinch, E.P. 

Macdonald, H.F. 

Shaw, K.B. 

UNITED STATES OF AMERICA 

Hopkins, D.R. 

Eckermann, K. 

Swedish Nuclear Power Inspectorate, Stockholm. 

National Institute of Radiation Protection, 

Stockholm. 

Department of Transport, London. 

Central Electricity Generating Board, London. 

Central Electricity Generating Board, Berkeley, Glos. 

National Radiological Protection 3oard, Chilton, 

Oxon. 

U.S. Nuclear Regulatory Commission, Washington DC. 

Oak Ridge National Laboratories, Tennessee. 

INTERNATIONAL MARITIME ORGANISATION 

Wardelmann, Capt. H. Marine Safety Division, London 

Scientific Secretary: 

Barker, R.F. Department of Nuclear Energy and Safety, 

International Atomic Energy Agency, Vienna, Austria. 

- 21 -
P.27 

APPENDIX B 

Intake Factors for Type A Package Transport Accidents 

In this Appendix the fraction of the contents of a Type A package which 

may be released and inhaled by persons in the vicinity of an accident is 

-3 

considered. In the 1973 Regulations it was assumed that 10 of the package 

_3 

contents might escape as a result of a "median" accident and that 10 of this 

material might be taken into the body of a person involved in the accident [2]. 

This results in a nett intake factor of 10 of the package contents and, 

following discussions at- the London SWG Meeting, this value has been retained 

within the Q system. However, it is now recognised as representing a range 

of possible release fractions and uptake factors, as discussed in the main 

text of this report, and it is convenient to consider intake factors in terms 

of these two parameters independently. 

The range of release fractions now recognised under the Q system, namely 

-3 -2 10 - 10 , covers that represented by the earlier assumption in the 1973 

Regulations and the original proposal within the Q system [3, 4]. Underlying 

this there is tha tacit assumption, also contained in the 1973 Regulations, 

that the likelihood of a "major accident" which could cause the escape of a 

large part of the package contents is small [2]. To a large extent this 

approach is borne out by the behaviour of Type A packages in severe accident 

environments [28-30]. 

Data on the respirable aerosol fractions produced under accident conditions are generally sparse and only available for a limited range of 

materials. For example, for uranium and plutonium specimens under enhanced 

oxidation rate conditions in air and carbon dioxide respirable aerosol 

fractions up to M % have been reported [31], However, below this level the 

aerosol fractions showed wide variations dependent on the temperatures and 

local atmospheric flow conditions involved. In the case of liquids higher 

fractional releases are obviously possible but here the multiple barriers 

provided by the Type A package materials, including absorbent and lead 

shielding pot, remain an effective containment system even following severe 

impact or crushing accidents [30]. Indeed in a recently cited example of an 

131 

I source which was completely crushed in a highway accident, less than 2% 

of the package contents remained on the road after removal of the package 

debris [32]. 

- -?? -
P.28 

Potentially the most severe accident environment for many Type A 

packages is the combination of severe mechanical damage with a fire. However, 

even in this situation the role of debris may be significant in retaining 

released activity, as appeared to happen in the 1979 Athens DC8 aircraft 

accident [29, 30]. Frequently fires tend to produce relatively large particulate material which would tend to minimise any intake via inhalation, 

while at the same time providing a significant surface area for the absorption of volatile species and particularly vaporised liquids. A further 

molifying factor is the enhanced local dispersion associated with the convective air currents due to the fire which would also tend to reduce intake 

via inhalation. 

Based on considerations of the type outlined here a release fraction 

-2- 3 . . 

in the range 10 - 10 was agreed by the SWG meeting as appropriate for 

the determination of Type A package contents limits within the IAEA Transport 

Regulations. 

-4- 3 . . 

The 10 - 10 range of uptake factors now used within the Q system 

is based upon consideration of a range of possible accident situations, both 

indoors and out-of-doors. The original Q system proposals considered 

3 

exposure within a storeroom or cargo handling bay of 300 m volume with 4 

room air changes per hour. Assuming an adult breathing rate of 

—4 3—1 . -3 

3.3x10 m s this results in an uptake factor of M 0 for a 30 minute 

exposure period. An alternative accident scenario might involve exposure in 

3 

a transport vehicle of 50 m volume with 10 air changes per hour, as employed 

in the determination of the Type B package normal transport leakage limit in 

the 1973 Regulations [33]. Using the same breathing rate and exposure period 

-3 

as above this leads to an uptake factor of 2.4x10 , of the same order as 

the value obtained above. 

For accidents occurring out-of-doors the most conservative assumption 

for the atmospheric dispersion of released material is that of a ground level 

point source. Tabulated dilution factors for this situation at a downwind 

-4 -2 -3 

distance of 100 metres range from 7x10 to 1.7x10 s.m [34], corresponding to uptake factors in the range 2.3x10 - 5.6x10 for the adult 

breathing rate cited above. These values apply to short term releases and 

cover the range from highly unstable category A to highly stable category F 

weather conditions; the corresponding value for average category C/D conditions is 3.3x10 , towards the lower end of the range quoted above. 

- 23 -
P.29 

Extrapolation of the models used to evaluate the atmospheric dilution 

factors used here to shorter downwind distances is unreliable, but reducing the 

exposure distance by an order of magnitude to 10 metres would increase the 

above uptake factors by about a factor of 30. This is indicative that as the 

downwind distance approaches a few metres the uptake factors would approach 

-4- 3 • • 

the 10 - 10 range used within the Q system. However, under these 

circumstances other factors which would tend to reduce the activity uptake 

come into effect and may even become dominant. The additional turbulence to 

be expected in the presence of a fire has been mentioned earlier. Similar 

reductions in airborne concentrations can be anticipated due to turbulence 

originating from the flow of air around any vehicles involved in an accident 

or from the effects of nearby buildings. 

-4 -3 Thus on balance it is seen that uptake factors in the range 10 - 10 

appear reasonable for the determination of Type A package contents limits. 

Taken in conjunction with the release fractions discussed earlier, they result 

in the overall intake factor of 10 agreed by the London SWG meeting. 
P.30 

APPENDIX C 

Concentration Limits for Consignments of Tritium Transported 

in Type A Packages 

As noted in the main text of this report, tritium and its compounds are 

the subject of special consideration under the Q system. This arises becasue 

the majority of transport requirements for tritium are in the form of 

tritiated water and, in addition, tritiated water is likely to be formed 

during the transport or storage of tritium gas or other compounds of tritium. 

For these reasons the remaining discussions here relate to tritiated water, 

although it should be noted that a factor of 50 reduction is applicable in 

the special case of organic forms of tritium [10]. 

In all cases the Q values determined for tritium exceed the arbitrary 

40 TBq cut-off limit, cf . QD > 370 TBq, Q_ = 3000 TBq for tritiated water 

and Q„ = 60 TBq for organic forms of tritium. Since Type A packages designed 

for the transport of liquids are required to have double containment or 

absorbent material and to withstand a 9 metre drop test, the assumption of 

a release in the range 0.1-1% used in the derivation of Q is not unreasonable, 

Indeed at the London SWG meeting the view was expressed that it might be 

unduly conservative in many practical accident situations. 

It is inappropriate to calculate a Q_ value for tritium since the 

radiation dose to the sensitive basal layer of the skin would be insignificant for this low energy beta emitter. However, contamination of the skin 

leads to two potential internal exposure routes, namely ingestion or 

adsorption through the skin, and these result in a specific activity limitation for tritiated water transported in a Type A package. The considerations 

outlined above concerning the integrity of such packages make extensive 

wetting of the skin unlikely and in the treatment presented here it is 

assumed that only the hands are involved; this follows the original proposals 

for the derivation of tritium limits within the Q system [4]. 

If the specific activity limit for tritium is defined as C Bq.l , 

this may be determined in terms of the dose commitments due to absorption 

through the skin, D , and that via ingestion, D_. The former is given by 

[35]: 

D = 190 {1 + 0.2 x t'} x f x C/100 x 3.7 x 101° 

= 5.14 x 10~11 {1 + 0.2 x t'} x f x C mSv, 
P.31 

where t' is the tritiated water skin contact time in minutes, and f is the 

fraction of the body skin wetted. The dose via ingestion may be estimated on 

-3 ? 

the assumption that a person may ingest all the tritiated water from 10 m" 

9 

(10 cm") of skin, with the thickness of the moisture layer on the hands taken 

_ / Q 

as 10 m, and noting that an intake of one ALI (3 x 10 Bq) results in a 

committed effective dose equivalent of 50 mSv [10]. This leads to: 

D, = 50 x 10"* x C/3 x IO9 

- P =1.67x1 0 ~x C mSv. 

Further, limiting the total dose, D + D_., to 50 mSv and noting that the skin 

2 . 2 

area of the hands is 0.06 m" and that of the whole body is 1.75 m [23], and 

assuming that the skin remains wetted for a period of up to one hour, one 

obtains: 

C = 2.0 x 1012 Bq.l"1 = 2 PBq.m"3 (^50 Ci.l"1). 

An alterantive exposure scenario is that involving a Type A package involved 

in a fire in a confined space in which it is likely that tritiated water would 

evaporate rapidly. Assuming an initial ambient temperature of 38°C at a 

relative humidity of 50%, the evaporation is likely to increase the relative 

humidity to M00%. This would result in a tritiated moisture content for 

-3 -1 

air at 38°C of 23 g.m [36], leading to an intake rate of 27.3 g.h for a 

-4 3 -1 

person breathing at the typical adult active rate of 3.3.10 m .s [23]. 

This rate would be increased by 50% due to absorption through the skin giving 

a nett intake rate of 41.0 g.h . Further, if it is assumed that a person 

is unlikely to be exposed under these conditions without breathing protection for a period of more than 30 minutes, the maximum intake of tritiated 

water is 20.5 g. Equating this intake with the ALI thus leads to a specific 

activity limit for tritiated water as follows: 

C = 3 x l09/20.5 Bq.g"1 

s 0.14 x 1012 Bq.l"1 = 0.14 PBq.m"3 (^ 4 Ci.l"1). 

A more detailed treatment of a similar scenario, taking into account the 

finite evaporation rate of 200 litres of tritiated water spilt over a floor 

2 

area of 100 m [37], leads to a somewhat higher specific activity limit of 

9 - 1 1.25 x 10' Bq.g . This result is based upon use of the long-term average 

-4 3- 1 adult breathing rate of 2 x 10 m .s , and use of the higher short-term 

- 26 -
P.32 

rate cited above reduces the calculated specific activity limit for tritiated 

water to 0.77'*Bq.g~1 or 0.77 PBq.m"3 (^20 Ci.l"1). 

Finally, detailed assessments of the likely releases of tritiated water 

under accident conditions from strong industrial containers used for road 

transport have been carried out [36]. These studies considered the probability 

of occurrence of accidents of varying severity and of various meterological 

conditions. They showed that in ^90% of accidents involving tritiated water 

of 20 Ci.l specific activity contained in 55 gallon type CTC-5B drums the 

resulting dose commitments to exposed persons were 5 rem or less. Consideration of the results of this study, noting that Type A packages are 

capable of withstanding conditions of much greater severity than strong 

industrial packages and the cases considered earlier, led to the London SWG 

meeting recommending a Type A package specific activity limit for consign-

-3 -1 

ments of tritiated water of 1 PBq.m (27 Ci.l ) . 

- 27 -
P.33 

TABLE 1 - Type A Package Contents Limits : QA, Q !B» ^C> etc. 

— - -, c- - _,- - ^A, ^3, *^, 

Values and Limits for Special Form (A-^) and Non-special 

Fora (A2) Materials 

Radionuclide 

Q. or q,+ 

(TBq) 

QB(TBq) Qc(TBq) On or QE* 

(TBq) 

A1(TBq) A2(T3q) 

Ac-2250 

Ac-227 

Ac-228 

Ag-105 

Ag-108m 

Ag-llOm 

Ag-111 

Al-26 

Am-241 

Am-242m 

Am-243 

Ar-37 

Ar-39 

Ar-41 

Ar-423 

As-72 

As-73 

Aa-74 

As-76 

As-77 

At-211 

Au-193 

Au-194 

Au-195 

Au-196 

Au-198 

Au-199 

3a-131 

Ba-133m 

Ba-133 

Ba-1400 

Be-7 

Be-10 

Bi-205 

Bi-206 

Bi-207 

Bi-210m^ 

Bi-210 

Bi-2120 

7.5 

4.3x103 

1.1 

1.9 

0.62 

0.36 

38 

0.37 

2.0+ 

2.0+ 

2.0+ 

4.4x103 

-

0.78 

3.6 

0.56 

62 

1.3 

2.3 

110 

26 

6.3 

0.96 

12 

2.2 

2.5 

11 

2.2 

15 

2.5 

0.43 

20 

-

0.61 

0.31 

0.65 

3.9 

-

0.82 

0.63 

>370 

0.63 

>370 

>370 

>370 

0.63 

4.2 

>370 

21 

21 

>370 

21 

0.63 

0.21 

0.21 

>370 

21 

0.21 

21 

1.0xl0"z 

2.0x10"5 

0.40 

40 

0.90 

3.0 

30 

2 

2.0xl0_t* 

2.0X10-1* 

2.0XKT1* 

— 

-

-

-

50 

60 

30 

50 

200 

0.46 

>46 

0.46 

>46 

2.3 

0.93 

0.46 

0.46 

>46 

0.46 

0.46 

1.4x10 

2 0 * 

5'6* 2.2* 

0.46 

>46 

0.46 

0.46 

0.46 

>370 2.0 >46 

>370 

>370 

>370 

>370 

4.2 

>370 

>370 

>370 

>370 

0.63 

>370 

21 

>370 

>370 

>370 

0.32 

0.63 

0.32 

700 

200 

20 

200 

60 

100 

300 

300 

30 

40 

700 

0.50 

50 

30 

10 

3.0xlO-2 

1.0 

9.0 

>46 

>46 

>46 

>46 

0.46 

0.93 

>46 

0.93 

>46 

0.46 

>46 

0.46 

>46 

9.3 

9.3 

0.46 

0.46 

0.46 

0.6 

40 

0.6 

2 

0.6 

0.4 

0.6 

0.4 

2 

2 

2 

40 

20 

0.6 

0.2 

0.2 

40 

1 

0.2 

20 

30 

6 

1 

10 

2 

3 

10 

2 

10 

3 

0.4 

20 

20 

0.6 

0.3 

0.7 

0.3 

0.6 

0.3 

10-2 

2X10"3 

0.4 

2 

0.6 

0.4 

0.5 

0.4 

2x10--

2x10--

2X10--

40 

20 

0.6 

0.2 

0.2 

40 

0.5 

0.2 

0.5 

6 

1 

10 

2 

0.5 

0.9 

2 

0.9 

3 

0.4 

20 

0.5 

0.6 

0.3 

0.7 

3x10"2 

0.5 

0.3 
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rHrHrHrHrHrHrHrHrHr H 

I I I I I I I I I I 

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• • r-l * 

CM CO rH VO 

OV 

v© CM 

CM « • CO rH 

r-l O r-l r-t r-l 

CO. O co 

UO • uo * 

0 0 V© -tf r-t 

CO 

-tf 

r H • 

-tf O 

0 g 

r H CO -tf 

r-t r-l r-t 

r H r H r H 

1 1 1 

C C C 

M I-H |- H 

0 

U O 

r H 

r H 

CJ 

M 

Ov 

00 

r H 

1 

r l 

M 

0 

O CM CO -tf 

Ov Ov Ov Ov 

r H r H r H r H 

lll l 

u u tj u 

M M t-t M 

CM 

-tf 

1 

W 

CO 

-tf 

1 

fern 

oo 

1 

u 

Ni 

fl 

uo 

CO 

1 

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r*i 

m 

00 

1 

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f s 

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brs o 

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1 1 

Cfl fl) 

rJ *J 
TABLE 1 (Continued) 

P.37 

Radionuclide 

QA or Q F

+ 

(TBq) 

QB(TBq) Qc(TBq) Qn or QE* 

(TBq) 

A:(TBq) A2(TBq) 

Lu-172 

Lu-173 

Lu-174m 

Lu-174 

Lu-177 

0.54 

7.8 

16 

7.9 

28 

>370 

>370 

>370 

>370 

>370 

40 

10 

8.0 

4.0 

80 

9.3 

>46 

9.3 

>46 

0.93 

0.5 

8 

20 

8 

30 

0.5 

8 

8 -* 0.9 

Mg-2-2S8 0 0.56 0.21 50 0.46 0.2 0.2 

Mn-52 

Mn-53 

Mn-54 

Mn-56 

0.29 

720 

1.2 

0.59 

>370 

>370 

>370 

0.21 

30 

Unlimited 

30 

600 

2.3 

Unlimited 

>46 

0.46 

0.3 

Unlimited 

1 

0.2 

0.3 

Unlimited 

1 

0.2 

MFP See formulae for mixtures of radionuclides. 

Mo-93 

Mo-99 

93 

6.7 

>370 

0.63 

7.0 

50 

>46 

0.46 

40 

0.6 

7 

0.5 

N-13 0.98 0.63 0.46 0.6 0.5 

Na-22 

Na-24 

Nb-92m 

Nb-93m 

Sb-94 

NT)-95 

Nb-97 

Nd-147 

Nd-149 

Hi-59 

Ni-63 

Ni-65 

Np-235 

Np-236 

Np-237 

Np-239 

Os-185 

0s-191m 

0s-191 

0s-193 

Os-194^ 

P-32 

P-33 

0.46 

0.24 

0.67 

520 

0.64 

1.3 

1.5 

7.1 

2.7 

410 

-

1.8 

140 

7.4 

2.0+ 

5.8 

1.4 

110 

13 

14 

11 

-

-

>370 

0.63 

>370 

>370 

>370 

>370 

0.63 

4.2 

0.63 

>370 

>370 

0.32 

>370 

>370 

>370 

21 

>370 

>370 

>370 

0.63 

0.21 

0.32 

>370 

20 

200 

90 

6.0 

0.60 

40 

3.0x103 

30 

900 

Unlimited 

30 

600 

50 

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TABLE 1 (Continued) 

P.41 

Radionuclide QA or Q ?

+ 

(TBq) 

Q3(TBq) Qc(TBq) Qn or QE* 

(Tbq) 

A!(TBq) A2(TBq) 

Te-118*' 

Te-121m 

Te-121 

Te-123m 

Te-I25m 

Te-127m0 

Te-127 

Te-I29m0 

Te-129 

Te-131m 

Te-132^ 

Th-227^ 

Th-228i9 

Th-229 

Th-230 

Th-231 

Th-232 

Th-234*1 

Th(oat) 

Ti-44* 

Tl-200 

Tl-201 

Tl-202 

Tl-204 

Tm-167 

Tm-168 

Tm-170 

Tm-171 

U-230 

U-232 

U-233 

U-234 

U-235 

U-236 

U-238 

U-(nat) 

U(enr<5%) 

U(enr>5%) 

U(depl) 

V-48 

V-49 

1.2 

4.6 

1.7 

6.8 

28 

89 

210 

17 

17 

0.73 

0.45 

9.4 

0.82 

0.30+ 

2.0+ 

39 

0.40+ 

88 

0.60 

0.47 

0.78 

11 

2.1 

880 

6.9 

0.80 

180 

1.5xl03 

100+ 

3.0+ 

10+ 

10+ 

6.5 

10+ 

20+ 

13+ 

ll+ 

10+ 

19+ 

0.34 

l.lxlO3 

0.21 

>370 

>370 

>370 

>370 

21 

21 

0.63 

0.63 

21 

0.63 

>370 

0.32 

>370 

>370 

>370 

>370 

0.21 

4.2 

0.63 

>370 

>370 

>370 

4.2 

>370 

>370 

4.2 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

>370 

20 

7.0 

100 

8.0 

20 

9.0 

600 

9.0 

2.0xl03 

20 

7.0 

l.OxlO"2 

4.0xl0-1* 

3.0xlO-5 

2.0xl0-1» 

200 

Unlimited 

6.0 

Unlimited 

0.20 

400 

800 

200 

80 

70 

20 

8.0 

10 

l.OxlO-2 

3.0xl0-1* 

l.OxlO-3 

l.OxlO"3 

Unlimited 

l.OxlO-3 

Unlimited 

Unlimited 

Unlimited 

l.OxlO-3 

Unlimited 

20 

700 

0.46 

>46 

>46 

>46 

9.3 

0.46 

0.46 

0.46 

0.46 

0.46 

0.46 

>46 

0.46 

9.3 

>46 

0.93 

Unlimited 

0.46 

Unlimited 

0.46 

>46 

>46 

>46 

0.46 

9.3 

>46 

0.46 

>46 

>46 

>46 

Unlimited 

Unlimited 

Unlimited 

Unlimited 

Unlimited 

Unlimited 

Unlimited 

Unlimited 

Unlimited 

0.93 

>46 

0.2 

5 

2 

7 

30 

20 

20 

0.6 

0.6 

0.7 

0.4 

9 

0.3 

0.3 

2 

40 

Unlimited 

0.2 

Unlimited 

• 0.5 

0.8 

10 

2 

4 

7 

0.8 

4 

40 

40 

3 

10 

10 

Unlimited 

10 

Unlimited 

Unlimited 

Unlimited 

10 

Unlimited 

0.3 

40 

0.2 

5 

2 

7 

9 

0.5 

0.5 

0.5 

0.5 

0.5 

0.4 

IO- 2 

4xl0-1* 

3xl0-5 

2xl0-1+ 

0.9 

Unlimited 

0.2 

Unlimited 

0.2 

0.8 

10 

2 

0.5 

7 

0.8 

0.5 

10 

IO- 2 

3xl0-^ 

IO-3 

IO-3 

Unlimited 

10"

 3 

Unlimited 

Unlimited 

Unlimited 

IO-3 

Unlimited 

0.3 

40 
P.42 

TABLE I (Continued) 

Radionuclid e 

QA or QF

+ 

(TBq) 

QB(TBq) Qc(TBq) Qn or qE* 

(TBq) 

A-/TBq) A2(TBq) 

W-17810 

W-181 

W-185 

W-187 

w-iss10 

Xe-127 

Xe-I31m 

Xe-133 

Xe-135 

Y-87 

Y-88 

Y-90 

Y-91m 

Y-91 

Y-92 

Y-93 

Yb-169 

Yb-175 

Zn-65 

Zn-69m0 

Zn-69 

Zr-88 

Z rr 9 3 

Zr-95 

Zr-97 

0.98 

25 

1.8X101* 

2.1 

17 

3.6 

50 

22 

4.0 

2.2 

0.37 

5.9xl0 5 

1.9 

280 

4.0 

11 

3.2 

25 

1.7 

2.4 

1.7xl0 5 

2.5 

-

1.4 

1.5 

>370 

>370 

>370 

4.2 

0.21 

>370 

>370 

>370 

4.2 

>370 

>370 

0.21 

>370 

0.32 

0.21 

0.21 

>370 

>370 

>370 

4.2 

4.2 

>370 

>370 

>370 

0.32 

700 

l.OxlO3 

Unlimited 

300 

50 

— 

-

-

-

100 

9.0 

20 

6.0xl0 3 

4.0 

300 

90 

30 

100 

10 

300 

5.0xl0 3 

8.0 

0.20 

5.0 

50 

2.3 

>46 

0.93 

0.46 

0.46 

o n * 

28 56* 

K 11 

>46 

>46 

0.46 

>46 

0.46 

0.46 

0.46 

9.3 

0.93 

>46 

0.46 ' 

0.46 

>46 

Unlimited 

0.93 

0.46 

1 

30 

40 

2 

0.2 

4 

40 

20 

4 

2 

0.4 

0.2 

2 

0.3, 

0.2 

0.2 

3 

30 

2 

2 

4 

3 

40 

1 

0.3 

1 

30 

0.9 

0.5 

0.2 

4 

40 

20 

4 

2 

0.4 

0.2 

2 

0 . 3 

0.2 

0.2 

3 

0.9 

2 

0.5 

0 . 5 

3 

0.2 

0.9 

0.3 

Qr, tabulated in place of Q-. 

Qp tabulated in place of QAA-^ and/or A2 value limited by daughter product decay. 

q. limited by spontaneous fission. 
P.43 

'•^v, 

QA 

, P 

*i 

Q B 

Qi ID 

FIG. .1. Schematic Representation 

of Exposure Pathways 

Employed in the 

Q System. 

Q. 
P.44 

U 

O 

LL. 

C 

0-1 VO 10 

Max. Beta Energy (MeV) 

FIG. 2. Shielding Factor as a Function of Beta 

Energy [191. 
P.45 

103 

o 

c 

> l 

2 -* 

*-* 10 

a 

tn 

O 

Q 

CD - 5 

10 

-i—i—r i i i i 1—i—I I i i i 

Beta dose rate at 1m from 

a point source of 37 MBq 

r~_5****< J 

4*x 

• ' i 'ti n ' I I t I ! I I 

0.1 1.0 10 

Max. Beta Energy (MeV) 

FIG, 3. Variation of External Beta Dose Rate 

to Skin with Energy [20]. 
P.46 

10' 

1 1— i : i ; t ; | 1 1—i i ! ; i 

Beta dose rate at depth of 

Z 5-10mg.cm-2 in tissue for 

- plane source of 0.37 TBq. m"2. -

in 

in 

r 10' 

a 

cc 

a> 

g 101 

CD 

10v 

1JL- - -

i , i • i 111 1 • • ''' ' 

0-1 1-0 10 

Max. Beta Energy (MeV) 

FIG. 4. Variation of Beta Dose Rate Due to 

Skin Contamination with Energy [22]. 
P.47 

3 a. 

3 

< 

G 

3 Q_ 

& <3 a> 5< 

0) 

3 cr o 

o at 

to o at 

O U 

o 

3 tn 

tn 

O 

m 

g 

a 

pvl 

< CM 

a 

1 

JS 

___3 

_^___ . S 

__^__£ 

K ^____a 

fc_ 

s 

^ 

_ ^ 

_s 

l\\^\\\\\\\\v^ 

O 

•*< 

a i_ <" 

^m 

s 

J L 

y^NSWvl S 

H_- _ ^ 

1 

KY\YY\\V_ 

S 

KVvYWvNXVN , _S j _L 

in 

tn 

b 6 

o!;Da V/ D 

to -C 

•*- » 

c tn 

3 "O <D uat _ i 

o > UJ tn 

•«- » Wl - J tn 

«»-» 

c 

*»• 

C 

o 

o 

(1) ai 

a 

._* u Qa 

. 

< 0; 

o 

l_ 

R tn 

• c 

a 

i-

< UJ 

< CO a> 0) 

-_J 

c 

• ^ 

to in 

o _c 1— 

JZ 

•+* 

i 

'In a> _D "a 

> 

^ tn lue DA c* ^ ! J^ < in 

c 

o 

3 cn 51 

crl 

c 

o cn o. Q. E 

o 

o 

LO' 

o 

£ 

to In 

>* CM 
P.48 

DISTRIBUTION TPRD/B/0340/R83 

Dr E 

Dr D 

Dr J 

Mr D 

Mr M 

Mr H 

Mr H 

Dr 

Mr 

G 

J 

Mr 

Dr 

Dr T Broom 

Acting Laboratory Manager 

Dr P F Chester 

W Carpenter 

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Oates 

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G 

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Health & Safety Department 

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P.49 

TPRD/B/O340/R83 

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J 

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N 

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P 

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F 

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W 

A 

P 

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D 

M 

G 

D 

K 

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B 

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P.50 

TPRD/B/0340/R83 

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P 

G 

S 

A 

I 

P 

R 

R 

D 

R 

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P.51 

TPRD/B/0340/R83 

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P.52 

REPORT SUMMARY SHEET 

Central Electricity Generating Board 

Berkeley Nuclear Laboratories 

TPRD/B/0340/R83 

Job No. XB255 

The Q System for the Calculation of A and A- Values 

within the IAEA Regulations for the Safe Transport 

of Radioactive Materials 

- b y -

H.F. Macdonald and E.P. Goldfinch* 

September 1983 

SUMMARY 

This report contains a comprehensive description of the Q system, an 

improved method for evaluating the Type A package contents limit A^ and A2 

values within the IAEA Regulations for the Safe Transport of Radioactive 

Materials. The work reported here was performed in collaboration with the 

Board's Health and Safety Department under the terms of Research Agreement 

No. 2291/CF, which forms part of the IAEA Coordinated Reseach Programme on 

the Safe Transport of Radioactive Materials. The Q system was adopted by the 

IAEA in March 1982 for the calculation of A-| and A2 values to be published 

in the 1984 revised edition of the Transport Regulations. This present 

report was prepared at the request of the Agency for publication at the same 

time as the revised regulations later next year. For this reason the format 

used is somewhat different from that of standard TPRD reports, following 

instead that used by the Agency in its publications. 

The Q system represents a development and extension of the earlier 

A-|/A2 system. It incorporates the latest recommendations of the ICRP and, 

by explicitly identifying the dosimetric considerations underlying the 

derivation of Type A package contents limits, provides a firmer and more 

defensible basis for the IAEA Transport Regulations. Overall the use of the 

Q system introduces relatively minor variations in the A^ and A2 values 

compared with those in the 1973 Edition of the Regulations. Where major 

variations occur, say by a factor of greater than 2-3, these are generally 

associated with reductions in the Type A package contents limits for beta 

emitters determined by dosimetric pathways not previously considered. 

Isolated relaxations in the contents limits also occur due to changes in 

dosimetric data introduced in ICRP Publication 30 and more generally in the 

case of special form alpha emitters due to the revised definition of the Ai 

value for such materials. 

Contents 

Approved 

Date: 

Classification and Distribution 

Approved X^l", _ 

Date: /