ML20085H661
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| Issue date: | 10/10/1991 |
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Commission of tha European Communitics nuclear science and technology .
Assessment of management modes l for graphite from reactor decommissioning l
1.F. White, G.M. Srnith '
L.J. Saunders '
C.J. Kaye, T.J. Martin 2 G.H. Clarke
- l M.W. Wakerley *
' National Radiological Protection Board, Chilton, Didcot, UK
' Central Electricity Generating Board. Barr, wood UK
' Central Electric!ty Generating Board Maacheater. UK
' United Kingdom Atomic Energy Authorr.y. Aleley, UK
' Orltish Nuclear Fuels Limited. Sellafiel'.f. UK l
This work was performed in the framework of the European Communities research programme on decommisaloning of nucler.r power plants, project No .4 ' Treatment of specific waste materials' Contract No DE D-001-UK
-y Directorate-General Science. Research and Development 1984 EUR 9232 a y,y02$$$$[c $$$h,f(7 4 P ,
l 1
i CONTENTS Pepe Nn.
- 1. INTRODUCTION I i ARI LNG AND COMPOSITION OF IRRADIATED CRAPMITE 4 2.1 Introduction 4 2.2 Activation of Graphite and tapurities 4 2.2.1 Elemental concentrations 4 2.2.2 Production and decay routes and reection cross-sections 5 2.2.2 Neutron flux levels 5 2.2.4 Reactor invwi.' alculations 5 2.2.5 Magnon programs. .ventory 6 2.2.6 Resulte 6 2.3 Fission of 'Jrsnium Ispurity 1 2.4 Circuit Contamination 7 2.5 Major y Enitters 8
- 3. PACKAGING FOR SEA DISPOSAL 24 3.1 General Requirements 24 3.2 Disposal in Drum Containers 24 3.2.1 Rmference Magnox reactor 25 3.2.2 Reference AGE 26 3.3 Disposal in Shielded Overpacks 26 3.3.1 General requirements - 26 3.3.2 Reference Magnox reactor 27 3.3.3 Reference AGE 27 3.4 Disposal in Large Concrete Containers 28 3.4.1 General requirements 28 3.4.2 Reference Magnox reactor 30 3.4.3 Reference AGE 31 3.5 Conclusiano 31 '
4 PACKACING OFTIONS FOR LAND DISPOSAL 35 l- 5. INCINERATION 36 l
5.1 Combustion 36 5.2 Treatment of Furnace Of f-Geses and Recovered Ash 37 5.3 Trausport/Disyvaal Requirements f or the Immobilised Ash 38 I 5.4 Cos t s 38
- 6. LEACHING OF IRRADIATED GRAPHITE 42 l
l 6.1 Introduction 42 l
l l
t CONTENTS (con Page No.
6.2 Empe riment si 42 6.2.4 Pressure testa 42 6.2.2 Leaching teste 42 6.3 Kabults 43 6.3.1 Pressure tests 43 6.3.2 Les;hing teste 43 6.4 Summa ry 44 6.3 !a911catfons for Radiological Aesessments 65 7 KADIOLOGICAL ASSESgMENT 62 7.1 Int roduction 62 7.2 Options Considered 62 1.3 tadiologiaal Concepts 62 7.3.1 Decision-making in ra' inactive waste management 62 7.3.2 Requirements for radiological assessments 64 7.4 General Features of the Models Used 65 7.5 Clobal Circulation 6e 7.5.1 Latroduction 66 7.5.2 Methodologies and modele 67 7.5.3 tasulte 68 7.6 Disposal on the Deep Ocean led 68 7.6.1 Methodologies and models 68 7.6.2 Raruits 70 7.7 Dispossi iu a Deep Geologic Repository 71 7.7.1 Methodologies and models 71 7.7.2 Result s 72 7.8 Shallow Land lurial 73 7.8.1 Methodology and models 73 7.8.2 Lasulta 76 7.9 Lnciseration 78 7.9.1 Methodology and models 78 7.9.2 Leeults for atmospheric discharges 81 7.9.3 Disposal of solidified ash arising f rom incineration 81 7.9.4 Total radiological impact of incineration 62 7.10 Effects of the Presence of 0.1 ppa Natural Uranium in the Graphite 82 7.10.1 Introduction 82
- _ - _ _ - - . - _ - _ _ _ _ - - - - _ - - _ - - _ - . . . _ . - - - . . - - - - . - - - - - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - -
_ -- ^-
CONTENTS (contd )
Page-No.
7.10.2 Deep ocean bed disposal b7 7.10.3 Deep geologic disposal R3 7.10.4 Thallow land - burial 63 7.10.3 Incineration F.
7.10.6 Concluelone 84 7.!! Susmaries of Results for Magnox Reactor. Magnox Programme and AGR 84 7.11.1 Introduction 84 7.!!.2 Individual doses 8) 7.11.3 Collective doses 85
- 8. DISCUSSION: CH0!G AND ACGPTABILITT OF MANACDENT OPTIONS !!)
8.1 Fossibility and Coste !!3 8.2 Radiological Impact !!6 8.2.1 Individual doses 114 8.2.2 Collective doses !!4 8.3 Uncertainties !!5
- 9. CONCLUSIONS !!S 4
e
- l. INTRODUCTION The decommiseloning of nuclear f acilities involves the generation and management of radioactive wastes. As de fined by t he I AEA II and by common usage. Stage _1 of decommissioning involves the removal of the last fuel charge and the coolant, Stage 2 the dismantling of structures peripheral to the reactor core and its containment, and $tage 3 the dienentling of the core and containment themselves. Most o. the radioactive waste sanagement 4ctivities related to Stage 1 may be regarded as extensions of stallar activities that took place Phile the reactor 'fac operating. In nome reactor syntese. Stage 2 of decommiseloning may involve the generation of rew typea of wastes f rom decontamination operations and f rca the reenval and perhape dismant!!cs of contaminated plant. However, it is not until Stage 3 that the more radioactive decommiseloning wastes will arise. These will include activated and contaminated asterials such as steel and concretei and for some reactor systese, grapnite.
The principal Western staphite-soderated power reactors are the Prttlah Magnon and the stallar French systes, both of which use metallic natural uranium fuel clad in magnesius alloy, and the gritish Advanced Cee-Cooled Reactor ( ACR) fuelled by enriched uranium oxide clad in stainless steel.
All these reactore are cooled by carbon dioxide gas under pressure.
This study examines the management options for irradiated and ponethly contaminated graphite that will arise as waste from Stage 3 of the decomatheter.ing of Magnox reactors and ACRs. In order to do so, the quantittee of graphite arising are estimated (section 2), together with the associated inventories of radionuclides. There are seversi dif f erent designs of both Magnos reactors and ACRa, so for each systes a representative " reference" reactor has been defined for the purposes of thia study.- Detailed calculttione are then carried out to predict the radio-nuclide inventory of the graphite in the reference reactors et the end of their assumed 40 year lifetime. The earliest time for which inventories are calculated is 10 y after shutdawn of the reactor, for it le unlikely that Stage 3 of decommiseloning would be reached wuch sooner.
Throughout the study, results are presented for the reference Maanns reactor, for the reference ACR, and for all the UK Magnon reactorea 18 twin-reactor stations owned by the Central Electricity Generating Board (CECn) and the South of Scotland Electricity Board plus 2 quadruple-reactor stacions owned by British Nuclear Fuels Ltd (BNTL).
Even after years of irradiation, staphite retaine most of the good mechanical properties which allowed it to be used as a structural material i
1-e
~ . .
I l
'for reactor cores, and it le relatively insoluble and not otherwtee particularly.cheetcally reactive. In other words trradiated graphite appears to fulfil acet of the general requiremente for a solid radionettve vaste tuttable f or disposal. Most of the canagement optione essained in thin study do not therefore call for any form of chemical processina, and only involve euttable packaging of the graphite before disposal by the chonen route. rackaging options for see disposal are examined in sectinn ),
and those for land burial in section 4.
As a contrast. section S deals with the option of incinerating the graphite in a purpose-designed f acility, so that the bulk of the graphite la disposed of to atmosphere se carbon d10mide. IbC would follnw the same route. together' with tritium as water vapour and small quantitles of tha less volatile radionuclidee in ontde fore which remain af ter filtrat tnn of the incinerator.off gases. The rest of *he non-volatile radtosctivity vnutd be retained in a such reduced Volume of eah, which to then assumed tn he dispaced of by one of the options already considered for untreated l graphite.
Section 6 represents the results of a serive of teaching testa on saarles of irradiated graphite. A variety of teachente were used under eleulated conditions of disp 0eal, including staulated seawater at the reduced temperature saJ high pressure encountered on the deep ocean bed.
More extensive teste of the mechanical behaviour and uptake of seawater at high pressures were carried out using inactive graphite.
Section 7 is a radiological adessement of the sansgeset.t options for graphite identified in sections 3-5 and section 8 is an overview of both the technological ~and radiological aspects of graphite managoesnt. Finnity.
the conclusions of the study are given in section 9.
The inclueton of a management option in this study, and the choicem nf particular persaeter values such as the assuesd reactor lifetime or the elapsed time berore stage 3 of decommiseloning, in no way reprenect pottev decisions on rest $r decommissioning.or radioactive waste mannaement by any of the organtauti .a concerned.
Costs have been estimated for three packaging options for asa disponal, and ateo for. incineration. The primary als in presenting the estimaten.lg to illustratu Sow the cost of each separate packaging / disposal option la af f ected by alloping time for radioactive decay. The settmates cannot he used to make valid intercomparisons, because they have been prepared by r different organisations whose procedures and assumptions are not tne name, l-i .
_.~__ . . ~ . _ _ . _ . _ .-. __- . . _ . _ . . . - .. .___. m.. _ .. . _ ..m..__
- med ' because the estimate for current see-dumping practices le historical while all 'h = re6t are speculative.
References for Section t-l
- 1. Internettonel Atomic Energy Agency. Factore relevant to the
' decoentestoning of lend *beced nuclear reactor plante. Safety Set t en No. 32 Vienne, 1AAA (1980).-
s d
' i
+ - < w
- 2. ARIS!NCS AND COMPOSITION OF IRRADIATED CRAPHITE 21 Introduction In order that assessmente may be made of the packaging requireeents for graphire f rom Magnoa and ACR stations (section 3 and 4) and of the radiological impact of the various methods of disposal (section 7), the quantittee of the various radionuclides contained in irradiated graphite must be estimated. This section describes the calculations and data used in the estimations and presents the results.
Three contributione to the total inventory of radionuclides in the graphite have been considered. These are activation of carbon and of stable .
tapurities in the graphite, fission of natural uranium present in the graphite se an impurity, and the possibility of surf ace contestnation f rom elsewhere in the reactor. Each of these contributions has been calculated or estimated for a reference Magnos reactor, a ref erence ACR, and also f or the entire UK Kag'nos decommissioning programme of 16 Generating Board reactors plus 8 owned by gNFL.
2.2 Activation of Craphite a,' Impurities The calculation of the activity level in the graphite requires the following information to be specified, separately f or Magnon reactors and AGRs.
(a) The elemental concentrations of impurities in che graphite -
(b) the production and decay routes of the leportant tootoposi (c) the reaction crose-sectional and (d) the neutron flus levele in the graphite.
In' addition a scheme is required to integrate the results of point emicula-tions over a whole reactor core, and to extrapolate the results . f rom one reactor to another in the case of the Magnon deconstesioning programme.
2.2.1 Elenestal concentrations Available data were reviewed and reference impurity concentrations
- specified for the purposes of this study. The main sources of data are chemical analyses of unitradiated graphite, though some ispurity levele have been deduced from measurements on irradiated graphite samples with known irradiation histories.
A notable feature of the impurity concentrations is the large variation from saap1' to sample. The most appropriate reference impurity levels for use in this study are the best estimates of the mean values.-as these will quite accurately define the total activity to be disposed of when large masses of graphite are involved. When drawing up specific rather than O
genetic decommissioning plans these estimates would be supplemented by teasurements on the reactor in question.
Various grades of graphite are used in Magnox and AGR cores. but the dif f erence between grades f or the same reactor type is generally no large r than the difference within grades. Therefore only two sets of impurity data have been specified, one f or Magnox and one f or ACR. The ref erence dat a ar e given in Table 2.1, 2.2.2 Production and decay routes and resetion cross-sections developed by the For the activation calculations the FISPIN code United Kingdom Atomic Energy Authority (UKAEA) was used with a three group cross-section library supplitd by BNFL (Sellafield). For details of the Liber *". application should be made to that organisation. Neither the code nc e t. . library has yet been validated or f traally endorsed f or use in this context, chough the results are generally consistent with those of an earlier study (P. Woollen. CEGB internal report) af ter making allowances f or dif f erences in input data.
2.2.3 Neutron flux levels Two group flux data f or the active core region of the ref erence Magnon reactor. Dungeness A were supplied by the South East Ragion of the CECB.
They refer to a core churmal power of 730 MW. Flux data f or the reflector regions were derived f rom data used f or the previous inventory assessment.
The fluxes used f or the activation calculations, including an answmed station load f actor of 0.7. are shown in Figures 2.1 and 2.2.
Fluxes for the moderator and reflector regions of the reference AGR, Hartlepool, sere supplied from 3-D core calculations (E.A. Rill. CECB internal report). They are summarised in Figures 2.3 and 2.4. again in:1uding an assumed station load f actor of 0.7.
2.2.4 Reactor inventory calculations Curves of specific setivity of each isotope as a f unctit n of enermal and epithermal flux were produced, covering the ranges encountered in the reactor moderator and reflect'or. In all cases the f ast flax was set at 80%
of the epithermal flux. The reactor core was then divides into elements Th e assuming astauthal symmetry. and the fluxes defined f or each elemen;.
specific activity of each isotope in each element was interpolated and the spect.tc activities summed over the whole core to obtain the total activity.
No loss of I"C to the carbon dioxide collant was assumed.
This process was carried out f or the Magnon and AGR fluxes given in those Figures 2.1 - 2.4. and repeated for Magnox fluxes 15% above and below in given in Figures 2.1 and 2.2 so that the inventories f or other reactore the Magnox programme couth be estimated by interpolatiQn.
-5
2.2.5 Maanos programme inventory In order to derive the inventory of activation produ:ts for the whole l' Magnox decouaissioning programme, it was assumed that all reactors are stallar to the ref erence reactor, both geone trically and regarding the flux shapes. For the core graphite, therefore, the inventory was taken to be proportional to mass, and the flux dependence was given by the three flux estimates made for the reference reactor, assuming flux to be proportionel to mean fuel rating.
For the reflector regions the flux dependence was derived as above, but it was not necessary to taka account of the variation between reactors of reflector thickness because the majority of the activity is produced wi- hin the innermost 0,5 s of reflector material. 152Eu is exceptional in that most of this nuclide is formed in the outer reflector regions. The mesh used for the inventory integration was not considered fine enough f or an accurate estimate of 152Eu distribution, and in any case the extrapolation to reactoie with different reflector thicknesses would not have been possible.
For this reason the specific activity of 152Eu in the entire mass of reflector graphite beyond the first 0.5 m has been assumed to be the maximum predicted to be achievable anywhere within the reflector sone. This gives an upper estimate of the total 152E u inventory.
2.2.6 Results All the calculations were initially carried out for a reactor tife of 40 y at the assumed thermal powers and load f actor, f ollowed by a decay period of 10 y after shutdown. Inventories at later times, including the post-disposal period, were extrapolated separately f rom these results.
Specific activity es a function of fluz and energy Large variations were observed between the' dependecces of the inventories of dif ferent radienuclides a.'ter 40 y on both neutron flux and energy distribution. Many l
of the inventories had a linear flus dependence, og as shown f or A*C'in
! Figure 2.5, thout at some wire more sensitive to the neutron energy distribution that,others.
An importart effect revealed by these detailed calculations is that of
" burn-out", in which the inventory at first increases in the normal way owing to activation of the stable precursor, and is then reduced by dce, +
l as the stable precursor is depleted. This means that simple multiplications of flux, time and cross-section (which may be accurate enough for estimating inventories of some radionuclides in fuel) can be totally misleading in i
decommissioning studies since the timescale is an order of magnitude longer.
For exampl3, Figure 2.6 shows that tritium is subject to burn-out; even more
so ara the europium isotopes, especially 152gu (Figure 2.7). As n(ted abos., af ter 40 y most of thle nucitda is uupected to be found in the outer reflector regions since that f orseo earlier in the inner regioris will have been burnt out in the higher fluxes.
Reference core inventottes Table 2.2 shows the calculated isotopic invaritsry of graphite in the ref erence Magnos reactor, and Tabla 2.3 gives the total inventory for all UK Magnos reactors.
Table 2.4 shows the calcul sted teotopic inventory of a single ref erence ACR.
2.3 Fission of Uranius Inpurity Besides the etable impurities whose activation contributes to the radionuclida inventories calculated . bove, reactor graphite may contain traces or natural uranium which would undergo fission. There is no product specification n.. this respect, and neutron activation analysis of a single sample of Magnon graphite indicated only that the level of natural ursuium was below 0.1 opa (F. A. Fry, NRFB, private communicat'.on).
In order to estimate the effect of the presence of natural uranium on the radiological impset of disposal of the irradiatad graphite, a level of 0.1 ppa was assumed. Simplified calculations f or .isgnon and f or iGR were carried out using the CECB inventory code RICE ( assuming a thermal flux averaged across the whole core and corrected f or load f actor, and taking the ratica of test and epithersal fluses together with all other nuclear data f rom the internal libraries of RICE. 40 } reactor lif e and a 10 y initial decay period were assumed, as f or the more couples activation calculatione.
The results are given in Tables 2.5 and 2.6. The simplified calculati no are justified by the finding in section 7.10 that the predicted radiological impact of disposal of these quantities ut activity is generally una11 compared with that of disposal of the activation products.
2.4 Circuit Contamination It is possible that contamination of the reactor cooling circuit sight contribute to the radionuclide inventory of the graphite. Poten:ial source s of contasination ares uranium present in the circuit f rom f uel el.mente externally contanit sted during manuf acturet spalled oxidse trom f uel cladding; apalled oxides f rom mild steel circuit components; and volatile fission products f ree def ective fuel. These sources are discussed in turn.
According to the fuel specification the external contamination of a Ma gno n f uel element may not ascoed 10 ug U, so the total uranium in the l
1 e
reactor f rom this source could not exceed about 1 g at the end of lif e; the corresponding figer f or an ACR is about 7 g. These levels are negligible compared with the caount of uranium possibly present in the graphite itself (section 2.3).
Few data are available on the quantities of spalled oxides f rom f uel elements but estimates by CECB suggest that the quantities of Fe. Co and Be present in the Magnox circuit f rom this source will be of the order of grams, much less than is present as an impurity in the graphite; the same applies to oxides f rom mild steel circuit components. There is less operating experience for AGR4 than f or Magnox, so the situation regarding cont amination is less clear. Spalled oxides f ree the stainless steel f uel cans are an additional potential source of contamination, and preliminary indications are that this source together with citcuit steel oxides may make a significant contribution to the total radionuclide inventory of ACR graphite; however, that would not necessarily also be a significant contribution to the total radiological impact of disposal.
Measursaa.nts of surf ace contamination by 137Cs have been made on fuel element compenents f rom one Kagnox reactor (CEGB. internal report), and imply that about 1012 Sq of this radionuclide might be present as a surface contaminant on core graphite at the end of the reactor lif e. It would no t be correct to generalise f rom a single aussurement, f or the levels of contamination f rom f ailed fuel will depend on the design of the reactor and of the fuel element (the Magnox f amily of reactors includes several quite dif ferent designs) and the operating history of the reactor in question.
However it appears that the quantity of I37Cs on the surfaces of core graphite may be similar to that present throughout the graphite due to fission of the uranium tapurity (Sactica 2.3). Fission product contamination of graphite in ACRs is currently low.
2.$ Major v Esitters In order to estimate the packaging requirements f or irradiated graphite or incinerator ash (sections 3 and 5) only the inventories of the major T-emitting nuclides 50Co. 152Eu and 15"Eu need be considered. Howe ve r, further analysis shows that europium isotopes can also be neglected for the following reasons.
In the Magnox case the europium isotope specific actirities are strcngly dependent on flux, so their relative importances vary according to location in the reactor. In the core graphite, the euLopium isotopes neither contribute significantiy to the y dose at 10 y decay nor do they signif t-
-3 D
cently af fect the_ time at which a package will no_ longer require shielding.
In reflector graphite, II"Eu contributes less than 201 to the unshielded y dose level at:10 y decay. Snd increases the period f or which the container requires shielding by no more than 4 y. - The contribution of IS2Eu is even -
1ess significant, so for the purposes of sections 3 and 5 of the report only 80 Co need be considered. 1he calculated 60Co specific' activities f or the various core regions of the reference Magnon reactor are give'n la Table 2.7 together with notes on their derivation.
For the AGR, the major y-emitting isotopes are again 60co, IS2Eu and L S*Eu.- and their inventories have been estimated on the basis of- the ,
ref erence-impurity levels f or ACR graphite. Because the cobalt content of -l AGR graphite is higher than that of Magnon graphite the europium teotopes can be neglected in the ACR case also. The estimated 60Co specific activir.ies for the various core regions are given in Table 2.8.
References for Section 2
- 1. Burstall. R F. FISFI:1 - a computer code f or nuclide inventory calculations. UKAEA. Risley. ND-R-328(R) (1979).
- 2. Nair. S. RICE - a reactor inventory code for calculating actinide and fission product arisings using a point source model. CEGB. RD/B/N6138 (1977).
S
Table 2 1 Reference Crashite lemurity Levels (ppa)-
Element Magnos ACE Li- 0.05 0.05 to' O.02 0.02 5 n.1 0.$
N . 10 la 1.0 4.0 Ps 0.; 0.4 A1 1.0 4.0 31 35 35 S 50 60 C1 2.0 4.0 Ca- 35 25 Ti 3 0.7 Y- 12 0.4 Cr. 0.35 0.4 Ma 0.04 0.25 Fe 10 28 Co 0.02 0.70 NL 1.0 6.0 En 0.13 1.0 Sr 0.4 0.4 No 0.1 2.5 Ag 0.001- 0.001 !
Cd 0.04 0.07 la 0.05 '0.06 Sa 0.05 1.0 la 1.5- 0.5 Se -- 0.04 0.05 tu- 0.004 0.00$
Gd 0.005 0.01 Dy 0.004 0.006 W 0.12- 0.15-Pb 0.12 0.8 31 0.08 0.05 4
I 4
Table 2.2 Activation Inventory (le) of Reference Magnos Reactor Af ter 40 years reactor operation followed by 10 years decay 3R 1.2 101 " '3Ho 8.5 los 1%e 7.1 1010 93%b 5.5 los I"C 8.5 1013 9"Nb 1.0 105 3'C1 9.5 1011 5'Tc 1.7 10e utCa 7.3 1011 10e*As 2.3 1010 5"Mn 2.7 los 113*Cd 1.0 1010 55Fe 1.5 1013 12185n 4.5 1010 styg 9,3 1o10 133ba 5.6 1011 60Co 2.7 1013 152Eu 2.2 1011 63Ni 1.3 1013 15"Eu 5.2 1012 651n 2.1 los 155Eu 1.6 1012
+
Total mass of graphito = 2233 t Table 2.3 ,
_y Activation Inventory (Eq) for t!K Magnox Programme 1 (9 twin reactor statione plue 2 quadruple resetor stations)
Af ter 40 years reactor operation followed by 10 years decay 1.
3M 2.6 1015 93y, t,9 1010 1%e l'.6 1012 _93%b 1.2 1010 I"C 1.9 1015 '"Nb 2.2 1p6 3'C1 2.1 1013 Tc 3.9 l'd htCa 1.6 1013 10 seas 5.3 1011 5"Mn 6.0 109 113*Cd -2.2 1011 55Fe 3.5 10 61 121*Sa 1.0 1012 5%i 2.1 1012 1335a 1.3 1013 80Co 5.9 10 14 152Eu 4.9 10 l 63Ni 2.9 1014 15"Eu 1.1 10 "
1 l
652n 4.8 109 155Eu 3.5 1013 l Total mass of graphite - 50,000 t 11 -
e
Table 2 4 Activation Inventory (Bq) of Ref erenc e ACR Af ter 40 years reactor operation followed by 10 years de:ay pai
'3 Mn ES 3H 7.6 1013 5.6 1010 q ID3 a 2.3 1011 '3*Nb 3.4 1010 ,,
=
I"C 149 1016 '"Nb 8.6 106 3 Eri 2.3 1012 ' 'Tc 3.6 10' "i Ca 1.2 1012 10esAs 3.6 1010 5"Ma 5.2 10' Il3*Cd 4.6 1010 557 9,3 1013 121 w 1,3 1012 ,
- 59yg 3,5 goll 1333 ,. 3,. 1011 60Co 9.6 101 " ***Eu 1.0 1011 LIN1 1.1 101 " 15"Eu 2.9 1042 652n 3.7 109 156tu 9.41011 w.
Total sans of artphite = 1633 t
l 1
Table 2.5 Inventory (sq) f ree irradiation of 0.1 pre natural uranium in referer.co Magnon reactor graphitt Af ter 40 years te.ctor operation followed by 10 years decay "Se 1.80 106 152tu 3.19 107 10$r/'4Y 1.36 1011 156tu 1.87 1010 elKr 6.44 109 Illtu 6.59 10'
'3Z r 1.30 107 13'u 5.33 105
'IaNb 1.03 107 2360 2.68 106 "Tc 6.67 107 tt? Np 8.26 105 106Ru/106Rh 4.19 los 21eru 3.53 1010 107Pd 1.04 106 219Pu 1.09 10' IIIRCd 6.41 107 2kOFu 2.2b 10' 12$$b 7.11 100 24t Pu 3.30 1011 125eTo 1.74 tot th2Pu 3.09 10 7 1291 1.87 105 261 Aa 1.18 1010 13"C s 6.96 109 253As 3.22 los 1350s 5.74 106 2""Cm 2211010 137Cs 2.99 1011 245Cs 3.24 106 1""Ce 4.96 107 266Ca 1.62 106 1265n 3.70 106 1"7Pm 1.26 1010 1515m S.56 10'
___-._____._____________.__._______m___ _
Table'e.1, investory (le) free irradiation of 0.1 pre natural vranium in I reference $ 1 nraphite !
Af ter 40 years reactor operation followed by 10 years decey .
MSe 2.98 106 ist tu 2.33 107 to$r/80Y 2.0$ 10ll II"tu 3.34 1010 85Kr 1.0$ 1010 lintu 1.50 1010 ;
'32 r 2.07 107 236U 3.29 lol [
' 88!tb 1.37 197 238g g ,g, to6 [
MTe 9.04 107 23?Np 1.49 107 106tu/106th 4.32 los llePu 4.07 1010 ,
107Pd 219 los 23'Pu 1.09 10' 1338C4 1.37 tot thoru 2.13 10' 125sb 137 tot thiru 4.07 1011 12Sete 3.34 los 262Pu S.81 107-IMI 3.31 105 26tAs 1.14 1010 t
li"Ce 2.30 1010 263 A g,o, got i 1350s 8.44 106 2""Ce 3.47 1011 137Ce 5.52 1911 265Cs 2.40 107 16"Ce 9.48 107 266Ce 3.50 108 1253 n 7.15 106
!"7Pr 1.70~1010 !
183S e 8.48 108 I
14 -
l.
+.em.w-y%9- . v-..e p,- e t y m em-m---+m'yy4 y u "W '* p me-gp.-'T qag & Ir vvrww
Table 2.7 Ref erence Magnon Rosetors Crephite Masses and 60Co Specific Activities Region Mass ( t) 60Co (Bq/t)
P Core plus low'er 1613 1.7 x 1010 axial reflector ,
Raoial and upper 620 0.8 s 1010 exisi reflectore l
Derivation trradiation period 40 y at load f actor 70!! decay period 10 y af ter shutdown.
- 1. Moderator (core) graphite: 1550 t Because of burn-out effects the 60Co specific activity in largely independent of flus in the core region. The valus of 1.7 x 1010 sq/g can be soeumed to be representative of tbs core graphite f rom al.1 Magnox stations,
- 2. Radial reflector graphite: 160 t The activating neutros ilus is more or Lees constant over two-thirds of the reflector this 9ese, se the graphite to arsumed to have the 80Co specific 4;tivity corresponding to the average flus, namely 0.8 s 1010 Se,'t.
- 3. Upper axial reflectors 160 t Thu upper axial reflector le the top graphite layer. Over the lower half sette of bricks the 60C e epocific activity will be similar to that of the radial reflector, and that value has been taken for the whole layer.
4 Lower axial reflectors 63 t This region corresponde to about half of the botton graphite layer. As it caanot easily be separated from the core graphite during decommiestonios, it has been included with the core and assigned the ease 60 Co specific activity. *
! 15 -
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Table 2.5 Reference ACpt Crephite Messes and 60Co $pecific Activitiet Region Mass (t) 40Co (Sq/t)
=
Moderator 9.3 a 10!!
1322 lettom reflector 6.7 x 101?
Radial reflector 7./ x 1011 321 Top reflector 2.5 x 1011 i
Derivation Irradiation _ period 40 y at load f actor 701: decay period to y after shutdown.
Because of the sistlerity of their specific activittee, the moderator.
bottom reflector and radial reflector staphites can all be treated identically.
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- 3. PACKACING FOR srA Di$PCSAL 3.1 Cene ral Requi t ement e The graphite moderators of Magnos reactore and AGRe are built f rom blocks. about 0.2 a 0.2 1 0.8 e for M.enom and about 0.6 x 0.6 x 0.9 a f or AGR, which are keyed together with sm&11 graphite tiles. A possible method of disposal of this graphite is see-dusping under the provisions of the Londor. Conventinn " and the NEA Coneuttative Mechanien This section .
examines the tuplications of the sea-duging option f or graphite, in teres of total numbers of packages, weight and cost. Wree alternative package concepts are examined. De first is the drum-like disposal package, stallar to those used at present. he second conceps envisages an unshielded disposal package containing the graphite, inside a separate shielding overpack which to returned af ter the dumping operation. he third is the concept of a large self shielded package, the graphite betet encapsulated within a concrete monolith.
The minimus time .if ter reactor shutdown considered to 10 y (section 1).
Longer decay times, efter which the sackaging method could perhape change, are also considered where appropriate.
3.2 Dispoest in Drum contSinert Drua containers suitable for een dumping are described in the NtA One of the largest Guidelines and in a paralle1 ITK Code of Practice .
. concrete-lined steel drums ta assumed to be used: the " Type 1803". of ,
enternal d1auster 0.84 a and height 1.1$ u. t- graphite vastes, this drum could be used in.either of two ways. If there were no shielding requiremente, the graphite could simply be packed into the drum with some iora of cement or concrete grout to make a package with no voids . no 3
the available capacity would be that of the drum itself, abou 0.64 m . If some shielding ween needed, the drum could be given an internal precast I
concrete liner at least 75 es thick,U'Nin whit.h case grouting would not be necessary but the available capacity would be reduced to 0.37 al or less.
l I The achievable capacity for graphite depends on two further f actors, the volumetric packing efficiency and the thickness of any additional internal shielding raquired to limit the surf ace dose rate. For this study, a patking ef ficiency of $01 has been assumed, a value which to considered achievabic f or either granulated graphite or whole blocks. In practice somewhat higher or lower values say prove possible, accordir.g to the design and phyencal state of the graphite fios the deconstestoning of any particular reactor. C.elcuist ons of internal shielding requirements f or
- 26 - a 4
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i drummed graphites have been based on a does rate of 0.5 mSv/h at the surface of the drus, this being within the limite suggested in the UK Code of !
Practice I '). It hee _ been estimated that tMe dose rate would be fourd 'st ,
the surf ace of an unlined Type 1803 drur ~1th a 501 packing ef ficiency, containing graphite having a 60 Co specih! ac?.1vity of 1.5 CBq/t. The ;
neceteary attenuation factors and the actual shielding thicknesses say thus [
ta turn t,e estimated.
3.2 1 Roference Maanos reactor 10 y af ter reactor shutdown, the core and ref)ector graphites whose ,
'OCo inventories are given in Tabla 2.7 would all require shielding for sea ~l dumping. For the core plus lower antal reflector graphite a thickiese of 90 ma of heavy concrete (density 4.8 t/r* would be needed. For radial plus upper antal reflector graphite a notiont.1 .f'ekness of 60 mm is indicated.
though in practice the liner would be at least 75 me thick, se mentioned I above.
I The 90 mm thick liner would reduce the capacity of the Type 1803 drum.
and f rom Ta' ale 2 5 the number of drume required would be 466J and the total ,
ases (core and lover asial reflector graphite + shielding + druss) weuld be 9094 t. A further 1590 drums with the normal 75 mm liners would be required-for.the radial and upper asial reflector graFhite, contributing ar.other 2910 t. The totale for all the graphito from the refersace Magnnu reactor would thus be about 7WO ,dm and 12.000 t. ;
35 y af ter reactor shutdown,_ shieldtra would b. required _ for neitner-the core nor .aa reflector graphite,' oc the entire 2133 t of graphite ceald be disposed of in monolithic contataere using OrJ16ery portland Cament (OPC) [
as an encapsulant, which would also bring the epoc%fic grevity above 1.2 es required f or _ saa-dumpungI ' I. bperience within CECs to that crushed tra- y phite will not float in cement, despite the density di,f ference, unJose the sia is vibrated., If whole bricks de float the encapsulation process cculd be done in two attges or et alternative encopoulant could be' used. As s un t .a the use of OpC, the numb. of cetainers for the graphite _ from the reference .
i-
- Magnos reactor would be about 3400 end the total mese about $3001 Costa The following wait costs are soeused (1980 prices)
Shielded contateer with heavy concrete 2750 Unshielded drum with OpC encapsulant (100 Transport and disposal 1 80/t
!- )
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l These lead to the ft119 wing costs per reference Magnos reactor 10 7 after shutdown ES.$N 35 y after shutdown 10.76M plus 1125M per resctor f or ;
encapsuistion plant.
The total quantf ttee of graphite, numbers of drums and estimated costs for the Magnon deepmaisotoning programme are given in section 3.5.
3.2.2 Reference ACR The consideratione for packaging graphite f rom the ref erence ACR are similar to those for the Magnon reactor, described above. However, the specific activity of 6bCo at 10 y after shutdow, to so high (Table 2.8) that concrete connot sensibly be considered as a shielding agiarial. For a type 1803 container, the thicknesses of steel ehtolding required to reduce surf ace dose rates to 0.$ nav/h would be 140 me for moderstor and bottos-reflector graphite, and 110 mm for radis1 and top reflector graphite. The numbers of drums and masses for disposal would be 5748 and 23.278 t for th<
former graphite, and 1067 and 3735 t for the latter, giving approminate totals for all tht graphite from the reference ACR of 6400 druse and 27,000,1 Not until 60 y af ter shutdown could all the A9A graphite be see-dumped in unshielded drums (cf. 35 y for Magnos). About 2500 druss would then be required, and the total mare would be about M001 costs The following unit costs are assumed (1980 prices) steel shielded drum (cast) 1400/t Unshielded drum with OPC 1100 transport and disposal L 30/t These less to the following roots per reference ACR 10 y after shutdown 112.3M 60 y after shutdown 10.$6M plus 21.2SH per reactor for encapsulation plant.
3.3 Disposal in shielded overpacks 3.3.1 C,eneral reautremente The current types of sea dumping containers are smaller than to desirable f tpe the viewpoint of shtolding. Larger packages would have a greater degree of self-shielding and require less to be- provided enternally.
- l. Also, if wastes could be transported to the dumping ette in returnable shielded overpacks, the packaged wastes themselves being unshielded. then the cost of shielding materials loct by dumping could be replaced by the l
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perhaps lesser cost of returning the overpacus f or re-use. The savings would in principle would be greater for larger packages, though the osckage site would in practice be limited by other constraints. In this study the shielded overpack has been assumed to be c.ylindrical, of 2.5 e dissetes and height, sade of steel, and weighing about 16 t. It could carry either 21 x 0.36 m3 drums or a single 9.37 s3 liner. In each case the graphite would be encapsulated in OPC, so the der.Jity of the dumped package would exceed '
i-2(3'4) .
For the purposes of transport, all the grapt ite vastes could probably be classified as low level Solid according to the current 1AEA Transport i Regulations and hence say be packed in strong industrial containers with a dose rate at 2 e f rom the surf ace not greater than 0.1 mSv/h. For the site and shape of overpack engtsaged, the corresponding surface dose rate would be 0.2 e5v/h.
3.3.2 Reference Magnot reeStor A shteided overpack would only be required f or decay periode less than 30 y af tar shutdown (section 3.2.1). co the 10 y case is examined here.
Only the option of the single 9.37 53 packege is presented since this represents the other end of the cost spectrum from conventional Type 1803 Aruas.
If all the grades of graphite f rom the ref ereeca Magnox reactor (Table 2.7) were included together, the total nsaber of overpsck journeys would be about 280, and the total mass to be transported would be about 9.700 ,t, coopered with 12,000 t f or the shielded drums (section 3.2.1).
Costs he cost of the liner and OpC encapsulant has been estimated to be
($00. De costs of transport, diaposal of the graphite in the liner, and return of the overpack have been taken to be (80/t (ths ease as f or transport and disposal of shielded drumet the savings due to handling larger packages are estimated to offset the cost of returning the overpack). The total cost for the reference Magnon reactoe, f rom these items only, would be about to.92M. However, there would be additional capital costs: L2.5M per 2-reactor station for encapsulation plant. 1.32.000 for each overpack (based i
on (2000/t for fabricated mild steel), and 145.000 for esen special rail wagon. The numbers of overpacks and wLsons required would depend on the transport ard disposal scheduling, and some of the costs might be attrit.utable to ther diaposal operations ehering the same resources.
3.3.3 Ref erence ACR Not until 60 y af ter shutdown could all the graphite f rom the ref erence l
AGR be transported ut. shielded (section 3.2.3), so as in the previous sectio, i
U .
only the case of 10 y decay ti, examined h6te. If all the graphite given n r.
Table 2.8 were trested the sese, the number of overpack journeys would be about _240, 4 and the total mass to be transported would be about 11,800 t,.
compared with 27,000 t f or disposal in shielded druat.
Oost) On the same assumptions as f or Magnox (section 3.3.2), the costs of liners, OPC, transport, disposal and return of overpacks woulo be about it.lM for the refurence AGR. Additional costs of encapsulation plant, overpacks and wagons would apply at the same unit rates.
3.4 Disposal in 1.arge Concrete Containers 3.4.1 Ceneral reautrements As noted in section 3.2, the existing arrangements f or the disposa) !
packaged radioactive waste into the sea rely asinly upon the material being encapsusted into steel drune. The upper limits of volume and weight .
0.73 33 and 3 t, are dictated by the capacity of the existing transport and handling facilities.
During the UKAKA study of decommissioning plans, a review of the type and quantity of materials arising indicated that the use of larger packages would have advantages., The principal gains might be a reduction in the ncaber of packages required, reduc 61on of the doses received during handling of the waste and packages, better control of contamination, and using the self shielding properties of larger packages to further taprove the efficiency of the operation. It was theref ore decit.ed to study larger packages in more detail.
UK railway replations dictated the maximum size of tre containers.
l The external dimenatone selected were length 2.36 a plur, lifting l attachments, width 2.21 a, and height 2.21 m. Two containers weuld be carried simultaneously, with the lif ting attachments placed f ore and af t upun the rail wagon. The maximum weight of each container could be 50 t.
The containers are designed to be transported under the requirements of l
a " strong industrial" package as defined in IAEA Saf ety feries No .and l carried as a full load. Whilst in transport the radiation level aunt not exceed 0.1 m Sv/h at 2 m f rom tN surf ace of the container .
The general f eatures of the contatuer incorporate v'ere practicable the recom endations described in the NEA (bidelines and the UK Code of Practice f or sea dumping. ISO standard f reight containers could 'Je used f or wastes requiring little or no shielding, but f or the more active vasten a new type of container was designed to the above size and weight specification.
1
Precset reinforced concrete 23 cm thick wim.'d fore the esternal shielding valls of the container. Normally the concrete would have a density of 2.6 t/a ,I though a denser (4.3 t/s3) material could be veed, the choire depending on the attenuation factor required to achieve a persissible enternal dose rate.
The containers (Fig. 3.1) would be clad externally with 12 se a11d steel. (A possible anti-corroaton finish by means of metal spraying is being considered for a long tern land storage situation if the need arises.)
The mild steel cladding would serve a fourfold purposet (a) to prevent accidental damage to the concrete and assist in retaining th. shiniding properties, it damaged during handling and storaget (b) to act se a foster for the pouring of the precast concrete shieldt (c) to at J decanteatuationi and (d) to improv. the appearance vn11st being transported through the public sector.
The outer placast concrete container wov'd be filled wirb the graphite wasce incorporated in a monolithic cos.?r*" ntria, with appr priate precautions to prevent the graphit. from floating in the concrete before the concrete sets. The six for both precast and infill concretes has been designed to have a compressive strength of not-less than 20 L./es2, m #
completed assembly would have a specific travity of not less than 1.2 to ensure that it sinks to the sea bedI ' I. Preparation of the internal face of the precast concrete container would be required, to ensure an adequate bond between the infill concrete and the precast concrete. This would also assist in retaining the concrete shioiding under accident conditions.
At each Magnon/ ACE site a weets packaging f acility would have to be built or adapted from asiating buildings. This facility would be used for assessment of the activity contant of the waste and the dose rate, and for placing the infill ccacrete. lawluded in the building would be a bay for sonitoring the conta16.ere and if necessary decontaminating thes before dispatch. The capital cost sould be about f.2M per reactor ette, but could be apportioned anons all decommissioning wastes f rom all stations on the site and therefore is not included in the cost estimates given below; all other operating costs attributable to graphite disposal have- been included.
Rail transport for the containers has been assumed, to a buf f er store at the dockside. A train would comprise a maxistas of 6 wagons, each carryins 2 containers. The cost of special 8-axle rail wagons has been estimated at 1165,000 eae.h. though for tte 130 containers standard "Flatrol" wagons could be used at 145,000 each. As above, these capital costs have
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i not been included in the estimates, but the running costs have. There ,s a relativalf large tuitial cost la running a train so the estimates are ..t very sensitive to the distance travelled.
A dedicated disposal vessel for this type of container has been assuesd. The design is shown in Figure 3.2 and is bas 6d on a veso+1 called a " pipe carrier" in use in the North Sea oilfields, which has a deck cargo capacity of about 1600 t. The vessel would be equipped with a $0 t dertien crane for loading the containers, and once on deck the containers would be soved by the gantry ceaue. On reaching the dump site the gantry crane would sove the containers one by one to the hinged platfore at the stern of the ship, and they would be discharged over the stern by hydraulic rams. The capital cost of the vessel sind onboard cransse would be about I&.5M. and ras not been included in these estimates. However, the operational costs have been included, and have been estinated to be L2500 a day for the 10-36y return trip f rom a UK port to the NEA dump site (about 900 km south
- west of Land's End) plus (1000 a day for t e aseused 2-day turnaround in port.
3.4.2 Reference Hannon Reactor The 2233 t of core and reflector graphite whose 60 Co inventory is given in Table 2.7 would all require shielding for ses disposal af ter a 10 y decay pe riod . The standard (denoity 2.4 t/m 3) reinforced concrete container would be adequate, and would give a dose rate of G.23 e Sv/h at the surface. This dose rate has been estimated using a 74% packing efficiency.
The container has a not capacity of 5.7 m 3. which will acconeodat-7.3 t graphite in the Isra of bricks. These bricks would be incorporated in 3.6 t of concrete infill to form a package with no voidsI') . The completed package would weigh 25 t. From Table 2.7 the total number of containera required would be 307 with a total sans of 8600 c.
It decaenissioning use delayed unc1135 y no shielding would be required either for the core or reflector graphite. so the whole of the 2233 t of graphite could be eleply encapsulated in Ordinary Portland Cement (OPC). ISO cargo containers could be used for this purpose providing their capacity was lietted en 30 t. The number of containers f or the graphite f rom the reference Magnos reactor would be approximately 140 with a total ease of approximately 3990 t.
Costs The costs for sea disposal of large concrete containers have been estiested in some detail. All costs are given using 1980 prices. As I
sentioned above, they include all running costs of packaging, sonitoring, transport and disposal. but not the capital elements. For disposal of the 10 yent-cooled graphite frort the reference Magnon reactor the cost would Se
13.25M. Disposal af ter 35 y could be in 150 containers and the cast would fall to Lt.3M.
3.4.3 Reference ACR By using the higher density concrete container (6.3 t/m 3) and reducing tne packitig ef ficiency to 651, the ACR graphite could be ses-dumped 10 y after shutdown. This reduction in packing efficiency allows an increase in shielding thickness to 32 cm by adding 9 cm of infill concrete (also 4.3 t/m3 ) to the 23 cm precast liner. The dose rate at 2 a f rom the surf ace of the containst would be less than 0.1 a Sv/h. The numbers of containers and the total sees for disposal would be 2,51 and 10,900 t.
Beyond 35 y af ter reactor shutdown, the grcphite f rom the ref erence AGR could be incorporated in a stonderd density concrete container (2.4 t/33) with a packing efficiency of 741. This would require the use of 220 cortainers with a mass of 6330 t. Beyond 60 y af ter reactor shutdown AGP.
graphite could be grosted into ISO contatiers as described above for Ma g nox.
Costs 10 y after reactor shutdown, disposal costs for the reference ACR graphite would be L3.5M. These costs would be similar to those tot the ref erence Kagnox reactor at the same time, the lesser mass of the ACR core being of f set by the need to use the more expensive high-density concrete f or better shividing. 35 y af ter shutdown normal-density concrete would be adequate and the cost would then te L2.3&M. The cost of disposal in 150 containers 60 y af ter shutdown would be LO.97M.
3.5 conclusione All three packaging and sea-disposal methods ( A.41ned are logistically f easible, and could be implemented at reasonable tows. The cost estimates have been made on different basse and therefore cannoc be used to make detailed comparisonst however, it appears that the larger packages would be more cost-effective. Further and more detailed cost studies are required.
References for sectiog 3,
- 1. Convention on the Prevantion of Karine Pollution by Dumping of Wastes and Other Matter. London, 1972.
- 2. CECD Nuclear Ertegy Agency. Decision of the Council establishing a multilateral consultation and surveillance mechanism f or sea damping of radioactive wasta. Paris, 1977.
- 3. OECD Nuclear Energy Agency. Guidelines f or sea dumping packages of radioactive waste. Ravised version (1979).
e
- _ .~__ __ _ _.
4, llK Ministry of Agriculture, fishertes and Food. Ses disposal of radioactive wastes packaging requirements. DSCP 1 (1979),
- 5. International Atomic Energy Agency. Regulations f or the saf e t ranspor t of radioactive materiale (revised edition). Saf ety Series No. 6, IAEA, Vienna (197f4)
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6 PAcKAc!NG OPT 10Ng TOR 1.AND DISPOSA1. ;
At the present time there is no land disposal f acility in the UK f or the disposal of graphite f rom reactor decommissioning. A number of alternative types of f acility have been definod( ' and are the subject nf I generic studies ( '. '
However it has already been shove in the previous section that graphite is not unduly dif ficult to package, being a strong, stabte solid material, and that many of the constraints on possible packaging methods arise f rom the need to transport the finished packages overland f rom the reactor site to the dockside. Much the same applies to packaging and tra tsport for land disposal, although the surf ace dose rate limit in the
! AIA Transport llegulations la not as stringent as that for ses disposal.
Although the general concidstons of section 3.5 apply equally to ses and land disposal, the specific requirements could dif fer in respect of overall size and weight limits. and in the importance of achieving high volumetric packaging efficiency. As always, these two requirements conflict arid must be balanced. The optinu for sea disposal might not be the same us f or land disposa . and the optisms f or deep geologic disposal night be dif f erent i
f rom that for shallow 'and burial. It is not yet clear whether graphite f or ,
land disposal would have to be grouted to form a sono 11thic package in order to bear tne overbunden stresses in shallow burial or the geoststic stresses in deep geologic disposal.
A festibility study by CEGS on the shallow land burial of radioactive wastes has estimatsd that che cost (per toene including transport) would be about- M1 higher thn. 'or sea disposal. -
References for Section 4
- 1. Duncan. A G and Brown. S R A. batities of weets and a strategy for disposal. Nucl. Energy. R (3),162 (1982).
- 2. Cinniff N E. The atting, design and construction of f acilities f or low and intermediate level radioactive vastes in Great Britain. IAEA.
International Conf erence or. Radioactive Waste Management. Seattle. USA .
May 1983 (to be published).
- 3. International Atomic Energy Agency. Regulations for the saf e transport of radioactive asteriale trevised edition). Saf ety Series No. 6. I AEA.
Vienna (1973).
,,. -- ,, . _ , . -- , ~ - e - - - - -
I S. INCINERAT!g
!! graphite wastes were to be incineratod, the total volume of solid wastes for disposal would be einsiderably reduced. However, the l*C and N content would be discharged to atmosphere, and the incinerator would constitute an additional cost. This section examines the technological and cost aspects of one conceptual method of graphite incineration.
S.1 Combustion The graphite recovered from the cores of Magnos reactars and MRs will be in the form of large blocks and tiles (section 3.1). For a reasonably rapid and more easily controllable combustion process it is preferable to site-reduce the graphite, og by hammer silling or staged jaw crushing.
Cooling the graphite below 100 C would increase the ease of size reduction but may not be necessary. A roughly cubic product, about 2.5 cm in face size, could be produced by commercially available equipment but this would need to be suitably shielded and modified fer active use. To accept the larse graphite blocks a mill capable of crust.ing tens of tonnes per hour would be required. A esaller aised product would require another stage of crushing, with more severe dust control problems.
Enlargement of the graphite pores due to radiolytic corrosion by the reactor coolant gas, and deposited metal oxide dust and the ef fects of radiation, are all likely to make the graphite slightly more reactive towards oxidation (a factor of 2 at 10000 C) than unirradiated gr.phite.
Craphite can react with oxygen in the air thus C+O 2 = CO 2 M = -94.0 kes1/ mole (180 C, 1 sta)
C + \02 = C0 M = -26.6 " " "
CO + 02 = CO2 2 = ~67.4 " " "
C + CO2 = 2CO M = +40.8 " " "
The minus sign denotes the anothermic reactions. The relevant equilibrius constants are +15.9, +9.1, +6.8 and +1.3 at 1000 0C (assuming the gas pres-sure is seasured in atmospheres): this indicates that the reactions are thermodyneutcally favourable. Operation with air at 10000 C would have the advantage of a reasonably rapid combustion, minisat production of carbon monoxide and a reasonable service life for furnace romponents. At about 0
1000 0 the speed of graphite oxidatioe is controlled mainly by boundary layer diffusion. is by the rate of diffusion of the oxygen across the gaseous boundary layer covering the graphite surface. Oxidation of irradt-ated graphite p16ces (2.5 cm cubes) at 10000 C would be virtually complete e
within a few hours, so it is estimated that a bed volume of about 233 would be required to handle the combustion of 10 t graphite per day.
A notional flowsheet for the combustion of 10 t per day of 10 y cooled graphite f rom the ref erence Kagnos and ACR stations is shown in Fig 5.1. It is sasuand f or the flowsheet that the graphite from the reflector and soder-ator areas of ar individual reactor would be randomly sized before f eeding to the process, but that graphite from individual reactors would be cam-paigned separately. A propane burner would ts required for start up the f requency of start-up being detersined by the method of working employed.
Continuous shift operation should enable about 2,000 t graphite per year to be burnt. A vertical furnace is envisaged and the need for hot air to be blovn in at the base of the furnace means that a grid, even a non-metallic one probably could not be fitted to support the burning bed. Cooler air would have to be fed in via several inlet points to limit the temperature of the walls and base, and to minimise C0 production. the graphite pieces are envisaged as being fed in via a double-doored hopper vertically above the furnace.
The flowsheet shove a daily arising of about $$ kg of furnace ashi in practice this ash would probably be only .*emoved weekly and at a time when the furnace inventory of graphite was at a minimum. In the ash, the flov-sheet shows a quantity of I"C corresponding to 0.2% unburnt graphite, though this figure is very uncertain, In order to separate the taitological impact of disposal of the non-volatile radionuclides in the ash f rom that due to any accompanying I"C the assessment in section 7.9 deals first vith the non-volatiles and then considers th? ef fect of the presence of bC.
5.2 Treatment of Furnace of f-Cases and Recovered Ash it is envisaged that the of f gases would leave the furnace at abaut 1000 C, and would pass upwards through a short ceramic-lined duct into a 0
vertical quench chamber, down which a water spray would be directed so as to cool the gases to about 250 0C. The gases, now containing water vapour.
would pass on to the primary filtration unit, consisting of 10 um porosity sintered steel filters. These filters would be regularly e.nd sequentially blown back with heated compressed air to resove the collected dust, which would Se collected in a hopper beneath. The filtered gases would then undergo secondary filtration through high capacity KEPA filters operating at i about 150 C. 0 The combus:*.on of f-gases would then be discharged to atmosp- I here by a f an via a 40 m stack to provide adequate dispersion. All the e
_______u___
l I
coolicg vetor introduced in the optsy would be dischstged as vapour. so there would be no liquid effluents.
The used IEyA filters would be compacted individually in a remotely operated baling press and placed in 2001 drums. The drums containina ash.
compacted KEFA's or possibly single used sintered filters would then be filleJ with cement slurry containing the furnsee residues and the blown-back dust. Immobilisation of the drum contents with a cement grout would be necessary because of the dusty and easily dispersable form of the waste.
Each 2001 drum would effectively contain the non-volatile radioactive species (plus traces of unburnt I"C) from about 10 t of graphite. The radiological impact of the activity discharged to etstophore is sesessed la section 7.9.
5.3 Transport /Disposs! Requirements for the Immobilised Asb It has been assumed that the technical requiremente for transport are given by the LAEA Regulations for the Safe Transport of Radioactive Materials.Il} and additionally for sea disposal by the NEA Guidelines for Sea Dumping Packages of Kadioactive Vaero(2) . According to the torner the cemented graphite ash weste may be classed as Low 14 vel Solid and packed in strong industrial containers with a surf ace dose rate not greater than 2 mSv/h and dose rate at 2 e from the surface not greater than 0.1 aSv/h.
and no further limit is assumed to apply in the esse of land disposal. If the dose rate at the avrface of the ses dumping container were to be further limited to 0.5 asv/h, as assumed in section 3 for direct disposal of graphite without incineration, then stallar calculations indicate that attenuation f actors of about 240 and 1200 are required for Magnos and AGR ash respectively. For the Magnox case, the attenuation could be provided by about 0.18 m of heavy concrete (density 4.8 t/m3 ), or at greater cost by 0.09 m of cast steel. yor the ACR case, the necessary thickness of heavy concrete shielding could not be accomodated around a 2001 drus within the standard sea disposal drum, so steel 11nere 0.18 a thien would be necessary.
The grose masses of the lined and filled sea disposal drums would be 4.3 t f or ACR ash, and 2.7 t for Magnon ash. Each drum would contain the non-volatile activity associated with 10 t of graphite, so the By specific activities would be 1.6 TBq/t for the ACR ash and 81 GBq/t for the Kagnon ash.
5.4 costs incineretion and ash immobilisation It is difficult to cost tb8e systen since there is little experience with radioactive plant of tra
regaired duty \ Taking into .ccount the considerable development work and coste and the need to safely and roostely maintain the plant. it is unlikely that a combastion/aeh immobilisation unit could be constructed and approved t for use at a cost less than 130 M. With such a high capital cost it might he thought that a few regional facilitise or a national one could be more cost-ef f ective then building numerous combustion /soh immobilisation '
f acilities, one at each of the UK stations. Estimated conte range f rom a 3s41Run of (15,000 per tonne of graphite for numerous smaa or f acilities down to perhaps 1500/t plus tronoport for a national facility. Varying the plant throughput f rom the notional 10 t/ day assumed here would af f ect the capital costs to some dsgree, but any optisisation to cover the range of sitec, treatment rates and plant lifetimes le outside the scope of this report. 1 Transport and disposal For tonee of graphite fed to the combustion process, costs of transport and disposal have bcsn estimated as about f40 f or ACR ash and 120 for Magnos ash. These settmates are more uncertain than those for stephite itself (sectione 3 and 4) but the uncertainties and the dif f erences between the costs of sea and land disposal are not important. l since the costs of the incineration step are so much greater.
Overall costs It can be seen that the capital cost of the combustion /seh immobilisation f acility is the hajor ites. Depending on the number of f acilities needed, the overall cost per tonne of graphite for the incineration option could be up to 215,000.
References for Section 5
- 1. International Atomic Energy Agency. Regulations for the safe transport of radioactive materiale (revised edition). Saf ety Series No. 6.
IAEA, Vienna (1973).
- 2. OECD Nuclear Energy Agency. Cuidelines for sea dumping packages of radioactive weste. Revised version (1979).
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6 LEACHING OF fiULAD1ATED CitApHITE 6.1 Introduction i The leaching behaviour of radiologically significant radionuclides f rom small scale samples of reactor irradiated graphite has been investigated in support'of the radiological 1spect sesessment. The test conditions have been cimed at simulating possible disposal environments, in particular those of shal .ow land buttal and deep ocean disposal. In addition, a control tent environment has been considered from which, if desired limited comparisons
,ay be made between the leech rates f roe graphite and other waste f orms.
In support of the assessment of the deep ocean disposal option, the ef fect of high hydrostatic pressure on the structural integrity of 16tae blocks of unitradiated graphite has been separately investigated.
6.2 Experimental 6.2.1. Prersure teste Patre of unitradiated blocks of graphite, of nominal volumes 14 and IM litres, were tested at the British Teleconsunications Research Cent re.
Martlenham, to assess the durability *f the material under simulated deep ocean pressures. The blocks were measured and weighed before and af ter testing to assess any gross physical changes which might have occurred.
Jigns of structural damage were determined by visual examination, Du rin g th,se touts the blocks were immersed in water and hydrostatic pressure applied up to a aaximum operating pressure of 928 bar, appreximately twice that which would prevail at a disposal; depth of 40C0 m, *)0 bar.
6.2.2 Leaching tests Leaching tests were carried out using small-3cale samples of graphite which had been irradiated in the core of a CICS Magnox reactor for approximately 13 y at about 16.000 mwd /t. Af ter about 3 y post-t rradiation decay the blocks were machined to remove the surf ace layer. In orde r to einimise interference from surface contesinants. The finished camensions u the samples are given in - Table 6.1.
The testa perf ormed f ollowed the guide 11nus prerosed by the I AEA for their standard leach test method with certain modificationst in particular, the entire sctf ace area of the sample was exposed to the teachant in order to increase the sensitivity of the test. Th e t e s t enviroasents and permasters are given in ?tble 6.1. The compogations of the seawater and groundwater stuulates are given in Table 6.2. The deep sua ~
disposai environment was staulated by a specially designed high pressure rig which enables leachate sampling and renewal to be carried out with only a=small loss in operating pressure, amounting to less chan 101.
42 -
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l In each environment two separate graphite samples were tested and the I teachates bviked for analysis. Graphite saeples esintained at atoospheric pressure were wesghed at each change of leechste af ter ext.ess surf ace water had been removed. The leachstes were filtered to remove any particulate mitier before being assayed. Y-esitters being determined directly by T-ray vectrometry, and S emitters by liquid scintilla. ion counting af ter sultahle ra,'iocnemical treatment to teolate 3H sad I"C .
In order to derive meaningful leech r:te data, a radionuellde inventory of the graphita easples _was determined before leech testing was carried out.
t emitters again being determined directly by Y-ray spectrnmetry. Th e significant 8-suitters. 3H and I"C. were detersindd by a procedure involving thermal decomposition of reeresentative f ractions of the graphite samples followed by liquid scintillation counting.
6.3 Resulta 6.3.1 Pressure tests The results obtained on _ subjecting large unirradiated graphite blocks to high hydrostatic pressures are presented in Table 6.3. In all cases structural integrity was maintained with no gross change in block dimenensions or evidence of cracking being obt svu 11 the test blocks exhibited an increase in weight as interstitial air n were filled. 9:311-scale irradiated samples als'o showed a progressive gain in weight, si-marised in Table 6.4 6.3.2 1.eaching tests The mean raitonuclide idventory of the graphite samples is given in Table 6.5. Although absent from thogratical inventories. II"Ce 'has been detected 'in all the graphite samples (and leachates). Other workers have also detected 13"Ca in irradiated ACR graphite (D. Hatton, CEGB. privat e communication). It would appear that this nuclide originates f rom the activation of 133 Ca present as a trace impurity in the graphite, rathe r than as a fission product, f or II7Cs has not been detected in these or other analyses. Further work would be required to confirm the mechanisse of production of the observed II"Ca.
The f ollowing nuclides have been detected in all the leechate samples:
3 H. 1"C. 1335a. 40Co. 13"Cs further nuclides listed in Table 6.5, if present in the samples. lie below the 1,imita of detection. ~
The cumulative percentages of activity released
- f or each environment
- This represents: f t activity Isached )
(initial activity of sample)
- s
are shown in Figuree 6.1 - 6.4. The curves all exhibt shar, intttai rise over the l'. ret 10 daye or so before approaching equilibrium salues. The enception to this behaviour is 60 Co, the curve for which continues to rise throughout t t.g duration of each test.
Incremental leech rates are given in Figures 6.5 - 6.8. Each curve exhibits a characteristic initial sharp fall, possibly due to lasching of nucitdes pre int on the surf ace of the graphite, followed by a more proaressive decline towards apparent equilibrium. Assin, 60Co shows anomalous behaviour in all environeetits. Nc tuplanation can as yet be advanced to account for this behaviour. The 'esch rates for all four environments are comparables variations in pressure, temperat tre. and teachant composition would appear to have little affect on the obse rved teaching behaviour. The results of the Isach tests are summertsed in Table 6.6 which shove incremental leach rates and cumuistive - f ractions nf activity released for the observed nuclides in each test environment at day 100. Further work to required to understand the physico-chealcal mechan 19mm governing leaching free graphits.
heximus likely leech ratas for the rossining nuclides have been calculated f rom choir einimum detectable activities, and are given in Table 6.7.
6.4 Summary
- 1. Large scale blocke of irradiated graphite and easil scale hiocks of irradiated graphite maintained their structural integrity af ter being subjected to the hydrostatic pressures that would prevail at a typical deep csean disposal site.
- 7. The fo1&owing radionuclides were leached f ree samples of irradtaced graphite I"C, 3 H, A3"Cs, 133FA, and 80Co. With the exception of 80 Co all showed similar leas.hing behaviour in the four test environmente considered.
- 3. For the radiologically agnificant nuclides l'c and 3 H. the mastmum observed incremental . ssh rates did not exceed 10"' and 10*S 1, cm/ day respectively, and fell during the course of the tests tend-l ~~ ing towards apparent equilibrius values. The lose in activity from the graphite samples ovat the period of the tests (up to L$0 days) corresponds to 0.ns for 1"C and C.3% for 3 d a The unite of (ca/ day) for the leach rate represent
'f raction of initial activity leacned per day) x (vol ze/ surf ace area ratio of sample) l
6.5 !aplications for Radiolt Assessments The rates of release of in. .. dual radionuclides f rom the graphite as a f unction of time are the fundamental quantities required for radiological assessments. In practice it is convenient to empress release rates as fractions o* the original activity released per unit time (f ractional rel*ese rates). These quantities may be calculated f rom the incremental leech rstes such as those given in Table 6.64, given the volume / surf ace irea ratio of the ptoces of graphite. Although quite specific sesumptions mif ht he made about the shapes and c;tes of the pieces actually disposed of, f or preliminary calculations ihe values /curf ace area ratio can staply be expressed as a characteristic dimension typifying the site of the pieces.
If for example this dimension were 10 cm, the fractional release rate of 3*C implied in Table 6.6a would be about 2 x 10-5 y-t if t he leachant was simulated groundwater: the corresponding figure for 60 Co would be about 3 10"3 y*1 These end stallar data f rom Table 6.6a would only strictly apply at day 100 of teaching, under the sene conditions as the laboratory tests , and nted to be extrapolated over the far longer timescales of tLe radiological
-assessments in section 7. Several competing ef fects might be envisage *.
One would be that dissolution of the graphite would progressively reduce the surf ace area of each piece, and hence reduce the rates of release of radionuclides. Counteracting this effect might be a progressive mechanical deterioration of the graphite le& ding to crumbling into a more finely-divided fors, uhich would tend to' increase the release rate, perhaps drastically. A further possibility is that under some conditions of
<ionasal the amount of groundwater available to costset the saste might not or enough to leach it at the rate predicted by the experiments in which the quanti y of water was essentially unlimitedt under such circumstances the release rates would be ove r-estiested.
Given these uncertainties acout both the propertise of the graphite and the disposal environment it is difficult to juecif y detailed assumptions about leaching behaviour in the very long ters. But generally this is not an important probles, for in section 7 it will be shown that the predicted radiological impacts of the verious management options f or graphite are seldom greatly af fected by the assumed release rates. ;
For the radiological assessmenta, the value of the experimental work described in this section of the report has bean to establish that there are
-no dramatic changes-in the mechanical or leaching properties of irradiated graphite under hydrostatic pressure, and thet the ef f ects of dif ferent
leachante and temperatures are also not .very marked. , Th e r e f o r e in section '
the ef f ects of leaching have been modelled by accusing a range of constant fractional roleses rates, the game fot every radionuclidel and the range of assumed release rates, 10** - 10*l y"I, adequately represents the observed rates at day 100 for typical particle elsee (Table 6.6a) as well as the initial transient release of up to a few percent of the initial activity (Ta ble 6.6b). The sensitivity of the results to the assumed teach raten (=
discussed in section 8.3.
References for Section 6
- 1. Haspe 1 D. Leach testing of immobilleed radioactive waste entida .
Atomic Energy Review j[ p 195 (1971).
- 2. Kaye. C J. The design and operation of a rig investigating the leaching behaviour of radionuclides under conditions of high hydrostatic pressure. CECB , NW Re gion. Repo r t No. NW/S S D/R R /R 3 / A l .
(1981).
- 3. Loveridge, 8 A.
The determination of tritius (r ef fluent. AERE.
Harwell AERE-M1219 (1963).
4 Cay, h P, Overv, M J.
The determination of I"C in saarles aristna f rom
' nuclear power stations. CECB, SE Region, Research Note No. 40/6 7 (1967).
- 5. Ridgway, D. Mcrtin. T J. Kaye, C J. The determination of I"C and lit in reactor core graphite. CECB, NW Region, Service Note No. NW/SSD/SNIO; (1961).
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'I Table 6.2 Ceeposition of teachates Steulated See-water (to 38 3900) _
gne Sodium Chloride 26,500 Magnesium Chloride 2,400 Magnesium Sulphate 3,300 Calcium Chlorids t.100 Potasetum Chloride 730 Sodium Ricarbonate 200 Sodium Broside 280
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Table 4 4 Wetthe Cain of trrediated Crephite Blacks Over the Period of teach Testina test conditione laseroton time Mean weight gain
.(days) (t)
Simulated groundwater 130 12.4 (L bar, 23'C)
Desinere11eed water 150 12.5 (1 bar, 20'C) simulated seawater 106 IS.1
.(450 bar,-2.5'C)
Staulated seawater 137- 9.1 (1 bar. 20'C) k
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Table 6 1-Measured Ra41onuclide leventory of Crechtte semples Radionuclide Mean*epecific activity:
(S c./ s) .
3N 2.7 los I"C 21 10" S Sre 1.9 10" 1855u 9.3 101 13 %a 2.2 10E 182tu a 10 100 sag c3-IS"Bu 9.6 102 60Co 1.0 10" 13"Ce 41 102
(
Mean of two semples, cospesed of traphite tros each leach test ossple group (see test)
b Table 6 64 incremental teach R4tes at Day 10,0 ;
Incremental leech rate (cm/ day)
- Nuclide simulated. Dentueralised Simulated eenvetor simuisted ,
groundwater water (2.5 0 C, 450 bar)' seawater IH 2.7 10*8 2.7 10-6 3.4 10*' 11 to'* i I"C 6 '. 0 10*I 1.4 10*8 9.0 10*7 54 in ' . i !
1338a 8.0 10*" 1.4 10*" 4.0 10*" 5.n in '
'Oco 7.0 10*l 2.0 10** 5.4 10-5 9,c i
'- i 13"ce 6.0 40*3 7.0 10-8 1.7 10*l 4.4 10-5 I
Table 6 6b Cumulative Practions of Activity teached at Day 100 Cumulative fractions of activity leached Nuclide Simulated Debineralised $1muisted esawiter simulated groundwater water (2.3 0 C, 450 bar) seawater j
--t 3
H 4.4 10*8 52 10*3 2.4 10*3 2.0 10-8 l
!"C 1.4 10*8 8.8 10** 4.8 10*" 5.0 10-'
133Ba 1.6 10*1 2.7 10*1 1.4 10*1 2.a lo*l 80Co 1.3 10-2 2.0 2 - 1.2 10-2 t,7 te-2
- 13"Cs 1.8 10*1 4.5 10-2 c.o 10-2 .,
t.3 to-t-For details of test conditions see Table 61 a
9
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Nuclide -
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40 60 80 100 lb 10 3ho 18 0 nme , e-Figure 6.3 Cumulative fraction of activity leached in simuloted see water at high pressure-( 450 bar . 2.5'C )
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7 RADIOLOGICAL ASSESSMENTS 7.1 Introduction In this section the methodologies and models used to calculate the radiological impact of each management option are discussed and results are presented. For each option there are several potential disposal routes.
The models used to predict doses f rom each disposal route are described, and detailed results are presented f or the disposal of graphite f rom the ref erence Kagnon reactor, decayed fot 10 years af ter shutdown.
At the end of the section, summary results f or graphite arising f rom the ref erence Kagnos reactor, the reference ACR and the complete Magnt x decocaissioning programme are presented and discussed.
7.2 Options Considered The management options considered are shown in Table 7.1. They are disposal on the deep ocean bed, two kinds of dsep geologic disposal, shallow land burial, and incineration. All options except incineration consist of packaging af the graphite followed by disposal. Incineration would involve
+
atmospheric dia;harge of 3R and I"C (plus traces of other radionuclides) followed by disposal of a much reduced volume of packaged asi by one of the above methnds. These schemes and the environmental pathways which would lead to the exposure of aan are broaaly outlined in Figure 7.1.
The half-lives of esveral of the major radionuclides arising in the graphite vaste are greater than 1000 y, much longer than any reasonable surveillance period. Thus storage of activated graphite can only be an interia measure en routu to disposal. and its radiologi:al 1spect is not considered in this study.
7.3 Radiologi, cal Concepts 7.3.1 Decipfun-makint in radioactive waste management The predictet radiological impacts of the alternative management options f or graphite wastus will be a major f actor in choosing between thst , and since the decision-making process itself has an important ef f ect on the objectives and structure of the radiological assessments it wili now be briefly discussed.
Ra ': slogical tapact is only one of the wide range of ^ echnical and non-technical factors which need to be considered when asking a oecision on radioactive waste management . Some of these f actors will be specific to the reactors at which the westes will arise, and to the available disposal facilities; therefore the optimum waste aansgement mode will depend on the specific circumstances, and there is no single universal optimu- strategy.
The role of this radiological assessment is to demonstrate generic metnocolssten which may then be applied in specific co* tents, and to identif y where more information is needed in order to do so.
The radiological aspects of deelston making are heavily influenced by the systes of dose limitation recommended by the ICRP . _The first ICRP principle is that a practice involving the use of ionising radiation must be
" justified", le it must provide a positive net benefit. However, justif ic a-tion of radioective waste management . takes place in the wider context o f justification of the practices that generate the waste in this case, the generation of nuclest energy by gar graphite reactors. Given that the graphite weetne must be sanaged, the second and third ICRP principles are more relevant to deciding how this may best be done. These aall for all e ,posures to be as low as reasonably achiavsble, econoeic and social f actor s being tak en into account (the so-called ALARA principle), within an over-riding cot.straint that doses to individuale must not exceed the prescribed limits.
There are dif ficulties in directly applying the ICRP system of dose limitation to the disposal of solid radioactive wastes. The ICRP system wa s conceived mainly in the contexts of occupational exposure and the exposure of the public due to routine discharges of radioactive ef fluAnts. The s e exposures occur either concurrently with the practices that cause them, nr commence soon af terwards and steadily decrease in ceatreet, the radiological 1spect of most modse of solid waste dispssal is expected to be almost sete in the short term, and the main taiiological fopact le predicted to arise o,,y much 1 ter. l.lso, exposures of tl 5 public due to effluent discharges are-virtually inevitable once the dischstges have been made, whereas the occdrrence of any radiological tapact f rom solid waste disposal cay depend on a series of utlikely (or at least uncertain) events,-the
, probabilities of which would vary with t 14e af ter d ;soosal.-
1he ICRP systee of dose limitation ( is based on the f undamental principle of keeping riska to ecceptable levels, and in order to extend it to de il also with solid waste sanagement this principle needs to s retained. Risk in this contert is the probability of occurrence o" r. arm, botn to individuals and to the population that is exposed owing to t he practice in quemtion.
Given that re should be limits on the riska to individuals arising is os solid waste disposal, it is generally acc9pted that the same limit s i
e
should apply to individual risks arising either now or at any time in the future. The main role of sn individual risk limit is to exclude unaccep t-able waste management options froe the subsequent stages of decision-making.
The ALAKA ptinciple takes account of "all exposures"( - or in extended form, all risks and embraces both individual and collective risks. In radioactive waste managssent its role is to identif y the option which, in the specific circumstanceh is the optimum f rom the radiological protection point of view. There are several techniques for doing this, but first it is necessary to carry out radiological studies of the proposed options and to predict the individual and collective desse and risksi more detailed technical requirements for radiological assessments are discussed below in section 7.3.2.
Costs are also an important input to decision-saking, and by making ,
judgements on the cost of radiological detriment it becomes possible to compare the radiological impacts and costs or the respective options within ,
a coe. mon f ramework of cost benefit analysis . By making f urther judgements on the cost of future detriment, options whose costs and radiological detriments are predicted to occur at widely dif f erent times ma y also be compared .
However, quantitative analytical techniques such as cost benefit analysis can not be used ditectly to aske decisione( }: they are only aids to decision-making. In this role they can provida queatitative inf ormatton on certain aspects of the probles, and can help identify some of the judgements which have to be made in reaching a decision, but which otherwise might remain implicit and unexamined. Ove r-reliance on quantitative techniques any lead to biased decisions, since such techniques (or their proponents) tend to concentrate only on those aspects of the probism that are quantifiable and consensurate - implying a decision to omit other aspects which could be of at least equal relevancs. More comprehensive aida to decision-a, ing are those techniques which allow widely dif f ering aspect s of the problem to be balanced against each other . But none of these techniques can supply ready-made decisionst they serely aid decision-makers.
7.3.2 Requirements for radiological assessments At preseni there are no incarnationally agreed radiological pro'ectiua c riteria for solid waste disposal against which the results of the assess-mente can be judged. The approach adopted in this study is to celculate the maximum potential individual doses, whenever they might occur and without IN
a; l
l I
indications of their probabilities of occurrence. 'hese potentia 1 doses a re l then comparad with the dose 11 tilt reco;smended by the ICRP f or members of the l
- 'public -(an annual ef fective dore* equiva).ent of 5 mSv). and against the lJ _
objective that the annual- dose received by an individual over a term if i .several years should not exceet. I sSv . If the potential doses from a grapnite management option are such smaller than either of these critetta, that option can pass forvald te the next stage of decision-making. Eve n i f
, the potentini ihdividual doses are close to or above :he dose limit, they may still represent accept 6bly low individual risks if their prchsbilities are low enough. In this generic assessere.t. potential individaal doses are calculated assuming that specific sequen'es of events take piece, while recognising that these doses and their l robabilities can only be estimateo
, satisf actorily in site-specific con:extr .
Collective doses to the entire expo ted population are also calculated an functions of time after disposal, for ..ch of the graphics discosal
. options.- 14 a context-specific assessment these twiculations should be combined with predictions of the probabilities of the doses being received (again-as functions of time), resulting in risk predictions which could then be used as an aid to deciding on the optinua option. An important f actor in judging the incortancs of a collective dose (or risk) it its distribution with respe; +=dividual dose rates the same collective dose may be judted more imports. t' is distributed among a relatively small population, implying high. ..w ese individual risks ( . Using exis ting methodologies ,
it has not been possible to predict distribuitons of individual dose rates for all the graphite. disposal optiens considerad.
Estimations s' the uncertainties in the above predictions of individual Land co11cetive riske should also be nede, and could be used to indicate the importance to be attached to the predictions when weighing all the relevant factors in the overall decision .
7.4 ~Ceneral Features of the Models Used All the pathwsys which could lesJ sw U.s significant exposure of man f or each option'should be considered, and those considered in this study are Itsted it. Table 7.2. The initial dispersion of activity into the environ-ment is treated (with two exceptions) by physical models which describe
In this report "does equivalent" and "ef f ective dose equivalent" are contracted to " dose" wherever possibie.
e
.e explicitly the processes of dispersion and' reconcentration during t rar. sport l through the environment to man. Such models give the magnitudes'arJ time l distributions of doses received via dif f erent pathways by individuals and =
population groups. The exceptions are the transfers of I"C afd 3n ihtough f ood chains f ollowing dischstge to atmosphere. For these two n adionuclides ;
the dispersion of the discharge plume is endelled physically, but the subsequent transport through food chains to man is estimated using a I Edif f erent approach which implicitly takes account of all the significant food chain pathways.
Amonf the radionuclides produced in significant amounts in the graphite, I"C and 3 H are particularly mobile. Separate models, described in the next section, are used to assess the radiological impact of their global circulation. The totti radiolcgical impact of a given management option is obtained by adding the contributions of all radionuclides, calculated f rom the appropriate initial utspersion models plus the global circulation models where necessalf, 7.5 Clobal' Circulation 7.5.1 Introduction The gloeal circulation models for I"C and 3H are common to the assessment of all the disposal options considered, and are therefore presented first.
The necessity for a global circulation model arises when a radionuclide can potentially become globelly dispersed f ollowing disposal, bef ore it has decayed to negligible levels of activity (Section 7.A). The two radionuclides present in significant quantf. ties in activated graphite f or which this may be the case are I"C and 3H The sichal circulation models taka advantage of the pervasive nature of carben and hydroten by assuming that I"C and 3 H enter natural global transport cycles, mixing rapidly with stable carbon and hydrogen to establish unif orm specific activities throaghout large compartments of the environment. Man's intakes of I"C and 3
H are inf erred f rom the total intakes of carbon and hydrogen. thus implicitly taking account of all pathways.
The assumption of rapid mixing in the global models is invalid in the short ters. and the global modet has little ability to distinguish betweer.
different modes of release hence the need for a range of regional models for the initial dispersion of activity. For each waste sanagement option, collective doses calculated by the appropriate regional model are added to g
l those calculated by the global sodel, taking care that no contribution is coitted er counted twice.
7.5.2 Methodolcates and models The global ciraulation of I"C has been calculated using the model in ref erence (4). 1"C is assumed to be uniformly and instantaneously mixed with the erstire carbc't content of the environmental compartment into which activity is released and then to follow thei natural global carbor cycle.
This model is auch less elaborate than those which seek to predict the relatively short-ters bahaviour of industrially-released fossil CO2 and of weapons 1*C. The short-tors behaviour of I"C is treated in this study by separate tagional sodels and little reliante is placed on the global model until the end of the initial dispersion of activity. In the longer term, all global l*C models tend towarda agreement, since the ultimate distribution c f I"C is determined by the sites of the major carbon reservoira anc the rates of a few tiow processes, all of which are quite well established and are adequately represented in the chosen model.
The irradiation of aan by I"C occurs chiefly through ingestion and inhalation, ei:ternal doses being trivial because of the low penetrating power of the il radiation emitted. Han's intake of I"O has been determined assuming that the specific acti-ity of all the carbon ingested or inhaled t o equal to that of the " circulating carbon" compartment of the sopropriate heulsphere. A fixed global population of 1010 has been assused, distributed in the same proportion between Northern and Southern hemispheros as the current popi.iation, ie, 82% in the North,181 in the South ( } . Aver ag e annual intakes of carbon have been taken froe reference (6). Otter relevant 1*C data are given in Table 7.[" .
The model ased for the global circulat1&n of 3H is sta11ar to than given in reference (10), and has been modified te increase the accuracy of mass balance. 3 H released into the er.vireinsent (whether to the atsoepheric or the aquatic environment) is assus.ed to be inssediately dispersed and exchanged with the hydrogen content of the circulating waters of tre hetaisphere into which the discharge is made. The subsequent circulat.icn of tritium is then datermined by the exchanga of the circulating waters between the two hemisphersa and the deep oceans.
The intaka of 3 R by man is determined assuming that all the water taken into the body has a specific activity of 3 H equal to t, hat in the circulating waters ot the appropriate hemispherd. Relevant data corresponding to those given above for I"C have been taken from the same sources.
l
s
-1 i
I 7.5.3 Re sult s Doses arising f rom global circulation f ollowing the release of unit activity (1 73q) of 1*C intu the troposphere, the surf ace oceans, or the deep oceans of the northern hemisphere have been evaluated. Similar&y doses arising from the release of 1 Tlq of 3 d into the circulating waters or the deep oceans of the northern hemisphere have been evaluated.
Individual dose rates from 1 T3q releases of either radionuclide are very low at all times (f or 1"C e 10-10 Sv/as and for 3He 10-15 Sv/s).
The development of collective dose commitment is shown in Table 7.6 for l*C and Table 7.5 for 3 H For equal releases of 1 TBq. the predicted global circulation doses from 1*C are very much greater than thoas from 3H.
Figure 7.2 shows the development of the collective dose comattmant f rom releases of I"C to _different cospcetnests of the environment.
In the f ellowing sections, appropriate models f or each graphite management option are used to calculate botit the doses due to initial dispersion'of activity and the rates of entry of 3H and I"C into global circulation. The latter are used as an input to the global-titculation-models. and the results f or initial dispersion and global circulation are then combined to give the tossi radiological impact.
7.6 Disposal on the Deep Ocean led 7.6.1 Methodologies and sadels Two models ari used in this study to eatinate the movement of activity after release from a disposal site on the bed of the deep ocean.
Pluse sodel The plume model, which is based on suggestione"" made-by an Mvisory Group set up to review the oceanographic basis of the IAgA " Definition and -
Reconnendations* under the landon' Convention ( , is used to assess the ef f ects of the possibility of upwelling or plusing, by which a 1s"go volume of water night be rapidly transported from the deep ocean to the surf ace waters.- It is assumed that these possible adverse dispersion conditions may arise at say time af ter disposal, so long as loaching of activity f rom the disposal site continues.
It een be shown(I that the activity concentration in the pluse can be approximately related to the rate of release of activity, under a wide range of disperaton conditione, by a constant f actor of' 10 Iq/m3 per Sq/s. Thia relationship embodies no assumptions about the mechanissa by which a plume might be generated. A recent report ( considers that such plues formation le unlikely. The plume model does not take account of sedimentation, e tnce e
the removal if activity to sediments is slow :ompared with advection rates in the plume.
Lse r- t s rs e nd el in evaluating the long-ters disposal of activity in the oceans a f our cespartment audel to used whtch takes account of tht interchange between the surface waters of the Northern and toutbarn hestepheres, the interchange between these surf ace waters and the deeper waters below the thermocline ,
and the interchange between the deeper waters of the two hentophores.
Except for I"C. activity is soeused to be transported f rom one part of the ocean to another in proportion to the water flow. Thus transfer coef ficients have been deterstnad from stailar considerations to those given in ref erence (15) for the global transport of 3 H However a more appropriate allowance is made for the water volumes of four boxes so that at equ111b71um, the concentration of activity is the sano in all four boxes. The vertical distribution of carbon in the ocean to not uniformO0I, so the use of water transfer coefficients for 1*C is inappropriate. DC transfer coef ficients are determined f rom ocean transport data given in reference (15) and carbon distribution data given in reference (16).
Allowance is made for the loss of activity to sediment using a particle scavenging model(10! The fractional annual losses of activity f r,em the deep ccean boxes to sediment are determined using the data given in ref arance (10) for large oceans, and the distribution coefficient f rom water to sediment of each particular radionuclide. However ref erence (10) makse no allovence for the high levels of sedimentation in coastal unters. The inclusion of the effect of coastal vatare in the North East Atlantic is to increase the rate of lose to sediments in that area by about en ordsr of magni t ud e. This being the caaw, the large ecean values derived from reference (10) are increased likewise, 3 H is assumed not to be af fected by sediment processes, which are also too slow to have an appreciable ef fect on 1*C. Radionuclides which are predicted to be removed to bottoa sediments
~
are assumed to remain there. This assumption is known to be unrealistic for some long-lived radionuc11desE I but it does not significantly affect the collective dosa couaitsente calculated for graphits disposal.
Both long-tors and plume models lead to estimates of the activity concentration in the surf ace w,*ers of the ocean, which are applied as follows. Plume advoction could theoretically lead, for brief periode ',
to abnormally high activity concentrations in relatively eas11 areas of surf ace waters, so the plume model is used to esiculate the maximum potential doses to individuals. Transient effects of this type could occur at any tins within the period et release but would not contributo e
. __--_.-o. - _ _ - - . _ _ _ _ - _ _ _ _ _ _ _ . _ - - _ - _ _ _ _ _ _ _ _ - _ . _ _ . _ _ . _ _ _ . _ _ - - _ _ _ _ _ _ _ _ . _ _ - _ _ _ _ _ _ _ . - _ . _ _ _ _ _ _ _ _ _ _ _
oignificantly to the collective dose cosuniteent in the long teria. The long ters model is assumed to be valid at all times af ter o'.sposal, and is used to calculate collective doses and to indicate the long tera trend in individual dose.
The time dependence of the concentrations of activity in surf ace waters
-is determined not only by transport from the disposal site on the ocean tcd.
but also by the rate of release of activity into seawater f rom the solid waste. In this study leaching and diseslution of graphite or of incineratst ,
ash (section 7.9) by seawater are assumed to begin immediately af ter disposal, and since the long-tera relsace characteristics of activity f rom neither waste fers have been fully established (section 6) the results are calcalated for an assumed range of constant fractional release rates f rom 10-1 y~l to 10-* y*l.
Given the time dependence of activity concentrations in surf ace waters, the calculation of doses to man f rom ingestion of fish, crustacea and sollusca f ollows the methodology of reference (13), using conentration f actors listed in Table 7.6 10,13 .17,18.1 M . Seafood ingestion rates are
-determined from total world fisheries statistica . . a s s uming t ha e the ,
cdible f ractions of fish, crustacea and solluece are 0.5. 0.35 and 0.15 respectively (D.F. Jeff ries, Flaheries Research 1Aboratory, MAFF. Ivestot t :
private communication). Maxiseus potential annual individual doses are calculated assuming an individual intake of 500 g/d of fish, or 100 g/d of crvatacos or acliuses.
Doses arising from thu sediment pathways have been calculated using
- the watertsediannt concentration f actors giver in Table 7.6. The cociponen t j; due- to external gaens irradiation has been calculated using the sethodology given in reference (21). Collective dosee are estimated assuming a. mean annual exposure ties of 2 x 1010 perestr hours, and f or maataua pctential.
annual individual doses an exposure time of 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br /> t's assumed. Idees arising f rom thu inhalation of resuspended beach sediment have been calculated assuming an individual sediment _ inhalation rate of 2 i.gid l Foi collective doses an exposed population of 4 x 104 is assumed.
L 7.6.2 Results L ' Maximum potential annual individual doses arising f rca the deep ocean bed dispossi of tha 10 year-cooisd graphite f rom the reference Macou
,, reactor are listed in Table 7.7. The external samma dose is do ainsted by
~
80 Ca and would be auch reduced for the lor.ger decay times before disposat, but doses via ingestion pathways are dominated by the longer-lived l*C and
!- thus would not be significantly reduced.
! - to -
l.
. . . =
The developeent of collective dose consiteent arising f rom marine par
- ways f ollowing " regional" dispersion in the oceans is dominated by doses from 1"C. The results presented in Table 7.8 include the global circulation of I"C and 3 H again these doses are dominated by I"C. Because 3"C la not significantly lost to bottom sediments the osission of return f rom bottom ,
sediments (settion 7.6.1) is not isportant. The co11cetive dose f rom all the other radionuclides assuaifig no loss of activity to sediments is still lower than that due to l*C alone.
7.7 Disposal in a Deep Coologic Repository 7.7.1 Methodologies and models There are many conceivable circumstances in which activity disposed of in a deep geologic repository could re-anter the biosphere and irradiate man. However, the majority of possible scenarios including those which are considered to be the more probable (22) , share the common f eature of
< transport' to the biosphers in moving groundwater. Only this mechanism to e x amine d.
Consideration has been given to the disposal of irradiated graphite itself, and to the solidified ash e asts arising f rom the incineration of
, irradiated graphite (section 7.9). Since the long ters release che.racteristics of activity f rom neithe waste f ore have been fully determined (section 6), calculations are perf ormed assumin5 that all waste is subject to the leaching action of groundwater tunedtately f ollowing disposal at the same range of *ractional release rates considered f or oceam bed disposal.'is from 10-1 y-1 to 10-* y*l. Subsequent to teaching, the mapr transport . processes taking place in the geosphere are advection.
dispersion and sorption. These processes are represented te the geosphere transport model ' by the groundwater webcity, the dispersion conflicient, and the reta7dation coefficien~..
Two dif ferent types of disposal site are considered with dif f erent hydrogeological enaracteristics, and pathways for-the return of activi':y t o man throwgh the biosphere.
I Inland site-The inland repository is assumed 'to be sited in a clay otratua it, whttn the groundwater velocity is very low. The dispersion coef ficient is essentially constant at these low groundwater velocities and has been
! determined from reference (23). Values assumed f or the hydrogeologic l parameters are given in Table 7.9. Ratardation coefficients have been
~
determined from published sorption data . assuming the ratio of l
l n
4 -
denstty to porosity for clay to be 3.5 g ca* . and are given in Table 7.10. Once the groundwater la released f rom the clay st ratum it is 4snum d to be rapidly transported to a f reshwater body used by man. The wat e r bod y is assumed to have a flow rate of 6 x 106 ,1 y-t serving a population of 1.5 x 10*; this corresponds to a supply f rom a er.411 reservoir. Th e ra t e s )
of discharge of radionuclides into the reservoir, together with the asseeri flow rate, detr:1mine the radionuclide concentratione in the water.
l l
Collective doses are calculated f or the drinking water pathway only, i assuming that the popi.lation drinks a total 30 m /d;3 collective dose n i arising f rem the ingestion of contaminated f resh fish are unlikely to be l significant . ladividual doses are calculated f or both drinking water ano f reshwater fish pathways, asstating individual intakes of 2 t/d of wate r and 110 g/d of fish.
Coastal site l The coastal repository is assumed to be sited in shale. He re w.s t e t movenent can he relatively f ast and the dispersion coef ficient i s a s s un.e d t o be proportional to the grounc'wster velocity (Table 7.9).
At a coastal site the hydraulic gradient is assumed to be towards .ne coast, leading to a discharge of contaminated groundwater into coastal w ters via the seabed. Doses arising from the ingesticia of seafoods, and f rom beach sediser.t pathways, are esiculated using the methodology given i n reference (13). Marine data have been chosen consistent with those acopted f or deep ocean bed disponsi (section 7.6). -
As regards the sensitivity of. results to parameter values, the main ef f ect of varying the hydrogeologic parameters would be to vary the time before agrise of activity into the biosphere. Flow path length and ground-water veloc.ity have been varied together to increase and to decrease the y transit time f or groundwater f rom the repository to the biosphere by a factor of ten for both sites.
7.7.2 Results l Resulta for the inland disposal of the ref erence Marnom graohite cooled f or 10 y are preseni.ed in Table 7.11. Maximum potential annual individual doses arise f rna the f reshwater fish pathway and are ses..l. Re gional collective doses are dominated by the presence of 36 C1, und are sensitive t o l the groundwater transit tics but not to t!ae f ractional release rate. The final column of the table shows the ef fect of including roses arising f rom global dispersion of IC which is released into the envi:onment before decay occars (all 3 M decays before release). This is only algnificant f or the l
l l .
shorter groundwater transit time, the quantity of 1*C released is then also af f ected by the f ractional release rate.
Results for the coastal disposal of the same material are presented in Taole 7.12. Individual and collective doses are dominated by the ingeselon at l'C in fish Doses are sonettive to groundwater transit time and to the f ractional release rate, particularly for the shorter groundwater transit times. As bef ore, the contribution f rom the global circulation of l*C released to the environment is added into the final coluan of the table.
Because of the shorter groundwater trenett time f or the coastal repository, the D C is released to the environment before decay and plays the dominent role f or al'. combinations of assumptions, 7.8 Shallow Land Burial,
?.3.1 Methedology and, models The shallow land burial f acility le assumed to be a f ully enntneered concrete structure incorporating engineered barriers against infiltration of unter (Figure 7.2), and located in a clay f ormation witti relativelv tow froundwater flow (Figure 7.3). The f act11ty and the methodology f or the radiological aesssement are described in detail elsewhere *
. and only a summary is given here. A eleplified verision of the original methodology has been used, which omits the probability calculations f or consistency wit h the other methoGelogise in this study.
Once disposal operations have ceased and a bur 131 site has been closed, the use of the land may still be controlled by local ur govercasental legislation. However, any initial rastrictions canrot be guaranteed f ar all time and the area night eventually be built on, f armed, forested or used f:r recreation or other purposes. Also, grouncitator f rom the site may run of f into stresse eed for drinking water or for fishing.
For the este characteristica assuaec, only the building pathway could lead to significant doses in the first decadue af ter dispoesi, and it **
feasible to guarantee restricted land use at those cines. The soproach adopted in 'this study has theref 3re beau to calculate the minimum restriction tiate needed to prevent desem via the ct1 ding pathway in excess of the 5 mSv dose limit f or members of the public (including building worke r s ); and f or the f arming and water pathways to calculate potential individual doses arid collective doses assuring that the least f avourable patnwdys are operative.
i 1
_ - _ _ _ - - . _ - _ - . _ _ . _ _ _ - - _ - - - - - - - - - - - - - - - - - - - - - ~- -^~ ~ ' ' ^ ^ ~ ~ ~ ' ~ ~
l l
Sullding fathways if the site were tn be built on, the radiological impact would depend on the entent to which the building operations disturb the waste and the surrounding clay strata, which any have become contaminated by groundwater transport of radionuclides. Nine claesou of building were considered.
according to the depth to wh.4h the foundations must be excavated, and the results are substantially the same for all classes of building requirina scre extensive encavation than residentisi dwellings.
The individual does that could be rwesived in the absence of sitt restrictione has been es1cu14ted f or a series cf times af ter disposal, arid ts the annual dose received by building wotters f rom inhalatinn and externti irradiation while excavating the trench containing the waste. From this inf ormation is derived the minimum restriction time required tu ensure tha t annual doses f rom the building pathway would be less than 5 mSv.
Farming pathwave Contact of the graphite waste by groundwater is assumed to teach activity at one of a series of fixed f ractional release rates ranging from 10"' y*1 (suction 6.5) to 10-2 y-1, ;he groundwater is predteted to f ollow flow paths to the surf ace se iadtcated in Figure 7.3 and miar.st ton of the leached radionuclides is predicted by a linear sorption model .
The rising groundwater enters the ruroff tone close to the surf ace, where it is diluted by rainwater infiltrating f rom above. Some of the grounowater rises to the plant rooting sone closer to the surf ace, and thus the radionuclides can enter crope and antral products consumed by isen af the site is farmed.
Doses to iadividual seabers of critical groups are estimated using the petsizistic assumptions that the individuals take their et s tre suopises nt the f oodssuf f s in questice f rom the area contaminated by the aroundwater f rom the sit e; the site of the critical group to thus limited by the agricultaral productivity of that area. Collective doses f or ch a respect t ve food pathways are esiculated by escuaing that only one type of f ondstuf f would be produced on the entire area, and that all the produce would be co ns ume d.
The nature of the graphite waste requirea one 6eparture f rom the oechodology desertbed in references (30) and (31). In predteting the concentrations of radionuclides in the rooting zone it can normativ be assumed that they are present in trace quantities which do not perturb the ambient soil cheatstry. However, that assumption is not always vaind !or b
C lanching from graphite, which is essentially 120 As s uming consruent
- 74 ,
_ _ _ _ _ _ - - - - _ - - - - - - - - ~~
- . - .. - ~ . . - -.- _ -- , , ~ . ,
n
!eaching of I'C 'f roo the I2C metria leech rales greater than about 10*5 y"'
would lead to increases in the normal concentration of soluble carbon in the rooting tone. In such cases tho'sanimum possible.speelfic activity of l'c available for root uptake would be that in the graphite itself, and stis sets an upper limit on the possible doses via f arming pathways (all other assusprions tenaining the same). *he specific activity of l*C available f or root uptake may well be very such lower than assumed, because the caroonate in the cuncrete of the vaste packaging and the trench structure wouad alan cnntribute to the isotopte dilution of I*C. However, in the absence of
- tr, formation on the relative rates of dissolution of ennerete and graphite thts effect cannot be modelled at present. Similar ef f ects may cont ribute to teotopte dilution of 3'C1, though these are likelv to be less tecortant ,
because the repository would not be the dominant source of dissolved chlorine in the soil.
Al though' the presence of ' the large gutntity of !2C to an imporeant >
fcctor in the prediction of peak individual doses from I"C in shit tnw tend burial of graphire, it has lees ef f ect on col',ec* tve doses. .f or two reasons.
44rstly, although isotopic dilution reduces the race of uptake ot' C f r9m the soil. nd thus reduces peak individual ioses f rom this radionuci tde, 1*C would be tatained in the soil f or some time and cc.1d still contribute t a collective cones. However, the magnitude and time-dependence of this ef f ect
- cannot os predicted by the simple auil modela prasently available.
Secondly, the sollettive dose commitment (integrated over all f uture time) -
would be dominated by-the. subsequent contribu' tion f rom global circulation of
!"C. Thersf or e the 'ef f ort of I"C isotopic dilutisu upon collective dowes has been ignored..
The output f rom the madel fnr radionuclidn migra'1un in the not; and
- rooting sor,es la veed to iterive the concentrations of radionuclides ta gran s and crops grown on-the soil b Relevant concentration f actors are given in Table 7.13. . Six diff erent typse of f arming have seen consideredt detry, beef : sheep root erwpa, grcin and green vegetables. There a re ersortat ed t M_the following routes for the asposure of man:
(t) :onsumption of aest, liver and milk taken f rom animals grazing on the sit 4; (11) cu mumption of crope grown on the site.
It to azaused that only one type of f arming would be practised on the atte at any ane time. Individual doses are calculated assuaing that the
- individual. derives all his food associated with i particular f arming -
practice from the site. Maaisua individual doses are those obta13ed f rom l
l i.
l- g
. - . .- . - . . , - - . . . ~- . _ . . -
i
-I l
the- f arming practice giving rise te the largest individval does. Collective 1
-deses.. art calculated assuming that all the food produced is consumed by l
-)
man, and only the results f or root crope (which give the largest rollective ;
dose). are presented.
Water pathways 1
The hydrogeology assumed f or the anallow burial f acility (Figure 7.3) envisages the presence of local stresse into which the contaminsted groundvater will eventusily flow. The f ollowing exposure pathways arising f rom use of these streena nave been considered: 4 (1) consumption of drinking wateri (11) consumption of fish 4 (111) conausption of produce taken f rom animals which drink f rom the strees Individual'and collectiva doses are calculated, the latter by ass:alna'that a fixed f raction (0.171) of the total flow is consumed by man an drinkinu vater. ' Appropriate allowance has been made for the isotopic' dilucion of i'C-by the 12C in the graphite. Further isotspic dilution by 12C in the st ream does not alter the predicted doses from I"C8 nor does dilution by 35 C1 in 4 the strema siter the predicted doses f rom 38C1.
7.8.2 Results Results f or the shallow land burial of the ref eren+ e Magnom venctor graphite cooled for 10 y are presented in Tables 7.14 and 7.15.
Table 7.16 refers co the groundwater-related pathways, and f or a teach rate of 10*2 y'l gives the maximum potential annual individual doses. Th e
" total" collective dose coenitaant is that predicted f rom f arming (root crops -only) plus the surface-water pathw&ys, plus global circulation of I'C.
Although the ' predicted peak annual. individual doses are of the order of the dose limit f or moabers of= the public recoansaded by the ICRP(2) .' the predictions are bcsed en a' series of poesialetic seausptions (discussed below). Therefore no specific conclusion _about the acceptability of shallow
- lar.d burial of graphite wastes can be drawn f rue such a result artsina f rom -
a generic assessments site-specific assesseents with appropriate models and data would be required.- The collective doses given in Table 7.14 are
. dominated by the contribution of 1"C via root cropei if there were no f arming on the utte the collective dose consiement woald be essentially that due to global circulation of tne I"C, to about 5000 man Sv.
The results presented in Table-7.14 are based on a series of pessimist'ic aseuapcions. Even so the doses predicted to arise f rom most of L
e-
- ~ . . . . -
t, I
the radionWC11 des in the graphite are very low. This is not necessarily the case f or some of the. ingestion pathways for !"C and - 36C1. and the uncertsar-ties requiring the adoption of pessimistic soeusptiens are . .asiderable. No allowances have been cada for retention of 1"C either is the repusitory or tn.the gevaphere, thou2h each could have a significant effect in reducing d:se s; for example there are clays in which the retention of I'C would be sufficient to schAeve very low releases to the biosphere. It he6 also been assumed that l*C in the soil is isotopically diluted unly by the carbon f uund as dissolved bicarbonnte (which comes entirely f rom the graphite f or the range of leseh rates assumed). In f act ther.' would be some exchanite Ntween dissolved l'C and the !!C in the sol) solids, and eithough the extent of this exchange-is= ttnt clear it could result in fatther significant
" eductions in predicted dosse. Similar uncertainties are present in the
.no6elling of I'C1. though these are lese important than for I'C because N Cl becomes the dominant radionuclide only at the h.ghest teach r$te assumed, h ' effects of lesch rates in the range 10 10" y"I on the ma ximum potenttal individus1 desee are diff arent f or the f arming and surf act-water pathways.. If the f Arming pathways vero operative at the timon when activit y 15 being teltased into the subsoil, a small reduction in leach rate below i
-2'y-1 would aske 1"C (!n grain) the fon61nent radionuclide but further reductions would nos change the predicted peak annual dose f rom the limittw v41un of 1.1 x 10~ Sv determined by isotopic dilution of l*C in the graphite itself. If the f arming pathways were not operative, maximum potential individual doses would occur de the surface water pathwayn and L would- be ruduced.approximately in proportion to the ;e.cn rate.
If drinking water was the only operative pathway and the leach rate wan j ,. 10** y-l. t5en the peak potential individual dose would. be' 4.3 x ;0-' sv I- 'from M C1 vis the' fish pathw6yg this is the basis f or the lower ends 01 the ranges of predicted potenti .1 ind3vidual doses f ree hallow land bw.tal 41ren later in the ausamry tables 7.21 - 7.23. The ef f ects of teach r nii io.wr'tean 10-2 y- t on collective doeos are 1eri. pronounced than f or individual dossa, owing to the overriding influence of the global eteculatton of l*C.
t l Tabis 7.15 presents the results f or the individual doses tha: could arise at various times af ter disposal, urless building was restricted, h entern.1 trradiation contribution, from 60Co. 15=Eu and 10 8 mag - is much greater than that via' inhalation.
It is clear f rom Table 7.15 that restric-
- ton of the use of 'he site f or a f ew decades af ter decommisatoning would
.creat ly mitigat e the individual risks due to building, and the ar...uai dose vould f al t below 5 mSv af ter about 30 y. le 40 y af ter reactor shutdown.
l 1
i
- 7. 9 Intineretton 7.9.1 Mer,hodology end models incineration of activated grr.phite will lead to the discharge of some a activity to the steosphere and an accumulation of the r,st of the settva;v
- n ash and fllter material. in this sec. ton the radiological impact o f i n.,
discharges to atmosphere 11 assessed. The radiological impact of the disposal of ' the solid wastes is estimated f rom sections 7.6, 7.7 and 7.5.
It is clear f ree section 3.1 that the bulk of the activity released t o the staosph'ere is A"C and 3 H Releases of other radionuclides are very man smaller because of the generally smaller arisings and the high proportion which is retained in the ash and filtete. Consideration has oesn limited t nuclides chosen on the basis of activity released, doses per unit iniake, sean gamma energy poi disintegrettun and radioactive half-lif e. The s e
. criteria ansure that the significant radionuelldes assoc 14sted with each pathway listed-in Tabls 7.2 havs been considered.
The methodology used to evaluate radiation exposure resultica f rom
~
atmospheric discharges -is dtvided into two porte a model to calialat e the concentration of activity in the atacephere as it disperses.f tom the potni of .releaset and models to describe the terrestrial pathways which bea t n wt 5 tha deposition of activity from the air.
Collective dossa to ths' population of the EC havc _ been calculated f or each pathway. The maximum annual does to en individual will depena upon ' ne rate at which activity is released to the atmosphere. For this report it as assumed that all' the graphite f rom a ref erence Magnon react,or is inciner.ir ed in 1 year. The critical indiridual is 4ssumed to spend the whole of the relvese patiod within 5 km of the re16sse point. and to derive all hte f ood of- the types considered (but not drinking water) f rom within that area.
The- stoospheric dispersion model ta described in ref erence (10) ana to applied with the following aaendments.. Mixing layer depths, windspeeds a t height 10 m, and corrections to windepeeds to sake allowance f or the hetant of the release are all chosen to be constatent with the recornendationn- or a UK working group on atmospheric diapersion codelling . The value o f
- roughnese length, a seasure of *he roughness of the ground surf ace. is taken to be 0.3 e .
The value of depostaton velocity f or all nucaldes except N C and 3 H is taaen'to be 10'l s/s ; 1*C and 3 N are treated dif fvrent t v (see later).~ Air eencontrations are Evaluatud SS functions of dtStance trum en assumed discharge locattun in north-west England, for a range of dispe s> 4 aton parameters uhich depend on esteorological conditions, and are then avataged with respect to the $nnual distributions of meteorological
- 7A -
Aub , -+
~ . - . .- -
F l
i l
o conatt.ons and wind directions appropriate to that location. Release f rom .
60 o' stack to assumed (section 3.1).
The air concentrations and appropriate Josimetric tioJels have been combined with the spettal distribution of the repulstlun of the CC with '
resnect to the location of the discharge, in order to estimate the collective dose to the EC popalation tros inhalation of activity and f rom enternal gamma irradiation f rom the cloud.
The second model, describing terrestrial pothways following the deposition of activity onto 1 sod surfaces is also described in reference (10). Radioactive asterial depentred onto land surf aces can gtve rise to irradiation of men by three main routes external irradtation, inhalation following resuspension. and ingestion of contaminated foodstuf f r .
in each case the esposure of man depends on the rate of removal of activtty from the surface into deep soils. Following ref erence (10) activity is assumed to eigrate f rom the top 30 cm layer into deeper solla with a half-time of 100 y.
The determination at doses for-all the 'radionu:11 des encept l'C and 3 tl la briefly described below. Where pathway-dependent date are varied f rom those given in ref erence (10) section to made in the tent.
1*C~and:3 P are tenated diff erently and calculation of doses due to these nuclides is discussed later.
The external treadiation to which man is subjected f roa surf ace deposits is evaluated assuming undisturbed soil. The dose rates are come hined wit'n the spatial distribution of the population of the EC countries.
pessimistically assumed to be outdoors continuously, to evalusie collective deem retas and commitments to the EC population.
Resuspenaton of activity from ground surf acts can necur by wind-driven processes or other physical disturbances, caused particularly by man, such as digging and aweeping. Surf ace deposits appear to be.nore readily resus-pended shortly after deposition, 40 resuspension is represented by a tisc~
d epet -lent resuspension f actor initially 10-8 ala and decree =tne v'ta a
- half-ctma of $$ days to a constant value of 10*' m*L- at two years and' be- '
yon #.
AAr con.wntrations arising f rom the resuspenaton of activity are eu.noined with the population distribution to svaluase collective dosee to tne F.C population.
-In ar<delling the movement of activity f rom the inttial surf ace teposit into f oodstuf f s a half-time of 100 y. for the residence of activity in the . -
top 30 cm of soil is acaused, as f or external irradiation and resuspension
'The units of casu.pansion factor are (Bq/mI in att above the ground) pet (Bq/m2 on the surf ace), is-m*I L
l 9
9
, -r
- _. = . - . _ .
l l
pathways. Soperimposed on this depletion term are appropriate ellowances for, inter alta, mixing of the top 30 cm layer of soil by ploughing, deposition end removal f rom plant autf aces. foot uptake, and the J1 rect ingestion of soil by animale. ' Activity in animal pteducts is considered tu nrise fro 6 =;th ingestion and inhalataon of activity by the anincl.
Element-depetsent parameters not given in reference (10). relating to transf er to crops and pasture, and cheap and cattle, are taken f rom-references (34). (35) and (36). Non elenant-dependent parameters varied f ree those given in reference (10) are taken f rom reference (37). Th es e include pasture yield, the food intake c-f cattle, the reter. tion time for activity deposited on plants, and the f raction of activity deposited on plant surf aces as cpposed to soil surf aces. The combination of deposition rates with movement of activity in the soil to agricultural produce. and with sgricultural production erstistics, permite the evaluation of collective dose rates and consitsents to the EC population.
-The transfer of I"C and 3 H between the atmosphere and the terrest rial environment is somewhat more complea than that described above, sainly ,
because of the fundamental roles playsd by hydrogen and carbon in bi o-logical systeem_(10). The modele described above are therefore not approp-inte for I"C and 3R , and a relatively simple specific activit y approach is-adapted to evaluete the transf er through the terrestrial environment to man.
It is assumed that the terrestrial environment comes into rapid equilibrium with the 1"C and 3 R in the atmosphase, so that the specific activity of ' carbon and hydrogen in f oodatuffs is equal to that in the atmosphere as the point of productica. Thus man's intake of activity via ingestion la determined f rom the amounts of carbon and hydrogen in each of the various f oodstuf fs, using the same agricultural production statistics a in the models for the other nuclidae. Han's intake of activity by inhalacion is determined from his intake of carbon and hydrogen (as carbon dioxide and water vapour) in-the air .
The use of a specific activity approach f or 3*C and 3 H gives no infornation un the temporal distribution of the dose, which in reality may be delivered over an extended period. 'It is irplicit in the modul that any h doas is only delivered while tha specific scrivity in the air is maintained.
!. te.during the release.- la reality doses are likely to be lower than pre-i dicted during the relean'a but activity la likely to be incorporated int o L
V plant detritus and soil systema, giving rios to exposure f rc soun time af ter the release.
l'
~l l
l
-l 7.* ','Ra'ults a for senospheric _discherats 1
The maximum attnual dose to an individual f ollowing the incineration in 1 y of all the 10 year-cooled graphite f rom the reference Magnon roattor is
-3 x 10-1 S v. - This dose is predoeinantly due to the ingestion of 3"C. Itere is a minor' contribution frou bl and the.other radionuclides give rise to very.small domes indeed.
-The collective dose' commitment arising f rom the regional dispersion of ,
the same releau A s 4.2 x 10'l man Sv. This valve is attaineJ by the end of the year of relente (according to the assumptions of the model used) and is ;
again dominated by I"C.
Results show that there vould have to be a complete 4 f ailurn of the filtering system for the contribution of other radionalides
=to be important..
~ The development of total collective does consitzent, including the cc.etribution f rom the global dispersion of I"C and 3 H is presented in
-Table 7.16.-
7.9.3 Dispossi of solidified coh arisina f rom incinaration As hoted in section.S. the ash from combustion of the graphite contains essentially all the non-volatile radiocuclides but say also contein t races of unburnt graphite. The exact quantity of the latter is uncertain.. bu t it s .
I"C content may be radiologically significant. In view of this uncertainty
. the radiological . impact of disposal of. totally-burtit ash is assassed firs t, and then the effects of traces of I"C are considered.
Dispos,a1 on the deep ocean bed Maximum potential annual individual doses arising f rom the deep ocean bed: disposal of the ask following incineration of the 10 year-cooled graphite f rom the reference Magacz reactor are very similar to those pre-dicted for the same disposal of the graphite itself. The development of
-collective dose comattaent arising f rom marine pathways f ollowing dispe*sion in the ocean in presentcW bble 7.17. and in the absence of I"C these doses would be dominated by JIC1.
Disposal in a deep zooloale recosito'ry o
Results .f or inland disposal of the ash are presented in Table 7.18.
- taximas potential annual individual doses arias f rom the f reshwater fish
. pathway and are small. . In the absence of I"C. desse would be doolnated by the presence.of M C1; the' predicted doses are sensitive to the groundwater b
c transit time but not to the fractional release rate.
Results.for the coastal disposal of the same saterial are presented in ll .. Table 7,19.
In the reference case tha individus1 doses are dominated by the tngestion of (Je sag la fish and the collective dose by 3'C1. For the I
~
L -
longer groundwater tranett time the individual and collectivo desse are jostnated by 30 C1. and for the shorter groundwater transit time the individual dosse are desinated by 10$ sag and the collecttve dose by 188 8Ag and 38C1. Again potential individual doses are small, and in this caso doses are sensitive both to groundwater tratait tise and to f ractional release rate.
Shallow 14tjl byrist The highest potential doses would artes f rom building on the alte during the first f ew decades ef ter deconsiostoning, unless the une of the ette wee restricted and since the radionuclides responsible (mainly 'CCo) wuld have remained in the ash the results are virtually identical to those f or the packaged graphtte tr. Table ?.13. The attivity concentration in the sah would of course be grsater than in the graphite, but if th4L part of the f act11ty was excavated the time taken would bn correspondingly shorter.
leading to sesentially the saae esposure. At later times the mantmuu potential doses would be trea 38C1.
7.9.6 total radiolemical innant of ietineration _
The totsi radiolatical impaat of incinerattee la that of the atmospheric discharge plus that of the chosen disposal option f or the ash.
In acet cases the former deatnates bsth individual and callective desse.
However. If the ash were disposed of by shallow land buttal then according to the present predtstions the mentava potential individual dose would onceed that due to atmospherie dispersion. The pegelble presenu of L raves of unburnt D C in the ash would not sigr.if tsantly shange this situation.
7.l0 Ef f ect, of tha-pronenee of 0.1 nem Natural tiranius in the Orachtes 7.10.1 Introduction in section 2 the peesth111ty was noted of there being e small asuunt of natural uranium prevent as tapurity in the graphite et the beginning of reactor 11te. and an apprentaste triventory resulting f rom the irradiestun or U.! pga uran.ua Was calculetsd f or Magnes and f or ACR graphite, in suis section the radiologie61 tapest af the disperal of that sattvity in the teradtated graphite Magnen is canaidered and compared with the tispect su, to activation products. Radionus11de data relevant to each model have-generally even taken f rom the esse seurass as those f or the acttvetton products.
The results are presented for the Mageon reactor.unty, but etallar concluetone apply to ACR.
1.10.2 Dave ocean bed dissonal Dosas have been t 'aulated using the same methodology dnd modelt o
_ _ __ - _ _ _ _ - - - - - - - - - - - ~^ - ~~~
_. , _- - ._m_ .~ -._ _. - ..-. _ . _ _ - - - _ _ , . . _ > . _ _ . _ _ . . . . .
~
4
- 4' _
'I stven in seetton 1.6. Meet of the environmental eencontration f actors f ort
. the various elements in fish, trustacea, selluste and'eediment have been taken f rom referense (34). The sedteent sensentration f arter f or plutontum v was sihan_froe reference (39).
t.eeuelns the fastest rolesse rate (10*l y*l), mantmas potential annual in tvidus1Ldosee-are absus-l 10*' Ov/y, le such less than these due to activation prodette. The solleettve dose seestteest, due mostly 'to 3*lA s. .
c to:ltkowisslvery eas11 sempered_te that for settvatten products, for all release rates. As noted in section.F.4 1, the escan models incorporate ,
steple assumptions-about the long ters behaviour ofl soes taportant elemente to!!awins their_ depeeltten with sediment en to the seabed, although thte le not important in the'reestle for.settvetten prodvete einse the dominant redlunucilde, l*C.. is _not s'tgnif teently'inet. to.bettee endimente. The >
- return of- radiesettvity f ree the soebed late the water solumn could-increase the collective does seamatoont~ f ree Shl A s and other redtonvalidee. but the-total co11estive deoes preststed assuming no lose of settvity to sedimente ,
ers _still substanstally lower than these due to setivation products. !
f.10.3 ~ hdenia diesenal 4
- Donos have been-satsulated veing the ease methodvlogy and modele se in e esction 1.7.. .)
-tnland' site For tholfull'rsnee of fresttonal reisese rates and aroundwater trennit i
times stven in Table f ell the individuel= end sc11estive doses are small cuepared le.'the serresponding deses ettelag f reslaelivetton products. For aussete, for the reference:eese the_ aestaus potensten annual individual does
'in about 5 10*l' ev (of -1.8 a 10 tv) and the sentributton to the- <
callestive dese-seasitaent~ to obeut il-s 10*l aan tv -(et 9.3 s-10*l ma n : sv ).
Cnastal~aita
' The same env6teneontal eensentratten f astere have been used as in I
..ction).lu.8. - As-fer the inland ette, reev1te for all sets of pa arster s ?
- alues are oush. Lower then these for the settvetten products _ even if no i tues ef aattvity torsediment te assised.
-O10.4 Shatlov innd burial ,
Doses' have been saleutated using the same methodology and sodate as n n.-
eastlun h a. TFeriforming'and drinking water pathways'the moutove potential annual individual dese!!s about 4 a 10*8 Sv -- The, as11ostiva ~ dose comett-
._ .sents for these pathways are eseh of the order of.0.1 aan 8v, and are 'due
- . mainly tw 33Np. leth andtvidual and se11estivc doses are very such lower thar _ ine-.estresponding deces f ree activetten produsse. Maximus potenstal
= 83 -
e n ' '
, , , ,- ., 4 , _ i~ - - - . . ~ . .
~~
m- ,- . _ . , _ , - _.J~
k*
annual ' individual doses atteing through the building pathway would be lower then those f rom activatten products in the first decade (Table 7.15). but at 1 ster times' the potential dose f rom II'Ce and !=l As would be of- the neder ,
of 1-agv, and would reastn a significant f raction 'of the annus1 dose limit untti the -IIice had decayed. However. there le some tesson to believe t hat 0.1 ppa to' an oversettaste of the true-level of natural urentum, and that the does froe IIICs hae been likawtet everestimated. If the samples of
. graphite that were analysed for the teach teste (section 6) had contained 0.1 ppe uranius, then the level of III Cs should have been about an order of magnitude above the limit of deteattent but none was observed.
7.10.$ _!ncinerggg, Assuming that the ease high decentaatnation f actore apply to the non-volatils elemente as for the activetten products (see section 5) the only volatile radionvalide atteing f ree the irradition of urantum in the erachtte to be teleased to etneophore would be 86tr. Howeve r. t he dos e e a r t e t n a f rnm such rolesses would be very esall compared with those due to the 1"C released .
7.10.6 Conclusione In seneral the disposal of irradiated graphite (or seh) would not give rise to significantly larger desee !! they contain irradiation producce f rom 0.1 ppe uraniums and there to reason to believe that. O.I ppe to an overestimate of-the natural uranius level.
7.11 Summeries of neeults for Mannen Raaater Mannon proarsene end- ACR
- 7.11.1 Introducties
- The results for the Magner deseantestoning programme and-the reference ACh were obtained by the saae methods as have been described f or the tef oreace Magnon roaster in eestions 7.3 - 7.10. Summarise of result s f o r
.all three cases are given in this eastion. In all cases the graphite is
- sesumed to be dispeeed of 10 y af ter shutdown.
-1; all three cases the esse radienus11 des, petticularly 40Co and l*C.
- are 'responettle for the sejer contributions to individual and collect tve
-doses f rom a given disposal option. ~ Thus the general explanatione already given for the form of the Magnes reactor results apply also to the result s f or the Magnon programme and for the AGR.
The doses estimated for a given disposal option are roughly propertional to the inventories of those two important radionuclideo (with some exceptione for individual doses, see below). In the esse of the Magnon programme the ins.atories of 14C and SO Co are both a f actor of about 22
higher than for the single reference reactor. The IC inventory of the ref erence ACR to about 2.2 times higher than for Kag tot, while that of "0 Co to about 36 times highet. (There to less graphite tu en ACR then in a Magnon reactor, but the ref erence tapurity level of C) to much great e r i e
- ACR graphite.)
Tables 7.21, 7.22 and 7.23 esmearles the results f ir the ref erence Magnon reactor, the Kagnon programme and the referente ACR. Ranaes o f results are given, corresponding to the ranges of assumptions sede.
7.11.2 -Individual doses As stated in section 7.3, the individual doses f rom the various optirna have different probabilities of actually being received, and this sust be taken into scrount when judging their significance. The ranges of estimated maximum potential individual doses given in Tables 1.21 - 7.23 correspon* s to the ranges of assumed values of parameters, f or example radionuclide retene.
rates and groundwater tranett times.
For shallow land burial the restriction periodo neceeeery tu ensure that annual doses due to encayntion while building on the site would not exceed 5 m$v would be a f ew darades, anu it is f easible to guarantee thin hv appropriate administrative ' controls. In the case of the Magnon decommiestaning progreene (Table 7.22), maximum potential individual doses predicted to arise in the longer term via farming pathways are lialted by the amount of graphite that could _be accommodated in a trench of the assumed alte, although if incineration were the chosen option all the ash could be accommodated and the predicted maximum potential individual dose is eccordingly higher.
Even if- all the graphite f rom the Magnon deconsiseloning prograsse could possibly be incinerated in one site in one year, as assumed in Tabis 7.22 the annual individual dose f rom one stack discharges would oniv reach about 101 of the dose limit.
7.11.3 Collective doses fallective does consitmente (integrated over all future time) f rom all the options considered are dominated by I"C, mainly through its global circulation. The dif f erent collective does conaiteents f rom the various optione reflect -the extents to which the I"C would have decayed before emerging into san's environment. .Only in the cases of long groundwater transit times for geologic disposal would the even longer-lived radionuclide 36 C1 become more taportant than I"C. The predicted collective dose commitment would be greatest in the case of shallow land burial, owing to the significant predicted contribution f rom groundwater-related I"C S
l pathways (though as sentioned in a atloa 7.8.1 this predicticn in b..med m m
peas tatstic essumptione). As already noted it hoe not been posathln tn estiasts the probabilities of the predicted collective doses beina received, or their distributions with respect to individuel dose rates; both voeld he important when comparing the rediological lupacts of the tempective dispo=al options.
References for Section 7
- 1. Webb. O A N and Kill. M D. 3 election of waste management eyetema ami ettstegies. IAEA International Conference on Wadioactive Weste Managseent. Seattle. USA, Kay 1983 (to be published).
- 2. It.ternational (bumieston on Radiological Protection. Pocommand a c i on s of the ICRP. Quf ord Pergamon Prese. ICRP Publication 26. Ann.
ICRP. 1,. No. 3 (1977).
i.
International Conaiselon on Radiological Ptotection. Co e t -te ne f a t analyste in the optintastion of radiation protectton. Caford. Pe r g amon Presa. ICRP Publication 37 Ann. ICRJ l_CL (2/3) (1983).
4 Bush R P. Whits. I F and Smith. C H. Carban-14 Weste Hanssement.
to be published. CLC .
5.
The Times Atise of the World, comprehencive edition. John Bartholocew i and *- Ltd. , Edinburgh (1973).
J 6.
Internettonsi Commission on Radiological Protection. Re comme nd a t i o n e of the ICRP. Oaford, Pergamon Press. ICR* Publication 2 3 (19 7 $).
7.
Nichslo. A L. Redicactive nuclide decay data f or reactor calculations.
Activation products and related isotopes. Ka rve11. UKAEA. AERE-8903 (1977).
S. Tohtaa. A and Davies, & $ J. UKFPDD2 t a reviaed fisaton product decav data fila in FKDP/1-IV format. CEGS toport RD/B/N4942 (1980).
9.
International Comeisson on Radiological Protection. Racommenda t ion e n t the ICRP. Oxford. Pergaton Press. ICRP Publicatien 30. Ann. ICRP. 7 (1-3) (1982).
10.
National Radiological Protection Board and Commissaria t 4 1
- Energt e Atomique. Methodology for evaluating the radiological consequenews ut radioactive affluents released in normal operatione. CEC document V/3865/79-EN.FR (1979).
- 11. 1AEA. 'The oceanographic basis of the I AEA revised defint t ton and recorsendations concerning high-level radioactive waste unaut table t ur dumping at sea. Vienna. IAEA 210 (1978).
12.
Convention on the Prevention of Marine Pollution by Dumping of Wasto s and Other Matter. London.1972.
_ _ _ _ _ _ _ _ _ _ _ . _ _ _ -- - - - - - - - - - --- ~
. _. . .. . _ . -_ =. .. - - _- - > - . - _ .
?
i i
Ih . Camp 11n. W C. Griswood, P D_ and White.- 1 P. The ef f ects of set tntdc ,
, separar' ton on the radiologith1 consequences of disposal of hish-level l radioactive waste on the ocean bed. Rarve11. National Radiological l Prottetton board, NRP3-R96 (1980). !
1% CESAMP.- - Group of taperte on the Scientific Aspects of Marine Pollution. An Oceanographic model f or the dispersion of westes
! disposed if-in the deep sea, IAEA heporte end.$tudies No.-19. Vi e nn e !
-(1903). 4
- 15. Kelly G N. Jones J A, Bryant, P M and Morley, P.- The piedicted radia*tiotemposure of the population of the European Comeunity resulting f rom discharges of krypton *45, trittua. carbotr-te and todtne-
- 29. :Lunesbourg. CBC doc. V/2676/73 (1973). ,
, ? l t. Ekdahl, C A. Escastow, R and Keeling, C D. Atacephuric carbon dien t de and radiocarbon in the natural carbon cycle 1N Proc. Srsposium on
- Carbon in the Bloop 5ere. CONP-720$10 (197 2).
11 IAEA. The radiological beste of the IAEA revised definition and i recoseendatione concerning high-level radioective waste unsultable f or duoping at sea.: !AEA 211- (19 74).
- 18. Thoepson. S E. Rurton. C A,- Quinn, D J and Ng, Y C. Conce at tat t n n f actore of cheetcal eineents in edible aquatic organisme. Univ.
Calif ornia. Lawrence Liversore Laboratory, OCRLe 5-566 Re v. 1 (19 7 2 L ,
- 19. -Bowen. H J M. Trace Elemente in biochemistry. tondon. Ac..drst e Pre s a (t966).
- 20. Consell' Internatf onal peur~ l'imploitation do la Mer.
Bull. Stat.- Peches Maritimes, 10,1975 Copanhagen (1976).
1
- 11. McKay. M' A C,- Miquel. P and White,1 P. . Manastaant modes f or todine-1129. IN Proc. opecialists' seeting on todine-129 sahamesent. CEC. EUR-7953 (1982).
. 1 22.; H111~, M D and Crismoed, P D. Preliminary assessment of the - radio-logicaliprotection aspects of disposal of highFlavel weste in geologic fornations. Harwell _- National- Radiological Protection Board, NRPR-Rb9 (1974);
- 2 3. Nye .P H, and Tinker, P 4. Solute sevement in the soil root system.
Studies in Ecology volume'6 Blackwell Scientific Publications ;
-(1977). - '
26.1 Little. Arthur D 'Inc. Technir.si support f or standards of high-level radioactive waste management. Volume C. sigration pa thways. EPA-
.520/4-79-007C (1979).
- 25. Meyer R E. IN Proc. of task 6'wasta isolation saf ety assessment program
- second contractor meeting. PNL-SA-7352 vol-2 (1978).
'26. Hill. M D. Hobbe. 5 F. and White ! P. An sesesseentLof the radin-
-logical consequences of disposal of intermediate level wastes in argillaceous rock. f ormations. Chiltor. National: Radiological Protection Sosed. 'NRPS-R126 (1981).
- 87 -
e
- 27. Erdal. 8 R. 1.aboratory studies of radionucinde distributtone haiween selected groundwaters and geologic media. Los Al amo n se t en t t i t e 1.aboratory, annual report LA 8088-PR (1979).
- 28. Francts. C W. IN Proc. of ta sk 6 wast e teolation enf ety a==v= =mant program second contractor seeting. PNL-5 A-7 352 Vol I (19 70.
- 29. Hill. M D and 14voon. O. An assessment of the redtonnatcal cnneequences of diposal of high level weste in coasts 1 saintoalt formations. Harve11, National Radiological Protection Anar d. WPH-win" (1980).
- 30. Pinner, A V and 11111 M D, Radiological protection aspect . of =halio.
land burial of PWR operating wastes. Chilton Na t innel Rati t olog t ral Protection Board. NRP3-R138 (199 3).
- 31. Finner, A V, Hemeing, C R and Mill, M U. An assessment of the radiological Protection aspects of shallow land burial of radianct tve wastes. Chilton, NRFB. Bapcet for the CEC (to be published).
- 32. Clarke, R H. A model for short and medium range disperston of radio-nuclideo rolesehd to the steosphete. Ra rwell, National Rad tnloalcal Protection Board, NRPB-R91 (1979).
- 33. Kelly. C N. Jones, J A and Broomfield. M. The radiation espovue, n f the l?K population f rom att* borne af fluente dietharged f rom civil nuclase installations in the UE in 1974 Chilt on. Nat iona l Rad t n-Logical Protection BaarJ. NRPS-Ril8 (1982).
34 Ng Y C, But ton, A. Theapeon. 5 E. Tandy. R K. Ke v t ne r . H K e tt l'y i t i M W. Predictions of the marinnan desage to man f rom the f =!!not at nucleat devices. IN Handbook for Eattaating the Montmum Internal ih r from Radionuclides Released to the 31oschere. Univ. Cali f orn t a , a t a l.-
501F1. Part IV (1968).
- 35. Ng, Y C, Colshar, 0 8. Quinn, D J and Thompson. 5 E. Tr a n = f e -
coef ficients for the prediction of the does to man via the f orage cow - alik pathway f rom radionuclides released to the biosphere. totv.
Calif ornia, UCEL-51939 i 977).
- 36. Final environmental statement, Waste Management Operations. lunf ocit Reservation, Volums 1. Washieston DC, EKDA-1135 (1975).
37 Simmonds, J 1 and Crick, M J. Transfer parameters for use in terrestrial foodchain acdels. Chilton, National Radiological Protection teard, NRPE-te3 (1982).
- 38. Camplin. W C, Clark, M J and Lelow C t. The radiation exposure of the UK population f roe liquid ef fluents dischargad t'rca civit nuc tuar installations in the UE in 1978. Chilton, Natit.nal Radiological Protection Board. NEP8-R119 (1982).
- 39. Pentreath. R J. The biological availability to marine orgentama ot transuranius, and other long-lived nuclides. 1N Pr oc. o f S ymp. on impacts of radionuclide releases int o the marine environment . Vtunna, Oct. 1980. Vienna. IAEA. (1981).
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - _ - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - ^ - - - -
_ . __. . . _ _ . _ . .. .. _ _ . . ~ , _ . . .. __ _.___ . _ _ . . _ .. _ _ . . ~ .-. .__-.. . - . . . .
table 7.1- Menenoment options for which the redtoloalcel tapact is~esiculated-l.- Dispo. 1 on the deep eteen bed. -
- 2. Geologic dispa.a1 at en inland . site.-
- 1. Coologic disposal at a coastal ette.
6 Disposal r.t a shallow land burial ette.
- 5. -Inciceration and disposal of ash and filters tot a) the deep ocatn bed, b) e deep taland repeettory,
.c) .a deep coastal tcpvattery.
'd) a shallow. land buttal ette, 4 6
E 3
i
\
Af s
f i
I' l
I
\.
e
, , - - + . my . . _
. _ - . . _ _ . . , . . , , , . . . . . . . - r _ , 4~-
- . - -~ . . . . . _ __ . _ . _ _ . . _ . .
Table 7.2 Environmental pathways f or each disposal rt.ote ,f or which the radiological impact is considered in thie studv
- 1. DISPOSAL ON THE DEEP OCEAN BED Internal irradiation due to ingestion of activity inn fish, crustacea and molluscs. 1 Internal irradiation due to inhalation of activity resuspended f ron beach sediment.
External samma irradi.stion froe contaminated beach sediment.
- 2. CE0 LOGIC DISPOSAL AT AN INLAND SITE Internal irradiation due te ingestion of activity int drinkins water and freshwater fish.
- 3. CE0 LOGIC DISPOSAL AT A COASTAL SITE Internal irradiation due to ingestion of activity inn fish, crustacea and sollusts.
Internal irradiation due to inhalattoi of activity resuspended f rom beach sediment.
External samma radiation from contaminated beach sediment.
- 4. DISPOSAL AT A SRALLO'J LAND 3URIAL SITE
-Internal irradiation due to ingestion of activity in:
root vegetables, green vegetables, grain, sheep eent a nd
. liver, ecw mest and liver, silkt drinking water and freshwater fish.
. Inhalation and esternal irradt.ation during building operations on the site..
- 5. DISCHARCE TO ATM0,5PMIRE Internal irradiation due to inhalation.of activity in the cloud, and inhalation of resuspended activity.
Internal irradiation due to ingestion of activity int root vegetables, greenivege?&bles, grain, sheep meat and liver, cow east and liver, silk and milk products.
External gaanna irratistion f rom the cloud and deposited activity.
- 9
Tabla 7.3 tadteene.lide de*.a for astivettee steducts home per amia 1stok.4, 99/94 Insas genums energy Radio- ,,g gggg ' I(7.8) nuclide tapesties laanlattee pne dictategrettee, leaf 35 1.23 101 1.71711 1 7 10*ll 0 l' be 1.60 108 1.1 tr' 9.3 10*' 0 l'c S.69 103 S.6 10*I' 4.3 1713 0 3'*1 3.04 108 7.8 10*18 3 3 10*' 2.3 10-6
'!ca 103108 3.3 1718 3.3 1 r 88 0 5'tta 8.53 10*1 7.3 10*18 1 7 10* 8.3 10*3 55te 2.70 1.6 1r38 3.3 10*1' 1 6 10*3 5'N1 7.50 10' S .4 1r!! 3 6 to-te 0 "Co S.27 f.9 10*' 4.1 Ig-a 3,3 83N1 1.00 102 g,3 gg.it 8.4 10*l8 0
'Sta 6.69 trl 3.9 10*' 3.0 10*' S.8 10*!
'tto 3.30 103 3.3 10*' -7.6 10*' 1 1 10*3
'*n 1.64 101 1.4 10*l' 7.7 10*' 1.6 10*F
n 2.03 10" 1410*' 9.0 10** 1.6 "Te 2.13 108 3.4 10*18 2.0 10** 5.7 10*7 3 "' As - 1.27 101 1.0 10*' 3.3 10*' 1.7 183*Cd 1.36 101 4.0 Ig=6 1.0 10*' 1.6 10*S 121*$a 3 30 101 3 7 10*l' 2.3 10*' 2.1 10-2 133&e 1.05 101 8.6 10*Il 1.9 10*' .1.6 10*1 -
Illte 1.33 103 1 6 10"'- S.9 10*' !. 15 15*te : 8.60 2.4 1r' 7.0 10*' 1.21 l'Itu 6.63 3.7 10 ale g,g tre 6.0 173 l
l I
I 6
= 91 I
L 4
Table 7.4 Collective dose commitment arising from the globst dispersion of 3*C Incomplete collective dose commitment, men Se, following a release of 1 789 of 3*C to:
Time after the troposphere of the surface oceans the deep oceans of the northern hemisphere in*:
release the northern . of the northern commences, y heatsphere' in 1 y heelephere in 1 y 10 y 102 y g g3 y gg4 y 1 5.1 10-1 1.6 10-2 2.1 10-' 2.1 10-7 2.1 10-8 2.1 to-'
2 1.3 4.1 10-2 2.3 10-5 2.3 10-' 2.3 10-7 2,3 10-8 5 2.8 5.3 10-1 6.2 10-* 6.2 10-5 6.2 10-' 6.2 10-7 1 IDI 4.4 1.3 6.8 10-3 6.8 10-* 6.8 10-5 6.8 10-'
, 2 101 6.1 2.5 5.5 10-2 6.2 10-3 6.2 10-* 6.2 10-5
. 5 101 8.0 4 t3 3.7 10-1 8.2 10-2 8.2 10-3 8.2 to-*
" 1 102 9.0 6.8 1.0 10 ' 4.5 10-1 4.5 10-2 4.5 10-3 8
2 102 1.0 101 6.1 2.2 1.7 2.1 10-1 2.1 10-2 5 102 1.4 103 9.8 5.9 5,3 1.5 1.5 10-1 1 103 2.0 10I ! 1.6 101 1.J 103 1.1 101 5.5 5.5 10-3 2 10 3 3.1 101 2.6 10I 2.7 10I 2.2 10I 1.6 10I 2.2 5 103 5.6 103 5.1 101 4,3 got 4.7 101 4.1 lol l1.210 1 1 10' 8.2 101 7.7 101 - 7.3 103 7.3 10I 6.7 101 2.6 10I 2 10" 1.0 10 2 9,9 10I 9.5 10 I 9.5 10I 8.9 103 5.0 10I 5 10 4 1.1 102 1,3 gg2 1.0 102 1.0 102 9.5 101- 6.0 101 1 105 1.1 102 g,1 302 g,o go2 1.0 102 9.8 89I 6.0 101
- These reicese times correspond to the times ever thich release occur e for the raese of fixed fractional release ratem assumed f or the t eacning of act ivity f rom the waste form disposed of on the deep tween bed. Account has been taken of radioactive decay while activity is r2tsined in the vaste.
i
- ...-.m. . . . _ .
}
i
.p, Table 7.5
+
CollectIwe dose cemettsent arieien frem the alohol diepereIon ef IH 35 ter incomplete collecttwo dose ceaeitsent, see Se.' felleetes a ret ees af 1 Tag of the deep eceses of the serthern hemisphere le*t flee efter .
g telesee them circelettes estere of the 10 y 107 y 103 y I. 10" y ceseences, y sorthere hentophore se 1 y 4
4.8 10-II 4.8 10-12 4,3 ;g-13 e
6.7 10-' 4.8 10-I8
'" 1 1.8 10-5 3.5 10- 3.5 10-I' 3.5 10-II 3.5 10-12 8 2 4.3 10-* 4.3 104 4.3 10-I' 4.3 s0-II 5 4.0 10-5 2.3 104 2.3 10-38 3 3.6 gg-5 2.3 30'I 2.3 10-8 10 S.2 10-7 9.5 10-8 S.S 10-' 9.5 10-18 70 6.5 10-5 3.1 10-8 3.1 10-9 6.8 10-5 1.7 10-' 3.1 10-7 30 1.8 10-6 4.2 10-7 4.3 10-e 4.3 10-'
100 6.8 10-5 g for the range of fined fractiesel reIcese
- These release times correepend to the times over dich release occure. Accoent hoc bee =,
rates seeased for the leechtsq et activity f ree the weste fore disposed of on the deep ocess bed.
taken of sodioactive decoy Alle activity is retelmed in the onete.
i i
i
l l
l table 7.6 Concentration f actore for seafoods and serine sediments Concentration f actors.(10.13.17,18,19),3f g, tienent Fish Crustacea Molluace Marine sediments 0
Tritium 1 1 1 103 103 103 106 Bery111un 5 103 $ 103 102 Carbon 5 103 1 10I Chlorine 1 1 Calcium 1 to 1 5 102 gow go w Manganese 5 102 tok tron 103 103 103 10' g02 go*
Nickel $ 103 102 102 103 10" Cobalt i Hlobium 5 102 tot 103 10' 2 103 $ 103 10I 10" Z1ne 10 103 103 10" Technetius Stiver 103 5 103 5 10" 10' Cadmium 103 3 103 106 10' go w Tin 103 3 102 102 Barium 10 2 30 3 103 Europlus 102 103 103 10" Molybdenum 20 20 50 250
- The concentratten f actors given here are the ratio of the quantity of element per tonne of material (fish, crustacea, etc) and the quantity of elemoet per m 3 of filtered water. These are based on the dry weight of sedimente and the wet weir.ht of the edible parts of other satorials.
- 94 =
Table 7.7 deep ocean Maximum potentiste , anual individual doses arising from the reactor, decayed for to y bed disposal of the graphite f rom the reference Magnos
- Maximum potential annual indivdual dose for each pathway, Se Beach sediment pot hways Ingestion of External Fractional Inhalation release rete of suspended games Crustaces Ms11uses y ~3 Fish sediment i
e w
5.0 10-s 3.9 10-1" 3.0 10-7 8.8 10-7 4.9 10-8 3
10-3 3,9 gg-15 3.0 10-s 3,3 30-s 4,9 30-9 5.0 10-9 10-2 3.9 10-8' 3.0 10-'
1.8 10-' 4.9 10-to 5.0 30-18 10-3 3.C 10-38 4.9 10 38 5.0 _ *3 3.9 10-37 10-* t.8 10-18 l
adecction may arise at any time during the period that
- Doses arising from plume The maximaa 'otential doses presented activity is being leeched f rom t he wast e. year after disposal.
here would arise if plume advection occurs during the flast
)
i 1
I Table 7.8 f c
- Total ,. atessi slun'alebal) es11estive dose tossiteent ertelna from all pathwere fellevtaa the' does tesen bed diesesel of the arethite f ree the gence Maanes reacter. decered f or 10 y l
Collective dose cousitseet, een Sv. f or the followins Time after fractional release ratest
' P"* ' I
, 10*l y*l 10*3 y*l 10*3 y*l 10" y*!
1 1.7 10** 1.7 10*9 1.7 10*' 1.7 10*'
2 2.2 10*3 2.2 10** 2.2 10-5 2.2 10-6
$ l.6 10-3 S.6 10*3 S.6 10** 5.6 i C*'
$.9 10**
1 101 S.9 10-3 S.9 10*3 S.9 10-3 ;
2 104 4.8 3.3 10-1 5.3 10*3 5.3 10*'
S 101 3 2-103 7.0 7.1 10~1 1.0 10-3 1 102 8.6 101 3.9 108 3.9 3.9 10-1 j 2 102 g,9 go2 1.$103 1.8 108 1.8 3 103 S.0 103 4.6 102 1.2 102 1.2 101 9.6 103 4.7 102 6,7 - g ot 1 103 1.0 103 2 103 1.9 10* 1.9 103 1.6 108 1.9 103
$ 103 4.1 103 4.0 103 3.5103 -1.0 101 1 10" 6.3 103 6.3 103 $. 8 103 2.3 103 2 10" 8.2 103 0.1 103 7.6 103 4.3 103
$ 10" 8.9 103 8.8 103 8.4 103 $. 2 103 1'105 .8.9 103 4.8 103' 8.4103 !. 1C' ,
~l
- i h
i m
t i
- 96-t e
.,c,-., n,~. - - - - - -., .,4 e -. , ,,,,-. .+..,, - - ,,a,,,..,,,.- . - , . , , , , -,rn, ,<-w,..-.,n,. , ,-,,----w,- , - - - - -
i '
i i
t-l' Table't.9 Assamed hydtcgeologic parameters j
teedisse Groundwater .
Diepere1oe . Flces path lengt h, a i_
velocity, a y'A coefficleate, e2 71 reference ve rtat tees reference wo r t et t e=ne reference vertattoes -
Cl</ 3'IS~' 8.30-*,1 10-3 3 10-3 -
50 150. 17 Shale I O.3. 3 10 3. 30 1000 3000, 300 ,
I {
4 Table 7.10 3 i
Betardettee coef ficiente med le aseleEle d!sseest calculertene l 9
c5 e
Element Betardettee coefficient .J t
i' De 3500 se 260 +
f C S ,
C1 8 Ce 700 see 20 51 70 Tc I Sa 875 Mb 7000 ;
Ag I ;
i i
)
Notes f or Table 7.10 :
) ll ) Caly those elements have been considered for wh ich a relat ive ly long-lived radiot sot ope l j
ts present in the gre tte.
l{2},Thesame values haw gen assawd f or tanth clar and shale.
i s
I e
f h
- m., y , -. 6 ..r.5~.- -
-+.U- v .-. , . ...,omw,e - + = . - , . ,. ,-m
--+. ,.r--- ..'w,n... ,-.-e .m..,.. .---.e-- _ .-.
Table 1.11 reactor graphite detsyed f or 10 y Doses atteing f ollewlag f.he inland deep geologic Jterosal of the ref rence Magnou Regional collective Begiosel and global Maatsum poteat tel T2me of maalous cwilective dose Groundwater
- Fractional d os e, y dose comaltment, transit time, y release rate, soeuel individwal men Se comettment, men So y-3 does. Su 9.3 10-8 9.3 10-3 5 10-1 1.2 10-' 8.5 10' 9.3 gg-s I.i 10 8.5 10' 9.3 10-1 S p-2 1.2 10-' 9.3 10-1 9.3 10-1 1.2 10-' S.5 10'6 9.3 10-1 (reference) 10-3 9.0 10 9.3 IG-1
[ 10-* 1.2 10-5
( 1.4 10-8 l 5.0 10 S 5.4 10-3 f
10-1 3.7 10-81 1.4 10-1 1.4 10-1 l
1.7 10' 3,y 30-81 5.0 10 5 l 10-2 1.4 10-1 1.4 10-1 3.a 10-18 5.0 10 5 1.4 10-3 (a 10) 10-3 5.0 10 5 1.4 10'8 l 10-* 3.7 10-11 E 1.0 10 8 3.0 10 6 1.4 s
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- The groundwater trenett time further af fected by serption anJ grecadweter velocity. Tr ans por t of radicauclides le l dispersion.
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Table 7.13 Boet coecentratio_n factors, and f ractions of deity 1stehe treesferred to setest prodoch menened for shelles two becist Reet concentratlee facter Fraction teameterred, d kg-I or d 1-1 bcItde Neet Liver Boot weg Crain Crees veg Feature Milk ,
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l Table 7.16 i Development of collective does c;--8toont erf stria f rom t he atsoomherte discherav durtria the tricaneration of -the 10 v-derayed ref eretice Memnos reactor' arschite
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Tine after release Incomplete collective dose -
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table 7.17
,Tjt_el (rettonal + alotel) collective does ecweittent arteter from all ethways foliovtyr the _ deep ocean hed disposal ol __the ash arts'en from tn; incineration of the graphite from the reference Mornon tenetor, Jecayid, for_10 y Collective dose c6asitzent. men Sv. t ot the f ollowing Time ef ter f ractional telease ratees disposal.y 10-1 y*l 10-2 y-l 30 3 y-t t o-a y-t
! 6.$ 10-6 6.$ 10-7 6.$ 10-8 6.$ 10**
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- Table 7.21 ,
Reference Magnon reactor: summary of estfested potential individual doses and collective doses
- Ranges correspond .to the ranges of answeptione made Baege of eetnested man. potential Emege of estiested collective Graphite esangement opties dose commitment. osa Sv ;
' tedividal doee, Sv/y ,
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Oceam-bed disposal 3.0 10-15 to 3.0 10-7 5.2 103 to 8.9 10 3 Inland geologic disposal 3.7 10-33 to 5.7 10-8 8.4 10-3 to 10 3 3.1 Coestal geologic disposat 3.7 10-33 4.3 10-7 toto4.710-'i1.2]
3.7 10-2 1.8 10 2 toto8.710{I}
2.3 10' Shallow land nortal i g <
Incineration g,, 303 steoepherte discharge 3.0 10-5 Flus disposal of ash to:
(negligible la compartess with (megitgible la comparisco with l ocean bed the atsoeg4 eric-diapersion [
inland geologic the atmospheric-dispersion F cometal geologic dose from tocineration) dose fromtotactaeratt 4.3 10-7 1.7 10-23 ).2]
shallow land i l
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[1] These estteates are uncertain and probably pesele*stic: see sectior F.8.2. >
[2] Restriction of building. en the site would be required for at least 30 y atter disposal (40 y af ter shutdown) to prevent anneal Joees in excess of 5 a5v to individual butIdt"E workers. :
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Severe! aspette of the sanagement of gtsphite f ree reactor decommis*
sioning have been vnamined in earlier parts of this report, and this sectice discusses their influentes upon the acceptability of the sanagement options.
The Sain criteria of acceptability are logistical feasibility. cost end g radlological 14 pact. Toasibility and cost are closely linked and att discussed qualitatively in section 5.1. Radiological tepact to discussed in section 8.2 and uncertainties effecting the over 11 results in section 8.3.
8.1 Fearibility and Costs The sea-duoping option to feasible f or graphite under the terms of the tendon Convention ' end the 1ALA Definition .
Wf th present f acilities the capacity f or handling drummed graphite would be limited, but any f uture f acilities ate expected to be able to handle either the present 1803 containers (singly or in multiples) or the contents of a shielded overpack.
A feasible system f or see-dumping larger containers is described in section 3.4 The deep geologic and shallow burial optione depend on the availability of auttable disposal attee within reasonsbit access of the reactor sitia.
Inct. oration poses logistical problems. An incinerator of the reasonable else envisaged in section 1 wou14 be able to deal with one graphite cote (moderator plus reflectors) in about a year. The greatest number of incineratore required would probably be to, one at each UK Magnon/ACR attet and conalderations of the else and dura 61on of the wormioad suggest that the least pose 1ble number would be 2. The strategy of latte numbers would require only the off* site traceport of the reduced quantity of ash, but every alte would require an incinerator and at single-station ettee the cost-ef f ectiveness would be very low. If f ewer incinerators were used it would be necessary to transport packaged graphite f ree other ettee. But the packaging requirements for transport are sesentia!!y the sees as those
! for transport followed by direct disposal, so the wane packaging cost and lomistic effort would be better spent on direct dispo.a1 without incurrina l the additional costs of socineration. The only attuation in which Locineretton eight become note attractive would be if other disposal facilities were so restricted that the vtlues reduction became a slanificant
!' advantage.
As noted in section 3. the costs of the packaging options f or direct disposal are not directly cooperable, becauce the costs of the packages for the present sea-dumptna practices are based on emperience while those f or 1
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the two larger types of packages are conceptual, and eleo because the coste f or the latter have been estimated on dif ferent bases by two dif f erent
>rgantentions. Esactly the ease kinds of objections arise to the detailed t
comptrison of the costs for ese and land dispoest methods. Howeve r , ti appears that none of the methods would have prohibitively high costs.
9.2 Radiological tepect for each graphite disposal option, the two aspects of radiological txpact to be considered are doses to individual members of critical groupe.
and collective doses to the entire esposed population.
1.2.1 Individus1 doses >
Individual doses calculated in section 7 are potential doses f ollowing particular sequences of evente, and the relative lake 11 hoods of these doses deleg received should be taken into account when coupering thee with the appropriate limite. In f act the potential individual doses predicted f or .
all the options but one are veil below the 5 mSv annual dose limit so it le not necessary to consider their probabilities in more detail.
In the esse of shallow land burial. the earliest individual esposure toute would be f rom escsvetion if the ette were built on af ter closure, but this can be avoided by restricting the use of the land during the firer I w ,
decades af ter shutdown, or by delaying decoundesioning untti the short-lised radionuc11 des responsible (particuj > rly 60C o) had decayed. In the lonscr ters 36C1 and 16C transported f rot?the engineered trench by moving groundwater are predicted to enter jood chaine and could give rise to appreciable individual doses.
However. the shallow land butidt assessment to uncertain in sose respects (section 7.4.2) and to prettably peestmistic. In particular, dissolved 1*C may in f act be intotopically diluted, or ret 61ned in the Similar, repository or the geosphere, to much greater antants than assuesd.
though probably lesser, eff ects may mean that the predicted doses f ros.36 C1 are slao pesetnistic. A eire dets11ed and ette-specific assessment could find the individual riche f ree shallow burial of graphite acceptable.
l Theref ore none of the optione considered in thte study need be ruled out by eacessive individual dosses all scrit further investigetton.
8.2.2 Collective doigt 1
All of the options assessed in section 7 are predicted to release some i of the l*C from the graphite into global circulation, and for many optione
- this component would then dostnate the collective does cos eatment (taken to f
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Inf inite tise), es shown in Figure 8.1. The only disposet ortions !
conaldered which could appreciably reduce the collective dose conniteert are inland a.d some esses ot coastel geologie disposal in which the stedicted geosphete tranett times for I"C are much steater than its radicactive half-11te. 3'C1 would then become the major contributor to collective doses. i When using the predicted collective doses se an aid te cueyering the graphite management options and choosing the optinue with the aid of such techniques se cost berellt analysia, more importates mer be attached to the earlier componente of the collective dose coseitsent and also to co!!ects ve '
doses received at higher individual dose rates (although it has ne* been pissible to make est1Estee in the latter respect).
There are very merked dif f erences oetween the pred1?ted time-distributtone of collective # oses f ree the various graphite disposal aptions. These dif f erences artes setecially in the earlier ters and are i enaracterited by considerable uncertainties, as shown by the shaded bando of results in Figure 8.l. ($1a11er uncertainties apply also to the results f or incineration, although only a single set of assumptiont has been e/aluated for that option.) These and other uncertainties are discussed below.
8.3 ync,ertainties Although some of the uncertainties associated with this study concet9 the f t)sibility and cost of the graphite management optione considered, they generally have sore marked ef fects on the predictions of radiological twpset. Many of the uncertainstes arise f rom the seneric nature of the study, and could readily be resolved by investigations related to a specific content. og to a specific reactor or disposal ette. This le particularly true of f actors which can be determined by contemporary measurements or require only short-term entropolatione.
Some other uncertainties need not be resolved f or the purpose of semins compartoons between options, because they are common to oli the optione considered. For emaaple. I*C in the environment can be cansidered t a be in l 31nbal circulation at times beyond about 10 3 y af ter tre rolesse and although the globst-circulation modelling of l*C contains uncertainties they need not enter into compartoona between options.
The remaining uncertainties are more dif ficult to ree91 wet they may f or enasple involve site-specific f actors which are dif ficult to determine, or require entrapolation of contemporary observations into the f or f uture.
i ?n this study some of the uncertaintine have bein dealt with by
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analyste of the sonett141ty of the resultr to variations in parameser P
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values. Sensitivity analysis can also indicate where f urther tr'vestigettone are required, and radionuclide inventuries and teach rates require more specific discuselon in this respect.
Radionue11de inventories _
The radionuclide inventories calculated f or Mar.nos and f or ACR in section 2 are ?eneric. ILepresentative character 16tice were aneuned f or the ref erence reactors, and representative values f or the concentrations of stable tapurittee. Werefore wh116 the predicted activation inventory say be accurate when averaged over en entire reacto't systen, variatione must ce empccted between individual reactore cwing to 41sterentes in design and in operating historieel these dif0erences will also affect the levels of contesination by circuit crud end volatile f1selon products eleased f ron failed fuel cledding. The urentus-fieston inventory was talculated soeuaine en initial tapurity level of 0.1 ppe natural vranius in the graphite, and while the predicted effect on the radiological 1spect of disposal to generally small there le also reason to believe that the true uranium leols are lower than assumed.
On a enslier scale within the graphite of a given reactor there may be even wider variatione in local radionuclide inventories f rom all sources.
Theref ore it would be .tecessary to comple the graphite et e number of locations within the reswer when devising detailed plane for its decommissioning. How3ver, given the flexibility of the pacLaging methode described in this report, variations about the seen activity should not afiset the overall choice of disposal route for the graphite.
Leach rates In many of the cases er.anined, the radiological impact of graphite disposal arises through leaching of the redtocuclides by groundwater or reawater, and some of the lay 11 cations of the aussured leach rates were discussed in sectica 6.S. We radioingical aseesemente in section 7 show that the measured leach rates f or short-lived radlinuclides such so 6cCo are so low that almost all their inventerise can confidently be tapected to decay within the graphite under conditions of disposa18 but the half-lives of some radiologically toportant nuclides such as 1*C (5690 y) and 36 c1 0 00.000 y) are so long that lasch rates say have some effect un their eventual contributions to individual and collective doses.
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102' 10 10' 10 10 10 Time, y Figure 8.1 Development of collective dose equivalent for the disposci of the 10-year-cooled graphite crising from the decommissioning of a . reference Magnox reactor for the various disposal options
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- 1. The radionuclide inventories of Magnoa and ACR graphites 10 years af ter the end of reactor lif e have been calculated. The sain contribution to f rom W activat. ion of the carbon steelf and the stable impurities.
- 1. A lesser contribution to the total inventory any arise f rom fieston of natural utantum preeect as an is, purity in the graphite. A essautrient on e .
single esople, and other indirect evidence, suggest that the urantue level in unlikely to exceed 0.1 ppe. Evsn at this level. the contribution to the total radiological impact sf disposal would be reistively minor.
- 3. Surf ace contaAination of the graphite by circuit crud or by volatile firston products both depend on the design and operating history of the particular reactor. While these contributions to the tc,tel tedionuclide inventory snd the redtological tapact of dispopal cannot be settested in a generic study, they will probably be velatively small.
4 Three dif ferent packaging methods for graphite have seen described.
Each is logistica11y feasible and could be used f or either ses or land disposal, though the type at present used f or sendumping would probably be less suitebis for land disposal than the two larger packages.
- 5. Cost estimates are presented for each of the three packaging methods.
and any of the three could be implemented at reasonshle cost. 1he estisetes have been made on different bases and cannot be used to make detailed intercomparisons between the packaging methods. However. it appears that larger packages are mere costMffective. Further studies are required in this area.
- 6. Untreadiated graphite in large blocks retains its mechanical integrity at pressures up to 900 bar. This finding was confirmed during leaching teste on ses11 samples of irradiated graphits at 450 bar and 2.5'C. the conditions on the deep ocean bed at a depth of 4 km. Staulated seawater was l
used as the.leachant, and a parallel test was carried out under ambient l
laboratory conditions. Other teste under ambient conditions used staulated groonowater and domineralised water. In all cases I"C had the lowest leach
! rats-and 133Sa or II*Cs the highest. but the ef f ects of the dif f erent teaching conditions were found ts be small.
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l 7 Ocean-beJ disposal of graphite le predicted to lead to ar.ceptably low i potential individual doses.
$. Two forms of deep geologic diseosal of graphite have been considered:
at an inland ette in clay, and at a coastal site in shale. For a wide tante of assumed teach rates and hydrogeologic parameters, either type of generic site could lead to acceptably low potential individ.ial doses.
9 In the esse of shallow land buttal it would be necessary to teatrict bulliting on the ette f or some decades af ter disposal, untti the 40Co content has appreciably decayedi such restrictions are quite feasible. The long-tera radiologital impact of graphite disposal by shallow land burial requitos further investigation, mainly because of the present dif ficulties in modelling tne behaviour of I"C in soils and shallow groundwater.
Ultimately these investigationi must be on a site-specific beste.
- 10. A f aaelble altersative to direct disposal of packaged graphite to incineration, followed by disposal of a ess11er values of ash. But incineration presente logistical difficulties additional to those of direct disposal either seversi incinerators must be built at dif f erent sites or the graphite eust be transported to a smaller number of centralised incinerators 4 the corresponding coste could very widely. The radiological tapact of incineration would arise almost entirely f rom the atmosphetic discharge of I"C and is greater than that of many other options, especially in the early ters. Incineration does not seen a desirable alternative to direct disposal, unless the reduced volume of seh to judged to override all other consideratione.
!!. All the graphite manageoest mode a considered in this generic study .
merit f urther investigation. The choice of t he optimus management mode must be nada in 'some epocific content, and the optimum will depend on the ,
specific circunstances. When making suc'n a choict (poselbly with Ihe eid of cost-benefit analysis) further information would be needed on the time-varying prob 4b111ttee that predicted individual and collective desse would be received, and on the individus1 dose rates et which collective doasa alght accrue.
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