APPENDIX 1 ELECTRIC POWER PRODUCTION IN GAS-GRAPHITE REACTORS

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP88R01225R000100310003-2
• Release Decision: RIPPUB
• Original Classification: S
• Document Page Count: 267
• Document Creation Date: December 22, 2016
• Document Release Date: September 1, 2010
• Sequence Number: 3
• Content Type: REPORT

Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
Electric Power Production in
Gas-Graphite Reactors
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01: CIA-RDP88RO1225R000100310003-2
PAGE 001
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Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88RO1225R000100310003-2
FILE:
BLUENOSE FT25FOOl A
BUFFING COMPUTER CENTER
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UNCLASSIFIED
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
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Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
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Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
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UNCLi , FIED
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88RO1225R000100310003-2
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0. 153000. 262000. 297000. 62000. 351000. 284804.
0. 150000. 9000. 268000. 2000. 342000. 232342.
0. 46000. 0. 1000. 1000. 337000. 124375.
0. 63000. 0. 1000. 1000. 163000. 271155.
0. 152000. 0. 1000. 60000. 0. 278348.
0. 152000. 94000. 1000. 158000. 0. 275937.
0. 151000. 241000. 39000. 11000. 0. 269913.
0. 158000. 135000. 143000. 220000. 0. 262689.
PAGE 006
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RUFFING COMPUTER CENTER
PAGE 007
8311
0.
0.
0.
0.
152000. 144707. 150342. 257517.
115110.
269623.
8312
0.
0.
4191.
0.
156813. 130589. 79932. 229530.
319680.
280844.
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 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
Gas-Graphite Fuel Fabrication Data
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1968
Production of fuel destined for the EDF reactors was limited to about 130 tonnes
of uranium, and 80 tonnes for Chinon 1.
1969
This was a year of transition. 330 tonnes were produced for EDF.
In 1970 and 1971, CEA will have to supply Saint-Laurent 2, Bugey and Vandellos;
production will be in the order of 800 tonnes a year.
1970
620 tonnes produced for graphite-core reactors, like Chinon 3.
Fuels are currently being produced for Saint-Laurent 2, Vandellos and Bugey.
1971
Initial delivery of fuel (natural uranium) was made to Vandellos; no amount given.
Since its inception in 1962, CEA has supplied EDF with more than 3,000 tonnes of
fuel for the graphite-gas system.
1972
Total supplies to EDF and Vandellos amounted to only 260 tonnes, as opposed to
nearly 600 tonnes in 1971. At the same time, there was a 10% decrease in pro-
duction for G2 and G3.
1973
Production for EDF and Vandellos--585 tonnes-- was more than double that of
1972 (260 t). On the other hand, production for G2 and G3 decreased by half in
relation to 1972.
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Production for EDF and Vandellos was about 550 tonnes, 10% lower than for 1973.
1975
Production was 680 tonnes of uranium, approaching the maximum produced in France
(717 tonnes in 1970). This was 10% more than in 1974, despite the decline in
production for G2 and G3. To this is added the production, always important, of
fuels needed for G2 and G3.
1976
Production for EDF and Vandellos was 650 tonnes, to which are added the fuels for
G2 and G3.
1977
Production for EDF and Vandellos was 570 tonnes, to which are added fuels for
G2 and G3.
376 tonnes for EDF and Vandellos.
491 tonnes-for EDF and Vandellos.
1980
471 tonnes for EDF and Vandellos, to which the supplies destined for G3 must be
added.
A manufacturing plant for fuels for water reactors is under construction; its
planned initial capacity is 500 tonnes/year, to be increased to 1,250 tonnes.
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APPENDIX 3
UP2 Reprocessing Data
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REPROCESSING OF GRAPHITE/GAS FUELS
From 1966 through 1982, the UP2 facility reprocessed 4112
tonnes of fuel from EDF graphite-gas reactors and the
Franco-Spanish Vandellos reactor. Until 1974 the facility
operated at less than design capacity because capacity exceeded
the amounts of irradiated fuel available for reprocessing. The
quantities reprocessed grew with fuel deliveries through 1976
(Table 1). Since then oxide fuel reprocessing activities have
limited graphite-gas fuel reprocessing to about 6 months per
year. Table 2 shows the types of fuel reprocessed since 1966.
25X1
25X1
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NUGG Fuel Reprocessing at UP2
Year
Delivered*
Reprocessed*
Stored*
1966
53
52
1
1967
150
98
53
1968
166
187
32
1969
235
228
39
1970
197
136
100
1971
101
164
36
1972
291
250
78
1973
430
213
295
1974
555
635
215
1975
532
441
306
1976
326
218
414
1977
333
351
396
1978
388
372
412
1979
238
264
386
1980
160
253
293
1981
185
251.
227
* In tonnes of initial uranium
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Types of Metallic Fuel Reprocessed
Type/Weight Reactor Quantity Reprocessed (te)
U/Mo U/Mo MoSnAl SiCrAl
0.5% Mo 1% Mo
Tubular EDF-1
10 kg EDF-2
EDF-3 191 1816 126
SL-1
Tubular EDF-3
Graphite-Core SL-1
10 kg SL-2
Vandellos
Annular Burley
24 kg
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TRANSPORT OF IRRADIATED FUEL TO THE LA HAGUE REPROCESSING PLANT
B. Lenail
COGEMA, Velizy tillecoublay Cedex, France
H. W. Curtis
Nuclear Transport Limited, Risley, Warrington, United Kingdom
The story of the transport of irradiated fuel to the La Hague
reprocessing plant of COGEMA is a story of oassive experience,
comparable throughout the world only with that of BNFL's Sellafield
plant.
Natural uranium fuel transports
The story began as long ago as 1966 with the transport of
natural uranium fuel from French Graphite Gas Reactors. These
transports have continued at verying annual rates to the present
time in relation to the distribution of fuel between La Hague and
the natural uranium reprocessing plant at Marcoule. It is expected
that all natural ur.,nium transports will ultimately be directed to
Marcoule as oxide fuel reprocessing at La Hague increases. The
following Table I shows the annual tonnages transported to La Hague.
Metric tons of natural uranium fuel
from graphite-gas reactors
1Q66
1967
196
1969
1970
1971
1972
1973
1974
12Z1,
53
150
166
235
197
101
291
430
555
532-*
1976
1977
1978
1979
1980
1951
1962
326
333
3S8
238
160
155
111
The flask used is of -'ibic shape of approximately 2.3m wide and
a loaded weight of some 54 ?ietric tons. The shielding consists of
lead totally unclosed in stainless steel cladding.
Light water reactor fuel transports
Light water reactor transports to La Hague began in 1973 but
remained at a level of about 100 a+ttric tons per year until 1981
when the annual amount transported tripled In comparison with 1980.
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I M1982 there was a further increase of 60% over 1981. The tonnages
transported are shown in Table :1.
1973
1974
1975
i976
1977
1976
1979
1960
1981
1982
10
26
97
60
101
42
112
135
405
647
Total - 1685 metric tons
gyropean countries outside France, but in 1979 fuel began tomarrive
frog Japan and In 1981, from the large French domestic pro
reame of
g
b: pYBs.
Light water reactor fuel Is transported to ',a Hague in a large
rt variety of flasks - dry flasks,?wet flasks, small 3E tonne flaska
d f
designe
or road transports up co 110 to monsters transported by
i A standard
~~ --?_ _ _. _ _ _
~
rang
ra
of
^ degree o automatic nanaif.ng in the UP3A plant and the present
yybrid family will act be extended. The ;8 tonne road flasks have
p:?s capacity of approximately one tome of uranium in irradiated fuel
i
steed cas
ng. The standard flasks am of the following types:
TABLE III
La Hague standard flask types
Loaded Weight
Capacity
to
Bh'R
PWR
TN 17/2
72
17
6
TN 12/2
102
39
12
TN 1./2
110
-
11
:r, The design. consists of a thick walled steel cyli:?ider clad
internally with stainless steel and externally with multiple copper
fins. Fuel is transported dry in these flasks, of which over 50 are
~Aow in operation.
The wet flasks weigh 70 tonnes and have a capacity of 17 BWR or
't pWR elements. A lead liner is enclosed in a thick walled steel
cylinder with external c1rcubf:)c4r.?1al dteel fins. The flasks are
.water filled.
~, a... a..c a..a.co t.u vrlglnal concepts: urloadieg in a dry cell in
'epatrast to a pond, and automatic nnndling. The hybrid family of
' glasks described above
..
are
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continue to be handled in the NPR. Vhilst ullsstand which,
desirable, there are many older
of dimensional or weight limitations, can not handle the larger
standardized flasks.
Fast reactor fuel transports
Irradtated fuel from the Phenix fast reactor has been transported
to La plus Hague since i ea year. at a Trate of aut 1 flask weighs 18 metric
pons plutonium per y f
tons and is lead shielded with external neutron shielding by an
annulus of water. Removable fins of copper and stainless steel
are two four sm ntled fuThe capacity o el assemblies= the flask is the
equivalent of
Transport methods to La Hague
Transports to La Pague are effected by road, rail or sea. Road
transports for the complete journey are used only for light flasks
of the NTL 8 or NTL 9 type which weigh 38 uetric tons and are
provided with their own semi-trailers. The use of this type of
transport is now limited to four reactors in Belgium, Hland,
Switzerland and Frac a where dimensional or geographical cot.sider-
been t as e followsfl~ual number of
nrensports to light La Hague flasks have advantage.
road t
Road transports to La Hague
1973
1974
1975
1976
1977
1978
1979
1960
1981
1982
9
21
90
45
70
54
56
57
74
90
The normal annual number of road transports has now:.,rcached
equilibrium around 55, but 1981 and 1952 saw an unusual irzprease due
to the removal of fuel from the Cundre?ingen i reactor, which has
been permanently removed fror service.
TrNUspor_s by sea to La Hague are performed
Pacific Nuclear Transport Limited and from Sweden by a
Special double-hulled ships discharge the flasks in the port of
Cherbourg onto rail wagons which are transported 20 km to the rail
terminal at Valognes. From Valognes they are handled like all
other rail transports to La Hague. The numbers of sea transports
to La Hague are as shown in Table V.
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road vehicle. Tne rlasHs tnen travel the rlnal 4u ea between Tne
lelognes terminal and La Hague along quiet country roads. The
14 Hague site has no connection to the railway system and the terrain
would make such connection difficult and costly. The Valognes
germinal has therefore been established to provide a marshalling yard
capable of handling many wagons simultaneously, with two travelling
bridge cranes of 130 metric tons capacity. The terminal also includes
facilities for maintenance of the road and rail vehicles. The annual
probers of rail transports to La Hague are shown in Table VI.
1966
1967
1968
1969
1970
1971
1972
1973
1974
1975
27
75
83
117
98
50
V#5
215
282
266-
1979
1980
1981
1982
13
14
36
20
29
32
99
106
The basic method of transport to La Hague is by rail, since most
of the flasks weigh from 80 to 110 metric tons. A special 8 axle
.ii wagon hss been developed for this traffic which is authorised
for normal rail freight speeds. The irradiated fuel flask wagons
r form part of normal freight trains. The trains arrive at the Valognes
The peaks in 1974 and 1975 correspond to the peaks in the transport
of natural uranium. Whilst transports of graphite-gas elements have
declined steadily since 1974, transports of light water reactor
elements by rail have steadily increased and will continue to do so.
to site trolleys which are used for storage on a large open-air
ilaek handllnR at La Hague
All flasks arrive at La Hague by road and are delivered to a
trolley loading plant known as A.L. Here the flasks are transferred
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corresponding to 2,000 metric tons or uranium, and also to the RAO plant.
The average flask hcndling capacity of NPR is 350 flasks per year
working through two separate unloading lines. In NPH, the flask is
lowered into the pond and fuel elements are extracted under water.
The future UP3A will include a unit which extracts the elements dry
from the flasks and transfers the- to. individual pits prior to
underwater storage in the pond.
Fuel elements are stored bare but rigorous tests are carried out
to ensure that failed fuel is segregated at the reactor and transported
only under special cond'tions of encaps2latitn. At the reactor site,
procedures exist to identify failed elements by activity measurement
in an enclosed environment with a known temperature increase. Where
this procedure is not yet in use, tests are made by means of fuel
pping flasks,transport,
dry flasks. Failurei~
sampling initheocaseeofflask
to meet the criteria involves rejection of the fuel at the reactor site.
Transport organisation
With the exception of the sea transports referred to above, all
transports to La Hague are performed by COGEMA, for French domestic
transports and Nuclear Transport Limited (NTL) for the rest of Europe.
NTL is an international organization with branches in England, France
and Germany. The management of transport operations from Germany,
Belgium, Holland and Switzerland to La Hague is centralized in the
Paris office. Most non-French transports originate from Germany where
NTL has a base at Hanau. NTL's services commence at the reactor site,
where technicians attend to provide advice to the reactor operator,
to verify that the fuel elements comply with the COGEMA acceptance
criteria, to check the fuel element identification numbers and to
accept the flask for transport after it has passed controls on
leak-tightness, radiation and ciistamination. Transport is then
organized by the appropriate NTL branch acting in concert with their
colleagues who will take over responsibility at an agreed hand-over
point.
. NTL acts as the sub-contractor of the reprocessor and forms the
bridge between the reactor and the reprocessing plant. Statistics
such as metric tons transported or number of flask movements
sometimes fail to present an adequate picture of the magnitude of
the activity. The number of light water reactor irradiated fuel
elements transported by NTL is now in excess of 5,000 - the
equivalent of more than 30 full reactor cores.
r
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MINISTRY OF RESEARCH AND INDUSTRY
FRANCE
HIGHER COUNCIL FOR NUCLEAR SAFETY
APPENDICES TO THE REPORT
OF
THE WORKING GROUP
ON
THE MANAGEMENT OF IRRADIATED FUELS
DECEMBER 1981 - NOVEMBER 1982
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DECEMBER 1981 - NOVEMBER 1982
Tonnaoe retraite ?
Lip
Annee ,
Taux moven de combustion
Dose 0
Teips
collective
d'arrets
^S
I 6
O
06o p
L ~rZ ?
homme /reins
(moi s )
1968 188,7
-
1 166
-
223,3 8,5 1
1969 157,4
-
986
-
222,8 10
1970
244,7
-
1 079
-
372,4 S
1971
126,3
- Y,
2 287
-
362,1 8,5 }
1972
250,4
-
2 164
-
344,2 5
1973
212,9
-
2 385
-
507,4 8
1974
634,5
-
2 331
-
543,5 4
1975
442,7
-
3 038
-
714,4 4,5
1976
217,8
14,62
2 783
15 803
700,8 5
1977
354,7
1 SS,02
2 947
28 077
673,1 4
1978
~ 371,4
38,22
3 345
27 211
_ 633,8 3
1979
240
79,41
3 590
20'375!
561.9 3
980
252
104.,86
3 317
20 980
643,2 3.
TABLE VIII: OPERATING PARAMETERS OF LA HAGUE ^+J
* Data of La Hague radiation protection department [probably
applies only to 1981 -- cut off in original -- -.ranslator).
Key: 1. Year.
5.
1- lean. burnup, MWd/t
2. :onnaae reprocessed
6.
Collective dose, man-reins
3. GGR
4. Light-water PWR~
7.
Stoppage periods, months
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Table ? n' 1 - REPROC S-3114-AT LA HAGUE
- Tonnage Reprocessed and Dose Indicators -
Co1?lectiv
Dose/Power
Year
Tonnage Rep
rocessed
Average Bur
n-up
Dose
Ratio
?
(tor-n
es)
(t".1 days/t
onne)
(rran-rem)
(re'J!!We yr)
metal"
oxide"
metal
oxide"
1968
189 -
-
1 170
-
223.3
1,106
1969
158
-
990
-
222.8
1.560
?
1970
245
-
1 080
-
372.4
11541
1971
126
-
2 290
-
- 362,1
1,374
1972
250
-
2 160
-
344,2
C.699
1973
213
-507,4
11091
1974
635
.543.5
C,402
1975
443
-
3 038
714,4
C,581
1976
218
14,6
.-2;783
15 800
700,B
0.916
1977
351
17.3
:,.2'947
26 C60
673,1
C1465
1978
372
38,2
3 345
27 270
633,8
0,304
1979
240
79,4
3 590
20 375
561,9
C,248
1980
252
104,9
3 317
20 960
643.2
0,232
1981
250
101,3
3 672
25 420
727,5
0,228
1932
226
153,5
3 720
-21 095
602,8
C,162
1983
117
221,3
3 727
23 230
590.7
01116
1968-1963
4 285
730.5
8 424
0,373
Sources : /ZER79/, /CAS22/ (1969 to 1981)
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SE RET
French Reprocessing Activities
This is a complete list of all the data provided to us by LULL on oxide fuel
reprocessing activities at Cap de la Hague. The column headings have the following
meanings:
NUM-ASBL number of fuel assemblies reprocessed
KGU-TOTAL mass of uranium in spent fuel in kilograms
I% U235 weight percent uranium-235 before assemblies inserted into reactor
INSRT DATE month/year assemblies inserted into reactor
DISCH DATE month/year assemblies removed from reactor
BURNUP total irradiation received by assemblies
D%-U235 weght percent uranium 235 in discharged fuel
KG Pu-TOTAL mass of plutonium in spent fuel in kilograms
A blank spot in a column indicates that LLNL does not have data for that parameter.
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LWR FUEL REPROCESSED AT LA HAGUE
FIRST CAMPAIGN
APRIL - JUNE 1976
NUM- KGU- I%-
INSRT DISCH RURNUP
D%-
KG Pu-
ASBL TOTAL U235
DATE DATE MWD/TE
U235
TOTAL
MUEHLEBEPG (SWITZERLAND)
1
196 2.39
3/71 1/74 8400
1.62
0.7
12
2346 2.39
3/71 8/74 13000
1.30
12.3
63
12080 2.39
3/71 8/74 14200
1.24
67.7
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SECRET
LWR FUEL REPROCESSED AT LA HAGUE
SEODND CAMPAIGN
NOVEMBER 1977 - MARCH 1978
NUM-
ASBL
KGU-
TOTAL
I%-
U235
INSRT D
DATE D
ISCH BUR
ATE M
NUP
D/TE
D%-
U235
KG Pu-
=AL
3
1056
2.38
1/72 7
/73 15
790
1.12
6.4
1
352
3.19
1/72 7
/73 18
136
1.64
2.3
33
11616
3.38
1/72 6
/74 23
196
0.78
97.3
16
5632
2.53
1/72 6
/74 23
726
0.83
47.3
1
352
3.19
1/72 6
/74 26
427
1.16
2.9
18
6336
2.53
1/72 5
/75 27
993
0.67
53.7
32
11264
3.19
1/72 5
/75 31
051
0.94
100.0
4
1408
2.38
1/72 7
/73
7/74 5
/75 21
839
0.83
11.8
5
1760
2.18
1/72 7
/73
6/75
4/76 23
000
0.68
14.6
2
704
2.53
1/72 7
/73
6/75
4/76 25
000
0.78
5.6
3
1056
3.19
1/72
7/73
6/75
4/76 26
237
1.17
8.6
2
704
2.53
9/73
4/76 25
000
0.78
5.6
42
14868
3.18
9/73
4/76 31
500
0.92
132.8
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LFWR FUEL REPROCESSED AT LA HAGUE
THIRD CAMPAIGN
DECEMBER 1978 - APRIL 1979
NUM-
ASBL
KGU-
TOTAL
I%-
U235
INSRT
DATE
DISCH
DATE
BURNUP
FVD/TE
D%-
U235
KG Pu-
TOTAL
30
9420
2.5
6/73
1/75
15000
1.26
47.3
8
2512
2.80
6/73
1/75
16160
1.44
14.9
32
10048
2.80
6/73
2/76
25600
0.93
81.1
5
1570
2.50
6/73
2/76
22900
0.86
13.2
27
8208
3.10
6/73
1/77
30800
0.90
72.5
9
1140
2.20
8/66
7/69
7/70
5/73
12214
1.20
5.9
5
621
2.24
7/70
5/73
15597
1.02
3.8
9
1156
2.40
7/70
5/73
15597
1.10
7.1
1
128
2.42
8/69
6/71
7/72
5/73
15735
1.13
0.8
1
127
2.20
8/66
5/74
21459
0.73
0.9
1
127
2.20
8/66
7/67
21244
0.74
0.9
23
2939
2.42
8/69
5/74
20271
0.89
20.9
15
1863
2.42
8/69
5/74
20283
0.89
13.3
10
1267
2.20
8/66
7/69
12956
1.16
6.8
16
1987
2.24
7/70
5/74
18160
0.91
13.3
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 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
SECRET
25X1
?
LWR FUEL REPROCESSED AT LA HAGUE
FOURTH CAMPAIGN
DECEMBER 1979 - JUNE 1980
NUM- KGU-
I%-
INSRT D
ISCH BURNUP
D%-
5
KG Pu-
ASBL TOTAL
U235
DATE D
ATE MWD/TE
U23
TOTAL
4 733
2.39
3/71
8/73 5600
1.82
2.2
21 4027
2.39
3/71
8/74 14200
1.24
22.6
2 391
2.39
3/71 1
0/74 24541
0.73
3.1
6 1165
2.39
3/71
8/74
10/74
8/75 19034
0.97
1.3
42 8195
2.39
3/71
8/75 19034
0.97
55.8
70 13459
2.39
3/71
5/76 17875
1.02
88.1
4 782
2.39
3/71
8/74
9/75
5/76 17875
1.02
5.1
2 384
2.39
3/71
1/74
WURGASSEN
9/74
5/76 19034
0.97
2.6
91
17745
2.20
10/71
9/76 13581
1.11
99.1
2
390
2.20
10/71
8/77 17590
0.91
2.5
12
1490
2.24
7/70
5/74 18160
0.91
10.0
1
124
2.42
8/69
5/70 14723
1.23
0.7
3
373
2.42
7/71
5/74 16530
1.10
2.3
2
253
2.20
7/72
5/74 23266
0.67
2.0
1
128
2.42
8/69
6/71
7/72
5/74 18529
1.01
0.8
1
124
2.42
8/69
6/71
7/72
5/74 18541
1.01
0.8,
1
128
2.42
8/69
4/72
6/73
5/74 19108
0.99
0.9
1
124
2.42
8/69
4/72
6/73
5/74 19120
0.99
0.8
1
128
2.42
8/69
5/75 21269
0.86
0.9
13
1615
2.24
7/70
5/75 19530
0.85
11.3
1
128
2.42
8/69
5/70
-7/71
5/75 20137
0.89
0.9
2
257
2.40
7/71
5/75 18562
1.00
1.7
44
5465
2.42
7/71
5/75 18624
1.01
36.6
3
383
2.42
8/69
6/71
7/72
5/75 20018
0.90
2.7
6
745
2.24
7/70
6/71
7/72
5/75 17810
0.92
4.9
6
745
2.42
7/72
5/75 15992
1.12
4.5
10
1242
2.40
7/72
5/75 15952
1.11
7.6
1
124
2.42
7/70
4/72
6/73
5/75 26547
0.64
1.0
SECRET 25X1
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2 ~
 Sanitized Copy Approved for Release 2010/09/01: CIA-RDP88R01225R000100310003-2
LWR FUEL REPROCESSED AT LA HAGUE
FOURTH CAMPAIGN
DECEMBER 1979 - JUNE 1980
NUN- KGU- I%-
INSRT
DISCH
BURNUP
D%-
KG Pu-
ASBL 707"AL U235
DATE
DATE
KaD/TE
U235
TOTAL
GUNDREMMINGEN-A (cont)
2
248
2.24
7/70
4/72
6/73
5/75
18376
0.90
1.7
2
257
2.40
7/70
4/72
6/73
5/75
18973
0.98
1.7
4
514
2.40
7/70
5/73
6/74
5/75
18287
1.01
3.4
1
124
2.42
7/71
5/73
6/74
5/75
15833
1.13
0.8
10
1242
2.42
7/71
5/76
20104
0.89
8.8
6
745
2.42
7/72
5/76
18263
1.02
4.9
33
4099
2.40
7/72
5/76
18217
1.02
27.1
8
994
2.41
6/73
5/76
16329
1.11
6.1
14
1739
2.42
6/73
5/76
16367
1.10
10.7
1
128
2.40
7/70
5/73
6/74
5/76
19819
0.94
0.9
4
514
2.40
7/71
5/73
6/74
5/76
18064
1.02
3.4
4
516
2.40
7/71
1/77
21122
0.85
3.7
27
7837
3.10
9/68
8/71
10/72
9/73
30308
0.92
68.7
3
831
3.10
9/70
9/73
30909
0.89
7.4
1
255
2.85
9/70
9/73
31172
0.73
2.3
1
291
2.50
9/68
8/70
9/71
9/72
24495
0.70
2.4
2
554
3.10
9/70
9/72
9/73
8/74
27696
1.04
4.6
1
291
2.80
9/68
8/71
24527
0.98
2.3
27
7856
2.50
9/68
8/70
9/71
9/72
24495
0.78
63.7
4
1161
3.10
9/68
9/72
29754
0.93
9.0
1
291
3.00
9/70
9/72
18053
1.48
1.9
1
290
3.10
9/68
9/73
37480
0.64
2.8
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
qprprT
LWR FUEL REPROCESSED AT LA HAGUE
FOURTH CAMPAIGN
DECEMBER 1979 - JUNE 1980
NUM-
KGU-
I%-
INSRT DISCH BURNUP
D%-
KG Pu-
ASBL
TOTAL
U235
DATE DATE MWD/TE
U235
TOTAL
DOEL-1
35
9240
2.01
7/74 2/76 13522
0.95
49.9
DOEL-2
35
9272
1.99
8/75 11/76 13716
0.98
51.0
TIHANGE-1
39
17745
1.95
4/75 9/76 15562
0.82
108.3
32
11264
2.18
1/72 7/73 15305
1.00
66.1
2
704
2.53
1/72 7/73 16177
1.22
4.4
1
354
3.18
9/73 4/76 31500
0.92
3.2
11
3894
3.18
7/74 4/76 21739
1.41
28.3
1
352
3.19
1/72 5/75
5/76 4/77 35631
0.76
3.3
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
LWR FUEL REPROCESSED AT LA HAGUE
FIFTH CAMPAIGN
FEBRUARY - JULY 1981
NUM-
ASBL
KGU-
TOTAL
I%-
U235
INSRT
DATE
DISCH
DATE
BURNUP
M'D/TE
D%-
U235
KG Pu-
TOTAL
22
6900
2.96
10/66
6/73
25800
1.02
58.0
28
8800
3.76
10/66
6/73
23200
1.58
66.4
1
300
2.50
6/73
1/77
22100
0.89
2.7
5
1600
2.50
6/73
11/77
22500
0.89
12.6
9
2800
2.80
6/73
2/76
25600
0.93
22.8
1
300
3.10
6/73
2/76
26300
1.11
2.5
12
3600
3.10
6/73
1/77
30800
0.90
32.2
1
300
3.3
4/75
2/76
12900
2.10
1.6
4
1300
3.30
4/75
1/77
21000
1.54
8.9
'
4
1300
2.95
4/75
11/77
33400
0.73
11.8
4
1200
3.10
4/75
11/77
29500
0.93
11.1
10
3100
3.30
4/75
11/77
30000
1.02
29.5
1
300
2.50
9/68
8/74
26500
0.70
2.4
6
1700
3.10
9/68
8/74
27500
1.04
14.5
1
300
2.85
9/70
9/73
31200
0.73
2.3
19 -
5300
3.10
9/70
8/74
27700 .
1.04
44.0
3
800
3.10
9/70
9/73
30900
0.89
7.4
16
4400
3.10
9/71
6/75
24300
1.20
34.2
2
600
3.10
10/72
6/75
27900
1.03
4.6
5
+1800
2.18
1/72 ?
4/76
23000
0.68
14.6
2?
700
2.53
1/72
4/76
25000
0.78
5.6
1
400
3.19
9/73
4/77
34200
0.81
3.3
4
1400
3.18
9/73
4/77
34100
0.81
13.2
1
400
3.18
9/73
4/77
30600
0.96
3.1
1
400
3.18
9/73
4/77
24500
1.25
2.8
20
7100
3.18
14/74
4/77
29400
1.01
61.2
6
2100
3.18
7/74
4/78
33200
0.85
19.5
.4
1400
3.18
7/74
4/78
28200
1.06
12.0
7
2500
3.18
6/75
4/78
29300
1.01
21.4
1
400
3.18
6/75
4/77
24200
1.26
2.7
44
15700
1.90
5/76
7/77
12800
0.89
89.0
5
1800
1.90
5/76
8/78
20800
0.60
14.4
31
11000
2.50
5/76
8/78
24000
0.81
90.4
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
LWR FUEL REPROCESSED AT LA HAGUE
SIXTH CAMPAIGN
7 DECEMBER - 30 JUNE 1982
NUM- KGU-
ASBL TOTAL
I%-
U235
INSRT
DATE
DISCH
DATE
BURNUP
KID/TE
D%-
U235
KG Pu-
TOTAL
DOEL-1
2 500
2.05
7/74
2/76
13500
0.95
2.9
32 8500
2.85
7/74
2/77
25500
0.96
66.1
2 500
3.42
8/75
11/76
11200
2.39
2.5
6 1600
1.99
8/75
10/77
21200
0.65
11.7
22 5800
2.84
8/75
10/77
24100
0.96
46.3
2 500
2.85
8/75
10/77
24400
0.95
4.3
3 800
2.01
2/77
10/77
20900
0.68
2.0
4 1800
3.10
2/75
9/76
19200
1.50
12.5
3 1400
1.95
2/75
9/76
15600
0.82
8.3
2 900
2.55
2/75
7/76
17400
1.17
6.3
4 1800
1.95
2/75
1/78
24200
0.52
9.8
21 9600
2.55
2/75
1/78
26200
0.76
57.1
TAKAHAMA-1
44
20200
2.00
3/74
11/75
15600
0.87
124.0
BORSSELE
.8
2500
3.3
4/75
11/77
30000
1.02
23.6
2
600
3.30
4/75
2/76
2/77
10/78
30500
1.02
5.4
28
8800
3.30
4/76
10/78
30500
1.02
72.2
9
2800
3.30
2/77
10/78
22500
1.40
20.6
25
11500
2.10
3/77
3/79.
15100
0.96
72.0
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01: CIA-RDP88R01225R000100310003-2
LWR FUEL REPROCESSED AT LA HAGUE
SIXTH CAMPAIGN
7 DECEMBER - 30 JUNE 1982
NUM-
ASBL
KGU-
TOTAL
I%-
U235
INSRT
DATE
DISCH
DATE
BURNUP
MSWD/TE
D%-
U235
KG Pu-
TOTAL
1
200
2.39
10/73
5/76
17900
1.03
1.3
8
1600
2.30
9/74
8/77
19400
0.91
10.4
68
12500
2.47
9/74
8/77
21300
0.95
90.1
4
800
2.30
9/74
8/78
20000
0.88
5.2
4
700
2.47
9/74
10/74
9/75
8/78
22600
0.84
5.2
36
6600
2.47
,9/74
8/78
25600
0.76
52.4
16
2900
2.74
9/75
8/78
23900
1.03
22.8
1
200
2.20
10/71
9/76
13600
1.11
1.1
104
20300
2.20
10/71
8/77
17600
0.91
132.0
23
4500
2.20
10/71
3/79
21200
0.76
32.2
4
500
2.42
7/71
1/77
21300
0.85
3.7
2
200
2.42
7/72
1/77
19900
0.85
1.8
24
2900
2.41
7/72
1/77
19900
0.94
21.3
1
100
2.42
7/71
4/72
6/73
1/77
20100
0.89
0.9
4
500
2.41
6/73
1/77
18000
1.03
3.3
102
28700
29000
260.0
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88RO1225R000100310003-2
LWR FUEL REPROCESSED AT LA HAGUE
SEVENTH CAMPAIGN
1 MARCH 1983 - 30 JUNE 1983
NUNI- KGU-.
ASBL TOTAL
I%-
U235
INSRT
DATE
DISCH
DATE
BURNUP
MWD/TE
1 460
1.95
02/75
01/78
24200.
9 4120
2.55
02/75
01/78
25628.
6 2740
2.55
02/75
01/79
34500.
3 1370
3.20
02/75
01/78
28535.
9 4120
3.10
02/75
01/79
33700.
1 460
3.20
03/78
01/79
11700.
7 2450
3.18
07/74
03/79
31000.
*
18 6290
3.18
06/75
03/78
30340.
5 1750
3.18
06/75
03/97
33500.
4 1400
3.18
06/75
03/79
31000.
16 5600
3.18
05/76
03/79
32700.
1 310
3.30
04/76
10/78
30500.*
*
3 940
3.30
04/76
10/78
30500.
*
28 8790
3.30
02/77
10/79
31600.
12 3770
3.30
12/77
10/79
24200.
4 480
2.40
07/72
01/77
15734.*
*
34 4120
2.40
06/75
01/77
10013.
*
31 3760
2.41
06/75
01/77
10032.
*
27 3270
2.41
10/76
01/77
3021.
ISAR-1 (KKI)
204
37830
1.94
11/77
04/80
12925.
D%- KG Pu-
U235 TOTAL
0.52 2.4
0.78 33.8
0.47 26.3
1.07 11.8
0.79 38.6
2.11 2.21
0.95 22.3
0.97 . 56.6
0.84 16.6
-0.95 12.7
0.88 52.5
1.02 2.7
1.02 8.1
0.99 79.6
1.37 28.2
1.12 3.0
1.47 17.5
1.48 16.0
:2.08 5.2
0.95 207.8
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
L1R FUEL REPROCESSED AT LA HAGUE
SEVENTH CAMPAIGN
1 MARCH 1983 - 30 JUNE 1983
NUM- KGU-
I%-
INSRT
DISCH BURNUP
D%-
KG Pu-
ASBL TOTAL
U235
DATE
DATE MWD/TE
U235
TOTAL
10 3300
3.20
7/75
6/78 27,400
1.03
29.30
1 300
3.21
7/74
5/77
6/78
6/79 29,000
1.08
3.1
4 1300
3.04
7/76
6/79 24,800
1.15
11.4
8 2600
3.20
7/75
6/79 34,200
0.88
24.6
8 2600
3.21
7/75
6/79 35,000
0.81
25.1
4 1300
3.21
7/74
5/76
7/77
6/79 33,200
0.88
12.2
BEZNAU 2 (Switzerland)
2 700 3.21
9/73
7/75 19,200
1.61
4.7
1 300 2.50
10/71
7/73
8/74
7/76 20,800
0.96
2.8
16 5300 2.78
10/71
7/74
8/75
7/76 24,100
1.02
45.1
12 4000 3.51
10/71
7/75 29,200
1.26
36.1
25 8300 3.22
9/73
7/76 24,600
1.30
84.5
11 3600 2.78
10/71
7/75 26,000
0.93
31.7
TAKAHAMA 1/2
21000
16,000
BUGEY 2
16 7300 2.10
4/78
4/80 14,400
1.01
40.
NECKARWESTHEIM-1 (FRG)
1 370 1.90
05/76
07/79 21183.
0.59
2.9
8 2950 2.50
05/76
08/78 24000.
0.81
24.3
7 2580 2.50
05/76
08/80 30000.
0.59
22.3
4 1480 3.20
05/76
08/80 29500.
1.03
12.6
49 18080 3.20
05/76
07/79 32418.
0.90
160.8
1 370 3.20
05/76
08/80 37400.
0.71
3.5
14 5170 3.20
09/77
08/80 29381.
1.03
44.5
Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
 Sanitized Copy Approved for Release 2010/09/01 : CIA-RDP88R01225R000100310003-2
LWR FUEL REPROCESSED AT LA HAGUE
SEVENTH CAMPAIGN
1 MARCH 1983 - 30 JUNE 1983
NUM-
ASBL
KGU-
TOTAL
I%-
U235
INSRT
DATE
DISCH
DATE
BURNUP
MWD/TE
D%-
U235
KG Pu-
ZCYTAL
BORSSELE (Netherlands)
1 310 3.30
4/76
10/78
30,500
1.02
2.7
3 940 3.30
4/76
10/78
30,500
1.02
8.1
28 8790 3.30
2/77
10/79
31,600
0.99
79.6
12 3770 3.30
12/77
10/79
24,200
1.37
28.2
4 480 2.40
7/72
1/77
15,734
1.12
3.0
34 4120 2.40
6/75
1/77
10,013
1.47
17.5
31 3760 2.41
6/75
1/77
10,032
1.48
16.0
27 3270 2.41
10/76
1/77
3,021
2.08
5.2
ISAR-1
204 37830 1.94
11/77
4/80
12,925
0.95
207.8
STADE
7 2450 3.18
7/74
3/79
31,000
0.95
22.3
18 6290 3.18
6/75
4/78
30,340
0.97
56.6
5 1750 3.18
6/75
3/79
33,500
0.84
16.6
4 1400 3.18
6/75
3/79
31,000
0.95
12.7
16 5600 3.18
5/76
3/79
32,700
0.88
52.5
TIHANGE-1 (Belgium)
1 460 1.95
2/75
1/78
24,200
0.52
2.4
9 4120 2.55
2/75
1/78
25,628
0.78
33.8
6 2740 2.55
2/75
1/79
34,500
0.47
26.3
3 1370 3.20
2/75
1/78
28,535
1.07
11.8
9 4120 3.10
2/75
1/79
33,700
0.79
38.6
1 460 3.20
3/78
1/79
11,700
2.11
2.2
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CO2 Reactor Fuel Element Data
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CONCEPT, DEVELOPMENT AND RELIABILITY OF--
FRENCH CO2-COOLED REACTOR FUEL ELEMENTS
D. BASTIENs
1. INTRODUCTION
The natural uranium-graphite-gas reactor series consists of 8 power plants
in France, 6 of which are still in operation, and one reactor in Spain.
The first, Marcoule G.2, whose first criticality took place in 1958 and
that was shut down on February 1st 1980, had a capacity of 40 MWe. The
last one to start up in 1972, BOGEY 1, is 540 We net.
Improvements of the nuclear steam supply system have been accompanied by
successive transformations of the fuel elements which have now reached a
high degree of reliability.
2. DESIGN AND DEVELOPMENT OF FUEL. ELE'ID4TS
Before passing on to the power plant stage France had built an unpressu-
rized air - cooled reactor, G.1, where spectacular deformations of the
unalloyed uranium were observed. This fuel consisted of magnesium clad
unalloyed uranium rods.
2.1. G.2-G.3 fuel element
For power reactors it was necessary to define a not too deformable uranium
alloy and a cladding material behaving well under CO2 at high temperature.
C E A - DEDR - C EN /SacZay - F-91191 GIF-sur-YVETTE - CEDEI -
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c
The choice of cladding material quickly-settled on a low-zirconium (0.6 %)
magnesium alloy. Easy to transform, amenable to argon arc welding, ductile
when hot and not very absorbent neutronically it possessed-all the
essential qualities required.
Its maximum service temperature still had to be proved compatible with
the increased in-pile performance of the fuel elements. This temperature,
set first at 400 ?C, was gradually raised to 51S ?C (Mg-Zr melting point
660..?C) while at the same time the longitudinal cooling fins gave way to
herring-bone fins.
Having given entire satisfaction this Mg-Zr material was kept for all
other types fuel elements in the series. Its only disadvantage is its
permeability to plutonium, a defect corrected by the interposition of a
thin graphite lining between the cladding and the fuel.
For the uranium rods a relatively non-absorbent alloy SICRAL F1 (0.07 % Al,
0.03 % Fe), deforming little under irradiation, was chosen. _
2.2. CHINON 1 fuel element
To increase the specific power extracted from the fuel element the rods
were replaced by tubes. At given maximum uranium and cladding temperatures
it is possible to extract more heat per unit channel length from a
tubular fuel than from a rod, which means that for a given reactor power
the number of channels necessary is reduced.
Since SICRAL F1 is not mechanically resistant enough to withstand Veep
under compression a new alloy was needed for this tube closed at both ends
by welded caps. A compromise had to be found conciliating neutron
absorption, mechanical resistance and swelling properties, and an 0.5 %
molybdenum- uranium alloy was finally adopted.
The CEIINON 1 reactor, shut down after 10 years' service for economic
reasons, was the first one designed with vertical channels in which the
fuel elements were stacked directly one on top of another. The ends of
the cans were thus subjected to considerable stresses, limiting the
possibilities of this type of longitudinally finned fuel element.
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2.3. CEIINON 2, GUNON 3, ST-LAURFNT 1 tubular fuel elements
To gain more specific power the uranium tube diameter was increased (43 x
23 mm) and consequently the molybdenum content of the alloy had to be
raised to 1.1 % to improve its mechanical resistance ; at the same time
the can was fitted with herring-bone fins, the geometry of which has
gradually been optimised through very detailed thermal studies in which
the fin height, profile shape, spacing and angle of inclination were
varied. Moreover, the fuel element was housed in an individual graphite
sleeve to limit mechanical stresses on the ends, each element supporting
only its own weight. This arrangement has contributed greatly to the
reliability of the fuel elements, especially at the time of handling which
takes place under running conditions.
On the other hand the resistance of this type of fuel was limited by creep
in the uranium tube and end caps, and the alloy used was going to be replaced
by a quaternary alloy, MOSNAL, containing 1 % Mo, 0.05 % Sn and 0.05 % Al
just then however a new design of fuel was conceived, a timely event
because MOSNAL, loaded in small amounts in ST-LAMB T 1, proved difficult
to reprocess industrially.
2.4. Graphite core fuel elements
This kind of fuel element has the same geometry as the tubular fuel and
the two are interchangeable, but the new version is different in 2 respects :
- the graphite kernel from the casting process has been left inside the
43 x 23 mm uranium tube (whence its name "graphite core fuel element")
- the uranium alloy is different : since the graphite kernel is there to
take pressure stresses, the mechanical properties of the uranium can be.
less stringent and the U-1.1 % Mo alloy is replaced by SICRAL F1 ,already
used for G2 and G3.
Other component such as plugs, cans, sleeves are the same as those of the
tubular element.
These graphite core elements have three main advantages :
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1) Higher_burn_up
Whereas for the U- 1.1 % Mo tubular element the technological and
neutronic limits are pratically the same (5 000 MWd/t) the graphite core
element, which uses a less absorbent fuel, offers greater neutronic possi-
bilities. It was therefore possible to increase the fuel irradiation level
to 6 500 MWd/t without overstepping the technological limits of the element.
This represents a fuel saving of 30 % over and above the 20 % saved
by the axial rearrangement of 3 elements out of the 15 contained in each
reactor channel.
2) Higher_working_temperatures_and_pressures
The working temperature of this type of element (tubular or graphite -
S core) is limited by the maxim nn temperature admissible on the uranium.
Owing to the presence of the graphite kernel this temperature can be raised
from 640 ?C to 650 ?C and the CO2 pressure from 26.5 to 28.5 bar, corres-
ponding to a possible increase of about 12 % in the reactor power.
3) Greater safety
Reducing.the-free volume inside the element reflects to a large extent
on how the oxidation of the uranium tube develops after a cladding failure
The presence of the graphite core, chosen non-porous, is thus an important
safety factor.
Besides possessing these three advantages the graphite core element
is simpler to manufacture than the tubular element and the fuel is therefore
noticeably cheaper.
M~ .
For all these reasons this element was chosen for the first fuel load
of ST-LAURENT 2 and VANDELLOS reactors and as replacement element for
QUNON 2, CHINON 3 and ST-LAURFNT 1.
2.5. BUGEY 1 annular fuel element
To obtain even higher specific powers an annular fuel was designed: The
principle is to cool a large uranium tube by outside and inside cladding,
which also means that no internal volume remains and high coolant gas
pressures can be reached.
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For BUGEY the diameters of the SICRAL F 1 uranium tribe were fixed at 95 x 77 mm,
providing 12 W/g specific power and high reactor power with few channels
(2 900 channels for 1 700 MWth at ST-LAURFNT 2 against 852 channels for
2 000 MWth at BUGEY 1).
This element took longer to develop. It was necessary in particular to bind
cladding and fuel together metallurgically in order to avoid detachment of
the inner can in certain thermal transients. This was achieved via an
.aluminium layer deposited by Shoop process, which diffuses into the cladding
and uranium to give a metallurgical bond.
This fuel eventually proved almost as reliable as the graphite core element.
3. RELIABILITY OF FUEL ELEMENTS
3.1. For 51 000 nominal fuel elements loaded in CHINON 1 six cladding failures
were'observed, representing a failures rate of 10/100 000.
These failures were mainly caused by the stacking method of fuel element
loading.
3.2. Of the 211 400 nominal U-1.1 % Mo tubular fuel elements loaded 22 cladding
failures occurred, a failure rate of 10/100 000. These were largely due
to localized inward tube deformation resulting from uranium creep.
3.3. In spite of its enhanced performances (maximum cladding temperature 515?C,
maximum uranium temperature 650 ?C, specific burn-up 6 500 MWd/t) the
graphite core fuel only included 7 failures amongst the 466 000 elements
loaded, i.e. 1.5/100 000. Manufacturing defects are responsible here.
3.4. The annular fuel element appeared slightly less reliable with 3 cladding
failures for 76 S00 elements loaded, which represents a failure rate of
less than 4/100 000. These again are due to manufacturing faults.
The favorable trend of these figures has been obtained by in-loop and in-pile
irradiations of experimental and standard fuel elements, examined afterwards
in the CEA hot laboratories, and by strict supervision of their manufacturer
COG2IA to maintain the same high standards throughout.
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HEAVY WATER TRITIUM/PLUTONIUM PRODUCTION REACTOR FUEL
Celestin-1 diverged on 15 May 1968. Celestin-2 diverged on
30 October 1968. These reactors were originally designed to
produce tritium. They are heavy water reactors with a thermal
power of 200 MW each. The reactors do not generate
electricity.
The original fuel for the reactors appears to have been a
PuAl alloy. It was reported that PuAl fuel from Celestin-1 was
reprocessed in the second half of 1970 and that 50 kg of
plutonium was recovered. Reprocessing of UA1 alloy fuel from
Celestin reportedly began in 1973.
In 1976 the reactors "received a plutonium breeding
vocation", like that of the G2 and G3 reactors. The same
document, from the early 1980s, stated that the irradiated fuel
was stored for "over 9 months".
STAT
STAT
STAT
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Fast Reactor Reprocessing Data
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FAST REACTOR REPROCESSING DATA
MARODULE: SAP plant - 10-30 kg/day capacity
50 kg Fortissimo MOX (30% Pu)
14-76 GWD/t
1976
1,650 kg KNK
U02
3.5-7 G -JD/t
1977-1978
2,300 kg Phenix
U02
38-45 G[mD/t
1979
150 kg Phenix
220 kg Phenix
MOX (18% Pu)
MOX (25% Pu)
37 cMD/t
36-65 GWD/t
10-30 months cool
14-43 months cool
1980
840 kg Phenix
MOX (25% Pu)
36-65 G4D/t
1981
780 kg Phenix
MOX (25% Pu)
36-65 c D/t
14-43 months cool
1982
9,000 kg total reprocessed (implies 3 t in 82)
JAN-JUN 1983 1,600 kg Phenix MOX (25% Pu)
CAP DE LA HAGUE
AT-1 Plant (capacity 1 - 2 kg/day)
1969-1979 250 kg Rapsodie MOX (25% Pu) 40-55 Gti'D/t 6-12 mths cool
658 kg Rapsodie MOX (30% Pu) 50-120 G6'D/t 5-24 months cool
177 kg Phenix MOX (18% Pu) 8-44 cJD/te 18 months cool
UP2 Plant (by dilution with gas graphite fuel)
2,200 kg Phenix MOX (18% Pu) 21-42 GUD/te 38-50 months cool
OCT 1980
-JAN 1981 2,100 kg Phenix
OC'r-tJOV 1981 1,600 kg Phenix
NOV-DEC 1983 900 kg Phenix
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NOTICE: This material may.
be projected by copyright
law (Titlg A UA C04
114 Reprocessing Technology
bottles previously filled with mixtures of charcoal and graph-
ite.
Overall DFs are measured b% a comprehensive monitoring
system to assess the limits of the technical feasibility with re-
spect to the cost of each proposed technique. The off-gas
treatment plant is supervised from a central control room
with computer-assisted operation.
The extraction part of the HERMES facility is housed in
a second hot cell complex consisting of ten hot cells divided
in two main groups based on the fuel contents and on the
activity of the treated solutions and in three analytical cells.
Human intervention in some cells for maintenance has been
kept as an option.
Basically, the PUREX solvent extraction consists of code-
contamination of uranium and plutonium from the bulk of
the fission products, partition of uranium and plutonium, fol-
lowed by final decontamination of uranium and plutonium.
Solvent recovery and recycle is 'an important item in the
facility as well as waste concentration, tritium removal, and
concentration from the low-level liquid waste (LLW).
Five to 30%7r TBP in kerosene is used as solvent in the de-
contamination pulsed column. This column has three feed
inlets to parametrically investigate the influence of the resi-
dence time and the pulse conditions on the DF and on the
plutonium contents of the fission product flow. The feed rate
of this column is -40 Q/h and the feed solution is about I ,tf
in heavy metals.
The choice of the partition column has not yet been de-
fined and will result of the study of two possible reduction
methods of fu(l)l) to Pu(IV): reduction by hydroxylamine
nitrate or by electrolytic reduction.
Final purification of uranium and plutonium is also per-
formed in pulsed columns, the second and the third cycles
being performed in the same equipment, with necessary
rinsings between the two cycles.
Fission products are concentrated in an evaporator, the
distillate of which will be fractionated in LLW and recycled
acid in a second evaporator. Climbing film evaporator types
equipped with a stripping tower are used, and lFs of 101 are
expected.
Recognizing that a sodium carbonate wash will probably
be inadequate for washing sovent when used with such
-highly irradiated fuel and will generate tot) large amounts of
waste. salt-free methods will he applied, such as the use of
hydrazine carbonate. Other indispensable operations. such as
rework of solutions and waste storage, received considerable
attention in the planning stage.
The proposed HERMES facility has been designed under
the following bases of selection:
I . Minimum volume of waste is generated through the use
a standard Purex solvent extraction process with non-
waste-reducing auxiliary operations and reduction of
the cladding material volume
2. Fluorides used in the system will concern only a little
fraction of the fuel solutions and will he icnioved prior
to the addition of these solutions in the main stream.
3. Near-zero release of gaseous fission products to the en-
vironment.
4. Great versatility in the processes to be used in the ex-
traction part with the possihility of Inunan interven-
lion.
2. Status of Fast Reactor Fuel Reprocessing in
France, Jean Meg;', Jean Sauzeron, Michel Bour-
geois (CE-i/CEN-France)
INTRODUCTION
The French program fur the reprocessing of fast reactor
fuels forms part of the logical implementation of this reactor
system, in which the reactor has reached the industrial stage.
passing through the following three main phases:
1. Experimental phase. with the Rapsudie reactor at
Cadarache. The Rapsodie fuel cycle has been closed several
times, thanks to the reprocessing of its fuel in the AT I facil-
ity at La Hague.
-'. Demonstration phase. with the Phenix 250-MW(e) re-
actor at Marcoule. The Phenix fuel cycle is currently closed.
thanks to the reprocessing of the core-2 assemblies in the
Marcoule pilot plant (SAP = Service des Ateliers Pilotes) and
the reprocessing of the cure-1 assemblies in UP2 Plant at La
Hague. The TOR facility will serve to handle all the Phenix
fuels.
3. Industrial phase initiated with the I200-MW(e) proto-
type power plant under construction at Creys-Malville in co-
operation with FRG and Italy. The PL-RR (Prototype d'Usine
de Retraitement des Rapides-Fast Fuel Reprocessing Plant) is
under study for reprocessing of a small series of industrial
fast breeder reactors. comprising Creys-Malville.
The French research and development (R&D) programs.
specific for fast reactor fuels reprocessing, are oriented to-
ward the PURR prc~iect. the TOR project being one of the
most important.
FRENCH EXPERIENCE IN FAST REACTOR
FUEL REPROCESSING
The quantities of fast reactor fuel< reprocessed as of July
1. 148 1I. in the French facilities are given in Table I.
:1 TI I-virility- La Hague. The ATI facility. specially de-
signed to reprocess Rapsudie fuel with a capacity of I kg/day.
went on stream in 190 and by July 1971). when it was
definitively shut down, had reprocessed more than I ton of
heavy metals front mixed oxides irradiated to a hurnup of
120 000 MWd, tun, and sometimes only slightly "cooled"
(5 months, and 1.5 months for a small number of assemblies
).
thus achieving closure of the Rapsodie fuel cycle. Part of the
reprocessed fuel (more than 150 kg) was obtained from
mixed oxides of Phenix.
,Ilareoule Pilot Plant (SA P). The Marcoule Pilot Plant was
adapted to reprocess fast reactor oxides. It processed a hatch
of 50 kg of highly irradiated fuel from Rapsudie/Fortissimo
in 1975, a slightly irradiated but far larger batch (1650 kg)
from the German KNK reactor in 1976. and a more irra-
diated (over 45 000 MWd/ton) batch of 2.3 tons of Phenix
enriched-uranium fuels in 1977 and 1978. It is currently re-
processing Phenix plutonium fuels from core-2 irradiated to
-55 000 to 'o 000 \IWdton and cooled one year or more.
Up to July 1. 1981. nearly 3 tons of Phenix fuels have been
reprocessed at a rate of- 10 to 20 kg day.
However. due to the obsolescence of the present installa-
tions and to their extremely limited possibilities in terms of
capacity and R&D, the TOR project was launched. .
UP` Pianr-La Naiue More than 4 tons of heavy metals
of Phenix cure-1 fuels have been reprocessed in the UP2 plant.
treated by dilution with gas-graphite fuel dissolving solutions
after chopping and dissolution in IIA( Ihead end for light
water reactor i LWR) fuels I-
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 ..?. .: ,. _,._,.._?.....c....... a..w._;a~ra~uo.e: ,nesu,l
Special transport services. This depart-
ment called STS is involved in the design
and manufacture of casks, cask manage-
ment and the transport of radioactive mat-
ter. In liaison with STS, a department of
radioactive fuels transport (STCI) is res-
ponsible, in liaison with STS, for all trans-
port operations in the direction of La Hague
and Marcoule. It assures the supervision of
development, manufacture and mainte-
nance of casks used for radioactive fuels,
plutonium, waste and residues. It operates
railway terminals, road transport and the
cask maintenance workshop at La Hague.
Future. Cogema has 70 foreign clients,
some of whom are in the United States,
where it has achieved a spectacular
breakthrough. Japan is however the com-
pany's most important foreign customer.
Cogema also relies on its engineering sub-
sidiaries. Company chairman Francois de
Wissocq expects turnover will fall slightly
this year (because of the renegotiation of
the contract under which the USSR enri-
ched uranium on behalf of France). But the
cashflow will be "substantial". Invest-
ments will be increasing considerably this
year to reach 5.5 billion francs for Cogema
alone and 6 billion for the group as a
whole. E.L.
Subscription form see
. page 106
Aout 1984
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La Hague, the biggest work site
in Europe
(from our special correspondent in La Hague)
Arriving near the La Hague site is an ex-
perience, even a shock, with the change in
landscape. It is true that it is the biggest
work site in Europe and that the infrastruc-
ture are at the same scale. For the con-
struction requirements, Cogema set up the
economic interest grouping Gifab to,
operate a concrete batching and mixing
plant which can be taken down later. In
five years it will produce some 800,000
cubic metres of concrete, a quarter of
France's total production in a year.
The main work site - The plant which
already exists at La Hague has belonged to
Cogema since 1976. Under extensions
planned both for French and foreign fuel,
two major complexes are being built:
- UP3 A, a plant which is being completely
financed by foreign customers and which
will eventually include: a cask discharging
facility; two pools (C and D) with a capaci-
ty of 2,000 tons apiece (the C pool receiv-
ed its first basket of radioactive fuel in
April, while the D pool should be in service
towards the end of 1 985); a T 1 shearing/
dissolution unit, of which the civil
engineering work is currently in progress;
four separation units, treatment of fission
products, separation of uranium and
plutonium (civil engineering in progress); a
waste vitrification unit IT 7) with a capaci-
ty three times greater that of AVM at Mar-
coule (civil engineering work started in
May 1984); an STE 3/T waste treatment
plant. This UP 3 complex, the construction
of which is progressing at full speed, will
be in service in the middle of 1988, except
for the vitrification shop which is due to
start up at the end of March 1989.
- UP 2 800, the plant which is to come
onstream in 1990, two years after UP 3.
Already several shops of this complex, in-
tegrated to the facilities of the current
plant, are being built or in some cases have
even been completed. The BST 1
plutonium storage shop (completed at the
beginning of 1983); the AD1 decontamina-
tion shop (completed at the beginning of
1984); SPF 5, the storage unit n? 5 for
fission products (due to come onstream
before the end of the summer); NCP 1, a
facility for concentrating fission products
(start up planned for the end of 1984);
finally, R 7, the UP 2 800 vitrification shop
for vitrifying fission products resulting
from the reprocessing of 800 tons a year
of fuel from this plant. The start up is due
to take place in a little more than two
years.
Facilities that are common to UP 3 and
UP 2 800 - These include the new ST3
waste treatment station, which is due to
come into service at the end of 1986; the
site, that is to say all the infrastructural
facilities which need to be planned around
the nuclear buildings (roads and distribu-
tion networks, conduits, production and
distribution of fluids etc). This work site is
La Hague. Aerial view. UP3 programme: work in progress on the
"medium activity" building. (Doc. Cogemal.
being carried out alongside work being
done on the two plants. As part of "major
work site" procedure, Cogema is financing
a certain number of works and municipal
facilities in the region.
A lightening visit - In the pools in 10
metres of water lie the racks containing
the radioactive fuel assemblies. The latest
of these ponds to come into operation, the
C pool , the first facility built in the UP 3
extension, is suspended on big neoprene
blocks so as to assure an increased safety
in the event of an earthquake. The follo-
wing pools (D and E) are also suspended.
Along the wall of the pool can be found the
"Nympheas" built by SGN, which are in
fact very sophisticated exchangers desi-
gned to maintain the water at a constant
temperature and quality.
Everywhere, in the maintenance period
(July-August), it is the cleaning and main-
tenance of the material which produces a
sizeable amount of technological waste:
cotton, woven paper, polyethylene sheets
and industrial grade gloves. The La Hague
plant uses 7 million pairs of gloves a year.
For the casks there is an outside storage
area. Different models can be seen there:
cubic casks for graphite gas fuel, cylindri-
cal ones for LWR fuel, including Transnu-
cleaire casks or the recent Lemer model.
During treatment programmes around one
cask is received every day.
86 Energie Nucleaire Magazine
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APPENDIX 9
Marcoule, COGEMA: French Plutonium Site
(Translation)
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MARCOULE,COGEMA: FRENCH PLUTONIUM SITE
UNKNOWN
[Text] The Marcoule Mystique (1952-1958)
Marcoule's birth certificate is dated sometime during the winter of
1952-1953, when the [French] Atomic Energy Commission selected a flat stoney
hilltop, 15 by 20 meters, bordered on the east by the Rhone [river],
for the creation of the first French plutonium production center. The
site, dominated by the Dent de Marcoule, a steep hill 220 meters high,
is in the canton of Bagnols-sur-Ceze, in the northeast of the Gard departement,
at the crossroads of the Languedoc, Provence and Rhone-Alpes regions.
The geographic and hydrological reasons that caused its selection had
already made it the choice in Gallo-Roman antiquity for the construction
of a vast manufacturing and shipping complex for wine amphora from the
region.
In 1952, the CEA [French Atomic Energy Commission] was seven years old
Seven years that had been used to make up for the delay that France had
experienced due to the war in the area of nuclear energy. 70 percent of
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of the :f inancial resources of the first five-year plan for industrial
development of atomic energy were devoted to the new Marcoule center.
Its construction took place in the extraordinary climate that characterizes
the work of "pioneers."
This enthusiasm accomplished miracles. The center, built in less than 3
years, was dedicated 10 October 1955 by Mr. Guillaumat and Mr. Perrin.
At the beginning of the following year, on 7 January, was the divergence
of the Gl reactor, which was built in 18 months. On 25 September 1956
the first French ldlh of nuclear origin were produced.
The construction of the G2 reactor began in March 1956 along with its
large associated power plant, while construction began on the G3 and the
plutonium reprocessing plant.
The Plutonium Route (1958-1969)
Marcoule entered the production phase. During this period, plant activity
was basically devoted to the production of the nuclear material required
for the first generation of devices for the French deterent power.
.Some-'key dates:
-January 1958: Startup of the UP1, first French reprocessing plant;
--July 1958: divergence of the G2 and entrance of:the first load of
irradiated uranium into the plant;
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--20 February 1959: production of the first plutonium lingot;
--15 May 1967 and 30 October 1968: divergence of the two Celestin
reactors, designed to supply the tritium used for thermonuclear
arms.
During this period, facilities for waste treatment, storage of fission
products (1960) and the bitumen coating shop (1966) were created at the same
time.
Starting in 1969, after 12 years of uninterrupted expansion, Narcoule
went through a difficult period: needs for "strategic" plutonium were
falling off and the reprocessing capacity exceeded needs, due to the
delays of the nuclear electricity program.
Therefore, a diversification of the firm's activity was undertaken:
reprocessing of fuels from European research reactors, production of
radioactive elements, miscellaneous services.
The only notable success during that time when France was rethinking its
nuclear future: the construction of the Phenix.
This reactor, the industrial prototype of the "fast neutron" type diverged
on 31 August 1973 and produced its first billion kWh on 23 October 1974.
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Marcoule Today
The rebirth of Marcoule can be dated from 1975. That year, the French
nuclear apparatus was simplified and organized. On 19 January 1976,
COGEMA_[General Company for Atomic Materials] was born, since the CEA
had obtained the agreement of the government to "give to its industrial
means in the fuel cycle the form of a full-fledged company." On 1 June
of the same year, Marcoule became an establishment of the reprocessing
branch of COGEMA.
Despite the inevitable problems raised by such a change, the first signs
of relaunching were quickly perceptible: investments, activity programs,
hiring of a young and qualified personnel (2,579 workers on 31 December
1979 compared to 2,096--CEA plus COGEMA--3 years earlier). During this
period, COGEMA alone recruited 664 persons.
In order to live in the present, Marcoule is creating a new image, of
which the principle traits are the following:
The Gl, G2 and G3 reactors responded to the desire to associate with a
considerable plutonium production--the first purpose of Marcoule--a
significant production of electricity.
At the present time only G3 remains in operation, with Gl having been
shut down in October 1968 for economic reasons and G2 in February 1980,
the twentieth anniversary of its going on line. The perfect regularity
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of operation that always characterized this reactor gave it the world
record for continuous operation.
Its load factor was an average of 82.5 percent for its first 20 years.
During that period, G3 produced 5,269 billion kWh. In normal operation,
G3 and its twin brother G2 produced about 600 million kWh annually, or
the equivalent of 140,000 TOE (tons oil equivalent).
The industrial irradiation reactors Celestin 1 and Celestin 2 were built
in 1967 and 1968 to produce tritium, the heavy hydrogen isotope used for
the requirements of the nation's thermonuclear armament and designed to
serve other irradiations for the production of radioactive elements and
transuranians.
With a power of about 200 MWe each, they are cooled and moderated with
.heavy water and supplied with either enriched uranium or with plutonium.
These two reactors-exhibit a high degree of safety in operation and
their flexibility in adapting to extremely diverse production for industrial,
medical and pharmaceutical use (cobalt 60, Pu 238 for cardiac stimulators,
specifically) has been remarkable. Since 1976, these irradiation tools
have received a plutonium breeding vocation whose purposes are identical
to those of the G reactors.
The breeder reactor Phenix, installed on the north end of the Marcoi4le
site, is a plant related to the CEA-Rhone valley firm and run by a joint
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CEA-EDF team. Placed in operation in 1974, Phenix is the intermediate
step between the experimental reactor Rapsodie and the 1200 MWe commercial
prototype Super-Phenix being constructed at Creys-Malville (Isere) which
is supposed to diverge at the end of 1983.
- Reprocessing
When France chose the "plutonium route" in 1952 and decided to make
Marcoule a complete center for the production of plutonium, the plan was
to associate a product extraction plant with the primary reactors. This
last extraction link, what is now called the "reprocessing" link, is
today composed of a unit made up of storage pools, cladding removal
facilities and the reprocessing plant itself.
There are four storage pools where the short-lived radioactivity of the
fission products, which are particularly radioactive, is allowed to
decay before the operation:of reprocessing irradiated fuels from reactors
begins.
Fuel from the G reactors is stored from 5 to 6 months. [Fuel] elements
from the Celestin reactors is stored over 9 months, and the fuels from
the EDF [French Electricity Company] reactors of the natural uranium
graphite gas type (UNGG) is stored over a year.
Cladding removal is the operation which consists of removing the cladding
from the irradiated uranium bars prior to their entry into the plant
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itself. Considerable modifications, which are aimed at the improvement
of working conditions and process reliability, have been made in the
operation of this shop.
In 1983, MAR 400 will relay the current facility for the processing of
fuels from the EDF reactors, integrating the experience acquired in the
operation of similar facilities at La Hague and Marcoule, specifically
in three areas:
--dry unloading of casks in a shielded cell with robotization of
the cask decontamination operations;
--the creation of two storage pools;
--doing the mechanical processing of the fuel cartridges in shielded
cells using remote control.
The reprocessing of irradiated fuels, the purpose of which is to separate
materials that can be recycled from the radioactive wastes, is done at
Marcoule in plant UP1 which uses the Purex process. This process,
developed during the first world war by the Americans, is today the most
widely used in the world. It is characterized by a succession of separation
in an aqueous medium of the products placed in solution and uses the
property of certain solvents to extract uranium and plutonium selectively.
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At first, the plant received exclusively the fuel bars from the G
reactors. Year after year, the facilities have been improved. Today
UP1 has carried out the first phase of its evolution which now makes it
possible for it to reprocess a portion of the UNGG fuels from the EDF
reactors and, in the near future, all of the fuels from this reactor
type.
The fission products from fuels reprocessed at Marcoule represent about
one percent of the weight of the irradiated fuel and almost all of its
radioactivity, or about 99 percent.
Their storage was done at the beginning, and continues to be done, in
specially designed vessels built for this purpose. Their inspection is
very stringent and numerous safety measures guard against any defect.
But that can only be a temporary solution for a few decades. The long-
term safety of the storage (as well of the handling and transportation)
require the solidification of these products, which are in the form of
concentrated acid solutions, after reprocessing. Glass has been selected
In fact, the glass selected, a borosilicate, lends itself particularly well to
incorporation of all of the oxides of fission products, or about 40
elements. Vitrification also has the advantage of reducing the waste
volume, which is variable depending on the type of fuel involved.. Finally,
this glass, because of its very low lixiviation [leeching] rate, which
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reflects its very low solubility, is the ideal material to eliminate
practically any risk of contamination of the environment.
The nominal production capacity of the AVM, which is about, 200 containers
per year, was determined to satisfy the regular needs of the plant and
also to absorb, over several years, the liquid stock of fission products
that has been accumulated over a period of about 20 years.
The storage of the containers filled with radioactive glass, which are
not contaminants, but rather extremely irradiating, takes place in 220
wells 10 m high, each capable of holding 10 containers. These wells are
constructed in a concreted envelope, in an enclosure contiguous to the
AVM. The total capacity of the storage hall is 330 m3 of glass. It
covers the operating needs of 11 1000 MWe PWR reactor units for 10
From its inauguration in 1978 until May 1981, the AVM vitrified 380 m3
of concentrated solution of fission products representing the equivalent
of 7,300 tons of fuel. In all, 172 tons of radioactive glass, packaged
in 506 containers have been produced.
Thanks to the demonstration made by the proper industrial operation of
the AVM, the predominant position of France is the basis of considerable
commercial fallout. To date, contracts involving the, continuous?vitrification
process have been signed with FRG, Belgium and the United Kingdom.
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The Marcoule reprocessing plant is completed by laboratories whose basic
purpose is to monitor the production of the plant and by the industrial
chemical engineering department, created to assist the production services
in the study and development of processes and new techniques.
The cultural and athletic clubs in Marcoule offer numerous activities
and are open to the local population since 40 percent of their 4,500
members are from outside the [Marcoule] firm. The plant also has relations
with schools and the university (700 visits and 50 annual apprenticeships,
payment of a FR 400,000 apprenticeship tax, lectures at the facilities,
participation in the university-industry association) and the socio-
professional organizations (the Nimes Chamber of Commerce and Industry,
the Gard Committee for Economic Expansion and Productivity).
Relations are also excellent with the wine-producing community. This
renowned vineyard, with the appellation "Cotes du Rhone" (wines from
Chusclan, Orsan, Tresques) is the oldest in France, With 890 growers
and 6920 hectares planted with vines, the canton of Bagnols-sur-Ceze is
in the second rank in the departmental inventory, after that of Vauvert.
Both from the production standpoint and from the standpoint of production
surface, being near Marcoule has never brought any harm to this famous
vineyard.
In sum, the average financial flow induced by the Marcoule economic
activity is on the order of Fr 350 million per year (1980 economic
conditions), almost all of which benefits the immediate vicinity.
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Marcoule Tomorrow
Marcoule believes in the future because it has been able to forge its
own future. An early worker in the great French nuclear thrust, the
firm could have rested on its laurels. In 1976, with the appearance of
COGEMA, it decided to wake up, to get back into the mainstream.
This effort that has been made already assures that it will continue on to
the'end of the century. While the reprocessing activity is still half
civilian and half military, construction to increase the capacity of the
plant will allow it to relay La Hague in treating all of the fuels from
the EDF UNGG reactors, until its extinction around 1995.
All of the facilities under construction or planned-bear witness to
Marcoule's will to live.
--MAR 400 (storage and cladding, removal of fuels).
--New continuous oxalate shop (end of the reprocessing process).
--New station for the treatment of liquid effluents.
--New water network and electrical supply.
--Renovation of the laboratories assuring the on-line monitoring of
the reprocessing plant.
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P. 1 G2 and G3 under construction. While the G1 reactor was designed
to allow quick construction, the G2 and G3 reactors are the fruit
of the experience acquired on this first prototype.
P.. 2 The G2 and G3 reactors. Each building could house three Arcs de
Triomphe side by side. G3 is the only one still in operation.
P. 3 The industrial irradiation reactors Gelestin 1 and 2. They
manufacture tritium as well as products for medical and pharma-
ceutical use.
p. 4 The breeder reactor Phenix
(top)
. p. 4 Handling of a cask for the transport of gas-graphite fuels
(lower . after their use in the reactor.
left)
p. 4 The storage pool. The fuel cartridges are stored here at least
(lower
right)
5 months in order for their radioactivity to decay.
p. 5 The UP1 reprocessing plant where the separation of uranium, plutonium
(top) and fission products is done. It processes irradiated fuels from
gas-graphite power plants.
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p. 5
The cladding removal facility. The irradiated fuels have their
(bottom) magnesium cladding removed with a hydraulic Jack.
p. 6 UP1 interior view: reprocessing control board
(top)
p. 6 Exterior view of the fission products vitrification shop (AVM).
(middle)
This is where the fission products are incorporated in glass.
p. 6 Glass pouring during inactive tests done in the pilot shop.
(bottom)
p. 7
In this storage hall containers containing vitrified fission products
are stored. (Here, an exhibit container).
P. 8 MAR 400 during construction.
p. 9 MARCOULE: REGIONAL CROSSROADS Marcoule is at the crossroads of the
Rhone-Alpes, Languedoc and Provence regions.
P. 3 (top)
G3 Reactor
Reactor Building Dimensions: Height 50 m
Length 75 m
Width 45 m
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Fuel: Load: 130 tons natural uranium in magnesium-clad bars. Diameter 3.1 cm,
length 28.2 cm.
Moderator: 1,200 tons of graphite bars
Cooling: Carbonic gas, pressure 15 kg/cm2
Input temperature: 140?
Output temperature: 320? to 365?
P. 3 (bottom)
Celestin 1 and 2: Industrial irradiation reactors. They produce tritium
and plutonium
Xel: Enriched uranium or plutonium
Moderator: Heavy water in closed circuit.
Rate 9,000 m3/hr
Cooling: Heavy water
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p. 4 (left)
Phenix Breeder Reactor, begun in 1968, placed in service in 1974
Power: 250 MWe
Fuel: Mixed oxide of uranium and plutonium U02-Pu02
Moderator:
Cooling:
None
Liquid sodium
Input temperature:
400?
output temperature:
560?
p. 4 (right)
The principle: the breeder reactor uses plutonium 239 as the fissile
material and uranium 238 as the fertile material. Once the reaction has
started, the uranium 238 placed around the core is transformed into plutonium
239 by capture of a neutron.
Contribution: Its provides a solution for energy supply. It uses two
by-products of thermal reactors: the depleted uranium discarded by the
enrichment plants or by the reactors and the plutonium. It ups the energy
efficiency of natural uranium, which it alone consumes completely, by a factor
of 50. Once the reaction has started, it allows production of more plutonium
than it consumes.
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p. 5
GAS-GRAPHITE POWER PLANTS
UP1: reprocessing plant for irradiated fuels from gas-graphite power plants.
St-Laurent-des-Eaux I: 480 MWe
St-Laurent-des-Eaux II: 515 Me
Chinon III: 480 MWe
Bugey I: 540 MWe
Marcoule G3: 38 MWe
p. 6
AVM: Fission product vitrification shop
Vitrification capacity: 200 containers per year
Storage capacity: 380 m3 of glass in 220 wells 10 meters high
From 1978 to May 1981: 172 tons of radioactive glass have been produced.
9969
CSO: 8119/0983
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La ((mystique de Marcoule> (1952-1958).
'acte de naissance de Marcoule se situe clans
- ,. le courant de I'hiver 1952-1953, lorsque le
Commissariat a I'Energie Atomique choisit pour
implanter le premier centre de production de
plutonium fran4ais, une terrasse caillouteuse de 15 a
20 metres de niveau, longee a I'est par le Rhone. Le
site - domine par la Dent de Marcoule, colline
abrupte de 220 metres - se trouve clans le canton
de Bagnols-sur-Ceze, au nord-est du departement
du Gard, au point de jonction des regions
Languedoc, Provence et Rhone-Alpes. Les raisons
geographiques et hydrauliques qui l'ont fait elire,
I'avaient deja fait choisir clans I'antiquite gallo-
romaine pour ('installation d'.un vaste complexe de
fabrication et d'expedition d'amphores de vin de la
region. ,
En 1952, le CEA avait sept ans. Sept
annees employees a rattraper.le retard pris par la
France Bans le domaine de f'energie nucleaire durant
la guerre.70%des ressources
financieres du premier
plan quinquennal de
developpement industriel
de I'energie atomique
sont affectees au nouveau
centre de Marcoule.
Sa construction se
deroule clans le climat
extraordinaire propre aux
travaux de o pionniers )>.
Cet enthousiasme
fait des miracles. Le
centre, construit en moins
de trois ans, est inaugure
le 10 octobre 1955 par
MM. Guillaumat et Perrin.
Au debut de I'annee
suivante, le 7 janvier,
divergence du reacteur
G1, dont la construction
s'est faite en 18 mois. Le
25 septembre 1956 sons
produits a Marcoule les
premiers kWh fran4ais
d'origine nucleaire.
En mars 1956
commence 1'edification
du reacteur G2, avec sa
grande centrale associee,
tandis clue sont ouverts
les chantiers de G3 et
de I'usine de retraitement
de plutonium.
A partirdel969, apres
douze annees d'expansion
ininterrompue, Marcoule
traverse une periode
difficile: les besoins en
plutonium o strategique>
sont en baisse et la
capacite de retraitement
est excedentaire du fait
du retard pris par
le programme electro-
nucleaire.
Aussi, une
diversification de I'activite
de I'etablissement est-elle
entreprise: retraitement
des combustibles des
reacteurs de recherche
europeens, production de
radio-elements,
prestations de services
divers.
Seule reussite
r '"'~?~ notable, clans ce temps ou
la France repense son
l
G2 et C_a t:n cur s de c: r!5'rrrciio !. Akr
r~'cicterr Ci ,a t. iC c.on~ it l.L !n~.r il!i'! : ci pl'rni.. ,
realisation trey !apide, les reacteurs G2 et G3 son;
fruit de ('e;i.erience aceulse surce premier pr-_?'0',?;It2
eaire. a
avenir nuc
construction de Phenix.
Ce reacteur,
prototype industriel de
la filiere ((neutrons
rapides> diverge le 31
aout 1973 et produit son
premier milliard de kWh,
le 23 octobre 1974.
La voie du Plutonium (1958-1969).
Marcoule entre clans la phase de production.
Pendant cette periode, l'usine est essentiellement
axee sur la production de la matiere nucleaire
necessaire a la premiere generation d'engins de la
force de dissuasion fran4aise.
Quelques dates cle :
- janvier 1958: demarrage de UP1, premiere usine
fran4aise de retraitement;
- juillet 1958: divergence de G2 et entree clans
l'usine de la premiere charge d'uranium.irradie;
- 20 fevrier 1959: production du premier lingot de
plutonium;
- 15 mai 1967 et 30 octobre 1968: divergence des
deux reacteurs Celestin, destines a fournir le
tritium utilise pour I'armement thermonucleair-e,
Parallelement, sons mises en place, au tours
de cette periode, les installations de traitement des
dechets, le stockage des produits de fission (1960)
et ('atelier d'enrobage de bitume (1966).
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e redemarrage de Marcoule peut etre date de 1975. Cette annee-IA, le dispositif nucleaire
fran4ais est simplife et ordonne. Le 19 janvier 1976, la Cogema voit le jour, le CEA avant
_ obtenu I'accord du gouvernement pour > et
decide de faire de
Marcoule un centre de
production de
plutonium complet, le plan prevoyait d'associer aux
reacteurs primaires une usine d'extraction du
produit. Ce dernier maillon d'extraction, disons
maintenant < de retraitement>, est aujourd'hui
constitue d'un ensemble forme de piscines de
stockage, d'installations de degainage et de l'usine
de retraitement proprement dite.
II existe quatre piscines de stockage ou, avant
que ne commencent les operations de retraitement
des combustibles irradies en provenance des
reacteurs, on laisse decroitre la radioactivate des
produits de fission a vie courte, particulierement
actifs.
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Le sejour est de 5 a 6 mois pour le
combustible des reacteurs G, de plus de 9 mois
pour les elements provenant des reacteurs Celestin,
et de plus d'un an pour les combustibles des
reacteurs EDF de la filiere uranium naturel graphite
gaz (UNGG).
Le degainage est ('operation qui consiste a
debarrasser de leur gaine les barreaux d'uranium
irradies avant leur entree dans I'usine proprement
dite. D'importantes modifications - qui vont daris
le sens d'une amelioration des conditions de travail
et de la fiabilite du procede - ont ete apportees au
fonctionnement de cet atelier.
En 1983, MAR 400 prendra le relais de
('installation actuelle pour le traitement des
combustibles des reacteurs EDF, en integrant
I'acquis de ('experience d'exploitation d'installations
similaires a La Hague et a Marcoute, notamment
clans trois domaines:
- le dechargement a sec des chateaux clans une
cetlule blindee avec robotisation des operations
de decontamination des chateaux;
- la creation de deux piscines de stockage;
- la realisation des operations de traitement
mecanique des cartouches de combustible clans
des cellules blindees a telecommande.
Le retraitement des combustibles irradies qui
a pour but de separer les matieres recyclables des
dechets radioactifs est effectue a Marcoule clans
I'usine UP1 qui utilise le procede Purex. Ce
procede mis au point durant la premiere guerre
mondiale par les Americains est aujourd'hui le plus
repandu clans le monde. II se caracterise par une
succession de separations en milieu aqueux des
produits mis en solution et utilise la propriete de
certains solvants d'extraire preferentiellement
('uranium et le plutonium.
L'usine a recu exclusivement, clans un
premier temps, les barreaux de combustible des
reacteurs G. Annee apres annee, les installations
ont ete ameliorees. Aujourd'hui UP1 a mene a bien
la premiere phase de son evolution qui Iui permet
de retraiter des maintenant une par-tie des
combustibles UNGG des reacteurs EDF et, Bans un
proche avenir, ('ensemble des combustibles de
cette filiere.
Les produits de fission issus des combustibles
retraites a Marcoule representent environ 1 % du
poids du combustible irradie et la quasi totalite de
sa radioactivite, soit environ 99 %.
Leur stockage s'effectuait au depart - et
continue de I'etre - Bans des cuves specialement
con4ues et elaborees a cet effet. Leur surveillance
est tres severe et de nombreuses securites parent a
toute defaillance.
FICHE TECHNIQUE
Reacteur G 3
Dimension
de la nef : Hauteur 50 m
Longueur 75 m
Largeur 45 m
Combustible: Charge: 130 tonnes d'uranium
naturel en barreaux gaines de
magnesium. Diametre 3,1 cm,
longueur 28,2 cm.
Moderateur: 1200 tonnes de barres de
graphite.
Refroidissement: Gaz carbonique, pression
15 kg/cm'.
Temperature d'entree:140?.
Temperature de sortie: 320?
a 365?.
Puissance : 38 MWe.
Le I'cdCt'.",: ; d U'radlatlon Industrlelic Celestin 1 et 2.
ma is aussi des produ is a usage
medical et pharmaceutique.
FICHE TECHNIQUE
Celestin I et 2 : reacteurs d'irradiation
Industrielle.
Its produisent du tritium et du plutonium.
Puissance: 200 MWe chacun.
Combustible: Uranium enrichi ou Plutonium. .
Moderateur: Eau lourde en circuit ferm6.
Debit 9000 m'/h.
Refroidissement: Eau lourde.
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FICHE TECHNIQUE
Surgenerateur Phenix:
commence en 1968,
mis en exploitation en 1974.
Combustible: Oxyde mixte d'uranium et
de plutonium UO.-PuO:.
Refroidissement: Sodium liquide.
Temperature d'entree: 400?.
Temperature de sortie: 560?.
LA FILIERE DES SURGENERATEURS
Le principe: le surgenerateur utilise comme
matiere fissile du plutonium 239 et comme matiere
fertile de ('uranium 238.
Une fois amorcee la reaction, ('uranium 238
dispose autour du cceur, se transforme en pluto-
nium 239 par capture d'un neutron.
Son interet: it apporte une solution en matiere
d'approvisionnement energetique.
- II utilise deux sous-produits des reacteurs ther-.
miques: l'uranium appauvri rejete par les usines
d'enrichissement ou par les reacteurs, et le plu-
tonium.
- II multiplie par 50 le rendement energetique de
('uranium naturel qu'il est seul a consommer tota-
lement.
- II permet une fois la reaction amorcee de pro-
duire plus de plutonium qu'il n'en consomme.
~.? i ?iF^.ii: 1;(1 :'t.:'a.: r: .'df,l di1 u'lf':sp 'i ii E'S L.%: p sc{r ,i :~i :r i )Ur
GOfll~ ri_;1Hes gtr