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J.E.N.509Sp ISSN 0081-3397
MEMO JEN/TCR/A 04-81
HYBRID REACTORS WITH MAGNETICCONFINEMENT.PRELSMINARY ANALYSiS AND CALCU-LATIONAL MODEL.
por
Caro, R.
Mínguez, E.
Perlado,J.M.
JUNTA DE ENERGÍA NUCLEAR
CLASIFICACIÓN INIS Y DESCRIPTORES
E36; F51HYBRID REACTORSCONFINEMENTBREEDING BLANKETSBREEDINGTRITIUM RECOVERYOPTIMIZATION
Toda correspondencia en relación con este traba-jo debe dirigirse al Servicio de Documentación Bibliotecay Publicaciones, Junta de Energía Nuclear, Ciudad Uni-versitaria, Madrid-3, ESPAÑA.
Las solicitudes de ejemplares deben dirigirse aeste mismo Servicio.
Los descriptores se han seleccionado del Thesaurodel INIS para-describir las materias que contiene este in-forme con vistas a su recuperación. Para más detalles con_súltese el informe ISEA-INIS-12 (INIS: Manual de Indiza-ción) j IAEA-INIS-13 (INIS: Thesauro) publicado por el Or-ganismo Internacional de Energía Atómica.
Se autoriza la reproducción de los resúmenes ana-líticos que aparecen en esta publicación.
Este trabajo se ha recibido para su impresión en
Octubre de 1. 981
Depósito legal n° M-37466-1981 I.S.B.N. 84-500-4989-x
INDEX
Page
1. INTRODUCTION 1
2. FUSION-FISSION SYSTEM, DIFFERENT CONFIGURA-TIONS 2
3. CALCULATIONAL METHOD .. . 7
3.1. Calculationai model evaluation ........ 16
4. RESULTS AND COMMENTS 18
4.1. UO2 and UC cases ....... 20'
4.2. U3Si systems analysis 30
4.3. Homogeneous blankets 44
REFERENCES 74
— 1 —
1. INTRODUCTION
This report is a consequence of two precedent ones (1,2)
presented since 197 9 to different meetings of the Spanish Nu-
clear Society.
Begining with a general revisión, of the technical concept,
an initial amount of calculations on typical geometrical confi-
gurations is given in order to compare the results with those
found in the references, validating so the ased methods, and
obtáxning an rmportant experience on the sensitivity of different
effects in the working system.
Is our main interest the analysis of the blanket as energy
mul.tiplier and breeder of tritium and fisssionable material. The
tritium breeding is an important part through the used D-T cycle
in the fusión zone.
Although the specific fusión component chosen would become
a significative condition in the dimensions, materials and source
intensity of the system, in this work is not aborded, imposing
rather the fusión boundary condition through the neutrón sourc&
strength and the distance to the first wall.
The calculations are focussed on heterogeneous and homoge-
neous blankets. The first ones are considered with a fast neutrón
spectrum and the homogeneous with a thermal spectrum blanket. Typi-
cal ÜO2, UC and U,Si fuels are considered, and so are different
locations of the breeding and energy zones in the global project.
Different geometric dimensions have been analyzed in order to
obtain the general sensitivity on the diverse influencing perfor-
mance parameters.
The initial work is centered in U0_ and UC fuels, with 1 m
total thickness of the blanket. It seems that thicknesses of 20
or 30 cm of the fueled zone appearing in the internal positions
are good to obtain interesting hybrid
- 2 -
Another typical solution is considered? the U_Si as fuel
which permits to reduce the general dimensions and therefore the
material quantities involved in the concept..
2. FUSION-FISSION SYSTEMS. DIFFERENT CONFIGURATIQNS.
In figure 2.1, the nuclear reactions • involved in
the fusion-fission systems are shown. In that figure the emergent
reactions with the uranium cycle and the thorium cycle have been
well considered» This typical 9roup_ of reactions together with
the burnup of the fission product and actinides transmutation
are mainly the base of the technological step wich invoked these
systems.
A good enough classification could be the following:
- Augean systems- Symbiotic systems- Hybrid systeras,
whose definition should be:
Augean Systems;
Those where fission products coming from fission reactor
should be transmuted intro less dangerous ones.
Symbiotic systems;
In these systems the generated fuel is removed towards se-=
parated fission reactors where it is consumed. The fission reac-
tions in the blanket should not have a decisive importance.
Furthermore, it would not produce any important energy addi-
tíons. - '
Hybrid systems;
Every system where the fission and fusión reactions are ira-
portant from the energy balance point of view> while the breeding
capacity is decisive too.
- 3 -
Fast and Thermal
2 3 5Ü + n - (FP)
2 3 9Pti + n * (FP)
Fast Fission
110Th + n * (FP)
2 3 8 Ü + n * (FP)
Fusión
D + T ->• n + a
f 3"*" He + nD + D <
r -*-1* + o • -*•1 r
{"-»- He + n
D+D Vl*T + p *
Breedina
Fission
+ vn
+ vn
+ vn
+ vn
(D+T -*
* (D +
(D+T *
2 3 2 T h - + n ^ 2 3 3 P a ^ 2 3 3
2 3 3a +"n - 239Np
6Li + n ;t a + T
Li + n -*- a + T + n
a
3
a
U
Pu
E
CE
CE
E
+
He
= 200 MeV
t h r e s h o l d - 1 ' 4 M e V )
threshold 55 0,8 MeV)
= 20 MeV
n)
•*• a + p )
n)
E th resho ld = 2,8 MeV
FTGURE Z.JL
_ 4 -
In this classification, three basic requirements are
well defined. These are: waste destruction, breeding of fissil
fuel and energy production. The first of those types is suffi-
ciently explained, and it seems to be the final step of the
fuel cycle. Hewever, the conection of either two types of sys-
tems with the current reactors generation is considered impor-
tant. Looking at figure 2.2 it is possible to see that the
Symbiotics would only try to obtain fuel for the external fission
reactors, while in the hybrid systems that purpose is added to
the intrinsic generation of energy.
In order to show the different isotopic generation capa-
city of the two above mentioned types, figure 2.3, is included,
where it should be noted again the unsimilar energy generation M
(basic criterium).
Prom another point of wiew two different kinds of hybrid
reactors result: homogeneous and heterogeneous.
When it was possible to carry out the reaGtor anaiysis as
a global concept, the homogeneous concept was focussed. When
the system could be considered as a repetitive sequence of unit
elements, the heterogeneous one was considered. From the techno-
logical point of wiew, a simplified, conclussion should be to assi-
milate-a.' circulatiñg-fuel with. the homogeneous concept, and to
get the heterogeneous to a conveniently cladded and patterned
fuel.
According to the neutrón spectrum of the blanket, these
reactors could be classified in fastr and: thermarl-. In general, the
fast blankéts show more interesting simpler and neutrón "characte-
ristics, while the thermal ones could introduce the typical LWR
assemblies in the loading of the blanket. The general features
of the mentioned different concepts can be found in references
(.23, 24, 25, 26, 27, 28, 29, 30, 31) and the historical (32, 22).
- 5 -
SYMBIOTIC HYBRIDS
Fusión
D + T (Fusión) D + T (Fusión)
N(neutrons) 14 MeV N(14 MeV)
Blankefc
Li(n,n'a)T9
y/o Be(n,2n)
}Li(n,a)T
i232ThCn,Y)
233ü
238U(n,£±ss) (nf2n) (n,3n)
232y/o Th(nffiss) (n,2n) (nr3n)
y/o 232Th(n,Y) 233U
233
V
232
U y/ó
U(n,fiss)
SepárateFissionReactors
V
y/o
FIGURE 2.2
- 6 -
Blanket
Symbiotic
Li Th|
Be + Li + Th
Hybrid
—
-
—
-
U Li|
Th Li|
U Th + Li|
U+PU+ | Th+Li
U+Pu| ThJ33U+Li|
Blanket performances
233U/n
^0,3
0,8
-
M),6
^0,7
>l,0
0
239 ,Pu/n
_
—
1,5
-
M),6
>0,5
0
M
1,0
^1,6
11
3,4
A,8
>10
>30
FIGURE 2 . 3
- 7 -
3. CALCULATION METHOD.
The required calculation least to the connection of trans-
port codes, with the source option incorporated, and the typical
isotopic burnup codes.
The burnup module is not difficult if the constancy of the
neutrón fluxes in the given time interval is allowed.
However, the problem could arise in the transuranides li-
brary, in order to follow adequately the involved isotopic chains.
In figure 3.1 is shown the operation diagrara followed in
this work, using the code ORIGEN (13) to consider the isotopic
evolution of the different elements in three energy groups. In
this case, the spectrum considered has been a typical fast one.
The data bank. considered was the ENDF/B 'in -tiie_
versions III and IV (3), wich has been processed by the CODAC
(4, 5) code in order to obtain a multigroup library in TIMOC
format, wich will be used as transport code.
A good enough description of the fast zone of energy is
given, being its group limits those represented in table 3.1.
The minimum energy (200 eV) is good enough for the exact treatment
of case 1 shown below, but it was insufficient in case 2, where
there is a good agreement in the fuel zone of the system but an
error in. the tritium breeding zone of 20 % in the flux and fluen-
ce results. This problem has been solved analitically as will be
shown later. The weighting function considered was a typical
fission spectrum; observing, in the two cases, that the specific
omission of the distortion effects in the fission spectrum by
the 14 MeV- neutrons did not need a special consideration.
In the energy multigroup structure of the library (table
3.1) it was intented to consider in detail the resonances of the
different materials typical of these hybrid systems. Especifically,
the well known resonances of the uranium and plutonium isotopes,
- 8 -
ENDF/B - III y IV
CODAC
Multi'grotap . Ixbrary C23)
x, j, I
i ,N j , 2 ,
TIMOC - RATE
1
j
ORIGEN
-r*
<Tg .
FIGURE 3.1
- 9 -
TABLE 3.1 - MOLTIGROUP STRUCTURE OF THELIBRARY
Group
1
2
3
4
56
7
8
9
10
11,
12
13
14
15
16
17
18
19
20
21
22
23
• E inferior (ev>
.20 + 3
.50+3
.20 + 4
,50 + 4
.10 + 5
.20 + 5
.50 + 5
.80 + 5
.10 + 6
.20 + 6
.30 + 6
.40 + 6
.50 + 6
.60 + 6
.80 + 6
,10 + 7
.20 + 7
.40 + 7
.60+7
.80 + 7
.10 + 8
.14 + 8
.16 + 8
- 10 -
those of the Li-6/Li-7 with a máximum of llb centered at 0.25
MeV with AE-rO.lQ MeV in the elastic scattering cross section;the
elastic scattering resonance of =íl00b in the 50 KeV energy for
Ferand that of Ni where the: energy is from 15 KeV to 70 KeV.
The CODAC code C4, 5) has been used to process the
ENDF/B-III-IV in order to obtain the simplified data library.
A general formulation as
g- =
where W(E) = Weighting function, is used to obtain the smooth
a's with different angular distribution.
The (2,2n) reactions are added to the inélastic scattering,ne-
cessarily "-to tile fissüe material and optionaly to the other ones.
In the multigroup library so obtained it is necessary to
consider the following:
- there is not inclusión of the (n,n')ct,T reaction of the
Li , wich must be added to the radiative capture (CT ) inc
TIMOC; the CODAC code didn't include this reaction between
its processes. Its influence is not ímportant in the trans-
port calculation, and the integral flux results should be
good, but it is absolutely necessary in the later evalúa-
tion of the reaction rate by neutrón source obtained by
the TIMOC-RATE (6) as a modified and new versión of TIMOC
code (7, 8, 9) .
- in the tritium breeding zone, the use of a library with a
lower limit of 200 eV leads to a 20 % error in the inte-
gral flux calculation in the second problem considered.
Therefore, looking at the structure of the a's of that re-
gión, it is possible to conclude that an important contri-
bution to the reaction rates should be lost. A reason to
- 11 -
maintain that 23 groups library was to observe its good
results in many cases -with a hard spectrum involved
(case n° 1)- and consider good enough the analitical
evalúation shown below.
In connection with the first point, and following the fona
(figure 3.2) (10) of the microscopio cross section cr(n,n')a,T in
Li , cr(n,n')a,T was approximated with linear components/ as
a = .1129E - .335
a » 3. C-2)E +0.2
a = -.0175E + .58
a = -.0137E + .535
This consideration leads to a reaction rate expression for
the tritium breeding by Li as follows,
follows:
26
68
812
1220
.82 MeV •*•n +
M e V •>n ^.
M e V •*•11 -*-
M e V -»•11 -*-
10"3b0.38b
0.440.37b
0.44b0.37b
0.37b0.26b
< T >4,Li-7 J _ . - _ „. ,
a(E)dE
where subscript 4 should represent the zone number where Li-7
should be included. In that formulation, the flux constancy with
energy is assuméd.This hypothesis is not valia froia the theore-
tical point of wiew, but is sufficiently justified by the smooth
slope of CT , , in the more important región of energy, permittingHf n
without a lar ge error to determine /<j> (E) cr(E) dE in a direct way.
The assumed error in the group 17(2-4 MeV[, 18(4-6 MeV¡ does not
seem to be important, further considering these two groups and the
19th as the less contributors to the final result. The following
results are obtained with the integration on the adopted linear
function,
10
Elástic;
- 12 -
'Li (MAT = 1272)
258 kV From Ref. 10
Slástic
0.1
0.01
Qc=2,O3710
e.01
EC=O,55
0.1 1
Energy (MeV)
,n')a,T
(n, 2ñ) a,D
nr2n)Li5
(nfd)He
E =9.0 i {Jlrá-]
10 20
FIGURE 3.2
- 13 -
Group
17
18
19
20
21
22
23
/ (aE+b dE
.843
.522
.02
.845
.775
.7125
.6875
The second problem mentioned above, has been solved with
a 1/E shape for the flux in the energy interval less than 200 eV.
The microscopio cross section of Li Cn,a)T Cfigure 3.3) (10)
follows the 1/T/E" law, and It has been treated under this conside-
ra tion.
Beginning with the knowledge (MonteCarlo transport resolu-
tion) of the integrated flux in the 23 energy group, the induc-
tion of that result under the 200 eV is almost trivial. The 20 %
error is adopted in the second case Csee reference 11}., where
this effect is appointedrwhil¿ no problem exists in the first
systeía. The proporcionality constants are obtained in each case
by,
íE2kl<<f»= j g=- dEEl
where the higher limit of energy is E_ = 200 eV and E = 2 eV
is taken as the lower limit where the flux contribution can be
neglected.
The microscopic cross section (nra)T in that energy range
can be assumed to be of the form k~/VE (see in the graphic the—0 5logaritmic linearity, In a = A inE+B = In k_E * , k_ - 16 0.0).
Writing again the unitary reaction rate expression for the
(n,a)T reaction in the Li ,
- 14 -
10
Q=4.786[n,a)T
Elástic
0.1
0.01
Qc=7.252
10-30.01
6Li (MAT=127l)
2^7 keV
From Ref.'10
0.1 iEnergy (MeV)
Elástic
Cn,n')a,D
f2n)a,p
(nfa)
\ín,p)Ee
10 20
FIGURE 3•3
- 15 -
<T> = N4(Li-6)Z * *° -1/2 -1/2
dE = 2N .. ,k_1 ¿
El -E2
About the general scheme of calculations, the neutrón trans-
port results in the blanket are obtained through the TIMOC-RATE
CODE (6) including furthenaore to the transport calculations with
external source, the unitary rates per time and source neutrón.
The external source is suppo.sed to be centered in the system, with
14 MeV emerging neutrons (fusión reactions). The reaction rates
with are given in the model are:
- Capture and fission by U-238. Capture producing Pu-239.
- U-235 fission.
- Energy generation by fission reactions.
- Tritium breeding by Li-6 and Li-7..
The simple expresión
T h . - N * Y <J,h ax,i x s y g»x,i.
where, N. = atomic density of isotope i in región h
<j> = neutrón flux in the energy group g in región h
a . = microscopio X.S. of channel x, isotope i and^' fl energy group g,
is taken here, resulting the reaction rate T . normalized byx, ±
source neutrón and time. The actual valúes can be easily obtained
through the first wall power adoptedf which produces a certain
source intensity. This first wall power could be taken as a design
limit coming from radiation damage to the involved material. The
relation mentioned before is,
- 16 -
Q = P x — ^ x 6.24 x 1012 x ij x 4TTR2
x 6.34 x 1012 x
2where, P = fixst wall power (MW/ra )
R = first wall radius (cm)
Q = -14 MeV neutrón source (neutrons/sec)2
J = neutrón flus in the first wall (n/cia rx sg) .
The use of the XSDRN code (12) has not been extensive In
the complete results, because of some convergence difficulties
in its source mode, and it was only considered in a few cases
to test the TIMOC code with a S_, technique code.N
The last part of the scheme calculations is that of isoto—
pie evolution with time, i.e. the burnup module.
The ORIGEN code (13) was chosen as the tool to give the iso-
topics. Chosing adequately the time intervals, the flux could be -
assumed to be constant in them, therefore the neutrón fluxes by zo-
ne given by TIMOC are necessary. The extensive library and the com-
plete solution of the differential equations involved in the exis-
ting time balance are at least two important characteristics of the
program. The consideration of •fchree energy groups is certainly the
darkhouse of the model, although an effort is made iñ order to
incorpórate into the TIMOC-RATE code a general burnup module wich
could process a greater library. Some comparisons between the two
models seem to be important, reducing the inaecuracy relative to so
few number of energy groups. In order to produce a correct analysis
of the actinides transmutation and waste destruction, in general,
it should be very intersting to have a library with more energy
groups for the unusual elements which need to be considered.
3.1. Calculational model valúation.
Two different checks were adopted to stablish the working
efficieney of the calculational system. Comparisons with experi-
mental benchmark and other techniques have been undertaken. These
are,
- 17 -
1) test with a natural uranium system (.14). wiiose results areexperimental and are well proved.
2) check upon a source problem with the XSDRN code.
i) Characteristxcs of the benchmark,
. Diameter 99 cíaCylxnder
p = 0.85 p tMaterial Natural Uranium ... p = real dénsity
p^ = theorical n.
Neutrón source at the center with 14 MeV of emerging energy.
The results are given in Table 3.2, compared with experimen-
tal ones and those offered by J.D. Lee from Lawrence Livermo-
re (14). The agreement is very good.
ii) The XSDRN code has been used as the S technigue processing
code to compare with the MonteCarlo TIMOC. Its capacity to
treat source problems together with 'an extensive library in-
corpora ted, and the resonances processing makes of it the most
interesting tool, at least in this kind of problems. Difficul-
ties have appeared" with. the source problems, however some cal-
culations have been carried out to obtain some comparisons.
The used benchmark assembly was:
- Geometry: sphere divided divided in three zones with radii;R- = 16 cm,
Zone 1 :
Zone 2 :
R_ = 36 cm,sillín
cuum»
ElementFeNiNbCU-235ü-238
FeNiNb
R_ = 116 cm.
Atomic density (at/cm x b)
.00323
.00077
.00266
.193(-3)
.137(-3)
.0191
.00323
.00077
.00213
Zone 3
- Source neutrons of 14 MeV centered in the sphere.
The results on flux shape are given in figure 3.4, showing
a good agreement.
- 18 -
4. RESÜLTS AND COMMENTS.
In the previous paragraphs, the calculational model and•its
valúation has been shown. In the following the results obtained
will be given and some comments made.
The aim of the study is to show some qualitative conclusions
on the reactor design, penaltting to stablish a global coiuparison
with other different developments in the fusión or fission energy
generation. Henee, the total energy balance, where the fission
blanket acts as a multiplier, must be an important objective and
a specific parameter will be provided to qualify this aspect..
The breeding capability looks as an important one, that can jus-
tify and complement the hybrid design too? consequently the bree—
ding rate of the system will be looked at as a design parameter.
The use of the D+T eyele as fusión burners means
the maintenance of a nearly unitary rate of tritium in the system.
This undoubtfully important necessity may be considered from two
point of wiewt
i) the tritium radioactivity Cg emitter, máximum energy
18.9 KeV, period 12.4 yearsl , which limits. the global
stock of that element,
ii) the different associated problems which could arrive
to a configuration where the mentioned condition is
not obtained,
The generation^ reactions of tritium through the Li and Li
neutrón capture has been showed before.
From this general perspective, the results presented below
can be divided in two large groups:
1] Results on UO- and UC fuels with geometrical configurations
cióse to the technological indications of some hybrid design
- 19 -
E-4
£-6
2-7
S-3
. 1 ¡
1 1
i
1 > 1 •- : ! t t ! i . t . i • t . ) 1
1 ¡ ! ! 1 • .' ' : ' ! 1 I I li ! !11 í t í I ¡ t
l i !
III
1!
' ' _i i, ; ¡"'¡ J
1 i
1
I
1
i ; ¡ ! ; m i :
1 í i i
i 1 ! -1
1
1
1
l !
i í !
III
III !
' ! 1
iI
II
I! ' j 1 i i
! i 1| ¡ i
Íl
! i ! • ' i > . < 1 i
H i i
t
1 • |
1 Hi
Mi!i l
1
! ' ' ' 't - t : | | . . .
! i '> i • i . !
1 i ii
; ii-: ! ; • ; ; » • . ; - , . • •
i !
¡IIIÜil
II
1 U¡ : 1 ! ; i 1¡ ii
| i | ! i ! ':
> ! í i| | |
ü
I!
i ii
i I
¡1! ' ! ' ' ' • " / i W • i : ' W < > • • • • • • • • • • ;: i i i i i > I I 1 ' i - \ • í . » n / / X i ^ ' : i i >• . ' ' • • • •
1 ! i L - * " \ l i l i / I :
!
Ilf
f1III1
i
l
|
t . IW!: *
i
ii
: • ¡
lli• 1
•
iÜ1
1, ip / j • •• l i > • ' : i ! i i ! . ii ¡ i I i i • •
1 i!/ ' : i i ¡ i iMi'/ .
i,
íM ; Í
1 1 1
11
, ' : . i : — . : , J f i , , 1—: • i : ;
' ¡ M i l ! ' i i '. 1 1 1 1 i i ; !
1
i i 1!! i i ! ! i i •
!H
lli
i
il
i
i
! 1 it
1 1 . : • • ¡ : i l l J • i ¡ . 11 !
1 i p • . : ¡ 1 . • • 1 ! | . i ! |
1 i
!
i
t
¡MÍ : 1 í '
lili ! !
III
i !
i¡ ¡
i
ií
. s •• ; í i i : : |
ii i i
il 1 i !
=11 ¿ Ik
i 11 i \\\í
i 1
i1
V
; M¡!
/iw
' Vi— n 'i
i • i ¡ i i ; : : i : ¡ i : *. .*¡ i i: ¡ i ;
!
i
II11
i
!
! i
1 1
i ¡ • ,
H
!
i • t
i| !
i
II
i ! i i i t • i i i < i : i i
1
' ' • \
1
i ! : ¡ ! • i¡. 1 !
1 Ii
i; i
• i< i
i !
II1
! . ' • • ! • • i
• ! • i ' 1
\ • • : , ¡ . :
.
1
'i
i ;1 i
i ¡i
i i
• '. I Ü
ii!;1i
i '. i ! . •
i • i; i: ' • i > ¡
; • ; i i ii i • i ; i
¡ - i
) ! •• M I ! ! 1 ' i i i : • . i i i i • : : ¡ ; . • • • • . •• •
i i 1
! 1i
;,:zzi:",ii
, t j !¡ | i '
1 11 i
i 'H !
•
ii
t
m
1 ! '< ni ;
i¡ •¡ j
i ¡II
! ! • ü :. • ' ; !
' i '. 1 ¡ i
1 >
i ;• i
! !1: . • . ii
! 1 M
i ! ! i i [
I I !| !
I I
i
' I 1 1 : II
i11 ¡1
mumi 1
i : ¡
1 !
i ! ! ! ; ¡ . i ! 1 : : i • i ii.
: i ! : i i • i j j | ¡ ¡ ! i i i ;
¡i
!1
| ;1i
1i
IIIi 1
1
1 i|
: , - 1
l i l i ¡ ! • ' i : ! '
i! ' 1 •
1 h: ! i f
E+6
TIMOC
-- XSDRN
Zone Spectra. Fissile material (2)
FIGURE 3.4
- 20 -
(15, 16, 17, 18), but where an. original spherical treatment
of the systeiu is considered.
2) Results on U_Si fuels with other geometrical data CU) under
the use of the proposed model and a spherical consideration
as will seen below.
Each group of results nave been studied iaodifying the rela-
tion between fission and tritiura breeder zone thickness, and under2
different valúes of the real power (MW/m I in the first wall. Ex-
posure results will be offered in some cases with some extensión.
4.1. U02 and UC cases.
The geometrical configuration adopted to represent the
fusion-fission reactor of references 15, wich would allow to
valúate the possible conditions of these concepts is shown in
figure 4.1 (p. 28) where each zone have a material composition
as follows:
Zl Vacuum and central source of 14 MeV neutrons.
Z2, Z3 On a total blanket thickness of 100 cm and adopting
different geometrical relations of the zones; the mate-
rial composition of those zones have been changed to
analyze their effect on the different design parameters
mentioned above. However with a general denomination of
tritium and fissile breeder zones, the material densi—
ties are:
Element At. densities Cat./cm x b)
Fissile Breeder Zone:
depending of case
Tritium Breeder Zone:
FeNiNbLi
UCuo2
FeNiNbLi
0.003230.000770.002660.00608
0.01930.0225
0.003230.000770.002130.03268
- 21 -
with a natural lithium isotopic percentage of Li ( 92.7) and
Li ( 7.3), the atom densities of those isotopes are,
Li7 (at/cm x b) = 0.005624 ; Li6 (at/cia x b) = 0.000456
in the fissile breeder zone, and:
Li7 (at/cm x b) = 0.030229 ; LI6 At/cm x b) = 0.00245
in the tritium breeder zone.
The constancy of the material concentration in every case
supposes for different volumes adopted (Table 4.1}, the conside-
ration of unequal masses of fissionable material in each case
CTable 4.2).
The zone numbered four is a leakage one iraposed by the
MonteCarlo procedure.
Beginning with the criticality problem, comparative results
with ref.15 are presented in two different moments of the core life:
Case
Beginning of10 % burmip
lifeuranium
Keff CRef.
0.280.73
15) Keff
00
(JEN).
.309
.70
The breeding features of the system bring the larger valué
of multiplication factor in time.
Significant results are offered by the foilowing chosen pa-
rameter design:'
- Plutonium generation rate per source neutrón from U-238neutrón capture.
- U-238 fission reactions per source neutrón.
- U-23 5 " " " " "
- Tritium production per source from Li neutrón captura.
- Tritium production per source from Li neutrón capture.
- Total energy generation per source neutrón.
The tables 4.3, 4.4, 4.5, 4.6 and the figures 4.2, 4.3, 4.4
4.5, 4.6, 4.7 give the amount of those results.
- 22 -
TABLE 4.1
Thicknessn
n
Case
Zone
Zone
Zone
2
2
3
= 10
= 20
= 30
cmcía
cía
VolumeZone 2
29,56 m3
60,35 "
92,40 "
VolumeZone 3
324,47
293,68
261,63
TABLE 4.2
-Thick-;
* (có)
10
20
30
Case
Zone 2. FissileZone 3. Pissile
Zone 2. PissileZone 3. Pissile
Zone 2. FissileZone 3. Fissile
Uranium mass(UC) (TM)
225,42474,13
460,182239,35
704,561994,9
Uranium massÜO (TM)
262,772884,35
536,472610,65
821,382325,74
TABLE 3.2
vsyst«a = °'816
N U-238
NU-235 / f
Reactions by 14 MeV source neutrón, integrated in volume.
Reactions
U-238(n,
U-238(n,
U-235(n,
Leakage
Y)fiss)
fiss)
Experiment
4,08±0,24
l,18±0,06
0,281+0,017
0,42±0,02
J.D.LEE'
4,36
1,11
0,266
0,504
. J.E
3,
1,
o,
.N.
85
20
268
0,39x0,004
TABLE 4.3 CASE, ZONE 2 (INTERNAD WITH FISSILE MATERIAL. Material UC
Thickness(cm)
InternalZone
10
20
30
U-238/(nrYV
JEN
0,4568
1,13
1,64
JEN*
0,625
1,56
2,29 .
LEE
0,68
1,50
2,12
U-238/(n,fisa)
JEN
0,336
O,'533
0,6217
LEE
0,360
0,536
0,620
U-235/(n,fiss)
JEN
0,0268
0,06174
0,0868
LEE
0,036
0,074
0,10
Energy (MeV)
JEN
72,63
120,0
1*1,7
LEE
9O,O
130,0
151,0
Li-6(n,t)
JEN
0,86
O.72
0,45
JEN8"1
1,15
0,95
0,58
LEE
1,28
0,99
0,69
CASE 2, (INTERNAD WITH TRITIUM BREEDER MATERIAL
10
20
30
1,6124
1,179
0,928
2,24
1,64
1,29
1,17
1,03
0,5325
0,3955
0,3126
0,384
0,289
0,0815
O,o6o
0,047
0,065
0,047
122,8
91.1
71,9
93
75
o,4o
o,44
0,53
0,46
0,50
0,61
O,58
0,66
- Blanket thickness is 100 cm in every case.
- JEN* = Using the 23 groups library chose to BNL-325 and Yiftah references.
- JEN** = Modified library in obtaining reaction rates, in order to come asthe graphics of ENDF/B-rv.
• I
I
TABLE 4.4 - CASE, ZONE 2 (INTERNAD WITH FISSILE MATERIAL. Material; UC>2
Thickness(cm)
InternalZone
10
20
30
ü-23fl/'(ii,Y)
JEN
.6<i 6
1,363
1,735
JEN*
.eo'i1,672,39
LEE
0,50
1,20
1,60
U-23Ü/(n,fiss)
JEN
.3679
.51'*
.570
LEE
O,2fl3
0,427
O,ítfl7
U-235/(n,flsa)
JEN
.03*16
.0674
.0020
LEE
.026
.O57
.077
Energy (MeV)
JEN
«O,5
116,3
130,5
LEE
7'i,*!
107,0
122,0
6Li(n,T)
JEN
0, OO
0,50
0,30
JENHM
1,05
.05
LEE
1,14
.901
.621
CASE, ZONE 2 (INTERNAD WITH TRITIUM BREEDER MATERIAL
10
20
30
1,99
l,/i91
1,179
1, 51,00o,06
-
.*I5O
• 33
.265 -
0,0657
.051
.odo -
103,5
76,2
61,0 -
0,37
O,k'k
0,50
-
I
I
- Blanket thickness is 100 era in every case.
- JEN* = Using the 23 groups library chose to BNL-325 and Yiftah references.
JEN** = Modified library in obtaining reaction rates, in order to come asthe graphics of ENDF/B.
— 25 --
= 4,8 m.
= 5,8 m.
Blanket = 1 0 Q
FIGURE 4.1
TABLE 4.5 - INTERNAL ZONE = FISSILE MATERIAL CüCl
Intemal zonethlckness Ccm)
10 .2030
. LJüCíl-Il' Ct-rtL
JEN
.336
.149
.064
LEE
.212
.095
.055
Bred Tritíum
JEN
1,196.869.514
JEN**
1,4861,099.64
LEE
1,4921,085.74
EXTERNAL ZONE = FISSILE MATERIAL (UC)
102030
.146
.272
.390.271. 326
.546
.712
.92
.606
.7721,00
.85
.986
TABLE 4.6 - EXTERNAL ZONE FISSILE MATERIAL (D02)
Intemal zonethickness (cm)
10-2030
Lüln,!!1 ,a>tl
JEN
• 354.070.022
LEE
.212
.094
.055
Bred Tritiiam
JEN
1,154• 57.322
JEN**
1,404• 92.472
LEE
1,352• 995.676
EXTERNAL ZONE = FISSILE MATERIAL (UO2)
102030
0,2700,137'0,382
1,070,64.68
1,32• 99.83
-
- 2S. -
3.0
2.5
2.0
1.5
1.0
0.5
0.0
Generated Pu (at),U-238(n,T)
10 20 30
FIGURE 4,2
40Thickness (cm)
5 0 Internal Zone
140.
130.
120.
110.
100.
90.
80.
70.
Energy (MeV)
10 20
Material; UC
o - Fissile Material in theinternal Zone.
A - Fissile Material in theexternal Zone
Results are g.iven byneutrón source
30 40 50 Thickness (cm)Internal Zone
FIGURE 4.3
-,27- ~
Generated Tritiinn Cat)
1.5
1.0
0.5
Material UC
50 Thickness (cía)Internal Zone
•pTGURE 4 . 4
2.5
2.0
1.5
1.0
0.5
0.0
Generated Fu Catl rU-238 GI ,
Mater ia l UO.
10 20 30
FIGURE 4.5
40 50 Thicícness (cía]I n t e r n a l Zone
- .28 -
140.1.
130.
120.
110.
100.
90.
80.
70.
60.
Energy^
10
• Material UO^ •
Fisslle Material in theinternal Zone
Fissile- Material in theexternal Zone
Results are gxvea by neutrónsource.
20 30 40
FIGURE 4.6
50 Thickness (.cía)ínternal Zone
1 .5
1.0
0 .5
Generated Tritium Cat[
10 20 30 40
FIGURE 4 .7
50 TítxcknessInternal Zone
- 29 -
In addition , flux results (figures 4.28 to 4.38) have
been given from the ÜC case because they can be adopted as re-
presentatives of the general evolution in both fuel cases consi-
dered here.
In respect to the employed method, two comments must be
made:
- a good general agreement was obtained with the results
from the references with other geometrical and calcula-
tional model.
- some little discrepancies in the Cn,t) reaction of Li
can be observed. Henee it was proved that a minor change
in the original 23 group library in that reaction permitted
to obtain good results.
- some results on the U-238 capture have been given through
the use of some valúes in specific energy groups from
refs 19 and 20 in substitution of those from CODAC code,
It is possible to see a difference by that mentioned effect
which brings in question the sensitivity of the systenu
Some comments need to be made on the general valuation -
given here:
i) in both UC and UO_ cases, an inner location of.'the
fissile material permit to obtain better results. Somewhat
less tritium breeding capacity and energy generation are
got when the tritium zone is internally placed
ii) in the UC case, an -20 cm inner thickness of the fissile
material zone could be an adequate proposition,which leads
to a generally good reproduction and energy generation.
When the UO^ is considered, the thickness must be reduced
in order to permit a good tritium breeding wich could jus-
tify the concept. Observe how with internal 20 cm thickness,
- 30 -
a tritium reproduction rate of =0.85 can be obtained with
at least an -1.3 fissile generation rate.
iii) the results worth is quite valid to conclude the optimal
valuation of this reactor concept as presented in the ge-
neral introduction. Through the amount of numerical re-
sults, it is possible to deraostrate the optimal energy
generation besides a fissile breeding levéis as well higher
than in the typical puré fission breeder reactor; in the
meanwhile a tritium breeding up to the unity can be obtained.
An important characteristic is the subcriticality condition
of these systems.
Some . results with exposure variation have been pro-
cessed with these two initial configurations. A more detailed
analysis is offered with the U_Si concept. The possibility to
obtain interesting fluxes to allow transuranides burnup and to
arrive to high levéis of exposure keeping the required explota-
tion conditions -forces to evalúate the time evolution of the
hybrid reactor.
In this case, results to the 2 years of exposure level are
presented in order to see the correct agreement with other calcu-
lations. The ORIGEN code was used to made these calculations,
considering a neutrón fusión source intensity of 1.2 x 10 n/sfwhich means an average flux in the fissile material zona (20 cm)
14 2of 5.1 x 10 n/cm . s. In the following, the indicated results
from a continuous exposure of 2 years are given in at/cm x b:
U-235 U-236 U-238 U-23 9Expo sure —(years)- JEN Ref JEN Ref JEN Ref JEN Ref
2 0.0126 0.0127 .00023 .0002 1.89 1.89 0.0170 0.0187
4.2. ü,Si system analysis.
The analysis of results about the U_Si fuel concept,
belonging to the Lawrence Livermore Lab. basic proposal 11, are
reported in the following. A mirror reactor is taken as the fu-
- 31 -
sion part of the concept.
Although the general view of the obtained performances Is
almost similar to the U02 and UC cases, it will be interesting
to note in this study a more extensive burnup analysis which has
been developped. Parametric results of the design qualitative
indicators have been obtained, in varying the relation of tiiick-
ness zones with a constant blanket width.
The purpose now will be to describe the geometric and mate-
rial configuration of the system. The proposed hybrid reactor of
ref. 11 will be adapted in our case to a sphere in order to re-
present adequately the core. That means that a general valuation
of the reactor is considered, but not an individual element. The
inner cavity is supposed to confine the plasma material with a
radius of 250 cm.
The first wall of the system is placed at 350 cm radius.
The neutrón fusión sources is centered in the system, considering
a vacuum sphere of 350 cm radius.
Following the adopted model, a top zone of cooling (He)
flow appears in the reactor, wich represent the individual current
of each of the elements composing the system, where the cooling
is flowing in backward before the heat removal. The zone thick-
ness on the mentioned región is 25 cm, with helium as coolant.
The blanket is considered next. As before,two subzones are
viewed heres fissile breeding energy generation, and tritium
breeding. The total thickness of the blanJcet was 92 cm which has
been considered constant in all the analysis. Nevertheless, the
subzonal relation in the blanket has been varied in order to
perform the mentioned parametric study. In figure 4.8. the geome-
trical data are gíven. . in the case of 25 cm thickness of ura-
nium región. The positional relative arrangement of the regions
is not changed in this problem in opposition to that considered
in the UO_ and ÜC valuations? that means that uranium is alwaysin the internal subregion of the blanket while lithium is in
the external one.
- 32 -
euo
II
37
5
sU
Oo-3-
GO
vO
ai oí
oosoHOS
SOw
!co
asD
H
- 33 -
The atomic densities by zone are:
Zone 1 . - Vacuum, introducing the neutrón fusión source.
Zone 2 ,- Described before and formed by the heat removal ele-
ment He.
Densí.ty,p = 0.00518 g/cm„ , j- . . , -. N = .00078 at/cm x bVolumen fraction =1.0
235Zone 3 .- Considering the fuel and breeding material = U ,
U , Si The structural elements have noy been taken.
The U-Si fraction in the individual assembly is 0.6 92
and the atomic densities are:
235N(U ) = 6.78852C-5) at/cm x b
N(U238) = 0.027071
NCSi14) = 0.009
Zone 4 .- Tritium breeding zone, where natural lithium is allowed
with the percentage of Li - Li given above. The support
materials of the assembly are taken into account. With a
LiH density of 0.82 g/cm , the atomic densities are:
N(Li^) = .00336 at/cm x b
NCLi7) = .04117
NCAl) = .00574
N(Pe) = .00898
N(H) = .0445
In table 4.7 and figures 4.9 and 4.10 are represented the
valúes of the parameter M C= MeV/14 MeV x neutrón source) for
different conditions. In viewing those results, some aspects must
be remarked:
- in a general look, the high valué of M for the thickness and
- 34 -
TABLE 4.7 - M PERFORMANCES VERSUS BURNÜP
uranium
(cm)
20
25
30
41
E (MW/year/m )
0
9.84
10.87
11.95
12.39
2
12.12
14.06
15.71
17.15
4
*
15.02
19.40
22.58
26.47
6
19.19
27.10
36.80
42.90
8
24.71
44,34
69.91
140.2
M =MeV
14 MeV x neutrón source
M
1O.
Energy inultiplication factor, M, veraus uranium thicknesa
MoVM = , iiíLí .
1*1 MoV x neutrón aource
I
FIGURE
L'O 35R
fio
uranium (cm)
- 36 -
- 37 -
exposure spectrum of the study. Observe that M reaches to a
valué of 10 when the fissile zone width is 20 cm
- a máximum is noted in the variation of M versus thickness.
That máximum valué (M=13) appears into the 35 cm width.
- for any fissile zone dimensión, gradually higher valúes of
M will be reached when increasing is given. This aspect
could be seen as a guarantee to have a positive energy ba-
lance with time.
The fissile material generation, whose identification para-
meter is represented by net.at.Pu/neutron source, has been pietu—
red in the table 4.8 and figures 4.11f 4.12. An almost linear in-
crease of the fissile parameter is noted when the fissile dimen-
sions varies in the same increasing sense.
Another characteristic must be emphasised. A constaney in
the fissile parameter versus exposure is observed when a little
thickness is taken, while an increase tendeney appears with
higher width of the fissile zone.
Table 4.9 and figures 4.13 and 4.14 show the tritium
breeding gain (at.tritium/source neutrón}. The necessity to go
to valúes equal of higher than one for this parameter indicates
that valúes of at least =70 cm for the lithium región must be
allowed, which is equivalent to a máximum of =30 cm as fissile
thickness with a blanket total dimensión of =100 cm. In figure
4.14, the net tritium production evolution with exposure is repre-
sented in the specific case of a 25 cm as fissile región width. A
positive increase tendeney is remarked, but some coraments need to
be made from the quantitative point of wiew. The hypothesis ta-
ken in the calculation model, supossed a 20 % of total flux
above 200 eV. This aspect was aborded by an analytical procedure
but it seems this flux percentage could be changed with the expo-
sure in modifying the obtained tritium performance. Confirming
this point, a hardening of the spectrum appears with the exposu-
re in the lithium región.
- 38 -
TABLE 4.8 - Net. at. Pu .Neutrón source > v e r s u s b u r n u P
uraiíium
(cm)
20
25
30
41
E CMW/year/m2} .. ;
Capt U-238
Fis Pu-239
Capt Pu-239
Pu neto
Capt U-238
Fis Pu-239
Capt Pu-239
Pu neto
Capt U-238
Fis Pu-239,
Capt Pu-239
Pu neto
Capt U-238
Fis Pu-239
Capt Pu-239
Pu neto•
0
1.2667
1.2667
1.667
1.667
2.04 .
2.04
2.56
2.56
2
1.36205
.1103
.0332
1.21855
1.8858
.1513
.0457
1.680
2.3762
.1900
.0632
2.123
2.998
.209
.0700
2.72
4
1.553
.263
.075
1.21704
2.228
.397
.119
1.712
2.877
.512
.167
2.198
3.911
.612
.200
3.00
6
1.7362
.47
.128
1.14
2.759
.785
.237
1.737
4.038
1.177
.356
2.5
5.3756
1.4 27
.487
3.46
8
2.035
.761
.22
1.054
2.945
1.6225
.58
1.7425
6.366
2.785
.840
2.7
14.8
6.0
5'.0
3.8
- 39 -
EO
tn3
T3(3U
0)c0N
0)
SO
•H
• i -
304
•
•
0)s
oS-t30
c0
-u30)C
\
\\\
\
\
\
(N \S \
\ \
i \o \
\II \
f - iT-i
*
«
DH
\\\\
\
\
ir»
om
m
CN
- 40 -
SumCN
üo
aasD
Ocu0}
0)
3c
- 41 -
TABLE 4.9 - Tritium production , b u r n u pNeutrón source r
uranium
(cm)
20
25
30
41
2E (MW/year/m )
Li-6 •
Li-7
Total T.
Li-6
. Li-7
Total T.
Li-6
Li-7
Total T.
Li-6
Li-7
Total T.
0
1.0752
.1259
1.2012
1.0084 '
.1416
1.1500
.9249
.0551
0.97
.0177
2
1.1581
.1259
1.2864
1.17495
.07897
1.2539
1.0551
.05626
1.1114
.01678
4
1.3243
.1322
1.456
1.4709
.0925
1.5634
1.3209
.0621
1.383
.0276
6
1.5413
.1418
1.683
1.843 .
.0942
1.937
1.9789
.0658
2.045
.0202
8
1.7880
.1420
1.930
2.7298
.1144
2.844
3.523
.0812
3.605
- 42 -
s
tnm
MV
oN
0)
tn
S
o
II
02
63
D
ct - iT30)<US-t
S3
• H
0)U5-130
c0
CN
O
CN
4.
3.
2.
1,
i
Tritium breeding .Neutrón source
-
•" 2
E^posure
4
2
(Mwy/m J
e2
FIGURE 4,14
i
6
r 25 era y
1
//
¿8
/
10
I
- 44 -
The flux shaoes of the fissile zone are shown in figures2
4.15 to 4.19, in neutrons per cm and second, and following the
multigroup structure adopted. The significantly fast nature,of
the spectrum is overtaken, noting the singular valué of the fu-
sión neutrons at 14 MeV. Is should be noted hov an almost verti-
cal translation appears in the flux shape when different temporal
moments are taken into account.
In figure 4.22, the integrated fluxes of the fissile región
versus the exposure are represented parametrically with the thick-
ness valúes; a progressive increase is remarked. Also the figure
4.20 and 4.21 are pictures of flux shapes, nevertheless in these
figures the lithium zone is considered.
Finally, the isotopic evolution with time is attempted in com-
parison with other reactor designs whose objectives could be to
obtain fissile breeding. Table 4,10 gives the isotopic densities
versus exposure referred to uranium and plutonium elements. In
table 4.11 the percentages of the plutonium isotopes are provided
and through them a nearly constant relation with time is regarded;
further a high Pu-239 enrichment is given in any case. In order
to compare the breeding capacity of these reactors concepts with
other closer to the comercial state at present,table 4.12 is
shown.
In that table the global amount of plutonium CMT/GWe) yielded
by the hybrid reactor model just considered is compared with that
of the actual fast breeder fission reactors. From that table is
possible to induce the importance and high performances in hybrid
reactors even higíier to the proposed fission breeder ones.
4.3. Homogeneous reactors.
One of the analized options to yield a total burnup in the
system is that of homogeneous hybrid reactors, where the material
could be kept into the system for a long period of time, leading
to a better utilization of fuel and permitting an important des-
- 45 -
o
ea
CNe\
Mw
y
co
CN
e\3Svo
CNg\3S"v
CN
s
1CN
n-rrd•H4J•Ha
aj.'-in
0- CN4J £•
0 0w \H- C
U /
/ "
**••»,
in
inTH
+•
r-cn*
T-I
IDt
» - *
rot-i
T H
1—|
+IDincn•cn
cn-e-
m CN1 1
CN r»vo vn
•q< C N
ID CN1 1
0 CNT-1 VO
in CN
in CNi 1
CN in
r- vom CN
in CNl i
vo 00CN VO
VO CN
T t—11 1
00 r-co 0
VO CN• •
in cocn cnCN CNl 1D D
1
cnr-cn
l
t—1
CN
r-
1
cn00
1
cncn
CN
cn
CN13Cu
c
in1
CN00
00
inl
incn^
vo1
CNcn
co
VO1
vo(Ti
•
O*3>CN13
r-l
vocnvo
1
T—1
VO
CN
CO1
CNcn
cn1
r-roco
t-i^j<
CN13CU
T-I
-r
00
inT-I
+
T-I
T-I
inT—i
ro00
r-4
IDt-i
+incot—i
0•
t-H
ro-9-
m CNl i
0 vo
<sr CN
in C Nl l
mn0 vo
incN
in CNl l
r>. invo vo
in CN
in CN1 1
meoCN VO
VO CN
rr T-11 1
co r-co 0
VO CN• •
m coro roCN CN
1 t
-,
1
00O
T-í
1•l|«
00
r«
l
CNO
in
l
CN
cnroCN13cu
irCN
ínl
CN00
inl
0roCN
vo1
roO
cn
VO1
<v0
CN
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t—1
rocncn
1" • • "
vororo
CO1¡«CN
00
cn1
cnco
T—|
TCN13CU
m
CO
H
t-1
+t-1VO
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Ün
^co• H
T-I
intH4-
1—i
CNO•
TH
ro-0-
tn1
CN
m1
0
in
inl
r-voin
inl
CN
VO
1
0000r~vo
inroCN1CD
CN1
inin
CN
CN1
T-IVO
CN
CN1
invoCN
CN1
COvoCN
T-iI
«—•
r-0
CN•
COroCN1
cn1
inTH
TH
1
O00
r-
1
roO
in
• 9
i
ID-51'
CN
cnroCN13CU
Oro
inl
CNin
un
inl
0
CN
VO1
in0
cn
vo1
vo0
CN
O"3*CN13cu
VO1
CNCN
TH
i
CN
ro
cc-1
0ro
co
cn1
CNO
cn
r-i
CN13Cu
in
ín+
00
•
tH
ín
cn0cn
TH
14
)
VOCN[->•
rocn•
co
ro-S-
\n1
T¡ro
ini
tH
in
inl
r-r-in
m1
iH
rovo
1• w
COcoVO•
inro(N1D
CN1
IDID
CN
CN1
TH
VO
CN
CN1
invoCN
CN1
CO
voCN
tH
1• _ *
r-0r-CN»
COroCN1D
ro1
O
tH
1
00ro¡^
1
voID
1
inTH
CN
cnroCM13CU
0
m1
ro00
ID1
IDO
CN
VO1
COro
vo1
00in
T-i
O"51"
CN13Cu
vo1
0O
TH
l
rocoCN
CO1
CNtH
vo
cn1
cn0
vo
T-ixa<CN13Cu
- 46 -
TABLE 4 . 1 1
e (cm)
20
25
30
41
-rsotope
Pu-239
Pu-240
Pu-241
Pu-239
Pu-24 0
Pu-241
Pu-239
Pu-24 0
Pu-241
Pu-239
Pu-24 0
Pu-241
2 Mwy/m
99.2
0.8
99.16
0.83
0.01
99.16
0.83
0.01
99 .26
0.73
0.01
24 Mwy/m
98.3
1.7
98.21
1.76
0.03
98.21
1.76
0.03
98.39
1.59
0.02*
26 Mwy/m
97.3
2.6
0.1
97.08
2.87
0.05
96.97
2.98
0.05
97.26
2.70
0.04
TABLE
HYBRIDff i s s i l e
= 25 era HYBRID, = 30 cm
2E(Mwy/m )
2
4
6
8
Pu
7.00
15.0
24.0
33.0
E(Mwy/m )
2
4
6
8
Pu
8.6
18.7
30.0
44.0
(T/Gwe x year)
LWR
Uraniumcycle
0.22
Plutoniumcycle
0.86
HWR
Uraniumcycle
0.50
HTR
Downenrichment
0.11
G.G.
0.60
A.G.R.
0.22
FBR
Firstgeneration
2.21
Advanced
2.28
- 48 -
truction of actinides and fission products. In tíiis concept, the
cladding is eliminated, emerging. the homogeneous concept. In it
the f-uel is flowing in liquid phase across the primary coolant
loop, not needing any support or structural elements to confine
it.
The system just described takes te SO.UO2 + H-0 as fuel
and shows a configurational plant as pictured in figure 4.26.
The blanket región is composed by HLi as tritium producer»-
A 14 MeV neutrón source is taken, producing a significant •
3 MW/m in the first wall within the allowed tecnological limits.
In figure 4.23 is represented the parametric analysis to
know the composition of the mixture wich gives a higher multipli-
cation factor. From that picture, a 40 Kg of natural uranium in
13- liters of H-0 seems to be the searched valué. In any case a
subcritical system condition (Keff = 0.77} is obtained. The
following results nave been provided with that composition.
In the results of figure 4.24 can be seen. a fast to thermal;
flux ratio $-/<!>, = 1.38, whereas in a LWR that valué is =5. The
codes LEOPARD C21} and WIMS-TRACA (22) have been used in these cal-
culations. As a comparisonr a typical spectrum of a fast reactor
(SNR-300) is shown in figure 4.25, noting a flux relation of8
With the code XSDRN, some calculations have been made,
whose results are represented in figure 4.26b where the spectra
in different radii of the system are given. A very strong thermal
absorptipn by HLi is observed coming from the large cross section
( = 950b) o-f Li (nfa) reaction. This specific aspect has been
proved with WIMS code working on a slightly multiplicative system
with HLi viewing the hardness of the spectrum by this effect in
figure 4.27.
- 49 -
16
i o 1 5
i o 1 *
1
.1
1
1
•
t 1
i o 1 3
ca
<N 1 2
s i o 1 2
c-9-
i o n
i o 1 0
1
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1
11
1
f1 i
I
1
TÍVI
1!
• — > • •
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1
f i
r !
tí —Í
1
4 .
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r
i1
i!
' 4— i - - -
i
i
¡!¡
t-. i
i
i?
II^IIH
1
l !
—r+i-f——' í ¡ i1 í1
1
11
i
l iI¡
f
i
i¡ ,
i i
y1 -t-4
'? í•i-íi •
.1
li-
l'l •
i 1
i
""" • •
f!
t 1
i i'
i
i
1;!
1-f*-«-H4—
i . :
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f :¡
r- L ¿ . 1 J
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1
j
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n!
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1
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T"
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f i ¡ II• ¡ ; • i | ¡i
i M i !I,
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I
i; i • i
i ' ! , ' .'• i i!' í i ' i ' i í
J_ . i ., .,],;• 1
..,, I; ,„
i i ' !i i . I
. i i.L,
-li [frr w*
¡i
1 • 1 M i i! • • : 1 1 !
1 i i I1
! ! i
! i1
"IIEZi!:! •
i
il!• | 1 ; i) ii
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i i
¡ i
1 i ' i i 11
! l ' ¡ ' Ij" t" ' ! I1
1
i ;11,i
¡ .
: .
; ^ tf! i ¡ i :
i i
10" 10 .10 10' ECevl
n = 1 Mw/m • U - 25 era ; B • 0m
FIGÜBE 4.15 - FISSILE ZONE FLUX
- so -
10 E(ev)
d = 1 Mw/m2 ; e z o a e U = 25 cm ;Mwy
m
FIGURE 4,16 - FISSILE ZONE FLUX
- 51 -
1016
10
m
m
so
10
F1 :
•
i
I
5 !
f
k
\r
i
13
i "
o10
i
!
I
• 1
1 ;1 - í
1
1 1
j
'
i
i
i-—;~Ur r—p
i
1
1
í i
• " " " • i : '
- • . !
1 ¿. • ; ; • T
' ;
; ' ;
• ! •' ! ; í
¡ '• Ü¡ I ; i :
i i ; . .j 1
1' i '1 • i i '
¡ V
: ; ¡ ¡ . : ;! |Tt• * >
• í i
• i - * ' !'(•'
" i i. i
; i1'' !;,¡1 ii i1 " i
• ;
• ¡iI , 1 ; ; .
• ¡;|!
'¡1;
» - - t -
!
i j j i ¡ i
¡ ! 'i. •••• i !
'" Hj
.... i j
¡1¡ l
h
MÍ
1 i ' ' ! i
1;
11
1
1
.MI T
i ¡ • . „
l¡t • ;
— i i 1
. '•• ¡ it ' t
¡;í;
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i
I !1
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i 'H i1
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1 .1' i !
i 'i • •!
—•—V
iíil-rhl1 ' '
1
I
i _ . , _ . , „
t
> '
It !
- ^ - ' " • ' • " " » - —
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i
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i Í 'i "1 I !
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i
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| i j i t
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i ¡
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t
1
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11i
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lt i1
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i t
i
1 ' ! '¡ ii • ,
i i
i :
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i- —_. 1
i : 1 :
; I
1 !
i <! '
i ii¡
í!ni
. i 1 ii
—4! i!
iI
> i
i f !
; i, i '1i
; j
j '
!li ¡j
| i |
i¡l
1t
i
; !
i i
I |J
• , ,
i !
L1 '
í '•'•
¡ii
• i i
ill
__| ¡i
I1J
i i1 1
! ! !
ii
1
i i":! i1
i
¿ 7
10'10 JO' 10 10'
Elev)
a - 1 Mw/m2 ; e _ _ U = 25 cm ; Bzone
<>m
FIGURE 4.17 - FISSILE ZONE FLUX
- 52 -
,o« ^ ^
oc-e-
-'1 Mw/m2 ; eZ Q n e
U - 25 cm ; B - 6 ^ <> 6 years
FIGUEE 4.13 - FISSILE ZONE FLUX
- 53 -
1610 .-
!_..,' - • I I-Í;; i >r
i li
, o 1 5 .
1 o 1 3
, o 1 2
ion
i o 1 0
!
i
r ,1
t
1i
f
it
i
t—i—
11
l i l i
mililii
IÍh+tj—
iii'!¡¡I
,¡i!
':'
I
i !
• 4 H I —i1
t
ji ; „
H|• |
i !l
¡ ! | i |
'; i'i
'''Ili
!'
Ti
4 U
1PT J~
i l l
•
T
mmimk
|
Jj —
i
II
4-itr-i h
1
•
.±-—1_4-
11 ii ' i
1 i
i i
|
i
l
'
II
~ljjLlMI
jj|
! ,1
illI
1
• í
!¡
!
!
¡ ;
- T 4 - r
J
¡
—H~^¡
¡
i ¡
¡
i1
i
—:r-3i=—--*i—f -' -i '• •
•t trr ¡ '• i'"!"l•+ -yt i 1 1 - -
I , i
I
!
HJ|j j
1
! 1
¡ ) '
1
j
7 i
t i .
rá 'i ' ! j
; ti
i]
t r —L-1 |1 ¡ ' t
; :
^r i-
; ;; 1
1
l i• • 1
i ! i i
i
1!ilM
1
t
I |
1 1 '
.¡ ;• i ¡ *.,." [ i
| II' ! ¡ill
i1
1 1
• 1! iw !
ü;<
i >
r
rJ1
'i> 1
i1 ! 'i
I11
1 - • 1
l i|
1 i i
1
1¡iMil
i
Í i I í !
1
i ; ¡ 1
\
10" 10 10' 10 10' E(_ev)
s 1 Hw/ra ; ezoneU = 25 cm ; B = 8 <> 8 years
m
FIGURE 4J.9. - FISSILE ZONE FLUX
- 54 -
l
T"• "• r | !' ; ! • ; ' 1 : .!__ L _LL_LJ ; i '; |i : i i I I : . T..::;:-::
i
1 !
' ' í
; í i
i .
1-; • ; i !.• • • • ' . i •
1 • : . :...!..
i < . ; i
I . 1 •
; : : !
CM
- 55 -
:'I ~ ¡ ; r~ r—~i ~ i !
i ' - ' . - . i •_ _ . - r • . . . _
( S * UIO/U")c
3 + 15
2 + 15
1+15 '
9 + 1 k
6 + 1
25, 30
30
25
20
FIGURE 4<,22 _ SCALAR FLUX (n/cm ,s) í EXPOSURE (Mwy/m )
B
- 57 -
CN
OCN
as
4J(t3
en
o
Oco
es
mos3OH
4
OCN
OO
OCO
oo»
o
00
O
ID
O
CN
- 58 -
>O)
63
¡4o
3 H
a m
cQ 3
esCU í¿
ow o
M
3
00 voi-i r4
9,
8.
7.
6.
5.
4.
3.
2.
1.
3.68MeV
2,
5MeV 24,8KeV
WIMS-TRACA
SNR-30Q
Faat System
= 1,1 x 10
Kef » 1.262
FIGURE 4.25
!,eV
uivo
1.13 eV
0. 10, 14. 18.
,02 eV
20,
1025
4 0 Kg ü nat/13 1
Q = 3 Mw/m (n,14 MeVl
10
10
1016
1015
1014
1, 3. 5. 7,
10 Mey 5Mey 24.8MeV
9. 11, 13. 15. 17. 19. 21» n454.eV 22.6eV 1.13eV .02eV E
10 2 2
10 2 1
io 2 0
io19
iola
^>1 = 14
é = 2
'lll =0
*117 = 0
,_*121 =0
íof 6
•io14
?20
¿72
•l
XSDRN
.92
,23
454
.16
,06
MeV ••
MeV ~
eV -
eV -
eV -
,021eV -
- 61 -
I
FUSIÓN
13,50 MeV
2,02 MeV
353,6 eV
0,14 eV
0,05 eV
0,015 eV
" FIGURE 4.26 bis
Rl
iFissioA\\\
SO4UO2+E2M
40 Dnat^!l+ 13 1 H2oV^
\
\
\
\\
\
•
LiH
R3
Bla
nk
et
ww\\V1 \i1\
•
^72
•lll
^127
- 62 -
<<osí
en2
en
mino*
oo
mino
<nj.esvo•
V0
eny
•
w
in
mo*
co
in
ino*
vo1en«r»
aoíin
•m
MH
1Ti
a
VC1•H
X
co1P
m1
CN 00 <N
- 63 -
SCALAR FLUX Cn/cm2.s) VERSÜS ENERGY C36.20)
FAST BLANKET XN FUSION-FISSION CONCEPT
i ; ; : ; i : ¡ ! i / l¡ i! ! V ! • ; : \ 7 \i i ; \ A< * •• I ^
".¡ilii ¡|!
! i i i
¡ i !i ! i i i
. I I I !
Al\i ¡"¡vi)
\l i ! ! 11 ¡ ^
i ! !
I i i
! : i • - i ¡ il i ¡ i i ÜÜÍ ¡\ ÍVi
'lü! zm li 1 Mi ni; lUlWl KüliV!¡ I iÜ l í l i l i
i ! i1I í i Muí !
! Ü líl ! Niüi. I i' ; i ! •
• ¡ ¡ / ; i ; ¡
! i l i í l I M i
i i Ü! llü ! i ! i i I ' ! ' !
i
9 ! ! ! M¡ i
I I l i l ! i l l ü l ! ! I I I lü iüií_!_' * • ;
i M i¡¡¡!
a
!
1
ij
!
yA
.! .Á '
\ \
A '
: : ¡ ;
i i' i
• : • /
fi?>'••:
• M;
c
il
i!
/ i
/ !
¡
i
i
1
i
i
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| |
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• i i !! ¡ i i
Ül¡
i
i
i
i
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i!.
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:l\
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i
i
i
i
1
;
i
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;
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:
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i i
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i ¡
i! i
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i
i1
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1 i I• ; 1
i i 1i ; i
• ' ;
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i i >
• \
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i i; i
i ¡
•
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i! í l
¡i : i
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i
i
i
; •
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i !
i ;
í j
i :i •
! i !
1 i ]
j | ;
i i i
'•
i
i
j
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¡
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:..
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i'1
•
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;
i MU:i j I ! i ¡
' 1
! ¡
!¡i i
i
in '
1 ! !i i !
! ; . í
! '• • i
i \ \\\\\\
i ' ' 1
i ! i ii : ;
! ¡ - i
. ; ; i
; • i 1 1
i i ; ! i
t ' ' i 1
' • • ' ' • * . " -
• •• - a
i ! ¡ !
Mil
i - '
i ' i
i i!
iii
i • ; \ : ' • > : ;
1 1 • : '•• ! • { i
i ¡ i j ü j i I
¡ : ¡
¡ } }
• • '-- • 1
; ' i :!
i ; : •
• : ; i
R- = 4 f 8 cm
0,95 ^m
FIGURE 4.28A 236.20 CZone U)
4 - 336.20 CZone Li)
- 20 c m / e t o t a l B = 100 cm.
- 64 -
SCALAR- FLUX (n/cm2.sL VERSUS ENERGY C36,101
FAST BLANKET IN FUSION-FISSION CONCEPT
~n
¡ ¡ i : !
I ¡ • • ' * í i• M i l : i ! Milu í ¡ i i i i ü i
! ! ! iHJliili.j ; i ¡m i• I S I !
¡ < 1 1 !
Mil1 . !!: • • i \
• \ • / : :'••: A
i ! /' i ¡ ü ! ! \ / . i i \ - - U ' ¡\ 'V Ü i : ! IIi íi j / ! j / ' i Í | Í i j j j Mi ! i ! ¡ ! i : í !
I ! ! vwm i i !¡ i ' 1¡ : !
; ' I ; . ; ' .
í ! ¡ / ! ' , ' • !
i í ¡i' ! ¡ i ; i-.üi' (i
! / !I ¡ ! i : ¡ Í !
; i \\\\\\\ i i Ü Ü I - !: : ; i ü ü ! I ! ! M i i i ¡ ! i ; ¡:¡¡
• ! i . : • • > : I
- ! I ' ! i ¡ i ; : u ! i i ! I; ; J i ¡ i i ! ! i ! l ! ¡ : ' ! i i '
c . ! ; i M ! ¡. • : i í
¡ MU;
• : i ; : •
: i Í ' / • • >
' / : • :
• <
• . /
• / . •
¡ ¡ ' •
• : • ' ' • ; .
; \ ' \ ¡ \ \
'• ¡ ; ' : ¡ ; ¡
; ! i i \'\\
j i • i i ! í i¡ : ; • i : ' •
'• ' '• • \ ' • ; ! ; • ! ; : .
; i i i i 111 ' i ¡ MÍ' • • • ' . ' . ' » i • t
M : Í i : ¡ | ¡ '
I ; : ¡
i ! ! l : ¡ ! n ¡ M i i * : ; ; ¡ •: : í ' ' i : .i i l ¡
J'-\A H n '
; i i : : ¡ i |fí
í i i Üiíi
L = 4,8 cm
0 f 9 5 22m
eZiE = 1 0
>C236.10 CZone U)
<5> 336.10 (Zone Lil
"4*. "2 9
- 65 -
SCALAR FLUX Cn/cm .si VERSUS ENERGY C36.30)
FAST BLANKET IN FUSION-FISSION CONCEPT
7'/ •• X ' ; = ' ! ! !
l l ! ü ¡ ¡ ! i ! l
! l i l i! ! I l l l l
I í ! I
l/!!i¡l
Mil ü ¡
i! 2¡ii I i ! h ! i j
1 i i : i i
\ ¡ : •'< I ,i í '• , i ! ! | • /I ü i l / ; . \ • ! : ;
l / i l ) i ' > ! : I A Í Ü / \ l i
i ! M lllÍL: • i ; ¡ ' i ! i ; , \ ¡ : : !
i • ' •
iü! i i ^ i 1 !I i : l!i
üliü• \A v., w
: : : ; , i!; t ¡ \ '
• '• '• ' i ' \ í TTíi ' i ' '••
/ i / ¡ ü ' l í i ' i ! ; ' ! • • i !
• ; •! ! > ÍÜI ¡ i i ¡>-r
¡ ! ¡ üii!| i Mjír*í W\\\
i i !! ¡ ! Üií'l
'. : : I
! i i ; : ¡
i ; i i ! i ; ; í ii ¡ : ;
: ' • • • • /
i i ; i ¡ / ; ¡ ; ! i i ; n
i i ' i ; ; I ! ¡ | Í ¡ i i iü iü: i \ '• ' l \
¿ni i j ' ü ^ n i!
: ¡ I ; I : i
i ! H !:!
'in =! i
n ; !
• I I 1 .;
i l i i !
IL = 4 . 8 cm. FXGURE 4,3(1 Z 236 .30 CZone ü i
0,95 ~2~m
= 30 c m / e t o t a l B = 100 cm
Y 336.30 CZone Li)
- 66 -
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- 74 -
REFERENCES
1. CARO, R, MINGUEZ, E., PERLADO, J.M., Study of hybrid reactors:
high burnup and waste destruction, JEN/TCR/A 79-17, presented
to the V annual meeting of the Spanish Nuclear Society, 1979.
(In Spanish).
2. CARO, R., MINGUEN, E., PERLADO, J.M., Analysis of on advanced
hybrid reactor concept. A calculation model, presented to the
Fusión meeting of the Spanish Nuclear Society, -june
1980. (In Spanish).
3. ENDF/B-III-IV, GARBER, 0. et al., Data Formats and Procedures
for the Evaluated Nuclear Data File, ENDF, BNL-NCS-50496(1975}.
4. KRAINER, H., CODAC, A., Fortran IV programme to process a
TIMOC library from the ENDF/B, EUR-4521e C197 0).
5. PERLADO, J.M., Guide to the use of the CODAC program,
TRCN-22 (1975) . Un Spanish) . . J E ^ £irterna3l publicatidn.
6. PERLADO, J.M.r TIMOC-RATE, A program based on the TIMOC code
to calcúlate reaction rates,(to be published (1981)i.
7. KSCHWENDT, H., RIEF, H., TIMOC, A general purpose MonteCarlo
code for stationary and time dependent neutrón transport,
EUR-4519e (1970).
8. JAARSMA, R., RIEF, H., TIMOC-72, code EUR-5016e (1973).
9. JAARSMA, R, PERLADO, J.M., RIEF. H., ESP-TIMOC Code Manual,
EUR05794e (1977).
10. ARAGONÉS, J.M., HONRUBIA, J., Utility programs to represent
and. analyze the cross section from the ENDF/B, JEN/TCR/A 79-10
(1979) . (In Spanish) .
11. Reference design for the Standard Mirror Hybrid Reactor.
Lawrence Livermore Laboratory, UCRL-5?478,
12. GREENE, N.M., CRAVEN, C.W., XSDRN, A discrete ordinates
spectral averaging code, ORNL-TM-2500.
- 75 -
13. BELL, M.J., ORIGEN, The ORNL Isotope generation and depletion
code. ORNL-4628 (1973).
14. LEE, J.D., Mirror Fusion-Fission hybrids. Atomkernenergxe
32(1) 19-29" (1978) .
15. LEE, J.D., Neutronic Analysis of a 2500 Mwth fast fission
uranium falanket for a DT fusión reactor, CONF-740402-Pl (1974) .
16. HANS BOROUGH, D., WERNER, R.W., A modular fission-fusión
hybrid blanket, CONF-740402-Pl (1974).
17. HAIGHT, R.C., LEE, J.D., Calculations of a fast fission
blanket for DT fusión reactors, CONF-740402-Pl (1974).
18. LEE, J.D., Subcritical fast fission blanket, ÜCRL-73952 (1972).
19. Neutrón Cross Sections, Brookhaven National Lab., BNL-325.
20. YIFTAH, S., OKRENT, D., MOLDAUER, P.A., Fast reactor cross
sections, Pergamon Press (1960).
21. BARRY, R.F., LEOPARD: A spectrum dependent non-spatial deple-
tion code for the IBM-7094. WCAP-3269-26.
22. AHNERT, C , WIMS-TRACA, para el cálculo de elementos combus-
tibles. Manual de usuario y datos de entrada. JEN-461.
23. LINDSKY, L.M., Fission-Fusion system: Hybrid, symbiotic and
augean. Nuclear Fusión 15 (1975) .
24. WOLKENHAUER, W.C., et al., Conceptual design of a fusion-
fission hybrid reactor based on a mirror fusión reactor with
a subcritical gas-cooled fission blanket, CONF-740402-Pl (1974)
25. GREENSPAN, E., et al., Source driven breeding thermal power
reactors using D-T fusión neutrón source. Aromkernenergie
32(1) 30-38 (1978) .
26. LONTAI, L.N., Study of a thermonuclear reactor blanket with
fissile nuclides. MIT research laboratory of electronics,
Tech. Rep. 446 (1965) .
- 76 -
27. PARISH, T.A., DRAPER, F.L., Jr.f Neutronic and photonic
analysis of simulated fusión reactor blanlcets containing
thorium and natural uranium. University of Texas, ESL-16
(1973) .
28. LEONARD, Jr., A hybrid neutronics analysis. USAEC report,
BNWL-1685 (1972) .
29. LEONARD, R.B., WOLKENHAUERf W.C., Fusión-fission hybrids:
a subcritical fission lattice for a D-T fusión reactor.
ÜSAEC CONF-721111 (1974) .
30. KOLESNICHENKO, Ya., RESNICK, S.N., D-T plasma a source of
neutrons for the combustión of U-238, Nucí. Fusión 14 (1974).
31. WOLKENHAÜER, W.C., LEONARD, L.B., Jr., et al., Conceptual
design of a fusion-fission hybrid reactor based on a mirror
fusión reactor with a subcritical gas cooled fission blanket,
USAEC report BNWL-4865 (1974) .
32. Proposal for a driven thermonuclear reactor, California
Research Corp. LWS-24920 (1953).
33. IMHOFF, D.H., A driven thermonuclear power breeder, Cali-
fornia Research Corp. Report CR-6 (1954),
J.E.N. 509
Junta de Energía Nuclear. Sección de Teoría y Cálculo de Reactores. Madrid.
"Reactores hídridos para sistemas de fusión por con-finamiento magnético. Análisis preliminar y modelosde cálculo".
CARO, R.; MINGUEZ, E.; PERLADO, J.M. (1981) 76 pp. 45 f l gs . 33 refs.
Se ofrece una revisión general del concepto híbrido de reactor, junto a un amplio
conjunto de cálculos efectuados sobre configuraciones geométricas t ípicas. De esta
manera, se validan e"< modelo y los métodos de cálculo propios, y se obtiene una impor!
tante experiencia sobre la sensibilidad del sistema a diferentes perturbaciones en el
mismo.
El análisis del blanket como multiplicador de energía y reproductor de t r i t i o y mg
te r i a l f i s i b l e es el objetivo básico.
Los cálculos se centran tanto en sistemas homogéneos como heterogéneos.
J.E.N. 509
Junta de Energía Nuclear. Sección dejeor ía y Cálculo de Reactores. Madrid.
"Reactores hidridos para sistemas de fusión por con-finamiento magnético. Análisis preliminar y modelosde cálculo".
CARO, R.; MINGÜEZ, R.; PERLADO, J.M. (1981) 76 pp. 45 f l gs . 33 refs.
Se ofrece una revisión general del concepto híbrido de reactor, junto a un amplio
conjunto de cálculos efectuados sobre configuraciones geométricas t ípicas. De esta
manera, se validan el modelo y los métodos de cálculo propios, y se obtienen una Impor
tante experiencia sobre la sensibilidad del sistema a diferentes perturbaciones en el
mismo.
El análisis del blanket como multiplicador da energía y reproductor de t r i t i o y ma
te r ia l f i s i b le es el objetivo básico.
Los cálculos se centran tanto en sistemas homogéneos como heterogéneos.
J.E.N. 509
Junta de Energía Nuclear. Sección de Teoría y Cálculo de Reactores. Madrid
"Reactores hidridos para sistemas de fusión por con-finamiento magnético. Análisis preliminar y modelosde cálculo".
CARO, R.; MINGUEZ, E.; PERLADO, J.M. (1981) 76 pp. 45 f l gs . 33 refs.
Se ofrece una revisión general del concepto híbrido de reactor, junto a un amplio
conjunto de cálculos efectuados sobre configuraciones geométricas t ípicas. De esta
manera, se validan el modelo y los métodos de cálculo propios, y se obtiene una impoj;
tante experiencia sobre l a sensibilidad del sistema a diferentes perturbaciones en el
mismo.
El análisis del blanket como multiplicador de energía y reproductor de t r i t i o y nvg
te r ia l f i s i b le es el objetivo básico.
Los cálculos se centran tanto en sistemas homogéneos como heterogéneos.
J.E.N. 509
Junta de Energía Nuclear. Sección de Teoría y Cálculo de Reactores. Madrid.
"Reactores hidridos para sistemas de fusión por con-finamiento magnético. Análisis preliminar y modelosde calculó".CARO, R.; MINGUEZ, E.; PERLADO, J.M. (1981) 76 pp. 45 f l g s . 33 refs.
Se ofrece una revisión general del concepto híbrido de reactor, junto a un amplio
conjunto de cálculos efectuados sobre configuraciones geométricas t ípicas. De esta
manera, se validan el modelo y los métodos de cálculo propios, y se obtiene una Impor-!
tante experiencia sobre la sensibilidad del sistema a diferentes perturbaciones en el
mismo.
El análisis del blanket como multiplicador de energía y reproductor de t r i t i o y ma-i
te r ia l f is ib le es el objetivo básico.
Los cálculos se centran tanto en sistemas homogfneos como heterogéneos.
Se consideran combustibles UOn y UC, con espesores totales del bianket de 1 m.
Otras soluciones, como U3SI, son también analizadas.
Espesores de 20 ó 30 cm de la zona combustible dispuesta internamente en el
blanket, aparecen como soluciones válidas para éstos sistemas.
CLASIFICACIÓN INIS Y DESCRIPTORES: E36; F51. Hybrid reactors. Confinement. Breeding
blankets. Breeding. Tritium recovery. Optimization.
Se consideran combustibles IJO y UC, con espesores totales del blanket de 1 m.
Otras soluciones, como U3S1, son también analizadas.
Espesores de 20 6 30 cm de la zona combustible dispuesta internamente en el
blanket, aparecen como soluciones válidas para estos sistemas.
CLASIFICACIÓN INIS Y DESCRIPTORES: E3B; FBI. Hybrid reactors. Confinement. Breeding
blankets. Bresding. Tritium recovery. Optimization.
Se consideran combustibles UOT y UC, con espesores totales del blanket de 1 m.
Otras soluciones, como U-jSI, son también analizadas.Espesores de 20 ó 30 cm de la zona combustible dispuesta internamente en el
blanket, aparacen como soluciones válidas para estos sistemas.
CLASIFICACIÓN INIS Y DESCRIPTORES: E36; F51. Hybrid reactors. Confinement. Breeding
blankets. Breeding. Tritium recovery. Optimization.
Se consideran combustibles U0£ y UC, con espesores totales del blanket de 1 m.
Otras soluciones, como U3SI, son también analizadas.
Espesores de 20 ó 30 cm de la zona combustible dispuesta internamente en el
blanket, aparecen como soluciones válidas para estos sistemas.
CLASIFICACIÓN INIS Y DESCRIPTORES: E36; F51. Hybrid reactors. Confinement. Breeding
blankets. Breeding. Tritium recovery. Optimization.
•
•
J•1
1
}
J J . E . N . 509J• Junta de Energía Nuclear. Sección de Teoría y Cálculo de Reactores. Madrid.
i "Hybrid Reactors with Magnetic Confinement.
i Preliminar y Analysis and Calculational Model".J CARO, R.; MINGUEZ, E.; PERLADO, J.M. (1981) 76 pp. te f i gs . 33 refs.
| Begining with a general revisión of the technical concepts a great amount of calcu-
l lat ions on typlcal geometrical configurations Is given, comparlng these results with
i those found 1n the references. So, we obtain the validatlon of our own methods, and
¡ Important experience on the sens i t iv l ty of different effects in the hybrid system.
¡ Is our main interest the analysis of the blanket as energy inult lpl ier and breeder
i of trit'uiin and fissionable material.
¡ The calculations are focussed on heterogeneous and homogeneous blanketso
¡ The i n i t i a l work Is centered 1n ÜO2 and UC fuels, with 1 m tota l thlckness of the1 blanket. I t seems that, thicl<nesses of 20 or 30 cm of the fueled zone In the interna!
1 position are good enough to obtain Interesting hybrid solutions. Other solutíons, as
J.E.N. 509
Junta de Energía Nuclear. Sección de Teoría y Cálculo de Reactores. Madrid.
\ "Hybrid Reactors with Magnetic Confinement.
Preliminary Analysis and Calculational Model".
[ CARO, R.; MINGUEZ, E.; PERLADO, J.H. (1981) 76 pp. 45 f i gs . 33 refs. !1 Begining with a general revisión of the technical concepts a great amourrt of calcu- !1 lations of typical geometrical configurations i s given, comparing these results with j
those found In the references. So, we obtain the validation of our own methods, and1 important experience on the sensi t iv l ty of different effects In the hybrid system.1 Is our main Interest the analysis of the blanket as energy multipUer and breeder !1 of tr i t ium and fissionable material. '¡ The calculations are focussed on heterogeneous and homogeneous blankets.1 The I n i t i a l work Is centered In UO7 and UC fuels, with 1 m total thlckness of thei blanket. I t seems that, thicknesses of 20 or 30 cm of the fueled zone in the intemal (
1 position are good enough to obtain Interesting hybrid solutíons. Other solutions, as
J.E.N. 509 !
Junta de Energía Nuclear. Sección de Teoría y Cálculo de Reactores. Madrid. 1
"Hybrid Reactors with Magnetic Confinement. 1
Preliminary Analysis and Calculational Model". 1CARO, R.; MINGUEZ, E.; PERLADO, J.M. (1981) 76 pp. te f igs . 33 refs. ¡
Begining with a general revisión of the technical concepts a great amount of calcu- j
lations on typical geometrical configurations Is given, comparing these results with '
tiloso found In the references. So, we obtain the validation of our own methods, and 1
Important experience on the sens i t iv i ty of different effects in the hybrid system. 1
Is our main interest the analysis of the blanket as energy mul t ip l ier and breeder ¡
of t r i t ium and fissionable material. '
The calculations are focussed on heterogeneous and homogeneous blankets . 1
The I n i t i a l work is centered in UO2 and UC fuels, with 1 ni tota l thickness of the 1
blanket. I t seems that, tbicknesses of 20 or 30 cm of the fueled zone In the internal ¡
position are good enough to obtain interesting hybrid solutions. Other solutions, as '
1
J.E.N. 509» *
Junta de Energía Nuclear. Sección de Teoría y Cálculo de Reactores. Madrid. ¡"Hybrid Reactors with Magnetic Confinement. ¡
Preliminary Analysis and Calculational Model". |CARO, R.; MINGUEZ, E.; PERLADO, J.M. (1981) 76 pp. 45 f i gs . 33 refs. ¡
Begining with a general revisión of the technical conceptos a great amount of calcu-1
lations of typical geometrical configurations i s given, comparing these results with 1those found 1n the references. So, we obtain the validation of our own methods, and 1Important experience on the sensi t iv i ty of different effects In the hybrid system. |
Is our main Interest the analysis of the blanket as energy mul t ip l ier and breeder 'of t r i t ium and fissionable material. 1
The calculations are focussed on heterogenoous and homogeneous blankets. ¡The I n i t i a l work Is centered In UO? and UC fuels, with 1 m tota l thickness of the J
blanket. I t seems that, thicknesses or 20 or 30 cm of the fueled zone in the intemal ,position are good enough to obtain interesting hybrid solutions. Other solutions, as 1
fuels, are considored.
INIS CLASSIFICATION AND DESCRIPTORS: E36: F51. HyMd reactors. Confinement. Breedingblankets. Breeding. Tritium recovery. Optimization.
U3SI fuels, are considered.
INIS CLASSIFICATION AND DESCRIPTORS: E36; F51. Hybrid reactors. Confinement Breedingblankets. Breeding. Tritium recovery. Optimization.
U3SI fuels, are considered.
INIS CLASSIFICATION AND DESCRIPTORS: E36; FS1. Hybrid reactors. Confinement. Breedingblankets. Breeding. Trit ium recovery. Optimization.
U3SI fuels, are considered.
INIS CLASSIFICATION AND DESCRIPTORS: E36; F51. Hybrid reactors. Confinement. Breedingblankets. Breeding. Tritium recovery. Optimization.