ipen · 2015. 3. 30. · toda correspondencia en relación con este traba-jo debe dirigirse al...

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

- 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

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1 1

i

1 > 1 •- : ! t t ! i . t . i • t . ) 1

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1 í i i

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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

O«3*CN13

t—1

rocncn

1" • • "

vororo

CO1¡«CN

00

cn1

cnco

T—|

TCN13CU

m

CO

H

t-1

+t-1VO

•H

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

•• ii ! ', r

1

11

1

f1 i

I

1

TÍVI

1!

• — > • •

-J.

1

f i

r !

tí —Í

1

4 .

-4|J

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 . :

-i

f :¡

r- L ¿ . 1 J

MI

1

j

I

n!

' ' 11 —

1

_4Z_..

T"

-i

MHIHi i

I

Mii

1

' 1 *

—,—j±^

! i1 i ! i '• i i i

f i ¡ II• ¡ ; • i | ¡i

i M i !I,

¡;

-í

\ ¡ 1

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

• ! i 1' Ü

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;

„ X I ' ni. i

i

I !1

1üj 1—}„,„;. I ,„1 1 '

i 'H i1

• ! '•

! l i _ _; . !

Í!: '1

1 .1' i !

i 'i • •!

—•—V

iíil-rhl1 ' '

1

I

i _ . , _ . , „

t

> '

It !

- ^ - ' " • ' • " " » - —

_ _

'"'! " í

!

t1

II

! 1!

11

1.A

i

U

1

r*ir i — — —•u ¡ —rt

i || :

l |

'. i

! •! )

. i !

; i; ;

" ! ! •

\

•

i i '

í ! 1M ! L

i

i

- - 44

1

•

¡1r

T-;—:i |"

i Í 'i "1 I !

1

J- -i

¡

r •*"| ' 1 , | i

" ! " Í ! ¡!; ií!

JJtT*-1

i

i

- i 1

• j - - i" i '

___

—-t

!f

i

.,._ ., ; !_ . _,1! j • i i 1 1! ¡.¡1 ¡

| i j i t

:!

¡¡

| ¡) (J

i ¡

¡;i

¡i

¡

i

• ; "

• M i

j 1 " "

| |

¡ '!'

t

1

i i!¡ i 'i III

' 1 i i i

1 M|i !—-! li1

1 ' i¡

j

: IIj ,

11i

-iL" lili""'

II1

lt i1

! ¡

• r

' 1 !

~t 1"í 11

¡1, !

i t

i

1 ' ! '¡ ii • ,

i i

i :

-| i

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

! i

| |

;: • • !

• i i !! ¡ i i

Ül¡

i

i

i

i

!•:

i!.

!!!¡ '1

:l\

!;!

i

i

i

i

1

;

i

i!i

•

i !

: -

' i

i t! '

;

• i

! •

1 •

'. .

:

!

| i

J i

i i

!! 1

i ¡

i! i

•

!

i

i1

•' ¡

: • t

1 i I• ; 1

i i 1i ; i

• ' ;

| ij

i i >

• \

! j

i i; i

i ¡

•

¡i i i

i! í l

¡i : i

! ! ' ;

;i ! !

i

i

i

; •

¡:

i !

i ;

í j

i :i •

! i !

1 i ]

j | ;

i i i

'•

i

i

j

! ! i

¡

; i

' ¡ !

:..

!'

i'1

•

!

;

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)


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 -

S-GALAR FE8X- ( j i / c m 2 . s L VERSÜS ENERGY C26,10l


•3

i ! I I i ¡ | Íi ; i li ! i! i i I i ! ÍÜ i ! | | A

! l i M Ü i l ¡ i '• ! J J i i i I I I ! i ' i j ! / \ = ! i i i S l i j . l i l i i i ü j ; 1 M ¡ H ' !

¡H1

i i

Ül!i ! i l i AJ lllíii i ! Üliilli

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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.



manera, se validan el modelo y los métodos de cálculo propios, y se obtienen una Impor


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.


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.



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.


J.E.N. 509


"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.



manera, se validan el modelo y los métodos de cálculo propios, y se obtiene una Impor-!


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.



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.



•

•

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


\ "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.


INIS CLASSIFICATION AND DESCRIPTORS: E36; FS1. Hybrid reactors. Confinement. Breedingblankets. Breeding. Trit ium recovery. Optimization.


INIS CLASSIFICATION AND DESCRIPTORS: E36; F51. Hybrid reactors. Confinement. Breedingblankets. Breeding. Tritium recovery. Optimization.

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