www.themegallery.com
Company Logo
‫بسم هللا الرحمن الرحيم‬
‫﴿ َقالُو ْا ُس ْب َحا َن َك لَ ِع ْل َم َل َنا إِلا‬
‫َ‬
‫ْ‬
‫ْ‬
‫َ‬
‫اعلا ْم َت َنا إِ ان َك أنت ال َعلِي ُم ال َح ِِكي ُم‬
‫َم َ‬
‫صدق اهلل العظيم‬
‫سورة البقرة‬
‫اآلية )‪(۳۲‬‬
Thorium Fuel Performance In a
Tight LWR Lattice
Paper presented by
• Sayed saeed abdelfattah
2012
Prof. Dr. Esmat A. Amin
Prof. of reactor safety analysis ,Nuclear $Radiological ,
Regularity Authority,Cairo,Egypt
Prof.Dr.Ibrahim I. Bashter
Prof of nuclear physics ,physics department, faculty
of Science,Zagazig university
Sayed saeed abdelfattah
Faculty of science,zagazig university
Introduction
Thorium fuel in PWR reactors
Characteristics of thorium fuel:• Thorium is 3 to 4 more abundant than uranium ,widely
distributed in nature as an easily exploitable resource in
many countries .
• Thorium oxide is chemically more stable and has
higher radiation resistance than uranium oxide.
• Thorium oxide has an extremely melting point
which is about 3300 C
Thorium fuel in PWR reactors
• Th -232 is a fertile material like the isotope U-238 which can
absorb neutron and become Th-233 which decays to Pa-233 by beta
decay.Pa-233 decays by beta decay to U-233 which is a fissile
material as U-235.
• The absorption cross section for thermal neutrons of Th-232
equals 7.4 barn which is three times that of U-238 (2.7 barn).
Thus Th-232 is a better fertile material than U-238 in thermal
reactors.
• The high thermal capture cross section of thorium will
require more amount of initial fissile material as a
compensatory where thorium will reduce neutron capture
in moderator so, the neuron yield of fission probability of
U-233 in thermal region is higher than that of U-235 and
Pu-239
The aim of the work
The aim of the work can be summarized in the following
points
• The aim of this work is studying the feasibility of thorium fuel
in PWR reactors which takes up more 60 percent of the
nuclear power plants in the world.
• This can be performed by using one of the thorium fuels as
thorium –plutonium oxide.
• Pin cell model is carried out for (Th-Pu)OX and many
parameters have been calculated such as k-infinity , fluxes,
average energy per fission ,atom densities of fission products
and actinides and the absorption and fission cross sections of
the produced isotopes at different stages of burnup.
Methods of calculation
Methods of Calculations of pin cell model
.The burnup calculation of this work is performed
using MCNP5 and WIMSD-5 code
.MCNP5 is computerized analysis tool used for
designing the fuel rod of thorium-plutonium oxide
.WIMSD-5 is a general lattice cell program which
can determine k-infinity ,neutron flux, burnup of the
fuel .wims5b library of 69 groups is used in these
calculations.
.the burnup calculations are performed at a constant
specific power (37.7 W/gm HM)
Parameters and dimensions of pin cell model of thoriumplutonium oxide
Fuel pellet radius
0.47 cm
Cladding radius
0.54 cm
radius Water (moderator)
0.85 cm
Temperature of fuel
1023 K
Temperature of clad
923 K
Temperature of water
583 K
The initial atom densities of fuel isotopes (atoms/cm3)
isotopes
Zone 1
Th-232
2.11E+22
Pu-238
9.72E+18
Pu-239
5.99E+20
Pu-240
2.32E+20
Pu-241
7.69E+19
Pu-242
4.78E+19
Cr
Zone 2
Zone 3
8.14E+19
Mn
Fe
1.60E+20
Ni
Zr
4.37E+22
C
2.68E+18
H
4.80E+22
1-K-inf versus burnup for pin cell of (Th-Pu)OX
• k-infinity versus burnup is shown in the following
graph
(IAEA-TECDOC-1349)
(k-inf by ORIGEN-2 code)
The present work
1.15
1.10
1.05
k-inf
1.00
0.95
0.90
0.85
0.80
0
10
20
30
40
Burnup (GWd/ton)
50
60
**K-infinity values of the reference results and WIMSD5 results
are tabulated as following:Burnup( GWD/ton)
0
30
40
60
k-infinity Ref ( TECDOC-1349)
1.12479
0.925198
0.887499
0.84756
k-inf ref by ORIREN-2code
1.112
0.889
0.851
0.822
The present work
1.127744
0.916267
0.873056
0.82698
k-infinity difference(Ref TECDOC1349 - our present work)
-0.002954
0.008571
0.01444
0.02058
From the previous figure and results ,we can say that:
**k-infinity for all cases decreases due to the depletion of fissile isotopes •
and the production of fission products and poisons. The remarkable
decrease in the beginning is due to the production of Xe-135 which has a
high neutron capture cross section and it has impact on thermal
utilization factor and thus multiplication factor .It is an important poison
in the reactor operation.
**The sharp decrease in the values of k-infinity at the end of burnup is •
due to the production of fission products which have high absorption
cross sections which are not taken in consideration in the results of the
reference (IAEA-TECDO-1349)
• **Our work represents the blue curve which is compared to the other
curves of the reference results. The results obtained shows good
agreement with the reference. this is indicated in the previous table.
2-fluxes of fuel,clad,moderator versus burnup
• The following figure shows flux of fuel versus burnup
(IAEA-TECDOC-1349)
The present work
flux n/cm2.sec
4.40E+014
4.20E+014
4.00E+014
3.80E+014
3.60E+014
3.40E+014
3.20E+014
3.00E+014
2.80E+014
0
10
20
30
40
burnup (GWD/ton)
50
60
**The total fluxes of fuel,clad,moderator as a function
of burnup are recorded in the following table:
Burnup(GWD/ton)
0
30
40
60
Fuel (Ref values)
2.913131E+14
3.5005688E+14
3.6624260E+14
3.8775357E+14
Fuel by WIMSD5
3.04E+14
3.82E+14
4.01E+14
4.250E+14
Clad(Ref values)
2.925591E+14
3.5062147E+14
3.6663592E+14
3.8785461E+14
Clad by WIMSD5
3.05E+14
3.83E+14
4.02E+14
4.2531E+14
Moderator(Ref
values)
2.930024E+14
3.5120984E+14
3.6725799E+14
3.8851187E+14
Moderator by
WIMSD5
3.06E+14
3.84E+14
4.03E+14
4.2537E+14
**from the previous figure and table it is obvious:-
The total neutron flux of thorium-plutonium oxide increases with
burnup because the macroscopic fission cross section decreases mainly
due to the depletion of fissile nuclides this is a direct consequence of
the constant linear power assumed.
The reference flux results begins from 2.9E+14 to 3.8E+14 which are
near to our WIMSD5 results which begin from 3.04E+14 to 4.2E+14
n/(cm2.sec)
3-The average energy per fission versus burnup
• Energy per fission for (Th-pu)OX:
energy / fission
energy per fission( IAEA-TECDOC-1349 )
energy / fission (The present work)
212
Mev/fission
211
210
209
208
207
206
205
204
203
202
0
10
20
30
40
burnup (GWd/ton)
50
60
**The following table shows the energy per fission of reference
results and that obtained by wimsd5
Burnup(GWD/ton)
0
30
40
60
Energy per fission
(Mev/fission) Ref
values
207.891
205.775
204.411
202.009
Energy per fission
(MEV) by WIMSD5
code
211.439
207.312
205.433
202.34
As shown in the previous figure ,the average energy per fission
decreases with burnup this is due to the change of the fissile nuclides
where the smooth transition from plutonium fissioning to U-233 causes
the decrease in the average energy per fission .this is due to the fact
that the fissile plutonium isotopes release about 200
Mev thermal energy per fission and U-233 only release about 190 Mev
thermal energy per fission.
4-the atom densities of fission products and actinides (atoms/barn x cm)
**From the previous table we can see that there is a change in
the concentration of fission products and actinides and this will
be illustrated in the following notes:
• 1-for plutonium isotopes,Pu-239 decreases with burnup, where its
amount nearly has been burnt up at 60 MWD/kg HM.
• Pu-241 increases and then decreases, this is due to that Pu-241 is
produced via the capture of Pu-240 and on the other hand [Pu-241
is depleted due to its fission.
2-for protactinium and uranium isotopes, the fissile nuclide U-233
increases with burnup due to neutron capture in Th-232 and
subsequent decay of Th-232 to Pa-233 then to U-233 which
contributes more and more to the power.
Pa-233 increases slightly with burnup due to increasing the total
neutron flux.
For U-234,U-235,U-232:U234 also increases with burnup due to
neutron capture in U-233 and capture in Pa-233 and subsequent decay
of Pa-234.
For U-235 it increases slightly due to neutron capture in U-234.
For U-232 it increases because it is formed by (n,2n) reaction
For the concentration of major actinides,Am-243 and Cm-244 are the
most abundant in the fuel.Cm-244 increases with burnup due to the
neutron capture in Am-243 and subsequent decay of Am-244
Am-243 is produced by neutron capture of Pu-242 and subsequent decay of Pu243
5-the absorption and fission cross section of fission products and actinides
conclusion
**we can summarize the conclusion in the following points
1-MCNP5 and WIMSD5 are used for the calculation of a PWR fuel pin taken from
17X17 assembly in pressurized water reactor under specific conditions of its
•
operation.
•
2-pin cell model is carried out for (Th-Pu) OX and many parameters are calculated and the
obtained results are compared with the results announced in the reference. (Potential of
thorium based fuel cycles to constrain plutonium and reduce long lived waste toxicity) and a
good agreement is found between the two results.
3-the parameters calculated in this work are neutron multiplication factor, fluxes, •
average energy per fission, fuel compositions, absorption and fission cross sections
versus burnup which are important parameters in the reactor operation.
• 4-for( Th-Pu) OX, Pu isotopes decreases with burnup .but uranium
isotopes increases with burnup and actinides as Am-243, cm-244
increases with burnup this is explained previously.
Thank you