Neutron Activation Analysis NAA

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Thorium Fuel Performance In a Tight LWR Lattice
Paper presented by

• Sayed saeed abdelfattah 2012

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The authors
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
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Introduction
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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

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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

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The aim of the work
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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.

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Methods of calculation
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.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)
Methods of Calculations of pin cell model
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Parameters and dimensions of pin cell model of
thorium-plutonium 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
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The initial atom densities of fuel isotopes
(atoms/cm3)
isotopes Zone 1 Zone 2 Zone 3
Th-232 2.11E22
Pu-238 9.72E18
Pu-239 5.99E20
Pu-240 2.32E20
Pu-241 7.69E19
Pu-242 4.78E19
Cr 8.14E19
Mn
Fe 1.60E20
Ni
Zr 4.37E22
C 2.68E18
H 4.80E22
O 4.41E22 2.40E22
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Results and analysis
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1-K-inf versus burnup for pin cell of (Th-Pu)OX

• k-infinity versus burnup is shown in the
  following graph

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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 TECDOC- 1349 - our present work) -0.002954 0.008571 0.01444 0.02058
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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.

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2-fluxes of fuel,clad,moderator versus burnup

• The following figure shows flux of fuel versus
  burnup

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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) 2.913131E14 3.5005688E14 3.6624260E14 3.8775357E14
Fuel by WIMSD5 3.04E14 3.82E14 4.01E14 4.250E14
Clad(Ref values) 2.925591E14 3.5062147E14 3.6663592E14 3.8785461E14
Clad by WIMSD5 3.05E14 3.83E14 4.02E14 4.2531E14
Moderator(Ref values) 2.930024E14 3.5120984E14 3.6725799E14 3.8851187E14
Moderator by WIMSD5 3.06E14 3.84E14 4.03E14 4.2537E14
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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 2.9E14 to
3.8E14 which are near to our WIMSD5 results
which begin from 3.04E14 to 4.2E14 n/(cm2.sec)
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3-The average energy per fission versus burnup

• Energy per fission for (Th-pu)OX

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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.
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4-the atom densities of fission products and
actinides (atoms/barn x cm)
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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.
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For U-234,U-235,U-232U234 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 Pu-243
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5-the absorption and fission cross section of
fission products and actinides
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conclusion
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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.

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Thank you