Transcript Study of the Feasibility of a Small Scale Transmuter

Investigating the Feasibility of
a Small Scale Transmuter –
Part II
Roger Sit
NCHPS Meeting
March 4-5, 2010
Outline
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Quick Review of Part I
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Preliminary Transmuter Design
Base Cases for Transmutation
Radionuclides to be studied
Activation analyses methodology
Summary of transmutation results for the
different radionuclides
Shielding calculations
Heatload calculations
Conclusions
Preliminary Transmuter Design
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Basic source term
Evaluate material type for best
multiplication/reflection to optimize neutron flux
Evaluate optimum thickness of material
Evaluate optimum size of sphere
Evaluate mesh tally results inside the sphere
Evaluate neutron energy spectrum inside transmuter
by using different moderators and target sizes
Select transmuter base cases to carry out the
transmutation calculations
Neutron Source
RF-driven plasma ion source
Geometry:
26 cm diameter, 28 cm length
Transmuter Design Base Cases
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D-T generator, unmoderated sphere (DT-Unmod): lead
sphere, 25 cm thick, 50 cm inner radius, neutron source
strength of 3E14 n/s
D-T generator, moderated sphere (DT-Mod): Lead
sphere, 25 cm thick, 5cm thick teflon, 45 cm inner radius,
neutron source strength of 3E14 n/s
D-T generator, themalized sphere (DT-Thermalized):
lead sphere, 25 cm thick, 50 cm inner radius filled with
heavy water, neutron source strength of 3E14 n/s
D-D generator, moderated sphere: Lead sphere, 25 cm
thick, 5cm thick teflon, 45 cm inner radius, neutron
source strength of 1E12 n/s
1.0E-09
7.5E-09
5.0E-08
4.1E-07
8.8E-07
1.9E-06
3.9E-06
8.3E-06
1.8E-05
3.7E-05
7.9E-05
1.7E-04
3.5E-04
7.5E-04
1.6E-03
3.4E-03
7.1E-03
1.5E-02
3.2E-02
6.7E-02
1.2E-01
1.7E-01
2.2E-01
3.0E-01
4.1E-01
5.5E-01
7.4E-01
1.0E+00
1.4E+00
1.8E+00
2.5E+00
3.3E+00
4.5E+00
6.1E+00
8.2E+00
1.1E+01
1.5E+01
Neutron Flux (n/cm2-s)
On-Target Neutron Spectra
1.E+11
1.E+10
1.E+09
1.E+08
1.E+07
mDT-u
DT-therm
1.E+06
DT-u
DT-m
1.E+05
DD-m
1.E+04
1.E+03
Energy (MeV)
Radionuclides Studied
Nuclide
Activity
I-129
Tc-99
Cs-137
Sr-90
Am-241
Pu-239
Pu-238
0.032 Ci
1 Ci
1 Ci
1 Ci
1 Ci
1 Ci
1 Ci
Atomic weight
gm/mole
129
99
137
90
241
239
238
Specific act.
Ci/gm
1.60E-04
1.70E-02
98
150
3.2
0.062
17
Mass
gm
6.25E+03
5.88E+01
1.02E-02
6.67E-03
3.13E-01
1.61E+01
5.88E-02
density
g/cc
4.93
11.5
1.87
2.54
13.67
19.84
19.84
# atoms
9.33643E+23
3.57813E+23
4.48533E+19
4.46074E+19
7.80861E+20
4.06398E+22
1.48838E+20
T 1/2
(yr)
1.57E+07
2.13E+05
30.17
28.6
432.2
2.41E+04
87.75
Requirements for Activation
Calculations
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Neutron flux
Neutron energy spectrum
Dominant reactions and the energy
thresholds for these reactions
Nuclear reaction cross sections
EASY-2003, European Activation System, a
software package utilizing FISPACT
Activation Analysis – Fission Products
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Starting activity (1 Ci except for I-129 [0.032 Ci])
Ending activity: NRC 10 CFR 20; Appendix C values (quantities
requiring labeling)
Using the base cases, calculate fluence required to reduce the
target radionuclides to the ending activity level
Iterate on the base cases by increasing the source strengths by
factors of 10 to reach the ending activity in a “reasonable period”
of time (< 100 years)
Evaluate effective half-lives for each flux level
Evaluate activation products (number of radionuclides and total
activity)
Evaluate dose rate of activation products
Evaluate radiotoxicity of activation products (based on ICRP 72
DCFs)
Evaluate “cooling” of activation products (decay down to 1 mR/hr
surface dose rate)
Iodine-129: T1/2 = 1.57E+7 years
Starting: 1.2E+9 Bq Ending: 3.7E+4 Bq
DT-Unmod
DT-Mod
DT-Thermalized
DD-mod
Initial neutron Flux (n/cm2-s)
N/A
1.55E+11
1.83E+11
5.14E+08
Neutron Flux (n/cm2-s)
N/A
1.55E+16
1.83E+14
5.14E+15
Irradiation effective T1/2 (yrs)
N/A
1.67E+00
4.03E+00
2.19E+00
OM flux increase required
N/A
5
3
7
Number of radionuclides generated
N/A
541
103
200
Activation Products (Bq)
N/A
2.29E+15
6.76E+10
7.12E+14
Dose rate (Sv/hr)
N/A
2.15E+06
8.71E+04
7.52E+05
Ingestion dose (Sv)
N/A
4.23E+06
4.43E+05
2.64E+06
Inhalation Dose (Sv)
N/A
6.50E+06
6.08E+05
3.37E+06
Time to decay to 1 mR/hr (yrs)
N/A
~ 900
~ 500
~ 750
Technetium-99: T1/2 = 2.13E+5 years
Starting: 3.7E+10 Bq Ending: 3.7E+6 Bq
DT-Unmod
DT-Mod
DT-Thermalized
DD-mod
Initial neutron Flux (n/cm2-s)
N/A
1.55E+11
1.83E+11
5.14E+08
Neutron Flux (n/cm2-s)
N/A
1.55E+15
1.83E+14
5.14E+14
Irradiation effective T1/2 (yrs)
N/A
2.25E+00
6.86E+00
6.44E+00
OM flux increase required
N/A
4
3
6
Number of radionuclides generated
N/A
227
62
84
Activation Products (Bq)
N/A
1.64E+14
9.72E+13
4.62E+13
Dose rate (Sv/hr)
N/A
8.71E+04
5.33E+04
2.55E+04
Ingestion dose (Sv)
N/A
3.82E+04
2.34E+04
1.12E+04
Inhalation Dose (Sv)
N/A
1.53E+05
9.54E+04
4.55E+04
Time to decay to 1 mR/hr (yrs)
N/A
~ 1.1 E7
~ 11 yrs
~ 20 yrs
Strontium 90: T1/2 = 28.8 years
Starting: 3.7E+10 Bq Ending: 3.7E+3 Bq
DT-Unmod
DT-Mod
DT-Thermalized
DD-mod
Initial neutron Flux (n/cm2-s)
1.23E+11
1.55E+11
1.83E+11
5.14E+08
Neutron Flux (n/cm2-s)
1.23E+17
1.55E+18
1.83E+16
5.14E+16
6.1
26.4
23.9
20.7
6
7
5
8
535
750
315
353
Activation Products (Bq)
4.40E+11
3.15E+12
4.93E+10
1.20E+10
Dose rate (Sv/hr)
3.38E+06
8.06E+07
5.68E+05
1.18E+05
Ingestion dose (Sv)
1.54E+02
4.88E+02
2.27E+02
7.49E+02
Inhalation Dose (Sv)
1.58E+02
7.38E+02
7.61E+02
2.72E+03
~ 1.6E7
~ 1E8
~ 3E7
~ 5E7 yrs
Irradiation effective T1/2 (yrs)
OM flux increase required
Number of radionuclides generated
Time to decay to 1 mR/hr (yrs)
Cesium-137: T1/2 = 30.2 years
Starting: 3.7E+10 Bq Ending: 3.7E+5 Bq
DT-Unmod
DT-Mod
DT-Thermalized
DD-mod
Initial neutron Flux (n/cm2-s)
1.23E+11
1.55E+11
1.83E+11
5.14E+08
Neutron Flux (n/cm2-s)
1.23E+15
1.55E+15
1.83E+15
5.14E+14
Irradiation effective T1/2 (yrs)
193
422
23
29.4
OM flux increase required *
4
4
4
6
386
426
180
161
Number of radionuclides generated
Activation Products (Bq)
9.36E+09
8.30E+09
2.46E+10
8.60E+08
Dose rate (Sv/hr)
3.59E+04
7.78E+04
3.55E+04
1.90E+04
Ingestion dose (Sv)
3.31E+01
4.59E+00
1.69E+01
6.41E-01
Inhalation Dose (Sv)
7.36E+01
4.80E+00
2.18E+01
4.49E-01
~ 7.0E7
~ 2400
Time to decay to 1 mR/hr (yrs)
~ 1.2E5
~ 7.5E7
Activation Analysis – Actinides
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Starting activity (1 Ci )
Ending activity: NRC 10 CFR 20; Appendix C values (quantities
requiring labeling)
Run MCNPX for each base case to calculate on-target flux which
includes fission neutrons added to the spectrum
Using these fission-modified neutron spectra, calculate fluence
required to reduce the target radionuclides to the target activity level
Iterate on the base cases by increasing the source strengths by
factors of 10 to reach a “reasonable time” frame of transmutation (<
100 years)
Evaluate effective half-life as a function of flux
Evaluate activation products (number of radionuclides and total
activity)
Evaluate dose rate of activation products
Evaluate radiotoxicity of activation products (based on ICRP 72
DCFs)
Evaluate “cooling” of activation products (decay down to 1 mR/hr
surface dose rate)
Evaluate amount of other actinides generated
Amercium-241: T1/2 = 432 years
Starting: 3.7E+10 Bq Ending: 37 Bq
DT-Unmod
DT-Mod
DT-Thermalized
DD-mod
Initial neutron Flux (n/cm2-s)
1.26E+11
1.62E+11
1.34E+11
4.37E+08
Neutron Flux (n/cm2-s)
1.26E+16
1.62E+15
1.34E+15
4.37E+15
2.89
0.966
0.418
0.304
5
4
4
7
1066
1012
738
770
Activation Products (Bq)
3.22E+12
1.88E+12
2.85E+12
4.76E+12
Dose rate (Sv/hr)
4.59E+05
3.18E+05
4.79E+05
8.58E+05
Ingestion dose (Sv)
3.46E+03
1.87E+03
2.83E+03
4.20E+03
Inhalation Dose (Sv)
8.57E+03
1.38E+04
4.78E+04
4.15E+04
~ 1E+9
~ 2E+8
~ 1E+10
~ 5E+8
2.55E+05
5.83E+09
1.53E+10
3.11E+10
Irradiation effective T1/2 (yrs)
OM flux increase required
Number of radionuclides generated
Time to decay to 1 mR/hr (yrs)
Actinides Created (Bq)
Plutonium-238: T1/2 = 87.8 years
Starting: 3.7E+10 Bq Ending: 37 Bq
DT-Unmod
DT-Mod
DT-Thermalized
DD-mod
Initial neutron Flux (n/cm2-s)
1.26E+11
1.55E+11
1.59E+11
4.22E+08
Neutron Flux (n/cm2-s)
1.26E+16
1.55E+15
1.59E+15
4.22E+15
1.95
1.36
0.768
0.508
5
4
4
7
1013
983
717
728
Activation Products (Bq)
6.55E+11
3.56E+11
3.99E+11
8.91E+11
Dose rate (Sv/hr)
5.93E+05
3.01E+05
3.00E+05
7.33E+05
Ingestion dose (Sv)
7.46E+02
3.39E+02
3.60E+02
7.37E+02
Inhalation Dose (Sv)
1.82E+03
9.67E+02
1.19E+03
1.97E+03
~ 2E+9
~ 6E+7
~ 1E+09
~ 7E+7
7.25E+03
8.58E+07
8.17E+08
4.70E+11
Irradiation effective T1/2 (yrs)
OM flux increase required
Number of radionuclides generated
Time to decay to 1 mR/hr (yrs)
Actinides Created (Bq)
Plutonium-239: T1/2 = 2.4E+4 years
Starting: 3.7E+10 Bq Ending: 37 Bq
DT-Unmod
DT-Mod
DT-Thermalized
DD-mod
Initial neutron Flux (n/cm2-s)
1.34E+11
1.71E+11
2.40E+11
4.75E+08
Neutron Flux (n/cm2-s)
1.34E+16
1.71E+15
2.40E+15
4.75E+15
1.39
0.869
0.268
0.296
5
4
4
7
1101
1041
744
771
Activation Products (Bq)
2.03E+14
1.34E+14
1.18E+15
2.99E+14
Dose rate (Sv/hr)
6.14E+05
4.20E+05
3.17E+06
9.37E+05
Ingestion dose (Sv)
2.31E+05
1.32E+05
9.98E+05
2.53E+05
Inhalation Dose (Sv)
5.53E+05
5.37E+05
2.41E+06
1.03E+06
~ 1E+9
~ 9E+7
~ 3E+9
~ 1E+8
9.73E+06
1.37E+11
4.31E+11
4.70E+11
Irradiation effective T1/2 (yrs)
OM flux increase required
Number of radionuclides generated
Time to decay to 1 mR/hr (yrs)
Actinides Created (Bq)
Calculate Shielding
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Use ANSI/ANS 6.6.1 concrete composition
with a density of 2.3 g/cc.
Use two variance reduction techniques
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Geometry (splitting and Russian roulette)
Source biasing
Use ICRP 51 photon DCFs
Use NCRP 38 neutron DCFs
Result: need about 7 ft concrete to reduce
dose rate to about 5 mrem/hr at 1 foot
Calculate Heat Load
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Calculate heat load from neutron and photon energy
deposition (collision heating)in material using MCNPX
(0.305 kW)
Calculate heat load from activation products in material
using MCNP coupled with FISPACT (0.0453 kW)
Convert kW to J/hr and then using specific heat capacity
of lead, the resulting heat rise is 0.69 C°/ hr.
In the absence of any type of cooling, the transmuter can
operate 474 hours before reaching lead melting point.
So will require cooling.
Conclusions
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A rigorous calculation methodology for transmutation analyses was
developed by coupling the MCNPX radiation transport code with the
FISPACT activation code
The present neutron source strength of the D-T and D-D neutron
generators is not sufficient to perform transmutation in a reasonable
period of time as defined in this investigation
One single transmuter design is not sufficient to transmute all
radionuclides; (ie, fast neutrons are preferable for actinides, slow
neutrons are preferable for LLFP)
There is no major benefit from using the D-D generator as the
neutron source for a transmutation device
The long-lived fission product radionuclides, Tc-99 and I-129,
behave similarly with regards to transmutation characteristics due to
the fact that they have very similar neutron reaction cross sections
Conclusions
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The short-lived fission products, Cs-137 and Sr-90, behave similarly
with regards to transmutation characteristics due to the fact that they
have very similar neutron reaction cross sections.
This investigation confirms industry opinion that it is not beneficial to
treat short-lived fission products by transmutation
The actinides have behaviors that are very radionuclide specific
because of their complex neutron reaction cross sections.
Transmutation of actinides create more actinides; higher energy
neutron spectrum is advantageous because it creates less activation
products.
Thin targets are more beneficial for long-lived fission products; thick
targets for actinides.
Conclusions
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Radiation protection issues:
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Activation products are extremely “hot”, thousands of Sv/h
Activation products are more radiotoxic for LLFP, less for SFP, and different for
actinides
Significant shielding is required for the transmuter (but not unreasonable)
Cooling is required for the transmuter
The methodology used in this investigation can be applied to other
radionuclides; specifically other long-lived fission products of interest
such as Pd-107, Cs-135, Zr-93, and Se-79
The methodology used in this investigation can be used to analyze
the production of a radionuclide of interest from irradiating a target
radionuclide
Thank You
Questions?