h. narrog - fast chloride mcfr vs. moderated fluoride msr
TRANSCRIPT
Holger Narrog 23.10.2013
Fast Chloride MCFR vs. Moderated Fluoride MSR
Conclusion: A fast chloride molten salt reactor (MCFR) surpasses a moderated fluoride molten salt reactor significantly. The main difference is the graphite moderator that fills 90% of the reactor vessel of the moderated reactor, reduces the power density, and requires regular replacement. The fast chloride reactor is by far more simple and more compact. Another issue is the internal reprocessing unit. The internal reprocessing is very complex, due to a decay heat that is several kW/g! for fission products fissioned less than a day ago. It requires a very strong cooling if feasible at all. Even if it is feasible it can blow-up any cost calculation. A fast chloride reactor can be operated with a high share of fission products in the fuel and hence requires only degassing plating out noble metal fission products and a replacement of the fuel.
1. Nuclear Evaluation Fast Chloride vs. Moderated Fluoride MSR
Ref. (69)
In opposite to moderated reactors the fast neutron spectrum reactor can use nearly all actinides as 236U, 240Pu, 242Pu as fuel. The build-up of long living high level radioactive waste is by far smaller.
The capture cross sections become very small in the fast neutron spectrum. In the consequence fast breeders can be run with a significant higher fission product concentration as moderated reactors. Actually Fast sodium cooled reactors achieve burn-ups of about 155MWd/Kg hm FBR vs. 55MWd/Kg hm for LWR.
Conclusion: A fast neutron spectrum is an advantage in favor of the fast reactor.
2. Salt Property Evaluation Fast Chloride vs. Moderated Fluoride MSR
Salt Properties 650°C
Unit Moderated Molten Fluoride Salt Reactor
Fast Molten Chloride Salt Reactor
Salt Composition Mol% 71.5%LiF, 15.8%BeF2, 11.8%ThF4, 0.4%UF4,
0.5% Fission Product Chlorides
43% NaCl, 23% KCl, 25.5%UCl3, 4.5% PuCl3, 4% Fission Prod. chlorides
Fuel Costs $/MWh Reprocessing is by far the most significant cost
Reprocessing is by far the most significant cost
Liquidus Temperature
C° 525°C 525-538°C
Heat Capacity cv KJ/m3/K 5038 2780
Density Kg/m3 3250 3230
Heat Conductivity W/m/K 1.1 0.69
Dynamic Viscosity Kg/m/s 0.0071 0.0030
Corrosivity The Fluoride salts are less corrosive. It is expected that nickel based materials work
The chloride salts are more corrosive and require more expensive materials as Mo-TZM.
The data for the chloride salts are based on the study Nuclear Power Plant of the Future and ref. 23
Conclusion: The salt properties of the fluoride salts are due to better corrosion properties and a higher cv very much in favor of a Moderated Fluoride Salt Reactor.
2b. Nuclear Reactions of the Salts used According to the report Reaktorsicherheit und Stahlenschutz in Baden-Württemberg (2000), (ref.
371) the yearly tritium emissions of a PWR are about 30 GBq/yr (Philippsburg II) to 70 GBq/yr (GKN I) in Germany.
The license allows for example the German nuclear power plant Philipsburg II to emit 48000 GBq/yr ( 4.8 x 10 13) of tritium. According to the IAEA Handbook of Nuclear Data for Safeguards ref. 6 the share of 3H is 0.01% for 238U, 0.0142% for 239U and 0,0141% for 241U per fission.
The total yearly tritium production from fission is calculated for a 4444 MWth reactor to 55g or 1.96*1015 Bq/yr. Most of the tritium will form TCl (chloride), TF (fluoride) even if the reactor is run
underfluorinated. Some of the tritium will remain as 3H2. It is assumed that 90% of the tritium and 99% of the TCl/TF is extracted in the degassing. It is further assumed that the very most of the remaining tritium is diffused thru the walls of the reactor, the heat exchangers and emitted in the reactor building. This tritium is absorbed in the charcoal filters of the reactor building. It is estimated in accordance with the (ORNL-4541 ref. 42) MSR project of the 60ies that 0.1% of the tritium is passing the primary and secondary heat exchanger and emitted via the condenser in the cooling water. That means a tritium emission of 2 GBq is well within the tritium emissions of current PWR reactors. In a fluoride reactor using enriched lithium (99% 7Li) the 7Li reaction will take place.
The tritium creation will be x-fold the tritium creation of a chloride salt reactor. It might become an issue for a fluoride salt reactor. A lithium enrichment is necessary anyway. Another topic is the creation of radioactive chlorine in the MCFR. It is due to the excellent hard neutron spectra not necessary to enrich the chlorine. During the operation the following neutron induced reactions occur. 35Cl + n -> 36Cl 75.8% of the chlorine is
35Cl. The
36Cl is radioactive with a half-time of 301.000yr.
35Cl(n,α) ->32P[β-[14.2days] -> 32S Sulphur is corrosive 36Cl + n -> 37Cl (stable). The reaction is welcome. 37Cl + n -> 2n-> 36Cl The 36Cl is radioactive with a half-time of 301.000yr. 37Cl(n,α) -> 34P(β-[12.34s] ->34S Sulphur is corrosive 37Cl + n -> 38Cl -----37.24min 38Ar (stable) In the MCSFR it is foreseen to reuse the salts in the reactor. As the salts needs to be treated protected from the atmosphere to avoid impurities the radioactivity should not increase the costs significantly. The creation of radioactive sulphur in the chlorine salt is (ref. 46) 1/8 of the oxygen creation 19F[n,α] -> 16O reaction in the fluoride system. The tritium removal at the MSFR requires more efforts and costs than the handling of radioactive chlorine in the reactor. The tritium removal at the MSFR requires more efforts and costs than the handling of radioactive chlorine in the reactor. The tritium removal at the MSR requires more efforts and costs than the handling of radioactive chlorine in the reactor.
Tritium Generation Table AD333
Tritium generation/Kg of fissioned actinides: 0.0351 g %/Fission
Tritium generation/year (g): 55 g 238U 0.01 Ref. 6
Tritium generation/day/Bq: 5.3722E+12 Bq 239Pu 0.0142 Ref. 6
Tritium generation/yr/Bq: 1.9609E+15 Bq 241Pu 0.0141 Ref. 6
Average 0.14 Ref. 6
Tritium Generation by Fission
3. Structure Material
Data 700°C if available Hastelloy N (ref. 161) Molybdenum TZM (ref. 168)
Density (700°C) 8860 Kg/m3 10070 Kg/m3 (ref. 264)
Thermal conductivity(700°C): 23,6 W/mK 112W/m/K (acc. To ref 264)
Young Modulus Gpa (700°C): 169,6 217.9 (acc. To ref. 164)
Mechanical yield
strength/Mpa (700°C):217 700 (acc. To ref. 168)
Accepted mechanical stress in
this design Mpa (700°C):
80 definition based on the
data above
250 definition based on the
data above
Corrosion resistance against
Flibe salts:
acc. To (ref. 61) page 41 less
than 0.025mm/yr at
temperatures of more than
700°C
No Corrosion at 1100°C (ref.
167)
Corrosion resistance chloride
salts:
acc. To (ref. 61) page 42
1.1mm/yr at 850°C No Corrosion expected
Manufacturing: Very Good Challenging, limited experience
Table D Modified Hastelloy N is less expensive and easier to weld than Mo-TZM. That’s why it was developed for fluoride MSR. Two unknown issues with Hastelloy N did surface, one was corrosion induced by the fission product tellurium and the other was irradiation damage caused by (n,alpha) reactions in nickel and boron contaminants (ref. 203). The n, alpha reaction is not observed in the radiation tests with molybdenum alloys. Mo-TZM/TZC allows much higher temperatures, has superior corrosion properties and by far superior mechanical properties. Conclusion: The possibility to use Ni-based materials is an advantage in favor of the moderated fluoride salt reactor even when it limit its potential.
4. The Reactor of the Moderated vs. Fast Chloride MSR
Reactor Design
Unit Moderated Fluoride Molten Salt Reactor
Fast Chloride Molten Salt Reactor
Power Density 24 MW/m3 (core), Peak 70MW/m3 (ref. 42)
492 MW/m3 (fissile zone), 22 MW/m3 (fertile zone)
T(outlet) °C 740°C 800°C
Total Height m 7.57 5.74
Total Diameter m 9.17 4.13
Total zylindric volume
m3 500 77
Complexity Medium complex Inserts made of graphite
No inserts, no complexity
Maintenance The graphite core needs replacement about every 2-4years
No regular maintenance required
Risks Graphite – Salt – Hastelloy materials can cause electrolytical carburization and corrosion ORNL-3626 (ref. 34)
The moderated reactor requires a graphite moderator. The moderator requires nearly 90% of the space in the core. A high neutron flux damages the graphite. It limits the
power density and requires a replacement every 4 years at a peak power density of 70MW/m3
(ref. 42). The combination of graphite and nickel or molybdenum based materials might cause electrolytical carburization and corrosion. The mean thermal expansion coefficient of Hastelloy N is 1,47E-05 vs. 2,50E-06 for graphite that creates some design headaches. Conclusion: Even if molybdenum TZM is more expensive to manufacture the total fast chloride reactor will be by far less expensive. This is a main advantage of the Fast Chloride Molten Reactor.
5. The Primary Heat Exchangers
The heat exchangers of the
MSR are designed
conventionally as tube and
shell heat exchangers.
The new MSR concepts are
based on using compact heat
exchanger designs as micro
channel types as HEATRIX,
plate design types and
others.
The advantages are that
there is less fissile material in
the heat exchangers.
The share of delayed
neutrons in the circuit is
lower.
It is more compact.
6. The Power Plant Due to the bigger reactor the moderated fluoride reactor has a bigger reactor building. It requires space and equipment like a crane to replace the graphite once every 2 – 4 years.
Ref. 42
The reactor outlet temperature of the fast chloride reactor is 40° higher which compensates slightly the lower cv of the chloride salts. It allows as well a better thermal efficiency. The real advantage is that the fast chloride salt reactor structure material Mo-TZM gives room for development to significantly higher operation temperatures.
Ref. 42
Both concepts require a 3 - circuit system with its complexity. It is a major disadvantage of most of the MSR concepts. The usage of Mo-TZM in the fast chloride reactor allows the usage of liquid bismuth lead as intermediate coolant. The advantage is the very low m.p. of 125°C. Conclusion: The usage of Mo-TZM gives the Fast Chloride Molten Salt Reactor the potential of using higher temperatures and intermediate coolants as Bi/Pb with a low melting point of 125°C.
7. The Reprocessing Unit All MSR concepts need a regular degassing of the fuel. In the moderated MSR it is done by a helium bubbling. It seems a simple method to get out the gaseous fission products. In the fast MCFR it is done by a 10mbar vacuum distillation at 950°C. The method is more complex but seems suitable to extract about 35% plus due to further decay of fp to such with a low bp. in total 40% of the fission products. Another 20% of the fission products are extracted as metal by gravity.
The fast MCFR does not have a complete and complex reprocessing unit. A fraction of the fuel is taken out and shipped to an external reprocessing unit about 2 years after it is taken from the reactor. The moderated MSR includes a complete and very complex reprocessing unit.
Ref. 42
Corrosion risk!
Corrosion risk! Cooling of the fission
products
Multistage
Processes
The degassing of the fuel takes place in a
helium bubbling process. Gaseous fission
products as Xe, Kr, I are separated in a
helium gas flow.
The system is simple and would not create
technical or economic challenges.
Conclusion: The gas extraction system is required in all MSR reactors to extract gasses and perhaps even noble metals. The complete reprocessing unit with its challenges of the heat creation and its complexity is part of the political promise to avoid waste. It is complex if feasible and would blow up any cost calculation.
Calculation Moderated MSR:
1970 MSR Recalculation 2000 MW Version of this reactor
1 feet/m: 0,3048 1 in/m: 0,0254
Core Diameter: 18ft 5,49 Acc. To ref. Core Diameter/m: 7,22
Total Diameter/m: 6,77 Estimate Total Diameter 9,24
Core Height: 13ft 3,96 Acc. To ref. Core Height (ornl 4541)/m: 5,23
Total Height Reactor/m 6,41 measured Total Height Reactor/m: 8,23
Core Volume/m3: 93,63 Calculated Core Volume/m3 214
Total Mass Graphite/to 304 Acc. To ref. Total Mass Graphite/to 780
Density Graphit/Kg/m3 1750 Acc. To ref 195 Density Graphit/Kg/m3 1750
Volume Graphit/m3 174 Acc. To ref. Volume Graphit/m3 445,50
Side Reflector Thickness/m 0,76 Measured Side Reflector Thickness/m 0,76
Side Reflector Volume/m3 59,07 Calculated from estimatesSide Reflector Volume/m3 99,60
Top/Bottom Reflector estimate av./m 0.3 - 0.73m Average est.: 0,55 Top/Bottom Reflector av./m 1,2
Top/Bottom reflector Volume: 33,69 est.from calc. Top/Bottom reflector Volume: 160,85
Graphit in core/m3: 80,96 est.from calc. Graphit in core/m3: 185,05
Total cylindric volume/m3: 247,08 Calculated Total cylindric volume/m3: 551,59
Electrical Power/MW 1000 Acc. To ref. Electrical Power/MW 2000
Efficiency% 44,40% Acc. To ref. Efficiency% 44,00%
Thermal Power/MW 2252,25 Calculated Thermal Power/MW 4545,45
Average Power density W/cm3: 22,00 Acc. To ref. Average Power density W/cm3: 21,24
Specific heat capacity KJ/Kg/K: 1,55 ref. (64) Specific heat capacity KJ/Kg/K: 1,55
T inlet/°C: 600 T outlet/°C 740 T inlet/°C: 600 T outlet/°C 740
Mass Flow/Kg/s: 10379 calculated Mass Flow/Kg/s: 20947
Density/Kg/L: 3,25 ref. (64) Density/Kg/L: 3,25
Volume Flow/m3/s: 3,19 calculated Volume Flow/m3/s: 6,45
velocity in the Reactor/m/s: 2,6 Acc. To ref. velocity in the Reactor/m/s: 3
Flow Area/m2: 1,23 calculated Flow Area/m2: 2,15
Ref.: (24)
1 inch fuel channels acc. to ORNL – 3626
Ref.: (24)
References:
1. NGATLAS Atlas of neutron capture cross sections Prepared by J.Kopecky, Contributions by J.-Ch.Sublet, J.A.Simpson, R.A.Forrest and D.Nierop Web Design and Plots by V.Zerkin (IAEA, Vienna 1997)
6. IAEA INDC(NDS)-0534 Distr. SQ Handbook of Nuclear Data for Safeguards: Database Extensions, August 2008, A.L. Nichols, D.L. Aldama, M. Verpelli
23. INL Flour Chlor ANL-6792 Molten Salt Fast Reactors 24. Reactors with Molten Salts: Options and Missions Charles W. Forsberg Oak Ridge National Laboratory*
File Name: France.MoltenSalt.2004 Draft Date: August 3, 2004
Frederic Joliot & Otto Han Summer School on Nuclear Reactors “Physics, Fuels, and Systems”
Cadarache, France August 25–September 3, 2004
41. ORNL-TM-3832 Design Studies of a Molten-Salt Reactor Demonstration Plant, E.S. Bettis, L. G.
Alexander, H. L. Watts, June 1972
42 ORNL 4541 Conceptional Design Study of a Single Fluid Molten Salt Breeder Reactor, Molten Salt
Reactor Program Staff, compiled and edited by Roy C. Robertson, June 1971
46. Transactions Advanced Reactors, ENS Conference 2012, ISBN 978-92-95064-14-0
61. INL/EXT-10-18297 Engineering Database of Liquid Salt Thermophysical and Thermochemical
Properties. Manohar S. Sohal, Matthias A. Ebner, Piyush Sabharwall, Phil Sharpe, March2010
69. Assessment of LIquid Salts for Innovative Applications, ALISIA DELIVERABLE (D-50) REVIEW REPORT ON LIQUID SALTS FOR VARIOUS APPLICATIONS Lead authors: O. Benes, C. Cabet, S. Delpech, P. Hosnedl, V. Ignatiev, R. Konings, D. Lecarpentier, O. Matal, E. Merle-Lucotte, C. Renault, J. Uhlir February 20, 2009 (version V4) Date of issue
161. HASTELLOY® N alloyH-2052B ©2002, by Haynes International, Inc.
164. http://aries.ucsd.edu/LIB/PROPS/PANOS/moa.html MOLYBDENUM AND ITS ALLOYS
167 ORNL-TM-2724, Compatibility of Molybdenum base Alloy TZM .. at 1100°C, J. W. Koger, A.P.
168 Molybdenum Material Properties and applications, Plansee Company, 530 DE 05.04 (3000) RWF
195. On Graphite Transformations at High Temperature and Pressure Induced by Absorption of the LHC Beam
LHC Project Note 78/97, Jan M. Zazula. 203. Molten salt reactors: A new beginning for an old idea, David LeBlanc, Nuclear Engineering and Design,
doi:10.1016/j.nucengdes.2009.12.033
371. Reaktorsicherheit und Strahlenschutz in Baden Württemberg, Ministerium für Umwelt und Verkehr des
Landes Baden-Württemberg, Abteilung Reaktorsicherheit, Umweltradioaktivität, November 2000