Investigating the Feasibility of a Small Scale Transmuter – Part II Roger Sit NCHPS Meeting March 4-5, 2010

Outline Quick Review of Part I  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 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 Geometry: 26 cm diameter, 28 cm length RF-driven plasma ion source

Transmuter Design Base Cases 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

DT-therm

Radionuclides Studied

Requirements for Activation Calculations 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 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: T 1/2 = 1.57E+7 years Starting: 1.2E+9 Bq Ending: 3.7E+4 Bq DT-UnmodDT-ModDT-Thermalized DD-mod Initial neutron Flux (n/cm 2 -s)N/A1.55E E E+08 Neutron Flux (n/cm 2 -s)N/A1.55E E E+15 Irradiation effective T 1/2 (yrs)N/A1.67E E E+00 OM flux increase requiredN/A53 7 Number of radionuclides generatedN/A Activation Products (Bq)N/A2.29E E E+14 Dose rate (Sv/hr)N/A2.15E E E+05 Ingestion dose (Sv)N/A4.23E E E+06 Inhalation Dose (Sv)N/A6.50E E E+06 Time to decay to 1 mR/hr (yrs)N/A~ 900~ 500 ~ 750

Technetium-99: T 1/2 = 2.13E+5 years Starting: 3.7E+10 Bq Ending: 3.7E+6 Bq DT-UnmodDT-ModDT-Thermalized DD-mod Initial neutron Flux (n/cm 2 -s)N/A1.55E E E+08 Neutron Flux (n/cm 2 -s)N/A1.55E E E+14 Irradiation effective T 1/2 (yrs)N/A2.25E E E+00 OM flux increase requiredN/A43 6 Number of radionuclides generatedN/A Activation Products (Bq)N/A1.64E E E+13 Dose rate (Sv/hr)N/A8.71E E E+04 Ingestion dose (Sv)N/A3.82E E E+04 Inhalation Dose (Sv)N/A1.53E E E+04 Time to decay to 1 mR/hr (yrs)N/A~ 1.1 E7~ 11 yrs ~ 20 yrs

Strontium 90: T 1/2 = 28.8 years Starting: 3.7E+10 Bq Ending: 3.7E+3 Bq DT-UnmodDT-ModDT-Thermalized DD-mod Initial neutron Flux (n/cm 2 -s)1.23E E E E+08 Neutron Flux (n/cm 2 -s)1.23E E E E+16 Irradiation effective T 1/2 (yrs) OM flux increase required675 8 Number of radionuclides generated Activation Products (Bq)4.40E E E E+10 Dose rate (Sv/hr)3.38E E E E+05 Ingestion dose (Sv)1.54E E E E+02 Inhalation Dose (Sv)1.58E E E E+03 Time to decay to 1 mR/hr (yrs)~ 1.6E7~ 1E8~ 3E7 ~ 5E7 yrs

Cesium-137: T 1/2 = 30.2 years Starting: 3.7E+10 Bq Ending: 3.7E+5 Bq DT-UnmodDT-ModDT-Thermalized DD-mod Initial neutron Flux (n/cm 2 -s)1.23E E E E+08 Neutron Flux (n/cm 2 -s)1.23E E E E+14 Irradiation effective T 1/2 (yrs) OM flux increase required *444 6 Number of radionuclides generated Activation Products (Bq)9.36E E E E+08 Dose rate (Sv/hr)3.59E E E E+04 Ingestion dose (Sv)3.31E E E E-01 Inhalation Dose (Sv)7.36E E E E-01 Time to decay to 1 mR/hr (yrs)~ 1.2E5~ 7.5E7~ 7.0E7 ~ 2400

Activation Analysis – Actinides 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: T 1/2 = 432 years Starting: 3.7E+10 Bq Ending: 37 Bq DT-UnmodDT-ModDT-Thermalized DD-mod Initial neutron Flux (n/cm 2 -s)1.26E E E E+08 Neutron Flux (n/cm 2 -s)1.26E E E E+15 Irradiation effective T 1/2 (yrs) OM flux increase required544 7 Number of radionuclides generated Activation Products (Bq)3.22E E E E+12 Dose rate (Sv/hr)4.59E E E E+05 Ingestion dose (Sv)3.46E E E E+03 Inhalation Dose (Sv)8.57E E E E+04 Time to decay to 1 mR/hr (yrs)~ 1E+9~ 2E+8~ 1E+10 ~ 5E+8 Actinides Created (Bq)2.55E E E E+10

Plutonium-238: T 1/2 = 87.8 years Starting: 3.7E+10 Bq Ending: 37 Bq DT-UnmodDT-ModDT-Thermalized DD-mod Initial neutron Flux (n/cm 2 -s)1.26E E E E+08 Neutron Flux (n/cm 2 -s)1.26E E E E+15 Irradiation effective T 1/2 (yrs) OM flux increase required544 7 Number of radionuclides generated Activation Products (Bq)6.55E E E E+11 Dose rate (Sv/hr)5.93E E E E+05 Ingestion dose (Sv)7.46E E E E+02 Inhalation Dose (Sv)1.82E E E E+03 Time to decay to 1 mR/hr (yrs)~ 2E+9~ 6E+7~ 1E+09 ~ 7E+7 Actinides Created (Bq)7.25E E E E+11

Plutonium-239: T 1/2 = 2.4E+4 years Starting: 3.7E+10 Bq Ending: 37 Bq DT-UnmodDT-ModDT-Thermalized DD-mod Initial neutron Flux (n/cm 2 -s)1.34E E E E+08 Neutron Flux (n/cm 2 -s)1.34E E E E+15 Irradiation effective T 1/2 (yrs) OM flux increase required544 7 Number of radionuclides generated Activation Products (Bq)2.03E E E E+14 Dose rate (Sv/hr)6.14E E E E+05 Ingestion dose (Sv)2.31E E E E+05 Inhalation Dose (Sv)5.53E E E E+06 Time to decay to 1 mR/hr (yrs)~ 1E+9~ 9E+7~ 3E+9 ~ 1E+8 Actinides Created (Bq)9.73E E E E+11

Calculate Shielding Use ANSI/ANS concrete composition with a density of 2.3 g/cc. Use two variance reduction techniques  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 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 ( 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 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 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 Radiation protection issues:  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?