Neutronics Simulations of 237 Np Targets to Support Safety-Basis and 238 Pu Production Assessment Efforts at the High Flux Isotope Reactor

Transcription

1 Neutronics Simulations of 237 Np Targets to Support Safety-Basis and 238 Pu Production Assessment Efforts at the High Flux Isotope Reactor David Chandler and R. J. Ellis Oak Ridge National Laboratory Nuclear and Emerging Technologies for Space 2015 (NETS) Aerospace Nuclear Science and Technology Albuquerque, New Mexico, USA February 23-26, 2015 ORNL is managed by UT-Battelle for the US Department of Energy

2 Presentation Overview. Brief description of 238 Pu, its production path, and its uses Brief description of the High Flux Isotope Reactor Purpose of safety-basis calculations performed to support target irradiations Neutronics computational toolkit and methods Safety-basis analyses and results 238 Pu production assessment analyses and results Final remarks, summary, and conclusions 2 Chandler, NETS 2015, February 23-26

3 238 Pu is a unique isotope that is used as a heat source in radioisotope thermoelectric generators (RTG). 238 Pu Quick Facts: Half-life of 87.7 years 100% alpha decay Power output of about 560 W/kg Pu-238 U-234 α (He-4) 87.7y ~5.5 MeV 238 Pu Supply Chain History: Multi-Mission Radioisotope Thermoelectric Generator MMRTG (NASA) Produced at the Savannah River Site (SRS) via neutron irradiation in their large reactors until the mid-to-late 1980s Obtained from foreign sources since the SRS reactors shutdown US DOE currently tasked to reestablish domestic production capability A technology demonstration subproject has been initiated to develop and implement the technology required to establish a new domestic supply chain 3 Chandler, NETS 2015, February 23-26

4 The US DOE and NASA have undertaken a program to reestablish a domestic 238 Pu production program. Storage of 237 Np INL Irradiation of NpO 2 /Al pellets ATR at INL and HFIR at ORNL Pu powder PuO 2 LANL Target fabrication at ORNL REDC Power source (i.e., MMRTG) 237 Np 2.14E+06 Y 238 Pu 87.7 Y 238 Np 2.12 D 239 Pu 2.41E+04 Y Robotic rover (i.e., Curiosity) Chemical processing ORNL 4 Chandler, NETS 2015, February 23-26

5 The US DOE and NASA have undertaken a program to reestablish a domestic 238 Pu production program. Storage of 237 Np INL Target fabrication at ORNL Irradiation of NpO 2 /Al pellets ATR at INL and HFIR and ORNL The focus of this presentation is on the target irradiations in the High Flux Isotope 237 Np 2.14E+06 Y 238 Pu 87.7 Y 238 Np 2.12 D 239 Pu 2.41E+04 Y Reactor. Chemical processing ORNL Power source (Multi- Mission Radioisotope Thermoelectric Generator, etc.) Robotic rover (Curiosity, etc.) or satellite 5 Chandler, NETS 2015, February 23-26

6 The High Flux Isotope Reactor is located on the Oak Ridge National Laboratory campus and has served a broad range of science and technology communities since it reached full power in Isotope Production Californium-252 Plutonium-238 Tungsten/Rhenium-188 Selenium-75 Nickel-63 Neutron Scattering Study biology, physics, chemistry, materials science, engineering Cold neutrons to 7 state-of-the-art instruments Thermal neutrons to 8 state-of-the-art instruments Materials Irradiation Fusion energy Fission energy National security Neutron Activation Analysis Nuclear and criminal forensics Impurity analysis Geology Environmental studies 6 Chandler, NETS 2015, February Gamma Irradiation Qualify materials and components for the nuclear industry Characterize material behaviors in a radiation environment

7 HFIR is a versatile 85 MW research reactor. Operates with an average power density of 1.7 MW/L and a peak thermal neutron flux of 2.5x10 15 neutrons/cm 2 -s (highest in the western world) Two fuel elements (inner and outer fuel elements) contain 540 involute-shaped fuel plates Highly enriched uranium (~93 wt% 235 U) fuel in the form of U 3 O 8 -Al cermet in Al clad Pressurized, light-water-cooled, lightwater-moderated, beryllium reflected, flux trap design Reactivity controlled by two concentric poison-bearing control elements Fuel cycles typically vary between 24 and 26 days depending on the experiment loading Cycle 459 scheduled to startup on 2/24/ Chandler, NETS 2015, February 23-26

8 Safety-basis calculations are performed to support experiment irradiations. Will the irradiation experiment increase the probability of occurrence of an accident previously evaluated in the SAR? Will the experiment increase the consequences of an accident previously evaluated in the SAR? Study potential impacts on reactor performance Power tilts, reactivity penalty, neutron fluxes, etc. Time-dependent radionuclide inventories to support post-irradiation examination, storage, transportation, and dose consequence analyses Calculate needed input for follow-on thermalstructural (TS) and thermal-hydraulic (TH) analyses Heat generation (during operation and post-shutdown), Fission product gases (He, Kr, and Xe), and Accumulated fission densities 8 Chandler, NETS 2015, February 23-26

9 The neutronics toolkit includes the MCNP, VESTA, and SCALE code packages. MCNP5 for Monte Carlo-based neutron and photon transport Los Alamos National Laboratory Multi-group neutron fluxes and cross-sections for activation calculations Heat generation rates and fission reaction rates VESTA for depletion calculations Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) Couples MCNP to ORIGEN 2.2 Multi-group binning approach SCALE 6.1, 6.1.2, and for activation and source term analyses Oak Ridge National Laboratory ORIGEN for decay heat, delayed gamma sources, and nuclide inventories CSAS/COUPLE/ORIGEN for cross-section processing and activation calculations Python, FORTRAN, MATLAB, and EXCEL Interlink codes, post-process results, and plot results 9 Chandler, NETS 2015, February 23-26

10 Simplified flow of neutronics calculations. Time-dependent flux MCNP Time-dependent fission power VESTA Time-dependent MCNP inputs ORIGEN 2.2 Time-dependent nuclide inventories KCODE Fixed Source ORIGEN-S H(neutron + FP KE) H(prompt + capture γ) H(delayed γ) Delayed gamma source Post-shutdown inventories H(delayed β) Fission rates Decay heat 10 Chandler, NETS 2015, February 23-26

11 Detailed MCNP model used for neutron and gamma transport calculations. Pellet construction based on ceramic oxide with Al powder Large VXF A A Inner small Vertical Experiment Facility (VXF) 20 vol.% NpO 2 70 vol.% Al 10 vol.% void Pellet stack length is about 20 inches 7 targets loaded in a small VXF Pin 2 Pin 1 19 targets likely to be loaded in a large VXF 11 Chandler, NETS 2015, February Outer small VXF Pin 6 Pin 3 Pin 7 Pin 4 Pin 5 Section A-A

12 Neutron flux profiles. Target X-Y Thermal Flux Distribution HFIR R-Z Thermal Flux Distribution 12 Chandler, NETS 2015, February 23-26

13 Nuclear heat generation rates are calculated at various times into the cycle to support T-H and T-S calculations. Heat generation rates increase with increasing time into the cycle due to the production of the fissile 238 Np and 239 Pu isotopes Follow-on thermal-hydraulic and thermal-structural calculations are performed to demonstrate the pellets melting temperature is never exceeded, the target capsule surface temperatures never exceed the adjacent coolant saturation temperature, and other limits are not reached during bounding events defined in the SAR Heat generation due to: Fission product Neutron Beta + alpha Prompt gamma Capture gamma Delayed gamma Pin 2 Pin 3 Pin 7 Pin 1 Pin 6 Pin 4 Pin 5 Heat Generation Rate (W/g) 13 Chandler, NETS 2015, February Pin 1 Pin 2 Pin 3 Pin Cycle 2 results shown in VXF-15 5 Pin 5 Pin 6 Day 5 Day 10 Day 15 Day 20 EOC Pin 7

14 VESTA-calculated fission powers are converted into fission rates which are integrated over time to estimate fission densities. Fission densities are used to help characterize the pellets swelling properties as a function of irradiation time 238 Np and 239 Pu are produced during irradiation 238 Np reaches equilibrium ~10 days into the cycle 239 Pu continuously increases in concentration 238 Np decays away during outages between cycles VESTA fission power (W/g) VXF-3 End-of-cycle 1 Beginning-of-cycle x fission density (fissions/cm 3 ) Day 10 into 2nd cycle Beginning-of-cycle 2 End-of-cycle 2 red circles: pin 1 material 1 green asterisks: pin 1 material 2 pink diamonds: pin 4 material 9 14 Chandler, NETS 2015, February 23-26

15 Gases are produced in the pellets due to fission events. Fission gases released from the pellets enter the Hefilled plenum region Reduces the thermal conductivity of the gas between the pellet and clad Reduces the ability of the pellet to transfer heat to the clad and out to the coolant He (moles/cm 3 ) 15 Chandler, NETS 2015, February x fissions/cm 3 x Kr (moles/cm 3 ) Xe (moles/cm 3 ) x fissions/cm 3 x x 10-5 Cycle 1 Cycle fissions/cm 3 x 10 19

16 Impacts on reactor core performance. Distance from core midplane (cm) IFE OFE Radius (cm) Relative power profile in fuel elements at beginning-of-cycle Two holders with fully loaded targets have Negligible impact on core relative power profile No statistical change in core reactivity No impact on estimated startup control element position No impact on cycle length More analyses will need to be performed for production level irradiations Impact on core reactivity, Impact on gas production in beryllium reflector, Impact on beam tube neutron fluxes, Impact on pressure vessel embrittlement, etc. 16 Chandler, NETS 2015, February 23-26

17 Modeling and simulation of fully loaded targets in all VXFs for 238 Pu production assessments. Brief model description: Pellets subdivided into 3 equal volume rings (self-shielding in pellet) Pellet stack subdivided into 6 axial regions (axial cosine flux shape) 70 target rods in the 10 ISVXFs 35 target rods in the 5 OSVXFs 114 target rods in the 6 LVXFs Total of 219 target rods modeled A few notes: Other irradiation facilities exist for potential use (RBs, SPBs, etc.) Some VXFs may need to be left vacant for other users, to maintain neutron fluxes to the scattering instruments, etc. 17 Chandler, NETS 2015, February 23-26

18 VESTA and ENDF/B-VII.0 cross-sections used. The targets were irradiated for 8 cycles. No target shuffling, replacement, or rotation modeled. One neutron transport and depletion step per cycle Total: 904 g 800 Pu238 (grams) ISVXFs (x10) OSVXFs (x5) LVXFs (x6) Sum of 21 VXFs ISVXFs: 384 g LVXFs: 365 g OSVXFs: 155 g time (days) 18 Chandler, NETS 2015, February 23-26

19 Irradiating 237 Np produces 238 Pu and other Pu isotopes. Efficiency and product purity are key. Pu238 to Pu ratio ISVXFs (x10) OSVXFs (x5) LVXFs (x6) 18 Percent Conversion: Production efficiency decreases with irradiation time due to feed material consumption More target irradiations (replacing irradiated targets with fresh) increases waste, operator time, etc time (days) 238 Pu Purity ( 238 Pu to Pu ratio): 85% criteria Higher purity the better Purity decreases with irradiation time due to subsequent neutron captures 100x(Pu238(t)/Np237(t=0)) ISVXFs (x10) OSVXFs (x5) LVXFs (x6) time (days) 19 Chandler, NETS 2015, February 23-26

20 Modeling and simulation of 6 cycles with fully loaded targets in all VXFs and target replacement. 13 neutron transport and depletion steps per cycle. Core fuel, control elements, and NpO 2 /Al target depleted as a function of time to assess the impact of fuel depletion, control element withdrawal, and time-varying neutron fluxes on 238 Pu production Target replacement after every two cycles of irradiation A total of 394 pins modeled and 486 materials activated over the 6 cycles OSVXF set 1 ISVXF set 1 LVXF set 1 ISVXF set 2 ISVXF set 3 OSVXF set 2 Cycles 1 and 2 Cycles 3 and 4 Cycles 5 and 6 ISVXF set 1 ISVXF set 2 OSVXF set 1 20 Chandler, NETS 2015, February 23-26

21 A total of about 0.96 kg 238 Pu can be produced in HFIR s permanent beryllium reflector in 6 nominal cycles of irradiation with target replacement ISVXFs (x10) ISVXFs (x10) ISVXFs (x10) OSVXFs (x5) OSVXFs (x5) LVXFs (x6) Sum of 21 VXFs Total: 956 g (per 6 cycles) Pu238 (grams) LVXFs: 293 g (per 6 cycles) time (days) ISVXFs: 172 g (per 2 cycles) OSVXFs: 98 g (per 4 cycles) 21 Chandler, NETS 2015, February 23-26

22 Target replacement keeps the purity levels above 85% and 8 9% conversion is achieved. Pu238 to Pu ratio ISVXFs (x10) ISVXFs (x10) ISVXFs (x10) OSVXFs (x5) OSVXFs (x5) LVXFs (x6) Percent Conversion: ISVXF: ~8.0% conversion in 2 cycles OSVXF: ~9.0% conversion in 4 cycles LSVXF: ~8.5% conversion in 6 cycles ISVXFs (x10) ISVXFs (x10) ISVXFs (x10) OSVXFs (x5) OSVXFs (x5) LVXFs (x6) time (days) 238 Pu Purity ( 238 Pu to Pu ratio): 2, 4, and 6 cycles in the ISVXFs, OSVXFs, and LVXFs keep the purity above 85% 100x(Pu238(t)/Np237(t=0)) time (days) 22 Chandler, NETS 2015, February 23-26

23 A total of about kg 238 Pu can be produced annually in HFIR s permanent beryllium reflector. 238 Pu production per VXF set per cycle (grams) Cycle ISVXF set 1 ISVXF set 1 ISVXF set 1 OSVXF set 1 OSVXF set 2 LVXF Calculated annual 238 Pu production (kg) and estimated PuO 2 production (kg) Cycles/year Pu or PuO Pu PuO Pu PuO 2 ISVXF OSVXF LVXF SUM Chandler, NETS 2015, February 23-26

24 Final remarks, summary, and conclusions. A technology demonstration subproject has been initiated to develop a safe and efficient 238 Pu production infrastructure Neutronics calculations are being performed at HFIR to support the goals of this subproject to Design and qualify target irradiations Assess annual 238 Pu production capabilities It is estimated that kg 238 Pu (~ kg PuO 2 ) can be produced in HFIR s permanent beryllium reflector per year Validation/benchmark studies are needed to confirm results Post-irradiation 237 Np and Pu vector appear to be in good agreement Fission product inventory appears to be over-estimated Studies and experiments are being performed and planned to Evaluate Np and Pu cross-sections and fission yields Assess ENDF/B-VII.0, ENDF/B-VII.1, and JEFF 3.1 cross-section data with the VESTA and SCALE ORIGEN depletion codes 24 Chandler, NETS 2015, February 23-26

25 Thank you. The authors would like to thank C. Bryan, R. W. Hobbs, and R. M. Wham for their support and contributions to this work. This work has been sponsored by NASA s Science Mission Directorate and the US DOE Office of Space and Defense Power Systems. Cold and thermal neutron scattering Materials irradiation Isotope production Neutron activation analysis Gamma irradiation The High Flux Isotope Reactor is located on the Oak Ridge National Laboratory campus. Oak Ridge National Laboratory is managed by UT-Battelle for the US DOE. The High Flux Isotope Reactor is a U.S. DOE Office of Science User Facility. 25 Chandler, NETS 2015, February 23-26