slides - University of California, Berkeley

Advanced Reactors and Fuel CycleGroup Research at UC BerkeleyDepartment of Nuclear EngineeringEhud GreenspanFrancesco Ganda (Ph.D student)Max Fratoni (Ph.D student)Florent Heidet (Ph.D student)Tommy Cisnero (Ph.D student)Mathieu Hursin (Ph.D student; Prof. Downar)Kevin Cramer (Ph.D student; LLNL)Steve Mullet (M.Sc student)Alessandro Piazza (Visiting student)Filippo Bartoloni (Visiting student)Galit Weidenfeld (Visiting researcher)October 2008

Ongoing & recent research projects1. Use of Hydride Fuel for Improved LWR CoreDesigns. With MIT & Westinghouse. Funded by theDOE NERI Program from 9/02 to 2/06.2. Feasibility of Recycling Plutonium and MinorActinides in Light Water Reactors Using HydrideFuel. With MIT & ANL. Funded by the DOE NERIProgram from 3/06 to 12/08.3. RBWR4. Lead Cooled Fast Reactors Generation-IV Program.Following 3-year NERI project with ANL, LLNL,Westinghouse, KAERI, CRIEPI.5. Solid-Core Heat-Pipe Nuclear Battery Type ReactorModule. DOE NEER contract from 7/1/2005 through6/30/2008.

Ongoing & recent research projects (2)6. Support of the Advanced High Temperature Reactor(AHTR) Program (GEN-IV). Led by Prof. Peterson.7. Independent evaluation of a Deep-Burn GT-MHR.Funded by DOE8. Molten Salt Reactors for TRU Transmutation. Has beenfunded by AFCI program via LANL9. Highly Compact Accelerator-Driven Subcritical Assemblyfor Medical and Industrial Applications. with Profs. Vujic(PI), Kastenberg and Leung. Funded by the DOE NEERProgram from 6/03 to 6/06.10.Development of the SWAN-SCALE Code for k effMaximization and Critical Mass Minimization. Funded byOak Ridge National Laboratory from 1/98 through 9/06.

Researchers• Francesco Ganda (Ph.D student) 1, 2, 3, 9• Max Fratoni (Ph.D student) 1, 5, 6, 7, 8• Florent Heidet (Ph.D student) 4• Tommy Cisnero (Ph.D student) 7• Mathieu Hursin (Ph.D student; Prof. Downar) Methods• Kevin Cramer (Ph.D student; LLNL) LIFE• Steve Mullet (M.Sc student) 5• Alessandro Piazza (Visiting student) 1• Filippo Bartoloni (Visiting student) 3• Galit Weidenfeld (Visiting researcher) CFD of ENHS

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN_IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Use of hydride fuel for BWR can1. Greatly simplify the fuel bundle design– Eliminate water rods and partial length fuel rods– Reduce number of enrichment levels– Increase number of fuel rods per unit core volume: 96/71=1.35Reference oxide fuel bundlevery heterogeneousHydride fuel bundlenearly uniform

Use of hydride fuel for BWR can (2)0.92 1.03 1.07 1.09 1.08 1.10 1.08 1.04 0.921.03 0.78 0.98 0.78 0.73 0.80 0.99 0.79 1.041.06 0.97 0.76 0.91 0.98 0.99 0.79 0.99 1.071.09 0.79 0.91 1.07 0.98 0.80 1.101.08 0.73 0.99 0.98 0.73 1.091.10 0.80 0.98 1.07 0.91 0.79 1.101.08 0.99 0.79 0.98 0.98 0.91 0.75 0.97 1.071.04 0.79 0.99 0.80 0.72 0.79 0.97 0.78 1.030.92 1.04 1.08 1.10 1.09 1.09 1.07 1.03 0.91BOL pin –wise power distribution in reference OxF bundle1.04 1.02 1.02 1.01 1.01 1.00 1.01 1.02 1.02 1.051.02 1.02 1.03 1.00 0.98 0.98 1.00 1.03 1.02 1.031.01 1.03 CR 1.02 0.97 0.97 1.02 CR 1.03 1.021.00 1.00 1.01 0.98 0.96 0.97 0.99 1.01 1.00 1.011.00 0.98 0.97 0.96 0.96 0.95 0.97 0.97 0.97 1.011.00 0.97 0.97 0.96 0.95 0.96 0.96 0.97 0.98 1.001.00 0.99 1.01 0.98 0.96 0.97 0.98 1.02 1.00 1.011.01 1.03 CR 1.01 0.97 0.97 1.02 CR 1.04 1.031.01 1.00 1.02 0.99 0.98 0.98 1.00 1.03 1.01 1.031.04 1.02 1.01 1.00 1.00 0.99 1.01 1.02 1.02 1.05BOL pin - wise power distribution in HyF with IFBA2. Increase core power density by ~ 40%– Increase reactor power level by ~40% (ABWRII?)– Reduce core height (volume) by ~40% (ESBWR?)3. Reduce COE by up to 20% as compared to oxides

Use of hydride fuel for PWRBOL Spectrum (top) and effect of increase in fuel temperature (bottom)0.250.25Neutron flux per unit lethargy0.20.150.10.05Neutron flux per unit lethargy0.20.15UO 2U-ZrH 1.60.10.05010 -2 10 0 10 2 10 4 10 6Neutron Energy (eV)010 -2 10 0 10 2 10 4 10 6Neutron Energy (eV)332.52.5MostlyDopplerDiff. in Flux (Pert.-Nom.)3.5 x 10-3 Neutron Energy (eV)21.510.50Diff. in Flux (Pert.-Nom.)3.5 x 10-3 Neutron Energy (eV)21.510.50Mostlyspectralshift-0.5-0.5-1-1-1.510 -2 10 0 10 2 10 4 10 6-1.510 -2 10 0 10 2 10 4 10 6

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN_IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Use of hydride fuel for PWR cansignificantly improve Pu (MA) recycling ability• Compared to MOX fuel of same dimensions and cycle length,use of PuH 2 -U-ZrH 1.6 fuel offers:- 74% of the needed Pu inventory- 100% higher burnup: 103 vs. 50 GWD/TiHM- Larger fractional transmutation: 50% vs. 24% fraction of Pu- Worse Pu quality: 44% fissile isotopes vs. 63%- Larger MA/Pu ratio: 13.25 % vs. 6.76 %- Stronger neutron source intensity and decay heat per gm Pu- But lower neutron source intensity and decay heat per fuelassembly• Might be able to burn ~85% of own Pu in 2 cycles !

Transmutation performance with multi-recyclingNo fertile fuel70%TRU destruction fraction60%50%40%30%20%10%0%Pu+recycled UPu+Np+recycled UTRU+recycled U0 10 20 30 40Recycle Number• Pu in PuH 2 -ZrH 2 can be recycled un-limited # of times; MOX, on theother hand, can be recycling only up to 2-3 times• Pu-Np in hydride fuel can be recycled at least 6 times• TRU in hydride fuel can be recycled at least 2-3 timesWhat limits the number of recycling is the large void coefficient of reactivity

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Organizations/researchers participated in theENHS NERI projectE. Greenspan (PI), A. Barak, D. Barnes, M. Milosovic’, D. Saphier, Z. Shayer, H. Shimada and S. WangUniversity of California, BerkeleyN. W. Brown (co-PI), L. Fischer and Q. HussainLawrence Livermore National LaboratoryD. C. Wade (co-PI), E. Feldman, K. Grimm, R. Hill, J. J. Sienicki and T. SofuArgonne National LaboratoryM. D. Carelli (co-PI), L. Conway and M. DzodzoWestinghouse Electric Company Science & Technology DepartmentYeong Il Kim (co-PI) and Ser Gi HongKorea Atomic Energy Research Institute (KAERI)Soon Heung Chang (co-PI) and Kwang Gu LeeKorea Advanced Institute for Science and Technology (KAIST)Il Soon Hwang (co-PI), Byung Gi Park and Seung Ho JeongSeoul National UniversityI. Kinoshita (co-PI), A. Minato, Y. Nishi and N. UedaCentral Research Institute of Electric Power Industry (CRIEPI)

ENHS reactor layout30m27m17.625mSchematic vertical cut through the ENHS reactor6.94m (I.D.)8m3.64m (O.D; t=0.05)Steam generatorsUnderground siloReactor poolENHS module3m2m3m2mReactor Vessel AirCooling System (RVACS)Cross Section of StackNumber of Stacks = 4Seismic isolatorsReplaceableReactor module• no pumps• no pipes• no valves• factory fueled• weld-sealed• underground silo• >20 years core• no fueling on site• Module is replaced• shipping cask• no DHRS butRVACS

ENHS reactor layout (2)Expanded view of the ENHS reactor(not to scale)Steam generatorSecondary coolantPrimary coolantHeat exchangerPeripheral control assemblyCentral control assemblyCore

Highlights• 2001 – ENHS type reactors selected as one of 6categories of GEN-IV reactors (LFR)– Only new reactor concept– Only concept developed under NERI to get to GEN-IV– Has a number of novel features, including• Once for life core; no refueling hardware on site• No blanket elements and no access to neutrons• Nearly zero burnup reactivity swing (very little excess reactivity)• Natural circulation cooling; no pumps/valves• Simple to operate; autonomous load following capability• Deterministically safe

DOE adopted ENHS type reactors asone of 6 types of GEN-IV reactors“The LFR battery (like the ENHS reactor) is• a small factory-built turnkey plant• operating on a closed fuel cycle with• very long refueling interval (15 to 20 years) cassette core orreplaceable reactor module• meet market opportunities for electricity production on smallgrids, and for developing countries who may not wish todeploy an indigenous fuel cycle infrastructure to support theirnuclear energy systems• The battery system is designed for distributed generation ofelectricity and other energy products, including hydrogen andpotable water”

Highlights (2)2006 – One of the 7 goals of GNEP is: “Small scalereactors designed for the needs of developingcountries” …The U.S. will also encourage the GNEPconsortium to pursue the ultimate goal of developing anddeploying a small scale reactor that utilizes the same nuclearfuel for the lifetime of the reactor.”??? What reactor type??? IRIS. Particle fuel LWR, LFR???!!! Should be LFR !!!

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Heat-pipe version of ENHS (HP-ENHS)• Derives features from the ENHS nuclear battery typereactor concept and from the SAFE-400 space nuclearpower reactor concept• Fuel rods and heat pipes (HP) are horizontally oriented;there are 2 HP’s per 3 fuel rods – one HP serves the lefthalf and the other HP the right half of the core• Solid core; fission energy is transported out from the coreby HP’s that transfer this energy to a coolant flowing (bynatural circulation) through heat exchangers formed fromboth sides of the core by the HP’s that extend beyond theaxial reflector

HP-ENHS core schematicsIHXGasPlenumActive CoreGasPlenumIHXHeatPipesHeatPipes

HP-ENHS overall layout

Unique features of the HP-ENHS• Once-for-life core (> 20 EFPY) along with sustainability –fissile inventory is preserved• High temperature heat supply – secondary coolant outlettemperature > 800 o C; efficiency could be ~50%• Superb safety – In case of loss of secondary coolant, HP’sdeliver decay heat to the outer vessel wall from where it isremoved by, say, RVACS• No positive void reactivity coefficient (typical to fastspectrum reactors); loss of secondary coolant has anegative reactivity feedback

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Pebble Bed-Advanced HighTemperature Reactor (PB-AHTR)• TRISO fuel particles dispersed in 6 cm diameter graphite pebbles• 2LiF-BeF 2 coolant• Inlet/outlet coolant temperature 600 o C/700 o C• Upward pebble motion• Core diameter 6.8 m• Outer reflector diameter 9.0 m• Vessel outer diameter 16.0 m• Effective core height 6.4 m• Core power density 10.2 MW/m 3• Total core power 2,400 MWthDesigned by Prof. Peterson et al.

PB-AHTR: attainable burnup andreactivity coefficientsPB-AHTR can be designed to attainburnup up to ~130 GWd/tHM using10% enriched uranium while allreactivity coefficients remain negativeMaximum attainable burnup (GWd/tHM) as afunction of C/HM and fuel kernel diameterReference design:425 µm fuel kernel12.5% TRISOs packing factor127 GWd/tHM (664 EFPD)Feedback mechanismFuel temperatureCoolant temperatureModerator and fuel temperatureModerator and coolanttemperatureValue-3.85 pcm/K-0.34 pcm/K-4.18 pcm/K-0.84 pcm/KCoolant temperature reactivity feedback (pcm/K)as a function of C/HM and fuel kernel diameter

High temperature reactors comparison• PB-AHTR maximum discharge burnup is very similar to that of theother three design options for high temperature reactors• Compared to the PBMR, the PB-AHTR can operate at higher powerdensity, larger total core power and therefore lower leakage probability• The power generated per pebble for the PB-AHTR is ~2.5 times thatfor the PBMRFeature PB-AHTR PBMR LS-VHTR VHTRCoolant Flibe He Flibe HeTotal power (MWth) 2,400 600 2,400 600Power density (MW/m 3 ) 10.2 6.6 10.2 6.6Leakage probability (%) 3 12 3 7Fuel kernels packing factor (%) 12.5 5.0 15.0 11.0C/HM 363 960 846 1033Specific HM inventory (kg/MWth) 5.23 3.23 4.07 4.59Burnup (GWd/tHM) 127 126 131 125Fuel residence time (EFPD) 664 410 525 567Energy generated per pebble (MWd) 1.27 0.51 - -

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Deep Burn MHR• DB-MHR incineratesplutonium and minoractinides reachingburnups of ~600GWd/MT• High temperatures can beused for increasedthermal efficiency orprocess heat (H 2 )• However plutonium fuelcauses large powerpeaking factors that mustbe mitigated

Coarse resolution of powerdistribution in core

Research objectives• Quantify and reduce power peaking in DB-MHR through- Fuel Shuffling Schemes- High Density Graphite Moderator- Using Fertile Fuel as a Burnable Poison• Maximize burn up and actinide consumption• Support research in modeling and analysis ofthe DB-MHR fuel performance

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Molten salt reactors for TRU transmutationStudy objectives:• Feasibility of designing a once-through MSR fed with TRUfrom LWR spent fuel to be critical• Define transmutation capability of critical MSR• Compare transmutation capability of MSR versus LMR andLWR assuming identical fractional transmutation (0.99)MSR pool design:• Carrier salt LiF (15%), NaF (58%)and BeF 2 (27%)• A pool design (graphite-free core)gives the best neutron economyi.e. the highest k ∞

Spectra of systems inter comparedThe fast reactors (LFR, SFR) spectrum peaks at 100KeV; thePWR spectrum peaks in the thermal and the MeV range; theMSR spectrum spreads over the intermediate energy rangeFraction of fissionFast Epithermal Thermal>100keV

Comparison of waste characteristics• In the first 100 years after discharge, the radio-toxicities ofAc from PWR and MSR are higher than those from thefast reactors; afterwards there is no significant difference• Actinides from the fast reactors have lower decay heatthan Ac from PWR and MSR in the first 100 years;afterwards there is no preferred spectrumRadiotoxicity per gram Ac at discharge (left) and after 10 3 years (right) for MSR,PWR, LFR, SFR

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Compact mobile BNCT facilityLawrence Berkeley National Laboratory is developing aCompact D-D Fusion Neutron Source (CNS)having an intensity of 10 12 n/sec.This is 1 order of magnitude too small for BNCT applicationsObjectives of UCB work• Design a small, safe and inexpensive Sub-Critical Neutron Multiplier (SCM) to multiplythe fusion neutrons by an order of magnitude• Design a Beam Shaping Assembly (BSA)and reflector to optimize the neutron beam totreat deep seated brain tumors

Compact mobile BNCT facility (2)Possible advantages ofusing a SCM• Earlier commercialization of theCNS for BNCT• Making it possible to attain theneeded neutron source intensityusing a D-D CNS• (a D-T neutron source needstritium that is expensive, healthhazard. expensive to confine)• Reducing the total power neededfor operating a CNS

Compact mobile BNCT facility (3)Results• Optimal design of a passively cooled SCM made of 20%-enriched,aluminum clad metallic uranium fuel:• Required uranium amount is 8.5 kg and its cost is ~ $57,400. (U cost:50 $/kg and SWU (Separating Working Units) $ 110);• Power level: ~ 400 W when driven by a 10 12 D-D n/s neutronsource passive cooling• Consumption of the initial 235 U atoms during 50 years of continuousoperation only ~ 0.5%. Continuous operation for the entirelifetime of the machine without refueling• Two optimal BSA designs were identified:• one for maximizing the tumor dose rate (10.1 Gy/hour);• and the other for maximizing the tumor total dose (51.8 Gy).• The study concludes that the addition of a SCM makes it possible toincrease the treatment beam intensity by a factor of 18 – from 0.56Gy/hour to 10.1 Gy/hours, with a CNS intensity of 1x10 12 D-D n/s.

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

The SMORES sequence of SCALE-5code packageIs a new prototypic analysis sequence recentlydeveloped by UCB with ORNL for incorporation inthe SCALE-5 code packageProvides an intelligent, semi-automatic search for either• maximum k eff ( m k eff ) a given amount of given fissilematerial can have when in combination with givenmoderating/reflecting material• minimum fissile material mass for a given k eff

Optimization methodologyThat of the SWAN code:• Use first-order perturbation theory to calculate zonedependentreactivity worth of each constituent ofvariable concentration – ρ i (z)• Calculate Equal Volume Replacement Reactivity WorthEVRRW = ρ i,R (z) = - ρ R (z) + ρ i (z)• Use ρ i,R (z) to guide a change in volume fraction Vf i (z) ofthe constituents of variable concentration• Repeat iteratively until meeting optimality condition

Minimum critical mass• Minimum critical mass identified with SMORES aremuch lower than published values• Optimal constituents distribution is difficult to arrive at2.5E-03by “trial and error”Fissile + Polyethylene + BeFissiletype235UMCM (g)201.2Core radius(cm)11.5239Pu Volume Fraction2.0E-031.5E-031.0E-03Polyethylene20 cm sphere 239Pu + Poly + Be239Pu mass = 118.98 gBeryllium233U151.79.55.0E-04239Pu119.0120.0E+000.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 2Radius (cm)

Research highlights• Use of hydride fuel for improving the performance of PWRand BWR• Recycling in PWR using hydride fuel• Lead cooled GEN-IV fast reactors• Solid-core heat-pipe nuclear battery type reactor• Liquid salt cooled high temperature reactors• Deep-burn in GT-HTR• Molten Salt Reactors for TRU Transmutation• Compact mobile BNCT facility• The SMORES (SCALE-5) code for criticality safety analysis• Hi-fidelity methods development

Higher fidelity methods:the numerical nuclear reactor• Fuel Rod Config.• Loading Pattern• Core GeometryDeCART(MOC)Direct 3DWhole CoreTransport47 GroupCross SectionLibraryBerkeley/MichiganIntra-pin-wisePowerDistributionIntra-pin-wiseFuel Temp.ANLNNR validation includes:STAR-CD(CFD)Mechanistic and DetailedT/H CalculationFine Mesh CFDw/ConjugateHeat Transfer• ChannelGeometry• Inlet Flow Cond.• Water/FuelProperty1- Z. Zhong, et al. , "Benchmark analysis of the DeCART MOC code with the VENUS-2 criticalexperiment" Proceedings of the PHYSOR 2004. p21-24, ANS, ANS, Chicago, IL (2004).2- D. Pointer, et al., "Eulerian two-phase computational fluid dynamics for boiling water reactor coreanalysis", M&C 2007, 200745

Higher fidelity methods: thenumerical nuclear reactor (NNR)• NNR originally developed to support of EPRI Fuel Reliability Program.• Goal was to study crud-induced failure from fuel-duty perspectiveAtrium 10 fuel bundleVoid fractionPowergenerated

Continuing work• Perform coupled MOC/CFD transientcalculations– Application to control rod ejection accident– Investigate the effect on fuel performancecode of more accurate power prediction• Core equilibrium calculations using theneutronic module DeCART