advanced reactors and fuel cycle group research at...

Advanced Reactors and Fuel Cycle Group Research at UC Berkeley

Department of Nuclear EngineeringEhud Greenspan

Francesco 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 Core

Designs. With MIT & Westinghouse. Funded by the DOE NERI Program from 9/02 to 2/06.

2. Feasibility of Recycling Plutonium and Minor Actinides in Light Water Reactors Using Hydride Fuel. With MIT & ANL. Funded by the DOE NERI Program 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 Reactor Module. DOE NEER contract from 7/1/2005 through 6/30/2008.

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 DOE

8. Molten Salt Reactors for TRU Transmutation. Has been funded by AFCI program via LANL

9. Highly Compact Accelerator-Driven Subcritical Assembly for Medical and Industrial Applications. with Profs. Vujic(PI), Kastenberg and Leung. Funded by the DOE NEER Program from 6/03 to 6/06.

10.Development of the SWAN-SCALE Code for keffMaximization and Critical Mass Minimization. Funded by Oak Ridge National Laboratory from 1/98 through 9/06.

Ongoing & recent research projects (2)

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 PWR

and 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=1.35

Reference oxide fuel bundle Hydride fuel bundlevery heterogeneous nearly uniform

Use of hydride fuel for BWR can (2)

2. 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

BOL pin –wise power distribution in reference OxF bundle

0.92 1.03 1.07 1.09 1.08 1.10 1.08 1.04 0.92

1.03 0.78 0.98 0.78 0.73 0.80 0.99 0.79 1.04

1.06 0.97 0.76 0.91 0.98 0.99 0.79 0.99 1.07

1.09 0.79 0.91 1.07 0.98 0.80 1.10

1.08 0.73 0.99 0.98 0.73 1.09

1.10 0.80 0.98 1.07 0.91 0.79 1.10

1.08 0.99 0.79 0.98 0.98 0.91 0.75 0.97 1.07

1.04 0.79 0.99 0.80 0.72 0.79 0.97 0.78 1.03

0.92 1.04 1.08 1.10 1.09 1.09 1.07 1.03 0.91

1.04 1.02 1.02 1.01 1.01 1.00 1.01 1.02 1.02 1.05

1.02 1.02 1.03 1.00 0.98 0.98 1.00 1.03 1.02 1.03

1.01 1.03 CR 1.02 0.97 0.97 1.02 CR 1.03 1.02

1.00 1.00 1.01 0.98 0.96 0.97 0.99 1.01 1.00 1.01

1.00 0.98 0.97 0.96 0.96 0.95 0.97 0.97 0.97 1.01

1.00 0.97 0.97 0.96 0.95 0.96 0.96 0.97 0.98 1.00

1.00 0.99 1.01 0.98 0.96 0.97 0.98 1.02 1.00 1.01

1.01 1.03 CR 1.01 0.97 0.97 1.02 CR 1.04 1.03

1.01 1.00 1.02 0.99 0.98 0.98 1.00 1.03 1.01 1.03

1.04 1.02 1.01 1.00 1.00 0.99 1.01 1.02 1.02 1.05

BOL pin - wise power distribution in HyF with IFBA

Use of hydride fuel for PWR

10-2

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Neutron Energy (eV)

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rgy

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om.)

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BOL Spectrum (top) and effect of increase in fuel temperature (bottom)

UO2U-ZrH1.6

Mostly Doppler

Mostly spectral shift


and 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 PuH2-U-ZrH1.6 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 fuel

assembly• Might be able to burn ~85% of own Pu in 2 cycles !

Transmutation performance with multi-recycling

Pu in PuH2-ZrH2 can be recycled un-limited # of times; MOX, on the other hand, can be recycling only up to 2-3 timesPu-Np in hydride fuel can be recycled at least 6 times TRU in hydride fuel can be recycled at least 2-3 times

What limits the number of recycling is the large void coefficient of reactivity

0%

10%

20%

30%

40%

50%

60%

70%

0 10 20 30 40Recycle Number

TRU

des

truct

ion

fract

ion

Pu+recycled U

Pu+Np+recycled U

TRU+recycled U

No fertile fuel


and 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 the ENHS NERI project

E. Greenspan (PI), A. Barak, D. Barnes, M. Milosovic’, D. Saphier, Z. Shayer, H. Shimada and S. WangUniversity of California, Berkeley

N. W. Brown (co-PI), L. Fischer and Q. HussainLawrence Livermore National Laboratory

D. C. Wade (co-PI), E. Feldman, K. Grimm, R. Hill, J. J. Sienicki and T. SofuArgonne National Laboratory

M. D. Carelli (co-PI), L. Conway and M. DzodzoWestinghouse Electric Company Science & Technology Department

Yeong 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 University

I. Kinoshita (co-PI), A. Minato, Y. Nishi and N. UedaCentral Research Institute of Electric Power Industry (CRIEPI)

ENHS reactor layout30

m27

m

8m

2m

3m 2m

Number of Stacks = 4Cross Section of Stack

3m

3.64m (O.D; t=0.05)

17.6

25m

ENHS module

Reactor pool

Reactor Vessel Air Cooling System (RVACS)

Steam generators6.94m (I.D.)

Seismic isolators

Underground silo

Schematic vertical cut through the ENHS reactor

Replaceable Reactor 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 but RVACS

ENHS reactor layout (2)

Expanded view of the ENHS reactor (not to scale)

Steam generator

Secondary coolant

Primary coolant

Heat exchanger

Peripheral control assembly

Central control assembly

Core

Highlights

• 2001 – ENHS type reactors selected as one of 6 categories 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 as one 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 or replaceable reactor module

• meet market opportunities for electricity production on small grids, and for developing countries who may not wish to deploy an indigenous fuel cycle infrastructure to support their nuclear energy systems

• The battery system is designed for distributed generation of electricity and other energy products, including hydrogen and potable water”

Highlights (2)

2006 – One of the 7 goals of GNEP is: “Small scale reactors designed for the needs of developing countries” …The U.S. will also encourage the GNEP consortium to pursue the ultimate goal of developing and deploying a small scale reactor that utilizes the same nuclear fuel for the lifetime of the reactor.”

??? What reactor type??? IRIS. Particle fuel LWR, LFR???

!!! Should be LFR !!!

Heat-pipe version of ENHS (HP-ENHS)

• Derives features from the ENHS nuclear battery type reactor concept and from the SAFE-400 space nuclear power reactor concept

• Fuel rods and heat pipes (HP) are horizontally oriented; there are 2 HP’s per 3 fuel rods – one HP serves the left half and the other HP the right half of the core

• Solid core; fission energy is transported out from the core by HP’s that transfer this energy to a coolant flowing (by natural circulation) through heat exchangers formed from both sides of the core by the HP’s that extend beyond the axial reflector

HP-ENHS core schematicsGas

PlenumGas

Plenum IHXActive CoreIHX

HeatPipes

HeatPipes

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 outlet temperature > 800oC; 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 is removed by, say, RVACS

• No positive void reactivity coefficient (typical to fast spectrum reactors); loss of secondary coolant has a negative reactivity feedback

Pebble Bed-Advanced High Temperature Reactor (PB-AHTR)

• TRISO fuel particles dispersed in 6 cm diameter graphite pebbles• 2LiF-BeF2 coolant• Inlet/outlet coolant temperature 600 oC/700 oC• 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/m3

• Total core power 2,400 MWth

Designed by Prof. Peterson et al.

PB-AHTR: attainable burnup and reactivity coefficients

Feedback mechanism Value

Fuel temperature -3.85 pcm/K

Coolant temperature -0.34 pcm/K

Moderator and fuel temperature -4.18 pcm/KModerator and coolant temperature -0.84 pcm/K

Maximum attainable burnup (GWd/tHM) as a function of C/HM and fuel kernel diameter

Coolant temperature reactivity feedback (pcm/K) as a function of C/HM and fuel kernel diameter

PB-AHTR can be designed to attain burnup up to ~130 GWd/tHM using

10% enriched uranium while all reactivity coefficients remain negative

Reference design:425 µm fuel kernel

12.5% TRISOs packing factor127 GWd/tHM (664 EFPD)

High temperature reactors comparison• PB-AHTR maximum discharge burnup is very similar to that of the

other three design options for high temperature reactors• Compared to the PBMR, the PB-AHTR can operate at higher power

density, larger total core power and therefore lower leakage probability• The power generated per pebble for the PB-AHTR is ~2.5 times that

for the PBMRFeature PB-AHTR PBMR LS-VHTR VHTRCoolant Flibe He Flibe HeTotal power (MWth) 2,400 600 2,400 600Power density (MW/m3) 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 - -

Deep Burn MHR• DB-MHR incinerates

plutonium and minor actinides reaching burnups of ~600GWd/MT

• High temperatures can be used for increased thermal efficiency or process heat (H2 )

• However plutonium fuel causes large power peaking factors that must be mitigated

Coarse resolution of power distribution 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 of the DB-MHR fuel performance

Molten salt reactors for TRU transmutation

Study objectives:• Feasibility of designing a once-through MSR fed with TRU

from LWR spent fuel to be critical• Define transmutation capability of critical MSR• Compare transmutation capability of MSR versus LMR and

LWR assuming identical fractional transmutation (0.99)

MSR pool design:• Carrier salt LiF (15%), NaF (58%)

and BeF2 (27%)• A pool design (graphite-free core)

gives the best neutron economy i.e. the highest k∞

Spectra of systems inter comparedThe fast reactors (LFR, SFR) spectrum peaks at 100KeV; the

PWR spectrum peaks in the thermal and the MeV range; the MSR spectrum spreads over the intermediate energy range

Fraction of fission

Fast Epithermal Thermal >100keV <0.625eV

PWR 1% 5% 94%

MSR 4.5% 95.3% 0.2%

SFR 34.6% 65.4% 0.0%

LFR 41.5% 58.5% 0.0%

Equilibrium composition of Ac

Comparison of waste characteristics• In the first 100 years after discharge, the radio-toxicities of

Ac from PWR and MSR are higher than those from the fast reactors; afterwards there is no significant difference

• Actinides from the fast reactors have lower decay heat than Ac from PWR and MSR in the first 100 years; afterwards there is no preferred spectrum

Radiotoxicity per gram Ac at discharge (left) and after 103 years (right) for MSR, PWR, LFR, SFR

Compact mobile BNCT facilityLawrence Berkeley National Laboratory is developing a

Compact D-D Fusion Neutron Source (CNS)having an intensity of 1012 n/sec.

This is 1 order of magnitude too small for BNCT applications

Objectives of UCB work

• Design a small, safe and inexpensive Sub-Critical Neutron Multiplier (SCM) to multiply the fusion neutrons by an order of magnitude

• Design a Beam Shaping Assembly (BSA)and reflector to optimize the neutron beam to treat deep seated brain tumors

Compact mobile BNCT facility (2)

Possible advantages of using a SCM

Earlier commercialization of the CNS for BNCT

Making it possible to attain the needed neutron source intensity using a D-D CNS

(a D-T neutron source needs tritium that is expensive, health hazard. expensive to confine)

Reducing the total power needed for 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 1012 D-D n/s neutron

source passive cooling • Consumption of the initial 235U atoms during 50 years of continuous

operation only ~ 0.5%. Continuous operation for the entire lifetime 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 to increase the treatment beam intensity by a factor of 18 – from 0.56 Gy/hour to 10.1 Gy/hours, with a CNS intensity of 1x1012 D-D n/s.

The SMORES sequence of SCALE-5 code package

Is a new prototypic analysis sequence recently developed by UCB with ORNL for incorporation in

the SCALE-5 code package

Provides an intelligent, semi-automatic search for either

• maximum keff (mkeff) a given amount of given fissile material can have when in combination with given moderating/reflecting material

• minimum fissile material mass for a given keff

That of the SWAN code:

• Use first-order perturbation theory to calculate zone-dependent reactivity worth of each constituent of variable concentration – ρi(z)

• Calculate Equal Volume Replacement Reactivity Worth EVRRW = ρi,R(z) = - ρR(z) + ρi(z)

• Use ρi,R(z) to guide a change in volume fraction Vfi(z) of the constituents of variable concentration

• Repeat iteratively until meeting optimality condition

Optimization methodology

Minimum critical mass

12119.0239Pu

9.5151.7233U

11.5201.2235U

Core radius (cm)

MCM (g)Fissile type

• Minimum critical mass identified with SMORES are much lower than published values

• Optimal constituents distribution is difficult to arrive at by “trial and error”

Fissile + Polyethylene + Be

20 cm sphere 239Pu + Poly + Be239Pu mass = 118.98 g

0.0E+00

5.0E-04

1.0E-03

1.5E-03

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2.5E-03

0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 2

Radius (cm)

239P

u V

olum

e Fr

actio

nBerylliumPolyethylene

45

Higher fidelity methods: the numerical nuclear reactor

NNR validation includes:1- Z. Zhong, et al. , "Benchmark analysis of the DeCART MOC code with the VENUS-2 critical

experiment" 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 core

analysis", M&C 2007, 2007

• Fuel Rod Config.• Loading Pattern• Core Geometry

• Channel Geometry• Inlet Flow Cond.• Water/Fuel Property

47 Group Cross Section Library

Intra-pin-wise Power

DistributionIntra-pin-wise

Fuel Temp.

Fine Mesh CFD w/Conjugate Heat Transfer

Direct 3D Whole Core Transport

ANL

Berkeley/Michigan

Mechanistic and Detailed T/H Calculation

DeCART(MOC)

STAR-CD(CFD)

Higher fidelity methods: the numerical nuclear reactor (NNR)

• NNR originally developed to support of EPRI Fuel Reliability Program.• Goal was to study crud-induced failure from fuel-duty perspective

Atrium 10 fuel bundle

Void fraction

Powergenerated

Continuing work

• Perform coupled MOC/CFD transient calculations– Application to control rod ejection accident– Investigate the effect on fuel performance

code of more accurate power prediction• Core equilibrium calculations using the

neutronic module DeCART

advanced reactors and fuel cycle group research at...

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