Fast Reactor Physics - Meetings and...

Fast Reactor Physics

§  Neutron Modera,on in Fast Reactor –  Typical Neutron Spectrum of Sodium-‐cooled Fast Reactor (SFR) –  Where neutrons go – Neutron Balance

§  Two Faces of Fast Reactors –  Breeder –  Burner

§  Neutronic Codes for Fast Reactor Analysis §  Fuel Cycle Implementa,on as Burner or Breeder

25 4/15/12

Sodium Cooled Fast Reactor -‐ PHYSOR2012

Neutron Moderation in FR

§  In Fast Reactors (FRs), neutron modera,on is avoided by using high A materials –  Slowing down power of acKnides in FR is smaller than structural or coolant materials

–  Significant elasKc scaMering of the neutrons aNer fission, but very liMle energy loss from elasKcs scaMering

–  Slowing-‐down power of most FR materials is less than 1% of H’s slowing down power in PWR

§  Neutrons are either absorbed or leak from core –  What is typical FR Spectrum? –  Where neutrons go in a criKcal FR?

26

ScaMering XSs (barn)

Atomic density (#/barn-‐cm)

Slowing down powers (cm-‐1)

TRU 4.0 3.2E-‐03 1.1E-‐04

U 5.6 5.6E-‐03 2.7E-‐04

Fe 3.4 1.9E-‐02 2.3E-‐03

Na 3.8 8.2E-‐03 2.7E-‐03

H (PWR) 11.9 2.9E-‐02 3.5E-‐01

4/15/12


0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

0.40

0.45

0.50

1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07

Energy (eV)

Nor

mal

ized

Flu

x/Le

thar

gy

LWR (EPRI NP-3787)

SFR (ufg MC2 -2 metal)

Typical Neutron Spectrum of SFR

§  No neutrons below 1 keV and energy tail dependent on composi,ons §  Reac,ons with ac,nides beyond resolved resonance range (>20 keV); unresolved resonance is important §  Most probably neutron energy of ~400keV and lots of peaks due to resonances of intermediate weight

nuclides: resonance treatment of structural materials is important; neutron slowing-‐down model is important; number of neutron groups and group boundaries affect neutronic calcula,ons

27 4/15/12


Fe-56 Cross Sections

4/15/12


28

Red – total Green – elastic scattering Blue - capture

§  Many resonances of intermediate weight nuclides affects fast neutron spectrum

Where Neutrons Go - Neutron Balance

§  Neutrons are either absorbed or leak in Fast Reactors –  A long diffusion length increases neutron leakage more than 20%, while ~4% in PWR (reflector effects

are important) –  U-‐238 capture is significant (needs more fissile to achieve criKcality and increases fissile breeding) –  Fission product poison is small –  Excess reacKvity is small in breeder or break-‐even, but high in burner

PWR SFR

CR=1.0 CR=0.5 U-‐235 or TRU mass fracKon 4.2 % LEU 13.9% TRU/HM 33.3% TRU/HM

Source fission 100.0% 99.8% 99.9% (n,2n) -‐ 0.2% 0.1%

Loss

leakage 3.5% 22.9% 28.7% radial 3.0% 12.3% 16.6% axial 0.4% 10.6% 12.1%

absorpKon 96.5% 77.1% 71.3% fuel 76.7% 71.8% 62.2% (U-‐238 capture) (27.2%) (31.6%) (17.1%) coolant 3.4% 0.1% 0.1% structure 0.6% 3.7% 3.7% fission product 6.8% 1.5% 2.4% control 9.0% 0.0% 2.9%

29 4/15/12


Comparison of One-Group Cross Sections

§  One-‐group XS are significantly reduced in fast systems, but possible to achieve same power compared with thermal reactors (see M. Salvatores’ presentaKon)

§  High ra,o of U-‐238 capture-‐to-‐Pu-‐239 fission –  U-‐238 is a good neutron absorber in FRs –  IniKal core requires high fissile enrichment to achieve criKcality ( > 10% LEU for U-‐fuel) –  High internal conversion

§  Parasi,c capture is smaller compared to thermal systems –  ReacKvity penalty of fission products is small

ReacKon Thermal systems Fast systems

PWR VHTR SCWR SFR LFR GFR

U-‐238 capture 0.91 4.80 0.95 0.20 0.26 0.32

Pu-‐239 fission 89.2 164.5 138.8 1.65 1.69 1.90

Fe capture 0.4 0.007

FP capture 90 0.2

U-‐238 capture to Pu-‐239 fission raKo 0.01 0.03 0.01 0.12 0.15 0.17

ParasiKc capture to Pu-‐239 fission raKo 1.0 0.13

30 4/15/12


Possible w/ Low Fission Cross Section in FR presented by M. Salvatores at ANL, May 2009

4/15/12


31

Fermi saw it right: neutron spectrum should stay „close“ to fission spectrum to get an excellent neutron balance The average cross sections in a spectrum close to the fission spectrum (i.e. in presence of „weak“ moderators) are much lower than in „thermalized“ spectra (factor ~100) So, to reach the same power Nfast~10xNtherm (i.e. higher power density) and !fast~10x!therm (i.e. higher neutron doses and material damage) Moreover, longer neutron mean free path: Excellent neutron balance: neutron excess to be used for breeding (or burning")!

!"=!#÷ ff NP

t!÷ /1"

Historical introduction

Extra Neutrons in Fast Reactors

4/15/12


32

SFR (U/Pu Fuel) Oxide Carbide Metal

ProducKon

η of fissile isotopes 2.283 2.353 2.450

FerKle fission bonus (ε) 0.356 0.429 0.509

Available extra neutrons (η–1+ε) 1.539 1.782 1.959

Loss

AbsorpKon loss 0.231 0.199 0.218

Leakage Loss 0.046 0.051 0.082

Fissile decay loss 0.031 0.029 0.032

Net extra neutrons for breeding 1.331 1.503 1.627

§  High neutron yield (η & ε) and low loss are favorable for breeding §  η and ε increase as spectrum hardens due to η of Pu-‐239 and U-‐238 fission threshold –  See Breeding raKo per SFR fuels in the later presentaKon

Neutron balance (normalized to neutron absorp4on in fission isotopes)

Fast Spectrum Physics Distinctions

§  Combina,on of increased fission-‐to-‐absorp,on ra,o and increased number of neutrons per fission in U/Pu fuel yields more extra neutrons –  Enables “breeding” of fissile material

§  In fast spectrum, U-‐238 capture is more prominent –  Higher fissile content is required –  Enhances internal conversion

§  Reduced parasi,c capture and improved neutron balance –  Allows the use of convenKonal stainless steel structures –  Slow loss of reacKvity with burnup

§  Much longer neutron diffusion length (10-‐20 cm, as compared to 2 cm in LWR) –  Neutron leakage is increased (>20% in typical designs, reacKvity coefficient) –  Reflector effects are more important –  Heterogeneity effects are relaKvely unimportant

33 4/15/12


Two Faces of Fast Reactors – Breeder & Burner

§  Fast Reactors have been developed as breeder and/or burner –  Conversion raKo is flexible in fast reactors (see CR vs. TRU mass fracKon) –  In ABR-‐1000 scoping study, flexible CRs (0.2 – 1.1) was achieved without significant

impacts on safety features •  Intra-‐assembly design parameters were changed, while number of assemblies and control rods are

retained

§  Breeder – Vision of closed fuel cycle –  UKlize extra neutrons for conversion to Plutonium –  High number of neutrons per absorpKon (η) and ferKle fission bonus (ε) are favorable –  Low parasiKc absorpKon is favorable –  EffecKve usage of extra neutrons for high conversion (heterogeneous core, etc.)

§  Burner – MA management –  High fission-‐to-‐capture raKo of ferKle isotopes is favorable for MA management

4/15/12


34

Breeding Ratio per SFR Fuel

§  Hardened neutron spectrum plays an important role for high breeding raKo in U/Pu fuel cycle

§  Metallic fuel has hardest spectrum with low slowing down power

§  Metallic fuel is favorable for breeding due to its high fuel density and harder neutron spectrum

§  Heterogeneous core configuraKon uKlizes neutrons effecKvely, which increases breeding raKo

4/15/12


35

C.E. Till, et al, “Fast Breeder Reactor Studies,” ANL-‐80-‐40, ANL (1980)

ScaMering XSs (barn)

Atomic density (#/barn-‐cm)

Slowing down powers (cm-‐1)

TRU 4.0 3.2E-‐03 1.1E-‐04 U 5.6 5.6E-‐03 2.7E-‐04 Zr 8.1 2.6E-‐03 4.6E-‐04

O 3.6 1.4E-‐02 5.8E-‐03

C 3.9 1.6E-‐02 1.0E-‐02

K-infinite of Depleted Uranium in Fast Reactors

4/15/12


36

!"

#!"

$!"

%!"

&!"

'!"

(!"

)!"

*!"

+!"

#!!"

!,&"

!,("

!,*"

#,!"

#,$"

#,&"

#,("

!" '" #!" #'" $!" $'" %!" %'" &!" &'"

!"#$#%

&'()

*""(+,*'-#.

($#($#

$*/(0

12(3

((

45&.6.

&$7(

89,.9%(:3;(

-./01"-./01"234"56"7.$%*"68.$%+"

§  K-‐infinite increases iniKally as Pu is bred, but decreases as fission products are accumulated

§  Possible to maintain cri,cality w/o external fissile feed if FPs are removed –  Reprocessing (historical approaches) – remove FPs occasionally –  Moving depleKon zone (traveling wave/CANDLE concept) – escape from fission product dominant zone

Fission-to-Capture Ratios

§  Fissile isotopes are likely to fission in both thermal/fast spectrum §  Significant Pu fer,le isotopes are destroyed by fission in FRs

–  Minor AcKnide (MA) generaKon rate is reduced in FRs

0.000.100.200.300.400.500.600.700.800.901.00

U235

U238

Np237

Pu238

Pu239

Pu240

Pu241

Pu242

Am241

Am243

Cm244

Fission/Absorption

PWRSFR

37 4/15/12


Sequential Transmutation Rate from U-238

4/15/12


38

§  Fission-‐to-‐capture ra,o of fer,le isotopes (Pu-‐240, Pu-‐242) plays an important role to minimize MA genera,on

Nn+1 ! ! cn

! cn +! f

n Nn = 1

1+! fn

! cn

N n

!"#$"!!

!$#""!!

%&'(! )*&'+! )*&,"! )*&,$! )*&,&! -.!

!"#$#%

&'()*+',#-

(%.*(/.+',#-

(0*#1

(2-&$(2

3456(

/012345!678913!

:489!678913!

TRU Compositions

§  Genera,on of Pu-‐241 (key waste decay chain) is suppressed due to high fission-‐to-‐capture ra,o of Pu-‐240

§  MA content much lower in fast spectrum system

39 4/15/12


PWR 50GWd/t

U-‐238 repeated recycles CORAIL-‐TRU Eq. state

ABR-‐100 Eq. state thermal fast

Np-‐237 4.9 0.2 0.8 1.8 1.6 Pu-‐238 2.3 4.6 1.4 10.2 2.7 Pu-‐239 46.5 38.8 66.6 24.4 47.7 Pu-‐240 22.4 19.7 24.3 21.7 29.1 Pu-‐241 13.7 11.1 2.1 7.1 4.5 Pu-‐242 6.7 8.5 1.8 16.9 6.8 Am-‐241 0.5 1.9 2.1 5.9 3.1 Am-‐242m 0.01 0.1 0.1 0.1 0.2 Am-‐243 1.5 3.3 0.5 4.6 2.2 Cm-‐244 0.7 5.5 0.2 3.9 1.3 Cm-‐245 0.04 1.8 ~0.0 1.1 0.4 Cm-‐246 ~0.0 3.1 1.7 0.2 Cm-‐247 0.4 0.2 0.02 Cm-‐248 0.6 0.3 0.03 Cf-‐252 9.4E-‐4

Typical Volume Fractions

§  For SFR, fuel volume frac,on is maximized –  Fuel pins are Kghtly packed in hex-‐can with wire-‐wrap,

which increases structure volume fracKon –  High blanket fuel volume fracKon for addiKonal

loading of U-‐238 –  Lower fuel volume fracKon of burner

§  For LFR, high coolant volume frac,on is required to reduce coolant velocity

§  For GFR, high coolant volume frac,on is required due to inferior heat transfer

4/15/12


40

PWR

Fast reactors

SFR-‐Breeder SFR-‐Burner LFR GFR

Driver Blanket

Fuel 30 40 55 30 34 30

Coolant 59 40 17 44 55 57

Structure 11 20 28 26 11 13

H

Fuel Pinand Wire

CornerSubchannel

EdgeSubchannel

InteriorSubchannel

Duct Wall

Fuel Pin D

P

Wire Wrap

Typical SFR Core Layouts

4/15/12


41

26NRC Topical Seminar on SFR, May 3, 2007

Conventional 1000 MWt SuperPRISM (Metal Core)

! Internal and external blankets allocated

– Result in conversion ratio of ~1

! Only 12 control rod locations with very low burnup reactivity losses

! Blanket, two row reflector, and boron carbide for radial shielding

!

!

"

!

!

!

"

!

"

!

!

!

#$%&'$()*+,-

./0'$/12(321/4'0()56-

718%12(321/4'0()5,-

7'92':0;$()*<=-

">%'28()?<-

@AB)=-

" "':;/81$C(D;/0$;2()+-

! !$%E1$C(:;/0$;2()6-

F;012()5G*-

26NRC Topical Seminar on SFR, May 3, 2007

Conventional 1000 MWt SuperPRISM (Metal Core)

! Internal and external blankets allocated

– Result in conversion ratio of ~1

! Only 12 control rod locations with very low burnup reactivity losses

! Blanket, two row reflector, and boron carbide for radial shielding

!

!

"

!

!

!

"

!

"

!

!

!

#$%&'$()*+,-

./0'$/12(321/4'0()56-

718%12(321/4'0()5,-

7'92':0;$()*<=-

">%'28()?<-

@AB)=-

" "':;/81$C(D;/0$;2()+-

! !$%E1$C(:;/0$;2()6-

F;012()5G*-

P

P

P

P

S

P

P

P

P

S

S

P

P

S

P

P

P

P

P

Outer core (102)

Reflector (114)

Shield (66)

P Primary control (15)

S Secondary control (4)

Inner core (78)

§  Breeder has blankets, while burner does not §  Breeder has 12 control rods with low reac,vity swing, while burner has 20 §  High neutron leakage rate requires thick radial reflector, and radial shield

1000MWt S-‐Prism (metallic) 1000 MWt – ABR (metallic)

42

Homogeneous representation of fuel assembly or core

Heterogeneous fuel assembly

Whole core

Homogeneous representation of fuel assembly or core

Heterogeneous fuel assembly

Whole core

Conventional Approximations §  Detailed space-‐energy-‐direc,on analysis performed for a repeated por,on of the geometric domain (lajce physics) –  Lower order approximaKon of Boltzmann equaKon for whole-‐core analysis –  Assumed/approximate boundary condiKons –  Condensed space/energy cross secKons –  Detailed informaKon recovered by de-‐homogenizaKon methods

§  Nuclide deple,on and buildup using quasi-‐steady model

§  Faster transients modeled with condensed (oken single point) model for ,me-‐dependent amplitude –  Accuracy depends on frequency of kineKc-‐parameter and space-‐energy-‐direcKon flux shape recalculaKon

4/15/12


ENDF/B

Assembly, core design parameters

Operating Conditions

Fuel management

scheme

Transient scenario

Plant information

T/H, structural data Eqs. of state

RCT (Intra-assembly reconstruction)

VARI3D (Perturbation theory

code)

Kinetics and reactivity feedback

coefficients

System dynamics behaviors

Pin-wise information

Multi-group XS

Power, burnup, atomic density, etc.

ETOE2/MC2-3 (Slowing down Eq. Solver)

SASSYS/SAS4A (Coupled neutron kinetics, T/H and

Structural mechanics)

Flow and Temperature Distributions

SE2-ANL (Steady-state T/H Analysis)

ANL Fast Reactor Code Suite

43 4/15/12


DIF3D/VARIANT/REBUS-3 (Whole-core fuel cycle analysis)

Neutronics Analysis

4/15/12


44

Thermal Fast Lawce calculaKo

ns

Example Code WIMS-‐9 MC2-‐3

Geometry Single assembly or 2x2 color set assemblies

Homogeneous medium or simple 1-‐D unit cell

EquaKon Transport equaKon Slowing down equaKon

Neutron Groups 172 2082

Major outputs

Assembly homogenized, burnup-‐and temperature dependent few-‐

group XS and Intra-‐assembly informaKon

Assembly homogenized, temperature dependent broad-‐

group XS

Who

le core de

pleK

on Example Code Many opKons DIF3D/REBUS-‐3

Geometry Whole core with homogenized node

Whole core with homogenized node

EquaKon Diffusion (Transport) equaKon Diffusion (Transport) equaKon

Neutron Groups 2 -‐ 4 9 -‐ 230

Major outputs Power, isotopic distribuKon, etc. Power, isotopic distribuKon, etc.

v  There have been recent acKviKes to solve whole-‐core transport/depleKon equaKons directly without group condensaKon process (lawce calculaKons).

Roles of SFR in Fuel Cycle Options (Examples)

4/15/12


45

!Fuel Cycle Data Package (FCDP) System Datasheet PWR-UOX/PWR-MOX/SFR PWR-UOX to PWR-MOX and to SFR Burner for Full Recycling

!

FCDP Template Rev 0.1: December-15-2011 "!

NU FT-1.1 LEU Oxide driver fuel

LEU Co-extraction (Pu/RU)

DF Pu, RU

MA

FP

To ST-2 To ST-3

!RU

!"#$%&'!(!)*'+!

DU

From ST-1 FT-2.1 Mixed Oxide driver fuel

MOX Aqueous separation (TRU)

DF TRU/RU

FP

To ST-3

!

!"#$%&,!(!)*,+!

Pu, RU

From ST-1 FT-3.1 Metallic driver

fuel

Metal Electro-chemical separation (TRU)

DF RU, TRU

FP

To ST-3

!

!"#$%&-!(!)*-+!

RU, MA

From ST-2,3 RU, TRU

./"%#!!$%&'$()(*!+,-.(!/$$+0!&1**!2(!34%567(5!18!3(49%4:18;!),41%6-!96(*!7<7*(!9687.1%8-=! !

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

!

!

SFR

!

PWR

!

PWR

RU !

Material Flow Diagram

!! !

!Fuel Cycle Data Package (FCDP) System Datasheet PWR-UOX PWR-UOX Once Through

!


NU FT-1.1 LEU Oxide driver fuel

LEU

DF

!"#$%&'!(!)*'+!

DU

,-"%#!!$%&'$()(*!+,-.(!/$$+0!&1**!2(!34%567(5!18!3(49%4:18;!),41%6-!96(*!7<7*(!9687.1%8-=! !

$(;(85#!>?!!!!!@!>,.64,*!?4,816:! AB!!!!!@!A1-7C,4;(5!B6(*! D+E!@!D4(--641F(5!+,.(4!E(,7.%4! @!>67*(,4!+,-.(!A1-3%-,*!G(7C8%*%;1(-!!A?!!!@!A(3*(.(5!?4,816:! $H?!@!$%&'(8417C(5!?4,816:!! ! @!>67*(,4!I,.(41,*!J.%4,;(!G(7C8%*%;1(-!! ! ! @!>67*(,4!I,.(41,*!G4,8-3%4.!G(7C8%*%;1(-!!

!

!

PWR


!! !

!Fuel Cycle Data Package (FCDP) System Datasheet SFR – continuous recycling SFR - Continuous Recycling

!


NU FT-1.1 Metallic driver

fuel

Metal Electro-chemical separation (TRU)

DF RU, TRU

FP

To ST-1

!

!"#$%&'!(!)*'+!

From ST-1 RU, TRU

,-"%#!!$%&'$()(*!+,-.(!/$$+0!&1**!2(!34%567(5!18!3(49%4:18;!),41%6-!96(*!7<7*(!9687.1%8-=! !

$(;(85#!>?!!!!@!>,.64,*!?4,816:! AB!!!!!@!A1-7C,4;(5!B6(*! DBE!@!D%516:'7%%*(5!9,-.!4(,7.%4! @!>67*(,4!+,-.(!A1-3%-,*!F(7C8%*%;1(-!!E?!!!@!E(7%)(4(5!?4,816:! BG!!!!!!@!B1--1%8!G4%567.-! FE?!!@!F4,8-64,817-! @!>67*(,4!H,.(41,*!D.%4,;(!F(7C8%*%;1(-!! ! ! @!>67*(,4!H,.(41,*!F4,8-3%4.!F(7C8%*%;1(-!!

!

!

SFR


!! !

§  Once-‐through fuel cycle –  PWR(UOX)

§  3 stage con,nuous recycle –  PWR(UOX) –  PWR(MOX) –  SFR (Burner)

§  Con,nuous recycle –  SFR(Break-‐even)

Mass Flow Rates

4/15/12


46

Once through ConKnuous ConKnuous

PWR PWR(UOX) PWR(MOX) SFR-‐burner

Break-‐even SFR

Required natural resource Natural uranium 18825.0 7163.2 100.3 Products from fuel fab. Depleted uranium 16633.5 6322.1 Products from reactor Discharge fuel 2191.5

Products from separaKon Recovered uranium 727.4 Recovered FPs 105.5 99.1 Others (loss, etc.) 8.2 1.2

Uranium uKlizaKon, % 0.6 1.5 98.8

§  Once-‐through fuel cycle –  Require 18,825 t of natural uranium (NU) and produce 2,192 t of used nuclear fuel

§  3 stage con,nuous recycle –  SFR burns Transuranics (TRU) generated in PWRs. Required NU decreases to 7,163 t and 106 t of FPs is

sent to disposal

§  Con,nuous recycle with Break-‐even SFR –  Required NU is 100.3 t and 99.1 t of FPs is sent to disposal. –  About 99% of uranium resource is uKlized to generate energy

Metric ton to produce 100 GWe-‐year

TRU Recycling in PWR: CORAIL-TRU

4/15/12


47

!"#$!!!$%&'((

( ))

!"#$ %&'()*$(++,-./0$%123,45$

*+,-.,(/01(*2.(( 3043.5*( 267((30859.*.4.77:(

;2.(2.*.10<.4.0+7(=>?!@#(677.8A9B(-7(,.7-<4.,(/01(8+9*-1.3B39-4<(0/(59+*04-+8(CD+E(1.5963.( *207.(0*2( .41-32.,(?FE( /01( *2-7(( *2.( 4+39.61((67(F>&(/+.9:(<4(AB(9-8-*-4<(

4+8A.1( 0/( G>H( 5-47( -4( *2.( 677.8A9BI( *0( 1.*6-4(304J.4*-0469( 301.( 5.1/018643.:(!A0+*( K%L(0/( *2.( /+.9( 5-47( 61.(G>H(5-47I(M2-32( 61.(5963.,(6*(*2.(5.1-52.1B(0/(*2.(677.8A9B:(;2.(=>?!@#(677.8A9B(-7(,-7596B.,(-4(N-<:(O:):(!77.8A9B(,6*6(61.(51.7.4*.,(-4(!55.4,-P(!:(

(

(

(

(

(

(

(

(

(

;2.(59+*04-+8(01(;?F(-7(8+9*-1.3B39.,(-4(*2.(=>?!@#(3043.5*(M-*2(*2.(6-8(0/(7*6A-9-Q-4<( *2.( 86*.1-69:( R*6A-9-Q6*-04( -859-.7( *26*( *2.( 304*.4*( 0/( 59+*04-+8( 01( ;?F(-4-*-699B( 906,.,( -4*0( *2.( 677.8A9B( -7( 3047*64*( 0J.1( *2.( 1.3B39.( 7*6<.7:( ;2.( =>?!@#(677.8A9B(3043.5*(364(*2.1./01.(A.(.4J-7-04.,(67(6(S,.96B$9-4.T(/01(*2.(59+*04-+8(01(;?FI(

;2-7(M01U(/03+7.7(04(*2.(=>?!@#(677.8A9B(3043.5*(*26*(267(A..4(7!!!(V0M47.9.3*-04( R*+,-.7( W&IKIOX:(Y.36+7.( ,.*6-9.,( -4/0186*-04( 04( *2-7A..4(510J-,.,(-4(?./:(&(64,(KI(049B(6(A1-./(,.731-5*-04(0/(-*(-7(<-J.4(2.1.(/01

01( *1647+164-37( C;?FE( -4( 1.*10/-**6A9.( #Z?( 677.8A9-.7( *26*( 364( 1.6,-9B(3+11.4*9B( +7.,( -4( *2.( #Z?( -4,+7*1B:( ;2.( 677.8A9B( ,.7-<4( .8590B7( A+164-+8( 0P-,.( CF>&E( 5-47( 64,( 8-P.,( 0P-,.( CG>HE( 5-47( CM-*2( D+( 01( ;5+1507.:( ;2.( /+.9( *B5.7( 26J.( A..4( ,.J.905.,( 64,( +7.,( .P*.47-J.9B( -4-4,+7*1-.7(0J.1(*2.(M019,I(69*20+<2(*2.(+7.(0/(G>H(/+.9(-7(40*(67(.P*.47-J.;2.(.41-32.,(+164-+8(/+.9(-7(.8590B.,(/01(7+5501*-4<(*2.(1.3B39-4<(36856-A0*2( *2.( G>H$5-4( 50M.1( 64,( *2.(

(

(

( !"#$%&$'$%%(")%*+,-")#%(,../0)%+1%23456*%577/89:;%

!"#$%&'$

(")$%&'$

*+,'-$.+/-$!

!!

!

"

#!

#"

$!

$"

! # $ % & " ' ( ) * #! ## #$ #% #& #" #' #( #)

+,+-.

/01234567128

930:;/05928128<=

%!

%"

&!

&"

"!

""

>4??4@10420:;/<0=

/01234567128 :;/05928128 >4??4@10420:;/

5

!!!!!!!!!!!

!"#$%&'()*)'+,-./012-3'4$56"1%&7875"9#'"9'+:;&'*)'

!"#$%&'()<)'+,-./012-3'4$56"1%&7875"9#'"9'+:;&'<)'

!

"

#!

#"

$!

$"

! # $ % & " ' ( ) * #!

+A5@1

/01234567128

930:;/059281280<0=

%!

%"

&!

&"

"!

""

>4??4@10420:;/<0=

/01234567128 :;/05928128 >4??4@10420:;/

!!!!!!!!!!!!!!!!!

"#!

TRU Recycling in PWR: CORAIL-TRU

§  Con,nuous recycling of TRU in PWR is difficult because recovered fuel is gejng homer and dir,er –  High acKnide generaKon (10X)

§  Require external fissile feed –  As long as the fissile support is needed,

uranium uKlizaKon is less than 2%

4/15/12


48

!"#$!!!$%&'((

)'(

(((

!"#$%&$'$%()*+,-".)/%)0%!123%4,/53"/#%6/5"72.%,8%!,9-"7,8")/%:8,#2%)0%%;<=%>138"-27?73"/#%@()*+,-25%8)%(A<B6CDE1F%

!

"

#

$

%&'

'

(

#

%

&!

! ( " ) # *!"#$%

&'()*$+,%-.-%#*"./%*0.

$"

!&'(*%

*&'&(&#

!!"%

&

&'

&''

&'''

&''''

! ( " ) # *

!"#$%

&'()*$+,%-.1%20('1.3'2(#%

#

%

&!

&"&#

&*

'

(

#

%

&!

&)

&$

! ( " ) # *!"#$%

&'()*$+,%-.4*))*.%1%(4".

%)+33+'1.(*0%.

Values were compared to Pu-‐recycling case

Fast Reactor Physics - Meetings and...

Documents

Transcript of Fast Reactor Physics - Meetings and...