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Multi-recycling of plutonium and MA in PWR using hydride fuels

Thursday 16th March 2009

Francesco Ganda – Ehud Greenspan

The 5th Joint Reactor SeminarGoNERI, The University of Tokyo

andNuclear Engineering Department, UC Berkeley

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Is it possible to multi-recycle Pu in PWR?

MOX allows multi-recyclings only up to about 12w/o Pu: after that the large void reactivity coefficient becomes positive.This happens after 2-3 recycles.

Solutions proposed up to now:

CORAIL, MOX-UE, CONFU (all TRU).

All use 235U to reduce plutonium mass, either homogeneously or heterogeneously mixed.

Drawbacks-) reach Pu (or TRU) stabilization, not net destruction; -) large power peak for heterogeneous configurations;-) no substantial U or SWU saving over conventional UO2.

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Our solution:

Use hydride fuels, of the type:

Ux-ZrnH1.6n-ThmH2m-PupH2p (-MAqH2q)

The rationale:

The extra H in the fuel (Hfuel~Hwater) will increase plutonium (and MA) destruction through a softer spectrum and will

counteract the effects of large voiding.

10-2 10-1 1000

0.005

0.01

0.015

0.02

0.025

0.03

0.035

0.04

0.045

0.05

Neutron Energy (eV)

Neu

tron

flux

UO2P UZHMOX

Beginning of Life

10-2

10-1

100

0

0.005

0.01

0.015

0.02

0.025

0.03

0.035

0.04

0.045

0.05

Neutron Energy (eV)

Neu

tron

flux

UO2P UZHMOX

End of Life

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Design variables

-) Fuel cycle scheme (total or partial recycling of U, Pu, Np etc…)-) Fuel composition: mixture of ZrH1.6, ThH2, U in various amounts-) Possible use of burnable poisons-) Possible use of different U enrichment

Assumptions

-) Instantaneous reprocessing-) 0.1% of U and Pu is lost during reprocessing-) 0% losses during fabrication and irradiation-) Feed Pu vector is fixed (from 1)-) Feed MA vector is fixed (from 2)

1 Youinou, G. and Vasile, A., 2005. Plutonium Multirecycling in Standard PWRs Loaded with Evolutionary Fuels. Nuclear Science and Engineering: 151, 25-45. 2 ANL ICONE 10-22575, Table 1 col. b, based on “extended PWR benchmark with 10y cooling”

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Constraints

-) Negative core-average burnup-dependent reactivity coefficients (fuel temperature, coolant temperature, small void, large void)

-) Similar cycle time (1430 EFPD) and control system of standard UO2-) Same fuel D and P/D of standard UO2

Unit Cell Geometry and Specific Power

Hydride Fuels Oxide Fuels

Clad outside diameter 0.95 cm 0.95 cmP/D 1.3261 1.3261

Fuel diameter 0.8192 cm 0.8205 cmClad inside diameter 0.8357 cm 0.8357 cm

Pitch 1.26 cm 1.26 cmSpecific power 76.715 W/giHM 36.138 W/giHM

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Methodology

TRITON/NEWT code is used for depletion analysis, extensively tested for degraded plutonium and MA–bearing fuels, both in

numerical benchmark and in measured samples (long experience with SAS2H).

BONAMI/NITAWL: XSEC pre-processingNEWT: transport, 3-groups XSEC collapsing

ORIGEN-S: 0-D depletion and decay

Transport calculations performed with 238 energy group, directlyincluding 40 actinides, 187 fission products.

More nuclei are followed using ORIGEN-S default XS.

Pu amount is adjusted exactly to match desired cycle length Two steps approach for greater precision:

1. Two runs that over and under estimate cycle length;2. Final run to estimate discharged and cooled composition.

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Methodology – types of fuels analyzed for Pu recycling

PuH2-ZrH1.6-U (uranium varies from 0 g/cm3 to 3.72 g/cm3)PuO2-ZrO2-UO2 (uranium varies from 0 g/cm3 to 8.28 g/cm3)

In both fuel types the U amount is fixed*, the plutonium is adjusted to reach the desired cycle length, the remaining space is taken by zirconium in the appropriate chemical form (hydride or oxide). If the zirconium is zero and more space is needed, uranium is taken out in the desired amount.

For Th-based fuels, ThH2 replaces U. Same methodology.

* The depleted uranium amount is initially fixed at a pre-assigned value, after depletion the entire uranium amount is recycled, adding new depleted uranium if:-) needed to make up for the consumption-) needed to reduce enrichment below 20% 235U or 12% 233U

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Methodologyhydride fuels for Pu + Np or TRU recycling

PuH2-NpH2-(MAH2)-ZrH1.6-U (uranium varies from 0 g/cm3 to 3.72 g/cm3)

Same methodology as before to fill the fuel volume.PuH2 has a density of 10.4 g/cm3, not much is known of hydrides of MA, so it is assumed that the density will be the same* as for PuH2. The density is not adjusted to account for varying isotopic composition*.

When a given amount of Pu is taken from LWR spent fuel, a proportional amount of Np (or MA) is taken as well, according to the following vector**:

U, Np, Pu, Am, CmMA_frac=[0.003810517, 0.07691684, 1, 0.07117211, 0.006312254];

*recommended by prof. Olander**from ANL ICONE 10-22575, Table 1 col. b, based on “extended PWR benchmark with 10y cooling”

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Fuel cycle scheme with plutonium recycle

Fuel fabrication

Reprocessing (Pu, U separation)

Final repository

Reactor (irradiation)

Depleted U Spent LWR fuel

Pu

Reprocessing (Pu separation)

99.9% of Pu, U

100% of FP 0.1% of Pu and U100% of MA

10 y cooling

100% of FP 0.1% of Pu100 % of U100% of MA

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Fuel cycle scheme with plutonium and neptunium recycle

Fuel fabrication

Reprocessing (Pu, U, Np separation)

Final repository

Reactor (irradiation)

Depleted U Spent LWR fuel

Pu, Np

Reprocessing (Pu, Np separation)

99.9% of Pu, U, Np

100% of FP 0.1% of U, Pu, Np100% of other MA

10 y cooling

100% of FP 0.1% of Pu, Np100 % of U100% of other MA

Results: selected characteristics of PUZH and MOX with variable uranium loadings at First Recycle

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U % of total in PUZH* 0% 25% 50% 75% 100% Standard MOX

PUZH TRU destruction fraction 64.36% 54.13% 47.83% 42.39% 37.48%

32.56%

46.13%

49.65%

MOX TRU destruction fraction 63.65% 49.76% 42.65% 37.11%

--23.7%

--PUZH Fissile fraction at EOL 21.63% 32.87% 38.52% 42.74%

MOX Fissile fraction at EOL 24.50% 38.05% 43.39% 46.99% 60.3%* maximum uranium in PUZH: 3.7203 g/cm3

Fraction of maximum uranium in PUZH* Standard MOX

0% 25% 50% 75% 100% --

Rho U (g/cm3) 0 0.9301 1.8601 2.7902 3.7203 8.4612

Rho Pu (g/cm3) 0.7337 0.7564 0.7791 0.8018 0.8245 0.8038

Rho HM (g/cm3) 0.7338 1.6865 2.6392 3.592 4.5448 9.265

Rho fuel (g/cm3) 5.9415 6.6087 7.2759 7.9432 8.6104 10.465

H/HM 82.04 34.89 21.85 15.73 12.18 3.9

Burnup (GWD/MTiHM) 628.4 284.2 182.1 132.6 102.9 54.66

EFPD 1376.0 1430.1 1434.3 1420.9 1395.8 1512.45

Characteristic

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Results: TRU destruction fraction

20.00%

30.00%

40.00%

50.00%

60.00%

70.00%

80.00%

0% 20% 40% 60% 80% 100%

Volume percent of the maximum loadable

TRU

dest

ruct

ion

fract

ion PUZH

TPZHMOX

0.00%

10.00%

20.00%

30.00%

40.00%

50.00%

60.00%

70.00%

0 20 40 60 80 100H/HM

TRU

des

truc

tion

frac

tion

PUZHTPZHMOX

• TPZH = ThH2-PuH2-ZrH1.6; • PUZH = PuH2-U-ZrH1.6; • Maximum loading is Th for TPZH and U for PUZH and MOX,• 100% is for PUZH, not MOX (which can load higher U amounts)

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Results: Fuel temperature coefficient of reactivity (FTC)

-3

-2

-1

0

1

2

3

4

0 100 200 300 400 500EPFD

FTC

(pcm

/K)

FTC with HSeries3

-3

-2

-1

0

1

2

3

4

0 500 1000 1500EPFD

FTC

(pcm

/K) FTC with H

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Results: Fuel temperature coefficient of reactivity (FTC)

-3

-2

-1

0

1

2

3

4

0 500 1000 1500

EPFD

FTC

(pcm

/K)

FTC with HFTC with D

-3

-2

-1

0

1

2

3

4

0 100 200 300 400 500EPFD

FTC

(pcm

/K)

FTC with HFTC with DS i 3

10-5

100

105

1010

0

0.02

0.04

0.06

0.08

0.1

0.12

0.14

0.16

0.18

Neutron Energy (eV)

Neutron flux per unit lethargy

ZrD1.6-P uD2ZrH1.6-P uH2

FTC of PUZH Burnup-dependent FTC<0 for U>25%.

ButFTC>0 for U=0 during the 3rd batch.

Solutions-) burn only up to 2nd batch-) Use D instead of H

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Selected results: TRU destruction fraction as a function of recycle number

PUZH (left) MOX (right)

0 5 10 15 20 25 30 350.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

Cycle Number

TRU destruction fraction

U 0%U 25%U 50%U 75%U 100%

0 2 4 6 8 10 12 140.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

Cycle Number

TRU destruction fraction

U 0%U 11.2%U 22.4%U 33.7%U 44.9%U 64.2%U 83.6%U 100.0%

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Results: PUZH, 33rd recycle: core-average FTC, and LVRC (90%)

without soluble boron

0 50 100 150 200 250 300 350 400 450 500-3.4

-3.2

-3

-2.8

-2.6

-2.4

-2.2

-2

-1.8

EFP D

FTC (pcm/K)

U 0%U 25%U 50%U 75%U 100%

0 50 100 150 200 250 300 350 400 450 500-30

-20

-10

0

10

20

30

40

50

60

EFP D

LVRC (pcm/% void)

U 0%U 25%U 50%U 75%U 100%

Major finding:The LVRC of PuH2-ZrH1.6 remains negative throughout the recyclings!

NOTE: with 100% void LVRC becomes positive, but core leakage compensates

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Results: MOX, 13rd recycle: LVRC (90%)

with soluble boron (left) and core-average without (right)

0 500 1000 1500120

140

160

180

200

220

240

260

EFP D

LVRC (pcm/% void)

U 0%U 11.2%U 22.4%U 33.7%U 44.9%U 64.2%U 83.6%U 100.0%

0 50 100 150 200 250 300 350 400 450 500130

140

150

160

170

180

190

EFP D

LVRC (pcm/% void)

U 0%U 11.2%U 22.4%U 33.7%U 44.9%U 64.2%U 83.6%U 100.0%

whileThe LVRC of PuO2-ZrO2-(U) becomes positive at the 13th recycling

Conclusion:Possible To Indefinite Recycle Pu In PWR Using Hydride But Not Oxides

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Results: PuH2-NpH2-U-ZrH1.6 TRU destruction fraction (left) and

fissile fraction at discharge (right) as a function of recycle number

0 2 4 6 8 10 12 140.1

0.15

0.2

0.25

0.3

0.35

0.4

0.45

0.5

0.55

0.6

Cycle Number

TRU

des

truct

ion

fract

ion

U 0%U 25%U 50%U 75%U 100%

0 2 4 6 8 10 12 140.25

0.3

0.35

0.4

0.45

0.5

Cycle Number

Pu

fissi

le fr

actio

n

U 0%U 25%U 50%U 75%U 100%

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Conclusions

-) It is found possible to infinitely recycle Pu in PWR using as feed only depleted uranium and Pu coming from LWR spent fuel, provided that most of the excess reactivity will be compensated by means other than soluble boron.

-) MOX, on the other hand, only allows recycling up to 9-10 times.

-) The TRU destruction fraction, >60% for first recycle, stabilizes at ~20% at equilibrium.

-) It appears possible to multi-recycle up to 8-12 times Pu and Np in PUZH fuel (corresponding to 110-170 years).

-) All-TRU recycle in U-based hydrides appears feasible only up to few recycles (2-3).

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Thank you

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Results: TRU-H2-U-ZrH1.6 TRU destruction fraction (left) and fissile fraction at discharge (right) as a function of

recycle number

0 5 10 15 20 25 30 35 40 450

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

0.45

0.5

Cycle Number

TRU

des

truct

ion

fract

ion

U 0%U 25%U 50%U 75%U 100%

0 5 10 15 20 25 30 35 40 450.25

0.3

0.35

0.4

0.45

0.5

0.55

0.6

Cycle Number

Pu

fissi

le fr

actio

n

U 0%U 25%U 50%U 75%U 100%