Safety Analysis of Direct Recycling of Nuclear Spent Fuel ... · Safety Analysis of Direct...

Safety Analysis of Direct Recycling of

Nuclear Spent Fuel in

Light Water Reactor (LWR)

Abdul Waris, Indarta Kuncoro Aji,

Novitrian, and Zaki Su’ud

Asahi Glass 2012

Novitrian, and Zaki Su’ud

Nuclear Physics and Biophysics Research Division, Department of Physics,

Faculty of Mathematics and Natural Sciences,Institut Teknologi Bandung,

Jl. Ganesa 10 Bandung 40132, INDONESIA

Outline

� Background

� Objective

� Methodology

� Results and Discussion

Asahi Glass 2012 | Institut Teknologi BandungPHYSI S

� Results and Discussion

� Conclusion

Background

Nuclear Energy Industry grows with 3 main issues:

� Reactor safety: Multi-barrier System :

Defence at Depth

� Nuclear Proliferation: political aspects


� Nuclear Proliferation: political aspects

� High level wastes (HLW) management

� The truly problem in nuclear energy

if we can manage HLW, public acceptance will increase

Background …

�Once countries decide to “go

nuclear” they may think about “HLW

repository site” or at least

“underground research laboratory

(URL)”


(URL)”

Underground Research Laboratories

130 Level

Room 214


240 Level

300 Level

420 Level

VILKSISOv7

DIF10DIF8

DIF9

DIF12

DIF13

Room 421

Room 301

Room 413

DIF14

DIF7DIF6DIF5

Generic URL : Kolar Gold Field, India

Site specific URL:Lac de Bonnet ,Canada,

Background …

� They may to choose “closed cycle

strategy” to recycle HLW in any type

reactor or hybrid systems (including

ADS(accelerator driven system))

�reprocessing plant is required


�reprocessing plant is required

� Its very difficult to have country’s own

reprocessing plant, besides it is very

expensive.

� Even OECD countries likes Korea is not

allowed to has a reprocessing plant

The Nuclear Fuel CycleUraniumMining

UraniumconcentrateYellow Cake

Conversion to UO 2 or UF6

Waste

Recovered Uraniumand Plutonium


235U enrichment and Fuel fabrication

Nuclear Power Station – Electricity Generation

Used FuelStorage

Reprocessing and Recycling

Background …

� IAEA suggests to construct and operate some regional reprocessing plant (for example in east asia region)

� Non-proliferation strategy proposed not toseparate Plutonium with Minor Actinidesduring the spent fuel reprocessing


separate Plutonium with Minor Actinidesduring the spent fuel reprocessing

� Some countries (especially which notallowed to have any reprocessing plantand enrichment plant) should find a bestway to deal with their HLW.

DUPIC: Direct Use of spent PWR fuel In CANDU

reactors Strategy

CANDU

PWR

Spent LWR Fuel

Uranium Saving

DUPIC Fuel Fab

On-site Storage

Natural Uranium

Korea:

-Advanced NE Industry

-- Good budget


Spent LWR/DUPIC Fuel

No Disposal

Less Disposal

DUPIC

PWR once-through

DUPIC Fuel FabAFR Storage

Permanent Disposal

On-site Storage

Permanent Disposal

* DUPIC : Direct Use of spent P WR fuel I n CANDU reactors

budget

--Not allowed to have RP &

EP

-- More than 1 type of

NPP (LWR & CANDU)

Background …

� Country likes Indonesia (no NE Industry,

less committed, limited budget, not

allowed to have RP & EP, If “go nuclear”

might have only 1 type of NPP) should has

another alternative way � SUPEL: Straight


another alternative way � SUPEL: Straight

Utilization of sPEnt LWR fuel in LWR reactorsscenario for Nuclear Waste Recycling Strategy

� Best NPP candidate for Indonesia either PWR or

BWR.

SUPEL Scenario for Recycling Strategy

LWR

PWR

Spent LWR Fuel

Uranium Saving

SUPEL Fuel Fab

On-site Storage

Enriched Uranium


Spent LWR/SUPEL Fuel

No Disposal

Less Disposal

SUPEL

PWR once-through

SUPEL Fuel FabAFR Storage

Permanent Disposal

On-site Storage

Permanent Disposal

* SUPEL : Straight U tilization of sPE nt LWR fuel in L WR reactors

Objective

� In our previous study on SUPEL PWR, we found

that the reactor can achieve it criticality when

the U-235 enrichment in loaded fresh fuel is ≥ 4%

and the amount of spent fuel in the core is ≤ 20%.

However, the safety aspect is not evaluated yet.

In the present report, we have performed a


� In the present report, we have performed a

safety analysis of direct recycling of spent PWR

fuel in PWR system, by evaluating the influence

of changing moderator-to-fuel volume ratio (MFR)

of PWR.

Methodology

Thermal power output 3000 MWth

Average cell power density 100 Wcm-3

Fuel pellet diameter 8.0 mm

Fuelroddiameter 9.6 mm

Design parameter of studied PWR


Fuelroddiameter 9.6 mm

Pin pitch 11.8 mm

Fuel type Oxide

Cladding Zircaloy-4

Coolant H2O

Moderator-to-fuel volume ratio (MFR) 0.5 – 4.0

Nuclear Power Plant using Pressurized Water Reactor (PWR)


Diagram of SUPEL Scenario

Methodology …


Effective multiplication factor (k-eff) of 6.0% UO2

enrichment

1.05

1.1

1.15

1.2

MFR=0.5

MFR=1.0

MFR=1.5

MFR=2.0

MFR=3.0

MFR=4.0

Effe

ctiv

e m

ulti

plic

atio

n fa

ctor

(k-

eff) PWR can not

attain its criticality condition


0.85

0.9

0.95

1

1.05

0 200 400 600 800 1000 1200 1400

Effe

ctiv

e m

ulti

plic

atio

n fa

ctor

(k-

eff)

Operation time (days)

condition


enrichment

1.05

1.1

1.15

1.2

MFR=0.5

MFR=1.0

MFR=1.5

MFR=2.0

MFR=3.0

MFR=4.0

Effe

ctiv

e m

ultip

licat

ion

fact

or (

k-ef

f) PWR can attain its criticality condition for

MFR = 4.0, since k-eff > 1 after more


0.85

0.9

0.95

1

0 200 400 600 800 1000 1200 1400

Effe

ctiv

e m

ultip

licat

ion

fact

or (

k-ef

f)


k-eff > 1 after more than 2/3 of operation time (cycle length)


enrichment

PWR can attain its criticality condition for all MFR values, since k-eff > 1 after more

1.1

1.15

1.2

1.25

MFR=0.5

MFR=1.0

MFR=1.5

MFR=2.0

MFR=3.0

MFR=4.0

Effe

ctiv

e m

ultip

licat

ion

fact

or (

k-ef

f)


k-eff > 1 after more than 2/3 of operation time (cycle length)

0.85

0.9

0.95

1

1.05

0 200 400 600 800 1000 1200 1400

Effe

ctiv

e m

ultip

licat

ion

fact

or (

k-ef

f)



enrichment

1. PWR can attain its criticality condition for All MFR

values, since k-eff > 1 during the whole

operation time (cycle length).2. The criticality can be achieved easier as the

increasing of MFR

1.15

1.2

1.25

MFR=0.5

MFR=1.0

MFR=1.5

MFR=2.0

MFR=3.0

MFR=4.0

Effe

ctiv

e m

ultip

licat

ion

fact

or (

k-e

ff)


increasing of MFR

0.95

1

1.05

1.1

0 200 400 600 800 1000 1200 1400

Effe

ctiv

e m

ultip

licat

ion

fact

or (

k-e

ff)


Since in PWR, the moderator is also the coolant material,

more coolant means that cooling process becomes

faster. As a consequence, the safety of the reactor becomes

higher

Number density of actinides for 8.0% enrichment

for MFR=4.0

18

1020

1022

Nuc

lide

num

ber

dens

ity in

rea

ctor

(#/

cc)

The number density of Pu-239, Pu-240, Pu-241, and Cm-244 significantly reduce with the

augmenting of MFR.

In contrast, the number density


1012

1014

1016

1018

0 200 400 600 800 1000 1200 1400

U-235U-238Np-237Pu-238

Pu-239Pu-240Pu-241Pu-242

Am-241Am-243Cm-242Cm-244

Nuc

lide

num

ber

dens

ity in

rea

ctor

(#/

cc)


In contrast, the number density of Pu-238, Am-241, Am-243,

and Cm-242 increase with the enlarging of MFR.

Conversion ratio for 3.5% and 7.5% enrichment

0.85

0.9

0.95

1

Con

vers

ion

ratio

(#)

The conversion ratio decreases with the

increasing of MFR as well as the

increasing of UO


0.6

0.65

0.7

0.75

0.8

0 1 2 3 4 5

CR_3.5% U235CR_7.5% U235

Con

vers

ion

ratio

(#)

MFR

increasing of UO2enrichment in loaded

fresh fuel.

Neutron Flux for 3.5% and 7.5% enrichment

The neutron flux decreases with the

increasing of MFR as well as the

increasing of UO3.5 1014

4 1014

4.5 1014

5 1014

Flux neutron_3.5% U-235 Flux neutron_7.5% U-235

Flu

x ne

utro

n (#

/cm

2/s)


increasing of UO2enrichment in loaded

fresh fuel.

1 1014

1.5 1014

2 1014

2.5 1014

3 1014

0 1 2 3 4 5

Flu

x ne

utro

n (#

/cm

2/s)

MFR

Conclusion

� Preliminary study on safety analysis of direct

recycling of PWR spent fuel to support SUPEL

scenario has been carried out.

� The reactor can achieve it criticality for as a

minimum 6.5 a% of U-235 enrichment in the loaded

fresh fuel with the fraction of spent fuel in the core


fresh fuel with the fraction of spent fuel in the core

is 5.0 %.

� The criticality can be achieved easier as the

increasing of MFR. Since in PWR, the moderator is

also the coolant material, more coolant means that

cooling process becomes faster. As a consequence,

the safety of the reactor becomes higher.

Arigatou Gozaimasu

Terima kasih


Terima kasih

Thank you

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