Andrei Shadrin – ITCP «PRORYV», Moscow, 107140, Russia

Reprocessing of mixed U-Pu oxides and nitrides

spent fuel of fast reactors

ROSATOM goal in the back-and is a closed

nuclear fuel cycle

3

ЗСЖЦ

back-end

Очень велик

too long

Отбросить?

to fall off?

Замкнуть!

to close!

4

Что делать с хвостом?

what to do with the back-end?

Сделать так, чтобы хвост не рос to prevent SNF accumulation

Утилизировать наследие

to utilize the nuclear heritage

Вот, собственно, зачем мы создаем

ту самую систему обращения с ОЯТ

that’s why we develop the system of SNF management

2

Baryshnikov M.V., ATOMECO, Moscow, 16.10.2012

MOX and MNIT for fast

reactors

3

• U and Pu are involved in NFC and Np and Am will be involved • Cm will be stored for long time and could be involved in NFC • No limitation recycle of U and Pu

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MOX – mixed U-Pu oxide fuel

MNIT – mixed U-Pu nitride fuel

Dense fuel (metal, carbide,

nitride) is better an compare with

MOX.

However – MOX fabrication and

MOX UNF reprocessing are

industrial process

MOX and MNIT for fast reactor today in Russia

4

• U and Pu are involved in NFC and Np and Am will be involved

• Cm will be stored for long time and could be involved in NFC

• No limitation recycle of U and Pu

BREST-OD-300

Mixed U-Pu nitride (MNIT) fuel

500 μm Pilot scale manufacturing

of MNIT started on Siberian

Chemical Combine

5

1.Pu reproduction coefficient ~1.05 without blanket

2.0.5% ΔК/К during all company

3.Compatibility of MNIT is shells material

4.Compatibility of MNIT with Na and Pb coolant

5.Experience of use U nitride fuel

Reprocessing of irradiated mixed uranium-

plutonium fuel at RT-1 plant

6

RT-1 complex has been

operating since 1977.

The main task of the plant is

the regeneration of irradiated

nuclear fuel for:

• recovery of actinides;

• conditioning of fission

products in a safe form for

disposal form.

Flow-sheet of UNF

reprocessing at RT-1

Reprocessing of MOX SNF at RT-1 plant

7

Reprocessing of MOX at the RT-1 plant was performed twice, in 2012 and 2014.

8 MOX SNF assemblies

4 operations on with 1 assembly at a time in 2012

2 operations on 2 assemblies in 2014) were reprocessed.

The initial fuel composition for these assemblies is a chemically-deposited mixture of U and Pu.

Operation

number

Burnup

factor, GWt

day/t U

Cooling time,

years

Calculated amount of

Pu, kg U, kg

1 80.7 19 5.2 36.2

2 71.6 18 5.5 36.1

3 78.5 13 5.5 35.8

4 77.7 13 5.5 35.9

5 73.3 and 88.9 21 and 18 10.9 71.9

6 75.0 and 81.4 19 and 23 11.0 72.2

8

Dissolution on MOX UNF BN-600

Uranium oxide UNF of BN-600 is a standard fuel for RT-1.

No any difference for MOX UNF BN-600 for reprocessing technology

before dissolution stage

The operation of dissolution has been carried out in two or three stages.

• primary dissolution in 10 M nitric acid solution at a temperature below the boiling point (to

avoid the accumulation of hexavalent plutonium)

• one or two control dissolutions in nitric acid with 1 g/l fluoride ion at a similar temperature

• control dilution at the boiling point.

Stage Process characteristics

Solution composition Temperature, С

Primary dissolution 10 M HNO3 85-95

First control dissolution 10 M HNO3 + 1 g/l F- 85-95

Second control

dissolution 10 M HNO3 + 1 g/l F- 85-95 or 100-105

Yield of actinides on stages,%

Primary dissolution First control dissolution Second control dissolution

U Pu Neutron flow U Pu Neutron flow U Pu Neutron flow

86.7-96.4 87.4-93.7 98.0-99.5 99.4-100 95.3-100.0 99.9-100.0 91.6-100.0 98.9-100.6 100

Conditions of the dissolution stages of MOX fuel assemblies

Efficiency of the dissolution stages of MOX-fuel assemblies

Solution and residue composition

9

No increased loss of uranium and plutonium into waste has been detected during

dissolution of BN-600 MOX UNF an compare with reprocessing of VVER-440 UO2 UNF

Outflow Content of elements

U, g/l Pu, μg/l

Raffinate <0,005 <250

Strip solution of Pu <0,005 2,570,18

Compositions of the outflows of the MOX-fuel extraction solutions

The raffinate contains very low amounts of Pu residue its relatively high content in the initial fuel solution

Operation CU, g/l CPu, mg/l Activity, μR/(s∙l)

5 <0,0005 <0,10 240

6 <0,0005 <0,10 240

The results of the analysis of the solution obtained after washing the sediment

Composition of insoluble residue

10

a b

Composition of the insoluble residue according to gamma-spectrometric (a) and element (b) analysis

The activity of the washed sediment samples ranged from 0.02 to 0.03 Ci.

The main activity is caused by 137Сѕ, 106Ru and 125Sb.

The main components of the residue (excluding carbon and silicon) were found to be Fe, Cr, Ru, Rh, Pd

Using the special sampling device, the claddings were placed into a metal cup and

then into a transporting container.

The total volume of the claddings was about 200 cm3,

The total activity was around 0.6 Ci.

The radiation level from the cup with claddings was about 2000 μR/s.

Hydrometallurgical reprocessing of

BREST-OD-300 mixed uranium-plutonium

nuclear fuel

11

• The duration of external fuel cycle of BREST-OD-300 reactor with

mixed U-Pu nitride fuel (MNIT) including hydrometallurgical

reprocessing should not exceed 3 years.

• An average burnup of the fuel should be 6 % of heavy metal (HM)

with the potential increasing up to 10 % HM.

• Reprocessing technology should provide spent nuclear fuel (SNF)

reprocessing after less than 2 years cooling time and fissile

materials (FM) content 10 – 15 %. Pellets technology MNIT fuel

production has been requires high purification coefficient (~ 106)

for recycled actinides oxides.

MNIT SNF hydrometallurgical

reprocessing flowsheet and requirements

12

ReceivingandentranceSFA

controlSFAdismantlingonfuelrods

Fuelpinsfragmentation

SNFvoloxidationanddrycladdingseparation

LSGOofvoloxidationunit

Dissolutionunit,ClarifyingandcorrectionofSNF

solutions

LSGOofdissolutionunit

Extraction-crystallizationrefining

UO2obtainingunitExtractionfractionatingunitofREE-TPE

Ureextractevaporating

+LSGO

Mixedoxidesobtainingunit

U-Pu-Np

Mixedoxidesobtainingunit

UO2+AmO2

ChromatographicAm-Cmseparationunitandspent

sorbenthandlingLSGOofmixed

oxidesobtainingunit

ConcentratingunitILW+LSGO

ConcentratingunitHLW+LSGO

Cmfractionhandlingunit

+LSGO

Waterflowdetritiationunit

LSGOofSNFdismantlingunitsAndfuelpinsfragmentation

(inertatmosphere)

HLW

HLW

FRM

(U-Cm)Oxforstorage

Primaryoperations

Hydro

HLW

Cooling time 1 year

Recycle of fissile

materials 99,9 %

Type of SNF

Mixed U-Pu nitride

or

MOX

Separation U, Pu,

Np Not provided

U-Pu-Np

purification

coefficient

Up to 106 from 103

Transmutation of

minor actinides

Homogeneously or

heterogeniously

Laboratory and pilot mock-up for

hydrometalurgy

13

Laboratory and pilot mock-up for hydrometallurgy

Laboratoryandhotcellset-ups

Offgastreatment

Hightemperaturetreatmentincludingoxida6on

!

PuO2

dissolu6on

Extrac6on

Ac6nideoxidesprepara6on

Pilotmock-upбоксовогоисполнения

Tanks

BoxesfortechnologyoperaGons

AnalyGcalboxes

• Flow (membrane) clarification – 100 g/l sludge

• Dissolution – U-Pu-Np ingot растворен was dissolved

– Dissolution time is 6 hours

– final concentration of HM was 250 g/l

– There are Pu (IV), Np(V) and 10% Np(VI)

– Pesidue contains of mixed oxides (Pu0,89U0,11)O2, (Pu0,5U0,5)O2 and metal

Dissolution

14

а – ingot U-Pu-Np б – ingot U-Pu-Np separated in 3 parts

Dissolution of simulated fuel

15

To test the oxidized MNIT SNF dissolution process a simulated nuclear fuel based on

MNIT with addition of stable isotopes of Mo, Zr, Pd, Ru, Rh, Ba, Sr, La, Nd has been

synthesized

Element U Pu O C Zr Mo Pd Rh Ru Nd La Ba

Content, wt % 73,0 11,4 0,1 0,5 0,8 1,1 1,2 0,3 0,6 2,1 1,5 0,3

Simulated nuclear fuel composition

MNIT simulated nuclear fuel oxidized

powder at different magnifications

The undissolved residue at different magnifications (the

residue weight – 4,9 % by weight of the initial spent fuel,

the output of noble metal to the solution 99,7 %)

Recovery and separation of Am and Cm

16

Two systems have been chosen:

1.сarbamoylmethylenephosphineoxide (CMPO) -

tributyl phosphate (TBP) -

metanitrobenzotriflouride (F-3)

2. N,N,N’,N’-tetraoctildiglycolamide (TODGA) -

metanitrobenzotriflouride (F-3)

140 hours dynamic test of Am recovery by TODGA - F-3 solution was carried out:

•99,99 % recovery of americium from the simulant HLW

•americium purification degrees from the rare earth elements:

La - more than 300000; Ce - about 500; Pr - over 20000; Nd – over 4500;

Sm - about 600; Eu - over 1000; Gd - about 6000; Y - about 30000.

•Am losses accounted for less than 0,1 %

Furthermore, a pilot Cm and Am separation

tests using a ion exchange resin were

carried out.

Am fraction was approximately 65 g of

241,243Аm containing less than 0,8 wt.% of

Cm and less than 0,1% 154,155

Eu by activity

Flowsheet of dynamic test of the group REE and TPE separation

by the TODGA in F-3

Concentrating of technological HLW

17

Feedsolutionpreparing

Vapor-gasmixture

Noncondensablegasses

H2O

Condensate

Organicphase

Mixingofsolutions

Inorganicphase

Formalin

Vap

or

T-HNO3

12mol/l

Vapor

NaOH

Vapor

Tririumcontainingwater

Condensate

HLWrafinat

H2O2

O2

Vaccum

Evaporationresidueto

vitrification

ToILWhandlingfacility

Toorganichandling

Tocementa-

tion

TothecirculationofTcontaining

acid

Todetritiationsystem

Togaspurification

system

FormalinAir

Condensatesfromvitrification

Flow-sheet of HLW handling

The pilot evaporator test set-up

Nitrogen oxides absorption has been proposed for prevention of Ba and Sr nitrates

precipitation.

Test using a pilot evaporator unit showed:

at REE 100 g/l (aim of evaporation) there is no precipitation of Ba and Sr nitrates

95 % of nitric acid can be recycled (65 % directly and 30% via absorption)

.

Concentrating of recyclable ILW

18

Feed solution preparing Evaporation residue

to NH4NO3 distruction

Technological solutions (IML)

Vapor

Vapor

To clean water cycle

Vapor

HNO3

12 mol/LTo HNO3

circulation

NaOH

Evaporationresidue to

HLW handling

H2O2

To central gas purification

system

Gas vent

Local gas purification

system

Formalin(optional)

To cementation

Air

Water condensate

Technological scheme

of ILW handling

Tasks - nitric acid regenerating and ammonium nitrate and complexons destruction

•First step - destruction during evaporation in the circulation type evaporator with addition of

formalin (Ammonium nitrate destruction is 80 - 85 %.)

•Second step – destruction in the autoclave at f 1,5 – 5,0 MPa within 4-6 hours. (Distraction is

90%)

Conclusion

19

• Нydrometallurgical reprocessing of BREST-OD-300 MNIT SNF after

2 years cooling time and average burnup of the fuel 6 % HM with the

potential increasing up to 10% HM has been proposed.

•

• The SNF reprocessing technology provides fissile materials content

10 – 15 % and FM treatment coefficient at the level of ~ 106.

• Currently in laboratory conditions the following process stages have

been tested on the real products: actinide oxides production and rare-

earth and trans-plutonium elements separation.

• Furthermore, in a pilot scale the process of HLW and ILW

concentration by evaporation has been tested, as well as the Am-Cm

separation.

• In 2015, the design of the MNIT SNF reprocessing plant has been

started, it placed at the JSC Siberian Chemical Plant site as a part of

the pilot demonstration power complex with BREST-OD-300 reactor.

MNIT SNF reprocessing plant should be put in operation after 2020.

Main publications

20

1.A. Shadrin. Proc. Int. conf. Global 2013, Salt-Lake-City, USA, 2013, p. 7683.

2.A. Shadrin, Radiochimica Acta 103(3), January 2015

3.A. Shadrin. Proc. Int. conf. on Management of Spent Fuel from Nuclear, Power Reactors: An

Integrated Approach to the Back End of the Fuel Cycle, Vienna, Austria, 2015.

4.A. Shadrin. Proc. Int. conf. Global 2015, Paris, France, 2015, p. 5140.

5.O. Ustinov , Atomic Energy 117(6):409-414 · April 2015

6.S.A. Kuluykhin, Journal of Radioanalytical and Nuclear Chemistry 304(1):425-428 · April

2014

7.Yu. Kulyako, S.Perevalov, T.Trofimov et al. Radiochimiya, 2013, Vol. 55, Issue 6, p. 481.

8.Myasoedov, Yu. Kulyako, A.Fedoseev et al. Radiochimiya, 2013, Vol. 55, Issue 6, p. 487.

9.Y. Kulyako. Journal of Radioanalytical and Nuclear Chemistry, 2014, Vol. 299, № 3, p. 1293-

1298.

10.B.Ya. Zilberman, M.N Makarychev-Mikhaylov, A.Yu. Shadrin, D.V. Ryabkov Proc. Int. Conf.

"Global-2009" (Paris, France., 2009). SNF, Paris, 2009. p. 222-230.

11.A. Murzin, B. Zilberman, N. Mishina, M. Makarychev-Mikhaylov, D. Ryabkov, N. Ryabkova,

A.Shadrin, Patent RU № 2532396, 2014, bulletin №. 31