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Designing new types of fuel and structural materials for large-scale nuclear power industry in Russia

Russia, Moscow, 26-27.05.2010.

V.M .Troyanov, A.V. Vatulin, V.V Novikov, I.A.Shkabura

VNIIMN after A.A. Bochvar, OJSC

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INTRODUCTION

The presentation addresses 3 issues related to designing nuclear fuel to supply nuclear energy complex of Russia:

1 – fuel for VVER-1000 and VVER-1200 reactors,

2 – conceptual approaches to establishing production of mixed fuel for fast breeder reactors operating within a closed fuel cycle, 3- development of the dispersed type fuel elements for floating reactor units (FRU) and small nuclear power plants (SNPP) разработка твэлов дисперсионного типа для плавучих энергоблоков (ПЭБ) и атомных станций малой мощности (АСММ).

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VVER-1200Key parameters of the VVER-1200 reactor and nuclear fuel

ParameterVVER-1000 VVER-1200

Rated power, MW 3000 3200

Coolant pressure at the reactor outlet, MPa 15,7 16,2

Coolant temperature at the reactor inlet, С 291 298,6

Coolant temperature at the reactor outlet, С 321 329,7

Maximal line heat flow, W/sm 448 420

Refueling interval, months 12 - 18 12/(18-24)

Fuel column height, mm 3530 3730

UO2 mass, кг 80600 87065

All changes of the operating parameters of fuel have to be justified!!!

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VVER-1200

Fuel cycles of NPP-2006 – customer’s choice!

Fuel cycle 5х1 3х1,5

Number of make-up fuel assemblies, pcs. 36 78

Average enrichment, % 4,84 4,85

Duration of the fuel campaign, effective days (without power effect)

302 521

Burnup of the unloaded fuel assemblies, MW*days/kg U-average-maximal

57,264,5

45,664,0

Specific consumption of natural uranium (0.3% dump) 0,191 0,24

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VVER-1200 The fuel assembly design is based on the experience accumulated in reference projects FA-

A and FA-2

Fuel element of TVS-2006

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VVER-1200

Fuel composition and fuel cladding

Gadolinium monoxide integrated into the fuel matrix to the weight fraction of 10% is used as a burnable poison.

Fuel cladding is made of optimized E-110opt alloy.

Fuel pellets have outer/inner diameters of 7.6/1.2 mm.

An option to use 7.8 mm pellets without an orifice and properly adjusted thickness of the cladding 9.10х0.57 mm in the future is being analyzed.

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Development of the fuel design for VVER-1000

Increased fuel load due to optimization of the fuel core and cladding at fixed outer dimensions of the cladding.

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Parameter Actual dimension

New dimension

Cladding thickness, mm 0.65 0.57

Pellet diameter, mm 7.57 / 7.60 7.80

Central hole, mm 1.4 / 1.2 0

Average grain size, μm 10 25

Features that provide fuel rod’s service lifetime: – use of zirconium sponge (since 2009)– fuel pellets with specified structure– E110 alloy with optimized chemistry

L=3530 mm

L=3530 +150 mm

L=3530 +200 mm

Advanced Fuel Rods for VVERAdvanced Fuel Rods for VVER--10010000 FA FA

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21-st coreof the Kalinin NPP Unit 1 (2005-2006 campaign)

30 FA-A - 7,57/1,4

10 FA-A - 7,60/1,2

1 FA-A - 7,60/1,2 + 18 FA 7,6/0,0

1 FA-A - 7,60/1,2 +18 FA 7,8/0,0

- 7,8/0,0 or 7,6/0,0 (4,4 %) FA

- 7,6/1,2 (4,4 %) FA

- 7,6/1,2 (4,95 %) FA

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Fuel chart of the 22nd core of the Kalinin NPP Unit 1 (2006-2007 campaign)

18 FA-A - 7,57 / 1,4

18 FA-A - 7,60 / 1,2

6 FA-A - 7,80 / 0,0

1 FA-A - 7,60/1,2 + 18 FA 7,6/0,0 (~ 28 MW*day/kg U)

1 FA-A - 7,60/1,2 + 18 FA 7,8/0,0 (~ 28 MW*day/kg U)

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State of the surface of FA-A assemblies after 2 years of operation

Neighborhood of the 13th SG Neighborhood of the 2nd SG

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Fuel chart of the 24-th core of the Kalinin NPP Unit 1 (2008-2009 campaign)

36 FA-A - 7,80 / 0,0

6 FA-A - 7,80 / 0,0

1 FA-A - 7,60/1,2 + 18 FA 7,6/0,0 (~ 55 MW*day/kg U)

1 FA-A - 7,60/1,2 + 18 FA 7,8/0,0 (~ 55 MW*day/kg U)

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VVER-1200

Justification of the corrosive resistance made for the new reactor parameters including the steam content in the coolant increased to 11.4 weight%. The calculated weight steam content at the outlet of the “hottest” cell throughout the campaign (real parameters for 5x1 years fuel

cycle) is shown on the graph.

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E110standard

E110optimized

Correction of 2007. TS for Fe in E110

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Modernization of the E110 alloy – increase of O and Fe content

300

200

750

100 450

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Dependence of radiation-induced deformation upon iron content in the pressure tubes made of E110 alloy irradiated in the BOR-60 reactor.

Radiation exposure 4200 hours

Radiation induced growth

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Characteristics of the new generation and standard fuel assemblies

Major requirements to the fuel cladding material• Increased reliability of the fuel of new generation (zirconium sponge, wall

thinning 0.65 0.57 mm) • Ensuring competitive ability (corrosion properties, resistance to deformation)• Processibility

Parameter Fuel cladding of new design, tech specs TS 001.392-2006

Standard fuel cladding, tech specs TS 95 2594-98

Cladding material E110 alloy based on zirconium sponge

E110 alloy based on electrolytic zirconium

Surface processing:InnerOuter

GrindingJet etching

Through etchingThrough etching

Dimensions:Outer diameterInner diameterWall thicknessUnevennessRoughness:Outer surfaceInner surface

9.1 mm7.93 mmAt least 0.54 mmNo more than 0.05 mm

Ra ≤ 0.6 mkmRa ≤ 0.8 mkm

9.1 mm7.73 mmAt least 0.63 mmNo requirement

Ra ≤ 1 mkmRa ≤ 1.5 mkm

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Conclusions for VVER-1000: fuel evolution at the plants

FA structural materials ZircalloysE110 and E635

Increase of FA rigidity

Burnable poison UO2 – Gd2O3 (5% 8%) Decreased peak loads. Increased burnup.

Fuel enrichment 235U (4,4% 4,95%)

Increased burnup and power generation

Increased outer diameter of the fuel

pellet

7,57 7,60 7,8

Decreased central orifice of the fuel pellet

2,3 1,4 1,2 0,0

Increased length of the fuel column

3530 3680

Improved design of the cladding

9,1х7,73 9,1х7,93

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Mixed fuel for fast breeders

The Federal targeted program “New generation nuclear power technologies” defines a high priority goal of establishing a closed nuclear cycle (CNS) with regeneration of plutonium from the spent fuel assemblies to be used as fuel for the fast breeder reactors.

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Mixed fuel for fast breeders A justified option for involving Pu into fuel cycle is

production of MOX-fuel pellets for fast breeder reactors. “Mayak” production union has accumulated experience in producing regenerated plutonium dioxide at the RT-1 plant and pilot production of fuel assemblies with pelleted MOX fuel for BN-350 and BN-600 reactors.

Total of 53 such fuel assemblies were tested to maximal burnup of 11.8% of heavy atoms with the damaging dose on the cladding of up to 82 displacements per atom.

Three experimental FA with pelleted MOX fuel for BN-800 reactor are being tested in the BN-600, the major design difference being the presence of an absorber instead of an end shield.

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Mixed fuel for fast breeders A promising line of development of the fuel

technologies is transition to the so called dense mixed fuels: – Nitrides, – Carbides, – Metal alloys and metal-based composite fuels.

A wide range of experimental fuel assemblies with various types of dense fuel was tested in the research reactors, including nitride and metallic mixed fuel.

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Unification of technologies and machine-building complex; Ensuring cost-effectiveness of the production and systematic

decrease of the fuel component of the cost of kilowatt-hour;  Preparedness of the technologies for industrial-scale

implementation; Maximal use of the existing production facilities in order to

minimize capital costs: - integration of elements of the CFC into the existing fuel cycle; - minimization of the amount of RW for ultimate disposal; -  minimization of transport expenses; - ensuring possibility of exporting technologies, products and

services; - possibility of stage-wise improvement of economic performance

and environmental friendliness.

Principles of establishing industrial-scale production of mixed fuel

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VNIIMN has developed a universal technology of production of pelleted MOX fuel based on the eddy mill pulverization process (EMP-process). The main underlying principle is dry mixing of uranium and plutonium dioxides in the electrmagnetic eddy mill. The technology introduced at Mayak production association is patented in Russia (RF Patent # №2262756) and in a number of countries abroad (Germany, Belgium, France, China etc).

Transition to production of pelleted dense fuel (e.g. mixed nitride) is anticipated without changing the key process equipment. Only an additional module for producing the required initial materials have to be constructed.

There is no alternative to fuel assemblies with pelleted MOX fuel

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Realization of principle of universal pellet production for MOX-fuel and (U,Pu)N fuel fabrication

Granulator

Sintering furnace

Hydraulic press

Facility on the basis of electromagnetic mixer

initialpowders

finishedpellets

moldingpowder

(U,Pu)Npowder

UO2, PuO2

(U,Pu)O2

Module of synthesisof nitride powders

MOX(U,Pu)N

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BN-800

Fuel elements with pellet MOX-fuel:

BN-600Upper head

PlugLock Pellet T3 Distance lattice Pellet T3Pellet A3Cladding Lower head

Pot

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EP 450 Ferrite-martensitic

steel

Austenitic steels

Irradiation-induced swelling – criterion of structural material choice

Damage doze, dpa

Sw

ellin

g,%

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Parameter Current status Stage 1 Stage 2

Fuel cladding material 06 Cr16 Ni15 Мo2 Mn2 Тi W B

(ChS-68 cw)

06 Cr16 Ni15 Мo2 Mn2 Тi W B

(ChS-68 cw)

07 Cr16 Ni19 Мo2 Mn2 Nb Тi V

(EK164 cw)

Lifetime duration, effective days 560-585 592 710/770

Interval duration between refuelings, effective days

14030 14830 14830

Number of operation intervals of FA basic array

4 4 5

Installed power factor, % 0,77-0,80 0,81 0,81

Maximal local fuel burn-up, % h.a. 11,2-11,6 11,7 15,0

Maximal damage dose, dpa 82-86 87 110

Fuel element maximal HGR, kW/m 47 47 47

Fuel cladding maximal temperature, С 700 700 700

Prospects of fuel burn-up improvement in BN‑600

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Parameter Design basis core Prospects

Fuel cladding material 06 Cr16 Ni15 Мo2 Mn2 Тi W B (ChS-68 cw)

07 Cr16 Ni19 Мo2 Mn2 Nb Тi V (EK164 cw)

Campaign, eff. days 465 570-620

Fuel cycle between refuelings, eff. days 155 143-155

Number of fuel cycles during FA campaign 3 4

Maximal local fuel burn-up, % h.a. 10,3 12,5-13,5

Maximal damage dose, dpa 90 110-120

Fuel rod maximal HGR, kW/m 48 48

Fuel cladding maximal temperature, С 700 700

Key specifications on the fuel operation in BN‑800

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Key specification of the fuel operation in BN‑1200

Parameter Stage 1 Stage 2 Stage 3

Fuel cladding material 16 Cr12 W2 V Ta N B (EK-181)

20 Cr12 Mo W V Nb N B (ChS-139)

ODS ODS

Campaign, eff. days 1320 1650 1980

Fuel cycle between refuelings, eff. days

330 330 330

Number of fuel cycles during FA campaign

4 5 6

Maximal local fuel burn-up, % h.a. 14,4 17,6 20,6

Maximal damage dose, dpa 133 164 182

Fuel rod maximal HGR, kW/m 46 46 46

Fuel cladding maximal temperature, С 670 670 670

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Reactor tests to validate serviceability of BN-1200 pilot fuel elements

Material science assembly, BN-600

dpa

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Spent nuclear fuel processing

Unified production of pellets integrates with water extraction technology of processing SNF from the thermal neutron reactors.

Continuity of the fuel cycle technologies is ensured during the transient period of nuclear industry development.

High degree of purification of the SNF from fission products (107 – 108) by means of water extraction technology minimizes the environmental impact of the fuel cycle and ensures acceptable radiological conditions during fuel manufacturing and handling of fresh FA at all stages.

It has been demonstrated on the West and is being confirmed in Russia that water extraction technologies of SNF processing can yield minimal amounts of RW for ultimate disposal due to use of advanced LRW conditioning methods. Ultimate amounts of RW from SNF processing depend on the properties of materials used for their immobilization (glass, ceramics).

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Development of dispersed fuel for floating reactors and small nuclear plants

The design of the core for pilot floating reactor is based on the channel type ice-crasher core KLT-40.

The KLT-40 core makes use of the fuel assemblies with highly enriched uranium (containing over 20% of 235U). In order to ensure export potential of floating reactors units and small nuclear plants with KLT-40C, fuel with the uranium enrichment below 20% had to be developed.

Development of fuel for FRU and SNPP was performed via modernization of the fuel assemblies of the nuclear ice crushers with proven design and technologies.

The designed FA is based on “UO2+aluminium alloy” (“cermet” fuel) having much larger uranium content than the fuel of atomic ice-crushers.

A complex of pre-core tests of the fuel assemblies has been performed, and their parameters in non-irradiated state were defined.

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Tests of the dispersed fuel for FRU and SNPP The newly designed fuel is being successfully tested in the loops of MIR

research reactor (NIIAR) as a part of Garland exposure module and within a full-scale FA. Two modules have been tested, tests of two more and of a full scale FA are ongoing, integrity of fuel is not compromised.

Post-core tests of the fuel assemblies with the burnup rate of up to 0.98 g/sm3 (150 MW day/kg U) have been performed. The tests demonstrated reliability and operability under the operating conditions of KLT-40C.

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1Н акопление осколков деления, г/см 3

0

2

4

6

8

10

Изм

ен

ен

ие

пл

ощ

ад

и т

опл

ивн

ой

ком

пози

ци

и,

%

Д иаметр (м м) и материал оболочки, что изм ерялось

5,8 ; Э 110; Sтк5,8 ; Э 110; D5,8; Э 635; Sтк5,8 ; Э 635; D

5,8; Э 110; 6,2; Э 635; D

On the left – microstructure of the fuel composition at the burnup rate of 0,89 g/sm3; on the right – FA core growth as a function of burnup.

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Tests of the dispersed fuel for FRU and SNPP

Thermal tests of the irradiated fuel and tests of behavior of the leaky irradiated fuel elements have been conducted in the MIR reactor. According to the results of the tests, cermet fuel is not inferior to the fuel of atomic ice-crushers in terms of radiation resistance during beyond design accidents and in terms of corrosion resistance in the leaky state.

As a result of design and process refinement, as well as pre-core and post-core tests, the technical deign of the 14-14 FA for the core of the pilot FRU has been issued and properly approved in 2007.

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