Studies of a Possibility to Include Modular Helium … resistant coatings...

1

Studies of a Possibility to Include yModular Helium Reactors with High Burnup of Microfuel

in the Closed Fuel Cycle

presented by Peter FOMICHENKO, National Research Centre “Kurchatov Institute”

[email protected]

IAEA’s Technical Meeting on the Development of ‘Deep-burn’ Concepts Using HTGR Coated Particle F el for the Incineration of N clear Waste S rpl s Fissile Materials and Pl toni mParticle Fuel for the Incineration of Nuclear Waste, Surplus Fissile Materials and Plutonium

without Recourse to Multiple Processing Vienna, Austria, August 5-7, 2013

2

Strategic issues of Nuclear Power development and the role of MHRs

Different fuel cycles on the basis of a unified core design Different fuel cycles on the basis of a unified core design Peculiarities of fuel cycle studies (Pu disposition option) Introduction of MA in the loaded MHR fuel Introduction of MA in the loaded MHR fuel Workscope definition and approach for a Deep Burnup study Examples of conceptual developments p p p

3HTGRs in the Russian Strategies of Power Development

In 2009 the Russian Government adopted the Russian Federation EnergyStrategy for the period until 2030, in which goals and tasks of long-termpower industry development were determined

Th St t id d l t f h d d ti i- The Strategy provides development of hydrogen production usingnuclear power and construction of high-temperature gas-cooled nuclear reactors

The Strategy of nuclear industry development in the Russian Federation iscurrently under revisioncurrently under revision

- Previous version (2000) provides the concept of the closed fuel cyclebased on the fast reactor technologybased on the fast reactor technology.

- It requires estimation of the HTGR role and place in the nuclear power industry

4Modular Gas Cooled Reactor is a Heat Source with the Versatility to Support Multiple Missions and Multiple Fuel Cycles

5GT-MHR Mission Statement

The GT MHR is a jointly funded project aimed at solving one of the The GT-MHR is a jointly funded project aimed at solving one of the most important tasks in the area of non-proliferation weapons of mass destruction – disposition of weapons-grade plutonium (PMDAmass destruction disposition of weapons-grade plutonium (PMDA Agreement 2000)

The GT-MHR technology provides efficient plutonium disposition and may also be used for development of a new generation of nuclear reactors with efficient electricity and hydrogen production, industry applications, and disposition of actinides from light-water reactor spent fuel

6GT- MHR Technical Concept

GT-MHR Technical Concept isbased on the following advancedgtechnologies Modular helium nuclear reactors,

in which core destruction and melting is impossible

Ceramic fuel design in the form of small coated particles with s a coated pa t c es tthermo-radiation resistant coatings

Modern technologies based on Modern technologies based on large gas turbines

Electromagnetic bearings High efficiency compact heat

exchangers

Reactor building

7GT-MHR Fuel Design

Fuel particle

Dense pyrocarbonSilicon carbideDense pyrocarbonP b

0.

65 -

1m

m

Porous pyrocarbonFuel kernel (Pu oxide, other HM)

12.5 mm 800

mm

m

Fuel compactThousands of coated Fuel assembly

25-5

0 m

m

fuel particles in a graphite matrix

•Thousands of fuel •compacts•Structural material –reactor graphite

Fuel is operable at the temperatures up to 1600C About 1 billion fuel particles of this type were fabricated and tested in Russia

8GT-MHR fuel cycles: features and advantages

Coated fuel particles on the base of WPu with ceramic protective coatings ensure the retention of fission products upprotective coatings ensure the retention of fission products up to high temperatures 1600 oC)

Th GT MHR tilit i l ti t th f l l The GT-MHR possesses versatility in relation to the fuel cyclefrom the viewpoint of fuel type. Uranium, thorium, as well as civil or weapons grade plutonium can be used for core loading

The high level of nuclear safety is reached under using of different fuel without changing of core design

Coated fuel particles on the base of different fuel determines the deep burnup and proliferation resistancethe deep burnup and proliferation resistance

Possibility to Use Different Fuel Cycles in GT-MHR 9

Compositions of weapons grade Pu, civil Pu, Pu in MOX fuel, and also fuel on the basis of low- and high-enriched uranium (LEU and HEU) were considered

Isotopic compositions of fresh fuels

tion

Fuel type Fuel composition Enrichment on fissile i t i U P k l

Fraction of fertile material i f l iti

Op yp p isotopes in U or Pu kernels in fuel composition

1 WGPu Pu-238 (0.1%)+Pu-239 (91.7%) + Pu-240 (6.6%) + Pu-241 (1.2%) + Pu-242 (0.4%)

93 % (Pu fissile) 7 % (Pu-240 and Pu-242)

2 LEU U-235 (14 %) + U-238 (86 %) 14 % (U-235) ~ 86 % (U-238)

2 HEU + Th WGU (15 %) + Th (85 %) 93 % (U-235) ~ 85 % (Th) ~ 1 % (U-238)

4 Ci il P Pu-238 (1%)+Pu-239 (59%) + Pu-240 (24%) + 70 % 29 %(P 240 d P 242)4 Civil Pu ( ) ( ) ( )Pu-241 (11%)+ Pu-242 (5%) (Pu fissile) 29 % (Pu-240 and Pu-242)

5 MOX WGPu (50 %) + U-nat (50 %) 47 % (Pu fissile)

50 % (U-nat) + 3.5 % (Pu-240 and Pu-242)

10WPu disposition scheme

11WPu disposition features

Change of Pu isotopic content in GT-MHR vs other reactor types

Generated GWe*yr at disposition of 50t WPu

(without recycling)( y g)

Burn-out of Pu-239, in % from initial quantity

12GT-MHR has high potential for Pu destruction and proliferation resistance

High degradation of Pu isotopic composition at burnup makes it High degradation of Pu isotopic composition at burnup makes it unattractive for weapons

Fuel particle refractory coatings make fissile material retrieval Fuel particle refractory coatings make fissile material retrieval difficult

Low fissile material volume fraction makes diversion of adequate Low fissile material volume fraction makes diversion of adequate fissile material quantities difficult

Neither a developed process nor a capability for separating fissionable material from GT-MHR spent fuel

13Fuel cycle studies - Qualification of design decisions

Core design Fuel composition design Zoning of fuel and poison Zoning of fuel and poison

loadings in the core volume

CPS positions Fuel reloading scheme Axial shifts of fuel columns

Пирокарбид (PyC)

Карбид кремния (SiC)

Пористый углерод (PyC)

Т й Топливная частица

Топливный керн

ОКБМ

Топливный компакт ТВС I типа ТВС II типа

14Fuel cycle studies – Safety considerations

Temperature coefficient of reactivity Possible positive values for Pu fuel

without fertile and inert diluents

4567891E+5 400

д.

without fertile and inert diluents Choice of optimum burnable Er loading

for compensation (‘Pu 239+Pu 241 mass’ to ‘Er 167 mass’

56789

2

3

1E+4

Xec

135*1/100

Фтеп

300

(Фтеп)

, отн

.ед( Pu-239+Pu-241 mass to Er-167 mass

ratio)

89

2

3

45

1E+3ие, барн Er

a167

200 ых нейтронов

-1E-005

0

1E-005

ти, d

k/dT

. 105 ,

/К

2

3

45678

Сечен

9f

потока

тепловы

4E 005

-3E-005

-2E-005

иент

реактивност Область неопределенности

3

4567891E+2

K K

40a

100

Плотность

п

-6E-005

-5E-005

-4E-005

турный коэффици

начало циклаконец цикла

2 3 4 5 6 7 89 2 3 4 5 6 7 89 2 3 4 5 6 7 890.01 0.10 1.00 10.00E, эВ

2

1E+1

300

K

1200

K

0

300 400 500 600 700 800 900 1000 1100 1200Температура, К

-8E-005

-7E-005

Температу

Studies of reactivity temperature effects at Big Physical Facility (BPF) 15

Proving the possibility of performing reactivity

temperature effects studies for the test specimens in the central graphite insert at BPF.1. Sample outside the core graphite insert at BPF.

Creation of the critical configuration at BPF with a

Reflector ~ 400 mm

Neutron Spectrum

configuration at BPF with a central graphite insert, where the GT-MHR reactor neutron

Active core~ 1000 mm

2. Sample in thecore

SchemeScheme ofofexperimentsexperiments onontemperaturetemperatureeffectseffects

1.E-01

1.E+00

compact

91-3

91-4

spectrum is simulated.

Performing the experiments on

1.E-03

1.E-02

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

g preactivity temperature effects for the test specimens with different Pu isotopic compositions

Top view

Cooling air inflow 48 mm

Energy (eV) Pu isotopic compositions including the effects of Erburnable poison. 51 mm

B

A

16Fuel cycle studies - Operating conditions

Temperature distributions Estimate local power peaking factors Estimate effects of fuel zoning and Estimate effects of fuel zoning and

column shifting Provide data for TH analysis BOC

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5

Fuel temperature, oC

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5

radial direction from core center

coolant flow direction

EOC

core center

17Fuel cycle studies - Input data for fuel performance analysis

Fuel performance Fuel composition

vs. burnupFuel composition

p Generation of fission

products Fluence calculations

63 15364-gd-15868-er-166

54-xe-13155-cs-13758-ce-14160-nd-14562-sm-14763-eu-153

Fast Neutron Fluence

45-rh-10347-ag-10948-cd-11650-sn-12051-sb-12652-te-132 Fast Neutron Fluence

vs burnup

32 ge 7234-se-8237-rb-8539-y-90

41-nb-9443-tc-99

45 rh 103

Fission Products Histogram

ОКБМ

0 0.0004 0.0008 0.0012 0.0016 0.002

32-ge-72

18Fuel cycle studies – example: Pu fuel design alternatives

№ K l iti K l Di t № Kernel composition Kernel Diameter

0 PuO1.65 (100 %) – reference fuel, SiC in TRISO 200 µm

1 PuO (100 %) as reference but ZrC in TRIZO 200 µm1 PuO1.65 (100 %) – as reference, but ZrC in TRIZO 200 µm

2 PuO1.65 (15 %) + ZrO2 (77,5%)+ Ce2O3 (7,5%) 450 µm

3 PuO1 65(15%) + UO2(15%) + ZrO2(70%) 400 µm1.65( ) 2( ) 2( ) µ

4 PuO1.65(15%) + ThO2(15%) + ZrO2(70%) 400 µm5 PuO1.65(15 %) + ThO2(15%) + ZrO2(40%) + Ce2O3 (30%) 400 µm6 P O (15 %) Z O (85 %) Z C tt l 400 6 PuO1.65 (15 %) + ZrO2(85 %) + gZrC getter layer 400 µm

7 PuO1.65 (15 %) + ZrO2(85 %)+ gZrC in ВРуС 400 µm

8 PuO1 65 (25 %) + ThO2(75 %) + gZrC getter layer 300 µm8 PuO1.65 (25 %) + ThO2(75 %) + gZrC getter layer 300 µm

9 PuO1.65 (25 %) + ThO2(75 %) + gZrC in ВРуС 300 µm

Fuel cycle studies – Technical basis for selection of fuel design 19

0.01

0.11

ts

0 0001

0.001

0.01

н. ед.

3 Базовый4

re,

rel.

unit

ReferenceDiluted 50

43

1E-005

0.0001

азрушения

, отн

2

WP

oatin

gs fa

ilu

Diluted & Gettered30

40

%

1E-007

1E-006

Вероятность ра

6

8

9

abili

ty o

f co Gettered

20Доля ТК

, %

20

1E-009

1E-008

5

7P

roba

10

14

107

6

20 40 60 80 100Выгорание, % FIFA

1E-010

Fuel burnup, % FIFA100 200 300 400 500 600 700 800 900 1000

Выгорание, МВтсут/кг

0

60930

Fuel burnup, MW·day/kg Pup y g

20GT-MHR Selected Basic Fuel

The Coated Particle Design Achieves the RequiredHigh Burnup by the Use of

I t Dil t i th F l K l• Inert Diluent in the Fuel Kernel• Oxygen Getter Coatings• TRISO Coatings

SiC

SiC SiC

SiCSiC

SiC

SiC SiCZrO2

PuO2-x

ZrO2

PuO2-x

SiC

SiC SiC2 2

SiC Oxygen Getter LayerSiC Oxygen Getteron Kernelin Buffer Layer

21WPu disposition and MA transmutation in the GT-MHR

Why WPu to transmute Minor Actinides in the GT-MHR?

Disposition Program presumes effective use of WPu. It is reasonable to admix some Minor Actinides from VVER (LWR)spent fuel

Use of WPu is advisable to provide a stabile Pu isotopic content to use in the GT-MHR (it is more difficult in case of civil Pu)in the GT MHR (it is more difficult in case of civil Pu)

In future it is possible to realize the closed fuel cycle on the basis of Pu from blankets of Fast Reactors

Use advantages of GT MHR flexibility to use different fuel types Use advantages of GT-MHR flexibility to use different fuel types


VVER RRMA-(Am,Cm)

VVER RREECCYY Micro particles

Spent fuel,cooling

CCLLIINN

Loading90%PuW+

10%МАNNGG

PuS P f VVER

Initial stage: Initial stage:

Store Pu from VVER for FBRs start-up

GTGT--MHRMHR

Discharge - 25%After 2020 year

Pu from FBR blanketPuPu from wfrom weaponseapons

Back EndBack EndResidual storageResidual storage

g

Heavy Nuclides

Residual storageResidual storage


Example: Example: LoadingLoadingEfficiency of WPu and MA utilization

90 % WPu+10 % МА90 % WPu+10 % МА

Discharge Discharge –– 10 %10 %90

100

Isotopes Isotopes inin spent fuelspent fuelPuPu--238238 -- 7.5 %7.5 %

Heavy NuclidesHeavy Nuclides

70

80

fuel

, %

PuPu 238 238 7.5 %7.5 %PuPu--239 239 -- 1.01.0 %%PuPu--240 240 -- 3.53.5 %%PuPu--241 241 -- 1.51.5 %%50

60

Bur

nup

PuPu--242 242 -- 54.354.3 %%AmAm--241 241 -- 0.30.3 %%AmAm--242 242 -- 0.10.1 %%30

40

0 5 10 15 20 25

AmAm--243 243 -- 16.316.3 %%CmCm--244 244 -- 15.215.2 %%CmCm--245 245 -- 0.30.3 %%

MA in loading, %


MA activity for open and closed fuel cycle

Open cycle:Open cycle:by 2050 – 330 t (Pu-239 + Pu-241)

from VVERClosed cycle:Closed cycle:

by 2050 – 0,04 t (Pu-239 + Pu-241) from GT-MHR

Open fuel cycle Back End

- VVER StorageU

Open fuel cycle Back End

- VVER StorageU VVER StorageU Closed fuel cycle Back EndVVER RecyclingU

GT-MHR StorageMA

Weapons Pu

25Workscope definition – Studies of deep burning of actinides in MHRs

Consider possible scenarios of deep burning of actinides in MHR-type reactors (various sources of MA and driver fuel, fueltype reactors (various sources of MA and driver fuel, fuel management schemes before and after irradiation in a MHR, choice of the final disposition form)

Choose functionals characterizing the efficiency of transmutation. For the conditions of Russia, a variant of the use of WPu for burning the minor actinides from VVER spent fuel could be considered.

Develop the proposals for the MHR core layouts intended for deep burning of different MA, determine the possibilities of optimization aimed at increase of efficiency of transmutation (e.g. by the choice of fuel placement scheme).

26Workscope definition – Studies of deep burning of actinides in MHRs

Evaluate the uncertainties of the basic characteristics (e.g. neutron-physics or thermodynamic data) and determine their majorneutron physics or thermodynamic data) and determine their major sources

Evaluate the influence of introduction of MA-based compositions into microparticles on safety, e.g. reactivity feedbacks, including transient modes of operation

Analyze the possibility to use Bench Scale Facility for fabrication of coated particles with minor actinidesof coated particles with minor actinides

Determine the design data needs (e.g. in neutron-physics or Determine the design data needs (e.g. in neutron physics or fabrication technology) and evaluate the possibility to obtain them in experiments at facilities in Russia or elsewhere

27Incorporation of MA burning workscope

in GT-MHR Technology Demonstration Program

Technology Demonstration Program

V lFuelBench Scale Facility

PCU

Heat exchange Turbomachine

Reactor

Reactor graphite

CCCM

Vessel system

Design optimization

Fabrication ofU and Pu fuelIrradiation and post-irradiation examinations

equipmentTests of models for heat exchange elements

Fabrication

CCCM

Reactor and post-reactor studies of graphite and CCCM

Physical

Selection of materialsFabricationtechnologymastering

Fabricationtechnologymastering

Turbocompressor Rotor suspension t

Generator

Physical experiments and verification of codes

Investigations of fission products

Fabrication offorgings and samples

Certification tests

Tests of turbine modelsTests of compressor models

system

Mini mockup

Tests of sensors

Tests of EMB models

Insulation tests

Tests of electric penetrations

fission products release and transport

Investigation of materialsTests of stator seals

Coupling tests

Tests of EMB models

Tests of rotor scale model

Tests of friction couples

penetrations

Generator mockup tests

Coupling tests

Full-scale TC tests

pTests of catcher bearings

28Modifications in the Planning in GT-MHR Project

(or setting of a new Project planning)

From “Development Plans” to Workscope Descriptions with account for the

extension of the Project Mission by burning of MA

Example of a Closed Fuel Cycle based on Three Types of Reactors 29

Estimation of GT-MHR Efficiency for Burning of MAs and WG Plutonium 30

31Estimation of GT-MHR Efficiency for Burning of MAs and Civil Plutonium

32Estimation of GT-MHR Efficiency for Burning of MAs and Plutonium

The maximum possible content of MAs is about 25% forweapons-grade plutonium and 10% for reactor-grade plutonium,both are values are drastically higher than the MA contentassumed for VVER or fast reactor fuels (usually 5%).( y %)

Preliminary analysis of neutronic characteristics of the coreloaded with different-grade plutonium and additionally with MAsshows that:

in the absence of burnable poison in the core additionally loadedMAs make it possible to balance the reactivity change during the fuelMAs make it possible to balance the reactivity change during the fuellifetime and provide the negative temperature feedback.

average burnup of all transuranium isotopes is not less than50% h50% h.a.

initially loaded fissile plutonium in once-through pass in the corecan be burnt out by 60%. Burnout of all MAs varies from about 20% toy60% h.a.

33Technology Development Program. Reactor physics

ASTRA –critical test facilityt NRC K h t I tit tat NRC Kurchatov Institute,

Moscow

Test goalsTest goals Investigations of neutron characteristics of

annular cores (power distribution etc) Investigation of control rods effectivenessg After upgrade - Measurements of

temperature coefficients (20…600 оС)

34Technology Development Program. Reactor physics

Big Physical Facility (BPF) –critical test facilityt IPPE Ob i kat IPPE, Obninsk

Test goalsTest goals Investigations of characteristics for HTGR

fuel compositions with Pu and MA Special measures to simulate HTGR p

spectrum at fast critical facility

FUEL PROCESS DEVELOPMENTBENCH SCALE FACILITY REPRESENTS ALL FUEL FABRICATION STEPS

FUEL PROCESS DEVELOPMENTBENCH SCALE FACILITY REPRESENTS ALL FUEL FABRICATION STEPS

Solution and Fuel Kernel Fabrication Glove Boxes

Emission Spectrometer for Quality Control

Fuel Kernel Processingy

CoatersOff Gas Treatment System

Fuel Kernel Sintering Furnace Bochvar Institute, Moscow

36Technology Development Program. Fuel Quality Control

OSA facility at NRC Kurchatov Institute, MMoscow

Studies of the fission product release from irradiated coated fuel microparticles and HGTR fuel elementsHGTR fuel elements

- express-analysis of radionuclide migration in the process of fuel fabrication

- Xe133, Xe135, Kr85, I133, I135,- solid: Cs134, Cs137, Ba140, La140, Nb95, Zr95


QC equipments at NRC Kurchatov Institute, Moscow

- diameter and non-sphericity- coating thickness

density measurements- density measurements

т.

40

т. 25

30

исло

частиц,

шт

10

20

30

Число частиц

, ш

5

10

15

20

Чи

мкм360 380 400 420 440 460

0

Ч

Коэффициент несферичности1 1,02 1,04 1,06 1,08

0

5


Planning of Irradiation Tests

design of capsules- design of capsules- selection of samples and pre-irradiation control

Instrumented Irradiation Capsule

FUEL QUALIFICATIONIRRADIATION AND POST-IRRADIATION EXAMINATION FACILITIES

Post Irradiation Examination Facilities SM-3 and RBT-6 Test Reactors

52

6ID 3

Post Irradiation Examination Facilities

4

44

3

3

1

2

17

Instrumented Irradiation Capsule

Well-Equipped Hot CellsCapsule NIIAR, Dimitrovgrad

40Conclusions

Hope on the versatile role of MHRs in the strategy of Nuclear Power development should be confirmed by comprehensive studies

It is possible to study different fuel cycles on the basis of a unified core design

Peculiarities of fuel cycle studies and technology issues should be thoroughly addressed

Investigations of introduction of MA in the loaded MHR fuel could be started as a neutron physics exercise (including nuclide flows)be started as a neutron physics exercise (including nuclide flows)

Workscope definition and approach for HTGR Deep Burnup studies should be consistently elaborated taking into account nationalshould be consistently elaborated taking into account national nuclear policy

Examples of conceptual developments are very useful for scientific p p p ydiscussions and elaboration of recommendations

Studies of a Possibility to Include Modular Helium … resistant coatings...

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