Accelerator Driven Systems for Thorium Utilisation in India

• S.B.Degweker, Pitambar Singh, P.Satyamurthy and Amar Sinha

Bhabha Atomic Research Centre, Mumbai, India

International Thorium Energy Conference

October 27-31, 2013

CERN in Geneva, Switzerland

Program for development of sustainable nuclear power in India

• Three stage program – Ist stage

• thermal reactors fuelled with uranium

– Fast breeder reactors • using Pu from the reprocessed fuel of stage 1 • Breed Pu to expand the program • Produce U233 by irradiating Th in the blanket

– 3 rd stage • Fast / thermal reactors based on a Th-U cycle

• Waste transmutation – Fast reactors – Accelerator Driven Systems

• Th utilisation – As part of the third stage – Alternative scenarios for early introduction / improved breeding

• Accelerator driven systems • Molten Salt Reactors

ADS Program Objectives: Th utilisation

• World Nuclear Scenario o Plenty of U available o Large fissile Pu availability

Little incentive for breeding

o Large waste volumes Strong incentive for waste

transmutation o Little incentive for Th use

• Present Indian Scenario o Very limited U availability

o Small fissile Pu base Strong incentive for breeding

o Small volumes of waste Less incentive for waste transmutation immediately Likely to change with expected large scale expansion

nuclear power program

o Large Th deposits Strong incentive for Th use Low transuranic waste generation.

o Three stage program PHWRs: Pu for fast reactors FBRs: Pu and Th breeding Th-U233 fuelled reactors

• What can ADS can achieve? o Faster breeding of U233 for use in critical reactors o Simplification of Th utilisation: once through cycle

Program for ADS development in India

• High energy and high current proton accelerator • Development of 30 mA 20 MeV Linac injector (LEHIPA)

• Development of High energy Linac (1 GeV)

• Spallation target and materials • Computational codes development and nuclear data for spallation

reaction analysis in the target.

• Thermal hydraulics computational tools development for LBE

target simulations.

• Experimental loops for validation of thermal hydraulics codes and

corrosion studies on window materials.

• Reactor Physics Development of Computer Codes and Nuclear data for ADS

Other theoretical studies

Experimental facility and studies

Fuel Cycle and Conceptual Design Studies for Th utilisation

Development of High energy and

high current proton accelerator

Proton IS 50 keV

RFQ 3 MeV

DTL 20 MeV

DTL/ CCDTL

Super- conducting

SC Linac

1 GeV

200 MeV

Normal Conducting

High current injector 20 MeV, 30 mA

Scheme for Accelerator Development for ADS

Design completed & fabrication is in progress

ECR Ion Source LEBT RFQ Drift Tube Linac

60 kW RF System 1.3 MW Klystron

Phase 1

Phase II

Phase III LEHIPA

50 kW RF Coupler

High Yield Neutron Facility

IS DTL

LEBT MEBT

RFQ

Proton Current = 30 mA

1.00E+14

6.00E+14

1.10E+15

1.60E+15

2.10E+15

2.60E+15

3.10E+15

3.60E+15

4.10E+15

0 5 10 15 20 25

Proton Energy (MeV)

Yie

ld (

Ne

utr

on

s/s

ec

)

Reflector(Pb)

Moderator

Beryllium target

Proton Beam

(20 MeV, 30 mA)

S0(EP) = 4.476 x 1011 x EP1.886 x I n/sec

Neutron Yield for Beryllium target

Related Activities in Indian labs • Accelerator physics

• Spoke resonators

• Superconducting cavities

• Helium vessels and tuners for SCRF cavities.

• Vertical and Horizontal Test Stands.

• Cryomodules for SCRF cavities.

• High power solid state RF amplifiers.

• Infrastructure for SCRF cavity fabrication, processing and testing.

• LLRF system, RF protection system and related instrumentation.

• Beam diagnostics, monitoring and protecting systems

• Cryomodule test stand (CMTS)

Development of Spallation

Target and Materials

-Very High Heat Deposition Density by proton

beam ~ few kW/cm3

-Very High Radiation Damage ~100 DPA or

more/year Embrittlement

Irradiation Creep

Void Swelling

Hydrogen Generation

Helium Generation

Transmutation

Solution for both these issues – Use

circulating liquid target

Possible Liquid

Targets

Elemen

t

Atomi

c

Mass

(A)

Atomic

Numbe

r (Z)

A/Z

Melting

Temper

ature

(0C)

Boiling

Temper

ature

(0C)

Densi

ty at

room

Temp

(g/cc)

Pb

207

82

2.52

4

327

1725

11.36

Bi

209

83

2.51

8

271

1560

9.80

LBE

~208

~82.5

~2.5

2

125

Similar

to

Pb/Bi

~10.0

Hg

200

79

2.53

2

-38.36

357

13.54

U

238

92

2.59

0

1132.3

3818

19.07

Ta

181

73

2.47

9

2996

5425

16.6

W

184

74

2.48

6

3410

5930

19.3

Hg is not

suitable due to

low boiling

temp. for

reactors

Pb or LBE seem

to be most

suitable target

materials

Mercury Loop • Simulation of

Window/Windowless Target

• Velocity field mapping by UVP monitor

• Carry-under studies

• Two-phase flow studies by Gamma Ray

• Laser-triangulation for free surface measurement

• CFD code validation

• Gas-driven flow studies

LBE Target Module

CYCLOTRON

VAULT

Target section

First Major Mile Stone of Indian ADS Target Programme

Target Experiments - Coupling With Cyclotron Proton Beam 30 MeV and

500 µA (CW) Neutrons generated: 4.12 X 10^13

-Coupling of Beam with

Target Module

-Window heat extraction

-Radioactivity Issues

-Gas handling

-Irradiation studies

-Combined Control &

Instrumentation

-Remote Operation

-Remote Dismantling

Status: Civil works in progress

Beam line procurement in

progress

Prototype target under

installation

Simulates 1Gev, ~3mA Proton

Window Heating

ADS Target Loop (~ 7m tall) for 30 MeV Proton Beam (under installation at BARC)

Separator tank with

remote dismantling

common flange

Window

Dump tank

Two phase Thermal Hydraulic &

mechanical studies

Variable thickness

LBE Thermal-hydraulic experimental test facility

(suitable for 100 MW Class-ADS Reactor)

VERTICAL

CANTILEVER PUMPSEPARATOR

PRIMARY HX

SECONDARY HX

THT DUMP TANK

DUMP TANK

INLINE HEATER

MIXER

RISER

CFB-LBE Thermal-hydraulic simulation facility-Under

construction

Flow rate-120 kg/s – Pump

Flow Rate – 40 kg/s – (Gas Injection)

Temperature – 3500 C

Coolant – Diphyl-ThT

Window Heat simulation-Electron/Plasma Heat source

Scope

-Thermal hydraulics

code validation,

-Primary coolant

development

-Corrosion mitigation

studies

-Component

development

-Diagnostic and C&I

development

-Operational

Experience

Thermal-Hydraulic

parameters suitable

to drive ~100 MW

ADS

Status: LOOP Under

Fabrication

Sub critical Reactor Physics

Activities

Experimental ADS facilities and studies

ADS Reactor Physics Activities • Experimental ADS facilities and studies

– The Purnima Sub-critical facility

• Basic Theoretical studies and code devlopment

• Studies on Th Utilisation in ADS – One-way coupled ADS concept

– Studies on starting ADS with naturally available fuel

– Th Utilisation in Heavy Water Moderated ADSs

– Th Utilisation in fast spectrum ADSs

– Th Utilisation in Molten Salt Reactors

– Breeding U-233 in ADSs for use in critical reactors

Facility for carrying out experiments on physics of ADS and for

testing the simulations is being set up. This will use 14 MeV

neutrons produced through D+T reaction.

Simple sub-critical assembly (keff=0.89) of natural uranium is

chosen

Measurements of flux distribution, flux spectra, total fission

power, source multiplication, and degree of sub-criticality will be

carried out.

For this purpose a 400 keV RFQ is being

built . Presently a 400 keV DC accelerator

Is used

ADS Experimental facility

For deuteron current of 1mA at 400 keV,

14 MeV neutron yield is 1.0x 1011 n/s

D+T reaction

The Purnima neutron generator

Accelerating voltage: 400 KeV

Target: Titanium deutride /tritide on copper substrate

Neutron production with tritium target:

Originally ~ 1.0e9 /s

Upgraded ~ 1.0e10 /s

BRAHMMA subcritical core after fuel

loading

Sub-critical facility for ADS Experiments, D+D and D+T , keff= 0.87

Reflector: BeO, Moderator: High Density Polyethylene.

Neutron Multiplication measured ( A. Sinha et al.)

BRAHMMA - “BeO Reflected And HDPE Moderated Multiplying Assembly

Sub critical Reactor Physics

Activities

• Basic Theoretical studies and code devlopment

Computer codes for ADS studies

– Inhouse development

• Monte Carlo codes – High and low energy transport

– Continuous energy

– Burnup

– Noise simulation

• Transport theory codes – Multigroup two and three dimensional transport theory codes

– Burnup

• Space Time kinetics codes

– Other codes in use (high energy transport)

• Fluka

• Cascade

Theoretical studies related to sub criticality measurements

– Theory of Reactor Noise in ADS

– Methods for determining alpha modes

• Useful in deciding detector locations in pulsed neutron and noise experiments for sub-criticality measurement

– Noise simulator

• For planning and analysis of noise experiments

– Simulation of pulsed neutron and noise experiments

Deterministic methods for sub-criticality measurements

• Pulsed Neutron Experiment

– Neutron pulse introduced periodically

– Decay of counts recorded in short time bins

• For determining ‘α’

• and ρ/β

– Results of simulations

• Source jerk method

– Source switched off after steady state operation

– Decay of flux observed as a function of time

– Does not require pulsing

– Experiment may have to be repeated several times

Pulsed neutron experiment simulation Detector located in reflector

Time response of counts on introduction of a pulse

0.01

0.1

1

10

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24

Channel number

Co

un

ts Delayed background

Prompt response

Total Counts

Pulsed neutron experiment simulation Detector located at zeros of modes

Noise methods • Do not require pulsing or switching off • Can also work if source is pulsed • Use of the following methods has been reported

• Feynman alpha • Rossi alpha

• Other possible methods • auto and cross correlation • Psd and cpsd methods

• All methods possible by – By recording time history of detection events – Off line analysis

• Difficulty – High degree of sub-criticality

• High efficiency requirement • Contamination from higher modes

• Simulator Development and results of simulation

Reactor Noise in ADS: New Theory

• Radioactive sources are Poisson sources due to – Large number of radioactive atoms – Relatively small number decay independently

• Accelerator sources are different – Pulsing – Cw accelerators – Fluctuations in intensity

• Typically a few per cent • For Poisson source of 1e8 strength should be only 0.01%

– Correlations in these fluctuations

• And are therefore non-Poisson sources • The difference is important in the interpretation of

noise based measurement • Requires a new theory

Reactor Noise in ADS: New Theory • Such a theory has been worked out in BARC

– All statistical descriptors commonly used for analysis • v/m, Rossi alpha, acf, psd,cpsd

– Spatial effects

– Finite pulse widths

– Delayed neutrons

– Probability generating function approach

– Langevin approach

• Some experimental evidence available

• Measurement of statistics of source is required – Non-Poisson character

– Preliminary measurements indicate non-Poisson nature

ACF Results from noise simulator

V/m and ACF: Analytical vs finite difference

Sub critical Reactor Concepts for

Th Utilisation and Breeding

ADS for power production

• First proposed by Carlo Rubia et al

– Thermal and fast energy amplifier

– Utilises Th in a self sustaining cycle

– Keff in the range 0.95-0.98

• If we want

– higher breeding rates

– Once through cycles

• Keff must be lower

• Severe power peaking near the source

• Problem for solid fuelled reactors

• Possible solutions

– Small reactors

– Multiple targets

– Fluid fuel reactors

Th Utilisation in ADS: One way coupled ADS • Power in ADS is inversely proportional to

sub-criticality and directly proportional to neutron source strength

• In the control rod free concept, the operating keff is limited to the range 0.95-0.98

• This requires accelerator beam power of about 10 MW

• The one-way coupled booster-reactor concept can reduce this requirement five fold – Inner fast core with source at centre

boosts the neutron source – These neutrons leak into the outer

thermal (PHWR/AHWR) core where they undergo further multiplication

– This cascade multplication gives very high energy gain

– Due to the absorber lining and the gap very few neutrons return to the booster – i.e. there is a one way-coupling between the two

– The one-way coupling ensures that the overall keff is limited to the desired value

– Consequently, accelerator power requirement for 750 MW(t) is ~ 1-2 MW

• Similar ideas have been studied in Russia – As an example, there is a recent proposal

for a waste transmuting ADS driven by an electron accelerator

Th Utilisation in ADS: Th burner concept

To Ultimate disposal

Main fuel: Th

Initial Seed: Nat. U, spent fuel, or Pu

Th burning ADS

(keff~0.95)

30 MWe accelerator input power

400 MWe output

340 MWe (net)

60 MWe feedback

Th Utilisation in ADS: Once through Th cycle in PHWR: The thermal Th burner

• Initial fuel: Nat. U & Th

• Normal refuelling of U bundles (say 7 GWd/t)

• Th will reside longer – U-233 generation adds reactivity

– Compensate by replacing some U by Th

• Th increases and U decreases

• Ultimately fully Th core – In situ breeding and burning Th

• Advantages – Use of natural fuels only

– 140 tons U consumption during reactor life

– High burnup of Th ~ 100 GWd/t

• Disadvantage – Low K ~0.9 and gain < 20 with Pb target – Accelerator power ~ 30 MW for a 200

MWe ADS

Once through Th cycle: The fast Th burner ADS • Advantages

– Good breeding properties of fast reactors • Lower parasitic capture on FPs and structures • Higher value of h e

Disadvantages Higher U-233/Th ratio required ~ 0.1 [against

0.015] Greater loss of U-233 when fuel is discharged

needing higher breeding rate

Requires initial fissile charge of fissile material

Very long irradiation time and high fluence exposure

Will need Minimum possible absorption in structures and

coolant

Higher discharge burnup ~ 400 GWd/t

On power fuelling facility with good shuffling

Metallic fuel gives higher value k~0.95 Coolant Na, Pb

Accelerator breeding of U233 for use in critical reactors • How far do we irradiate Th before discharging?

– Interesting nonlinear problem with interplay of various factors • Loss due to burning of U233 • Breeding gain due to fission of U233 • Losses to fission product captures • Shows multiple solutions in some situations

– Good choice of irradiation time • Power production for accelerator and grid • No significant loss of U233 production rate

– Long irradiation time limit takes us to the Th burner limit • Compared to thermal systems, fast systems

– Can go to much higher concentrations of U233 • Reprocessing costs are reduced

– Better breeding properties, Lower parasitic capture – Higher K – Greater power generation in power producing system – However longer irradiation times and higher fluence

• Production rate ~ 2kG per mA of 1 GeV p on Pb

U233 production rate, U233 fraction, power, and keff, for thermal blanket driven with 30 MW proton beam on Pb

U233 production rate, U233 fraction, power, and keff, for fast blanket driven with 30 MW proton beam on Pb

• MSR ADS concepts studied for long at Los Alamos by C.D.Bowman and coworkers

• One such scheme starts with pure Th fuel driven by a modest power (~10MW) proton beam on Pb

• They expected that such a reactor will reach full power of 200 MWe in about one year

• Our studies showed that actually it will take more than 5 years

• If however we use a mixture of Th and U we can get full power from day one

• In heavy water reactors the time required to breed the necessary U-233 is much longer ~ 20 years

• Hence using a mixture of Th and another fuel with a fissile species (say) natural U is more appropriate

Studies on Molten Salt Reactor (MSR) ADS

hh

xx

st 1ln)2()1(

1

h

x

xk

Monte Carlo burnup for MSRs • Developments for MSRs

– Automatic shuffling of fuel due to mixing of the fluid salt – Mixing time scales much shorter than the burn up time scales. – Salt concentration assumed to be uniform throughout the reactor at any given

time – one or two burn up zone – High accuracy in the computation of reaction rate tallies to obtain effective

cross sections.

• Fissile / fertile components continuously added to the system. • Continuous removal of fission products and actinides such as Pa • McBurn modified to take into account these processes.

– each nuclides deemed to have a removal decay constant in addition to nuclear decay constant.

– To simulate the addition of fissile or fertile species • Source terms added in the burn up equations • source terms adjusted at the beginning of each burn up step

– to maintain a prescribed value of keff – total actinide content in the system.

Studies on MSRs: The thermal MSBR

• Vertical section of the one salt thermal 2500 MWt MSBR – Nuttin et al [16].

Thermal critical MSBR • Removal time for rare gases and noble metals

– 1 minute

• Chemical processing time for salt: 10 days – Efficiencies

• Pa: 100% • Halogens: 20% • Zr and Semi noble metals: 5% • Alkali and alkaline earth metals: 1%

• Thorium content – Initial mass: 67 tons – Thorium in-flow adjusted to maintain constant total actinide mass

• U233 content – Initial mass: 1.08 tons – In-flow adjusted to maintain criticality – Important because breeding ratio depends upon keff

– Results – Evolution time studied: 50 years

– net extra 233U produced ~ 1400 kg

– Initial breeding ratio ~ 1.037 – Initial doubling time ~ 34 years

Thermal MSBR ADS

• Same as one salt thermal system considered before

• U233 concentration reduced to maintain k~0.96

• Driven by central spallation source

– 1 GeV protons falling on Pb

• Evolution studied up to 10000 days

– Breeding ratio ~ 1.14

– for an operating power of 2500 MWt • very powerful accelerator (about 50 MW)

• linear doubling is about 9 years

– With the 30 MW accelerator being planned • system power ~ 1500 MWt

• the linear doubling time increases to 15 years

• exponential doubling time a little over 10 years.

Summary of Studies on MSRs

• Studies on thermal, fast and ADS MSRs.

– The fast critical MSR gives a higher breeding ratio than thermal MSR

• smaller specific power,

• no advantage in terms of doubling time.

• lower demands on the fission product removal capability,

– Accelerator driven thermal sub-critical MSR

• good breeding ratio as well

• High specific power (lower fuel inventory for a given power).

• more studies required to decide upon the best option with regard to thorium utilization by the MSR route.

Summary • Program for ADS development in India

– Accelerator development

– Target development and ADS

– Reactor Physics

• Th utilisation studies in ADS indicate

– Possibility of breed burn ADS reactors • Power production

• U-233 production for use in critical reactors

– Solid fuelled reactors • Capture losses in fission products

• Severe power peaking

• Good in the short run for small demonstration systems

– Fluid fuelled ADS reactors (MSR) • Solve many of these problems and more suitable in the long run

Thank You