P. Puthiyavinayagam

Associate Director, Core Design Group

Indira Gandhi Centre for Atomic Research

Kalpakkam, India

Status of Fast Reactor Development in India:

April 2012 – March 2013

46th Annual Meeting of TWGFR, IAEA

IAEA Headquarters, Vienna, May 21-24, 2013

Energy Source

Installed

Capacity

(MWe)

Capacity

(%)

Generation

Apr12-Mar13

(BU)

Generation

Apr12-Mar13

(%)

Thermal (Coal,

Oil & Gas)151530 67.8 760.4 83.9

Hydro 39491 17.7 113.6 12.5

Nuclear 4780 2.2 32.9 3.6

Renewables 27542 12.3 -- --

Total 223343 100.0 906.9 100.0

Current Electricity Scenario

2012-13

Planned capacity addition:17,956 MWe Achieved: 20,623 MWe

Planned Generation: 930 BU Achieved: 912 BU

20 NPP in operation : Capacity factor - 80%, Availability factor - 90%

7 reactors under construction:

2 PWRs expected this year; PFBR under construction)

Strategy for Nuclear Power Growth

Results of a typical case study

1980 1995 2010 2025 2040 2055 2070

0

100

200

300

400

500

600Installed capacity (GWe)

Year

Power profile of PHWR programme

Growth with Pu-U FBRs

Further growth with thorium

FBR Program in India

• Indigenous Design & Construction

• Comprehensiveness in development of

Design, R&D and Construction

• Synthesis of Operating Experiences

• Focus on National & International

Collaborations

• Emphasis on sustaining quality human

resources

• Concepts and innovations to enhance safety

of SFRs

PFBR• 1250 MWt • 500 MWe• Pool Type• UO2-PuO2• Indigenous • From 2014

10 09

07 08

06

04 05

03

02 01

11

12

Ø11950

CFBR• 500 MWe• Pool Type• UO2-PuO2• Twin units• Indigenous • From 2023…

Future FBR• 1000 MWe• Pool Type• Metallic fuel • Serial constr.• Indegenous• Beyond 2025

FBTR• 40 MWt • 13.5 MWe • Loop type • PuC – UC• Design: CEA• Since 1985

Metallic Fuel Development

Activities

FBR-1&2�Design finalisation of reactor

assembly comp.�Component level technology

development�Safety systems design &

development� Integrated RA technology

demonstration (reactor grade)�Capacity expansion – through

EU based SFR as an option

PFBR�Construction�Commissioning �Component testing

& qualification�Design of special

devices for tests�Demonstration of

higher safety margins

�Operation strategy

FBTR�Operation�Fuel & Material Irradiation and Data generation (PIE)�Testing of PFBR devices�Life Extension

SFR & Fuel Cycle Activities - In a Nutshell

Metal Fuel Program� Test pin irradiation (Nat U-Zr; EU-Zr, U-Pu-Zr)� Plan for irradiation in PFBR to test design & safety aspects � Reference core and fuel pin design for test reactor (320

and 115 MWt)� Strategy for MFBR

R & D • Component development &

testing• Testing for demonstration of

safety margins• Clad Materials development• Sensors - Development • Safety engineering

SFR & Fuel Cycle Activities - In a Nutshell

Fuel Cycle• FRFCF awaiting

Govt sanction• Reprocessing –

DFRP• Pyro-processing

development

Human Resources Development & Collaborations

� Reactor Power:20.3 MWt; High power operationfor 1678 h with TG (1406 h)

� Electrical power generated 4.0 MWe

� Energy produced: 32.5 GWh thermal (3.3 BU)

FBTR Operation Highlights

AERB has, in principle,

agreed for extension of

operating license up to

31st March 2018

• Testing of Prototypes of High Temperature

Fission Counters for PFBR

• Irradiation of three pins of sodium-bonded

EU-6%Zr metal fuel started

• Irradiation of structural materials for FR

• Yttria capsule Irradiation - discharged for

separation of 89Sr

• Irradiation of ferro-boron

• Scram drop time measurement and testing

the industrial version of KALMAN filter for

reactivity

20th Campaign Core

FBTR Future Activities & MOX Fuel Irradiation Performance

(a) Fission gas release and

gas composition

(b) Pellet central hole

(C) Fuel pin temperature 0

10

20

30

40

50

60

70

80

90

100

0 4 8 12

Ga

s c

on

ce

ntr

ati

on

(%

)

BU (atom %)

%He %Ar %Kr %Xe % release

MOX Fuel Performance• A 37 pins SA with PFBR type MOX fuel

irradiated in FBTR up to 112 GWd/t• Fuel performance analysis carried out

and compared with PIE results• Important parameters, like fission gas

pressure, clad strains, pellettemperature distribution, central holediameter comapred which isreasonably satisfactory

• Results indicate that good marginexists between the analysis and thedesign safety limits for the target burn-up of 100 GWd/t

Future Activities

• Effect of bowing on core reactivity – experiments

• Irradiation of sodium bonded ternary (U-Pu-Zr) metal fuel pins

• Other irradiations planned : TRISO coated particle type of fuel for CHTR, sol-gel fuel and structural materials for fast reactors.

• Post Fukushima retro-fits have been identified and action will be initiated shortly.

1.6

1.8

2

2.2

2.4

0 50 100 150 200 250

Axial distance from active core bottom (mm)

Fu

el

pe

lle

t ID

(m

m)

at

EO

L

0

500

1000

1500

2000

2500

0 1 2 3 4Tem

pe

ratu

re d

istr

ibu

tio

n a

t P

PL

(d

eg

C)

Fuel pin radius (mm)

BOL

EOL

a b c

PFBR Project – Current Status

Turbine FloorReactor Vault TopDummy CoreOverall Project

� Civil work for NICB, service water pump house, Horton sphere, ventilation stack, DG building and service building completed.

� Erection of major large diameter critical reactor components including rotating plugs completed

� Primary tilting mechanism and primary ramp has been fixed to grid plate. � Erection of fuel handling equipment in fuel building is nearing completion. � Pre-commissioning activities in progress for the completed systems. � All eight SG and four DHX erected. � Commissioning of all 4 emergency DG completed. � Outgoing transmission lines from switchyard have been charged and

connected to southern India grid. � Erection of turbine equipments are nearing completion� Commissioning of DM water, raw water, normal service water and emergency

service water systems have been completed.

PFBR - Major Activities

• Analysis towards demonstration of higher safety margins in design

• Design of special components for tests after criticality and designconfirmation

• Review of commissioning procedures (~ 60nos.) in progress by AERB SG

• Adequacy, appropriateness & format of the details provided for SafetyLimits, Limiting Safety System Settings (LSSS) and Limiting Conditionsfor Operation (LCO) for ‘Protective Instrumentation’ and ‘ReactivityControl & Shutdown Systems’ review by SG-Tech Spec

• Philosophy and Scheme of PSI & ISI Inspection for Components –Review by SG-ISI in progress

• Observations given by the PDSC – Action Taken Reports submitted

• Performance testing and qualification of major components such asInclined Fuel Transfer Machine, Transfer Arm, Absorber rod drivemechanisms and seismic qualification of many systems undertaken

• Lessons gained from PFBR - broad categories (i) material,(ii) fabrication,(iii) welding & (iv) inspection&testing –for future FBR design improvement

Methodology for Isothermal Testing of Sodium Systems During Commissioning Stage

Heat inputs & losses from Primary &

Secondary Systems

� Liquid sodium would be preheated to200°C and filled in primary andsecondary sodium systems. Afterinitial filling, it is essential to maintainthe sodium temperature above 150°Cto avoid freezing at any location.

� Before fuel loading, isothermal testingof primary & secondary systems at450°C planned

� Time required to bring the sodiumsystems to isothermal condition withthe only available heat source viz.operation of sodium pumps.

� Various heat losses from primary andsecondary sodium systems areconsidered during the heatingoperation and the scheme is finalised Evolution of temperature in primary and

secondary sodium during heating

Strategy for Criticality and Operation

� Currently, a strategy for criticality and operation of the reactor in stages is being evolved

� Initial test program, low power physics measurements, physics experiments at various power levels, engineering tests for design confirmation will be carried out.

� Two options are under discussion:

(i) Full core configuration with testing at various subset of power levels

(ii) Progressive core configuration with tests at every stage (30%, 50% and 100%)

� The reactor is scheduled to go critical in Sep 2014 and full power operation will be reached by April 2015.

� A comprehensive and detailed operation strategy and test programme is under discussion.

Design of DND & Sodium Void Special Assemblies

DND & Na void SA

• After commissioning, it is essential to testthe fuel failure detection system & todetermine sodium void worth

• DND system functioning & Calibration :Delayed Neutron Detector SA is designed.DND pin contains natural U+6%Zr metalslug in the place of actual fuel enclosed in aperforated clad. Rest of the pins are dummysteel pins.

• To validate the sodium void worth estimatedby theoretical codes, special SA for sodiumvoid measurement is designed. Twoassemblies are designed - one with sodiumfilled and the other with argon filled. Thevoid is simulated for 1000 mm of active fuelregion and can be loaded in any location inthe core.

Results of Few Studies

• Seismic qualification ofControl & Safety Rod DriveMechanism was furthercontinued to demonstratethe insertability of CSRduring seismic events.

• Fast drop tests wereperformed in water with andwithout seismic excitationunder OBE and SSE.

Drop Time for CSRDM (Without &

With Seismic Excitation) in Water

• Insertability of the control rod into the control subassemblywas demonstrated in all cases.

• Drop time, strain and acceleration were measured atimportant locations.

• Measured increase in drop time is 255 ms / 180ms underOBE/SSE respectively

Test facility

Seismic Qualification of CSRDM in Water

ISI System for Main Vessel & Safety Vessel - Testing

ISI vehicle being manoeuvred

through the knuckle-crown

transition during the trialsMock-up test facility

ISI vehicle into the inter-

space during the trials

ISI Vehicle

ISI Vehicle

MV & SV Sectors

Air Lock Chamber

Plug Handling

Chamber

� System integration completed.� Testing of modules at 150 °°°°C� Comprehensive mockup test

facility with 1/13th of MV and SV sectors

� Trials on the deployment conducted

� ISI vehicle lowered up to crown region with all designed motions including the cable-take up system

� Extensive functional testing and validation of the system in the mock-up facility is underway

Assessment of Core Stability with Reactivity Perturbations

• Stability of core with small and largereactivity pulse perturbations (step andsinusoidal)

• Reactivity pulses of 0.1$, 0.3$ and 0.5$ forone second duration considered - for fullpower operation and part load operation(40% & 20%). Sinusoidal reactivityperturbations of amplitude 0.5 $ andfrequencies 10 Hz, 1 Hz, 0.1 HZ and 0.01Hz – all reactivity feedbacks considered

• Range of reactivity values consideredenvelopes the entire possible range.

• The power does not show any oscillatorybehavior and stabilised in few secondsconfirming reactor stability in all cases

Analysis Towards Higher Safety Margins

Net Reactivity and Normalised Power with

Time (Power =100%,

Flow=100%, Perturbation = 0.1$

Analysis of UTOPA at Low Power Conditions

• UTOPA analysis was carried out at low power to check ifthe reactor comes to another steady state ultimately.

• The analysis was carried out for conditions

(a) at criticality with 50% primary flow through the core

(b) at criticality with 100% primary flow through the core

(c) at 20% electrical power operation (26.4% thermalpower and 50% primary sodium flow).

• From the study, it is concluded that for UTOPA initiated atlow power, the reactor ultimately goes to a steady statewith fuel, clad and coolant temperatures stabilising atlower values than that in case of UTOPA happening atfull power conditions.

• There is no fuel melting in this case.

Integrated Seismic Analysis of SGDHR

• To study the effect of equipment and vesselsconnected to SGDHR lines under seismicexcitation for both OBE and SSE.

• SGDHR main line, Dump line, DHX-A, AHX-A,Storage tank, and expansion tank are modeled

• Envelop response spectrum method is used

• The analyses considered both coupling anduncoupling the equipment and vessels from thelines

• The maximum stress intensity is 169.4 MPa

• The comparison between the coupled anduncoupled analysis reveal that the componentsmainly DHX and storage tank affected theresponse of the loop by affecting its modeshapes, while the other components did notaffect the line or had negligible effect.

Location of Maximum stress intensity

Fig.6:Von Mises Plot

Primary Containment Capacity against CDA

� Higher mechanical energy release during asevere CDA investigated from the structuralintegrity of primary containment and postaccident cooling aspects.

� Parametric study - range of work potentials100-1000 MJ - indicates that primarycontainment has high potential to withstandthe transient forces generated by energyrelease even more than 1000 MJ.

� The sodium level fall due to main vesselexpansion (large permanent deformation) isfound to be acceptable.

� Hence, high uncertainties in the energyrelease assessment are not of greatconcern.

� However the deformations of DHXimmersed in the sodium pool could limit theacceptable work potential - found to be 500MJ from the simulated experimental study

Decay Heat Removal Capacity

of SGDHR after CDA.

Vessel Deformation

Evolution of hot pool and AHX sodium outlet temperatures during prolonged station blackout

REACTOR

M

DUMP TANK

STEEL STACK

SODIUM-AIR HEAT EXCHANGER(AHX)

ARGON

EXPANSION TANK

AIR

NITROGEN

AIR

DAMPERS

Fig. 1: SAFETY GRADE DECAY HEAT REMOVAL CIRCUIT

AIR

T

ARGON

DHX

STEEL CASING

T

MHot pool

T

C

D

� It is essential to ensure that sodium does not freeze during a prolonged SBO.

�Two different SGDHRS operating strategies, viz., (i) controlled manipulation of sodium-to-air heat exchanger dampers of all the 4 AHX and (ii) sequential closing of AHX dampers one after the other

�Sequential closing of AHX dampers is found to be better, because of the number of damper operations required is less.

�By this strategy, with the available battery power and pneumatic power, each of the dampers can be operated 2-3 times and the decay heat removal operation can be continued for more than 10 days without any risk of sodium freezing.

Sodium Freezing during Prolonged Station Blackout

Evolution of neighboring SA thermocouple reading during TIB in a single fuel SA

0

2

4

6

8

10

0 10 20 30 40 50 60

Time (s)

Nei

gh

bo

rin

g S

A T

her

mo

cou

ple

Rea

din

g (

K)

1 2 3

4

5 SCRAM Limit

1 - End of sodium boiling in blocked SA

2 - End of clad melting in blocked SA

3 - End of fuel melting in blocked SA

4 - End of blocked SA hexcan melting

5 - Neighboring hexcan with 55 %

…..residual thickness

Subassembly power (P) = 8 MW

Hexcan thickness (δ) =3.4 mm

Investigation of Total Instantaneous Blockage in a Fuel SA

� During a TIB event, the clad, fuel and hexcan of the affected SA melt –damage radially propagates to other SA

� Knowledge of the number of SA that would get affected by this event (before automatic reactor shut down takes place by core temperature monitoring system) is essential to decide the thermal load on core-catcher.

� Various thermal hydraulic phenomena taking place during this process are complex, involving phase-change heat transfer, moving solid-liquid interfaces and progressive changes in the geometrical configurations of the affected SA.

� A robust one-dimensional thermal hydraulic model has been developed to understand the sequence of various events and the response of the core monitoring thermocouples.

It is established that detection of this event is possible at ~55 s after TIB and the residual thickness of neighbouring hexcan is 55% at that instant suggesting damage is contained within 7 SA.

Investigations of PSP Under Large Suction Re-circulation

Inlet

Outlet

Impeller

Diffuser

Axial velocity (m/s) at impeller eye

100 % speed & 100 % flow

(Negative velocity indicates flow is

downwards without recirculation)

Axial velocity (m/s) at impeller eye

100 % speed & 20 % flow (Positive

velocity around periphery indicates

flow is upwards with recirculation)

� Two primary sodium pumps operating in parallel supply sodium to coreat 7000 kg/s (head - 75 mlc)

� At the end of 50 hrs testing, erosion marks have been observed on thepressure side of all the 5 blades – possible reasons could be eitherrecirculation or non-uniform flow distribution at the impeller eye.

� Detailed CFD studies on (i) suction passage hydraulics, (ii) impeller-diffuser hydraulics and (iii) sump hydraulics, are carried out

� Suction passage renders a non-uniform velocity field at the exit- no.ofwebs has to be increased from 4 to 8 - Extra 4 webs are welded

� Further testing revealed that the performance is smooth.

700

800

900

1000

1100

1200

1300

1400

0 1 2 3 4 5 6 7 8 9 10

Time, s

Cla

d h

ots

pot te

mpera

ture

, K

8 pipe layout

8 pipe layout

SCRAM initiation time = 0.93 s (8 pipe layout)

SCRAM initiation time = 1.22 s (4 pipe layout)

4

Clad hotspot temperature (K) under a pipe rupture event

Finalization of Mechanical Design of Reactor Assembly Components of FBR1&2

• Thermal Hydraulic Analysis: Detailed multi-dimensional thermalhydraulic investigations for design verification:

• Pool mixing

• Gross temperature gradients across components

• Free surface sodium velocity & Gas entrainment

• Main vessel cooling

• Baffle configuration inside the header

• Integrataed thermal analysis of top shield and reactor vault

Temp. distribution on MV during normal operation

Stress distribution in inner vessel under normal condition

403525201596

-3-6

Von mises stress intensity of grid plate top plate under

mechanical loading

Structural Mechanics Analysis:

• Carried out for various loading levelconditions

• Design verification as per RCC-MR code

• Stress analysis of welded grid plate formechanical load

• Stress analysis of primary pipes with pumpheader and grid plate for the mechanicalloading and various thermal transients(slope in absorber location and max slopeof SA)

• Design satisfies RCC-MR code

• Effective damage due to creep-fatigueinteraction has been determined to be< 0.3.

• Design optimization studies of inner vessel

Finalization of Mechanical Design of Reactor Assembly Components of FBR1&2 …

Seismic Analysis:

• Integrated seismic analysis considers main vessel, inner vesseland top shield while the core subassemblies, core supportstructure and grid plate are incorporated by appropriate dynamicequivalent models.

• Analysis was done by response spectrum method to determinerelative displacement between vessels and stresses on thevessel. It is seen that the relative vertical displacements betweencore and absorber rods are acceptable with respect to reactivityvariations.

• The maximum vertical acceleration of core subassemblies is lessthan 0.9 g and hence there is no concern of detachment of wholecore.

• Also, the stress limits are respected at all the critical locations withcomfortable margin for OBE as well as SSE.

Finalization of Mechanical Design of Reactor Assembly Components of FBR1&2 …

Technology Development of Critical RA Components

• Technology development exercises have beencompleted for the key components of reactorassembly which have a bearing on themanufacturing time and in turn on the constructiontime.

• The tri-junction forging for dome shaped roof slab,large diameter bearing, thick plate welding forrotatable plugs and welded grid plate werecompleted in the last year.

• Inner vessel with redan of large single torus and30 m long tubes of steam generator have beensuccessfully completed in the current year.

• Successful completion of technology developmenthas demonstrated the manufacturing feasibilitiesand given confidence for the design improvementsincorporated in the reactor assembly componentsof future FBRs.

Optimised IV Shape

Die & Punch

Profile Check

Design of Ultimate Shutdown System

Ultimate shutdown system (USS) isenvisaged to meet the followingobjectives :

(i) to ensure shutdown of the reactorunder Anticipated Transients WithoutScram (ATWS) events, which in theabsence of shutdown will lead toCDA.

(ii) to limit the consequences followingATWS to Cat IV design safety limit

• Enriched lithium liquid / B4C granules is provided within a pressurizedchamber separated by a fuse plug. During reactor operation, the poisonmaterial is kept above active core.

• Under ATWS, the temperature of the coolant rises and melts the fuse plugonce the temperature exceeds the melting point of fuse plug. Six such SAcan achieve cold shutdown with (n-1) criteria.

• With B4C granules, the worth is more; experiment was conducted tounderstand the flow behaviour of the granules.

I & C Devices Development

Advantages: Less head space required for the removal of the transmitter and

Commercially available.

Radar Type Level Probes

Tests were carried out at different temperature.

Errors are with +/- 10 mm.

Endurance test was carried out at 550 °°°°C for 600 hrs.

Effect of sodium vapour deposit on antenna was observed to be

insignificant

Use of Wireless Technology

Advantages: (i) Avoid congestion in roof slab (ii) Avoid trailing

cable system (iii) Reduction in cable and connector box

Areas identified for deployment : (i) Reactor Assembly

Instrumentation (ii) Mobile fuel handling machines (iii) Trailing

cable system

Tests were carried out to establish its suitability with

commercially available products.

Metallic Fuel Development

• Three sodium bonded metal fuel pins of natural U- 6% Zr undergoingirradiation

• Testing of mechanically bonded natural U/ U- Pu fuel with Zr liner planned

• Need for data from fuel, subassembly behavior and fuel cycle relatedaspects

• Utility of existing facilities like FBTR and PFBR to address the needs

• Smaller dedicated test reactor for the development of total technologiesassociated with metallic fuel including reprocessing facilities may beconsidered.

• Design for reference cores of different power capacity done

• Engineering scale studies on pyroprocessing of irradiated metallic alloyfuels is in progress. U-Zr fuel pins were chopped and processed in anelectorefiner on 1 kg scale.

• An ambient temperature electrorefiner on 10 kg scale has beencommissioned.

• A plant for large scale fabrication of metallic fuel is being set up at IGCARfor partial conversion of FBTR core with metallic fuel SA

Reference Core Physics Design

Preliminary physics design ofreference cores

i. Sodium bonded pin designii. Relatively low critical massiii. Peak LHR close to 450 W/cmiv. Peak burnup of 150 GWd/tv. Single enrichment zonevi. Marginal breedervii. 3-4 experimental locationsviii. U-21%Pu-6% Zr for 320 MWtix. 15.5%Eu-23%Pu-6%Zr for115

MWt

Reactor Power: 320 MWt

: Core (30)

: RB (126)

: Reflector (102)

: CSR (6)

: DSR (3)

: Exp. SA (3)

: Exp. Special SA (1)

Reactor Power: 115 MWt

A. Performance Testing of Transfer Arm in Sodium

Tested at different conditions viz. air at room

temperature, hot air, hot argon and sodium at 200oC.

B. Sodium calibration of all PFBR probes – mutual inductance type (continuous and discreet type)

C. Fabrication and qualification of high temperature ultrasonic transducers for the Ultrasonic Under Sodium Viewer

D. Pin-on-disc type (rotary type) tribometer – To conduct tribology tests

E. Eddy current position sensor for DSR testing

F. Eddy current flow meter testing

G. Studies in Steam Generator Test facility

(test with plugged tube, heat transfer expt, flow instability tests)

H. Multi-purpose loop for testing of sodium components for future FBRs

Component & Device Development and Testing

A

C

D

E

IFTM Chain Test

Qualification of IFTM Chain:

• Two specimen - fatigue and break load testing -cyclic load 0.2 t to 1.5 t

• 97000 cycles completed (ASME, MandatoryAppendix:6) without failure

• Functionality not affected; After fatigue loadtesting, break load test carried out; lowest breakload is 27.1 t

• Accidental conditions – load may go up to 5 t –Large margin exists

Structural Mechanics R&D

Pressure Carrying Capacity of SGDHR Piping after SSE

• A typical pipe bend of the SGDHR piping systemwas tested to quantify the net margin availablein the design beyond SSE limits

• The pipe bend could withstand bending momentof 2.9 times the SSE without any leak in the pipebend.

• Subjected to 180 bars internal pressure –structural integrity reatined

Seismic Qualification of PFBR Components

• A number of components, I&C systemsand sensors have been seismicallyqualified in the shake table facility.

• Few major tests: IFTM gate valves, Fieldinstruments – pressure,flow, tempearture,level measurement

• Shake table test was carried out - 5 OBEfollowed by 1 SSE

• All components qualified

• Experiments were carried out with FerroBoron slabs having 10% and 5% naturalboron in KAMINI

• The foils used are Gold, Rhodium, Indium,Platinum, Copper, Hafnium, Cadmium andManganese.

• Fast, thermal and epi-thermal neutronattenuation pattern

Neutron Attenuation Experiments

Validation of Thermal Design of Top Shield

90

100

110

120

130

140

150

0 50 100 150 200 250 300 350 400

Circumferential Location (Angles in Deg.)

Tem

pera

ture

(oC

)

Roof Slab Inner Shell EL28300

Roof Slab Inner Shell EL28600

Circumferential Temp.

Distribution in Roof Slab-SRP

Annulus

• The Integrated Top Shield Test Facility whichsimulates reactor vault, reactor assembly andassociated cooling systems, was successfullyoperated continuously and the intendedobjectives have been fulfilled. (jet coolingsystem, effectiveness of wire mesh insulation,,tem. Evolution during loss of power, cellularconvection in narrow annulus etc.)

• Sodium aerosols will be released in to the atmosphere in case of Sodium leak from pipes of PFBR

• A test facility was constructed to study the atmospheric dispersion characteristics of sodium aerosols in open atmosphere

• Various sampling techniques used for characterization of the dispersed sodium aerosols

Test facility for Atmospheric Dispersion Studies of Sodium Aerosols during Sodium Fire

Sodium Fire Studies in SOCA facility

Objectives of SOCA facility

Evaluation of different scenarios of sodium firesresultant to CDAEffect of sodium fire on the integrity of importantsafety grade components like Decay HeatExchanger and its piping held in the TSP of PFBREffect sodium fire on the cable materials, and theevaluation resultant secondary fire consequencesDesign optimization of TSP with polar table likestructure for the future fast reactorsGeneration of benchmark data for the validation ofnumerical models developed

Salient features

Experimental Chamber to withstand hightemperature and pressure (500°°°°C and 10 bar)High Pressure Sodium Ejection System with 81nozzles to release sodium in few secondsIntegrated instrumentation system to monitor thetemperature and pressure rise during theexperimentHigh Speed Video Camera system to capture theevents for analysisEfficient Exhaust Gas Treatment System toremove the sodium aerosols from the exit air andrelease clean air into the atmospherePurging systems to conduct the experiments atdifferent oxygen concentrations

Argon Buffer Tank

Experimental Chamber Scrubber system

3D VIEW OF SOCA FACILITY

SOCA chamber Sodium release system

Performance Evaluation of Scrubber unit

Sodium aerosol inlet concentration = 945 mg/m3

Sodium aerosol outlet concentration = 1.3 mg/m3

Scrubber efficiency in removing sodium aerosol=99.9%

Sodium Spray fire Scenario Temperature and Pressure rise

Investigation on structural integrity of DHX

Experimental Results - SOCA facility

Peak temperature of air = 78oCPeak pressure of air = 1.07 bar

Time (s)

Co

nc. (m

g/c

u.m

)before after

Settling behaviour of sodium aerosol

Time (Days)

Co

nc

. (m

g/c

u.m

)

SOFISOFI Facility Facility –– Experimental Program Experimental Program

� Phase – I (Induction heating of notional mass)

� (U metal + SS) – Sodium system using melt mass ~ 1 kg

� (U oxide + SS) – Sodium system using melt mass ~ 1 kg

�Phase – II (Induction heating of small mass)

� (U metal + SS) – Sodium system using melt mass < 20 kg

� (U oxide + SS) – Sodium system using melt mass < 20 kg

�Phase – III (Plasma heating of large mass)

� (U metal + SS) – Sodium system using melt mass > 20 kg

� (U oxide + SS) – Sodium system using melt mass > 20 kg

Crucible, coil and release valve Crucible top assembly Facility view from control room

Sodium Safety Experimental Facilities Sodium Lab-View School

• Controlled & remotelyoperated water injection

• Hydrogen gas analysis

• Transient pressure andtemperature measurement

• SS vessel with toughenedglass viewing windows

• High speed opticalimaging

• SS steam vessel with 20 bar

design pressure at 200oC

• Remotely operated steaminjection

• Hydrogen gas analysis

• Transient pressure andtemperature measurement

• High speed optical imaging

• Ignition of sodium at

controlled atmosphere

• Leak proof Quartz chamber

• Gas analysis, Pressureand temperaturemeasurement

• High speed optical andthermal imaging

• Temperature and pressure

monitoring

• Controlled spraying ofsodium at 500oC

• Robust & compact sodiumsystem with Innovativesodium release valve

• Gas analysis, aerosol,temperature and pressuremeasurement

• High speed optical andthermal imaging

Sodium Pool Fire Sodium Spray Fire Sodium-Water reaction Sodium-Steam reaction

� Two sphere-pac test fuel pinsfabricated and qualified

� These pins contain coarser fractionof (U,Pu) MOX microspheres (780±70 µm) and fine fraction of UO2

microspheres (115 ± 10 µm)vibrocompacted into D-9 clad tube

� Six numbers of test fuels containingsodium bonded enriched U-6 wt.%Zr and alloy slugs were fabricatedas part of development of metallicfuels

Chemistry - Fabrication of Test Fuel Pins

Glove boxes for sphere-pac test pin fabrication

X-ray radiograph of sphere – pac test pins

Facility used for fabrication of sodium bonded test fuel pins with U-Zr and U-Pu-Zr alloy slugs

Set up used for top end plug welding in test fuel pins with metal fuels

� Spot technique equipment housed an argon atmosphere glove box and also qualified for radioactive material handling

� Solidus temperature of Mark-I carbide fuel containing 70% PuCmeasured

� Solidus temperature determined to be 2161 ± 4 K which is in agreement with the value ( 2148 ± 25 K) determined earlier at BARC, Mumbai by incipient melting technique. The present value is more accurate

Solidus-liquidus measurements on U-Zr alloys

Lid and orifice

Equipment used for solidus measurement on Mark-I carbide fuel of FBTR

� Chopping of sodium bonded test fuel pinscontaining natural U-6 wt.% Zr alloy slug in T-91clad tubes and Electrorefining of sodium bonded1 kg of U-6wt% Zr alloy carried out in thedemonstration facility

� U consolidated from the cathode deposit by distillation of salt and melting

� Lab. scale facility for studies on actinide draw down process commissioned in argon atmosphereglove box

� Equilibrations runs between LiCl-KCl-UCl3 and Cd-1 wt.% Li alloy at 500 deg. C being carried out

� Studies on equilibration of halide salts with zeoliteat 1 kg scale at 500 deg. C in a V mixer carried out

Studies related to Pyroprocessing of Alloy Fuels

Chopped pieces of sodium bonded U-Zr pins

Chopped pins loaded in anode basket

Expt. Set up for studies on Actinide draw down process

V- mixer used for studies on equilibration of salt and zeolite

• To recover 10B from nuclear grade boron carbide: Electroextractionmethod

• Process parameters were optimized and product was characterized forchemical purity, specific surface area, size distribution of particles andX-ray crystallite size

• Boron was recovered from boron carbide scrap (rejected pellets), Purityobtained: 92 wt. %.

Chemistry Studies on Boron Extraction

� Regeneration of cold trap by thermaldecomposition of sodium hydride

� Method based on monitoring H2 releasedduring thermal decomposition using a polymermembrane hydrogen sensor (PEMHS)explored

� Tin oxide based trace level hydrogen sensorfor monitoring hydrogen in argon cover gaswas demonstrated

Chemical Sensors

Materials & Metallurgy

� Ion Irradiation studies on ODS alloys

� Integrity Assessment of the Ferritic / Austenitic Dissimilar Weld Jointbetween Intermediate Heat Exchanger and Steam Generator

� Effect of Nitrogen Alloying on Low Cycle Fatigue and Creep-fatigueInteraction of 316LN SS

� Microstructural Analysis of 20% CW Alloy D9after Irradiation in FBTR

� Creep Crack Growth Characterisation of SS 316(N) and P91 welds

� Wall Thickness Loss in 316LN SS Due to Exposure to Sodium

� Long Term Corrosion Evaluation of Candidate Materials in SimulatedFBR Dissolver Solution using Mockup Zircaloy dissolver

� High Nitrogen 304L Stainless steels for Nuclear Waste Storage

� Thermal cycling studies on 9Cr-1Mo steel, 316LSS and Inconel 600

� Studies on Materials & Coatings for Pyrochemical Reprocessing

� Fabrication of irradiation capsules

• Several campaigns for reprocessing ofspent fuel of FBTR with a burnup of 155GWd/t were completed in CORAL

• An empirical model for Pu(VI)distribution coefficients in 30% TBP andits temperature dependency wereformulated. Comparison with literaturedata revealed a reasonably goodagreement between the reportedexperimental and model predictedvalues.

• For dissolution experiments in rotarycontinuous dissolver were carried outwith unirradiated UO2 pellets. Theresults were compared with batchdissolution data.

Comparison of Modeling results

with reported experimental Pu(VI)

distribution coefficients in PUREX

conditions

A view of 150mm 2-stage Rotary

dissolver inside fume hood

Reprocessing Studies

� CEA-IGCAR Collaboration on LMFBR Safety :

� Several collaborative projects undertaken

� Training in Phenix

� ERANOS 2.1, CAST2M, PLEXUS, TRIO-U – Access

� Implementing Agreements

International CollaborationInternational CollaborationInternational CollaborationInternational Collaboration

� AFCEN : RCC-MR codes

� India has adopted RCC-MR for PFBR

� Proposed few revisions based on PFBR experience

� CRP on Passive system reliability

� Control rod withdrawal test during Phenix EOL

� Participation in JHR

Participation in JHRParticipation in JHR

� Development of In-CoreSodium Loop for Irradiation ofMultiple Samples at hightemperature to generate datatowards numerical simulationof fuel pin behaviour(Advanced AusteniticStainless Steels& ODS andOxide fuel) for understandingof fuel safety issues

� Use in FBTR till JHR is readyto receive

� Performance tests at adedicated test sodium facilitydeveloped at JHR

In core sodium loop test facility

� Activities on development of science and technology forSFR and fuel cycle have been continued in an intensemanner.

� PFBR operation strategy is under evolution and supportfor commissioning, safety review, Tech.specs, ISI reviewactivities

� PFBR component testing completed� Analysis for establishing higher margin in design� CFBR design in progress – Mechanical design of RA

completed� Metal fuel program actively pursued� Fuel cycle activities catering to MOX and metal fuel are

in progress� Advanced material development studies� International collaboration in the domain of safety –

bilateral and through IAEA

SummarySummarySummarySummary

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