Fuel Safety Research Capabilities of the BR-2 Safety... · Reactor core performance of BR2 Design...

SCK•CEN/23581253 ISC: Restricted Copyright © 2017 SCK•CEN

Fuel Safety Research Capabilities of the BR-2

Brian [email protected]

GAIN Fuel Safety Research WorkshopIdaho Falls, May 1-4, 2017


Contents Introduction

BR-2 reactor

Fuel irradiation capabilities CALLISTO loop (past) Pressurized Water Capsule LUCIFER (concept)

PIE capabilities

Fuel irradiation experiments GERONIMO (past) ATTICUS (future)

BREASY

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SCK•CEN/23581253 ISC: Restricted Copyright © 2017 SCK•CEN3

BR-2 reactor

Hot-cellslaboratories

Radiochemistrylaboratory


General layout of the BR-2

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BR-2 geometry and characteristics

Light water cooled

Water+Be moderated

Inclined reactor channels Compact core Good access at cover

Reactor channels accessiblefrom top (all) and bottom (17)

Irradiation rigs in reactor channel axis of fuel element

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Reactor core performance of BR2

Design goal: thermal neutron flux up to 1015 n/cm²s Achieved byCompact core arrangement with central flux trapMaterial choice: Be moderator and metallic

uranium fuel

High overall core power up to 100 MW

Achievable flux levels (at mid plane in vessel) Thermal flux: 71013 n/cm²s to 11015 n/cm²s Fast flux (E>0.1MeV): 11013 n/cm²s to 61014 n/cm²s

Allowable heat flux 470 W/cm² is allowed for the nominal thermal

hydraulic conditions of the BR2 primary circuit Demineralised water: pressure to 1.2 MPa, temperature

35-50 °C 10 m/s flow velocity on fuel plate

Up to 600 W/cm² can be achieved in experiments

nth

nf

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Reactor configuration

Unique flexibility in configuration Position of fuel, control rods and experiments Type of fuel elements Reactor power and cycle length

Reactor load is optimised for each operating cycle 3D MCNP model with burn-up evolution of entire core Detailed model of experiment/production if required Verification by nuclear weighing before start

BR-2 reactor management and irradiations are ISO 9001 certified

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Typical core load of BR-2

Driver FuelElement Control Rod

Si NTD

99Mo PRF

192Ir Basket

FuelExperiment

MaterialsExperiment

Mid-plane cross section of a typical BR2 core

Thimble tube

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BR2 - then and now

Now: Major Overhaul completed in 2016 Replaced Beryllium matrix Performed major maintenance operations and inspections Updated instrumentation to meet future challenges Potential for improved operational regime (higher up-time)

BR-2 : in operation since 1963 Upgraded in 1968 to 100MW Refurbishment in 1977-1980 Decennial license review since 1986 Refurbished in 1995-1997

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First era of operation 1963-1978

Prototype experiments for Light water reactor Gas cooled reactor Sodium cooled reactor

First irradiations of MOX fuel

Production of isotopes for energetic applications

First replacement of Be matrix: 1978-1980

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Second era of operation: 1980-1995

Legislation change in 1984: no operational limit in license, periodicsafety re-assessment (10y periods)

Safety experiments Na cooled reactors Loss of Flow accident Post Accident Heat Removal

BR-3 shut-down (1987) PWR loop in BR2 LWR MOX studies

Instrumented material irradiations e.g. in-pile fatigue/creep testing

Second Be replacement 1995-199711


Third era of operation: 1995-2015

Irradiations in support of PWR: CALLISTO loop

Neutron Transmutation Si doping: installation of pool side facility for 6” and 8” crystals

MTR fuel test irradiations EVITA: qualification of JHR Fuel FUTURE series: LEU Fuel test

Fusion Material Irradiation

Third Be matrix change 2015-2016

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Fourth era of operation: 2016-2026+

Major Overhaul completed in 2016 Replaced Beryllium matrix Performed major maintenance operations

and inspections Updated instrumentation to meet future

challenges Potential for improved operational regime

(higher up-time)

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CALLISTO loop: past experience (1992-2015)

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Conditions 80-300°C, 10-157 bar Controlled water chemistry Peak power levels up to 430 W/cm Thermal balance, on-line power

Fuel types irradiated UO2, MOX, Thoria-MOX, IMF

Instrumentation Flux detectors Fuel centerline thermo-couples

(OMICO) Pressure transducers


LUCIFER: concept for LOCA bundle tests Fuel Fragmentation, Relocation and Dispersal for real ballooning

configurations LUCIFER bundle test captures effect of neighbouring rods Rod constraints on balloon extension Presence of guide tube Effect of burnable poison rods Effect of power / temperature gradients

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LUCIFER: experimental rig design

Capsule concept for LOCA simulation No forced cooling flow, No External cooling loop Safety strategy: Double wall irradiation rig

Pre-transient phase To ensure stable reactor power; LHGR of 1 to 3 kW/m Conditions: unforced cooling – Low system pressure

Transient phase Power level: representative of LOCA conditions Low flow rate (steam) to control oxidation (steam

starvation / oxidative conditions) Re-flood phase

Injection of water from bottom of bundle Reactor scram when close to COLD steady-state (pin fully

covered)

Double Wall Irradiation Capsule

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Pressurised Water Capsule (PWC) & Calorimetric Device (CD)

fuel pin


Pressurised Water Capsule (PWC) & Calorimetric Device (CD)

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Capsule and irradiation characteristics

Fuel pin dimensions Cladding diameters: 8 mm - 12.5 mm Fuel stack length: 20 cm - 100 cm (core height BR-2: 80 cm)

Capsule water pressure from 1 to 160 bar Heat transfer by natural convection at low power levels... ... combined with boiling and condensation heat transfer at high rod

power levels (depending on the pressure)

Steady state conditions or transient conditions Linear power levels up to ql,max = 750 W/cm Rod power variation by reactor power Power increase rate ∆ql/∆tmax = 100 W/cm/min Accuracy of the rod power can be measured within 4%

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PWC capsule temperature controlTclad(Prod,pPWC)


T0 T1 T2 T3 T4 T5 T6

Total : 3 – 4 months

2-4w 3-4w 1w 3-4w 2w 2w

LHMA Hot CellsVisual inspectionProfilometrygamma-spectro….

Proximity of laboratory allows for rapid project execution

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GERONIMO - Past experience with PWC

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Gundremmingen Evaluation & Research programme On Nine-by-nine MOX BWRfuel Base irradiation up to peak burnup 30-40

GWd/tHM

Ramp testing at final peak power levels 400-465 W/cm

Investigation of Fission Gas Release


GERONIMO core configuration

PWC loaded in position E30

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GERONIMO: Reactor power and pin linear power

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0 10 20 30 40 50 600

100

200

300

400

500R

od li

near

pow

er (W

/cm

)

Time (hours)

0

10

20

30

40

50

Rea

ctor

pow

er (M

W)


Gamma scan and spectrometry

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Profilometry-before and after ramp test

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Fission Gas ReleaseOM results

FGR Kr-85 before/after ramp 10.7 % / 20.9 %FGR puncture -/ after ramp - / 22.3 %

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PIE-Geronimo PIE performed at SCK-CEN Visual inspection Rod length Eddy current testing Oxide thickness measurement Rod profilometry Gross gamma, gamma

spectrometry (Ba\La, Cs-137, Kr-85)

Rod puncture Destructive-OM, EPMA, SEM,

radiochemistry

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Power-to-melt MOX

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Before ramp test After ramp test


Post Irradiation Examination capabilities at SCK-CEN

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TEM disc slicing

Fuel Rod NDT

XRD

XPS

TEM

SEM

EPMA

Non-destructive:• Visual Inspection• Clad Integrity• Oxide thickness• Rod length• Profilometry• X-ray radiography• Gamma scanning• Gamma spectrometry

• 85Kr• 137Cs• 140Ba/140La 106Ru

Destructive:• Rod punction• Mass spectrometry • Fuel density• Hydrogen content (clad)• Optical microscopy• Radiography• µ-hardness• SEM• EPMA• XRD (unirradiated fuel)• TEM (clad)• XPS (clad)

Radiochemistry:• Base actinides (U, Pu)• Minor actinides (Np, Am, Cm)• Fission products

• Cs, I• Sr, Mo, Tc, Ru, Rh, Ag, Sb• Ce, Gd, Pm, Nd, Sm, Eu


ATTICUSAccident Tolerant fuel Test on the Interaction of Coolant with Uranium Silicide

Part of collaboration between INL and SCK CRADA of November 2016

Goal: Quantify interaction between U3Si2 fuel and water under PWR irradiation conditions

PWC irradiation device

Currently in the experiment design phase A. Leaker rodlet experiment (U3Si2 with Zirlo cladding)B. High burnup experiment (U3Si2 Cr-coated cladding)

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General specs

“Leaker” or Water Rodlets Number of Rodlets` 1 or preferably 2Fuel Type U3Si2Fuel Density 95% Theoretical for all compoundsTarget Burnup 1 cycle, 2 cycle (1 ATR cycle = 40-50 days -> 2-4 BR-2 cycles )Enrichment U3Si2 per analysis 5% preferredEnrichment note Preference is for all tests to use 1 or at most 2 enrichmentsLHGR 250 W/cm to 500 W/cm (not a hard requirement)Centerline temperature <1500°COther temperature Requirement

Water / Steam <350°C

Cladding Optimized ZirloOD=0.3744" (9.5 mm), ID=0.3288 (8.35 mm) (desired)Cladding OD/ID

Fuel Stack Height 10.16 cm (this is flexible)Fuel Pellet Height 0.983±0.025 cm Fuel Pellet Diameter Per analysis

±0.0005 cm (±0.0002”)Suggest 0.3245” (8.24 mm)to match ATF-1WB

Fuel Stack Configuration First and last pellet in the stack is depleted (0.22 wt.% U-235), other pellets are enriched (if this is sensible per the analysis)

Temperature Monitor TBDFlux Monitor TBDCapsule Material Per analysis (stainless steel)Capsule ID / OD Per analysis (10 mm / 12 mm)

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BREASYBR-2 Experimental Access and Scientific Yield enhancement

SCK is looking for long-term partners, reserving capacity at BR-2 toassure a sustainable operation Capacity Participation Credit (CPC) system

CPC-holder chooses which party (CPC User) benefits from the credits SCK provides safety and feasibility review

Flexible use of credits: Convert the irradiation credits partially to PIE, manpower, education and

training, experimental waste handling, … Send us your people for follow-up of irradiation and PIE Use our existing irradiation rigs and instrumentation Design and construct your irradiation devices (BYOD) with our help for

safety review and licensing (get extra credits) Preparation of sample transports to perform your own PIE

More information: Sven Van den Berghe ([email protected])

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Conclusion

The SCK•CEN BR-2 reactor has reliably offered fuel irradiation services for over 50 years

BR-2 has now been equipped with a new Be matrix and is ready for another 20 years of operation

SCK is looking for long-term partners, reserving capacity at BR-2 to assure a sustainable operation

The design of the ATTICUS irradiation experiment of ATF fuel in the PWC capsule started

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Copyright © 2017 - SCKCEN

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SCK•CENStudiecentrum voor Kernenergie

Centre d'Etude de l'Energie NucléaireBelgian Nuclear Research Centre

Stichting van Openbaar Nut Fondation d'Utilité Publique Foundation of Public Utility

Registered Office: Avenue Herrmann-Debrouxlaan 40 – BE-1160 BRUSSELSOperational Office: Boeretang 200 – BE-2400 MOL

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Fuel Safety Research Capabilities of the BR-2 Safety... · Reactor core performance of BR2 Design...

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