MIT NUCLEAR REACTOR LABORATORY AN MIT INTERDEPARTMENTAL CENTER Irradiation and PIE Capabilities at MIT Research Reactor Lin-wen Hu NSUF User’s Meeting,

MIT NUCLEAR REACTOR LABORATORYAN MIT INTERDEPARTMENTAL CENTER

Irradiation and PIE Capabilities at MIT Research ReactorLin-wen Hu

NSUF User’s Meeting, June 22-25, 2015

Associate Director, Research Development and Utilization

MIT Research Reactor (MITR-II) Multi-purpose research reactor owned

and operated by MIT.

Constructed in 1958 (MITR-I), upgraded in 1975 (MITR-II)

Up-rate in 2010 to 6 MWth (2nd largest

university reactor in U.S.)

Operates 24/7, 10-week cycles

Tank-type, light water cooled and moderated

D2O and graphite reflector

Excellent track record in in-pile irradiation experiments including LWR loop, fuel, high-temperature materials.

MITR joined NSUF as first partner facility in July 2008.

MIT Nuclear Reactor

Laboratory

3

MITR Core HEU plate-type fuel 3 dedicated in-core experiment positions Beam ports for neutron experiments Fission Converter/Medical Beam facility

MIT Nuclear Reactor

Laboratory

4

MIT Reactor Top View

FissionConverter

Filter /Moderator

ConcreteShielding

StudentSpectrometer

PromptGamma Facility

NeutronDiffractometer

FissionConverter

Beam

BNCT MedicalIrradiation Room

Collimator

GraphiteReflector

4DH64DH1

4DH5 4DH2

4DH34DH4

6SH46SH1

6SH26SH3

12SH1

Irradiation Positions

Facility Size Neutron Flux (n/cm2-s)

In-core3 availableMax in-core volume ~ 1.8” ID x 24” long

Thermal: 3.6x1013 Fast: up to 1.2x1014

(E>0.1 MeV)

Beam ports Various radial: 4” to 12” ID

Thermal: 1x1010 - 1x1013 (source)

Vertical irradiation position

2 vertical (3GV) 3” ID x 24” long Thermal: 4x1012 - 1x1013

Through ports One 4” port (4TH)One 6” port (6TH).

Avg thermal: 2.5x1012 to 5.5x1012

Pneumatic Tubes

One 1” ID tube* (1PH1) Thermal: up to 8x1012

One 2” ID tube* (2PH1) Thermal: up to 5x1013

Fission Converter BeamFacility (FCB)

Beam aperture ~ 6” ID Epithermal: ~ 5x109

Thermal Beam Facility (TNB)

Beam aperture ~ 6” ID Thermal: up to 1x1010

Fast flux is ~ 20% of ATR.

6

Three In-core Experiments Installed

Water Loop

In-core Sample Assembly

Special Purpose

Facility

In-Core Sample Assembly (ICSA)

In-core section

Full ICSA

ICSA high temperature capsule demonstration (up to 850°C) with NSUF support.

ACI (Water Loop, NSUF)o SiC LWR cladding in PWR conditions

BSiC (Water Loop, NEET)o SiC channel box and guide tubes

WATF (Water Loop, NEET)o Accident-tolerant cladding and

coatings

HYFI (Fuel, NSUF)o U-Zr-H LWR fuel rods with liquid metal

bonding

AFTR (NEUP)o Internally- and Externally-Cooled

Annular Fuel

HTIF (INIE, NSUF for PIE)o Very high-temperature gas irradiation1000-1400 C.

Drexel (ICSA, NEUP)o MAX-phase materials at 300-700°C in

inert gas

LUNA-1 and LUNA-2 (ICSA, SBIR, STTR)o Fiber optics at 700°C in inert gas

FS-1 (ICSA, IRP)o FHR coupons in flibe salt at 700°C

FS-2 (IRP)o FHR coupons in flibe salt at 700°C

ULTRA (ICSA, NSUF)o Ultrasonic transducer and self-

powered detector test

ICCGM (Water Loop, INL)o Actively-loaded real-time crack growth

monitor

Recent In-Core Experiments

Magnetostrictive and Piezoelectric ultrasonic sensors operating in-coreoParticularly interested in fast neutron damageoReal-time monitoring throughout irradiationoTwo types of magnetostrictive magnetsoThree types of piezoelectric crystals

Self-powered detectors included for local power monitoringoVanadium neutron detector (SPND)oPlatinum gamma detector (SPGD)

Ultrasonic Sensors (INL & PSU)

First lead-out ICSA capsule demonstration!

ULTRA Design

AlN-1

BiTi

AlN-2

ZnO

TC2

TC1Galfenol-1 and Remendur-1

Piezo Drop-Ins

Meltwires

ULTRA Capsule Axial Layout

Magnetostrictive Drop-Insand Flux Wires

ULTRA Loading

Irradiation carried out February 2014 through May 2015, produced real-time transducer and SPD data

Two piezoelectric and both magnetostrictive sensors transmited good signals throughout duration of irradiation.

ULTRA Sensor Data

Unexpected SPD responses to changes in reactor power and capsule temperature

Investigating this with help from INL and CEA

ULTRA SPD Data

Water Loop Design

Exposing specimens to typical power reactor conditions

Using a loop to achieve 300°C, 10 MPa, typical LWR flux, H2 overpressure if desired for <5ppb O2

For SiC, have achieved max exposures >800 EFPD, >3000 MWd

Irradiation Campaigns

3 primary irradiations in water loop starting in 2006 and still in progress.

During intermediate shutdowns specimens can be exchanged or re-inserted after non-destructive PIE.

102 238 240

46 371

90

Initial SiC clad scoping

(Days of exposure)

SiC monolithSiC triplexBonded end caps SiC monolith

SiC triplexBonded shear blocks SiC composites

Creep specimens~1 ppm oxygen

Next gen triplex

Structural

Advanced Cladding IrradiationFacility is extremely flexible in accepting

different types of specimenso Primarily testing partial-length LWR cladding

tubes for corrosion and mechanical properties studies (hoop strength)

o Coupons for corrosion and creep measurement

o Bonded parts for measuring bond performance/shear strength

ACI Results

17

50 100 150 200 250 300 350 400 450 500

-1.2

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

F(u) F(i)

H(u) H(i)

Exposure Time (days)

Wei

ght C

hang

e (%

/mo)

SiC Triplex

Measurement error ±0.5%

Monolithic (α and β) SiC corrosion was negligible Composite is most vulnerable to corrosion, barrier coating provides protection Weight change rates are generally consistent over time Estimated surface rescission based on weight loss as low as 0.5 µm/mo against 100 µm

thick EBC for F and H (triplex tube) specimens

0 100 200 300 400 500 600 700

-0.10

-0.05

0.00

0.05

0.10

0.15

Alpha(u) Alpha(i)

M(u) M(i)

T(u) T(i)

Exposure Time (days)

Wei

ght C

hang

e (%

/mo)

SiC Monoliths

Hydride Fuel Irradiation (HYFI)

University of California, Berkeley project selected by the NSUF for irradiation at the MITR.

Investigating the use of U-Zr-H fuel with zircaloy cladding in light-water reactors.

o 19.75% enriched fuel is bonded inside cladding with lead-bismuth eutectic

o Improved thermal conductivity leads to lower fuel temperature, fission gas release

o Good neutronics properties

HYFI Rodlets

12/1/2014

Fuel pellets produced at UCB (Kurt Terrani, Don Olander) from TRIGA fuel supplied by INL

Pellets loaded into pre-oxidized zircaloy rods at UCB and back-filled with LBE

Fuel centerline and cladding surface thermocouples installed at MIT

Sheath TC Welded to SS Flange

SS304 CF Mini Flange

Zr CF Mini Flange

He PlenumSS302 Spring

Pb-Bi Alloy

Alumina Spacer

Zircaloy-2 Tube

U0.17ZrH1.6 Fuel

Zircaloy-2 End Cap1 cm

HYFI Rodlet Radiography

12/1/2014

Active Fuel RegionAlumina Spacers

Zirconium FlangeSS 304 Flange

302 SS Spring

Rod 1

Rod 2

Rod 3

Rod 4

HYFI Capsules

Zircaloy rodlets were sealed into titanium capsuleso Space between rodlet and titanium filled with LBE

Thermocouples pass through cover gas tubeEach capsule has independent cover gas volume

HYFI Irradiation

Target temperatures ~575°C fuel centerline, 450°C cladding outer28 kW/m linear heat rateLocal power increases over irradiation

HYFI Data

Capsules irradiated between 1400 and 3400 hours

Initial rise in conductivity followed by gradual decline

Capsules 1 (rodlet #1) and 2 (rodlet #3) were removed early due to increased fission gas release from the rodlet into the secondary containment

Cause of release not yet known

0.1

0.12

0.14

0.16

0.18

0.2

0 500 1000 1500 2000 2500 3000 3500

Capsule 1 (rod 1)Capsule 2 (rod 3)Capsule 3 (rod 2)

Time (hr)

HYFI PIE

12/1/2014

Capsules removed from core and cooled in SFP

Gas lines cut and capped in hot cell

Shipped to PNNL in individual casks for PIE (ongoing)

Funded by ATR-NSUF as rapid turnaround experiment in 2014

(PI: R. Ballinger, MIT) Specimens recently extracted in MITR hot cellTRISOs examined as part of FHR project

HTIF PIE

The only university research reactor that operates a LWR loop.

More than 10 years of experience in testing SiC/SiC composites for LWR applications (NSUF, NEET, NEUP, SBIR)

The only university research reactor that can irradiate fuel samples up to 100 gm U-235.

High temperature drop-in capsule experiments demonstrated for up to 850°C. (NSUF, NEUP, SBIR)

Very high temperature irradiation 1000-1400°C (INIE, NSUF).Successful demonstration of lead-out capsule experiment

(NSUF)First demonstration of fluoride salt (flibe) coolant and

materials irradiation at 700°C. (NEUP-IRP)

MITR Accomplishments

MIT-NRL staff contributed to fuel/materials/instrumentation irradiation experiments include:o Dr. Gordon Kohseo Dr. David Carpentero Dr. Michael Ameso Mr. Yakov Ostrovsky o Dr. Kaichao Suno Mr. Tom Borko Dr. Tom Newtono MITR Operations and Radiation Protection staff

MIT-NRL acknowledges funding support for fuel, materials irradiation, and instrumentation irradiation experiments from NEUP, NEET, NSUF programs.

Acknowledgments