experimental capability of nuclear safety research reactor ... safety... · doped fuel pellets...

Experimental capability of Nuclear Safety Research Reactor (NSRR)

Takeshi MIHARA*, Yutaka UDAGAWA, Masaki AMAYA

Fuel Safety Research GroupNuclear Safety Research CenterJapan Atomic Energy Agency

GAIN Fuel Safety Research Workshop, May 1‐4, 2017

Contents 2

1. Reactor facility (NSRR)

2. Test capsules and Instrumentation

3. Hot laboratory (RFEF)

4. Ongoing research programs

1. Reactor facility (NSRR)

3

RIA simulation tests in NSRR 4

NSRR at pulse operation

Pulse operation to simulate RIA

Time

NSRR po

wer

Fuel temp.

Power

10 ms

Fuel te

mp.

Max. inserted reactivity $ 4.7At max. inserted reactivity:

Peak power 23 GWIntegrated power 130 MJPulse width 4.4 ms

(A brochure of NSRR facility, JAEA)

Overview of NSRR 5

Hold-down device

Control roddrive

Verticalloading tube

Offsetloading tube

Water level

Neutrondetector

Reactorcore

Testcapsule

Sub-pile room

Neutronradiographyroom

~9 m

Core horizontal cross section

Reactor coreEffective height: ~38 cmEquivalent diameter: ~60 cmModerator: ZrH, H2O

Driver fuel rodFuel materials: U-ZrH1.6Enrichment: 20%Cladding: SUS 304Dimensions: 3.75cmD x 60cmLNumber of rods: 157NSRR vertical cross section

Driver fuel element

Transient rod

Test capsule

Test fuel rod

Experimental cavity

Safety rod

Regulating rod

Reactor typeTRIGA®-ACPR (annular core pulse reactor)


Operation modes of NSRR 6

Natural Pulse (NP)23,000MW

Inserted reactivity4.7$

Reactorpower

Time (min)

300kW

10MW110MWs

23,000MW

10MW

Steady State (SS)

Combined Pulse (CP)Shaped Pulse (SP)

Time (ms)

Time (s)Time (s)


• Reactor startup• Pre‐conditioning

• Simulate reactivity insertion

• Simulate abnormal power transient

Regulating rods are automatically controlled by computer program Combined mode of SP and NP

• Achieve pulse after high power state

1989

1975

2006

First criticality (June 11) Fresh fuel experiments (Phase Ⅰ Program) started for

investigating fuel behavior under RIA conditions.

Irradiated fuel experiments (Phase Ⅱ Program) startedin order to investigate PCMI failure behavior ofirradiated fuels.

History of NSRR 7

High burnup fuel and MOX experiments (Phase ⅢProgram) were conducted in order to understand theeffects of high burnup and MOX.

High temperature and high pressure experiments werecarried out for BWR-simulating tests and betterunderstanding of PCMI failure.

2. Test capsules and Instrumentation

8

Water column velocimeter

NSRR fresh fuel test 9

Top plug

Bottom plug

Plenumspring

Iron core to detectpellet stack elongation

Fuel pellets

Zry-4Cladding ~1

35 m

m(10% U-235)

(PWR14x14)

NSRR test fuel rod Test capsule

Test fuel rod

Coolant water

Pressuresensor

120 mm

Coolant condition：Room temperature, atmospheric pressure, stagnant water

orHigh temperature,high pressure,flowing water

Transient measurement：‐ Cladding surface temp.‐ Rod internal pressure‐ Capsule pressure‐ Pellet stack elongation‐ Cladding elongation‐Water column velocity

Post irradiation exams：‐ Energy deposition‐ Pellet ceramography‐ Cladding metallography

Phase Ⅰ


Irradiated fuel tests at NSRR 10

Test capsule

Transient measurement- Temp. at clad surface and coolant- Rod pressure, capsule pressure- Clad surface strain (hoop, axial)- Elongations of clad and pellet stack- Water column velocity- etc.

Instrumentation andcapsule assembling

Pellet stack~110 mm

Cutting andrefabricationat RFEF*

Test fuel rod

Time

Rea

ctor

Pow

er Temperature

Power

10 ms

Fuel

Tem

pera

ture

Pulse irradiation

Tota

l len

gth

~4m

Fuel rod

Detailed PIEsat RFEF

* Reactor Fuel Examination Facility in JAEA-Tokai

Power station

Phase Ⅱ


Test capsule for high temperature& high pressure condition

• New capsule for high temperature condition was developed to investigate temperature effect on fuel behavior.

• Pulse irradiation is carried out at the coolant condition of 286C and 7MPa.

• In some tests with the new capsule, availability of cladding surface thermocouple has been confirmed.

11

Coolant condition :

559K(286C)7MPa

Electric heater

Vacuum insulation

Inner capsule

Pressure suppression tank

Outer capsule

Test fuel rod

Phase Ⅲ

(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)

Main instrumentation 12

Capsule pressure sensor

Coolant water thermocouple

Cladding surfacestrain gauge

Rod internal pressure sensor

Capsule pressure sensor

Pellet elongation sensor (LVDT type)

Cladding elongation sensor(LVDT type)


Cladding surface thermocouple

Inner capsule

In case fuel failure is expected

In case fuel failure is NOT expected

Cladding surface thermocouple

• Quick response is needed to follow the fast transient.– Thin ( 0.2mm) wire is employed to minimize heat capacity.– Bare wire is spot‐welded directly on the cladding surface.

• Availability for high temperature range is needed.– R‐type thermocouple (Pt / Pt‐13%Rh) is employed.

• Irradiated fuel rod has oxide layer on the cladding surface.– Removal of oxide layer is needed before spot‐welding to achieve enough electric conductivity.

13

• These requirements must be satisfied with remote control technique.

Remote control technique I (removal of surface oxide)

• In the hot cell …• Oxide layer is removed with grinder.• Removal of oxide layer is confirmed

with…– Visual inspection– Electric conductivity check

14

Grinder

Test fuel rodOxide‐removed zone

Cross section ofirradiated cladding Oxide layer (to be removed)

(Y. Muramatsu, Y. Udagawa, Post‐Irradiation Examination and In‐pile MeasurementTechniques for Water Reactor Fuels, IAEA‐TECDOC‐CD‐1635, IAEA, Vienna, 2009)

Remote control technique II (spot‐welding) 15

TC cartridges (for Pt and Pt‐Rh) Reel

Nozzle

Test fuel rodTerminal

Oxide‐removed zone

TC wire

Powersource

Welding to cladding

Welding to terminal Cut by melting The other wire

Aspect of fuel rod after welding

Horizontal cross section of cladding

Terminal works as cold‐junction

Next, thermocouple wires are spot‐welded to oxide‐removed zone.

Pellet

Cladding

TC

TC is welded successfully.

(Y. Muramatsu, Y. Udagawa, IAEA‐TECDOC‐CD‐1635, IAEA, Vienna, 2009)

Measured cladding surface temperature

• With spot‐welded thermocouple, cladding surface temperature is successfully measured and DNB behaviors have been observed in numbers of NSRR tests.

16

0 2 4 6 8200

400

600

800

1000

1200

Cla

ddin

g su

rfac

e

tem

pera

ture

[K]

Time [sec]

Film boiling

DNB (Departure from Nucleate Boiling)


Water column velocimeter 17

• To evaluate generated mechanical energy at fuel failure

Water column

At fuel failure

SteamRing magnet

Float

CoilShaft


• Water column velocimeter is composed of a fixed shaftand a movable float.

• When fuel fails, steam is instantaneously generated and water column jumps up.

Float jumps up with water column.(Y. Muramatsu, Y. Udagawa, IAEA‐TECDOC‐CD‐1635, IAEA, Vienna, 2009)

Mechanism of water column velocimeter 18

Electromotive force

0

Time

Velocity0

At fuel failure

Ring magnet

Float

CoilShaft

• Shaft contains internal coils (inversed with 3mm interval).• Float contains a permanent magnet ring.• When float jumps up, sign-shaped electromotive force is generated in

the coil by electromagnetic induction.• Water column velocity can be evaluated from the coil interval and the

frequency of sign-wave.

AmplitudeFrequency

3mm

Corresponding to 3mm interval

Velocity


Measured water column velocity 19

0

10

20

0

50

100

150

-50

0

50

0

10

20

2

21 mv

vm

0 5 10 15 20

Water column

Axial center of pellet crack

Failure time

Mechanical energy :

NSRR po

wer [G

W]

Integrated

po

wer [M

J]Water colum

n velocity [m

/s]

Velocimeter

output [m

V] Velocimeter output

Power Integrated power

Time [ms]

• From the velocimeter signal, the velocity of water column is evaluated.• The mass of water column is evaluated with an assumption that

the water above the axial center of pellet stack jumps up.Generated mechanical energy is evaluated.

Velocity



• Strain gauge pressure sensor is comprised of a diaphragm and strain gauges.

• Strain gauges are mounted on the diaphragm.

• When the internal pressure of a fuel rod increases, the diaphragm deforms. Strain gauges measure the deformation and it is converted into a pressure value.

20

Strain gauge

Pressure

Fuel rod

diaphragm

Cable


Measured rod internal pressure

• The rod internal pressure can be measured.• The sensor is affected by pulse irradiation.

21

Pulse

0.2 0.3 0.4 0.50

2

4

Time [sec]

Rod

Pre

ssur

e (M

Pa)

(A report of ALPS program, 2005 )

3. Hot laboratory (RFEF)

22

Reactor Fuel Examination Facility (RFEF) 23

NPPs

RFEF

‐ Pre‐test Examination‐ Test fuel rod fabrication‐ Capsule assembling

NSRR‐RIA test

NSRR

(A brochure of RFEF)

‐ Capsule disassembling‐ Post‐test Examination

Before pulse irradiation

After pulse irradiation


Operation areaHere manipulators and apparatuses are controlled by operators

Next to NSRR building

Cask handling poolTransfer casks from NPPs are sunk and fuels are transferred to concrete cell No.1 (A brochure of RFEF)


Concrete cell No.1

• Visual observation • Assembly washing• Dimensional meas.• γ-scanning

For ‘initial’ non-destructive tests of accepted fuels from pool,

Concrete cell No.2

• X-ray radiography• Eddy current test• Puncture test

For subsequent non-destructive tests and fission gas measurements:



Devices for fuel re-fabrication and destructive PIEs:

• Cutting• De-fueling• Oxide layer removal• Plug welding etc…

Concrete cell No.3

• Tensile test• Burst test• Out Gas analysis(OGA)

• Integral thermal shock test for LOCA conditions

Concrete cell No.4

Concrete cell No.5



• Sample cutting• Surface polishing• Macro observation• Melting point meas.

• Metallography• SEM/EPMA• XRD• Ultra micro hardness

Lead cell No.1

Concrete cell No.6

αγ-concrete cell No.1,2

• Metallography• Densitrometry• High temperature

oxidation testetc…


4. Ongoing research programs

28

NSRR produced many failure limit data for burnup range from about 30-60GWd/t before 2000.

The Japanese fuel failure criteria against PCMI was determined for the fuelburnup range up to 75 GWd/t in 1998.

Failure limit for high burnup fuels

JMTRATR UO2MOXSPERT&PBFPWR

NSRRBWR

CABRI

non−failurefailure

Fuel

ent

halp

y in

crea

se

[J/g]

Burnup [GWd/tU]0 20 40 60 80

0

400

800

0

100

200

PCMI failure criteriaPCMI failure

PCMI failure criteria (Japan)

Subsequent issue has been updating the criteria taking into account burnupextension, MOX effect, improved performance of fuel/cladding materials, etc.

RIA study at JAEA using NSRR facility 29


2009

2014

2002

2010

2011

launched to obtain lacking data of high burnup fuelbehavior under RIA conditions

UO2&MOX fuels, Zry-2/4, M5, MDA, ZIRLO as claddingmaterials, burnup up to 78 GWd/t

Sponsored by NRA, JapanALPS(2002-2009) program

Fuel

ent

halp

y in

crea

se[J/g]

Burnup [GWd/tU]0 20 40 60 80

0

400

800

0

100

200

PCMI failure criteria

Progress of NSRR experiment program in phase Ⅲ 30


2009

2014

2002

2010

2011

Sponsored by NRA, JapanALPS(2002-2009) program

Made important contributions to RIA safety regulation,especially providing …

VA-3: first failure limit data as high-burnup PWR fuel testedunder high-temperature & water-cooled condition

LS-1: failure limit data at highest burnup as BWR fuel

Fuel

ent

halp

y in

crea

se[J/g]

Burnup [GWd/tU]0 20 40 60 80

0

400

800

0

100

200


15 data points from ALPS program



2009

2014

2002

2010

2011

Launched to obtain regulatory data for advanced orfurther high burnup fuels

UO2&MOX fuels Doped fuel pellets Cladding materials: M-MDA, LT-ZIRLO, M5, Zry-2(LK3)

ALPS-II(2010-) program Sponsored by NRA, Japan

Fuel

ent

halp

y in

crea

se[J/g]

Burnup [GWd/tU]0 20 40 60 80

0

400

800

0

100

200


Burnup range of fuel rods tested in ALPS-II program



2009

2014

2002

2010

2011

Great East Japan Earthquake in March, 2011 Reactor building was damaged RIA test schedule was modified

NSRR restart after repair of seismic damage (Dec.2013)

Five ALPS-II tests successfully performed2017 Under review and inspection for checking conformity to New

Regulatory Requirements of Japan


* rod average burnup

BurnupFuel Reactor Ass'y GWd/t Room temp High temp FGD

M-MDA(SR)

81 1 1

M-MDA(RX)

78 1 1

ZIRLO(low-Sn) 80 1

Gravelines-5(France)

M5 84 1 1 2

15x15 Ringhals-2(Sweden)

M5 68 (1)

Zry-2 49(doped)

1

Zry-2 91 (1) (1)

Oskarshamn-3(Sweden)

Zry-2 63(ADOPT) 1

MOX PWR 17x17 Chinon-B3(France)

M5 64 1 1 2

UO2 BHWR DiskHalden

(Norwary) - 130 - - 2

Total 6 (7) 5 (7) 4

Leibstadt(Switzerland)

Fuel type Reactor(Country) Cladding

NSRR test number

UO2

PWR17x17

Vandellos-2(Spain)

BWR 10x10

VA‐5

VA‐6 VA‐8

GR‐1

VA‐7

VA‐9

OS‐1

LS‐4

Performed in JFY2013‐2014 Planned in JFY2017‐2018

FGD‐1

CN‐1

LS‐5

RIA tests in ALPS‐II 34


Experiments performed in JFY2013‐2014

• Five RIA tests VA‐5, ‐6, ‐7, ‐8, and GR‐1 were successfully performed and the PIE of the tests are on‐going in RFEF.

• The failure data of VA‐5 and the non‐failure data of VA‐7, GR‐1.

35

0 20 40 60 800

400

800

0

100

200

Fuel

ent

halp

y in

crea

se [J

/g]


UO2MOXSPERT&PBFJMTRATRPWRBWR

CABRINSRR

failurenon−failure

Burnup [GWd/tU]

Fuel

ent

halp

y in

crea

se

[J/g]

GR-1

VA-7

VA-5

VA-8

VA-6

The current Japanese PCMI failure criteria is applicable for M5 and M‐MDA cladding fuels up to this high burnup of ~80 GWd/tU.


Test fuel

- Rigid chamber with minimum deformation against internal pressure increase

Requirements

Time

pow

er

Concept: Measurement of fission gas release history during RIA test

Pressuresensor

P (t) n (t)Gas

releaserate

Free volume: V

Rigid chamber

Fission gas releaseduring RIA condition

Thermo-couple

T (t)

- Transient measurement of temperature

- Transient measurement of pressure

Fission gas dynamics (FGD) test 36

0.2 0.3 0.4 0.50

10

20

Time [sec]

Rod

Pre

ssur

e (M

Pa)

Results of conventional pressure sensor

• Strain gauge pressure sensor is usually equipped.• The sensor is strongly affected by gamma field in the NSRR core.

39

Pulse

Need to develop a better type of a pressure sensorLVDT type

0.2 0.3 0.4 0.50

2

4

Time [sec]

Rod

Pre

ssur

e (M

Pa)

Could not measure

(Reports of ALPS program, 2005 and 2014)

Thermocouple for gas

Spring Bellows

Fuel chamberPressure

LVDT

Connectors

- neutron and gamma radiation on LVDT electric properties

1. Quick response to rapid pressure rise

2. Accuracy during pulse-irradiation in NSRR: effects of…

PassedQualification tests on…

- thermal expansion of sensor components on the LVDT core position Evaluated

- gamma heating of thermocouple elements on gas temperature measurement

- gamma heating of fuel chamber on gas temperature and pressure Evaluated

Evaluated

as the final qualification test

Development of a pressure sensor using LVDT 38

Pressure measurement chamber

Solenoid valve

Gas reservoir(N2, 4MPa)

LVDT type pressure sensor

Strain gauge typepressure sensor

TC for pressure sensor

50

50

Chamber surface TC #2

Water TC #1

Water TC #2

Chamber surface TC #1

-20 -10 0 10 20 30 400

1

2

3

4

Time (s)

Pres

sure

(MPa

)

LVDT type Strain Gauge type

Recently a qualification test was performed to check the effect of gamma and/or neutron radiation on the response of the LVDT pressure sensor with controlled pressure history.

The test demonstrated that the LVDT pressure sensor has much higher stability against the power pulse

than the SG pressure sensor. quick response at the onset of pressure rise

(key factor for the main objective of the FGD test: to capture gas release kinetics).

Strain-Gauge type

LVDT type

Pulse irradiation

Performance of LVDT‐type pressure sensor under pulse‐irradiation of NSRR

39

The qualification tests finished.Future plan2017: perform a test using a mock‐up of the capsule2018: conduct the first pulse test

(Y. Udagawa, et al., Proc. ICAPP 2016, San Francisco, CA, April 17‐20, 2016)

Summary A lot of NSRR experiments have been conducted to provide database for

regulatory judgment and to promote a better understanding of fuel behavior under RIA conditions, continuously adapted to changing regulatory and scientific needs, with flexibly designed test devices and strong support from the very good‐access hot laboratory RFEF.

The facility and associated experimental resources are currently dedicated to studies on LWR fuel behavior under RIA conditions; it covers• non‐irradiated and irradiated fuels as test rodlets• cooling conditions with stagnant water, temperature from RT to ~280degC, and pressure from 0.1 to ~7 MPa

• online measurements of rod‐surface temperature, water hammer velocity, rod‐internal pressure, coolant pressure, etc.

Recent NSRR experiments performed under ALPS‐II program have added important data points to the PCMI failure‐limit database that reinforces our understanding or prospect on the effects of hydrogen and cladding temperature.

Fission Gas Dynamics (FGD) test with special instrumentation is planned. NSRR restarted in JFY2013 after repair of seismic damage and then is under

review and inspection for checking conformity to new regulatory requirements of Japan from JFY2014.

40

Supplementary

RIA Test Plans for Advanced Fuels in ALPS‐II program 42

Reactor type Fuel type NPP Cladding material Burnup (GWd/t)

PWR17x17, UO2

Vandellos ZIRLO 80M-MDA 73-81

Gravelines M5 84-8715x15, UO2 Ringhals M5 6817x17, MOX Chinon M5 64

BWR10x10, UO2 Leibstadt

Zry-2/LK3 73-91

10x10, Doped-UO2Zry-2 49

Oskarshamn Zry-2 63HBWR Disk, UO2 Halden 130

In order to obtain regulatory data for the advanced fuels under accident conditions, JAEA started the tests in the new ALPS (ALPS-II) program.

UO2 and MOX fuels irradiated in European commercial reactors were gathered in a site in Europe, and test fuel transportation from Europe to Japan was successfully completed in January, 2011.

Due to the Great East Japan Earthquake in March, 2011, RIA test schedule was modified. RIA tests were re-started in February, 2014.

List of advanced fuels for ALPS‐II program


experimental capability of nuclear safety research reactor ... safety... · doped fuel pellets...

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