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)
(A brochure of NSRR facility, JAEA)
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)
(A brochure of NSRR facility, JAEA)
• 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
(A brochure of NSRR facility, JAEA)
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
(A brochure of NSRR facility, JAEA)
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)
Water column velocimeter
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)
(Y. Muramatsu, Y. Udagawa, IAEA‐TECDOC‐CD‐1635, IAEA, Vienna, 2009)
Water column velocimeter 17
• To evaluate generated mechanical energy at fuel failure
Water column
At fuel failure
SteamRing magnet
Float
CoilShaft
Water column velocimeter
• 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
(Y. Muramatsu, Y. Udagawa, IAEA‐TECDOC‐CD‐1635, IAEA, Vienna, 2009)
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
(Y. Muramatsu, Y. Udagawa, IAEA‐TECDOC‐CD‐1635, IAEA, Vienna, 2009)
Rod internal pressure sensor
• 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
Rod internal pressure sensor
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
Reactor Fuel Examination Facility (RFEF) 24
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)
Reactor Fuel Examination Facility (RFEF) 25
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:
(A brochure of RFEF)
Reactor Fuel Examination Facility (RFEF) 26
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
(A brochure of RFEF)
Reactor Fuel Examination Facility (RFEF) 27
• 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…
(A brochure of RFEF)
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
(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)
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
(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)
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
PCMI failure criteria
15 data points from ALPS program
Progress of NSRR experiment program in phase Ⅲ 31
(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)
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
PCMI failure criteria
Burnup range of fuel rods tested in ALPS-II program
Progress of NSRR experiment program in phase Ⅲ 32
(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)
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
Progress of NSRR experiment program in phase Ⅲ 33
* 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
(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)
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]
PCMI failure criteria
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.
(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)
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
(M. Amaya, et al., Proc. Top Fuel 2016, Boise, ID, September 11‐15, 2016)