LWR Stakeholder Survey Results and Gap Analysis Safety Presentations... · 2017. 5. 9. · Test...

45
© 2016 Electric Power Research Institute, Inc. All rights reserved. Ken Yueh Principal Technical Leader GAIN Fuel Research Safety Workshop May 1-4, 2017, Idaho Falls LWR Stakeholder Survey Results and Gap Analysis

Transcript of LWR Stakeholder Survey Results and Gap Analysis Safety Presentations... · 2017. 5. 9. · Test...

Page 1: LWR Stakeholder Survey Results and Gap Analysis Safety Presentations... · 2017. 5. 9. · Test results indicate strong temperature and loading rate dependence 0 2 4 6 8 10 150 250

© 2016 Electric Power Research Institute, Inc. All rights reserved.

Ken YuehPrincipal Technical Leader

GAIN Fuel Research Safety WorkshopMay 1-4, 2017, Idaho Falls

LWR Stakeholder Survey Results and Gap Analysis

Page 2: LWR Stakeholder Survey Results and Gap Analysis Safety Presentations... · 2017. 5. 9. · Test results indicate strong temperature and loading rate dependence 0 2 4 6 8 10 150 250

2© 2016 Electric Power Research Institute, Inc. All rights reserved.

Presentation Content

Recent transient research activitiesSurvey resultsGap evaluation

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3© 2016 Electric Power Research Institute, Inc. All rights reserved.

Background

Regulators became concerned with high burnup effects in the early ‘90s

– Reactivity initiated accident (RIA) test programs CABRI, NSRR, IGR/BIGR

– LOCA research test programs NRC ANL, JAEA, Halden IFA-650

US NRC initiated LOCA rule making and issued draft regulatory guidance on RIA

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4© 2016 Electric Power Research Institute, Inc. All rights reserved.

Reactivity Initiated Accident

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Fuel Response During a RIA

Reactivity Initiated Accident – Sudden control rod ejection causes a brief power surge

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Proposed RIA PCMI Limits

Limits based on NSRR test data (low temperature and short pulse width)

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Modified Burst Test

NRC suggestion a “go/no-go” test– Support new alloy performance qualification

Driver tube simulates pellet expansion

Improvement over standard burst tests– Partial displacement driven deformation– Loading rate control by drop weight velocity/height– Desired deformation by piston travel– Live diameter measurements– Uses only 1” of cladding

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Modified Burst Test Results – PWR SRA Cladding

Test results indicate strong temperature and loading rate dependence

0

2

4

6

8

10

150 250 350 450 550 650 750

Burst S

train (%

)

Hydrogen Concentration (ppm)

25°C, 2.3 ms25°C, >=3.1 ms280°C, 2.3 ms280°C, >=3.1 ms

775630

620660

775600

618653 671

0.0

0.5

1.0

1.5

2.0

0 2 4 6 8 10

Burst S

train (%)

Loading Time To Burst (ms)

NS

RR

Burst strain increases from 1 to 1.5% relative to NSRR condition

280°C, H>600 ppm

Temperature Dependence

Proposed NRC temperature

adjustment basis

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9© 2016 Electric Power Research Institute, Inc. All rights reserved.

MBT Benchmark Results

Failed NSRR tests (and a few non-failed tests)

– Known power history included in the benchmark

MBT test data directly correlate with calculated NSRR failure strain– A few RXA BWR tests depart slightly (both

parent rods operated at very low power in the last cycle)

Some tests survived beyond cladding burst strain limits– Potential explanation with fuel flow at high

temperatures

VA‐4MH‐3

HBO‐6

VA‐1

VA‐2

VA‐3

HBO‐1

HBO‐5 OI‐11

FK‐6FK‐7

FK‐9

FK‐10

FK‐12

LS‐1

0.0

0.5

1.0

1.5

2.0

0.0 0.5 1.0 1.5 2.0

Calculated

 NSRR Pellet E

xpansio

n (%

)

Measured Cladding Ductility (%)

PWR ‐ No Failure ‐ Maximum dHPWRBWRExpected Failure1:1 Correlation

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Technical Limits Suggested by Test Data

Additional in-pile verification would be useful

0

25

50

75

100

125

150

175

200

0 100 200 300 400 500 600 700 800

Enthalpy

 Rise

 (cal/g)

Hydrogen Concentration

EPRI Temperature & Loading Rate Corrected

NRC Proposed LimitEPRI ‐ 2010PNNL ‐ 2013NSRR corrected to 280°C @ 10 msEPRI‐2016REPNa‐3 ‐ did not fail 0

50

100

150

200

0 100 200 300 400 500 600Energy Dep

osition

 (cal/g)

Hydrogen Concentration (ppm)

NSRR Temperature & Loading Rate Corrected

SRP InterimEPRI‐2010NRC‐2015PNNL‐2013NSRR Temp/Strain Rate CorrectionEPRI‐2016

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Proposed Over Pressure Failure Limit

Requires consideration of transient fission gas release ~20% depending on burnup

No transient fission gas release data from prototypical rod pressure

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Loss of Coolant Accident

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Phases of a LOCA

(Assume 72 hr passive cooling)

(MAPP code)

Autocatalytic Oxidation

0

200

400

600

800

1000

1200

0 100 200 300 400 500 600

Peak

Cla

ddin

g Te

mpe

rtur

e (º

C)

Time (s)

Design Basis Accident

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14© 2016 Electric Power Research Institute, Inc. All rights reserved.

ANL LOCA Study

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Proposed Ductility Limit

What about the ballooned and burst region?

Limiting time at temperature

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NRC LOCA and Bend Test – Severe Fuel Fragmentation

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Sample Halden Tests

Test BU

4

800°C PCTPWR

92

5

1050°C PCTPWR

83

6

830°C PCTVVER

56

7

1150°C PCTBWR

44

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Severe Fuel Fragmentation Impact

NRC staff specifically identified a need to “….define the boundary of safe operation for key fuel design and operating parameters, the staff is challenged to evaluate the acceptability of future fuel design advancements and fuel utilization changes.”

Industry is interested in extending the burnup limit to reduce operating cost– 1/3 of the fleet is burnup limited

Characterization of the fuel behavior is needed to enable disposition of the issue

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Parametric Pellet Heating Test

NRC Tested Rod75 GWd/MTU

15 kw/m Last Cycle Power

Alternative Tested Rod67 GWd/MTU

5 kw/m Last Cycle Test sample ~1” long Axial slit to remove radial

restraint – to simulate ballooning condition

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Pre-transient Power Effect

8

9

10

11

12

13

14

15

0 50 100 150 200 250 300 350 400

Ru106Ac

tivity

 Cou

nt

Elevation (cm)

NRC-189

NRC-192EPRI-Inert Gas

EPRI Temperature threshold

Survive Better?

Fuel rod operated with significant power tilt in the last cycle of operation– 106Rh half-life ~1 year– Fuel rod irradiated for 3 cycles in the

original host fuel assembly– Discharged and 3 years later

transplanted to a fresher assembly and irradiated for a 4th cycle

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Fuel Fragmentation Threshold

Research is limited by availability of fuel with required pre-transient power history

0

5

10

15

20

25

30

35

50 60 70 80 90

Pre‐Transie

nt Pow

er (kW/m

)

Burnup (GWd/MTU)

EPRI DataHalden DataNRC DataProbable Transition

0

5

10

15

20

25

30

35

50 60 70 80 90

Pre‐Transie

nt Pow

er (kW/m

)

Burnup (GWd/MTU)

EPRI DataHalden DataNRC DataProbable Transition

0

5

10

15

20

25

30

35

50 60 70 80 90

Pre‐Transie

nt Pow

er (kW/m

)

Burnup (GWd/MTU)

EPRI DataHalden DataNRC DataProbable Transition

No Data Here

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Fragmentation Mechanisms

Grain boundary fission gas generally accepted by scientific community

Internal stress due to operational temperature gradient may also be a factor

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Grain Boundary Fission Gas Measurement

Oxidize fuel at low temperatures in the presence of oxygen– Preferential oxidation of grain boundaries– Pellet interior contains significant grain boundary

fission gas

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Fuel Fragmentation Mechanisms

No difference in the grain boundary fission gas content detected– Canadian Nuclear Laboratory (CNL) personnel indicates grain boundary fission

gas saturation

0

5

10

15

20

25

30

35

50 60 70 80 90

Last Cycle Pow

er (kw/m

)

Burn‐up (GWd/MTU)

Close to 50% of fission gas inventory in the grain boundary

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Intra-Grain Fission Gas Evaluation

Focused ion beam milling to expose pores

Incremental Slice Building Pore Profile

Use Software to Model X-ray Emission

Adjust Gas Density to Match Measured X-ray Emission

0

200

400

600

800

1000

1200

300 800 1300 1800 2300 2800 3300 3800

Intensity

 (cps)

X‐ray energy (eV)

OxygenUraniumXenon

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Intra-Grain Fission Gas

Verified expected lower intra-grain fission gas in the pellet interior Limited number of bubbles analyzed

– Bubble pressure difference at differential radial positions is small

Consistent with higher operational fission gas

release of E08 rod

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Thermal Stress 1/2

Thermal stress arise because during operation the pellet interior is hotter than pellet periphery– In a LOCA fuel pellet would be at nearly uniform temperature– A 3-dimensional finite element analysis was performed

Solid Pellet Model

Pellet Interior in tension

Cracked Pellet ModelLocalized stress

near surface

– Local surface stress concentration was indicated in the cracked quarter pellet model

– Follow-up evaluation needed to determine if the local stress concentration could lead to crack propagation

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Thermal Stress 2/2

Quarter pellet interior do not see local stress concentration Lower internal stress for lower pre-transient power

‐800

‐600

‐400

‐200

0

200

400

600

0.0 1.0 2.0 3.0 4.0 5.0

Stress (M

Pa)

Radial Location (mm)

Radial‐HP Radial‐LPHoop‐HP Hoop‐LPAxial‐HP Axial‐LP

• Quarter pellet too conservative for axial stress calculation since pellet is cracked

Axial stress should be lower can calculated

Fuel reconditioning at lower power needed to change the pellet

condition

LOCA test to test thermal stress hypothesis

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Survey Results

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30© 2016 Electric Power Research Institute, Inc. All rights reserved.

Survey Questions 1/2 What are your existing or emergent fuel transient performance issues that are not being

adequately addressed?

What methodology do you currently use to resolve fuel safety issues associated with– Emergent regulatory questions– Advanced fuel designs – Introduction of new operating regimes (i.e. power uprates, load following)– Optimize current regulatory/operating methods (i.e. apply risk informed methods)– Other, please explain

What additional access to in-pile transient testing capability do you need to support timely resolution of issues?

What capability should test devices have to meet your testing needs? (e.g., what thermal hydraulic environment (BWR/PWR static or flowing loops), pin configuration (single or multiple pins), and transient types (gradual power ramps, RIA/LOCA simulation)?

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Survey Questions 2/2 What complementary separate effects testing capability would be useful to first evaluate

issues at a phenomenon level before committing to integral scale, in-reactor testing?

What is the highest priority candidate separate effects study/studies that could be used to isolate phenomena of interest to improve codes and better design in-core testing?

To enable/support fuel safety research described above, – What source material is required to conduct these experiments (fresh fuel, pre-irradiated fuel)? – What fuel pedigree/pre-characterization do you need to support evaluation? – What post-test examinations (PIE) are required to collect the desired data for NDE? – What PIE is required to collect the desired data for destructive examinations in a hot cell?– What small-sample PIE is required to collect the desired data?

If you need access to historical reports and/or data, please list. Please add any additional information and/or questions.

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32© 2016 Electric Power Research Institute, Inc. All rights reserved.

Existing or Emergent Issues

Transients Conditions RIA/LOCA performance

– U2Si3 fuel clad with SiC or SiC coated zirconium cladding

– Fission gas release– Higher burnup– Doped pellets (Cr2O3 , BeO2, Al2O3/SiO2,

Gd2O3, etc. doped pellets)

Power ramp– No capability gap, ATR/TREAT/Halden

Power cycling - load follow– EDF has extensive experience– Some data exist within the US industry

but will require extensive evaluation

Normal Operation Crud phenomenon understanding and

modeling– No gap, EPRI and CASL programs

Feed-water nickel concentration– No gap, EPRI program

Fuel degradation on failure– Data gap on advanced fuel

Fission gas release– Data gap on advanced fuel

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MethodologyWhat methodology do you currently use for

Regulatory Advanced fuel designs New operating regimes Optimizing reg/operating methods

Other, please explain

• Calculation• Fuel vendor• PWROG• CE reload

methods as modified to use CASMO4/SIMULATE3

Calculationparticipating in EPRI, PWROG and CASL

• Halden test data with calculation

• Licensed method• Vendors/EPRI/CASL• Talk to fuel vendors,

transient thermal analyses

• Calculation• look for ways to gain

process efficiencies while staying within our approved methods.

• Industry feedback with EPRI, reactor vendor

• Using previous NRC guidance for LWRs

• fuel performance analysis using advanced computational techniques.

• Studsvik codes coupled to T/H system codes (RELAP5/RETRAN/VIPRE)• We generally use calculations, based on previous testing when

applicable/needed. When conditions are entirely new (e.g. advanced fuel designs) and testing has not been done in the past, we do testing

S3R core model integrated into a full scope simulator

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In-pile Testing Access and CapabilityAdditional In-pile Access In-pile Test Capability

• LOCA and RIA • Pressurized static capsule• Prototypical commercial reactor condition – flow loop• Water and steam cooling• Multiple pins and whole small bundle

• Normal operation and AOOs • Power ramp and cycle• Fuel degradation from failure under normal conditions• Debris

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Separate Effects Tests and PriorityComplementary Separate Effects Testing Capability

Highest Priority SE Testing

• Electrically heated test train - no gap, ANL, ORNL, Studsvik, etc)

Steam cooling of high performance metallic fuel, 550C peak fuel temperature assuming Zr cladding

• SiC creep and burst test – no gap, SiC creep too slow, MBT at ORNL used to test SIC

U3Si2 swelling, melting point and conductivity as a function of burnup

• U3Si2 interaction with water/steam Modified burst testing on advanced cladding concept

• Fuel annealing studies (unclad fuel) to measure fission gas release – may be difficult

• Fuel swelling measurement (U3Si2) – ATR

• Measure effect of irradiation on U3Si2 melting point and conductivity

• Modified burst testing on advanced cladding concept – Studsvik and ORNL

Page 36: LWR Stakeholder Survey Results and Gap Analysis Safety Presentations... · 2017. 5. 9. · Test results indicate strong temperature and loading rate dependence 0 2 4 6 8 10 150 250

36© 2016 Electric Power Research Institute, Inc. All rights reserved.

Source Material and PedigreeSource Material Required Pedigree Need to Support Evaluation

Fresh and irradiated U3Si2/UO2 fuel in SiC or SiC coated cladding

Fuel grain size distributionCladding mechanical properties, hydrogen concentration, oxide thickness

Uranium nitride fuel Cladding mechanical properties, hydrogen concentration, oxide thicknessPower history, burnup, material composition and manufacturing processComplete characterization

Standard pedigree per INL standard

Late 2nd and 3rd cycle fuel from typical BWR/PWR

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Post Irradiation Examination

Non-destructive Examinations• Chemical and isotopic analysis

• Optical microscopy

• Electron microprobe

• X-ray diffraction

• Gamma scanner

• Alpha scanner

• TEM

• SEM with FIB

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Small Sample Examination

Small Sample PIE Destructive ExaminationsBow and length Optical microscopy

Eddy current Mechanical (tensile, compression, bend, micro-hardness)

Neutron radiography Density

Gamma scanning Composition

Radiation mapping Fission gas measurement

High resolution visual High temperature furnace (accident conditions)

Large plate/element checker Blister annealing testing

Metrology

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Historical Report and Additional InformationHistorical Report Additional Information / Questions• Complete set of exp results AND

testing conditions used by Shimizu for U3Si2 testing (available literature has limited info)

• Also, access to UN irradiation data in both thermal and fast spectrum is valuable.

• For current fuel, research needs to include a REALISTIC examination of the phenomena to ensure that it's of real world concern, rather than a change to a decimal point value in an analysis that is already grossly conservative.

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Operational and PIE Gap Evaluation

No capability gap on post irradiation examination– Additional capability may need be developed as need arise

No capability gap on normal operational issues

Data gap on operational issues– New fuel operational fission gas release– New fuel degradation in contact with coolant– New fuel / cladding power ramp performance– Fuel load follow limits/guidance

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41© 2016 Electric Power Research Institute, Inc. All rights reserved.

RIA Capability Gap

Some tests could be conducted using pressurized static capsule

Pressurized flow loop representative of commercial reactor conditions is needed in most tests– DNB potential at partial power– High temperature failure – Over pressure failure– Fuel-coolant interaction– Transient fission gas release

Prototypical commercial reactor RIA pulse width is between 25 and 65 ms -Cladding ductility is pulse width dependent– TREAT pulse width is > 60 ms

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RIA Data Gap

PCMI performance of doped and new fuel designs

PCMI performance of non zirconium cladding– Mechanical tests could generate most of the test data

PCMI performance of zirconium based cladding under some conditions to verify mechanical characterization data

Fuel-coolant interaction

Transient fission gas release

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43© 2016 Electric Power Research Institute, Inc. All rights reserved.

LOCA Capability Gap

Lack fully realistic test conditions– 50% external electrical heating used in Halden LOCA tests– 100% external heating utilized by others– TREAT could provide 100% nuclear heating

Pressurized flow loop desirable– Remove heat from initial conditioning at pre-transient power to close fuel-

cladding gap– Multiple rods to evaluate flow blockage due to balloon/burst and potential local

temperature excursions

In situ characterization of fuel relocation / dispersal

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LOCA Data Gap

Fuel fragmentation threshold and mechanisms– Fission gas and grain boundary role– Pellet stress contribution from operational temperature gradient

Fuel relocation and dispersal

Fuel cladding balloon and burst behavior

No data on new fuel designs– Pellet behavior– Limited cladding characterization

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Together…Shaping the Future of Electricity