INL/EXT-20-57844
Revision 4

High-burnup Experiments inReactivity InitiatedAccidents (HERA)May 2022

INL is a U.S. Department of Energy National Laboratory operated by Battelle Energy Alliance, LLC

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INL/EXT-20-57844

High-burnup Experiments in Reactivity InitiatedAccidents (HERA)May 2022

Idaho National Laboratory
Idaho Falls, Idaho 83415

http://www.inl.gov

Prepared for the
U.S. Department of Energy
Under DOE Idaho Operations OfficeContract DE-AC07-05ID14517

Page intentionally left blankABSTRACTHERA is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s (NEA’s) framework for
irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performanceat high burnup under reactivity-initiated accidents (RIA). In-pile RIA experiments have been performed on high burnup
fuels (above 60 GWD/MTU) in the CABRI reactor in France, and the NSRR reactor in Japan. However, most of theseexperiments have taken place with heavily corroded Zircaloy claddings with pulse widths that are narrower (5ms – 30ms
full-width-half-max (FWHM)) than what would be likely in a commercial LWR (30ms – 80ms FWHM). A few tests were
performed in the CABRI facility with FWHM up to 76 ms though only one “blistered” rod experienced cladding failure.
Heavy waterside corrosion and narrow pulse widths are both known to increase the vulnerability of LWR fuel to pellet
cladding mechanical interaction (PCMI). The HERA proposal is designed to (1) quantify the impact of pulse width on fuelperformance, offering new insight into the applicability of existing data, (2) generate new data on high burnup fuel underpulse conditions prototypic of LWRs (3) quantify the additional margin provided by modern cladding alloys to PCMI
failure limits and (4) offer improved data for modelers using specially designed tests that eliminate key uncertainties inhigh-burnup fuel tests.

iii

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CONTENTS
ABSTRACT............................................................................................................................................................................iii
1.

PURPOSE AND OBJECTIVES .................................................................................................................................... 1

2.

PROJECT ORGANIZATION ....................................................................................................................................... 1

3.

BACKGROUND ........................................................................................................................................................... 24.

FACILITY DESCRIPTION .......................................................................................................................................... 4

5.

6.

4.1

THE TREAT REACTOR ................................................................................................................................... 4

4.2

WATER CAPSULE TESTING IN TREAT ....................................................................................................... 6

4.3

THE NSRR REACTOR...................................................................................................................................... 8

PULSE WIDTH (PRE-HYDRIDED) IRRADIATIONS ............................................................................................ 10
5.1

FABRICATION AND CHARACTERIZATION OF PRE-HYDRIDED SPECIMENS ................................. 10

5.2

PRE-HYDRIDED TRANSIENT IRRADIATIONS ........................................................................................ 11

5.3

POST TRANSIENT EXAMINATIONS OF PRE-HYDRDIED IRRADIATIONS ........................................ 12

HIGH-BURNUP IRRADIATIONS ............................................................................................................................. 13
6.1

PRE-TRANSIENT ANALYSIS AND TEST ROD FABRICATION OF HBU RODS .................................. 13

6.2

HIGH BURNUP TRANSIENT IRRADIATIONS ........................................................................................... 147.

HERA MODELLING AND SIMULATION EXERCISE .......................................................................................... 15

8.

PREPARATIONS FOR HERA PHASE II .................................................................................................................. 17

9.

SCHEDULE BUDGET AND DELIVERABLES ....................................................................................................... 18

10.

REFERENCES............................................................................................................................................................. 20v

FIGURES
Figure 1. In-Pile RIA Testing Database as a function of Pulse Width. ................................................................................... 3
Figure 2. Burnup Distribution of Ballooning and Rupture Failures in RIA Experiments ....................................................... 4
Figure 3. Cutaway view of the TREAT reactor ....................................................................................................................... 6
Figure 4. SERTTA Water Capsule .......................................................................................................................................... 7
Figure 5. Configuration of the NSRR ...................................................................................................................................... 8
Figure 6. Test fuel rod and test capsules .................................................................................................................................. 9
Figure 7. Cladding hoop stress during PCMI phase of RIA transients of varying pulse widths. .......................................... 16

TABLES
Table 2. Pre-hydrided test rod characteristics........................................................................................................................ 10
Table 3. Pre-hydrided test matrix. ......................................................................................................................................... 11
Table 4. High-burnup test matrix........................................................................................................................................... 14
Table 5. Past RIA experiments to be simulated using advanced tools as part of HERA. ..................................................... 17
Table 6. HERA Schedule ....................................................................................................................................................... 19
Table 7. HERA Budget .......................................................................................................................................................... 19vi

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High-burnup Experiments in Reactivity InitiatedAccidents (HERA)
1.

PURPOSE AND OBJECTIVESHERA is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s(NEA’s) framework for irradiation experiments (FIDES). The HERA JEEP is designed to understandlight water reactor (LWR) fuel performance under reactivity-initiated accidents (RIA). Specifically, the
JEEP is intended to:
1. quantify the impact of pulse width on fuel performance, offering new insight into the applicability ofexisting data;
2. generate new data on high burnup fuel under pulse conditions prototypic of LWRs;
3. quantify the additional margin provided by modern cladding alloys to PCMI failure limits; and4. offer improved data for modelers using specially designed tests that eliminate key uncertainties in highburnup fuel tests.
HERA will achieve these objectives through two experimental tasks as well as a modelling andsimulation exercise. The first task is associated with the identification of pulse width impacts on hydride
assisted Pellet Cladding Mechanical Interaction (PCMI) failure in RIA transients and will be achieved
through systematic experiments on pre-hydrided cladding. The second task is to extend the existing
empirical database of high-burnup RIA test data particularly to assess post failure phenomena. Thissecond task will involve the transient irradiation of material, which has been previously irradiated in acommercial LWR. Most experiments in HERA will take place at Idaho National Laboratory’s (INL’s)
Transient Reactor Test Facility (TREAT) with some experiments taking place at the JAEA’s Nuclear
Safety Research Reactor (NSRR).
The modeling and simulation exercise will be directed at evaluating experiment design performanceand evaluation of key experiment objectives, focused on both blind predictions and post-test follow-up
evaluations. While the HERA project will provide management and logistical support for the modellingand simulation exercise, the actual modelling work will be sought through participation of the broader
FIDES community.

2.

PROJECT ORGANIZATIONThe HERA JEEP organization consists of a core group, operating agent(s), and a project coordinator.
The core group will consist initially of the following five organizations. Consistent with the definition of
the JEEP Core Group in the FIDES Agreement, each of these organizations will be providing resources
needed for the successful execution of the HERA JEEP either through funding or in-kind contributions.
1. the Department of Energy of the United States of America (DOE), also signing on behalf of Idaho
National Laboratory (INL);
2. the Nuclear Regulatory Commission of the United States of America (NRC);
3. Westinghouse Electric Company (WESTINGHOUSE);
4. the Japan Atomic Energy Agency (JAEA);
5. the French Institut de radioprotection et de sûreté nucléaire (IRSN).

1

Roles and responsibilities for the various core group members are discussed in this proposal. The
Idaho National Laboratory (INL) will act as the principal Operating Agent as the majority of theexperimental work will occur there. Two of the transient irradiations with fresh fuel and pre-hydrided
cladding will take place at the NSRR reactor. For those experiments, JAEA will be the Operating Agent.
The HERA JEEP core group will nominate one individual to serve as the HERA JEEP coordinator.
The HERA JEEP coordinator will have the principal responsibility of managing and overseeing the
project as well as serving as the liaison between the HERA core group, and the FIDES Governing Board,
the NEA Secretariat, and the JEEP Advisory Group (JAG) if the FIDES Governing Board decides toestablish such a group.

3.

BACKGROUND

During RIAs LWR fuel cladding can be breached by one of two mechanisms; pellet-cladding
mechanical interaction or high temperature balloon and rupture. Pellet-cladding mechanical interaction(PCMI) failure occurs in the early phase of an RIA transient where the fuel pellet expands rapidly into thecladding prior to meaningful heat transfer from the pellet to the cladding. This results in high stresses in
the cladding resulting in crack propagation through the base Zircaloy metal and axially along the cladding
tube. Failure by PCMI is a brittle failure mode and occurs at low cladding strains, approximately 2% hoopstrain.
If the cladding survives the PCMI phase of the transient, it may still fail due to ballooning and rupture
induced by extended exposure to high temperature. Thermal transport from the fuel cladding to thecoolant is significantly reduced if boiling transition occurs on the cladding surface. This results in both arapid increase in fuel temperature, which drives fission gas release, and a spike in cladding temperature,
which results in loss of mechanical strength. If the internal pressure of the fuel rod is greater than thesystem pressure, the cladding can inelastically deform and eventually rupture. Failure by swelling andrupture occurs at larger strains, generally greater than 5% hoop strain. While PCMI failures are generally
seen as more limiting, the dominance of one failure mode over the other can be affected by both the
cladding material conditions, and the transient evolution.1
Due to the complex multi-physics interactions, fuel performance in RIA is best studied in test reactors
especially suited for replicating the rapid nuclear heating conditions in the fuel. In-pile RIA experiments
have been performed on high burnup fuels (above 60 GWD/MTU) in the CABRI reactor in France, andthe NSRR reactor in Japan2 .However, most of these experiments have taken place with either heavily
corroded claddings or with pulse widths that are narrower (5ms – 30ms full-width-half-max (FWHM))
than what would be likely in a commercial LWR (30ms – 80ms FWHM). A few tests were performed in
the CABRI facility with FWHM up to 76 ms though only one “blistered” rod experienced cladding
failure. Heavy waterside corrosion and narrow pulse widths are both known to increase the vulnerabilityof LWR fuel to pellet cladding mechanical interaction (PCMI).3
In traditional Zircaloy claddings, thick oxide layers and zirconium hydrides form on the outer edge ofthe cladding where flaws and cracks can be easily nucleated. The severity of these oxide layers andcladding hydrides directly effect the fuel’s vulnerability to PCMI failure. The relationship between PCMI
failure threshold and the extent of waterside corrosion and hydrogen pickup is clearly seen in both out of
pile separate effects tests5,6 and in the in-pile experimental database.2 However, recent alloy developments
by nuclear fuel vendors have resulted in zirconium alloys that are highly resistant to both waterside
corrosion and hydrogen pickup7. Efforts are also currently underway to coat the cladding with a thin
chrome layer which could all but eliminate corrosion concerns during normal operations.8,9 PCMI failure
thresholds derived from experiments on heavily corroded legacy claddings may be unnecessarily
conservative when applied to these more modern alloys, which may allow PCMI failure threshold to be
replaced by less limiting high temperature phase failure thresholds.

2

In a recent update to its “State of the Art Report on Fuel Performance in Reactivity InitiatedAccidents” the NEA’s Working Group on Fuel Safety (WGFS) identified representativity of experiments
to prototypical RIA as a vital consideration.10 The transient pulse width is an important factor to consider
when comparing experiments to prototypical RIA conditions. It is theorized that longer pulses should
decrease the likelihood of PCMI failure due to two factors. First the smaller strain rates should result in
lower cladding stress by giving the material time to deform plastically. Additionally, longer transientsshould allow more time for heat transfer to occur from the pellet to the cladding, increasing the cladding’s
temperature and further promoting ductile behavior and decreasing the overall stress. The effect of pulse
width can be predicted using modern fuel modeling and simulation codes11 and has been confirmed in outof pile separate effects tests.12 However, the pulse width effect cannot be deduced from the existing inpile experimental database mainly due to the lack of experiments conducted with pulse FWHM between~60ms and ~500ms as seen in Figure 1.13 Modern 3-dimensional core physics calculations show
increasingly wider pulse widths highlighting the need for experimental data in this area.

Figure 1. In-Pile RIA Testing Database as a function of Pulse Width.
Because modern high burnup fuels are expected to be more resilient to PCMI failure, the dominantfailure mode may transition to ballooning and rupture failure. High burnup fuels contain a proportionallygreater amount of fission gas stored in the fuel matrix. This fission gas can be suddenly released in the
RIA transient due to high temperature exposure and high thermal stresses resulting in high plenum
pressures that promote ballooning and rupture failure. Very little in-pile data exists to support thedevelopment of a swelling and rupture failure threshold for high burnup fuels. Figure 2 below shows theburnup distribution for RIA tests conducted in water environments which either failed from ballooning
and bursting or did not fail.

3

Figure 2. Burnup Distribution of Ballooning and Rupture Failures in RIA Experiments
Additionally, ballooning and rupture failure occurs with high hoop strains creating balloons similar to
those often observed in loss of coolant accident (LOCA) transient simulations. In addition, it has been
observed in prior LOCA simulations on high burnup fuels that the loss of cladding restraint promotes
gross fuel fragmentation, relocation, and dispersal (FFRD).13 In high energy RIA transients this could also
lead to fuel melting in the ballooned region and upon quench, higher levels of fragmented and molten fuelcoolant interaction than what has been observed in experimental programs to date. These interactions
could pose a challenge to maintaining coolable geometry. Due to limitations of the experimental facilities
previously available for testing, the current database of in-pile RIA experiments on high burnup fuelscontains very few experiments with peak radial average enthalpies above ~100 calories per gram.
Therefore, investigation of peak radial-average fuel enthalpy up to the current fuel failure limit of 150
cal/g-UO2 and to the core coolability limit for fresh and low burnup fuel of 230 cal/g-UO2 is needed toestablish a high temperature failure limit for RIA events.

4.

FACILITY DESCRIPTION

The FIDES HERA program would ideally extend to multiple existing transient reactor facilities thatwould eventually form the foundation of FIDES’ RIA ‘core group’ including the CABRI facility inFrance, the NSRR in Japan, and the TREAT facility in the USA. In the initial three-year program of
HERA, in-pile RIA irradiations will take place primarily at the TREAT reactor at Idaho National
Laboratory (INL) with two tests at the NSRR facility in Japan. Description of the NSRR reactor, and thestatic water tests that have occurred there are available in numerous technical articles and is summarized
below in section 4.3.14,15,16,17.

4.1 THE TREAT REACTOR
TREAT is an air-cooled reactor driven by a core of graphite blocks having a small concentration ofdispersed uranium dioxide shown graphically in Figure 7. Experiment assemblies are typically removed4

from or placed into the core through a slot in the reactor’s upper rotating shield plug. Pulse type transientsdesigned to simulate an RIA are initiated in TREAT by bringing the reactor to a low steady state power of
50 watts and then rapidly removing transient control rods, resulting in a step insertion of excess reactivity.
TREAT pulses initially have a nominally Gaussian shape followed by a decaying exponential tail. Larger
step reactivity insertions result in transient pulses that have higher peak powers, higher overall energy
releases, and shorter pulse widths. TREAT has the ability to re-insert the transient control rods andshorten the natural transient. These maneuvers are referred to as “transient clipping”. The clipping system
has the ability to reduce both the pulse width (FWHM), total energy released, and almost completely
eliminating the energy in the tail. As an example, for a 4.5% k/k reactivity insertion, clipping
capabilities could decrease the maximum energy released from the reactor from ~2800 MJ to ~630 MJ
and shorten the pulse width from 103ms to 95ms. This approach gives the ability to tailor transients to adesired duration and energy release. Tailoring is currently limited by the speed of the rod drive system(~355 cm/s). For TREAT’s ~1m reactor length, it takes ~280ms for the control rods to fully insert
themselves. However, pulses with FWHM of 89 ms have been demonstrated, and planned plantmodifications involving a He-3 clipping system are predicted to be able to achieve pulse FWHM as low
as 50 ms,18 within reach of representative PWR HZP RIA pulse widths and overlapping nicely to theupper end of pulse widths that CABRI is capable of producing, which more naturally overlaps with
NSRR capability.

5Figure 3. Cutaway view of the TREAT reactor

4.2 WATER CAPSULE TESTING IN TREAT
Currently, experiments in TREAT are performed in a static water capsule called SERTTA (static
environment rodlet transient testing apparatus) shown below in Figure 4. The capsule can accommodate atest rod that is of prototypic PWR radial dimensions (9.5mm outer diameter) and is 12cm long with a 8cm
fueled length. The capsule is equipped with a gas volume above the test specimen to accommodate
pressurization by rapid vaporization of the water during the transient. A high temperature crucible is thenplaced at the bottom of the capsule to safely accommodate partial or complete melting of the test
specimen. SERTTA fits inside a containment pipe which is lowered into the reactor as an integral unit
through the rotating shield plug. Twelve 1 mm instrument leads penetrate the capsule and containmentand can be used to accommodate a variety of different instruments. One possible configuration with four
cladding thermocouples, capacitive boiling detectors, one fuel thermocouple and two capsule waterpressure sensors is shown below in Figure 4.

6

Figure 4. HERA Water Capsule74.3 THE NSRR REACTORThe NSRR (Nuclear Safety Research Reactor ) as shown in Figure 5 is a modified TRIGA-ACPR(Annular Core Pulse Reactor) of which salient features are the large pulsing power capability whichallows the 10% enriched fuel to be heated by nuclear fission to temperatures above the melting point of
UO2; and a large center cavity of 220 mm in diameter with loading tubes, which enables the fuelirradiation experiments with high neutron flux and easy loading/unloading of the test capsule containing atest fuel rod. The core structure is mounted at the bottom of a 9-m deep open-top water pool and cooledby natural circulation of the pool water. The NSRR core consists of 149 driver uranium-zirconium
hydride (U-ZrH) fuel/moderator elements, six fuel follower regulating rods and two fuel follower safetyrods. The pulsing operation is made by quick withdrawal of enriched boron carbide transient rods bypressurized air. The pulsing power escalation is controlled by spectrum hardening caused by themoderator temperature increase, and the Doppler effect in the NSRR. During the maximum reactivity
insertion of 3.4%Δk/k ($4.67), the pulse power reaches 21.1 GW with a corresponding core energy
release (integrated reactor power) of 117 MJ with a minimum reactor period of 1.17 ms. The shape ofreactor power history depends on the inserted reactivity, and the smaller pulse becomes broader. While
the FWHM for a $4.6 pulse is 4.4 ms, and for a $3.67 pulse, it is 5.6 ms. The energy deposition in a test
fuel rod is controlled by the amount of reactor power and by the fissile content of the fuel.
Example of a test fuel rod and a test capsule with instrumentations for un-irradiated test fuel isillustrated in Figure 6. The test fuel rod is set at the center of the stainless-steel capsule with coolant
water. The test capsule is a sealed pressure vessel in which the diameter and the height of the inner roomare 120 and 800mm.

Figure 5. Configuration of the NSRR

8

Figure 6. Test fuel rod and test capsules

95.

PULSE WIDTH (PRE-HYDRIDED) IRRADIATIONSThe HERA pulse width studies are aimed at determining if there is a pulse width dependence or pulse
width threshold for PCMI failures in RIA conditions. These studies will involve six transient irradiation
tests at three different pulse widths (two tests at each pulse width to ensure repeatability). Four of thetransient irradiations will take place at the TREAT facility in the United States operated by INL, and two
will take place at the NSRR facility in Japan operated by the JAEA.

5.1 FABRICATION AND CHARACTERIZATION OF PRE-HYDRIDED
SPECIMENS
These transients will take place with unirradiated fuel rods that simulate specific aspects of highburnup, which are known to principally contribute to PCMI failure. These aspects are the presence of azirconium hydride rim on the outer diameter of the cladding, and the closure of the pellet cladding gap.
Characteristics of the pre-hydrided test rods are in Table 1.
Table 1. Pre-hydrided test rod characteristics.
Test Rod Characteristic
Cladding Material
Cladding Outer Diameter
Cladding Wall Thickness
Cladding Hydrogen Content
Cladding Hydride Rim Thickness
Fuel Material
Fuel Pellet Density
Fuel Pellet U-235 Enrichment

Value
Zircaloy-4
9.5 mm0.57 mm~500 ppm
50 - 90 m
UO2>95% TD
0.8% (TREAT Rods)~5% (NSRR Rods)
8.32 mm
20 m
8-15
8–18 cm
12–25cm
He~2 MPaFuel Pellet Diameter
Pellet/Cladding Radial Gap
Number of Fuel Pellets
Fuel Column Height
Overall Test Rod Height
Fuel Rod Backfill Gas
Fuel Rod Internal Pressure (Room Temperature)

The Zircaloy-4 and UO2 will be supplied by INL. INL will perform the pre-hydriding procedure prior
to fabrication of the test rods. The hydride rim structure will be similar to that described in Tomiyasu etal.16 as similar materials have been tested previously in transient irradiations in NSRR. INL will be
responsible for fabricating the test rods, which will be irradiated in TREAT and NSRR.
Some material and mechanical characterization of the pre-hydrided cladding will take place to verify
the characteristics of the material used in the transient irradiations. These will include analyticalmeasurements of the cladding hydrogen content, as well as optical micrographs of sister samples showing
the hydrogen distribution (rim thickness). Some mechanical testing of the pre-hydrided cladding may also
take place. Data generated in the pre-hydrided fabrication and characterization task that will be owned by
the core group and shared with the FIDES community subject to the confidentiality requirements in the
FIDES Agreement shall include but may not be limited to the following:
10

•
•
•
•
•

As built dimensions of the test rods including location of attached instrumentation;
Density and Enrichment of the UO2 fuel pellets;
Analytical measurements of cladding hydrogen content from similarly hydrided cladding tubes(distribution of this data may not be limited to the FIDES community);
Metallography of cladding hydrogen content/morphology from similarly hydrided cladding tubes(distribution of this data may not be limited to the FIDES community);
Results of any mechanical testing from similarly hydrided cladding tubes (distribution of this data
may not be limited to the FIDES community).

5.2 PRE-HYDRIDED TRANSIENT IRRADIATIONSFor Stress Relieved Annealed (SRA) cladding with ~500 ppm hydrogen, PCMI failure is expected tooccur at around 75 calories per gram (313.8 J/g) peak radial average enthalpy19. If the cladding does not
fail due to PCMI, then ballooning and bursting failure would be expected at around 150 calories per gram(629 J/g) peak radial average enthalpy19. As the purpose of the study is to determine if cladding failuremode transitions from a PCMI failure to a high-temperature failure at a given pulse width, all of thetransient irradiations will take place at the same target peak radial average enthalpy of ~650 J/g, which isslightly above this failure threshold. The two tests at NSRR will take place with a pulse width between
5 and 10 ms. The first two tests at TREAT will take place at TREAT’s current narrow pulse width
capability of ~90 ms. The second two TREAT experiments will take place following the installation of a
He-3 clipping system at TREAT, which will lower the pulse width to ~50 ms. The test matrix issummarized in Table 2.
Table 2. Pre-hydrided test matrix.
Test Number
Test Reactor

Pulse Width (ms)Target Peak Radial Average
Enthalpy (J/g)
HERA-PreH-1
NSRR
5-10
650
HERA-PreH-2
NSRR
5-10
650
HERA-PreH-3
TREAT
90
650
HERA-PreH-4
TREAT
90
650
HERA-PreH-5*
TREAT + He-3 Clip
~50
650
HERA-PreH-6*
TREAT + He-3 Clip
~50
650
* If transients 3&4 fail due to PCMI then the Pulse Width of transients 5&6 will be extended rather thancontracted.
Data generated in the pre-hydrided transient irradiation task that will be owned by the core group andshared with the FIDES community subject to the confidentiality requirements in the FIDES Agreement
shall include but may not be limited to the following:••••

Reactor power during the transient;
Energy calibration data necessary to convert reactor energy to specimen energy;
Raw data outputs from all rodlet and capsule in-situ diagnostics;
Results and methods used of any assessments or analysis interpreting the results of the raw data;
signals into meaningful fuel performance parameters (e.g. interpreting cladding temperature fromthermocouple data, determining moment of cladding rupture, etc.).

11

5.3 POST TRANSIENT EXAMINATIONS OF PRE-HYDRDIED
IRRADIATIONS
Following the transient irradiations, the test rods will be moved to examination facilities. Test rodsthat were irradiated at TREAT will be examined at INL facilities and test rods that were irradiated at
NSRR will be examined at JAEA facilities. The rods will undergo some analytical measurements toverify the amount of energy that was deposited into the fuel during the test. It is assumed that all rods inthe irradiations will fail (either by PCMI or balloon and burst). The failure areas of each of the rods will
be investigated nondestructively and the size of the rupture openings will be measured and documented.
Following the visual and dimensional measurements, the rods will be segmented. In rods that failed due
to PCMI, the fracture surfaces will be investigated. Evidence of the characteristic brittle fracture in
hydride rim and ductile shear in the substrate will be sought with both optical and scanning electron
microscopy of the cladding edges and fracture surfaces, respectively. In rods that failed due to ballooning
and burst, microstructural investigations will focus on the thickness of the oxide, alpha prime, and priorbeta layers in the cladding especially in the vicinity of the burst. Evidence of secondary hydrogen uptake
above and below the ballooned region will be determined by taking analytical measurements of claddinghydrogen at several axial locations near the balloon.
Data generated in the post transient examinations of pre-hydrided irradiations task that will be ownedby the core group and shared with the FIDES community subject to the confidentiality requirements in the
FIDES Agreement shall include but may not be limited to the following:
•
•
•
•
•

Analytical determination of total energy deposition;
High resolution photographs of the test rods following removal from test capsules;
Post transient dimensional measurements of cladding and any rupture openings;
Metallography images of cladding fracture surfaces;
Analytical measurements of cladding hydrogen content.

12

6.

IRRADIATIONS WITH PREVIOUSLY IRRADIATED FUEL

Following the irradiation of the pre-hydrided cladding, the irradiation of two high-burnup samples
and two moderate burnup Chromium Coated samples will take place. All four of these transient
irradiations will take place at the TREAT reactor at INL. The previously irradiated test rods will comefrom host material that was irradiated in the Byron Nuclear Generating Station (BNGS) near Byron,
Illinois, in the United States. It is anticipated that the rod segments will be in INL hotcells by Q1 of 2023.
The rods come from Westinghouse Optimized Fuel Assemblies and have an outer diameter of 9.1 mm
with a wall thickness of 0.57 mm. The rods were all initially fueled with UO2 pellets (one of the
Chromium Coated rods contains ADOPT™ UO2) enriched to 4.5% in U-235.
The two high burnup (HBU) transient irradiations will make use of material coming from parent rods,
which are clad in an advanced zirconium alloy cladding and have been irradiated to a rod average burnup
of ~73500 MWd/MTU (~81000 MWd/MTU segment burnup). One of the ATF rods comes from a Cr
coated Optimized ZIRLO™ rod with UO2 fuel and the other ATF rod comes from a Chromium coated
Optimized ZIRLO rod with ADOPT UO2 fuel. Both ATF rods are at ~33000 MWd/MTU segment
burnup)Background data on the high burnup rods will be supplied by WESTINGHOUSE working with
BNGS. This data will be shared with the core group and the FIDES community, but ownership of the data
will remain with Westinghouse and BNGS respectively. This data will include but may not be limited to
the following:••

Initial fuel dimensions and enrichments of the parent rods;
Rod average burnups and Effective Full Power Days.

6.1 PRE-TRANSIENT ANALYSIS AND TEST ROD FABRICATION OF
PREVIOUSLY IRRADIATED RODS
Prior to transient irradiations, the parent rods will be punctured to determine fission gas pressure andmakeup prior to being sectioned for pre-transient characterization and test rod fabrication. Test segments
ranging from 15 to 30 cm (10 to 24 cm fueled length) will be rebuilt into the test rods used in thetransients. The test rods will be backfilled to with plenum pressures approximately equal to that
determined in the fission gas puncture with a mixture of Ar and He that provides approximately equal
thermal conductivity to the original He/fission gas mixture. Test rods will be fitted with diagnostics andloaded into the static water capsules used in testing at TREAT. Water temperature and pressure sensors
will be included in the capsule diagnostics.
Adjacent cladding segments from the parent rod will be sent for analytical and microstructural
examinations. Key outcomes of the pre-transient characterization include analytical determination of thefuel burnup and the cladding hydrogen content. Microstructural examinations will determine the oxide
layer thickness, fuel/cladding interaction layer thickness, and the distribution and orientation of hydrides
present in the cladding. Some mechanical testing of adjacent cladding segments may also be undertaken.
Data generated in the pre-transient analysis and test rod fabrication of the previously irradiated rods
task that will be owned by the core group and shared with the FIDES community subject to theconfidentiality requirements in the FIDES Agreement shall include but may not be limited to thefollowing:••••

Gamma Scans of parent rods.
Plenum pressure and fission gas composition of parent rod;
High resolution photographs of the test segment at various phases of the refabrication process;Analytical measurements of fuel burnup from adjacent fuel segments;

13

•
•
•

Analytical measurement of cladding hydrogen content from adjacent cladding segments;
Metallography images of the fuel and cladding in adjacent fuel/cladding segments;
Profilometry data of the refabricated test rod

6.2 HIGH BURNUP TRANSIENT IRRADIATIONSConsistent with the data needs identified in NRC memo dated 31 March 2020 from Paul Clifford to
Joseph Donoghue20 testing objectives will focus on testing conditions at and beyond the current PCMI
failure limits. The initial peak enthalpy targets for the previously irradiated rods will be selected from thecurrent regulatory failure19 given their analytically determined cladding excess hydrogen concentration
determined in pre-transient characterization.
Following the initial HBU irradiation, the second HBU irradiation will depend on the outcomes of thefirst test. If the rod did not experience cladding failure, then the target peak radial average enthalpies will
be increased by 10–15% during the second irradiation. If the rods do experience a failure, then thetechnical advisory group will be consulted and determine if HERA objectives are best served byincreasing the target enthalpy further to evaluate post failure behavior or by decreasing the target peakradial average enthalpy to verify the validity of existing failure limits. For example, if rod failure occurs
early in the transient, at a lower-than-expected radial average enthalpy, then the second test should
employ a lower enthalpy target to verify that the test rod can survive in a less severe transient. However,
if rod failure occurs as expected, late in the transient at or near the predicted failure enthalpy, then thesubsequent test may target an even higher enthalpy to evaluate the impact of the higher enthalpy onpressure pulses in the water capsule and the extent of material loss from the cladding failure opening. The
peak enthalpy target for both of the Chromium Coated rods will be the same to determine if there is achange in behavior as a result of the use of the ADOPT fuel pellets. A summary of the high-burnup tests
is presented below in Table 3.
Table 3. High-burnup test matrix.
Target
Pulse
Width
(ms)
50
50

Test Number
HERA-HBU-1
HERA-HBU-2

Cladding
Advanced Zirconium Alloy
Advanced Zirconium Alloy

Segment
Burnup(GWd/MTU)~81~81

HERA-ATF-UO2

Cr coated Optimized ZIRLO

~33

50

HERA-ATF-ADOPT

Cr coated Optimized ZIRLO

~33

50

TargetEnergyDeposition(J/g)~630 J/g*
TBD~630 J/g *J/g~630 J/g *J/g*Determined based on cladding excess hydrogen

Data generated in the high burnup transient irradiation task that will be owned by the core group andshared with the FIDES community subject to the confidentiality requirements in the FIDES Agreement
shall include but may not be limited to the following:•••

Reactor power during the transient;
Energy calibration data necessary to convert reactor energy to specimen energy;
Raw data outputs from all rodlet and capsule in-situ diagnostics;

14

•

Results and methods used of any assessments or analysis interpreting the results of the raw data
signals into meaningful fuel performance parameters (e.g. interpreting cladding temperature fromthermocouple data, determining moment of cladding rupture, etc.).

Post transient examinations are beyond the schedule and budget allotted for the first 3-year phase of
HERA and will be undertaken in the follow-on Phase II of HERA in 2024–2026.

7.

HERA MODELLING AND SIMULATION EXERCISE

Analytical support for HERA will use advanced modeling and simulation (M&S) tools to perform
blind predictions of experiment performance and post-test evaluations with the support of preliminary
data results from experiments. This exercise will be performed with reference to three extensive RIA
benchmark projects performed under the NEA Working Group for Fuel Safety22,23,24. Consistent with theobjectives of Phase I HERA project, the HERA M&S exercise will evaluate the impacts of peak pellet
enthalpies, pulse widths, and varying levels of environmental degradation. Participation in the simulation
exercise will be from the HERA core group and also sought from the broader FIDES community.
The focus of the simulations will be on the PCMI phase of the transient. It is theorized that hydrided
zirconium cladding fails in RIA PCMI conditions under rapid loading, when cracks initiating in the brittle
hydride rim propagate through the metallic substrate. Failure can be predicted if the stress intensity factor(K1) (as a result of crack initiation in hydrides near the cladding outer diameter) is greater than somecritical value (KIC). The stress intensity factor is expressed in the following form:
𝐾1 ∝ 𝜎11 √𝜋𝑎(1)where (a) is the crack length which can be assumed to be the depth of the hydride rim (in SRA cladding
types) and 11 is the normal (hoop) tensile stress field.
Thus, the vulnerability of a fuel rod to PCMI failure will depend upon both on the extent of theenvironmental degradation (depth of hydride rim) as well as the extent of the PCMI interaction. In
transients where the cladding has time to increase in temperature prior to experiencing a tensile load it ismore likely to survive the transient. The increase in temperature has two consequences. First the yieldstress of zircaloy cladding decreases dramatically with increasing temperature. Therefore, at a given
loading, the tensile stress field that develops in the cladding is much lower when the cladding temperature
is higher, thus the stress intensity factor is also much lower for a given crack depth. The effect of longer
pulses leading to higher cladding temperatures and thus lower cladding stresses is clearly seen in fuelperformance models of RIA transients of different pulse widths. A simple BISON calculation on fresh
fuel illustrates this phenomenon in Figure 7 below.

15

900

300

800

200

700

100

600

0

500
136 ms FWHM

-100

400

60 ms FWHM
10 ms FWHM

-200

Clad Surface Temperature (°C)

Hoop Stress (MPa)-solid line

400

300
0

0.1

0.2

0.3

0.4

0.5

0.6

Plastic Hoop Strain (%)

Figure 7. Cladding hoop stress during PCMI phase of RIA transients of varying pulse widths.
A second related consequence of the cladding heating rate being much faster than the loading rate (asis the case in longer transients) is that the cladding fracture toughness (critical stress intensity factor KIC)
begins to increase dramatically with temperature21. While precise determination of fracture toughness
values for irradiated and hydrided zirconium alloy claddings are difficult to determine experimentally,
numerous separate effects studies show at least qualitatively that the recovery of fracture toughness withincreasing cladding temperature is rather dramatic4,11.
The HERA test matrix has been adapted to develop a proposed modeling matrix, which will be split
into two waves corresponding with the prehydrided cladding tests and the irradiated segment tests.
Several workshops are intended to be held to coordinate the effort and report results with participants.
Updates will be provided to the project advisory groups. Details of the first modeling wave are provided
below as a blind prediction of test results. The second modeling wave will follow a similar structure,
however specific details regarding inputs and outputs will be described at a later time. The second wave
will be focused on post-test evaluations with as-run conditions and preliminary experiment data support.
A description of these cases are listed in Table 4. All model inputs will be based on the design targets
for the HERA project, with the exception of variations in energy deposition and hydride content and rim
thickness specified in Table 5. Comparisons of stress intensity factors versus cladding temperature forexperiments that experienced PCMI failure will be compared to those that did not experience failure to
determine if critical temperature dependent fracture toughness (KIC) values can be inferred. These will becompared with out-of-pile mechanical test results and will be used to provide a basis for the analyticalprediction of pulse width impacts and thresholds on the current PCMI failure criteria.

16

Table 4. Past RIA experiments to be simulated using advanced tools as part of HERA.
Case #
Corresponding Test Capsule/
Pulse Width
TargetHydrogen
HERA Test ID Specimen
EnergyContent/Rim(ms)
Deposition* thickness(ppm/μm)
(J/g)
1
2
3
4
5
6
7
8
9
10
11
12

HERA-PreH1,2

NSRR

5

~50

650
650
650
550
750
650
650
650
550
750
650

400/80200/40600/140
400/80400/80400/80200/40600/140
400/80400/80400/80HERA-PreH3,4

TREAT

90

HERA-PreH5,6
HERA-PreH5,6

TREAT
TREAT

~300

650

400/80* Targeted energy deposition within 3 FWHM past the time of peak power.
A model prescription report will be provided that includes all required model inputs and output
requests. This will include specimen geometry, thermal hydraulic conditions, and power information. A
template for providing output data will also be provided to all participants.

8.

PREPARATIONS FOR HERA PHASE II

The follow-on phase of HERA (HERA Phase II) will likely focus on PIE of rods from phase I, thetesting of more HBU fuel rods, as well as coated zirconium alloy claddings and doped fuel pellets. A
principal aim of HERA Phase II will be utilizing test specimens from the same parent rods in experiments
at TREAT, NSRR and CABRI. Some material that has already been tested in CABRI and NSRR as wellout of pile mechanical studies may be available in various hotcells. To achieve this goal, work may beundertaken in HERA Phase I to identify relevant material and make arrangements to ship the material to
hot cell facilities in the U.S., Japan, and France where it can be staged for HERA Phase II work. Thesearrangements will be discussed and agreed to by all HERA Core Group and technical advisory groupmembers.

17

9.

SCHEDULE BUDGET AND DELIVERABLES

The entire project was originally expected to take three years to complete and start in January of 2021 and conclude in December of 2023.
However, the project has been delayed ~ 6months and is now planned to conclude in June of 2024. An updated project schedule is presented
below in Table 5. Light yellow boxes indicate the initial delays in tasks, the light blue boxes indicate further delays. Navy boxes indicate the initialschedule slips and dark navy boxes indicate further schedule slips. The project budget is still estimated to be 8.1 million USD. 4.35 million USD
will be funded by the HERA core group and 3.75 million USD will be funded by FIDES. It was originally anticipated that the distribution of the
FIDES funds would occur on an annual basis by the end of the first quarter for a given calendar year. However, it is now anticipated that the
FIDES funding for the first year (2021) will arrive in the first quarter of 2021, and $200k of the second year funding will be sent to JAEA in Julyof 2022 with subsequent funding will arrive in the fourth quarter in years 2022 and 2023. A breakdown of the budget by organization is shown
below in
Table 6.
Deliverables for the HERA project will include an interim and a final report. The interim report will be delivered in March of 2023 A final
project report including all the project results and conclusions and will be delivered in draft form in March of 2024. After incorporating commentsfrom a three-month review period and any final results, the final HERA report will be issued in July of 2024. Two HERA workshops will be held,
one at INL and one in Japan where FIDES participants will have the opportunity to tour facilities where the experimental work is taking place. The
first workshop will take place in the third quarter of 2022 and the second in the first quarter of 2024. All data described in the proposal will be
included in the reports and made available for inclusion in any data preservation activities undertaken by the FIDES members.

18

Table 5. HERA Schedule
2021
Q1

Task

2021
Q2

2021
Q3

2021
Q4

2022
Q1

2022
Q2

2022
Q3

Pre-Hydrided Test Fabrication
Pre-Hydrided Test Irradiations
Pre-Hydrided Test Post Transient Exams
HBU Test Fabrication & Characterization
HBU Test Irradiations
Modelling and Simulation Exercise
HERA Phase II Preparations

Table 6. HERA Budget
Contributing
Organization
FIDES
CORE Group
Total Cost

2021

2022

2023

750
750

1500
1800

1500
1800

1500

3300

3300

Total
Contribution
3750
4350
8100

19

2022
Q4

2023
Q1

2023
Q2

2023
Q3

2023
Q4

2024
Q1

2024
Q2

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