Fuel Fragmentation, 

Relocation, and Dispersal 

During the Loss-of-Coolant 

Accident

Office of Nuclear Regulatory Research

NUREG-2121
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Fuel Fragmentation, Relocation, 

and Dispersal During the 

Loss-of-Coolant Accident

Manuscript Completed: February 2012

Date Published: March 2012

Prepared by 

Patrick A.C. Raynaud

Office of Nuclear Regulatory Research

NUREG-2121

iii

Abstract

In light of recent results from the U.S. Nuclear Regulatory Commission loss-of-coolant accident 

(LOCA) research program, the staff of the Division of Systems Analysis in the Office of Nuclear 

Regulatory Research conducted a comprehensive review of past research programs for 

observations related to the phenomena of fuel fragmentation, relocation, and dispersal. The 

goal of this investigation was to determine whether these phenomena occur during a LOCA, and 

whether they were or should be incorporated into the criteria used to evaluate the acceptability 

of emergency core cooling systems. The review of over 90 LOCA test results performed in 

eight different programs over the last 35 years prompted the staff to conclude that 

fragmentation, relocation, and dispersal of fuel could not be precluded as possible phenomena 

during a LOCA. In addition, a number of conditions for the occurrence of these phenomena, as 

well as trends aggravating these phenomena, were derived from the analysis of the data 

compiled. The report also presents a preliminary assessment of the consequences of fuel 

fragmentation, relocation, and dispersal. The topics discussed are core damage distribution, 

fuel-coolant interaction, hydraulic and mechanical effects with relation to downstream effects, 

and radiological consequences. The preliminary assessment concludes that the consequences 

of fuel fragmentation and dispersal are not likely to result in an imminent safety hazard. This 

conclusion was made in consideration of the anticipated low number of fuel rods expected to 

burst and the conservative manner in which radiological consequences for a postulated LOCA 

are calculated.

v

Contents

Abstract......................................................................................................................................iii

Contents .....................................................................................................................................v

List of Figures ...........................................................................................................................vii

List of Tables .............................................................................................................................xi

Foreword.................................................................................................................................. xiii

Abbreviations and Acronyms.....................................................................................................xv

1 Introduction.......................................................................................................................... 1

2 Definitions............................................................................................................................ 3

2.1 Fuel Fragmentation...................................................................................................... 3

2.2 Fuel Relocation............................................................................................................ 3

2.3 Fuel Dispersal.............................................................................................................. 3

3 Regulatory History of Fuel Fragmentation, Relocation, and Dispersal ................................. 5

3.1 Circa 1973 ................................................................................................................... 5

3.2 Circa 1980 ................................................................................................................... 6

3.3 Circa 1995 ................................................................................................................... 6

3.4 Circa 2004 ................................................................................................................... 7

3.5 Circa 2006 ................................................................................................................... 7

3.6 Circa 2008 ................................................................................................................... 7

3.7 Today........................................................................................................................... 8

4 Experimental Results of Fuel Fragmentation, Relocation, and Dispersal ............................. 9

4.1 Integral LOCA Test Programs...................................................................................... 9

4.1.1 Power Burst Facility—Idaho National Engineering Laboratory ....................... 9

4.1.2 FR-2—Karlsruhe Institute of Technology ......................................................14

4.1.3 National Research Universal—Pacific Northwest Laboratory........................21

4.1.4 PHEBUS-LOCA—Institut de Radioprotection et de Sûreté Nucléaire 

(France) .......................................................................................................26

4.1.5 FLASH Tests—Commissariat à l’Energie Atomique (France) .......................27

4.1.6 Argonne National Laboratory LOCA Test Program .......................................29

4.1.7 Halden Boiling-Water Reactor LOCA Test Series .........................................30

4.1.8 Studsvik LOCA Test Program.......................................................................41

4.2 Integral LOCA Test Data.............................................................................................51

4.2.1 Data Tables ..................................................................................................51

4.2.2 Data Trends..................................................................................................71

4.2.3 Summary ......................................................................................................75

5 Consequences of Fuel Fragmentation, Relocation, and Dispersal......................................77
vi

5.1 Consequences of Fuel Relocation ..............................................................................77

5.2 Consequences of Fuel Dispersal ................................................................................77

5.2.1 Core Damage Distribution.............................................................................77

5.2.2 Fuel-Coolant Interaction................................................................................78

5.2.3 Hydraulic and Mechanical Effects of Dispersed Fuel Material .......................78

5.2.4 Radiological Effects of Dispersed Fuel Material ............................................79

6 Conclusions........................................................................................................................83

7 References.........................................................................................................................85

A. Appendix: Core Damage Distribution Assessment: Methodologies and Results ..............A-1
vii

List of Figures

Figure 4-1 Schematic of the Power Burst Facility in-pile instrumented test rig (Ref. 18)..........10

Figure 4-2 Metallographic cross-sections of rods 11 and 12 from the LOC-6 test, showing 

cladding ballooning, fuel fragmentation, and fuel relocation (Ref. 18)..................11

Figure 4-3 Axial cross-section of ballooned area in test LOC-3, rod 2, showing fuel melting 

(white central area) (Ref. 17)...............................................................................12

Figure 4-4 Neutron radiograph of test LOC-5, rod 7B, showing axial fuel relocation in a rod 

with limited fuel fragmentation (pellet fragments were relatively large, but 

relocation was nonetheless observed) (Ref. 17)..................................................12

Figure 4-5 Neutron radiograph of test LOC-6, rod 12, showing extensive axial fuel 

relocation (Ref. 18)..............................................................................................13

Figure 4-6 Schematic of the FR-2 LOCA test rig (Ref. 20) ......................................................15

Figure 4-7 Pre (top) and post (bottom) LOCA transient fuel fragmentation for low-burnup 

rods (2.5 GWd/MTU, left) and medium-burnup rods (35 GWd/MTU, right) 

(Ref. 19)..............................................................................................................16

Figure 4-8 Axial cross-section of nontransient-tested irradiated fuel rods showing the

developement of a rim structure at medium burnup (Ref. 19)..............................17

Figure 4-9 Cross-section of A rods from FR-2 in-pile test (unirradiated rods) (Ref. 20) ...........17

Figure 4-10 FR-2 in-pile tests: cross-sections of the test rods from series C 

(2.5 GWd/MTU) (Ref. 19) ....................................................................................18

Figure 4-11 FR-2 in-pile tests: cross-sections of the test rods from series E 

(8.0 GWd/MTU) (Ref. 19) ....................................................................................18

Figure 4-12 FR-2 in-pile tests: cross-sections of the test rods from series F 

(20 GWd/MTU) (Ref. 20) .....................................................................................19

Figure 4-13 FR-2 in-pile tests: cross-sections of the test rods from series G2/G3 

(35 GWd/MTU) (Ref. 20) .....................................................................................19

Figure 4-14 (a) Neutron radiograph of rod F1 (20 GWd/MTU), comparing pre- and 

posttransient fuel location (Ref. 19), (b) neutron radiograph of rod E5 

(8 GWd/MTU) after the transient (Ref. 19)...........................................................20

Figure 4-15 Pellet stack reduction as a function of cladding hoop strain for preirradiated 

fuel rods (based on the increase in volume in cladding balloon with no axial 

elongation) (Ref. 21)............................................................................................21

Figure 4-16 NRU-MT cross-section of the test assembly for the LOCA simulation program 

(Ref. 22)..............................................................................................................22

Figure 4-17 Circumferential strain profiles for all rods in tests MT-1, MT-2, MT-3, and 

MT-4, showing coplanar ballooning pinned by grid spacers (Ref. 26) (X-axis: 

strain in percent (%), Y-axis: rod length in inches)...............................................23

Figure 4-18 Ballooned area in tests MT-1 (left; ballooned and ruptured rods, with mostly 

intact pellets remaining in the rod) and MT-2 (right; ballooned and ruptured 

rods with some large fuel pellet fragments remaining in the rod) (Refs. 22 and 

23).......................................................................................................................23
viii

Figure 4-19 Ballooned and ruptured rod 3C in test MT-3: minimal fuel pellet 

fragmentation, pellets remaining in rod (Ref. 24) .................................................24

Figure 4-20 Ballooned and ruptured rod 3D in test MT-3: some large fuel pellet fragments 

remaining in the rupture node (Ref. 24)...............................................................24

Figure 4-21 Ballooned and ruptured rods in test MT-4 showing rupture node devoid of fuel 

(Ref. 25)..............................................................................................................25

Figure 4-22 PHEBUS test 215-P metallographic cross-sections at elevations 285 mm and 

252 mm from bottom of fuel stack (Ref. 27).........................................................27

Figure 4-23 PHEBUS test 219 metallographic cross-section at elevation 437 mm from 

bottom of fuel stack showing radial displacement of rods (Ref. 30) .....................27

Figure 4-24 Fuel aspect at rupture plane in FLASH-5 test (Ref. 32)........................................28

Figure 4-25 Low-magnification images of the post-LOCA test ICL#2 fuel samples (a) at 

≈12 mm above the rupture center (15–35% strain), (b) at ≈50 mm above the 

rupture center (2–4% strain), (c) at ≈130 mm below the rupture center, and 

(d) prior to LOCA testing (180 mm from the LOCA sample) (Ref. 33)..................29

Figure 4-26 Ballooned region with relocated fuel in the Halden IFA-650.2 test (Ref. 61).........31

Figure 4-27 Through-wall cracks with brittle appearance, Halden IFA-650.2 test (Ref. 61)......32

Figure 4-28 Gamma scan of a very-high-burnup (approximately 92 GWd/MTU) fuel rod 

showing major loss of fuel material after LOCA testing (IFA-650.4).....................33

Figure 4-29 Gammas scans (Cs-137 and Ru-103 activity) of IFA-650.9 about 6 weeks 

after the test. Left: full scan at 0 degrees. Right: flask rotated 90 degrees 

counterclockwise. Resolution: 5 mm in vertical and 1 mm in horizontal 

direction. The rig drawing in the middle shows the instrument levels. 

(Ref. 45)..............................................................................................................35

Figure 4-30 Fuel residue from posttest disassembly of IFA-650.9 (Ref. 46) ............................36

Figure 4-31 Fragmentation of the rim structure in IFA-650.9 (Ref. 46) ....................................36

Figure 4-32 The rupture opening in IFA-650.10 (Ref. 49)........................................................37

Figure 4-33 Fuel fragmentation in IFA-650.10 (Ref. 49) ..........................................................37

Figure 4-34 Neutron radiography of IFA-650.10 (Ref. 49) .......................................................38

Figure 4-35 Fuel fragmentation and relocation as a function of cladding diameter increase 

in test IFA-650.10 (Ref. 49) .................................................................................39

Figure 4-36 Neutron radiography of test IFA-650.11 (Ref. 50).................................................40

Figure 4-37 Fuel fragmentation and relocation as a function of cladding diameter increase 

in test IFA-650.11 (Ref. 50) .................................................................................40

Figure 4-38 Rupture opening in Studsvik LOCA tests 189, 191, 192, and 193 (left to right), 

showing the absence of fuel in the rupture plane.................................................44

Figure 4-39 Rupture opening in Studsvik LOCA tests 196 and 198 (left to right).....................44

Figure 4-40 Particle size distribution from six integral LOCA tests...........................................45

Figure 4-41 Images of fuel particles collected from test rod (a) 192 and (b) 193 revealing a 

very small, sand-like fragmentation size ..............................................................45
ix

Figure 4-42 Images of fuel particles collected from test rod (a) 196 and (b) 198 revealing 

large fragments ...................................................................................................46

Figure 4-43 Fuel collected beneath the LOCA test train (a) just after the fuel rod is 

removed, and (b) 2 days after the fuel rod is removed.........................................46

Figure 4-44 Axial extent of fuel loss in comparison to measured final strain in test 189...........47

Figure 4-45 Axial extent of fuel loss in comparison to measured final strain in test 191...........48

Figure 4-46 Axial extent of fuel loss in comparison to measured final strain in test 192...........48

Figure 4-47 Axial extent of fuel loss in comparison to measured final strain in test 193...........48

Figure 4-48 Axial extent of fuel loss in comparison to measured final strain in test 196. No 

fuel loss was observed during the LOCA simulation for this test..........................49

Figure 4-49 Axial extent of fuel loss in comparison to measured final strain in test 198. No 

fuel loss was observed during the LOCA simulation for this test..........................49

Figure 4-50 Average fuel fragment cross-section as a function of burnup, showing 

increased fragmentation with increasing burnup..................................................72

Figure 4-51 Rupture width as a function of rod fill pressure, showing increasing rupture 

width with increasing rod fill pressure ..................................................................73

Figure 4-52 Length of ballooned section of the rod as a function of (a) rod fill pressure, 

and (b) rupture to plenum distance, showing that balloon length decreases 

with increase rod fill pressure, and increases when the rupture occurs further 

from the plenum. .................................................................................................73

Figure 4-53 Rupture pressure (a) and temperature (b) as a function of rod fill pressure, 

showing opposite trends, implying that rupture pressure and temperature are 

inversely correlated, as shown in Figure 4-54......................................................73

Figure 4-54 Rupture temperature as a function of rupture pressure, showing that rupture 

temperature decreases as rupture pressure increases........................................74

Figure 4-55 Rupture area as a function of (a) rupture length and (b) rupture width, 

showing that rupture area increases with longer and wider ruptures....................74

Figure 4-56 Rupture width as a function of rupture length, showing that rupture width and 

length increase together, with rupture width generally being between on half 

and one fifth of the rupture length........................................................................74

Figure A-1 Determination of fuel failure rate after LOCA with deterministic method 

(Siemens case) (Ref. A.2) .................................................................................A-2

Figure A-2 Probability for fuel rod failure as a function of rod power (W/cm) at different 

burnups (Siemens case) (Ref. A.2)....................................................................A-3

Figure A-3 Power and burnup distribution in a core with failure threshold. Each number 

gives the number of fuel rods at the particular power-burnup combination 

(Siemens case). (Ref. A.2) ................................................................................A-4

Figure A-4 Power and burnup distribution in BWR GE-14 core with best-estimate (blue) 

and conservative (green) failure thresholds, for (a) best-estimate core and (b) 

conservative core (Ref. A.5)..............................................................................A-6

Figure A-5 Peak cladding temperature response for deterministic assessment of extent of 

rupture (Ref. A.6)...............................................................................................A-8
x

Figure A-6 Flow chart for statistical damage assessment for German PWRs (Ref. A.7)......A-10

Figure A-7 Map of core damage extent in German PWR after a large-break LOCA (text is 

visible in magnified view) (Ref. A.7).................................................................A-10

Figure A-8 Number of failed fuel rods/failure rate during LOCA (Ref. A.7)...........................A-11

Figure A-9 Best-estimate failure thresholds for PWR fuel in a large-break LOCA (Ref. A.2)A-12
xi

List of Tables

Table 4-1 Measurements of fuel loss for each test conducted to date.....................................43

Table 4-2 Summary of Power Burst Facility (PBF) Test Data (1 of 4)......................................52

Table 4-3 Summary of Power Burst Facility (PBF) Test Data (2 of 4)......................................53

Table 4-4 Summary of Power Burst Facility (PBF) Test Data (3 of 4)......................................54

Table 4-5 Summary of Power Burst Facility (PBF) Test Data (4 of 4)......................................55

Table 4-6 Summary of FR-2 Test Data (1 of 8) .......................................................................56

Table 4-7 Summary of FR-2 Test Data (2 of 8) .......................................................................57

Table 4-8 Summary of FR-2 Test Data (3 of 8) .......................................................................58

Table 4-9 Summary of FR-2 Test Data (4 of 8) .......................................................................59

Table 4-10 Summary of FR-2 Test Data (5 of 8) .....................................................................60

Table 4-11 Summary of FR-2 Test Data (6 of 8) .....................................................................61

Table 4-12 Summary of FR-2 Test Data (7 of 8) .....................................................................62

Table 4-13 Summary of FR-2 Test Data (8 of 8) .....................................................................63

Table 4-14 Summary of PHEBUS-LOCA Test Data ................................................................64

Table 4-15 Summary of FLASH Test Data..............................................................................65

Table 4-16 Summary of Argonne National Laboratory Test Data ............................................66

Table 4-17 Summary of Halden Test Data (1 of 2)..................................................................67

Table 4-18 Summary of Halden Test Data (2 of 2)..................................................................68

Table 4-19 Summary of PNL/NRU Test Data..........................................................................69

Table 4-20 Summary of Studsvik LOCA Test Data .................................................................70

Table 4-21 Trends Identified from Past LOCA Test Programs (Green = Correlation, Red = 

Inverse Correlation, Gray = No Correlation)..........................................................71

Table 5-1 The Release from the Gap and Fuel for Radiologically Significant Nuclides............80

Table 5-2 Radiological Release in Halden LOCA Tests ..........................................................81

Table A-1 Best-Estimate Analyses of Percentage of Fuel Rod Failures (Ref. A.2) ................A-5

Table A-2 2004 Iberdrola Study Percentage of Failed Bundles, GE-14 Fuel (Ref. A.5) .........A-6

Table A-3 WCOBRA/TRAC LOCA Simulation Results for Peak Cladding Temperature 

Above 925°C (Ref. A.6) ......................................................................................A-8

xiii

Foreword

The purpose of this document is to capture our current understanding of fuel fragmentation, 

relocation, and dispersal during the hypothetical loss-of-coolant accident (LOCA), and the 

implications of these phenomena within the U.S. regulatory framework. The review captures the 

results of the U.S. Nuclear Regulatory Commission’s integral LOCA research program, as well 

as the results of other LOCA research programs carried out in the United States and abroad.

Research suggests that current knowledge and modeling technology may be sufficient to 

account for the phenomena of fuel fragmentation and axial relocation. However, current 

knowledge is not sufficient to determine the size of the rupture opening or to quantify the extent 

of fuel dispersal. Furthermore, a complete assessment of the consequences of fuel dispersal is 

not possible at this time, due to insufficient information about the downstream effects of fuel 

entrained in the coolant. Therefore, research suggests that initiating a detailed LOCA analysis 

with existing LOCA analysis codes and methods will not provide resolution or provide additional 

assurance of plant safety in the event of fuel dispersal during a LOCA.

Further understanding and resolution of this issue will require additional experimental work in 

order to develop models and support analysis assumptions. Programs at Studsvik and Halden 

are performing additional tests that should provide more information on fuel dispersal, including 

the behavior of medium-burnup fuel and particle-size analysis. The NRC will monitor these 

programs closely and will evaluate the results and conclusions against the observations and 

conclusions of this report.

xv

Abbreviations and Acronyms

AEC U.S. Atomic Energy Commission

ANL Argonne National Laboratory

ANP Advanced Nuclear Power

BWR boiling-water reactor

C Centigrade

CEA Commissariat à l’Energie Atomique (French Atomic Energy Commission)

CFR Code of Federal Regulations

cm centimeter

Cs cesium

CSNI Committee on the Safety of Nuclear Installations

ECCS emergency core cooling system

ECR equivalent cladding reacted

EdF Electricité de France (Electricity of France)

EG&G Edgerton, Germeshausen & Grier

F Fahrenheit

FCI fuel-coolant interaction

g gram

GE General Electric

GI generic issue

GSI generic safety issue

GWd/MTU gigawatt-day per metric ton of uranium

GWd/t gigawatt-day per ton

GWd/tU gigawatt-day per ton uranium

HBWR Halden Boiling-Water Reactor

I iodine

IBERINCO Iberdrola Ingeniería y Construcción (Iberdrola Engineering and Construction)

IFA instrumented fuel assembly

INEL Idaho National Engineering Laboratory

IPSN Institut de Protection et de Sûreté Nucléaire (French Institute for Nuclear 

Protection and Safety)

IRSN Institut de Radioprotection et de Sûreté Nucléaire (French Institute for Radiation 

Protection and Nuclear Safety)

K kelvin

KfK Karlsruhe Institute of Technology

Kr krypton

kW/ft kilowatt per foot

kW/m kilowatt per meter

KWU Kraftwerk Union (now part of Siemens)

LHGR linear heat generation rate

LOC loss-of-coolant

LOCA loss-of-coolant accident

LWR light-water reactor

m meter

mm millimeter

MOX mixed-oxide

MPa megapascal

MT materials test

MWd/t megawatt-day per ton
xvi

MWd/kg megawatt-day per kilogram

MWt megawatt thermal

NEA Nuclear Energy Agency

NNC National Nuclear Corporation

NRC U.S. Nuclear Regulatory Commission

NRG Nuclear Research and consultancy Group (Netherlands)

NRU National Research Universal

OECD Organization for Economic Co-operation and Development

PBF Power Burst Facility

PCT peak cladding temperature

PIE postirradiation examination

PNL Pacific Northwest Laboratory

psi pound(s) per square inch

PWR pressurized-water reactor

RES Office of Nuclear Regulatory Research (NRC)

RIL research information letter

Ru ruthenium

s second

TID technical information document

UO2 uranium dioxide

VVER water-water energetic reactor (Russian pressurized water reactor)

W/cm watt(s) per centimeter

wt ppm weight parts per million

Xe xenon
1

1 Introduction

In May 2008, the U.S. Nuclear Regulatory Commission’s (NRC’s) Office of Nuclear Regulatory

Research (RES) issued Research Information Letter (RIL)-0801, “Technical Basis for Revision 

of Embrittlement Criteria in 10 CFR 50.46 (Ref. 1), which provided a technical basis for revising

the loss-of-coolant accident (LOCA) cladding embrittlement criteria found in Title 10 of the Code 

of Federal Regulations (10 CFR) 50.46, “Acceptance Criteria for Emergency Core Cooling 

Systems for Light-Water Nuclear Power Reactors” (Ref. 2). RIL-0801 discussed axial fuel

relocation and the loss of fuel particles through a rupture opening and recommended further

research in these areas. Prompted by recent findings, the purpose of the current document is to

revisit the conclusions of RIL-0801 with respect to fuel relocation, fragmentation, and dispersal

during a hypothetical LOCA. This was accomplished by the review of historical and more recent

data to determine if there are trends or observations that can be made beyond those in RIL0801. RES then used these trends and observations to characterize the likelihood of fuel

fragmentation, fuel relocation, and fuel dispersal under LOCA conditions. This report also 

describes the potential consequences of these phenomena.

With respect to axial fuel relocation, historical attention has primarily focused on the

consequence of an increase in heat generation in the ballooned region, with a resulting increase

in cladding temperature and oxidation compared with an unballooned length of the fuel rod. 

RIL-0801 noted the low prioritization given to this issue in the NRC Generic Issues (GIs)

Program, where it was identified as GI-92, “Fuel Crumbling during LOCA.” The low prioritization

was apparently assigned because of compensating conservatisms in the analyses of 

Appendix K, “ECCS Evaluation Models,” to 10 CFR Part 50, “Domestic Licensing of Production 

and Utilization Facilities” (Ref. 2). RIL-0801 further suggested that, with the move to bestestimate methodologies, the low prioritization for this issue may no longer be appropriate. 

Regarding the issue of fuel particles being dispersed through a rupture opening in the fuel

cladding, RIL-0801 concluded that the current NRC burnup limit of 62 gigawatt-days per metric 

ton of uranium (GWd/MTU) (average for the peak rod) was probably low enough to prevent

significant fuel loss during a LOCA. Based on the current study, a burnup limit of 62 GWd/MTU 

by itself no longer appears sufficient to preclude fuel dispersal during the hypothetical LOCA. 

The following pages discuss this conclusion further.

3

2 Definitions

For the purpose of this report, it is important to define the following three terms: fuel

fragmentation, fuel relocation, and fuel dispersal. This section provides the definitions used in

this report. The staff’s assessments of the test results described in Section 4 of this report are

based on these definitions.

2.1 Fuel Fragmentation

Fuel fragmentation refers to any separation of the fuel pellet into more than one piece, 

regardless of when or why it occurred. During normal operation, oxide fuel pellets develop

many cracks because of thermal stresses. At higher values of burnup, fission gas production

and migration is postulated to generate a “rim” region in fuel pellets that is highly porous. In

addition, during LOCA conditions, additional fragmentation is postulated to occur because of the

thermal-mechanical response to the transient.

2.2 Fuel Relocation

If fuel pellets are fragmented and separated from each other, they could be free to move relative

to their neighbors. Simply stated, fuel relocation can be described as any physical movement of

fuel pellets or fuel fragments within the cladding. Generally, radial fuel relocation is described

as distinct from axial fuel relocation.

Radial fuel relocation is the movement of the fuel outward toward the fuel cladding. 

Measurements in instrumented test rods consistently show lower fuel centerline temperatures

than those predicted based only on fuel and cladding thermal expansion. Microscopic

examination of postirradiation fuel cross-sections has led to the conclusion that fuel pellet

cracking promotes an outward relocation of the pellet fragments that causes additional gap

closure. This process is widely recognized in fuel performance analysis. It starts at beginning

of life and quickly reaches equilibrium—by 5 GWd/MTU, according to the FRAPCON-3.4

computer code (Ref. 3).

Axial fuel relocation is the vertical movement of fuel fragments or particles within the cladding. 

Under normal operation, this process is usually limited by the fuel pellet immediately above or

below the pellet in question. For the purpose of this report, axial fuel relocation is said to have

occurred if postirradiation examination (PIE) reveals that fuel fragments have moved axially

relative to their original location. Evidence that would support this determination includes voided

regions of the cladding rod or the observation of additional fuel material in the enlarged volume

of the balloon region, or both. 

In the remaining discussion, “fuel relocation” refers to “axial fuel relocation.”

2.3 Fuel Dispersal

Fuel dispersal is the ejection of fuel fragments or particles through a rupture or opening in the

cladding. For the purpose of this report, fuel dispersal is said to have occurred if any fuel

material is found outside of the fuel rod. Even if the fuel material is small in quantity, the finding

will be noted and qualified by the nature of the dispersal (e.g., “only a small black powder on the

test chamber wall was observed”).

5

3 Regulatory History of Fuel Fragmentation, Relocation, and

Dispersal

3.1 Circa 1973

Much of this historical information has been selectively extracted from Hache and Chung’s 2000 

paper, “The History of LOCA Embrittlement Criteria” (Ref. 4).

In 1967, the U.S. Atomic Energy Commission (AEC) Advisory Task Force on Power Reactor

Emergency Cooling, appointed to provide additional assurance that substantial meltdown is

prevented by core cooling systems, concluded the following:

The analysis of (a LOCA) requires that the core be maintained in place and

essentially intact to preserve the heat-transfer area and coolant-flow geometry. 

Without preservation of heat-transfer area and coolant-flow geometry,

fuel-element melting and core disassembly would be expected.... Continuity of

emergency core cooling must be maintained after termination of the temperature

transient for an indefinite period until the heat generation decays to an

insignificant level, or until disposition of the core is made. (Ref. 5) 

This rationale made it clear that it is important to preserve both the heat transfer area and the

coolant flow geometry, not only during the short-term portion of the core temperature transient

but also for the long term.

Consistent with the conclusions of this task force, AEC issued general design criteria

(Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50 (Ref. 2))

such that “fuel and clad damage that could interfere with continued effective core cooling is

prevented.” At this time, AEC also issued interim acceptance criteria (Ref. 6) for emergency

core cooling systems (ECCS) for light-water reactors (LWRs).

These criteria were subjected to a rulemaking hearing in 1973, and the proceedings were well

documented in the Journal of Nuclear Safety in 1974 (Refs. 7 and 8). The peak cladding

temperature (PCT) and maximum oxidation limits, now found in 10 CFR 50.46(b) (Ref. 2), were

established during this hearing process.

During the rulemaking hearing in 1973, the following remarks were made part of the formal

record:

The purpose of these first two criteria [peak cladding temperature and maximum

oxidation] is to ensure that the Zircaloy cladding would remain sufficiently intact

to retain the UO2 fuel pellets in their separate fuel rods and therefore remain in

an easily coolable array. Conservative calculations indicate that during the

postulated LOCA, the cladding of many of the fuel rods would swell and burst

locally with a longitudinal split. The split cladding would remain in one piece if it

were not too heavily oxidized, and would still restrain the UO2 pellets.

[emphasis added] (Ref. 9) 
6

In further discussion on “coolable geometry,” AEC noted the following:

If there were no emergency core cooling after a LOCA, the core would probably

eventually fuse together into a large mass with insufficient external surface area

to allow the fission product heat generated within it to be transferred away. 

Intermediate steps in arriving at such a state might be the oxidation and melting

of the Zircaloy cladding, allowing the uranium dioxide fuel pellets to fall

together into a heap that would be difficult to cool. [emphasis added] 

(Ref. 9)

Examination of the historical record indicates that, at the time the cladding criteria were

developed, there was no expectation of fuel loss—at least, no expectation of significant

3.2 Circa 1980

fuel

loss—during a successfully mitigated LOCA.

Experiments conducted at several test facilities prior to 1984 (Power Burst Facility (PBF), FR-2, 

National Research Universal (NRU), ESSOR, and PHEBUS-LOCA; see Section 4) showed that

irradiated fuel could fragment (crumble) into small pieces during a LOCA and may relocate

axially, settling into ballooned regions (Section 4 of this report discusses these experiments). In

1984, the NRC classified this effect as GI-92. Edgerton, Germeshausen & Grier (EG&G)

evaluated this effect for the NRC in 1983 (Ref. 10). EG&G’s evaluation focused on the

consequences of an increase in heat generation in the ballooned region, with the corresponding

increase in cladding temperature and oxidation compared with an unballooned length of the fuel

rod.

At the time of the initial evaluation of this issue in July 1984, the existing ECCS performance

analysis codes did not account for fuel settling into ballooned regions. Thus, the lack of

inclusion of this effect was a nonconservatism. However, the EG&G study (Ref. 10) concluded

that known conservatisms in Appendix K to 10 CFR Part 50 would more than offset this effect,

and the issue was therefore given a low priority. This determination meant that there was

insufficient risk-based justification for starting a major re-review of existing ECCS performance 

analyses under Appendix K to 10 CFR Part 50. 

3.3 Circa 1995

By the early 1990s, there were ongoing efforts to develop and license ECCS performance

models that were more realistic (and consequently less conservative) than the models in use at

the time of the evaluation in July 1984. As summarized in NUREG-0933, “Resolution of Generic 

Safety Issues” (Ref. 11), with the move toward best-estimate models, it was no longer valid to

conclude that the effects of fuel crumbling and settling into ballooned regions could necessarily

be neglected in any new, more realistic models. Instead, the NRC expected that these effects

(which are real physical phenomena) would be appropriately addressed in the calculations. 

Moreover, the NRC staff determined that a separate generic issue on fuel crumbling was not

necessary; such work was best done within the scope of the review of the new calculational

methodology. Thus, the issue was given a low priority (see Appendix C to NUREG-0933

(Ref. 11)). NUREG/CR-5382, “Screening of Generic Safety Issues for License Renewal 

Considerations,” issued December 1191 (Ref. 12), concluded that consideration of a 20-year

license renewal period did not change this priority. Further prioritization, using the conversion

factor of $2,000/man-rem approved by the Commission in September 1995, resulted in an

impact/value ratio (R) of $50,000/man-rem, which placed the issue in the “drop” category. 
7

3.4 Circa 2004

In 2004, axial fuel relocation was the central issue in a hearing about the insertion of four lead

test assemblies of mixed plutonium-uranium oxide (mixed-oxide) fuel in the Catawba Nuclear 

Station plant. The contention was that axial fuel relocation would be worse in mixed-oxide fuel

than in standard uranium dioxide (UO2) fuel. The NRC staff successfully argued that there

would be no significant difference, but the testimony did not deny the importance of the effect

itself.

3.5 Circa 2006

The regulatory history recalled above is related to the phenomena of fuel fragmentation and

relocation. When the consequences of these phenomena were evaluated at different points in

history, the NRC determined that a new regulatory requirement or change in analysis

methodology was not needed. However, fuel designs and operating trends have evolved

significantly since these determinations were made. In addition, and arguably more significant,

there was no expectation from any prior research that fragmentation and relocation into the

balloon region could result in the loss of fuel particles through the rupture opening. When

integral LOCA tests were recently completed at Argonne National Laboratory (ANL) on

high-burnup boiling-water reactor (BWR) rods with a local burnup of 64 GWD/MTU, a small

amount of fuel loss was noticed (about the quantity of one fuel pellet). Because the amount of

material was small, this observation was not thought to be important. However, in April 2006, a

LOCA test, IFA-650.4, was run in the Halden reactor on a fuel rod segment with a very high

local burnup of 91.5 GWd/MTU (the Halden experiments are discussed in Section 4 of this

report). Results from this test recently presented to members of the Halden Reactor Project

showed gross loss of fuel material from above the rupture opening (Ref. 13). Online

instrumentation indicated that this fuel loss occurred during the temperature transient rather

than after the test was over. In this very-high-burnup fuel specimen, more than 40 percent of

the fuel material was in a nearly powdered form, referred to as a “rim.” This rim material was

able to flow freely under the influence of gravity and pressure differences out of the rupture

opening (about 5 centimeters or 2 inches long).

3.6 Circa 2008

In 2008, RES prepared RIL-0801, which discussed research findings in the area of high-burnup

fuel performance under postulated LOCA events. 

On the topic of fuel relocation, RIL-0801 noted this history of GI-92:

After the best estimate option was added to 10 CFR 50.46, “Acceptance Criteria

for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors,”

those conservatisms were no longer guaranteed and fuel relocation was again

prioritized. This time it was classified as “drop,” but the evaluation that led to this

classification may not have been adequate [Refs. 14 and 15]. That evaluation

appears to have overlooked the possibility that rapid cladding embrittlement

would occur at the assumed cladding temperature of 1427°C (2600 °F) and that

embrittled fuel rods might collapse. NRC and Halden programs are performing

additional testing to resolve this issue and the resolution of GI-92 will be

documented when the testing is completed.
8

With the results of Halden’s IFA-650.4 available, the RIL included a discussion of fuel dispersal. 

The RIL noted that additional integral testing was being performed and suggested that these

tests should provide more definitive information on fuel loss during a LOCA with high-burnup

fuel. However, RIL-0801 concluded that “the current NRC burnup limit of 62 GWd/MTU

(average for the peak rod) is probably low enough to prevent significant fuel loss during a

LOCA.” Rulemaking to revise ECCS criteria began with this conclusion. 

3.7 Today

As part of the NRC’s ongoing integral LOCA program, four integral LOCA tests were run at

Studsvik laboratory with high-burnup fueled rods. These fuel rods had a rod average burnup

near 70 GWd/MTU (Section 4 of this report discusses the Studsvik experiments). In these tests,

significant fuel loss was observed. The burnup of the rods tested at Studsvik is still above the 

current NRC burnup limit of 62 GWd/MTU; however, it is closer to the limit than that of the

Halden experiments and therefore generated greater concern than the findings of IFA-650.4. In

addition, the fuel fragmentation size of the dispersed fuel could be readily observed in the

Studsvik tests, and the fuel fragments were observed to be fine. 

The observations of the Studsvik tests led to two main questions:

(1) What conditions and variables are important in determining the likelihood of fuel

fragmentation, fuel relocation, and fuel loss under LOCA conditions?

(2) Are these results observed during testing of lower burnup fuel?

To answer these questions, RES reviewed a wide range of historical data, including the

Studsvik results, to determine if there are trends and observations available in the existing data

when evaluated as a whole and in light of more recent findings. 
9

4 Experimental Results of Fuel Fragmentation, Relocation, and

Dispersal

4.1 Integral LOCA Test Programs

This section describes integral testing programs designed to investigate LOCA-related

phenomena using a bundle of test rods (the Pacific Northwest Laboratory (PNL)/NRU and

PHEBUS-LOCA programs) or a single test rod (all other programs) submitted to a LOCA

transient. Only one program used a full-length fuel rod (PNL/NRU); the others used segments

of varying length. In all cases, the test rods contained fuel, sometimes previously irradiated. 

The majority of the LOCA test programs described below took place in test reactors, using

nuclear heating, but a few (ANL, Halden, and Studsvik) used external heating of the fuel rods. 

This report presents the LOCA test programs in more or less chronological order. For each test

program, a description of the test design and important parameters is provided, as well as test

results that pertain to fuel fragmentation, relocation, and dispersal.

This report presents only testing on actual fueled rods or segments thereof, so as to assess the

phenomena of fragmentation, relocation, and dispersal inherently related to the presence of

fuel. Other research programs have performed a number of tests on defueled cladding tubes in

order to assess thermal-mechanical behavior, but they are not presented here. This report

presents all of the testing programs that the staff is aware of and for which information was

available. Despite the staff’s efforts to find all of the information available, it is possible that

some testing programs are not documented here. However, although the information contained

in this report may not be exhaustive, the staff deemed it sufficient to draw a number of

conclusions about fuel fragmentation, relocation, and dispersal. 

4.1.1 Power Burst Facility—Idaho National Engineering Laboratory

Four transient tests were conducted in the PBF at Idaho National Engineering Laboratory (INEL)

between 1978 and 1983 (Refs. 16, 17, and 18). Each test included four rods that were each in

a separate shroud within an instrumented test rig, as shown in Figure 4-1. The shape of the

individual rod shrouds simulated the shape of adjacent rods. The fuel rod segment had an

active fuel length of about 91 centimeters (cm) and a plenum length of about 11.5 cm. The fuel

design was typical pressurized-water reactor (PWR) 15-by-15 array rods with Zircaloy-4

cladding but 9.6-percent enriched UO2 fuel, which is higher than the 5-percent limit currently

used for LWR fuel in the United States. Although the effect of enrichment has not been studied 

with regards to fuel fragmentation, relocation, and dispersal during a LOCA, it is possible that 

higher enrichment may lead to enhanced fragmentation and fission gas release in the rim region 

of the pellet. The PBF-LOC experimental program was designed to investigate rod ballooning

and failure in the event of a LOCA; therefore, the main test parameters were burnup and rod fill

pressure.
10

Figure 4-1 Schematic of the Power Burst Facility in-pile instrumented test rig (Ref. 18)

Test LOC-11 was the first test of the PBF-LOC series and consisted of four unirradiated

calibration rods. Two rods were essentially unpressurized and two others were pressurized at

350 pounds per square inch (psi) and 700 psi, respectively. These tests resulted in negative

cladding strains for the two low-pressure rods and small balloons (1.5-percent and 3-percent

strain) that increased in size with increasing rod fill pressure. The subsequent LOC-3, LOC-5,

and LOC-6 tests were designed to cause rupture at different temperatures: LOC-3 in the alpha

plus beta region of the zirconium phase diagram (approximately 915 degrees Centigrade (C)),

LOC-5 in the beta region (approximately 1.093 degrees C), and LOC-6 in the alpha region

(approximately 798 degrees C). A short preirradiation period in-pile preceded the actual

transient testing. The rods experienced variable linear heat generation rates during

preirradiation, on average about 12 kilowatts per foot (kW/ft) with peak ratings as high as

18 kW/ft, which would be expected to promote more extensive fuel cracking than normal reactor

conditions.

Flow shroud

Flux shaper (stainless steel, 

3mm thick

Flux wires

Inside shroud coolant 

thermocouples

Outside shroud surface 

thermocouples

Differential thermocouple

Inlet/outlet thermocouple

Rod-to-rod pitch – 49.39 mm

Cladding thermocouples 

(elevation given relative to all 

rods are identical)

Self-powered neutron detectors

Self-powered gamma detectors

Zircaloy-4 support tube – 10.7 

mm outer diameter
11

Figure 4-2 Metallographic cross-sections of rods 11 and 12 from the LOC-6 test,

showing cladding ballooning, fuel fragmentation, and fuel relocation

(Ref. 18)

As seen in Figure 4-2, fragmentation was observed after the LOCA transient even for fresh fuel

rods, and fragmentation generally increased with burnup. Furthermore, significant relocation of

fuel fragments was observed in the plane of the balloon, and the relocation phenomenon was

eased by increased fragmentation (at higher burnup) and by larger circumferential ballooning

strains. It should be noted that in test LOC-3, rod 2, the combination of relocation and burnup

(16 GWd/MTU) resulted in fuel melting near the rod centerline, as observed in Figure 4-3. 

Finally, fuel dispersal was also observed and reported for LOC-3, rod 2.

Cladding

Epoxy

Fuel

Cladding

Epoxy

Fuel

Burst Tips

Burst Tips

Fresh rod 11

31% total circumferential elongation

1098 K estimated burst temperature

Previously irradiated rod 12

74% total circumferential 

elongation

1066 K estimated burst 

temperature

1.82 mm

1.67 mm
12

Figure 4-3 Axial cross-section of ballooned area in test LOC-3, rod 2, showing fuel

melting (white central area) (Ref. 17)

Figure 4-4 and Figure 4-5 are neutron radiographs of LOC-5, rod 7B (fresh fuel), and LOC-6,

rod 12 (10.8 GWd/MTU), respectively. It can be seen that both tests resulted in rod ballooning

and fuel relocation. The fuel relocation was more pronounced in LOC-6, rod 12, which had a

higher burnup and consequently underwent more significant fuel fragmentation. The additional

fragmentation in LOC-6, rod 12, compared to LOC-5, rod 7B, combined with the larger

ballooning strain (74 percent versus 48 percent), facilitated the movement of fuel particles from

the portion of the rod above the ballooning plane into the ballooned region.

Figure 4-4 Neutron radiograph of test LOC-5, rod 7B, showing axial fuel relocation in a

rod with limited fuel fragmentation (pellet fragments were relatively large, but

relocation was nonetheless observed) (Ref. 17)

1.00 mm

Melted fuel region

Epoxy

Plenum region Fuel gap

End cap Cladding rupture Balloon Fractured fuel pellets

0.5 0.4 0.3 0.2 0.1 0 -0.02

0.9 0.8 0.7 0.6 0.5

Distance from bottom of fuel rod (m)
13

Figure 4-5 Neutron radiograph of test LOC-6, rod 12, showing extensive axial fuel

relocation (Ref. 18)

In summary, the staff draws the following conclusions from the PBF tests:

• Fragmentation is a real phenomenon that appears to increase with burnup.

• Fuel relocation can occur a result of fragmentation and rod ballooning.

• Fuel relocation into the balloon can result in fuel melting.

Distance from bottom of fuel stack (cm)

Bottom of fuel stack

Fuel at rupture location

Plenum region
14

• Fragmented fuel can be dispersed from ballooned and ruptured fuel rods.

• Fuel dispersal can occur as a result of a LOCA transient.

4.1.2 FR-2—Karlsruhe Institute of Technology

Concurrently with the PBF test program in the United States, the Karlsruhe Institute of 

Technology (KfK) in Germany conducted the FR-2 program (Refs. 19, 20, and 21). A total of 39

fueled rods were tested in the FR-2 reactor between 1978 and 1983. PWR fuel rods with UO2

fuel enriched to 4.9 percent and clad with Zircaloy-4 were refabricated into test rods about 1

meter in length, with 50 cm of active fuel length. The initial refabricated rod fill pressure was

43.5 psi, but the rods were irradiated to different burnups (from 0 to 35 GWd/MTU), resulting in

a wide range of rod internal pressures at the beginning of the LOCA transient. The linear heat 

generation rate during base irradiation varied from 20 kW/m to 45 kW/m. The preirradiation in

FR-2 resulted in low corrosion and hydrogen. The transient was initiated from a steady-state

steam environment around 300 degrees C and 6 megapascals (MPa). The blowdown ramp 

occurred in 8 to 10 seconds and resulted in a pressure around 10,000 Pascals (Pa). The rod 

heatup rate was between 6 degrees C per second (s) and 20 degrees C per second (s) and was 

obtained by maintaining around 4 kilowatts (kW) per meter (m) linear heat generation rate until 

the temperature reached 927 degrees C. The reactor was scrammed at 927 degrees C,

following which the rods reached the PCT within a certain period of time. The transient was

ended by a steam quench at 727 degrees C. Figure 4-6 shows the test rig used in FR-2. The

main parameter of the FR-2 tests was burnup, but a wide range of parameters were measured,

and the FR-2 database is one of the more extensive LOCA test program databases.
15

Figure 4-6 Schematic of the FR-2 LOCA test rig (Ref. 20)

The large number of micrographs from PIE of rods tested in the FR-2 program allows for a fairly

extensive study of the effects of burnup on fuel fragmentation and relocation. Figure 4-7 shows

cross-sections of very-low-burnup (2.5 GWd/MTU) and medium-burnup (35 GWd/MTU) rods

that have simply been irradiated to these levels without being transient tested, and others that

have gone through a LOCA transient. It can be seen that fuel fragmentation had already

occurred before the LOCA transient, and that the transient did not appear to have a marked

effect on fuel fragmentation. Regarding fuel relocation, the information in Figure 4-7 indicates 

that radial relocation increases when rod ballooning occurs, as is particularly visible when

comparing the micrographs of the low-burnup rods. It should also be pointed out that the rod

irradiated to 35 GWd/MTU and not transient tested shows the signs of rim formation on the

outer 50 percent or so of the fuel pellet as a result of irradiation.
16

Figure 4-7 Pre (top) and post (bottom) LOCA transient fuel fragmentation for low-burnup

rods (2.5 GWd/MTU, left) and medium-burnup rods (35 GWd/MTU, right)

(Ref. 19)

Figure 4-8 shows an axial cut of the same nontransient-test fuel rods shown in Figure 4-7. 

These micrographs show evidence that the fuel pellets contain not only radial cracks but also

axial and circumferential cracks. The medium-burnup rod (35 GWd/MTU) appears to show

increased fragmentation near the centerline of the fuel pellet and also shows the development

of the rim structure as a result of irradiation. The staff expects that the rim structure is more

porous than the center of the pellet and thus may fragment more easily if the fission gases are

released rapidly during a LOCA transient.

2.5 GWd/MTU, not transient tested

C6

35 GWd/MTU, not transient tested

G1.6

35 GWd/MTU, transient tested

Local circumferential elongation 5%

G1.1

2.5 GWd/MTU, transient tested

Local circumferential elongation 8%

C2
17

Figure 4-8 Axial cross-section of nontransient-tested irradiated fuel rods showing the

developement of a rim structure at medium burnup (Ref. 19)

Figure 4-10 through Figure 4-13 show metallographic cross-sections of series A, C, E, F, and G

rods tested in the FR-2 program. This series of figures illustrates the increasing fragmentation

that occurs as rod burnup is increased from 0 to 35 GWd/MTU. These observations are also

supported by fuel particle size measurements performed on some rods from the FR-2 tests,

which show that the average particle size varies by roughly a factor of two between 0 and

35 GWd/MTU. These micrographs also show extensive radial fuel relocation, as well as some

axial relocation, as evidenced by the fact that the amount of fuel visible in the ballooned plane of

some of the rods (A1.1, A2.1, C1, C3, C4, C5, F2, F4, G2.2, and G3.2) amounts to more than

one pellet. Further evidence of extensive axial fuel relocation can be seen in neutron

radiographs from FR-2 test rods, an example of which is shown in Figure 4-14. Finally,

although fuel dispersal was not documented in the FR-2 program, there are many cases in 

which the fuel fragments were smaller than the rupture opening. Therefore, fuel dispersal may 

have occurred in the FR-2 tests, but this was not the object of investigation in this test program.

Figure 4-9 Cross-section of A rods from FR-2 in-pile test (unirradiated rods) (Ref. 20)
18

Figure 4-10 FR-2 in-pile tests: cross-sections of the test rods from series C

(2.5 GWd/MTU) (Ref. 19)

Figure 4-11 FR-2 in-pile tests: cross-sections of the test rods from series E

(8.0 GWd/MTU) (Ref. 19)

Cross sections of the 

test rods from series E

8 GWd/MTU burnup

Cross sections of the 

test rods from series C

2.5 GWd/MTU burnup
19

Figure 4-12 FR-2 in-pile tests: cross-sections of the test rods from series F

(20 GWd/MTU) (Ref. 20)

Figure 4-13 FR-2 in-pile tests: cross-sections of the test rods from series G2/G3

(35 GWd/MTU) (Ref. 20)

Cross sections of the 

test rods from series F

20 GWd/MTU burnup

Cross sections of the test 

rods from series G2/G3

35 GWd/MTU burnup
20

(a) (b)

Figure 4-14 (a) Neutron radiograph of rod F1 (20 GWd/MTU), comparing pre- and

posttransient fuel location (Ref. 19), (b) neutron radiograph of rod E5 

(8 GWd/MTU) after the transient (Ref. 19)

Axial relocation of fuel in ballooning or disrupted fuel rods shifts the position of heat generation

and changes the temperature distribution. INEL developed an empirical model, based on PBF

and FR-2 data, to calculate the transient axial relocation (Ref. 21). 

The amount of axial fuel relocation in 18 ballooned fuel rods was determined from neutron

radiographs, niobium gamma decay counts, and photomicrographs. The examined fuel rods

had burnups from 0 to 35 GWd/MTU and cladding hoop strains varying from 0 to 72 percent.

This experimental study reported the following results:

• No axial relocation of fuel occurs until the cladding hoop strain exceeds a value of

8 percent.

• After the cladding hoop strain exceeds 8 percent, the fuel pellets crumble.

• As the cladding continues to strain, axial fuel relocation occurs so that the fuel void

fraction in the balloon region remains equal to the void fraction at the time of fuel pellet

crumbling.

The model developed for axial fuel relocation reflects the experimental results, as shown in

Figure 4-15. In fuel rods ballooning until rod-to-rod contact occurs, the model predicts that axial

fuel relocation causes a 35-percent increase in the linear heat generation rate on the ballooned

region of the fuel rod.

Pre-transient Post transient

Plenum 

spring

Upper 

end of 

heated 

zone

Lower end of heated zone

Cladding

Pellet axial 

relocation 83 mm
21

Figure 4-15 Pellet stack reduction as a function of cladding hoop strain for preirradiated 

fuel rods (based on the increase in volume in cladding balloon with no axial 

elongation) (Ref. 21)

In summary, the staff draws the following conclusions from the FR-2 tests:

• Fuel fragmentation occurs early in fuel rod life and tends to increase with burnup.

• Extensive fuel relocation occurred in ballooned rods where the diametral strain 

exceeded 8 percent.

• Fuel dispersal may have occurred when fuel fragments were smaller than the rod 

rupture opening on rods that had ruptured.

4.1.3 National Research Universal—Pacific Northwest Laboratory

The research program conducted by PNL at the NRU reactor in Canada was the first series of

tests performed on a bundle of rods instead of a single rod (Refs. 22, 23, 24, 25, and 26). The

objective of these tests was to study the thermal-hydraulic and thermo-mechanical behavior

(deformation, flow blockage) within a bundle of 32 full-length PWR rods. The test rig is shown in

Figure 4-16; outer rods formed a guard ring for the 11 or 12 inner pressurized test rods. The

test bundle was surrounded by a stainless-steel shroud to protect the loop pressure tube. The

test rods were fueled with fresh UO2 pellets but were preconditioned before the LOCA transients

by power cycling to full power, which caused the fuel pellets to fracture.

The four thermo-mechanical tests were labeled MT-1 through MT-4. The test sequence

consisted of a hold at 375 degrees C in steam, followed by cutting off steam to generate the

LOCA transient, then quench and scram. The test rods all went through a ballooning phase,
22

and 6 of 11, 8 of 11, 12 of 12, and 8 of 11 rods ruptured in tests MT-1, MT-2, MT-3, and MT-4,

respectively. Two important findings of the Material Test (MT) series were that (1) the

ballooning observed in the bundles of rods occurred in the same axial regions of the bundle for

all rods in the bundle, thus resulting in flow blockage ratios up to about 70 percent, and that (2)

grid spacers act to mostly prevent ballooning in the short section of the fuel rod that traverses

them, thus “pinning” the balloons and potentially acting as choke points for fuel relocation. 

Figure 4-17 shows this phenomenon with a double balloon observed in all four tests, with the

ballooned regions being pinned by the grid spacers.

Figure 4-16 NRU-MT cross-section of the test assembly for the LOCA simulation

program (Ref. 22)

Figure 4-18 through Figure 4-21 show pictures of the ballooned and ruptured regions for tests 

MT-1 and MT-2, and of individual ballooned and ruptured rods for MT-3 and MT-4. These 

pictures show that fuel fragmentation appears to have occurred in tests MT-2, MT-3, and MT-4, 

although to varying degrees in different rods. In fact, MT-3 rods 3C and 3D (adjacent rods) 

seem to have very different fuel pellet fragment sizes: almost-intact pellets for rod 3C and 

fragmented pellets for rod 3D. Tests MT-2 and MT-4 show extensively fragmented pellets and 

large rupture openings. It is possible that fuel dispersal occurred in test MT-2, but it is not 

documented. For test MT-4, the rupture region of some of the rods shown (4E, 5C, 5D) is 

empty, likely indicating that fuel dispersal through the rupture opening occurred.
23

Figure 4-17 Circumferential strain profiles for all rods in tests MT-1, MT-2, MT-3, and

MT-4, showing coplanar ballooning pinned by grid spacers (Ref. 26) (X-axis: 

strain in percent (%), Y-axis: rod length in inches)

Figure 4-18 Ballooned area in tests MT-1 (left; ballooned and ruptured rods, with mostly

intact pellets remaining in the rod) and MT-2 (right; ballooned and ruptured

rods with some large fuel pellet fragments remaining in the rod) (Refs. 22

and 23)

100

90

80

70

60

50

40

30

20

10

0

100

90

80

70

60

50

40

30

20

10

0

100

90

80

70

60

50

40

30

20

10

0

100

90

80

70

60

50

40

30

20

10

0

0 15 30 45 60 75 90 105 120 135 150

0 15 30 45 60 75 90 105 120 135 150

0 15 30 45 60 75 90 105 120 135 150

0 15 30 45 60 75 90 105 120 135 150
24

Figure 4-19 Ballooned and ruptured rod 3C in test MT-3: minimal fuel pellet

fragmentation, pellets remaining in rod (Ref. 24)

Figure 4-20 Ballooned and ruptured rod 3D in test MT-3: some large fuel pellet

fragments remaining in the rupture node (Ref. 24)
25

Figure 4-21 Ballooned and ruptured rods in test MT-4 showing rupture node devoid of

fuel (Ref. 25)

In summary, the staff draws the following conclusions from the PNL/NRU tests:

• In a bundle of rods, ballooning can occur such that all of the balloons are coplanar.

• Ballooning is mostly prevented in the sections of fuel rods that cross a grid spacer.

• Grid spacers appear to “pin” rod ballooning, potentially acting as choke points for fuel 

relocation.

• Under certain conditions, fuel dispersal due to ballooning and rupture of a fuel rod can 

result in regions of the rod that are devoid of fuel, indicating relatively large amounts of 

fuel dispersal.
26

4.1.4 PHEBUS-LOCA—Institut de Radioprotection et de Sûreté Nucléaire

(France)

The PHEBUS-LOCA program performed by Institut de Radioprotection et de Sûreté Nucléaire

(IRSN) consists of an in-pile LOCA program with integral tests on Zircaloy-4-clad PWR fuel rods

in a five-by-five bundle configuration (Refs. 27. 28, 29, and 30). Starting from an initial state

representative of PWR nominal conditions, the test transient aimed at reproducing a

hypothetical large-break LOCA transient from initial blowdown until the final quench of the test

device. IRSN conducted the PHEBUS-LOCA experimental program from 1980 to 1984. 

The test train consisted of a bundle of 25 five-by-five PWR-type rods containing fresh UO2 fuel,

maintained by four Inconel spacer grids. The fuel pins were 1 meter (m) long (0.8 m active

length) and could be internally pressurized. The fuel bundle was surrounded by a Zircaloy-2

shroud to link the square section of the bundle to the circular section of the concentric outer

structures. 

The test device was inserted in a loop in the PHEBUS driver core that allowed for nuclear

heating of the test rods. The test loop was designed to reproduce the initial steady-state

conditions of a PWR before transient initiation. The transient was initiated by isolating the test

section and rapidly opening valves on the upstream and downstream pipes to simulate breaks

on the cold and hot legs. Simultaneously, the power of the driver core was decreased to

simulate the nuclear power transient following plant shutdown.

The area of the breaks could be adjusted to simulate different transient histories. Four injection

lines, upstream and downstream from the test train, were used to refill the loop and the reflood

of the bundle after the blowdown phase.

Figure 4-22 and Figure 4-23 show cross-sections of test bundles 215-P and 219 at various

elevations. Although the PHEBUS program did not investigate fuel fragmentation, relocation,

and dispersal, it can be seen that fragmentation occurred in many of the fuel pellets that can be

observed in these cross-sections. In these tests, the large amounts of ballooning resulted in

radial fuel relocation but no apparent axial fuel relocation. It should be noted that the fact that

fresh fuel pellets were used in the PHEBUS tests caused the fuel fragments to be relatively

large (pellets broken in only a small number of chunks), which would make axial fuel relocation

more difficult.
27

Figure 4-22 PHEBUS test 215-P metallographic cross-sections at elevations 285 mm and

252 mm from bottom of fuel stack (Ref. 27)

Figure 4-23 PHEBUS test 219 metallographic cross-section at elevation 437 mm from

bottom of fuel stack showing radial displacement of rods (Ref. 30)

In summary, the staff concludes from the PHEBUS-LOCA tests that axial relocation of 

very-low-burnup fuel that shows minimal fragmentation is unlikely, even in the event of relatively 

large cladding diametral strains. Radial fuel relocation may nonetheless occur under these 

conditions.

4.1.5 FLASH Tests—Commissariat à l’Energie Atomique (France)

Five tests were carried out in the FLASH facility of the SILOE reactor at Commissariat à 

l’Energie Atomique (CEA)/Grenoble (Refs. 31 and 32). The main purpose of the tests was to

study the release of fission products during a LOCA transient. These tests were performed on a

preconditioned fresh fuel rod (FLASH-1 to FLASH-4) and a high-burnup fuel rod (FLASH-5). 

The PWR 17-by-17-type test rods with an active length of 30 cm were centered in an unheated

Elevation ~ 285 mm: blockage ratio ~ 36 % Elevation ~ 252 mm: blockage ratio ~ 48 %
28

shroud tube located on the reactor periphery, which induced large azimuthal temperature

differences.

For tests FLASH-1 to FLASH-4, a preirradiation phase at nominal PWR conditions (35 to

40 kilowatts per meter (kW/m), 13 MPa) made it possible to produce a fission product inventory

at a burnup ranging from 1.65 to 3.32 GWd/MTU. For the FLASH-5 test, the test rod was

refabricated from a high-burnup PWR rod (50.3 GWd/MTU) and was reirradiated for a few

weeks in the SILOE reactor at about 17 kW/m in order to re-form short-lived fission product

species, thus adding 1.412 GWd/MTU to the initial burnup.

The experimental transient began with a linear heat generation rate of 7 kW/m. The test train

was then depressurized between 0.5 and 2.1 MPa, in conjunction with the injection of helium. 

This resulted in a peak cladding temperature of 1,100 degrees C in FLASH-1 and FLASH-2,

1,270 degrees C in FLASH-3 and FLASH-4, and 1,350 degrees C in FLASH-5. The tests were

terminated by reactor scram (FLASH-1 to 3) or quenching at hot conditions while maintaining

the nuclear power for about 10 minutes (FLASH-4 and FLASH-5).

Rupture strain ranged from 16 percent in FLASH-5 to 62 percent in FLASH-2, the low strain

values in FLASH-5 being explained by the occurrence of large azimuthal temperature

differences exceeding 110 kelvin (K) at time of rupture. There was no observed fuel relocation

during the transient in tests with preconditioned fresh fuel. However, there were indications of

some fuel relocation following handling when removing the rod from the test device. In test

FLASH-5 on an irradiated rod, fine fuel fragmentation in the central part near the maximum flux

level was observed, as well as significant displacement of fuel fragments in the rupture plane, in

spite of the low clad strain (see Figure 4-24).

Figure 4-24 Fuel aspect at rupture plane in FLASH-5 test (Ref. 32)

The staff concludes from the results of the FLASH tests that axial fuel relocation of 

very-low-burnup preconditioned fuel is unlikely even for large cladding diametral strains. 

However, axial and radial fuel relocation can occur for diametral strains as low as 16 percent

and burnups as low as 35 GWd/MTU, even with very small rupture openings that mostly prevent 

fuel dispersal.
29

4.1.6 Argonne National Laboratory LOCA Test Program

Four single-rod LOCA tests were performed at ANL using high-burnup BWR rods

(56 GWd/MTU) (Ref. 33). These tests used external heating of the rods in a steam

environment. The temperature was ramped to 1,200 degrees C, held for a predetermined

period of time, and ramped down to 800 degrees C, at which point quench occurred.

The ANL program focused on the details of cladding oxidation, hydriding, and ductility, rather

than on the fuel behavior, and methods that could be used to freeze the fuel particles in place

(e.g., epoxy) conflict with cladding characterization. Therefore, for tests such as ICL#2, no

attempt was made to prevent fuel fallout during handling. 

(a) (b)

(c) (d)

Figure 4-25 Low-magnification images of the post-LOCA test ICL#2 fuel samples (a) at

≈12 mm above the rupture center (15–35% strain), (b) at ≈50 mm above the

rupture center (2–4% strain), (c) at ≈130 mm below the rupture center, and

(d) prior to LOCA testing (180 mm from the LOCA sample) (Ref. 33)

Figure 4-25 shows low-magnification images of the fuel structure of the ICL#2 sample at axial 

locations: (a) about 12 millimeters (mm) above the rupture center, (b) about 50 mm above the 

rupture, and (c) about 130 mm below the rupture center (45 mm above the bottom end-cap). 

Also shown in (d) is the fuel structure of the as-received fuel. The structures of Figure 4-25(c) 

and (d) are similar, except that the post-LOCA fuel shows a ring of circumferential tearing about 

mid-radius. This tearing may have occurred as the cladding tried to move a small distance 

(0.1 mm) away from the fuel and/or because the fission-product gases affected the fuel (see the 

dark ring near mid-radius for the pre-LOCA fuel in Figure 4-25(d)). At about 50 mm above the 
30

rupture, the circumferential tearing is enhanced as compared to the about 130-mm location, 

most likely due to the larger cladding strain. Some fuel fallout may have occurred during 

cutting, although this region of the fuel column was embedded in a soft epoxy prior to cutting. 

Smaller fuel particles are also observed. In Figure 4-25(a), a wide range of fuel particles is 

observed, although these particles and fuel chunks are not coplanar. The particles and chunks 

are held in place by soft epoxy. Because this photograph was taken after extensive handling of 

the sample, resulting in axial redistribution of particles and fuel fallout through the rupture 

opening, it does not represent the fuel condition near the rupture center during the LOCA test or 

after the quench. The most that one can glean from such a picture is that the wide range of 

fuel-particle sizes would allow some fuel to fall from less than 50 mm above the rupture center 

to the rupture region. In addition, it can be said that the fuel fragmentation observed in these 

tests allowed for easy and significant fuel relocation within the rod and for fuel dispersal as a 

result of handling. Finally, although the ANL LOCA tests did not measure the amount of fuel 

dispersal, it was reported that a fine dust of fuel particles was expelled from the test rods upon 

rupture during the ramp to 1,200 degrees C.

The staff concludes from the ANL tests that fine fuel fragmentation is likely to facilitate fuel axial 

relocation, as well as fuel dispersal in the form of fine particles that can easily migrate to the 

rupture opening and escape into the coolant.

4.1.7 Halden Boiling-Water Reactor LOCA Test Series

The Organization for Economic Co-operation and Development (OECD) Halden Reactor Project

is a series of in-pile experiments performed in the Halden Boiling-Water Reactor (HBWR), a

heavy-water-cooled and moderated reactor. The project is sponsored by about

20 organizations from different countries and is operated by the Norwegian Institute for Energy 

Technologies. The experiments in the Halden reactor are assigned instrumented fuel assembly

(IFA) numbers, but IFA-numbered experiments are not limited to fuel assemblies and can also

be structural materials experiments.

Two LOCA programs were run in the Halden reactor in the early 1980s: IFA-511.X (Ref. 34)

and IFA-54X (Ref. 35). These programs mostly aimed to study thermal-hydraulic and thermomechanical phenomena, respectively. The staff’s review did not find information about the 

phenomena of fragmentation, relocation, or dispersal of fuel; therefore, this report will not 

describe these programs in detail.

The currently (2003–present) ongoing Halden experiments numbered IFA-650.X are single-pin

tests and focus on effects that are different from those studied in out-of-reactor tests. A

prototypical bounding LOCA transient does not exist, and Halden project participants

recommended that the test conditions be selected to meet the following primary objectives:

• Maximize the balloon size to promote fuel relocation and to evaluate its possible effect

on cladding temperature and oxidation.

• Investigate the extent (if any) of “secondary transient hydriding at high temperature” on

the inner side of the cladding around the rupture region in the presence of a 

pellet-cladding bonding layer.

Target PCTs for the preirradiated rods have been set at 800 degrees C and 1,100 degrees C for

high and medium burnups. The new Halden experiments were deemed necessary because

industry trends to high-burnup fuel design and introduction of new cladding materials have
31

generated a need to reexamine and verify the validity of the safety criteria for LOCAs. 

High-burnup fuel rods irradiated in commercial reactors have been used in this series of tests.

To date, 11 LOCA tests have been conducted in the HBWR (Refs. 36, 37, 38, 39, 40, 41, 42,

43, 44, 45, 46, 47, 48, 49, and 50). These single-rod, in-pile tests are heavily instrumented and

the subject of extensive PIE after the test. Of the 11 LOCA tests conducted, two showed major

fuel relocation and dispersal: IFA-650.4 and IFA-650.9.

The first series of LOCA test runs of the IFA-650 test series (Refs. 36 and 37) were performed

in the Halden reactor using a fresh unpressurized and small-gap PWR rod with low-tin zircaloy-4

cladding. A total of six test runs were performed. The target PCTs of about 800 and about 

1,100 degrees C were achieved.

The second trial LOCA test in IFA-650.2 was carried out in May 2004. Before the LOCA test,

two power ramps were performed to cause decay-heat and fission-product buildup. The fresh

fuel rod was irradiated for 1.5 days to accumulate fission products. The test consisted of a

blowdown phase, heatup phase, hold at PCT, and termination by a reactor scram. The target

for peak cladding temperature was 1,050 degrees C. In order to achieve this target, the heat

rates of the fuel rod and heater were adjusted to 22 and 17 W/cm, respectively. These

parameters were chosen on the basis of earlier experience obtained during the first trial tests in

IFA-650.1. An extensive postirradiation examination of the IFA-650.2 test was performed and

presented at a LOCA workshop meeting in 2005 (Ref. 38). Figure 4-26 shows the balloon with

relocated fuel pellets. Although it is thought that the relocation mainly occurred due to transport

and handling (because the in-pile instrumentation did not reveal any clear indications of fuel

relocation), there was clear observation of pellet fragmentation into large fragments. This

indicates that fragmentation occurred at a very low burnup. The actual rupture was a totally

ductile failure. However, a more brittle crack through the cladding was observed close to the 

rupture location, as shown in Figure 4-27. Such cracks could be the result of hydriding, but

hydrogen analyses of the rod did not show more than about 100 weight parts per million 

(wt ppm) hydrogen near the balloon. 

Figure 4-26 Ballooned region with relocated fuel in the Halden IFA-650.2 test (Ref. 61)
32

Figure 4-27 Through-wall cracks with brittle appearance, Halden IFA-650.2 test (Ref. 61)

Test IFA-650.3, the first test with preirradiated fuel in the Halden project LOCA test series, was 

conducted on April 30, 2005 (Ref. 39). The fuel was provided by Framatome Advanced Nuclear 

Power (ANP) and had been irradiated in a commercial PWR to a high burnup, 82 GWd/MTU. 

The experimental arrangements of the third test were similar to the preceding LOCA tests. The 

target temperature (800 degrees C) was achieved, and a rod rupture occurred at around 

780 degrees C. The results from the PIE indicate that no large ballooning occurred and that the

cladding failure occurred close to the lower thermocouple weld. However, small local ballooning

had also started in the middle of the rod. The rod experienced some uniform cladding

deformation (7 percent). Bowing was also observed in the rod. However, no fuel relocation

could be detected in-pile, and the PIE results also indicate that no fuel relocation occurred.

Test IFA-650.4, a repeat of the first test with preirradiated fuel (IFA-650.3), was conducted on

April 25, 2006 (Ref. 40). The fuel was provided by Framatome ANP and had been irradiated in

a commercial PWR to a high burnup, 92.3 GWd/MTU, which resulted in about 40 percent of the

fuel exhibiting a rim structure. The experimental arrangements of the fourth test were similar to

the previous LOCA tests, especially to IFA-650.3. The target temperature (800 degrees C) was

achieved, and a cladding rupture with fuel relocation occurred. The results from the posttest

gamma scanning indicate that large ballooning in the middle of the rod and rupture with fuel

relocation occurred. The ballooning, rupture, and fuel relocation were detected in-pile and were

verified by the gamma scanning performed at Halden. Figure 4-28 shows the gamma scan of

fuel rod segment IFA-650.4 that had been subjected to LOCA conditions in the HBWR. The

entire upper portion (19 cm) of the fuel above the rupture opening was lost during rod

depressurization. Similar out-of-pile tests have been run at ANL on fuel that had a burnup of

about 57 GWd/MTU, with the observed loss of fuel amounting to less than that in one fuel pellet. 

Earlier LOCA tests at the Karlsruhe Nuclear Research Center on fuel with burnups no greater

than 35 GWd/MTU did not report any fuel loss, although it may have occurred but was not

reported because it was not the focus of those tests. Thus, very-high-burnup fuel with a welldeveloped rim structure may be susceptible to significant loss of fuel particles during a LOCA

transient.
33

It should be noted that the Halden tests IFA-650.3, IFA-650.4, and IFA-650.9 (discussed 

hereafter) had burnups far beyond the current burnup limit of 62.5 GWd/MTU rod average in the 

U.S. There is some evidence that higher burnup may aggravate the phenomena of fuel 

fragmentation and dispersal, thus it is expected that the extensive fuel dispersal observed in 

IFA-650.4 and IFA-650.9 would not be as severe for fuel with a burnup below the current U.S. 

limit of 62.5 GWd/MTU rod average.

Figure 4-28 Gamma scan of a very-high-burnup (approximately 92 GWd/MTU) fuel rod

showing major loss of fuel material after LOCA testing (IFA-650.4)

Postirradiation examination of three of the most recent Halden LOCA tests provides further

evidence of how fuel fragmentation, relocation, and dispersal occur (Ref. 47).

Test IFA-650.9 (Refs. 45 and 46) was carried out using low fission power to achieve the desired

conditions for high temperature, ballooning, and oxidation. The target peak cladding
34

temperature was 1,100 degrees C, and cladding failure occurred about 133 seconds after

blowdown at about 810 degrees C. As a consequence of the cladding rupture in IFA-650.9, fuel

from the upper part of the fuel stack dropped into the ballooned region. Figure 4-29 shows the

resulting images from the gamma scans. As can be seen in this figure, about 12–13 cm of the

original pellet stack was missing at the upper part of the rod, but at the top end of the rod, about 

25 mm of the fuel (which corresponds to two fuel pellets) remained. The relocated fuel had

dropped to the about-6-cm-long ballooned area at the midheight of the rod, where the diameter

had increased by almost twice the diameter of the original rod (i.e., the diameter increased to

about 18 mm). In this ballooned area, the Cs-137 and the ruthenium (Ru)-103 count rates were

respectively 30–70 percent and 20–30 percent higher than the general level of the rod. The fuel

rod midsection also appeared wider, by 1–3 mm, than at the top and bottom and showed a

Cs-137 count rate 30–70 percent higher than the general level (on the other hand, the Ru-103 is

similar to the general level). In addition, some fuel had fallen through the cladding opening,

confirming that fuel dispersal had occurred. Figure 4-30 shows dispersed fuel from the 

IFA-650.9 test.

The high-burnup (90 GWd/MTU) fuel shows a high-burnup rim structure with micrometer-size

pores and submicrometer grains. This rim structure was found over approximately one third of

the fuel radius. The fragmentation of the rim structure shows the fuel broken into both larger

and very small, micrometer-size particles as shown in Figure 4-31. The remaining two thirds of

the fuel radius seems to have a more robust grain structure, but fuel relocation and release of

fuel particles during this test was very extensive, and larger fuel particles (the inner two thirds of

the fuel diameter) relocated to the balloon and out the rupture opening during the test. From 

this, one may conclude that fuel fragmentation and relocation is not limited to the rim region.
35

Figure 4-29 Gammas scans (Cs-137 and Ru-103 activity) of IFA-650.9 about 6 weeks

after the test. Left: full scan at 0 degrees. Right: flask rotated 90 degrees

counterclockwise. Resolution: 5 mm in vertical and 1 mm in horizontal

direction. The rig drawing in the middle shows the instrument levels.

(Ref. 45)
36

Figure 4-30 Fuel residue from posttest disassembly of IFA-650.9 (Ref. 46)

Figure 4-31 Fragmentation of the rim structure in IFA-650.9 (Ref. 46)

Test IFA-650.10 was conducted in May 2010 (Refs. 48 and 49). The parent fuel rod, which was

provided by Electricité de France (EdF), was irradiated in a commercial PWR and had a burnup

of 61 GWd/MTU. The experimental conditions were similar to the fourth LOCA test in this

series. The target peak cladding temperature was rather low (800–850 degrees C), but

cladding rupture did occur. The rupture opening was rather small (15 mm by 5 mm) and fuel
37

fragments can be seen through the rupture opening, as shown in Figure 4-32. Fuel

fragmentation consisted of both large and small fragments, which became mobile with cladding

diametral expansion (see Figure 4-33).

Figure 4-32 The rupture opening in IFA-650.10 (Ref. 49)

Figure 4-33 Fuel fragmentation in IFA-650.10 (Ref. 49)
38

To varying degrees, the process of fuel fragmentation can be seen all along the test rod. In

Figure 4-34, a neutron radiograph shows compact, unfragmented fuel pellets away from the

rupture location but fragmented fuel present in areas where diametral cladding strain is

significant. It should be noted that the last four pellets in the fuel column (lower frame of the

image) have a central hole. Pellets of this design were used in the subsequent test, IFA-650.11. 

The point at which the diametral strain allowed fuel pellet fragmentation to occur was found to

be 13 to 15 percent, as shown in Figure 4-35.

Figure 4-34 Neutron radiography of IFA-650.10 (Ref. 49)
39

Figure 4-35 Fuel fragmentation and relocation as a function of cladding diameter

increase in test IFA-650.10 (Ref. 49)

Test IFA-650.11 (Ref. 50) was similar to its predecessor (IFA-650.10), but the burnup of the

parent fuel rod was lower (56 GWd/MTU) and the target PCT was higher (1,000 degrees C). 

Postirradiation examination, including gamma scanning, visual inspection, and neutron

radiography, was performed to discover the state of the rod following the test. Gamma

scanning and visual inspection confirmed cladding deformation along the lower half of the rod,

namely, an increase of the cladding diameter. The rupture failure was a 5-mm short split and a

1-mm narrow opening slightly below midrod location. Neutron radiography showed intact fuel

pellets in the upper part of the rod and the last two pellets in the bottom end of the active fuel

stack. Two minor fuel gaps in the upper part of the rod were found; however, the gamma scan

revealed only one gap.

Fuel fragmentation with fuel fragments in the mm range was found from midrod toward the

bottom end of the active fuel stack. The zone with fuel fragmentation and the zone with

measured cladding diameter increase coincided (see Figure 4-36). The threshold for fuel

fragmentation and relocation was a 17-percent diameter increase, as shown in Figure 4-37. 

This number is consistent with other LOCA tests in the Halden IFA-650 series.

Diameter from radiograph

Diameter from visual

Fragmentation and relocation

Burst

Radial and axial cracks

Radial cracks

Fragmentation and relocation
40

Figure 4-36 Neutron radiography of test IFA-650.11 (Ref. 50)

Figure 4-37 Fuel fragmentation and relocation as a function of cladding diameter

increase in test IFA-650.11 (Ref. 50)
41

In summary, the NRC staff draws the following conclusions from the Halden LOCA tests:

• Fuel fragmentation is consistently noted in the Halden IFA-650 series of LOCA tests.

• The degree of fragmentation (the resulting particle size) is not uniform. It appears to be

more pronounced in some regions of the fuel (e.g., the rim region) than others, and, in

those cases, both large and small particles may result.

• Restraint on the fuel, resulting from a tight fit between fuel and cladding

(i.e., undeformed cladding) appears sufficient to make cracked pellets appear intact—at

least when examined through neutron radiography.

• Visually apparent fragmentation and the subsequent relocation of the fuel particles occur

when the cladding outer diameter is deformed by the ballooning process. The threshold

for this effect is 13- to 17-percent diametral strain.

• A small rupture opening may be sufficient to prevent significant fuel dispersal, as shown

in tests IFA-650.10 and IFA-650.11, but smaller fuel particles may still escape.

• The lack of cladding restraint and the presence of a large rupture opening are sufficient

to result in extensive dispersal of fuel pellet fragments—particularly in high-burnup fuel

as shown in tests IFA-650.4 and IFA-650.9.

• Fuel fragmentation and relocation is not limited to the rim region.

The Committee on the Safety of Nuclear Installations (CSNI) asked the OECD Nuclear Energy 

Agency (NEA) Working Group on Fuel Safety to evaluate how the Halden LOCA tests affect 

regulation. The summary report of the working group’s evaluation provides valuable discussion 

and conclusions about the observations of fuel loss in the Halden results (Ref. 51).

4.1.8 Studsvik LOCA Test Program

The LOCA program at Studsvik recently completed six single-rod integral LOCA tests using the

same overall procedures as the ANL LOCA program described in Section 4.1.6 above. The first 

four tests in the NRC LOCA program at Studsvik were performed on rods with a burnup around 

72 GWd/MTU (tests 189, 191, 192, and 193), while the last two were performed on rods with a 

burnup around 55 GWd/MTU (tests 196 and 198). In each of these tests, a pressurized, highburnup, fueled rod segment was subjected to a temperature transient in a steam environment

through ballooning and rupture. Tests 189 and 196 were terminated just after rupture while the 

four other tests each experienced some degree of high-temperature oxidation, followed by

quench. The total voided length was measured for each segment using a wire probe. Following 

the LOCA simulation, four-point bend tests (4PBTs) were conducted to measure the residual 

mechanical behaviour of the ballooned and ruptured region. After the 4PBT, a “shake” test was 

performed to determine the mobility of fuel particles that remained in the fuel rod. The “shake” 

test consisted of an inversion of the two halves of the broken fuel rod followed by minor shaking 

to dislodge any loose fuel particles. This test was conducted approximately two days after the 

LOCA simulation.
42

While the experimental procedures for each test remained largely the same, there were 

important differences between the test segments, the observed burst response, fragmentation 

size and fuel dispersal. Table 4-1 below consolidates some of the characteristics of each test 

and the remaining subsections will expand on the observations of fuel fragmentation, mobility 

and loss.

In all four of the 72 GWd/MTU tests, significant fuel loss through the rupture opening was 

observed during the LOCA transient. Figure 4-38 shows the rupture opening after LOCA testing 

for the first four specimens tested, with burnup around 72 GWd/MTU. It can be seen that the 

rupture is completely devoid of fuel, indicating large relocation and dispersal of fuel. Figure 4-39

shows the rupture opening after LOCA simulation for tests 196 and 198. In these two tests, the 

rupture opening was significantly smaller than that in tests 189 through 193.

In contrast, during tests 196 and 198 at 55 GWd/MTU, essentially no fuel was found outside of 

the fuel rod following the LOCA simulation, and all fuel found outside of the fuel rod was 

measured after the broken rods were “shaken.”

Fuel fragment size measurements were completed by processing the total mass of fuel material 

found outside of the fuel rod after all steps of the experimental procedures for each of the six 

tests. The fragments were processed through a series of six sieves to determine the distribution 

of the fuel particles by size. Figure 4-40 depicts the results of this examination. One of the first 

observations that are apparent is a significant difference in distribution between tests 191 to 193 

and tests 196 and 198. In tests 191-193, all fuel fragments measured less than four millimeters 

(except a small amount from test 193). The mass of fuel fragments from these tests was 

approximately evenly distributed between the size groups separated by the six sieves used. In 

contrast, the fuel fragments in tests 196 and 198 were predominately larger than 4 millimeters.

Although it is not obvious by examination of Figure 4-40, Table 4-1 indicates that the total mass 

found outside of the fuel rod for tests 191-193 was about the same that measured in tests 196 

and 198, so the difference in size distribution is not due to a skewing of limited sample size. 
43

Table 4-1 Measurements of fuel loss for each test conducted to date
44

Figure 4-38 Rupture opening in Studsvik LOCA tests 189, 191, 192, and 193 (left to

right), showing the absence of fuel in the rupture plane

Figure 4-39 Rupture opening in Studsvik LOCA tests 196 and 198 (left to right)
45

Figure 4-40 Particle size distribution from six integral LOCA tests

To further illustrate the difference in the fragmentation of fuel observed in tests 191-192 and 

tests 196 and 198, Figure 4-41 and Figure 4-42 provide images of fuel particles collected from 

four tests. Figure 4-41 includes two images of fuel particles from tests 192 and 193 and the 

images reveal the sand-like quality of the fuel fragments collected from these tests. In contrast, 

Figure 4-42 includes two images of fuel particles from tests 196 and 198 and the images reveal 

much larger fuel fragments. In some cases the fragments appear to be as large as half a pellet.

(a) (b)

Figure 4-41 Images of fuel particles collected from test rod (a) 192 and (b) 193 revealing 

a very small, sand-like fragmentation size
46

(a) (b)

Figure 4-42 Images of fuel particles collected from test rod (a) 196 and (b) 198 revealing 

large fragments

It is important to note that most of the fuel loss in the tests at 72 GWd/MTU was noticed just 

after the LOCA transient and the collected fuel fragments were very fine and small. In addition, 

it is important to note that when the fuel rod was first removed from the test train, leaving a small 

hole at the bottom of the train, a small amount of wet, black fuel sludge fell out. An image taken 

just after segment 191 was removed can be seen in Figure 4-43(a). After two days, a far larger 

amount of fuel was found beneath the test train, even though no disturbance or activity took 

place with the test train. An image taken two days after segment 191 was removed can be seen 

in Figure 4-43(b).

(a) (b)

Figure 4-43 Fuel collected beneath the LOCA test train (a) just after the fuel rod is

removed, and (b) 2 days after the fuel rod is removed

Based on this observation, the staff assumes that as the fuel dries out, it becomes more mobile

and gradually empties out of the test train after it loses its adherence to the test train surfaces. 

An observation from the “shaking” test, is that the fuel that fell out during the shaking step

flowed out readily as soon as the rod was inverted and before any shaking. A video of this step

during this test captures this quite well. In the video, the shaking does not appear to dislodge
47

any almost-mobile fuel fragments, but rather the inversion allows already mobile fuel to just spill

out.

The magnitude of fuel mobility was examined by measuring the mass of lost fuel, as well as by 

a wire probe measurement indicating the length of “empty” cladding material. The length of 

“empty” cladding material was compared to the measured final strain in each test. Figure 4-44

through Figure 4-49 are an illustration of this comparison. In all figures, a region of the graph is

shaded purple to indicate the final measurement of the axial extent of fuel loss and the 

intersection of this length with the values of local, final strain are noted. The values range from 

1% to 9% strain. 

For tests in which the wire probe was used to measure the length of “empty” cladding just after 

the LOCA simulation, a blue region indicates the axial extent of fuel loss just after the LOCA 

simulation. In tests 191-193, there was significant fuel loss during the LOCA simulation and 

further fuel loss took place during the “shaking” test. There was no significant agitation of the 

fuel rod between the LOCA simulation and the “shaking” tests, so the additional fuel lost could 

be a result of the dry out phenomenon discussed above.

It should be pointed out that the effect of gravity was observed in tests 192 and 193, where the 

voided volume above the burst region was larger than that below the burst region. No 

measurements were performed right after the LOCA test for test 189, and no clear effect of 

gravity was observed in test 191. In the figures below, fuel dispersal in the volume below the 

burst region only occurred after the LOCA- tested fuel rods were broken in two pieces in a 

subsequent mechanical test, and the fuel was shaken out of the ends of the fuel rods.

Figure 4-44 Axial extent of fuel loss in comparison to measured final strain in test 189

0%

1%

2%

3%

4%

5%

6%

7%

8%

9%

10%

0 50 100 150 200 250 300 350

Circumferential Strain (%)

Test 189 - Axial Distance from end cap

0

90

burst width

fuel loss

Fuel Loss 

Region

Strain ≈ 3%

Strain ≈ 6%
48

Figure 4-45 Axial extent of fuel loss in comparison to measured final strain in test 191

Figure 4-46 Axial extent of fuel loss in comparison to measured final strain in test 192

Figure 4-47 Axial extent of fuel loss in comparison to measured final strain in test 193

0%

1%

2%

3%

4%

5%

6%

7%

8%

9%

10%

0 50 100 150 200 250 300 350

Circumferential Strain (%)

Test 191 - Axial Distance from end cap

0

90

burst width

LOCA fuel loss

Final fuel loss Strain ≈ 4%

Strain ≈ 6%

0%

1%

2%

3%

4%

5%

6%

7%

8%

9%

10%

0 50 100 150 200 250 300 350

Circumferential Strain (%)

Test 192 - Axial Distance from end cap

0

90

burst width

LOCA fuel loss

Final fuel loss Strain ≈ 4% Strain ≈ 5%

0%

1%

2%

3%

4%

5%

6%

7%

8%

9%

10%

0 50 100 150 200 250 300 350

Circumferential Strain (%)

Test 193 - Axial Distance from end cap

0

90

burst width

Final fuel loss

LOCA fuel loss Strain ≈ 4%

Strain ≈ 1%
49

Figure 4-48 Axial extent of fuel loss in comparison to measured final strain in test 196. 

No fuel loss was observed during the LOCA simulation for this test.

Figure 4-49 Axial extent of fuel loss in comparison to measured final strain in test 198. 

No fuel loss was observed during the LOCA simulation for this test.

It should be noted here that Studsvik Nuclear AB proposed several hypotheses regarding fuel 

fragmentation and dispersal to attempt to explain the differences between the 72 GWd/MTU 

tests and the 55 GWd/MTU tests. None of these hypotheses has been investigated to date. 

The first hypothesis is that the extent of fuel fragmentation prior to the LOCA tests may have 

been widely different between the two levels of burnup. Second, it has been postulated that the 

driving mechanism for fuel fragmentation is a tensile stress on the fuel pellet that is induced by 

the cladding strain when a strong fuel-cladding bond exists. Finally, another hypothesis is that 

the heat up and expansion of the fission gas bubbles in the fuel pellet causes fuel fragmentation 

during the LOCA simulation.

In summary, the staff draws the following conclusions from the Studsvik LOCA tests:

• Fuel dispersal is only possible if the rupture opening is larger than the fragmented fuel 

particles inside the fuel rod.

• Very fine fuel fragmentation can occur at high burnups around 70 GWd/MTU, such that a 

significant portion of the fuel resembles a fine powder. The extent to which the LOCA 

transient aggravated the fuel fragmentation is not known.

0%

1%

2%

3%

4%

5%

6%

7%

8%

9%

10%

0 50 100 150 200 250 300 350

Circumferential Strain (%)

Test 196 - Axial Distance from end cap

0

90

burst width

LOCA fuel loss

Final fuel loss

Strain ≈ 5%

Strain ≈ 3%

0%

1%

2%

3%

4%

5%

6%

7%

8%

9%

10%

0 50 100 150 200 250 300 350

Circumferential Strain (%)

Test 198 - Axial Distance from end cap

0

90

burst width

LOCA fuel loss

Final fuel loss

Strain ≈ 9%

Strain ≈ 4.5%
50

• Under conditions of extensive fragmentation of fuel, relocation was observed for 

diametral strains as low as 5 percent. Fuel relocation was generally observed for 

diametral stains in the 4–12 percent range.

• Fuel fragment mobility is enhanced as the fuel dries out (in the absence of steam or 

water).
51

4.2 Integral LOCA Test Data

4.2.1 Data Tables

The tables shown here summarize the data extracted from the reports describing past LOCA

testing performed with fueled rod segments. Although not always documented in past LOCA

research programs, information on fragmentation, relocation, and dispersal of fuel was extracted

through analysis of micrographs and other data collected as part of these programs. In cases in 

which the size of the rupture opening and the size of the fuel fragments were documented, RES 

compared the area of the rupture and the area of an average fuel particle. In cases in which the

average fuel particle cross-section was smaller than the area of the rupture opening, RES

concluded that dispersal was possible, although not always documented. Note that different

test programs measured different experimental parameters; consequently, the test results

displayed in the three tables shown below are grouped to optimize the layout of the tables.
52

Table 4-2 Summary of Power Burst Facility (PBF) Test Data (1 of 4)

PBF PBF PBF PBF

LOC-11 rod 1 LOC-11 rod 2 LOC-11 rod 3 LOC-11 rod 4

1978-1979 1978-1979 1978-1979 1978-1979

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

fresh fresh fresh fresh

Yes

No No No No

757 757 757 757

1205 1205 1205 1205

915 915 915 915

114 114 114 114

8.9-9.0 8.9-9.0 8.9-9.0 8.9-9.0

1.2 48.7 24.3 1.2

pressure (bar)

temperature 

(°C) 

length (mm)

width (mm)

-0.6% 3.0% 1.5% -0.8%

large: 2 to 9 mm

burnup (GWd/MTU)

Test Program

Test ID

Date

Cladding type

Fuel type

Rupture 

dimensions

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

in-pile unheated shroud, steady state 320°C, 15.2 Mpa, 450 W/cm, 3 

successive transients: PCT 557°C, 607°C, 757°C

Fragmentation size

Comments

Maximum circumferential strain 

(%)

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Transient ECR

Length of fuel between 

plenumand rupture (mm)
53

Table 4-3 Summary of Power Burst Facility (PBF) Test Data (2 of 4)

PBF PBF PBF PBF

LOC-3 rod 1 LOC-3 rod 2 LOC-3 rod 3 LOC-3 rod 4

1980-1981 1980-1981 1980-1981 1980-1981

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

fresh 15.96 fresh 16.62

Yes Yes Yes Yes

Yes Yes Yes Yes

Yes

917 917 917 917

1205 1205 1205 1205

875 889 876 889

114 114 114 114

7.79 10.6 7.66 11

24.5 23.8 49.2 47.5

pressure (bar) 16 10 51 48

temperature 

(°C) 917 1027 837 847

length (mm)

width (mm)

29.0% 40.0% 20.0% 41.6%

400 (5%) 100 (5%) 450 (5%) 550 (5%)

≈100 ≈75

<2% <5% <1% <3%

658 372 671 595

7% below 

1.18mm, 21% 

1.18-2mm, 72% 2-

5.6mm

4% below 

1.18mm, 22% 

1.18-2mm, 74% 2-

5.6mm

in-pile unheated 

shroud, steady 

state 320°C, 15.2 

Mpa, 450 W/cm, 

blowdown and 

decay heat ramp 

in-pile unheated 

shroud, steady 

state 320°C, 15.2 

Mpa, 450 W/cm, 

blowdown and 

decay heat ramp, 

fuel melting

burnup (GWd/MTU)

Test Program

Test ID

Date

Cladding type

Fuel type

Maximum circumferential strain 

(%)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

in-pile unheated shroud, steady state 

320°C, 15.2 Mpa, 450 W/cm, 

blowdown and decay heat ramp

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Transient ECR

Length of fuel between 

plenumand rupture (mm)

Fragmentation size

Comments
54

Table 4-4 Summary of Power Burst Facility (PBF) Test Data (3 of 4)

PBF PBF PBF

LOC-5 rod 6 LOC-5 rod 7A LOC-5 rod 7B

1980-1981 1980-1981 1980-1981

Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

17.66 fresh fresh

Yes Yes Yes

Yes Yes Yes

1077 1077 1077

1205 1205 1205

889 889 888

114 114 114

10 6.86 7.86

24.1 48.3 48.3

pressure (bar) 6 35 7

temperature 

(°C) 1077 887 1077

length (mm)

width (mm)

35.0% 19.0% 48.0%

520 (5%) 500 (5%) 500 (5%)

<6% <3% <2%

408 657 610

in-pile unheated shroud, steady state 320°C, 15.2 Mpa, 

450 W/cm, blowdown and decay heat ramp 

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Comments

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Transient ECR

Length of fuel between 

plenumand rupture (mm)

Fragmentation size
55

Table 4-5 Summary of Power Burst Facility (PBF) Test Data (4 of 4)

PBF PBF PBF PBF

LOC-6 rod 9 LOC-6 rod 10 LOC-6 rod 11 LOC-6 rod 12

1982-1983 1982-1983 1982-1983 1982-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

PWR 15x15 UO2 

9.6% enriched

fresh 10.8 fresh 10.8

Yes Yes

Yes Yes

No No

797 797 797 797

1205 1205 1205 1205

915 915 915 915

114 114 114 114

8.9-9.0 8.9-9.0 8.9-9.0 8.9-9.0

24.1 24.1 47.4 (120) 

inleakage 48.3

pressure (bar) 140 53

temperature 

(°C) 825 793

length (mm) 18.5 ≈22

width (mm) 7.5 ≈4

1.0% 13.6% 31.0% 74.0%

260 (5%) 180 (5%) 300 (5%)

541 555

5% below 1mm, 

25% 1-2mm, 70% 

2-5.8mm

5% below 1mm, 

25% 1-2mm, 70% 

2-5.8mm

burnup (GWd/MTU)

in-pile unheated shroud, steady state 

320°C, 15.2 Mpa, 450 W/cm, 

blowdown and decay heat ramp 

Test Program

Test ID

Date

Cladding type

Fuel type

Maximum circumferential strain 

(%)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

in-pile unheated shroud, steady state 

320°C, 15.2 Mpa, 450 W/cm

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Transient ECR

Length of fuel between 

plenumand rupture (mm)

Fragmentation size

Comments
56

Table 4-6 Summary of FR-2 Test Data (1 of 8)

FR-2 FR-2 FR-2 FR-2 FR-2

A1.1 A1.2 A2.1 A2.2 A2.3

1978-1983 1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

fresh fresh fresh fresh fresh

Yes Yes Yes Yes

Yes Yes Yes No

1002 1008 1050 1028 1015

973 973 973 973 973

500 500 500 500 500

30.642 30.524 30.466 30.541 30.596

25-100 25-100 25-100 25-100 25-100

pressure (bar) 50 88 58 25

temperature 

(°C) 810 820 860 1015

length (mm) 19 35 50 19

width (mm) 1.3 9 5.5 4.5

64.0% 36.2% 56.3% 34.7%

460 263 355 356

in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, steam steady state at 

300°C/6MPa, blowdown ramp 12°C/s, 40W/cm, scram at 927°C and steam quench at 727°C Comments

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Active Fuel Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

burnup (GWd/MTU)

Test Program

Test ID

Date

Cladding type

Fuel type
57

Table 4-7 Summary of FR-2 Test Data (2 of 8)

FR-2 FR-2 FR-2 FR-2 FR-2

B1.1 B1.2 B1.3 B1.4 B1.5

1978-1983 1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

fresh fresh fresh fresh fresh

No

No

1031 1010 985 1018 1009

973 973 973 973 973

500 500 500 500 500

30.581 30.637 30.55 30.593 30.726

55-90 55-90 55-90 55-90 55-90

pressure (bar) 52 45 61 45

temperature 

(°C) 900 915 845 910

length (mm) 41 11 36 45

width (mm) 9.5 1.8 8.5 3.9

29.0% 25.7% 34.2% 60.4%

210 162 220 341

Comments in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, steam steady state at 

300°C/6MPa, blowdown ramp 12°C/s, 40W/cm, scram at 927°C and steam quench at 727°C

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)
58

Table 4-8 Summary of FR-2 Test Data (3 of 8)

FR-2 FR-2 FR-2 FR-2

B1.6 B1.7 B3.1 B3.2

1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

fresh fresh fresh fresh

1015 890 1016 1011

973 973 973 973

500 500 500 500

30.54 30.61 30.229 30.293

55-90 55-90 55-90 55-90

pressure (bar) 80 61 79 50

temperature 

(°C) 825 840 825 915

length (mm) 28 49 27 33

width (mm) 9.5 9.6 9.8 8.2

38.0% 34.1% 36.9% 49.9%

210 242 261 258

Comments

in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, 

steam steady state at 300°C/6MPa, blowdown ramp 12°C/s, 40W/cm, 

scram at 927°C and steam quench at 727°C

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)
59

Table 4-9 Summary of FR-2 Test Data (4 of 8)

FR-2 FR-2 FR-2 FR-2 FR-2 FR-2

C1 C2 C3 C4 C5 C6

1978-1983 1978-1983 1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

2.56 2.56 2.56 2.56 2.56 2.56

Yes Yes Yes Yes Yes Yes

Yes Yes Yes Yes Yes

1017 950 773 1011 1003

973 973 973 973 973 973

500 500 500 500 500 500

30.562 30.56 30.754 30.734 30.606 30.821

25-110 25-110 25-110 25-110 25-110 25-110

pressure (bar) 46 30 98 65 22

temperature 

(°C) 900 945 749 815 916

length (mm) 31 25 33 42 18

width (mm) 6.9 7.9 10.5 9 2.1

51.2% 38.8% 36.7% 44.4% 62.2%

332 410 342 314 430

2.04% below 

1mm, 24.52% 1-

2mm, 69.08% 2-

5mm

2.04% below 

1mm, 24.52% 1-

2mm, 69.08% 2-

5mm

2.04% below 

1mm, 24.52% 1-

2mm, 69.08% 2-

5mm

2.04% below 

1mm, 24.52% 1-

2mm, 69.08% 2-

5mm

2.04% below 

1mm, 24.52% 1-

2mm, 69.08% 2-

5mm

2.04% below 

1mm, 24.52% 1-

2mm, 69.08% 2-

5mm

Fragmentation size

Comments in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, steam steady state at 300°C/6MPa, 

blowdown ramp 12°C/s, 40W/cm, scram at 927°C and steam quench at 727°C

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)
60

Table 4-10 Summary of FR-2 Test Data (5 of 8)

FR-2 FR-2 FR-2 FR-2 FR-2 FR-2

E1 E2 E3 E4 E5 E6

1978-1983 1978-1983 1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

8 8 8 8 8 8

Yes Yes Yes Yes Yes Yes

Yes Yes Yes Yes Yes

1009 999 1000 1005 933

973 973 973 973 973 973

500 500 500 500 500 500

30.634 30.75 30.755 30.74 30.739 30.772

25-120 25-120 25-120 25-120 25-120 25-120

pressure (bar) 23 113 49 72 24

temperature 

(°C) 910 708 860 781 929

length (mm) 13 17 14 31 6

width (mm) 3.4 7.5 6 4.3 0.1

30.4% 46.0% 30.9% 55.5% 67.4%

330 370 185 388 435

4.2% below 1mm, 

26.14% 1-2mm, 

67.37% 2-5mm

4.2% below 1mm, 

26.14% 1-2mm, 

67.37% 2-5mm

4.2% below 1mm, 

26.14% 1-2mm, 

67.37% 2-5mm

4.2% below 1mm, 

26.14% 1-2mm, 

67.37% 2-5mm

4.2% below 1mm, 

26.14% 1-2mm, 

67.37% 2-5mm

4.2% below 1mm, 

26.14% 1-2mm, 

67.37% 2-5mm

Fragmentation size

Comments in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, steam steady state at 300°C/6MPa, 

blowdown ramp 12°C/s, 40W/cm, scram at 927°C and steam quench at 727°C

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)
61

Table 4-11 Summary of FR-2 Test Data (6 of 8)

FR-2 FR-2 FR-2 FR-2 FR-2 FR-2

F1 F2 F3 F4 F5 F6

1978-1983 1978-1983 1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

21.91 21.91 21.91 21.91 21.91 21.91

Yes Yes Yes Yes Yes Yes

Yes Yes Yes Yes Yes

1016 1007 1017 1049 1006

973 973 973 973 973 973

500 500 500 500 500 500

30.817 30.82 30.734 30.709 30.675 30.746

45-85 45-85 45-85 45-85 45-85 45-85

pressure (bar) 56 53 42 72 60

temperature 

(°C) 890 893 932 835 880

length (mm) 62 14 20 28 31

width (mm) 6.1 2.6 6 9 8

59.0% 37.5% 27.3% 34.1% 41.2%

158 90 169 176 175

11.43% below 

1mm, 46.11% 1-

2mm, 37.64% 2-

5mm

11.43% below 

1mm, 46.11% 1-

2mm, 37.64% 2-

5mm

11.43% below 

1mm, 46.11% 1-

2mm, 37.64% 2-

5mm

11.43% below 

1mm, 46.11% 1-

2mm, 37.64% 2-

5mm

11.43% below 

1mm, 46.11% 1-

2mm, 37.64% 2-

5mm

11.43% below 

1mm, 46.11% 1-

2mm, 37.64% 2-

5mm

Fragmentation size

Comments in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, steam steady state at 300°C/6MPa, 

blowdown ramp 12°C/s, 40W/cm, scram at 927°C and steam quench at 727°C

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)
62

Table 4-12 Summary of FR-2 Test Data (7 of 8)

FR-2 FR-2 FR-2 FR-2 FR-2 FR-2

G1.1 G1.2 G1.3 G1.4 G1.5 G1.6

1978-1983 1978-1983 1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

33.78 33.78 33.78 33.78 33.78 33.78

Yes Yes Yes Yes Yes Yes

Yes Yes Yes Yes No

1010 1009 977 971 927

973 973 973 973 973 973

500 500 500 500 500 500

30.343 30.224 30.208 30.145 30.194 30.134

50-90 50-90 50-90 50-90 50-90 50-90

pressure (bar) 68 41 83 52

temperature 

(°C) 730 890 785 780

length (mm) 4 27 25 44

width (mm) 0.4 2.6 10 7.2

29.5% 62.3% 32.6% 40.8%

201 178 183 153

9.36% below 

1mm, 26.84% 1-

2mm, 51.76% 2-

5mm

9.36% below 

1mm, 26.84% 1-

2mm, 51.76% 2-

5mm

9.36% below 

1mm, 26.84% 1-

2mm, 51.76% 2-

5mm

9.36% below 

1mm, 26.84% 1-

2mm, 51.76% 2-

5mm

9.36% below 

1mm, 26.84% 1-

2mm, 51.76% 2-

5mm

9.36% below 

1mm, 26.84% 1-

2mm, 51.76% 2-

5mm

Fragmentation size

Comments in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, steam steady state at 300°C/6MPa, 

blowdown ramp 12°C/s, 40W/cm, scram at 927°C and steam quench at 727°C

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)
63

Table 4-13 Summary of FR-2 Test Data (8 of 8)

FR-2 FR-2 FR-2 FR-2 FR-2 FR-2

G2.1 G2.2 G3.1 G3.2 G3.3 G3.6

1978-1983 1978-1983 1978-1983 1978-1983 1978-1983 1978-1983

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

PWR UO2 4.7% 

enriched

36.52 36.52 36.52 36.52 36.52 36.52

Yes Yes Yes Yes Yes Yes

Yes Yes Yes Yes Yes No

952 940 930 940 947

973 973 973 973 973 973

500 500 500 500 500 500

30.713 30.653 30.102 29.942 29.955 30.229

60-125 60-125 60-125 60-125 60-125 60-125

pressure (bar) 37 66 33 57 111

temperature 

(°C) 869 846 900 838 750

length (mm) 6 33 29 39 27

width (mm) 1.5 10.9 7.2 9.7 11

31.7% 28.3% 45.7% 41.4% 32.4%

55 221 219 288 202

8.50% below 

1mm, 41.80% 1-

2mm, 50.04% 2-

5mm

8.50% below 

1mm, 41.80% 1-

2mm, 50.04% 2-

5mm

8.50% below 

1mm, 41.80% 1-

2mm, 50.04% 2-

5mm

8.50% below 

1mm, 41.80% 1-

2mm, 50.04% 2-

5mm

8.50% below 

1mm, 41.80% 1-

2mm, 50.04% 2-

5mm

8.50% below 

1mm, 41.80% 1-

2mm, 50.04% 2-

5mm

Fragmentation size

Comments in-pile, no heaters, pre-irradiation in FR-2: low corrosion and hydrogen, steam steady state at 300°C/6MPa, 

blowdown ramp 12°C/s, 40W/cm, scram at 927°C and steam quench at 727°C

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Length of fuel between 

plenumand rupture (mm)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)
64

Table 4-14 Summary of PHEBUS-LOCA Test Data

PHEBUS-LOCA PHEBUS-LOCA PHEBUS-LOCA PHEBUS-LOCA PHEBUS-LOCA

215-P 215-R 216 218 219

7/8/1982 5/6/1983 12/1/1983 7/19/1984 12/1/1984

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR PWR PWR PWR PWR

fresh fresh fresh fresh fresh

Yes Yes

1250 1050 1350 1360 hot rod, 

1200 average 1330

1000 1000 1000 1000 1000

800 800 800 800 800

40 40 40

33.5 except 2 

rods at 5 and 2 

rods lower

30 with 4 corner 

rods 

unpressurized

pressure (bar)

temperature 

(°C) 

44-54% hot zone, 

18-25% cold 

zone, 15-38 outer 

rods

20-49% inner 

rods, 17-39% 

outer rods

15-30%

11-27% inner 

rods, 12-36% 

outer rods

18.6-46.2% inner 

rods, 14.5-26.4% 

outer rods

150 150 60 inner rods, 80 

outer rods

80 inner rods, 20 

outer rods

75um and 62um 

outer and inner on 

max rod

2 sided oxidation 

up to 80-90 um 

on each side

2 sided oxidation 

up to 110 um 

total, alpha as 

large as oxide

5x5 bundle in 

PWR loop, 

320°C, 15.5 MPa, 

breaks simulated 

in hot or cold leg, 

reflood possible 

from top or 

bottom, driver 

core power 

reduced during 

transient to 

simulate scram, 

21 pressurized 

rods all burst

5x5 bundle in 

PWR loop, 

320°C, 15.5 MPa, 

breaks simulated 

in hot or cold leg, 

reflood possible 

from top or 

bottom, driver 

core power 

reduced during 

transient to 

simulate scram, 

19 rods burst, 3 

low pressure 

rods, 3 

depressurized 

inadvertently

5x5 bundle in 

PWR loop, 

320°C, 15.5 MPa, 

breaks simulated 

in hot or cold leg, 

reflood possible 

from top or 

bottom, driver 

core power 

reduced during 

transient to 

simulate scram, 

all rods burst

5x5 bundle in 

PWR loop, 

320°C, 15.5 MPa, 

breaks simulated 

in hot or cold leg, 

reflood possible 

from top or 

bottom, driver 

core power 

reduced during 

transient to 

simulate scram, 

20 rods burst

5x5 bundle in 

PWR loop, 

320°C, 15.5 MPa, 

breaks simulated 

in hot or cold leg, 

reflood possible 

from top or 

bottom, driver 

core power 

reduced during 

transient to 

simulate scram, 

large lateral rod 

displacements, 2 

rods unzipped 

over 60-80mm 

(#8, 18)

Comments

Maximum circumferential strain 

(%)

Length of balloon (strain > 10%) 

with respect to strain (mm)

Transient ECR

Active Fuel Length (mm)

Rod fill pressure (bar)

Rupture 

conditions

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

burnup (GWd/MTU)

Test Program

Test ID

Date

Cladding type

Fuel type
65

Table 4-15 Summary of FLASH Test Data

FLASH FLASH FLASH FLASH FLASH

FLASH-1 FLASH-2 FLASH-3 FLASH-4 FLASH-5

circa 1990 circa 1990 circa 1990 circa 1990 circa 1990

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR 17x17 PWR 17x17 PWR 17x17 PWR 17x17 PWR 17x17

1.65-3.32 1.65-3.32 1.65-3.32 1.65-3.32 51.712

Yes

Yes

1100 1100 1270 1270 1350

300 300 300 300 300

temperature 

(°C) 930 940 995

62.0% 16.0%

unheated shroud tube, irradiation at 35 to 40 kW/m, 13 Mpa, adjustment to 7 kW/m, 

blowdown to 0.5-2.1 MPa, injection of He, ramp rate 28°C/s, quench at hot conditions, 

nuclear power maintained 10 minutes, large azimuthal ΔT

Maximum circumferential strain 

(%)

Comments

Fragmentation observed?

Relocation observed?

PCT (°C)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)
66

Table 4-16 Summary of Argonne National Laboratory Test Data

ANL ANL ANL ANL

ICL1 ICL2 ICL3 ICL4

09/23/02-12/04/03 09/23/02-12/04/03 09/23/02-12/04/03 09/23/02-12/04/03

Zr-2 Zr-2 Zr-2 Zr-2

BWR 3.95% 

enriched

BWR 3.95% 

enriched

BWR 3.95% 

enriched

BWR 3.95% 

enriched

56 56 56 56

Yes Yes Yes Yes

1200 1204 1204

300 300 300 300

270 270 270 270

13 13 13 13

10 10 10 10

80 80 80 80

pressure (bar) 86 80 86 80

temperature 

(°C) 755 750 730 790

length (mm) 13 14 11 15

width (mm) 3 3.5 4.6 5.1

38+9 39+10% 43+9 36+9%

70 90 100 80

70 70 70 70

10 10 10 10

0 20 21 20

140 140

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Pre-transient oxide thickness 

(microns)

Transient ECR

Length of fuel between 

plenumand rupture (mm)

Maximum circumferential strain 

(%)

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

burnup (GWd/MTU)

Test Program

Test ID

Date

Cladding type

Fuel type
67

Table 4-17 Summary of Halden Test Data (1 of 2)

Halden Halden Halden Halden Halden

IFA-650.2 IFA-650.3 IFA-650.4 IFA-650.5 IFA-650.6

4/30/2005 4/25/2006 10/23/2006 5/18/2007

Zircaloy-4 - 

duplex

Zircaloy-4 - 

duplex

Zircaloy-4 - 

duplex E110

PWR 3.5% 

enriched

PWR 3.5% 

enriched

PWR 3.5% 

enriched

VVER 3.6% 

enriched

fresh 82 92 83.4 56

Yes Yes Yes Yes Yes

Yes No Yes No Yes

Yes No Yes Yes No

1050 800 800 1080 830

497 450

480 480 480 480

21.5 21.5 15 17

70 40 40 40 30

pressure (bar) 71 53 70 62

temperature 

(°C) 800 780 785 750 800

length (mm) 35 45 10

width (mm) 20 crack crack

54.0% 60.0%

>300 >200

5 250 50 650 100

0 25 11 65-80 5

0 0 0

265 220 ≈400

pressure decay 

from 70 to 55 

over 60 sec

slow 

depressurization 

after heatup

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Pre-transient oxide thickness 

(microns)

Transient ECR

Length of fuel between 

plenumand rupture (mm)

Comments

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)
68

Table 4-18 Summary of Halden Test Data (2 of 2)

Halden Halden Halden Halden

IFA-650.7 IFA-650.9 IFA-650.10 IFA-650.11

4/18/2008 4/17/2009 5/16/2010

Zircaloy-2 Zircaloy-4 - 

duplex Zircaloy-4

BWR 4.46% 

enriched

PWR 3.5% 

enriched

PWR 4.487% 

enriched

44 90 61 56

Yes Yes

Yes No Yes

Yes Yes No

1160 1200 840 1000

500 500

480 480 440

20 20

17 19 17

6 40 40

pressure (bar) 10.9 62 70

temperature 

(°C) 1100 810 755

length (mm) 5

width (mm) crack

44 30 150-220

4.4-10 8 20-30

15.3 0

200

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Pre-transient oxide thickness 

(microns)

Transient ECR

Length of fuel between 

plenumand rupture (mm)

Comments

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

Maximum circumferential strain 

(%)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Test Program

Test ID

Date

Cladding type

Fuel type

burnup (GWd/MTU)
69

Table 4-19 Summary of PNL/NRU Test Data

PNL/NRU PNL/NRU PNL/NRU PNL/NRU

MT-1 MT-2 MT-3 MT-4

Apr-81 Jul-81 Nov-81 May-82

Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

PWR 17x17 2.93% 

enriched

PWR 17x17 2.93% 

enriched

PWR 17x17 2.93% 

enriched

PWR 17x17 2.93% 

enriched

fresh pre-conditioned fresh pre-conditioned fresh pre-conditioned fresh pre-conditioned

Yes Yes Yes Yes

No Yes Yes Yes

Yes Yes

871 888 853 1186

3910 3910 3910 3910

3660 3660 3660 3660

250 250 250 250

32 32 39 46

pressure (bar) 94 105

temperature 

(°C) 872 883 794 821

65% (ave 43%) 72% (ave 43%) 94% (ave 47%) 96% (ave 72%)

750 total 750 total 375 total 550 total

0 0 0 0

1625 1625 1000 1000

large Medium-small

32 PWR rods (6x6 minus 

corners, 12 test rods, 

surrounded by guard rods), 

hold at 375°C in steam, cut 

off steam to generate 

transient, quench, scram, 

12/12 rods burst after start 

of quench, 2 balloons 

pinned by grid spacers

32 PWR rods (6x6 minus 

corners, 11 test rods, 1 

water rod, surrounded by 

guard rods), hold at 375°C 

in steam, cut off steam to 

generate transient, quench, 

scram, 8/11 rods burst 

after start of quench, 2 

balloons (1 double hump) 

pinned by grid spacers

32 PWR rods (6x6 minus corners, 11 test rods, 1 water 

rod, surrounded by guard rods), hold at 375°C in steam, 

cut off steam to generate transient, quench, scram, 

6/11 rods burst after start of quench, 2 double hump 

balloons pinned by grid spacers

Comments

Length of fuel between 

plenumand rupture (mm)

Fragmentation size

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient oxide thickness 

(microns)

Rupture 

conditions

Maximum circumferential strain 

(%)

Rod fill pressure (bar)

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Fuel loss observed?

Cladding type

Fuel type

burnup (GWd/MTU)

Plenum Length (mm)

Test Program

Test ID

Date

Fragmentation observed?

Relocation observed?
70

Table 4-20 Summary of Studsvik LOCA Test Data

Studsvik Studsvik Studsvik Studsvik Studsvik Studsvik

189 191 192 193 196 198

11/1/2010 2/7/2011 2/21/2011 3/11/2011 11/11/2011 11/16/2011

ZIRLO ZIRLO ZIRLO ZIRLO ZIRLO ZIRLO

PWR 4.00% 

enriched

PWR 4.00% 

enriched

PWR 4.00% 

enriched

PWR 4.00% 

enriched

PWR 4.94% 

enriched

PWR 4.94% 

enriched

72.6 71 72.6 71 55.2 55.2

Yes Yes Yes Yes Yes Yes

Yes Yes Yes Yes No No

Yes Yes Yes Yes No No

950 1185 1185 1185 700 1185

300 300 300 300 300 300

280 280 280 280 280 280

15 15 15 15 15 15

>10 >10 >10 >10 >10 >10

110 110 82 83 83 82

pressure (bar) 113 104 77 77 72 74

temperature 

(°C) 700 680 700 730 686 693

length (mm) 23.9 21.6 22.7 17.8 11 1.5

width (mm) 10.5 17.5 9 13.8 1.6 0.2

48.0% 50.0% 56.0% 51.0% 29.0% 26.0%

90 80 90 100 90 100

≈200 ≈200 ≈200 ≈200 ≈200 ≈200

0% 13% 11% 17% 0% 17%

≈150 ≈150 ≈150 ≈150 ≈150 ≈150

mostly fine, some 

up to 5 mm

mostly fine, some 

up to 5 mm

mostly fine, some 

up to 5 mm

mostly fine, some 

up to 5 mm

Large, up to half 

pellet

Large, up to half 

pellet

Comments Out-of-pile, external heating, single rod

Length of balloon (strain > 10%) 

with respect to strain (mm)

Pre-transient hydrogen content 

(wt.ppm)

Transient ECR

Length of fuel between 

plenumand rupture (mm)

Fragmentation size

Maximum circumferential strain 

(%)

Fragmentation observed?

Relocation observed?

Fuel loss observed?

PCT (°C)

Test Rod Length (mm)

Active Fuel Length (mm)

Plenum Length (mm)

Cold void volume 

(cubic centimeters)

Rod fill pressure (bar)

Rupture 

conditions

Rupture 

dimensions

burnup (GWd/MTU)

Test Program

Test ID

Date

Cladding type

Fuel type
71

4.2.2 Data Trends

RES analyzed the data extracted from the integral testing programs to look for any correlation

between the variables documented. The reviewers generated plots of the various combinations

of variables that were measured in the test programs, in an attempt to determine any existing

trends that could shed light on these phenomena. Many plots were inconclusive, but the trends

that were identified are listed in Table 4-21. Some of these trends were expected and obvious

but are nonetheless listed (green = correlation, red = inverse correlation, gray = no correlation).

Table 4-21 Trends Identified from Past LOCA Test Programs (Green = Correlation, Red =

Inverse Correlation, Gray = No Correlation)

Variable Change Definite Trend Possible Trend

PCT increase

ECR increase Rupture pressure decrease

Rupture pressure increase

Fragment size decrease

Burnup increase Fragment size decrease (Figure 4-50)

Rod length increase Balloon length increase

Active fuel length increase Balloon length increase

Void volume increase

Rupture length increase

Rupture width decrease

Fill pressure increase

Rupture pressure increase (Figure 4-53) Balloon length decrease (Figure 4-52)

Rupture temperature decrease (Figure 4-53) Rupture area increase

Rupture width increase (Figure 4-51)

Initial hydrogen increase

Rupture pressure increase

Rupture temperature decrease

Initial oxygen increase

Rupture pressure increase

Rupture temperature decrease

Rupture to plenum increase Balloon length increase (Figure 4-52)

Rupture pressure increase

Rupture temperature decrease (Figure 4-54) Rupture length increase

Rupture width increase

Rupture area increase

Rupture temperature increase

Rupture length increase

Rupture width decrease

Maximum strain decrease

Balloon length increase

Rupture length increase

Rupture width increase (Figure 4-56)

Rupture area increase (Figure 4-55)

Rupture width increase Rupture area increase (Figure 4-55)

Maximum strain increase Balloon length increase
72

Figure 4-50 through Figure 4-56 show the plots of the trends observed when analyzing data

from the LOCA test programs described above. The following trends are observable:

• Fuel fragmentation appears to increase with burnup.

• Rod fill pressure has a direct impact on the balloon and rupture characteristics, such that

increased rod fill pressure results in shorter balloons but wider rupture openings.

• Rod fill pressure has a direct impact on rupture pressure and temperature, such that 

increased fill pressure results in increased rupture pressure and decreased rupture

temperature.

• There is a strong inverse correlation between rupture pressure and temperature.

• As expected, the rupture area increases with rupture width and length.

• Balloon length increases if the rupture (i.e., the balloon) is further from the plenum.

Figure 4-50 Average fuel fragment cross-section as a function of burnup, showing

increased fragmentation with increasing burnup.
73

Figure 4-51 Rupture width as a function of rod fill pressure, showing increasing rupture

width with increasing rod fill pressure

(a) (b)

Figure 4-52 Length of ballooned section of the rod as a function of (a) rod fill pressure, 

and (b) rupture to plenum distance, showing that balloon length decreases 

with increase rod fill pressure, and increases when the rupture occurs 

further from the plenum.

(a) (b)

Figure 4-53 Rupture pressure (a) and temperature (b) as a function of rod fill pressure,

showing opposite trends, implying that rupture pressure and temperature

are inversely correlated, as shown in Figure 4-54
74

Figure 4-54 Rupture temperature as a function of rupture pressure, showing that rupture

temperature decreases as rupture pressure increases

(a) (b)

Figure 4-55 Rupture area as a function of (a) rupture length and (b) rupture width,

showing that rupture area increases with longer and wider ruptures

Figure 4-56 Rupture width as a function of rupture length, showing that rupture width 

and length increase together, with rupture width generally being between on 

half and one fifth of the rupture length
75

4.2.3 Summary

The following are the important conclusions that can be drawn from the available LOCA data

about fuel fragmentation, relocation, and dispersals:

Fuel fragmentation—that is, fracture of the fuel pellet into large fragments—appears to occur as

soon as any meaningful amount of burnup is accumulated, as low as a few megawatt days per 

metric ton uranium (MWd/MTU). Fragmentation in irradiated fuel occurred in all but one case

for which it was assessed. The only case in which it did not occur was in an unirradiated fuel

rod.

• The size of fuel fragments is not uniform but tends to become smaller with increasing

burnup.

• Axial relocation occurs in the presence of appreciable cladding diametral strain

accompanied by appreciable fragmentation. Although cracking could be observed, 

tests in which fresh fuel was used and the pellets essentially remained in their initial 

shape without fragmenting into several pieces generally did not result in axial relocation.

• Reported diametral strains resulting in fuel axial relocation are 8 percent in the FR-2 test 

series, 13 to 15 percent in the Halden LOCA IFA-650.X series, and 5 to 12 percent in the 

Studsvik LOCA test series. Axial fuel relocation was observed even in tests on fresh 

fuel, as long as the pellets were fragmented into several pieces. It should be noted that 

experience from the defueling of irradiated nonballooned fuel rods shows that it is very

difficult to remove the fuel from within the cladding, indicating that fuel relocation and

dispersal from punctured but nonballooned rods is unlikely.

• Grid spacers may “pin” rod ballooning, potentially acting as choke points for fuel 

relocation. In bundle geometries, ballooning tends to occur such that all the balloons 

are coplanar, but ballooning is largely suppressed in the sections of fuel rods that cross

a grid spacer.

• The rod fill pressure has an important effect on rupture, the characteristics of the 

balloon, and rupture opening sizes and shapes. Increased rod internal pressure 

generally results in shorter balloons but wider rupture openings.

• Some fuel dispersal has been observed in every case in which (1) rod rupture occurs,

and (2) the fuel fragments are small enough to get through the rupture opening.

• The amount of fuel that is dispersed can vary widely, from a puff of dust to large

amounts of fragmented and pulverized fuel. Although evidence points to likely fuel

dispersal in many tests, this phenomenon was not systematically investigated nor 

documented in the majority of test programs.

77

5 Consequences of Fuel Fragmentation, Relocation, and Dispersal

This section describes some of the potential consequences of fuel fragmentation, relocation,

and dispersal. These include a change in axial heat flux profile and temperature profile, fuelcoolant interaction (FCI), the hydraulic and mechanical effects (i.e., flow blockage) of fuel

material in the reactor coolant system, and the radiological consequences of fuel dispersal. Of

these effects, both change in axial heat flux profile (with the associated change in temperature 

profile) and flow blockage may occur without rupture of the cladding. The others require rupture

of the fuel rod cladding.

5.1 Consequences of Fuel Relocation

As discussed in Section 4 of this report, fuel fragmentation is exhibited by essentially all

irradiated fuel material to some degree. Under LOCA conditions, relocation of the fuel into the

balloon region may occur, with subsequent dispersal of fuel fragments outside the cladding if a

rupture occurs. Based on thermocouple measurements from the Halden reactor, fuel relocation

is coincident with the rupture of the fuel rod (Refs. 40, 45, and 51). The filling ratio of the

relocated fuel in the balloon is an important parameter associated with fuel relocation. This ratio,

also called the packing fraction, is the ratio of the volume of fuel in the balloon region to the total

volume of this region. There have been efforts to model this ratio, including the INEL empirical 

model based on PBF and FR-2 data (mentioned in Section 4 of this report), to calculate the 

transient axial relocation (Ref. 21).

The additional fuel accumulating in the balloon region may create a local increase in heat flux. 

Although a preliminary evaluation by the Halden Reactor Project indicated that this local heat 

flux increase will not cause coolability problems over the investigated range of parameters, it

was clear that under specific conditions the local heat flux increase can delay the cooldown of

the assembly during reflood (Ref. 51). The Halden Reactor Project noted that the effect of fuel

relocation on the local heat flux can be taken into account in current design approaches and

tools through an additional peaking factor (the axial power profile calculations already include

various sources of power peaking), together with a reduction or closure of the fuel-cladding gap

in the ballooned section. Finally, the potential effects of changes in core geometry (fuel rod 

ballooning with fuel relocation) on criticality analyses during post-LOCA recirculation should be 

evaluated as part of any future safety assessment.

5.2 Consequences of Fuel Dispersal

5.2.1 Core Damage Distribution

As discussed in Section 4 of this report, fuel dispersal has been observed in all simulated LOCA 

experiments in which the two following conditions are met: (1) rod rupture occurs, and (2) the 

fuel fragments are small enough to get through the rupture opening. The amount of fuel that is 

dispersed can vary widely. It is reasonable to state that the consequences of fuel dispersal 

depend on the amount of fuel dispersal. The amount of fuel dispersal depends on the number 

of rods that rupture and the amount of fuel that escapes each rod through the rupture opening.

The number of rods that rupture during a large-break LOCA is dependent on a number of 

factors. Foremost is the number of rods in the core that achieve high cladding temperatures 

during the transient. This is a function of the rod power census and the ECCS design capability. 

Another important factor is the cladding pressure differential for the various rods, which is 

affected by the initial helium fill pressure, the fission gas generation and fission gas release from 
78

the fuel with burnup, the fuel rod power, and the coolant pressure. A number of best-estimate 

analyses performed in the United States and abroad have shown that the number of fuel rods 

predicted to rupture during a LOCA may be quite limited. These predictions are a place to start 

when assessing the amount of fuel dispersed, but knowledge of the size of fuel fragments and

the rupture opening, as well as quantifying the axial mobility of fragmented fuel, would all be 

necessary to quantify the amount of fuel that escapes each rod through the rupture opening 

and, therefore, the amount of fuel dispersal. It should be noted that the inputs, assumptions, 

and modeling techniques used in these best-estimate analyses have not been reviewed by the 

NRC staff, and have not been endorsed by NRC. As such, these results should not be used to 

inform any safety assessment without prior evaluation and approval by the NRC.

The current state of knowledge about the controlling phenomenon and modeling capabilities for 

fuel dispersal in the event of fuel rod rupture is not sufficient to accurately predict the amount of 

fuel dispersal. However, reasonable discussion can be made of the potential consequences of 

fuel dispersal in general, including FCI, the radiological consequences, and the hydraulic and 

mechanical consequences.

5.2.2 Fuel-Coolant Interaction

FCI can occur when hot fuel particles rapidly release their energy into a surrounding coolant

with excessive amounts of vapor production. If this event occurs within a short timescale

compared to vapor expansion, it can cause local pressurization similar to an explosion and

threaten the surroundings by the subsequent high-pressure vapor expansion.

In the nuclear industry, the vapor explosion phenomenon has been an issue in safety analyses

for many years. Vapor explosions are a potential hazard in LWRs after a prolonged lack of

cooling allows reactor core materials to melt and contact residual water coolant within the

reactor vessel or below in the containment reactor cavity (Ref. 52).

Two important components are necessary to cause a vapor explosion in a reactor: hot

(e.g., molten) fuel and relatively cool reactor coolant. In a design-basis LOCA, fuel 

temperatures (which initially decrease as a result of negative void reactivity and control rod 

insertion) may increase and cause a cladding temperature increase until the peak cladding

temperature (1,204 degrees C) limit specified in 10 CFR 50.46 is reached. Although this

temperature limit does not preclude cladding rupture (around 800 degrees C), it is well below

the melting point of the fuel. Nonetheless, no regulatory criteria preclude fuel melting during a 

LOCA, and initial conditions and assumptions are not selected to maximize fuel temperature in 

LOCA analyses. However, to reach temperatures sufficiently high to cause core melting, the

coolant in contact with the cladding must also be marginal (e.g., dry steam). Consequently, for

a design-basis LOCA, neither component (molten fuel or liquid coolant) is available; thus,

energetic FCI is unlikely.

5.2.3 Hydraulic and Mechanical Effects of Dispersed Fuel Material

Generic Safety Issue (GSI) 191, “Assessment of Debris Accumulation on PWR Sump 

Performance” (Ref. 53), was initially identified to deal with the issue of emergency-sump 

blockage during the recirculation phase of a LOCA. Consequently, the licensees did some 

testing to demonstrate adequate flow and head-loss through sump strainers. In 2004, the NRC 

added the issue of core-inlet blockage to GSI-191, requiring the nuclear industry to perform 

tests to evaluate the impact of debris and chemicals on flow blockage through the lower vessel 

internals, core inlet, and fuel assemblies.
79

GSI-191 considered a wide range of debris and chemicals that could be entrained from various 

locations in the containment building to the sump screens and, from there, to the core inlet. 

GSI-191 did not address dispersed particulate fuel or the chemical species present in irradiated 

fuel. Thus, it is unclear whether the assumptions of GSI-191 could be applied to dispersed fuel 

during the recirculation of coolant through the emergency sump and the ECCS pumps with 

respect to chemical effects, strainer head-loss, or downstream effects. In addition, the impact of 

heat deposited by dispersed fuel into the reactor coolant system on timing of core reflood and 

overall ECCS performance is not yet known. The impact of fuel particle transport and 

deposition on coolability and criticality of deposited fuel fragments, as well as equipment 

qualification, are also unknown. The staff has submitted the issue of fuel dispersal for screening 

in the Generic Issue Program, and is considering what vehicle should be used to treat this 

issue.

5.2.4 Radiological Effects of Dispersed Fuel Material

A discussion of the radiological effects of dispersed fuel material must start with the regulatory 

context. Consideration of the radiological effects of postulated accidents is one of the

fundamental tenants of reactor safety analysis. The issue has been historically complex, and it

has been subject to considerable evolution in the regulatory review process.

As specified in 10 CFR Part 50 (Ref. 2), an applicant for an operating license must provide an

analysis and evaluation of a proposed commercial nuclear reactor in order to assess the risk to

public health and safety that would result from operation of that facility. Further,

10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants” (Ref. 54),

requires that applicants provide a similar analysis and evaluation of the proposed site.

For power reactor applications before 1997, the criteria for evaluating the radiological aspects of

the proposed site appear in 10 CFR Part 100, “Reactor Site Criteria” (Ref. 55). A footnote to

10 CFR 100.11, “Determination of Exclusion Area, Low Population Zone, and Population Center 

Distance,” states that “the fission product release assumed for these calculations should be

based upon a major accident...assumed to result in substantial meltdown of the core with

subsequent release of appreciable quantities of fission products.” As a result, the radiological 

aspects of plant safety analysis consider very severe accident conditions; that is, conditions well 

beyond simple estimates of fuel damage that may occur during a less severe event, such as a 

successfully terminated LOCA.

As a source of further guidance on these analyses, 10 CFR 100.11 cites Technical Information

Document (TID) 14844, “Calculation of Distance Factors for Power and Test Reactor Sites,”

issued March 1962 (Ref. 56). Although initially used only for site evaluations, the TID-14844

source term has been used in other design-basis applications, such as environmental

qualification of equipment under 10 CFR 50.49, “Environmental Qualification of Electric 

Equipment Important to Safety for Nuclear Power Plants.” . Other sources of guidance include 

NUREG-1465 (Ref. 57) and Regulatory Guide 1.183 (Ref. 58).

As noted in 10 CFR 100.11, the radiological source term is intended to be representative of a

major accident involving significant core damage. Such an accident is typically postulated to

occur in conjunction with a large-break LOCA but need not be bounded by other LOCA criteria

(e.g., the cladding embrittlement criteria in 10 CFR 50.46).
80

As examples of the assumptions that the NRC originally used for evaluating the radiological

consequences of a LOCA, Regulatory Guide 1.3 Revision 2, “Assumptions Used for Evaluating 

the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water 

Reactors” (Ref. 59), and Regulatory Guide 1.4 Revision 2, “Assumptions Used for Evaluating 

the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water 

Reactors” (Ref. 60), both state the following:

• Twenty-five percent of the equilibrium radioactive iodine inventory developed from

maximum full power operation of the core should be assumed to be immediately

available for leakage from the primary reactor containment...

• One hundred percent of the equilibrium radioactive noble gas inventory developed from

maximum full power operation of the core should be assumed to be immediately

available for leakage from the reactor containment.

Thus, the regulatory guides assume that a significant fraction of the volatiles and essentially all

of the noble gases have escaped from the fuel, the fuel rod cladding, and the reactor coolant

system and are therefore available for leakage from the reactor containment.

Even after a successfully terminated large-break LOCA, where all safeguard systems have

worked as planned, a number of fuel rods will have failed (Ref. 61). Large amounts of

radioactivity are assumed to have reached the containment in the form of volatile or noble

gases, and also particulate aerosols, which would transport other radionuclides trapped in the

fuel material. From a regulatory point of view, some of this radioactivity will leak out of the

containment and reach the environment outside the nuclear plant, where it will cause a dose

burden to the public. In the traditional design-basis regulatory calculation, it was assumed that

the containment leak rate is the highest permissible rate stated in the technical specifications or

license conditions of the plant.

While radiological safety assessments conducted in the United States assume that essentially

all of the fuel rods in the core rupture during a LOCA, calculations of best-estimate releases

from the fuel have been performed in Europe. Table 5-1 summarizes the findings of a 

European study for fuel with a burnup up to 50 GWd/MTU. Based on the results shown in this 

table, it is clear that, in regulatory space, the source term is independent of whether the fuel

actually failed or not and is independent of the manner or extent of the failure.

Table 5-1 The Release from the Gap and Fuel for Radiologically Significant Nuclides

Best Estimate Conservative

Nuclide Gap Release

(%)

Fuel Release

(%)

Total Release

(%)

Total Release

(%)

Kr-85 1.0 6.5 7.5 9.5

Xe-133 0.2 1.95 2.15 4.2

I-131 0.25 0.4 0.65 2.5

Cs-134 1.0 0.86 1.86 3.1

Cs-137 1.0 1.0 2.0 3.3

Regarding experimental measurement of radiological release during LOCA, the Halden Reactor

Project measured the release of radioactive fission products from failed fuel in the last three 
81

tests in the IFA-650 LOCA series (Ref. 62). The focus was on the release of iodine (I)-131 and

cesium (Cs)-137, as shown in Table 5-2.

Table 5-2 Radiological Release in Halden LOCA Tests

Test Iodine-131

Release (%)

Cesium-137

Release (%)

IFA-650.10 (PWR) 0.2 0.6

IFA-650.11 (VVER) 0.6 0.3

In both of these tests, the measured release fractions for both isotopes are quite low—less than

1 percent. For the IFA-650.9 LOCA test, in which extensive fuel dispersal was observed, the

measured release of I-131 was 0.4 percent of total calculated inventory (Ref. 63). This result

suggests that actual release fractions are small—even in the presence of very high burnup and

extensive fuel dispersal.

In conclusion, the current regulatory framework in the United States to determine the 

radiological consequences of a LOCA, which assumes that most of the volatile fission product 

inventory as well as some particulate aerosols are released from the fuel upon fuel rod rupture, 

is largely bounding based on scoping analyses and experiments performed in the United States

and abroad.

83

6 Conclusions

RIL-0801 discussed the technical basis for revising 10 CFR 50.46(b) rulemaking, axial fuel

relocation, and the loss of fuel particles through a rupture opening.

The purpose of this report is to revisit the conclusions of RIL-0801 in two areas: axial fuel

relocation and the loss of fuel particles through a rupture opening. The staff accomplished this

by reviewing a wide range of existing data to determine if there are trends and observations

available in the existing data when they are evaluated as a whole and in light of recent findings. 

The purpose of this review was to determine if the previous conclusions about axial fuel

relocation and loss of fuel particles through a rupture opening were accurate.

After completing this review and examining the experimental results for trends and observations

on the phenomenon of fuel relocation, the staff finds that it appears that the previous conclusion

related to axial fuel relocation remains founded. The review of existing data confirms that

fragmentation appears to almost always occur, regardless of burnup and other variables, and

that axial relocation occurs in the presence of appreciable cladding diametral strain. The data

also suggest that rod ballooning is partly inhibited at the location of grid spacers, such that the 

ballooned regions may be pinned by the grid spacers, which in turn could result in choke-points 

for axial fuel relocation. Said in another way, the review of existing data confirms that fuel

fragmentation and subsequent relocation are real physical phenomena expected under LOCA

conditions. Therefore, given that these physical phenomena are expected and with the move to

best-estimate methodologies, accounting for the impact of fuel relocation is appropriate. 

After completing this review and examining the experimental results for trends and observations

on the phenomenon of dispersal, the staff finds that it appears that fuel dispersal during a LOCA 

may not be prevented by a burnup limit alone. Specifically, the existing data suggest that some 

fuel dispersal occurred in all cases in which rod rupture occurred in conjunction with fuel

fragmentation, when the fragments were small enough to get through the rupture opening. In

this conclusion, two factors are at play: (1) fragment size, which the existing data show to

decrease with increasing burnup, and (2) rupture opening size.

While the fuel fragment size was observed to overall decrease with increasing burnup,

fragmentation size is likely to be a function of factors beyond simply burnup level, such as

operating history, fission gas retention, the mechanical properties of the pellet, transient 

temperature, and pressure history. No direct correlation was found between burnup and rupture

opening size. However, increasing burnup results in higher rod internal pressures, which

correlated with wider rupture openings in the database analyzed for this report. Rupture

opening size is not something that ballooning models have been designed to predict. For the

most part, ballooning models are designed to predict the maximum ballooning strain because

the models are designed to inform the conditions of flow blockage (for thermal-hydraulic

analysis) and wall thinning (for oxidation embrittlement analysis) during a LOCA. Given these

observations on fragmentation size and rupture opening size, it could be postulated that, with a

large enough rupture opening, fuel loss cannot be excluded below the current licensed burnup

limit of 62 GWd/MTU.

The full phenomenological assessment and evaluation of the consequences of fuel dispersal for 

plant safety were beyond the scope of the research programs reviewed here. However, this 

report identifies several potential phenomena, including FCI, radiological release, and hydraulic 

and mechanical effects. The staff developed preliminary assessments for these phenomena. 
84

Concerning FCI, the staff concluded that the simultaneous presence of molten fuel and liquid 

coolant during a design-basis LOCA is unlikely, thus preventing a violent FCI. Regarding the 

hydraulic and mechanical consequences of fuel dispersal, the assumptions used to resolve 

GSI-191 do not consider dispersed fuel; therefore, it is unclear whether the consequences of 

fuel particles being entrained in the coolant are bounded by the GSI-191 assumptions. The staff 

has submitted this issue for consideration in NRC’s Generic Issue Program, and has developed 

plans for additional analytical and experimental work. Concerning the radiological 

consequences of fuel dispersal, the current regulatory framework, which assumes that most of 

the volatile fission product inventory is released from the fuel, is largely bounding. In fact, a 

number of best-estimate analyses performed in the United States and abroad show that the 

number of fuel ruptures predicted to occur during a LOCA is quite limited. More research and 

detailed analyses are required to determine the extent of fuel loss, evaluate the identified 

consequences, and ensure that the identified consequences are comprehensive, complete, and 

within the regulatory envelope.
85

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87

27. B. Adroguier, “Synthèse de l’Essai PHEBUS 215 P,” Rapport Technique

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88

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A-1

A. Appendix: Core Damage Distribution Assessment: Methodologies 

and Results

The information presented in this appendix has been selectively extracted from “Nuclear Fuel 

Behavior in Loss-of-Coolant Accident (LOCA) Conditions—State of the Art Report” (OECD 

Report NEA/CSNI/R, (2009) (Ref. A.1). It should be noted that the inputs, assumptions, and 

modeling techniques used in these best-estimate analyses have not been reviewed by the NRC 

staff, and have not been endorsed by NRC. As such, these results should not be used to inform 

any safety assessment without prior evaluation and approval by the NRC.

A.1. A Review of Core Damage Assessment Practices in Europe

In 2000, the European Commission issued a report titled “Fuel Cladding Failure Criteria” 

(Refs. A.2 and A.3). The report presented the results of a collaborative exercise on the 

calculation of the extent of fuel clad failure following a large loss-of-coolant accident (LOCA) in 

light-water reactors. The partners in the collaborative study were Tractebel (Belgium), Institut 

de Protection et de Sûreté Nucléaire (IPSN) and Electricité de France (EdF) (France), Siemens 

and nuclear Reactor Safety Group (GRS) (Germany), Nuclear Research and consultancy Group

(NRG) (Netherlands), Iberdrola (Spain), ERI (Switzerland) and National Nuclear Corporation

(NNC) (United Kingdom). The objectives of the exercise were the following:

• Review the existing clad failure criteria and licensing approaches in each participant’s 

country.

• Form a consensus view on clad failure criteria.

• Determine the effect of the clad failure criteria on the extent of clad failure for a reference 

design in each participant’s country.

The report describes in great detail the methodologies used by the different organizations to 

calculate fuel rod failure. There are basically two types of failure criteria: (1) the empirical 

failure criterion based on NUREG-0630, “Cladding Swelling and Rupture Models for LOCA 

Analysis,” issued April 1980 (Ref°A.4), or (2) a more mechanistic failure criterion based on 

knowledge of creep properties and expected azimuthal temperature differences for the rods in a 

fuel assembly. 

The mechanistic model applies best-estimate creep models for each specific cladding material 

under consideration. The knowledge of rod power, burnup, and thermo-hydraulic events during 

the LOCA is taken as a basis for the failure calculation. All of these calculations rely on 

computer codes. Figure A-1 shows a typical flow chart for a deterministic calculation of fuel 

failure rates. 

In contrast, the empirical NUREG-0630 model ignores the cladding creep. It provides a simple 

relation between the mechanical load on the undeformed cladding and the rupture temperature 

of the cladding.

As an alternative to the deterministic analysis, the participants in the study performed 

probabilistic analyses. These analyses also use the deterministic codes to produce statistical 
A-2

data based on statistical distributions of the input data. However, it may be quite cumbersome 

to run through a code for each randomly selected set of input parameters. Therefore, a subset 

of the input data is run with the code to produce a “response surface” that establishes a 

regression relationship between the input and output data sets. This response surface is 

subsequently used for calculating the output data from random input data. Figure A-2 shows 

the probability of fuel rod failure as a function of rod power for different burnups. It is clear that 

the failure probability increases steeply with burnup. If the failure probabilities are known, or if a 

failure threshold as a function of burnup can be deterministically established, then overviews of 

core damage due to a LOCA can be constructed, as can be seen in Figure A-3. For the 

particular case calculated in Figure A-3, only a few rods at a burnup of 10–15 gigawatt-days per 

metric done of uranium (GWd/MTU) exceed the failure threshold. 

Figure A-1 Determination of fuel failure rate after LOCA with deterministic method 

(Siemens case) (Ref. A.2)
A-3

Figure A-2 Probability for fuel rod failure as a function of rod power (W/cm) at different 

burnups (Siemens case) (Ref. A.2)
A-4

Figure A-3 Power and burnup distribution in a core with failure threshold. Each number 

gives the number of fuel rods at the particular power-burnup combination

(Siemens case). (Ref. A.2)

A comparison of the results of the different participants led to the following conclusions:

• Mechanistic fuel failure models or the nonmechanistic NUREG-0630 fuel failure model 

can be used to evaluate best-estimate fuel failure fraction because they are in good 

agreement with experimental data. The NUREG-0630 model provides a conservative 

approximation of the fuel rod rupture compared to the mechanistic model because the 

mechanistic model allows consideration of the fuel rod ballooning, which is a 

consequence of the cladding creep under stress and temperature. 
A-5

• Of the parameters that cause fuel clad failures, the clad temperature is certainly one of 

the most important. It is thus not surprising that failures happen primarily in the hot rods.

• The other important parameter is the rod internal pressure. The lower value of internal 

pressure in boiling-water reactor (BWR) rods causes fewer fuel failures in comparison 

with a pressurized-water reactor (PWR). The internal rod pressure can also be 

influenced by modifying the volume of the gas plenum of the fuel rods. Consequently, 

the fuel design is of importance. For a given design, the internal pressure increases with 

the burnup. In contrast, the expected peak linear power decreases with burnup much 

faster than the threshold value of fuel failure.

• The calculations show that the combination of both effects causes rods to fail at a 

burnup lower than 35 GWd/MTU, except for mixed-oxide (MOX) rods, which could fail at 

higher burnups.

Despite the fact that all participants used different models and codes, there was agreement on

the best-estimate result of the number of failed rods: no failures at all, as can be seen in Table 

A-1. 

Table A-1 Best-Estimate Analyses of Percentage of Fuel Rod Failures (Ref. A.2)

When judging the results mentioned above, one should keep in mind that the results reported in

(Refs. A.2 and A.3) have generic character only. That is, the underlying thermal-hydraulic 

transient, the cladding material investigated (fresh Zircaloy-4, no hydrogen uptake in cladding, 

and so forth), and the fuel rod design are arbitrarily chosen.

A more recent study by Iberdrola applied the same assumptions as those used for the European 

Commission study to the specific case of the Cofrentes nuclear power plant, a BWR/6 using 

General Electric (GE)-14 fuel and having received an extended power uprate to 111.8 percent

of nominal power (Ref. A.5). As shown in Figure A-4 and summarized in Table A-2, this later 

study showed that best-estimate assumptions resulted in no core damage, while conservative 

estimates resulted in 1.3 percent to 25.6 percent failed fuel rods in the core.
A-6

(a)

(b)

Figure A-4 Power and burnup distribution in BWR GE-14 core with best-estimate (blue)

and conservative (green) failure thresholds, for (a) best-estimate core and (b) 

conservative core (Ref. A.5)

Table A-2 2004 Iberdrola Study Percentage of Failed Bundles, GE-14 Fuel (Ref. A.5)

From the common licensing practice in Germany, it is known that the failed fuel fraction 

sometimes reaches values close to the licensing limit of 10 percent. This is the case if 

conservative thermal-hydraulic boundary conditions and conservative assumptions for the 
A-7

operation mode of the reactor are combined with realistic power histories of the fuel rods in the 

normal operation phase before the LOCA event.

The latest trend in proving the licensing limit in Germany is to account for the power history of 

each fuel rod in the core in order to keep closest track of the best-estimate fuel rod internal 

pressure. This development of the licensing procedure is the consequence of the enhanced 

use of MOX fuel, of cores loaded with power-uprated fuel rods, and the permanent increase of 

the discharge fuel burnup.

A.2. Damage Assessments for a Westinghouse Pressurized-Water 

Reactor

A common practice for assessment of the radiological consequences of a LOCA in the United 

States is to assume that 100 percent of the rods in the core fail. On the other hand, it may be 

instructive to consider what a realistic failure fraction might be under more representative 

conditions. One such study has been reported by Nissley et al. for a four-loop Westinghouse 

PWR (Ref. A.6). The study was done in two parts. The first part dealt with a 3,600-megawatt 

thermal (MWt) reactor with 17-by-17 fuel assemblies and used deterministic calculations. The 

other part concerned a 3,216-MWt reactor with 15-by-15 fuel assemblies and used a statistical 

analysis. 

The first analysis assumed that a full train of emergency core cooling system was lost in order to 

ensure some cladding rupture. It also assumed that peaking factors were 15-percent higher 

than the maximum expected values. On this basis, cladding temperature response was 

calculated with the code WCOBRA/TRAC. A few examples are shown in Figure A-5. On the 

basis of cladding temperature distributions, the internal pressure was calculated and compared 

to an empirically determined rupture temperature versus hoop stress curve for the ZIRLOTM 

cladding used. With the core loading assumed, the fraction of assemblies with ruptured rods 

was about 10 percent.

This assessment of a conservatively assumed base load power distribution was compared to 

design-basis results using a U.S. Nuclear Regulatory Commission (NRC)-approved uncertainty 

methodology. The comparison showed that the realistic assessment resulted in a peak cladding 

temperature (PCT) of 944 degrees Centigrade (C) compared to 1,140 degrees C for the 

design-basis calculations. The equivalent cladding reacted (ECR) at rupture locations was

1.4 percent and 12 percent, respectively, and in nonruptured rods was 0.8 percent and 

6 percent. 

The statistical study used a sampling of 59 separate large-break LOCA transients, each with its 

own combination of randomly sampled uncertainty parameters. According to the statistical 

theory, the most limiting of the 59 cases will bound at least 95 percent of the actual PCT 

distribution, with 95 percent confidence. The goal of the assessment was to examine the extent 

of local oxidation within and away from the ballooned region for the most limiting cases, and to 

assess to what degree the limiting PCT elevation and the cladding rupture elevation were 

coincidental. Table A-3 shows the results for all of the cases above 925 degrees C. Below this 

threshold, oxidation levels are very low. The most limiting PCT case (1,037 degrees C) also 

corresponds to the maximum local oxidation case (2.1 percent). The maximum local oxidation 

occurred at the rupture elevation in this case, but the PCT did not.
A-8

Figure A-5 Peak cladding temperature response for deterministic assessment of extent 

of rupture (Ref. A.6)

Table A-3 WCOBRA/TRAC LOCA Simulation Results for Peak Cladding Temperature

Above 925°C (Ref. A.6)

Nissley et al. summarized the following observations and conclusions (Ref.°A.6):

• The extent of core-wide fuel cladding rupture which would actually be 

expected in a large break LOCA is far less than the 100% assumed by 

many US licensees in their radiological dose calculations. Even 

assuming the worst single failure and a conservative normal operating 
A-9

power shape with linear heat rates 15% higher than predicted, less than 

10% of the rods in the core were estimated to have cladding failures.

• Significant margins exist between realistic estimates of PCT and ECR, 

and those resulting from design basis analyses. Even assuming the 

worst single failure and a conservative normal operating power shape 

with linear heat rates 15% higher than predicted, the PCT was reduced by 

~ 200  ̊C, and the ECR was reduced to negligible amounts compared to 

the design basis analysis results.

• The rupture location tends to have the maximum ECR, due to thinning of 

the cladding and double-sided oxidation.

• PCT frequently occurs away from the rupture location, for plants that have 

a certain LOCA transient response (e.g., 4-loop plants with large dry 

containment designs). 

Finally, Nissley et al. noted that the information presented in these large-break LOCA 

assessments should be interpreted as illustrative and representative. Extent of rupture and 

degree of oxidation are highly dependent on the transient conditions, which are highly 

dependent on plant-specific parameters such as core power, nuclear peaking factors, 

emergency core cooling system capacity, and other factors.

A.3. A German Example of Core Damage Extent Analysis

In a recent presentation, Heins (2004) outlined the current procedure to calculate core damage 

extent in German reactors (Ref. A.7). The German requirements for LOCA analyses differ from 

most other countries in that it is necessary to show that less than 10 percent of the fuel rods 

rupture during a LOCA in order to secure that the radiological consequences will be limited. For 

each PWR, this requirement must be met in each fuel cycle. The analysis must be performed at

both the beginning and the end of the fuel cycle. According to Heins, future analyses will be 

performed with a statistical approach as shown in Figure A-6. The result of a core damage 

extent analysis is shown in Figure A-7.
A-10

Figure A-6 Flow chart for statistical damage assessment for German PWRs (Ref. A.7)

Figure A-7 Map of core damage extent in German PWR after a large-break LOCA (text is 

visible in magnified view) (Ref. A.7)

A further alternative, recently introduced in Germany, is a deterministic calculation of the lifetime 

of each individual fuel rod of the core and a subsequent LOCA transient. Such calculations use 

the material and geometrical data and the power history of the individual fuel rods. Based on 

the calculated actual strain and stress values, the failure/nonfailure is derived according to the 

mechanistic failure criterion implemented in the code. Performed for all fuel rods of the core of 
A-11

a specific reload, this procedure provides the number of failed fuel rods (respective failure rate) 

in a core through a LOCA transient (Figure A-8). 

Figure A-8 Number of failed fuel rods/failure rate during LOCA (Ref. A.7)

A.4. The Failed Fuel Fraction

For PWRs in the United States, Westinghouse studies (Refs. A.3 and A.9) have shown that on 

the order of 200 degrees C reduction in peak cladding temperature is calculated for realistic 

versus licensing-basis LOCA assessments. Further, Westinghouse indicates that less than 

10 percent of the rods in the core will experience cladding rupture (Ref. A.6).

For the fraction of failed fuel calculated in the European community, Westinghouse used the 

results from another study (Ref. A.2). Partners in this collaborative study were Tractebel 

(Belgium), IPSN and EdF (France), Siemens and GRS (Germany), NRG (Netherlands), 

Iberdrola (Spain), ERI (Switzerland), and NNC (United Kingdom). The primary objective of 

Westinghouse’s study was to review the existing cladding failure criteria and licensing 

approaches. Mechanistic and nonmechanistic cladding failure criteria were considered by 

Westinghouse. Mechanistic models had been developed by EdF, GRS, Siemens, and NNC. A 

nonmechanistic (empirical) model for cladding failure was taken from NUREG-0630 (Ref. A.3)
A-12

and also included as part of a methodology for cladding swelling and rupture for use in LOCA 

analysis.

On the basis of the mechanistic analyses, participants in the project determined failure 

thresholds in terms of linear heat generation rate for the fuel. Figure A-9 shows the bestestimate failure threshold. A similarly estimated conservative failure threshold was about 50–

150 W/cm lower, depending on the participant. 

Figure A-9 Best-estimate failure thresholds for PWR fuel in a large-break LOCA

(Ref. A.2)

The failed fuel fraction can be calculated based on the failure thresholds. In the best-estimate 

analyses, only one of the participants arrived at a non-zero value. More conservative analyses

obtained failed fuel fractions between 0 and 16.6 percent. These values were significantly 

below the 100 percent normally applied by the different participants (including the United 

States). Therefore, the participants suggested as a common position that in licensing 

calculations a failed fuel fraction of 33 percent should be applied for reactor designs with coldleg injection only. For reactors with injection into both hot and cold legs, a failed fuel fraction of 

10 percent was supported by the participants, which is the same as that adopted in the German 

licensing methodology (Ref. A.10).

A.5. References

A.1. “Nuclear Fuel Behavior in Loss-of-Coolant Accident (LOCA) Conditions—State of the Art 

Report,” OECD Report NEA/CSNI/R(2009)15.

A.2. “Fuel Cladding Failure Criteria,” European Commission, EUR 19256, 1999.

A.3. P.A.W. Bratby, L.M.C. Dutton, and L. Sutherland, “Fuel Cladding Failures Following a 

Large LOCA,” NNC Report C5665/TR/006, Issue 01, May 1999.

A.4. D.A. Powers and R.O. Meyer, “Cladding Swelling and Rupture Models for LOCA 

Analysis,” NUREG-0630, U.S. Nuclear Regulatory Commission, Washington, DC, 

April 1980. ADAMS Accession No. ML053490337.
A-13

A.5. R. O. Moreno, I. G. Cabezón, and P.J. García Sedano, “Fuel Cladding Failures 

Following a Large LOCA,” Paper 310, 11th International Topical Meeting on Nuclear 

Reactor Thermal-Hydraulics (NURETH-11), Avignon, France, October 2–6, 2005.

A.6. M.E. Nissley, C. Frepoli, and K. Ohkawa, “Realistic Assessment of Fuel Rod Behavior 

Under Large-Break LOCA Conditions,” in “Proceedings of the Nuclear Fuels Sessions of 

the 2004 Nuclear Safety Research Conference,” NUREG/CP-0192, U.S. Nuclear 

Regulatory Commission, Washington, DC, October 2005, pp. 231–273. ADAMS 

Accession No. ML052980524.

A.7. L. Heins, “Core Damage Extent Analysis To Fulfill an Additional LOCA Criterion,” Special 

Expert Group on Fuel Safety Margins (SEGFSM) Topical Meeting on LOCA Issues, 

2004, Argonne National Laboratory, United States.

A.8. M.E. Nissley, C. Frepoli, and K. Ohkawa, “Realistic High Burnup UO2 Response to a 

LOCA in a PWR,” SEGFSM Topical Meeting on LOCA Issues, Argonne National 

Laboratory, May 25–26, 2004, Nuclear Energy Agency Report NEA/CSNI/R(2004)19.

A.9. C. Frepoli, M.E. Nissley, and K. Ohkawa, “Comparison of High and Low Burnup UO2 

Fuel Response to a LOCA Using a Non-Parametric Statistical Method,” Paper 210, 11th 

International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-11), 

Avignon, France, October 2–6, 2005.

A.10. J. Keusenhoff, “Proposed Criteria for the Evaluation of Fuel Element Damage Due to 

Loss of Coolant in Pressurized Water Reactors,” Gesellschaft für Reaktorsicherheit 

Report GRS-A-1231, June 1986. The NRC’s English translation of this German report is 

available at ADAMS Accession No. ML111741202.




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-2121 Fuel Fragmentation, Relocation, and Dispersal 

During the Loss-of-Coolant Accident