NUREG–2214

Managing Aging Processes

In Storage (MAPS) Report

Draft Report for Comment

Office of Nuclear Materials Safety and Safeguards

Managing Aging Processes 

In Storage (MAPS) Report

Draft Report for Comment

Manuscript Completed: October 2017 

Date Published: October 2017

Office of Nuclear Materials Safety and Safeguards

NUREG-2214
COMMENTS ON DRAFT REPORT 

Any interested party may submit comments on this report for consideration by the staff of the 

U.S. Nuclear Regulatory Commission (NRC). Comments may be accompanied by additional 

relevant information or supporting data. Please specify the report number NUREG–2214 in your 

comments, and send them by the end of the comment period specified in the Federal Register 

notice announcing the availability of this report. Addresses: You may submit comments by 

any one of the following methods. Please include Docket ID NRC-2016-0238 in the subject line 

of your comments. Comments submitted in writing or in electronic form will be posted on the 

NRC Web site and on the Federal rulemaking Web site http://www.regulations.gov. 

Federal Rulemaking Web site: Go to http://www.regulations.gov and search for 

documents filed under Docket ID NRC-2016-0238. Address questions about NRC dockets to 

Carol Gallagher at 301-415-3463 or by e-mail at Carol.Gallagher@nrc.gov. 

Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch, 

Division of Administrative Services, Office of Administration, Mail Stop: TWFN-8-D-36M, 

U.S. Nuclear Regulatory Commission, Washington, DC 20555 0001. 

For any questions about the material in this report, please contact: John Wise, Senior Materials 

Engineer, 301-415-8085 or by e-mail at John.Wise@nrc.gov. 

Please be aware that any comments that you submit to the NRC will be considered a public 

record and entered into the Agencywide Documents Access and Management System. Do not 

provide information you would not want to be publicly available.
iii

1 ABSTRACT

2 This Managing Aging Processes in Storage (MAPS) Report provides guidance for the 

3 U.S. Nuclear Regulatory Commission (NRC) technical reviewer. It establishes a technical basis 

4 for the safety review of renewal applications for specific licenses of independent spent fuel 

5 storage installations and Certificates of Compliance for dry storage systems, as codified in 

6 Title 10 of the Code of Federal Regulations Part 72, “Licensing Requirements for the 

7 Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and

8 Reactor-Related Greater Than Class C Waste.” 

9 The MAPS Report evaluates known aging degradation mechanisms to determine if they could

10 affect the ability of dry storage system components to fulfill their safety functions in the 20- to 

11 60-year period of extended operation. The guidance also provides examples of aging 

12 management programs that are considered generically acceptable to address the credible aging 

13 mechanisms to ensure that the design bases of dry storage systems will be maintained. An 

14 applicant for a renewed license or Certificate of Compliance may reference the information in 

15 the MAPS Report to support its aging management review and proposed aging 

16 management programs.

17 Paperwork Reduction Act

18 This NUREG provides guidance for implementing the mandatory information collections in 

19 10 CFR Part 72 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et. 

20 seq.). These information collections were approved by the Office of Management and Budget 

21 (OMB) under control number 3150-0132. Send comments regarding this information collection 

22 to the Information Services Branch, U.S. Nuclear Regulatory Commission, Washington, DC 

23 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of 

24 Information and Regulatory Affairs, NEOB-10202, (3150 -0132) Office of Management and 

25 Budget, Washington, DC 20503.

26 Public Protection Notification

27 The NRC may not conduct or sponsor, and a person is not required to respond to, a collection 

28 of information unless the document requesting or requiring the collection displays a currently 

29 valid OMB control number. 

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v 

1 CONTENTS 

2 

3 ABSTRACT............................................................................................................................... iii

4 LIST OF FIGURES ...................................................................................................................ix

5 LIST OF TABLES......................................................................................................................xi

6 ABBREVIATIONS AND ACRONYMS .................................................................................... xiii

7 1 INTRODUCTION......................................................................................................... 1-1

8 1.1 Purpose and use of the MAPS Report ..............................................................1-1

9 1.2 Scope of report .................................................................................................1-2

10 1.3 Acknowledgments.............................................................................................1-3

11 1.4 References .......................................................................................................1-4

12 2 DEFINITIONS.............................................................................................................. 2-1

13 2.1 Materials ...........................................................................................................2-1

14 2.2 Environments....................................................................................................2-3

15 2.3 Aging mechanisms............................................................................................2-4

16 2.4 Aging effects.....................................................................................................2-9

17 3 EVALUATION OF AGING MECHANISMS ................................................................. 3-1

18 3.1 Introduction.......................................................................................................3-1

19 3.2 Casks and internals ..........................................................................................3-7

20 3.2.1 Steel (carbon, low-alloy, high-strength low-alloy)...................................3-7

21 3.2.1.1 General corrosion................................................................... 3-8

22 3.2.1.2 Pitting and crevice corrosion................................................. 3-10

23 3.2.1.3 Galvanic corrosion................................................................ 3-11

24 3.2.1.4 Microbiologically influenced corrosion................................... 3-11

25 3.2.1.5 Stress corrosion cracking ..................................................... 3-13

26 3.2.1.6 Creep ................................................................................... 3-13

27 3.2.1.7 Fatigue ................................................................................. 3-14

28 3.2.1.8 Thermal aging ...................................................................... 3-15

29 3.2.1.9 Radiation embrittlement........................................................ 3-16

30 3.2.1.10 Stress relaxation................................................................... 3-17

31 3.2.1.11 Wear..................................................................................... 3-17

32 3.2.2 Stainless steel ..................................................................................... 3-17

33 3.2.2.1 General corrosion................................................................. 3-18

34 3.2.2.2 Pitting and crevice corrosion................................................. 3-18

35 3.2.2.3 Galvanic corrosion................................................................ 3-19

36 3.2.2.4 Microbiologically influenced corrosion................................... 3-19

37 3.2.2.5 Stress corrosion cracking ..................................................... 3-20

38 3.2.2.6 Creep ................................................................................... 3-22

39 3.2.2.7 Fatigue ................................................................................. 3-23

40 3.2.2.8 Thermal aging ...................................................................... 3-23

41 3.2.2.9 Radiation embrittlement........................................................ 3-25

42 3.2.2.10 Stress relaxation................................................................... 3-26

43 3.2.2.11 Wear..................................................................................... 3-26
vi

1 3.2.3 Aluminum alloys .................................................................................. 3-26

2 3.2.3.1 General corrosion................................................................. 3-26

3 3.2.3.2 Pitting and crevice corrosion................................................. 3-27

4 3.2.3.3 Galvanic corrosion................................................................ 3-28

5 3.2.3.4 Microbiologically influenced corrosion................................... 3-29

6 3.2.3.5 Creep ................................................................................... 3-29

7 3.2.3.6 Fatigue ................................................................................. 3-31

8 3.2.3.7 Thermal aging ...................................................................... 3-31

9 3.2.3.8 Radiation embrittlement........................................................ 3-32

10 3.2.4 Nickel alloys ........................................................................................ 3-33

11 3.2.4.1 General corrosion................................................................. 3-33

12 3.2.4.2 Pitting and crevice corrosion................................................. 3-33

13 3.2.4.3 Microbiologically influenced corrosion................................... 3-34

14 3.2.4.4 Stress corrosion cracking ..................................................... 3-34

15 3.2.4.5 Fatigue ................................................................................. 3-34

16 3.2.4.6 Radiation embrittlement........................................................ 3-35

17 3.2.4.7 Stress relaxation................................................................... 3-35

18 3.2.4.8 Wear..................................................................................... 3-35

19 3.2.5 Copper alloys ...................................................................................... 3-36

20 3.2.5.1 General corrosion................................................................. 3-36

21 3.2.5.2 Pitting and crevice corrosion................................................. 3-36

22 3.2.5.3 Microbiologically influenced corrosion................................... 3-37

23 3.2.5.4 Radiation embrittlement........................................................ 3-37

24 3.2.6 Lead.................................................................................................... 3-37

25 3.2.7 Depleted uranium................................................................................ 3-37

26 3.2.8 Coatings.............................................................................................. 3-38

27 3.2.9 References.......................................................................................... 3-38

28 3.3 Neutron shielding materials............................................................................. 3-49

29 3.3.1 Neutron-shielding materials................................................................. 3-49

30 3.3.1.1 Boron depletion (borated materials)...................................... 3-49

31 3.3.1.2 Thermal aging ...................................................................... 3-50

32 3.3.1.3 Radiation embrittlement........................................................ 3-50

33 3.3.2 References.......................................................................................... 3-51

34 3.4 Neutron poison materials ................................................................................ 3-53

35 3.4.1 Borated stainless steel ........................................................................ 3-53

36 3.4.1.1 Boron depletion .................................................................... 3-54

37 3.4.1.2 Creep ................................................................................... 3-54

38 3.4.1.3 Thermal aging ...................................................................... 3-55

39 3.4.1.4 Radiation embrittlement........................................................ 3-55

40 3.4.2 Borated aluminum alloys and aluminum-based composites ................ 3-55

41 3.4.2.1 General corrosion................................................................. 3-56

42 3.4.2.2 Galvanic corrosion................................................................ 3-56

43 3.4.2.3 Wet corrosion and blistering ................................................. 3-56

44 3.4.2.4 Boron depletion .................................................................... 3-57

45 3.4.2.5 Creep ................................................................................... 3-57

46 3.4.2.6 Thermal aging ...................................................................... 3-58

47 3.4.2.7 Radiation embrittlement........................................................ 3-58

48 3.4.3 References.......................................................................................... 3-59

49 3.5 Concrete overpacks, support pads, and ceramic fiber insulation..................... 3-61

50 3.5.1 Concrete.............................................................................................. 3-62

51 3.5.1.1 Freeze and thaw................................................................... 3-62
vii

1 3.5.1.2 Creep ................................................................................... 3-63

2 3.5.1.3 Reaction with aggregates ..................................................... 3-63

3 3.5.1.4 Differential settlement........................................................... 3-64

4 3.5.1.5 Aggressive chemical attack .................................................. 3-65

5 3.5.1.6 Corrosion of reinforcing steel................................................ 3-67

6 3.5.1.7 Shrinkage ............................................................................. 3-68

7 3.5.1.8 Leaching of calcium hydroxide.............................................. 3-69

8 3.5.1.9 Radiation damage ................................................................ 3-70

9 3.5.1.10 Fatigue ................................................................................. 3-70

10 3.5.1.11 Dehydration at high temperature .......................................... 3-71

11 3.5.1.12 Microbiological degradation .................................................. 3-72

12 3.5.1.13 Delayed ettringite formation.................................................. 3-73

13 3.5.1.14 Salt scaling........................................................................... 3-74

14 3.5.2 Ceramic fiber insulation....................................................................... 3-74

15 3.5.2.1 Radiation damage ................................................................ 3-75

16 3.5.2.2 Moisture absorption.............................................................. 3-76

17 3.5.3 References.......................................................................................... 3-76

18 3.6 Spent fuel assemblies..................................................................................... 3-85

19 3.6.1 Cladding materials............................................................................... 3-86

20 3.6.1.1 Hydride reorientation (high burnup fuel)................................ 3-86

21 3.6.1.2 Delayed hydride cracking (high burnup fuel)......................... 3-89

22 3.6.1.3 Thermal creep (high burnup fuel).......................................... 3-91

23 3.6.1.4 Low-temperature creep (high burnup fuel)............................ 3-93

24 3.6.1.5 Mechanical overload (high burnup fuel)................................ 3-94

25 3.6.1.6 Oxidation .............................................................................. 3-96

26 3.6.1.7 Pitting corrosion.................................................................... 3-96

27 3.6.1.8 Galvanic corrosion................................................................ 3-97

28 3.6.1.9 Stress corrosion cracking ..................................................... 3-98

29 3.6.1.10 Radiation embrittlement........................................................ 3-99

30 3.6.1.11 Fatigue ................................................................................. 3-99

31 3.6.2 Assembly hardware materials............................................................ 3-101

32 3.6.2.1 Creep ................................................................................. 3-101

33 3.6.2.2 Hydriding ............................................................................ 3-102

34 3.6.2.3 General corrosion............................................................... 3-102

35 3.6.2.4 Stress corrosion cracking ................................................... 3-103

36 3.6.2.5 Radiation embrittlement...................................................... 3-104

37 3.6.2.6 Fatigue ............................................................................... 3-104

38 3.6.3 References........................................................................................ 3-104

39 4 ANALYSIS OF DRY STORAGE SYSTEMS AND SPENT FUEL ASSEMBLIES ......... 4-1

40 4.1 Introduction.......................................................................................................4-1

4.2 NUHOMS® 41 systems: standardized and standardized advanced.......................4-3

42 4.2.1 System description ................................................................................4-3

43 4.2.2 Dry shielded canister.............................................................................4-3

44 4.2.3 Horizontal storage module.....................................................................4-6

45 4.2.4 Transfer cask ........................................................................................4-8

46 4.3 HI-STORM 100 and HI-STAR 100 systems .................................................... 4-91

47 4.3.1 System description .............................................................................. 4-91

48 4.3.2 Multipurpose canister .......................................................................... 4-91

49 4.3.3 HI-STORM concrete overpack............................................................. 4-94

50 4.3.4 HI-STAR metal overpack..................................................................... 4-96
viii

1 4.3.5 Transfer cask ...................................................................................... 4-99

2 4.4 TN-32 and TN-68 systems ............................................................................ 4-151

3 4.4.1 System description ............................................................................ 4-151

4 4.4.2 Bolted metal cask .............................................................................. 4-151

5 4.5 NAC International Systems ........................................................................... 4-169

6 4.5.1 System description ............................................................................ 4-169

7 4.5.2 NAC-UMS ......................................................................................... 4-169

8 4.5.3 NAC-MPC ......................................................................................... 4-170

9 4.5.4 MAGNASTOR ................................................................................... 4-171

4.6 FuelSolutionsTM 10 storage system.................................................................... 4-259

11 4.6.1 System description ............................................................................ 4-259

12 4.6.2 W21 and W74 canisters .................................................................... 4-259

13 4.6.3 W150 storage cask............................................................................ 4-260

14 4.6.4 W100 transfer cask ........................................................................... 4-262

15 4.7 Concrete pad ................................................................................................ 4-291

16 4.8 Spent fuel assemblies................................................................................... 4-297

17 4.8.1 Spent fuel assembly description ........................................................ 4-297

18 4.8.2 Fuel cladding and assembly hardware .............................................. 4-297

19 4.9 References ................................................................................................... 4-303

20 5 TIME-LIMITED AGING ANALYSES ........................................................................... 5-1

21 5.1 Introduction.......................................................................................................5-1

22 5.2 Review..............................................................................................................5-1

6 EXAMPLE AGING MANAGEMENT PROGRAMS ...................................................... 6-1

23

6.1 Introduction.......................................................................................................6-1

24

6.2 Alternative approaches .....................................................................................6-1

25

6.3 Learning aging management ............................................................................6-2

26

6.4 References .......................................................................................................6-3

27

6.5 Localized corrosion and stress corrosion cracking of welded stainless 

28

steel dry storage canisters ................................................................................6-5

29

6.6 Reinforced concrete structures ....................................................................... 6-17

30

6.7 External surfaces monitoring of metallic components...................................... 6-29

31

6.8 Ventilation systems......................................................................................... 6-35

32

6.9 Bolted cask seal leakage monitoring............................................................... 6-43

33

6.10 Transfer casks ................................................................................................ 6-53

34

6.11 High-burnup fuel monitoring and assessment ................................................. 6-59

35

36

5.3 References........................................................................................................5-2
ix

1 LIST OF FIGURES

2 Figure 1-1 Use of the MAPS Report in the renewal process ............................................. 1-3

3 Figure 4-1 NUHOMS dry storage system.......................................................................... 4-4

4 Figure 4-2 NUHOMS-24PT2 DSC assembly–spacer disk design...................................... 4-4

5 Figure 4-3 NUHOMS-32PT DSC assembly–tube or plate design...................................... 4-5

6 Figure 4-4 Pressure and confinement boundaries for NUHOMS-32PT DSC..................... 4-5

7 Figure 4-5 Air flow diagram for a typical standardized HSM design................................... 4-7

8 Figure 4-6 Advanced NUHOMS horizontal storage module .............................................. 4-7

9 Figure 4-7 Side elevation and end view of the DSC support structure............................... 4-8

10 Figure 4-8 OS197L transfer cask ...................................................................................... 4-9

11 Figure 4-9 HI-STORM 100 and HI-STAR 100 systems ................................................... 4-92

12 Figure 4-10 Cross section elevation view of MPC ............................................................. 4-92

13 Figure 4-11 Cross sectional views of different MPC designs............................................. 4-93

14 Figure 4-12 MPC confinement boundary........................................................................... 4-95

15 Figure 4-13 Cross sectional views of the HI-STORM 100 and 100S overpacks 

16 with an MPC inserted..................................................................................... 4-95

17 Figure 4-14 HI-STAR 100 overpack elevation view........................................................... 4-98

18 Figure 4-15 HI-STAR 100 overpack cross sectional view.................................................. 4-98

19 Figure 4-16 Cross sectional views of the HI-TRAC 125 transfer cask with pool lid 

20 and transfer lid ............................................................................................. 4-100

21 Figure 4-17 Components of the TN-32 storage cask....................................................... 4-152

22 Figure 4-18 TN-68 cask confinement boundary components .......................................... 4-152

23 Figure 4-19 TN-32 cask seal pressure-monitoring system .............................................. 4-153

24 Figure 4-20 Radial cross section of TN-32 cask showing basket, basket rails, and 

25 gamma and neutron shields......................................................................... 4-154

26 Figure 4-21 TN-32 cask shielding configuration .............................................................. 4-155

27 Figure 4-22 NAC-UMS.................................................................................................... 4-172

28 Figure 4-23 NAC-UMS transportable storage canister for PWR fuel ....... ...................... 4-173

29 Figure 4-24 NAC-UMS VCC and transfer cask arrangement .......................................... 4-173

30 Figure 4-25 NAC MAGNASTOR TSC and concrete cask................................................ 4-174

31 Figure 4-26 NAC MAGNASTOR concrete cask .............................................................. 4-174

Figure 4-27 Typical FuelSolutions 32 TM W21 and W74 canisters ................ ...................... 4-260

Figure 4-28 FuelSolutions 33 TM W150 storage cask............................................................. 4-261

Figure 4-29 FuelSolutions 34 TM W100 transfer cask ............................................................ 4-263

35 Figure 4-30 Typical pressurized-water reactor fuel assembly.......................................... 4-298

36 Figure 4-31 Boiling-water reactor GE14 fuel assembly ................................................... 4-299

37

xi

1 LIST OF TABLES

2 Table 2-1 Use of terms for materials .................................................................................. 2-1

3 Table 2-2 Use of terms for environments ........................................................................... 2-3

4 Table 2-3 Use of terms for aging mechanisms ................................................................... 2-4

5 Table 2-4 Use of terms for aging effects............................................................................. 2-9

6 Table 3-1 Environment abbreviations................................................................................. 3-1

7 Table 3-2 Casks and internals aging mechanism evaluations ............................................ 3-2

8 Table 3-3 Neutron shielding materials aging mechanism evaluations................................. 3-4

9 Table 3-4 Neutron poison materials aging mechanism evaluations .................................... 3-4

10 Table 3-5 Concrete overpacks, support pads, and ceramic fiber insulation aging 

11 mechanism evaluations...................................................................................... 3-5

12 Table 3-6 Spent fuel assembly aging mechanism evaluations ........................................... 3-6

13 Table 4-1 Evaluated storage system designs..................................................................... 4-1

14 Table 4-2 Standardized NUHOMS dry shielded canister.................................................. 4-10

15 Table 4-3 Standardized Advanced NUHOMS dry shielded canister ................................. 4-22

16 Table 4-4 Standardized NUHOMS horizontal storage module.......................................... 4-33

17 Table 4-5 Standardized Advanced NUHOMS horizontal storage module......................... 4-52

18 Table 4-6 NUHOMS transfer cask.................................................................................... 4-69

19 Table 4-7 HI-STORM / HI-STAR multipurpose canister.................................................. 4-101

20 Table 4-8 HI-STORM 100 overpack ............................................................................... 4-110

21 Table 4-9 HI-STAR 100 overpack .................................................................................. 4-129

22 Table 4-10 HI-TRAC transfer cask ................................................................................... 4-139

23 Table 4-11 TN bolted metal casks.................................................................................... 4-156

24 Table 4-12 NAC-UMS transportable storage canister....................................................... 4-175

25 Table 4-13 NAC-UMS vertical concrete cask ................................................................... 4-184

26 Table 4-14 NAC-UMS transfer cask ................................................................................. 4-194

27 Table 4-15 NAC-MPC transportable storage canister....................................................... 4-200

28 Table 4-16 NAC-MPC vertical concrete cask ................................................................... 4-213

29 Table 4-17 NAC-MPC transfer cask ................................................................................. 4-223

30 Table 4-18 MAGNASTOR transportable storage canister ................................................ 4-228

31 Table 4-19 MAGNASTOR concrete cask ......................................................................... 4-241

32 Table 4-20 MAGNASTOR transfer cask........................................................................... 4-250

Table 4-21 FuelSolutionsTM 33 canister .................................................................................. 4-264

Table 4-22 FuelSolutionsTM 34 storage cask ......................................................................... 4-274

Table 4-23 FuelSolutionsTM 35 transfer cask......................................................................... 4-282

36 Table 4-24 Concrete pad.................................................................................................. 4-292

37 Table 4-25 Spent fuel assemblies .................................................................................... 4-300

38 Table 5-1 Examples of fatigue analyses contained within storage system design bases.... 5-2

39 Table 6-1 Example aging management programs.............................................................. 6-1

40 Table 6-2 Example aging management program for Localized Corrosion And Stress 

41 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters............... 6-6

42 Table 6-3 Example aging management program for Reinforced Concrete Structures ...... 6-18

43 Table 6-4 Example aging management program for External Surfaces Monitoring 

44 Of Metallic Components ................................................................................... 6-30

45 Table 6-5 Example aging management program for Ventilation Systems......................... 6-36

46 Table 6-6 Example aging management program for Bolted Cask Seal Leakage 

47 Monitoring ........................................................................................................ 6-44

48 Table 6-7 Example aging management program for Transfer Casks................................ 6-54

49 Table 6-8 Example aging management program for High-Burnup Fuel Monitoring 

50 and Assessment............................................................................................... 6-60

xiii

1 ABBREVIATIONS AND ACRONYMS

ACI American Concrete Institute

ADAMS Agencywide Documents Access and Management System

AISC American Institute of Steel Construction

AMP aging management program

AMR aging management review

ANSI American National Standards Institute

ASCE American Society of Civil Engineers

ASME American Society of Mechanical Engineers

ASR alkali-silica reaction

B&PV boiler and pressure vessel 

BWR boiling-water reactor

CAP corrective action program

CFR Code of Federal Regulations

CoC Certificate of Compliance

CISCC chloride-induced stress corrosion cracking 

CWSR cold worked stress relieved

DBTT ductile-to-brittle transition temperature

DEF delayed ettringite formation

DHC delayed hydride cracking

DOE U.S. Department of Energy

DSC dry shielded canister

DSS dry storage system

EPRI Electric Power Research Institute

FSAR final safety analysis report

HBU high burnup

HDRP

HSM

HBU Dry Storage Cask Research and Development Project

horizontal storage module

IFBA integral fuel burnable absorber

IN Information Notice

ISFSI independent spent fuel storage installation

ISG Interim Staff Guidance

MAPS Managing Aging Processes in Storage

MIC microbiologically influenced corrosion

MPC multipurpose canister

NDE nondestructive examination

NRC U.S. Nuclear Regulatory Commission

PCMI pellet-to-cladding mechanical interaction

PWR pressurized-water reactor

QA quality assurance
xiv

RIA reactivity-initiated accident

RXA recrystallized annealed

SCC stress corrosion cracking

SNF spent nuclear fuel

SSC structure, system, and component

TC transfer cask

TLAA time-limited aging analysis

TMI Three Mile Island

TN Transnuclear Inc.

TS technical specification(s)

TSC transportable storage canister

VCC ventilated concrete cask

VVM vertical ventilated module

Units of Measure

atm atmosphere (pressure)

C Celsius

dpa displacements per atom (radiation damage)

F Fahrenheit

g gram

gal gallon

GWd/MTU gigawatt-days per metric ton of uranium

in inch

K Kelvin

kGy kilogray (absorbed radiation dose)

ksi 1,000 pounds per square inch

L liter

mg milligram, 0.001 grams

MPa megapascal, 1x106 pascals (stress)

MeV megaelectron-volt, 1x106 electron-volts (energy)

mil one-thousandth of an inch, 0.001 inch

mpy mils per year 

mm millimeter, 0.001 meter

n neutrons

oz ounce

ppm parts per million

psi Pounds per square inch

rad (unit of absorbed radiation dose)

sec second

μm micrometer, 1x10-6 meter

yr year

1
1-1 

1 1 INTRODUCTION

2 1.1 Purpose and use of the MAPS Report

3 The U.S. Nuclear Regulatory Commission (NRC) licenses the storage of spent nuclear fuel 

4 (SNF) in dry storage systems (DSSs) under the regulations of Title 10 of the Code of Federal 

5 Regulations (10 CFR) Part 72, “Licensing Requirements for the Independent Storage of Spent 

6 Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C 

7 Waste.” To date, licenses for specific independent spent fuel storage installations (ISFSIs) or 

8 Certificates of Compliance (CoCs) for DSSs have been issued for initial terms of 20 years, 

9 although regulations currently allow an initial 40-year storage period. Licenses and CoCs can

10 be renewed for additional terms not to exceed 40 years. In accordance with 10 CFR 72.42, 

11 “Duration of License; Renewal,” and 10 CFR 72.240, “Conditions for Spent Fuel Storage Cask 

12 Renewal,” renewal applications must include:

13 i. time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and 

14 components (SSCs) important to safety will continue to perform their intended function 

15 for the requested period of extended operation

16 ii. aging management programs (AMPs) for management of issues associated with aging 

17 that could adversely affect SSCs important to safety 

18 NUREG–1927, Revision 1, “Standard Review Plan for Renewal of Specific Licenses and 

19 Certificates of Compliance for Dry Storage of Spent Nuclear Fuel,” provides guidance for the 

20 staff’s review of TLAAs and AMPs (NRC, 2016).

21 This Managing Aging Processes in Storage (MAPS) Report is a technical basis document that 

22 provides additional guidance to NRC staff to improve the effectiveness and efficiency of the 

23 renewal process for the dry storage of SNF. The MAPS Report provides a generic evaluation of 

24 the aging mechanisms that have the potential to challenge the ability of DSS SSCs to fulfill their 

25 important-to-safety functions. The MAPS Report also describes acceptable generic AMPs that 

26 an applicant may use to maintain the approved design basis of its storage system during the 

period of extended operation (from 20 to 60 years of storage1 27 ). An applicant for a renewed 

28 license or CoC may reference the information in the MAPS Report to support its design-specific 

29 aging management review (AMR) and proposed AMPs. 

30 The content of the report is as follows.

31 • Chapter 1 briefly describes how the MAPS Report is to be used by the NRC staff. 

32 • Chapter 2 defines the terms that are used throughout this report, including descriptions 

33 of materials, environments, aging mechanisms, and aging effects (the manifestations of 

34 aging mechanisms by degraded conditions or performance). 

35 • Chapter 3 evaluates the aging mechanisms that may challenge the ability of SSCs to 

36 fulfill their important-to-safety function(s). Those mechanisms that are shown to have

37 the potential to adversely affect an important-to-safety function in the 60-year timeframe 

1Because the NRC has granted, to date, initial storage licenses and CoCs for 20 years only, the MAPS Report 

considers the effects of aging for 40 years beyond the initial 20-year term (or 60 years total).
1-2 

1 are identified as “credible.” This chapter provides the technical bases for the aging 

2 management recommendations that appear in the AMR tables and AMPs in Chapters 4 

3 and 5, respectively.

4 • Chapter 4 describes selected DSS designs and provides AMR tables for those designs. 

5 The AMR tables identify the aging mechanisms and effects that could challenge the 

6 capability of each SSC to fulfill its important-to-safety function(s) in the 20- to 60-year 

7 period of extended operation. For those credible aging effects, the AMR tables 

8 recommend aging management approaches (i.e., AMPs, TLAAs, or other analyses). 

9 • Chapter 5 provides guidance for identifying and evaluating time-limited aging analyses

10 • Chapter 6 contains example AMPs that an applicant may use to address the credible 

11 aging effects identified in the AMR tables. 

12 Figure 1-1 provides a flowchart that shows how the guidance in the MAPS Report supports the 

13 renewal process.

14 The MAPS Report increases the efficiency of the licensing process by reducing redundant 

15 reviews of the same topic. If an applicant credits the information in the MAPS Report in the 

16 renewal application, the staff should ensure that the applicant demonstrates that the design 

17 features, environmental conditions, and operating experience for the subject ISFSI or DSS are 

18 bounded by those evaluated in the MAPS Report. Otherwise, the staff should ensure that the 

19 applicant revises its AMR and AMPs, as appropriate, to address the design or operating 

20 parameters applicable to its facility or storage system. 

21 The MAPS Report contains one acceptable method to identify and manage credible aging 

22 mechanisms and effects for specific-license and CoC renewals. An applicant may propose 

23 alternatives for staff review. As such, the staff should not use the MAPS Report as a 

24 requirement. Nevertheless, its use should facilitate both the preparation of a specific license or 

25 CoC renewal application by an applicant and a timely, consistent review by the NRC staff.

26 Finally, the MAPS Report does not address the scoping of SSCs for specific-license or CoC 

27 renewal; this is addressed in Chapter 2 of NUREG–1927, Revision 1. Although the MAPS 

28 Report generically addresses SSCs for several storage system designs, scoping is design and 

29 license specific. The inclusion of a certain SSC in the MAPS Report does not necessarily imply 

30 that the particular SSC is within the scope of renewal for all ISFSIs or DSSs. Conversely, the 

31 omission of a certain SSC in the MAPS Report does not imply that the particular SSC is not 

32 within the scope of renewal for any ISFSI or DSS.

33 1.2 Scope of report

34 The MAPS Report addresses the aging mechanisms and effects associated with the following 

35 DSS designs: Standardized and Advanced NUHOMS, HI-STORM 100, HI-STAR 100, TN-32 

36 and -68, the NAC UMS, MPC, and MAGNASTOR systems, and the FuelSolutions storage 

37 system. The selection of these systems addresses near-term renewal applications and a 

38 variety of storage system designs. Although this report was written to specifically address those 

39 designs, the staff may consider the general applicability of this guidance to other designs. 
1-3

Figure 1-1 Use of the MAPS Report in the renewal process

1 1.3 Acknowledgments

2 The NRC would like to acknowledge the contributions of the staff at the Center for Nuclear 

Waste Regulatory Analyses at the Southwest Research Institute® 3 for its role in developing the 

4 technical bases for the aging evaluations in this report. This includes the evaluations of the 

5 aging mechanisms in Chapter 3 and the associated AMR tables in Chapter 4. The staff at the 

6 CNWRA also assisted in the development of the introductory material and combining all 

7 portions of this document into a single, cohesive report. 

8 The NRC also would like to acknowledge the contributions of Argonne National Laboratory in 

9 support of the U.S. Department of Energy (DOE) Used Fuel Disposition Campaign. Portions of 

10 the storage system descriptions in Chapter 4 of the MAPS Report were taken from the 

11 information contained in Chapter V of the Argonne/DOE report, “Managing Aging Effects on Dry 

12 Cask Storage Systems for Extended Long-Term Storage and Transportation of Used Fuel” 

13 (Chopra et al., 2014). 

14

Scoping Evaluation

• Identify SSCs within the Scope of Renewal 

(see NUREG–1927, Section 2)

Aging Management Review

• Define materials and environments for each SSC (MAPS Chapter 2)

• Evaluate potential aging mechanisms and effects (MAPS Chapter 3)

• Tabulate aging evaluation results for each SSC (MAPS Chapter 4)

• Identify activities to manage credible aging 

mechanisms and effects (MAPS Chapter 4)

Time-Limited Aging Analyses or Other 

Supporting Analyses

• Evaluate time-limited aging analyses

• Provide other supporting analyses to 

demonstrate that SCCs will continue to 

perform their intended functions

(see NUREG–1927; MAPS Chapter 5)

Aging Management Programs

• Describe programs for the management 

of aging issues

(MAPS Chapter 6)
1-4 

1 1.4 References

2 Chopra, O., D. Diercks, R. Fabian, Z. Han, and Y. Liu. “Managing Aging Effects on Dry Cask 

3 Storage Systems for Extended Long-Term Storage and Transportation of Used Fuel.” 

4 FCRD–UFD–2014–000476. ANL–13/15, Rev. 2. Washington, DC.: U.S. Department of 

5 Energy. 2014. 

6 NRC. NUREG–1927, “Standard Review Plan for Renewal of Specific Licenses and Certificates 

7 of Compliance for Dry Storage of Spent Nuclear Fuel.” Revision 1. Washington, DC.: 

8 U.S. Nuclear Regulatory Commission. Agencywide Documents Access and Management 

9 System Accession No. ML16179A148. 2016. 
2-1

1 2 DEFINITIONS

2 This chapter defines the usage of terms in the technical basis discussions in Chapter 3, the 

3 aging management review (AMR) tables in Chapter 4, and the aging management programs in 

4 Chapter 5. Selected definitions and usage are provided for the materials of construction, 

5 service environments, aging mechanisms, and aging effects (the manifestations of aging 

6 mechanisms by degraded conditions or performance). 

7 2.1 Materials

8 Table 2-1 describes many of the terms used to describe the materials of construction for the dry 

9 storage systems (DSSs).

Table 2-1 Use of terms for materials

Term Usage in This Document

Aluminum Includes commercially pure aluminum 1100 and 

precipitation-hardened alloys 6061 and 6063.

BISCO NS-3 A castable cementitious material for neutron and gamma 

shielding applications that may be blended with boron fillers to 

enhance neutron attenuation. It is fully encased in a metal, such 

as aluminum or steel.

Boral® A laminate composite that is used as a neutron poison material. 

It consists of a core of aluminum and boron-carbide powder 

sandwiched between sheets of aluminum. The boron-carbide 

content in the core ranges from 35 to 65 weight percent.

Boralyn®, MetamicTM Two variations of boron-carbide aluminum metal-matrix 

composite for neutron poison applications, one with billets 

produced by vacuum hot pressing (Boralyn®) and the second 

produced by cold isostatic pressing followed by vacuum sintering 

(MetamicTM). 

Borated aluminum An aluminum alloy typically containing up to 4.5 weight percent

boron. It is used as a neutron poison material. The boron is 

incorporated in the aluminum matrix as discrete particles of AlB2

or TiB2 (for alloys also containing titanium). Aluminum alloys 

1100, 6063, and 6351 have been used as base materials for 

boron additions.

Borated polymers Borated polymers include borated polyester resin and 

polypropylene for neutron shielding applications. Borated 

polyester resin is an unsaturated polyester crosslinked with 

styrene and typically contains about 50 weight percent mineral 

and fiberglass reinforcement.

Borated stainless steel An austenitic chromium-nickel steel with boron additions up to 

2.5 weight percent. It is used as a neutron poison material. The 

boron in the form of borides is dispersed in the Type 304 

stainless steel matrix as an intermetallic phase.

Concrete A mixture of hydraulic cement, aggregates, and water, with or 

without admixtures, fibers, or other cementitious materials.
2-2 

Table 2-1 Use of terms for materials

Term Usage in This Document

Copper alloys Copper alloys used in DSSs include bronzes (copper alloyed with 

tin) and brasses (copper alloyed with zinc).

Holtite-ATM A Holtec neutron shielding material consisting of epoxy polymer, 

B4C added as a finely divided powder, and aluminum hydroxide. 

It is fully encased in a metal enclosure.

Nickel alloys Nickel alloys include Inconel 718 and X750. Inconel is a family of 

austenitic nickel-chromium-based superalloys. Both Inconel 718 

and X750 are precipitation-hardening alloys.

Stainless steel Stainless steel includes Types 304, 316, XM-19, SA193-Gr. B8, 

SA351-Gr. CF3, and Nitronic 60 austenitic stainless steels and 

Type 630 precipitation-hardening martensitic stainless steel. 

Type 630 stainless steel is commonly referred to as 17-4PH and 

contains 15–17.5 percent chromium, 3–5 percent copper, and 

3–5 percent nickel (in weight percent).

Chrome-plated stainless steel is also included in the category of 

stainless steel.

Steel Various carbon steels, alloy steels, and high-strength, low-alloy 

steels. Examples of steel designations included in this category 

are ASTM A36, ASTM A320-Gr. L43, ASTM F436, SA36, 

SA193-Gr. B7, SA203-Gr. D/E, SA266-Cl. 2, SA320-Gr. L43, 

SA350-Gr. LF2/LF3, SA414, SA508-Cl. 1A/3A, SA516-Gr. 70, 

SA533-Gr. B, SA537-Cl. 2, SA540-Gr. B23/24, SA620, and 

SA696-Gr. B.

Galvanized steel, aluminum-coated steel, and electroless 

nickel-plated steel are also included in the category of steel.

Zirconium-based alloys The materials of construction of fuel cladding and fuel assembly 

hardware. Various zirconium-based materials have been used in 

commercial reactor applications because of their low neutron 

cross section and excellent corrosion resistance to a variety of

environmental conditions. The cladding types Zircaloy-2, 

Zircaloy-4, ZIRLOTM, and M5® are included in this category.

Zirconium-based cladding in high burnup (HBU) spent nuclear 

fuel refers to assembly-average burnups exceeding 45 

GWd/MTU.

1 
2-3

1 2.2 Environments

2 Table 2-2 defines many of the environments to which DSS SSCs are exposed.

Table 2-2 Use of terms for environments

Term Usage in This Document

Air–outdoor (OD) Direct exposure to weather, including precipitation and wind; 

possibly salt laden.

The indoor air environment to which transfer cask components 

are typically exposed is conservatively evaluated as outdoor air.

Demineralized water (DW) Water that has been treated to remove dissolved minerals. 

Demineralized water is used as the liquid neutron shield in

transfer casks. 

Embedded in:

Concrete (E-C)

Metal (E-M)

Neutron shielding (E-NS)

When one or more surfaces of a component are in contact with 

another component or material. This may prevent ingress of 

water and contaminants to the embedded surface, depending on 

the permeability of the embedding environment.

Fully encased or lined (FE) The environment of some concrete structures that are fully

enclosed inside another component or fully lined by another 

material (e.g., steel), which prevents ingress of water and 

contaminants. Also, ceramic fiber insulation is fully encased in 

foil-facing or jacketing.

Helium (HE) The helium fill gas inside a canister or cask and trace quantities 

of other gases, such as nitrogen, oxygen, argon, and fission 

product gases. This environment applies to fuel, cladding, and 

other internal components inside a cask. 

Groundwater/soil (GW) Groundwater is subsurface water found in wells, tunnels, or 

drainage galleries, or water that flows naturally to the earth’s 

surface via seeps or springs. Soil is a mixture of organic and 

inorganic materials produced by the weathering of rock and clay 

minerals or the decomposition of vegetation. Voids containing 

air and moisture can occupy 30 to 60 percent of the soil volume. 

Below-grade concrete structures are assumed to be partially 

exposed to a groundwater or soil environment.

Sheltered (SH) The environment outside a sealed canister but within the 

confined internal space of a shielding structure (e.g., overpack or 

horizontal storage module). The sheltered environment is open 

to outdoor air, but it is shielded from direct exposure to 

precipitation. This environment may contain moisture, salts, and 

other contaminants from the outdoor air. 

3
2-4

1 2.3 Aging mechanisms

2 Table 2-3 defines the aging mechanisms that are evaluated in this report.

Table 2-3 Use of terms for aging mechanisms

Term Usage in This Document

Aggressive chemical 

attack

The degradation of concrete by strong acids. Chlorides and 

sulfates of potassium, sodium, and magnesium may attack 

concrete, depending on their concentrations in the 

soil/groundwater that comes into contact with the concrete. The 

minimum thresholds causing concrete degradation are 500 ppm 

chloride and 1,500 ppm sulfate.

Boron depletion The degradation of the neutron-absorbing capacity of neutron 

poison and shielding materials when they are exposed to 

neutron fluence.

Corrosion The electrochemical reaction of a metal or metal alloy in an

environment that results in oxidation or wastage of the material.

Creep Creep, for a metallic material, refers to a time-dependent 

continuous deformation process under constant stress. It is a

thermally activated process and is generally a concern at 

temperatures greater than 40 percent of the material’s absolute 

melting temperature. However, low-temperature creep is an 

athermal process that is considered as a potential degradation 

mechanism for some alloys, including zirconium-based alloys.

In concrete, creep is related to the loss of absorbed water from 

the hydrated cement paste. It is a function of the modulus of 

elasticity of the aggregate. 

Crevice corrosion Localized corrosion in joints, connections, and other small, 

close-fitting regions that develop local aggressive environments. 

Dehydration at high 

temperatures

Dehydration reactions of the hydrated cement paste in concrete 

when exposed to temperatures greater than 65 degrees C 

[149 degrees F]. Dehydration can degrade concrete strength and 

increase susceptibility to cracking. The degree of concrete 

degradation depends on several factors, including concrete 

mixing, aggregate type, curing, loading condition, moisture 

retention and content, and exposure time.

Delayed ettringite 

formation

During concrete curing, the naturally occurring ettringite (a 

calcium aluminum sulfate mineral) converts to 

monosulfoaluminate if curing temperatures are greater than about 

70 degrees C [158 degrees F]. After concrete hardens, ettringite 

will reform if the temperature decreases below about 70 degrees 

C [158 degrees F], resulting in concrete cracking and spalling. 

The conditions necessary for the occurrence of delayed ettringite 

formation are excessive temperatures during concrete casting, 

the presence of internal sulfates, and a moist environment.
2-5

Table 2-3 Use of terms for aging mechanisms

Term Usage in This Document

Delayed hydride cracking The propagation of a crack in zirconium-based cladding materials 

as a result of diffusion of hydrogen to a crack tip and the 

embrittlement of the near-tip region due to hydride precipitation. 

The operability of the delayed-hydride-cracking mechanism in fuel 

cladding depends on the stress imposed on the cladding.

Erosion Soil erosion, or removal, is primarily caused by rainfall and

surface runoff, floods, or wind. Soil erosion can affect the stability 

of concrete structures, resulting in scouring that is a localized loss 

of soil, often around a foundation element. Factors that affect the 

erosion rates include soil structure and composition, climate, 

topography, and vegetation cover.

Fatigue Also termed “cyclic loading” or “thermal/mechanical fatigue.” 

Fatigue is a phenomenon leading to fracture under repeated or 

fluctuating stresses having a maximum value less than the tensile 

strength of the material. Fatigue fractures are progressive and 

grow under the action of the fluctuating stress. Fatigue due to 

cyclic thermal loads is defined as the structural degradation that 

can occur from repeated stress/strain cycles caused by 

fluctuating loads and temperatures. After repeated cyclic loading 

of sufficient magnitude, microstructural damage may accumulate, 

leading to macroscopic crack initiation at the most vulnerable 

regions. Subsequent mechanical or thermal cyclic loading may 

lead to growth of the initiated crack.

Freeze-thaw Repeated freezing and thawing of water can cause degradation 

of concrete, characterized by scaling, cracking, and spalling. The 

cause is water freezing within the pores of the concrete, creating 

hydraulic pressure.

Galvanic corrosion Accelerated corrosion of a metal when in electrical contact 

with a more noble metal or nonmetallic conductor in a 

corrosive electrolyte.

General corrosion Uniform loss of material due to corrosion, proceeding at 

approximately the same rate over a metal surface.

Hydride reorientation and 

hydride-induced 

embrittlement

The precipitation of radial hydrides results in embrittlement of 

zirconium-based cladding materials under pinch-load stresses at 

low-to-moderate temperatures. Reorientation of hydrides from 

the circumferential-axial to radial-axial direction is caused by

heating and cooling of the cladding under sufficient cladding hoop 

tensile stresses.
2-6

Table 2-3 Use of terms for aging mechanisms

Term Usage in This Document

Leaching of calcium 

hydroxide

The dissolution of calcium-containing concrete components 

(e.g., calcium hydroxide) when water passes through either 

cracks, inadequately prepared construction joints, or areas not 

sufficiently consolidated during placing. Once the calcium 

hydroxide has been leached away, other cementitious 

constituents become vulnerable to chemical decomposition, 

finally leaving only the silica and alumina gels behind and 

lowering the strength of the concrete. The water’s 

aggressiveness in the leaching of calcium hydroxide depends on 

its salt content, pH, and temperature. This leaching action is 

effective only if the water flows through the concrete.

Mechanical overload The overload of fuel cladding due to fuel pellet swelling. Fuel 

pellet swelling is the result of decay gas production in the pellet. 

Pellet swelling can increase stresses on the cladding.

Microbiological 

degradation

Biodegradation attack of concrete by organisms growing on its 

surfaces under favorable environmental conditions 

(e.g., moisture, near neutral pH, presence of nutrients), causing 

an increase in concrete porosity and permeability and the loss of 

material by spalling or scaling.

Microbiologically 

influenced corrosion

Any of the various forms of corrosion influenced by the activity of 

such microorganisms as bacteria, fungi, and algae, and/or the 

products of their metabolism. For example, anaerobic bacteria 

can establish an electrochemical galvanic reaction or disrupt a 

passive protective film; acid-producing bacteria can produce 

corrosive metabolites.

Oxidation A corrosion reaction. In this report, oxidation also is a defined 

aging mechanism describing the reaction of zirconium alloy fuel 

rod cladding with water to form zirconium oxide..

Pitting corrosion A localized form of corrosion that is confined to a point or small 

area of a metal surface. It takes the form of cavities called pits.

Radiation damage and 

radiation embrittlement

The loss of ductility, fracture toughness, and resistance to 

cracking of metals that may occur under exposure to neutron 

radiation. In concrete, radiation exposure can cause dissociation 

of water into hydrogen and oxygen, leading to decreased 

compressive and tensile strengths. The extent of radiation 

damage to concrete depends on the neutron and gamma fluence.

Reaction with aggregates The presence of reactive alkalis in concrete can lead to 

subsequent reactions with aggregates that may lead to cracking, 

a loss of material, or an increase in porosity and permeability. 

These alkalis are introduced mainly by cement but also may 

come from admixtures, salt contamination, seawater penetration,

or solutions of deicing salts. These reactions include alkali-silica 

reactions, cement-aggregate reactions, and aggregate-carbonate 

reactions. 
2-7

Table 2-3 Use of terms for aging mechanisms

Term Usage in This Document

Salt scaling Salt scaling damage manifests as flaking of material from the

concrete surface. Salt scaling takes place when concrete is 

exposed to freezing temperatures, moisture, and dissolved 

salts (e.g., deicing salts). This degradation mode affects

mainly horizontal concrete surfaces where water ponding can 

be expected. 

Settlement Settlement of a concrete structure may occur due to changes in 

the site conditions (e.g., water table). The amount of settlement 

depends on the foundation material.

In soil, loss of form due to settlement can occur during the first 

several years of placement. Factors that control soil settlement 

include the type of soil particles and particle packing, the amount 

of water used during the compaction process, and the height of 

soil fill.

Shrinkage Shrinkage of concrete can result from cement hydration and loss 

of moisture during drying. Cracking and shortening of concrete 

due to shrinkage can occur early after concrete placement.

Stress corrosion cracking 

(SCC)

The cracking of a metal produced by the combined action of 

corrosion and a tensile stress (applied or residual). SCC is highly 

chemical specific in that certain alloys are likely to undergo SCC 

only when exposed to a small number of chemical environments.

Stress relaxation A loss of preload in a heavily loaded bolt. Over time, the 

clamping force provided by a bolt may decrease due to atomic 

movement within the stressed bolt material (analogous to the 

metallic creep mechanism at elevated temperatures).

Thermal aging Also termed “thermal aging embrittlement” or “thermal 

embrittlement.” Many materials are intentionally thermally aged 

during their manufacture to achieve desired mechanical 

properties. Continued exposure to elevated temperatures during 

operation can, in some cases, result in undesirable properties.

For example, at operating temperatures of 300 to 400 degrees C 

[572 to 752 degrees F], austenitic stainless steel welds that 

contain ferrite exhibit a spinodal decomposition of the ferrite 

phase into ferrite-rich and chromium-rich phases. This may give 

rise to embrittlement (reduction in fracture toughness), depending 

on the amount, morphology, and distribution of the ferrite phase 

and the composition of the stainless steel.

Wear The removal of surface material due to relative motion between 

two surfaces or under the influence of hard, abrasive particles. 

Wear occurs in parts that experience intermittent relative motion 

or frequent manipulation.
2-8

Table 2-3 Use of terms for aging mechanisms

Term Usage in This Document

Wet corrosion and 

blistering

A degradation mechanism for neutron poison plates with open 

porosity as a result of water entering pores in the material during 

loading, leading to internal corrosion. Blisters occur from trapped 

hydrogen produced from corrosion reactions. Wet corrosion and 

blistering can cause dimensional changes affecting criticality 

considerations due to moderator displacement and may also 

hinder the retrieval of fuel assemblies.

1
2-9

1 2.4 Aging effects

2 An aging effect is the manifestation of an aging mechanism, as evidenced by a degraded 

3 condition or performance. Table 2-4 defines the aging effects described in this report.

Table 2-4 Use of terms for aging effects

Term Usage in This Document

Changes in dimension A change in the size of a component resulting from creep of 

aluminum and zirconium-based alloys. Changes in dimension 

also can be caused by wet corrosion and blistering of Boral®

neutron poison materials.

Cracking Crack initiation and growth in metallic components as a result of 

SCC, fatigue, and delayed hydride cracking. Cracking in 

concrete is a complete or incomplete separation of concrete into 

two or more parts produced by breaking or fracturing.

Increase in porosity and 

permeability

An increase in the percentage of the volume of voids in a 

concrete material or an increase in the susceptibility of concrete 

to permit liquids or gasses to pass through.

Loss of bond A loss of the interacting force that prevents slip of the reinforcing 

steel bars relative to the surrounding concrete in a reinforced 

concrete member.

Loss of criticality control A diminishment of the capability of neutron poison materials to 

maintain the subcriticality of spent nuclear fuel. 

Loss of form A change in the shape or position of soil resulting from settlement 

due to poor soil consolidation. In addition, soil tends to absorb 

moisture with time and thus promotes loss of form.

Loss of fracture toughness 

and loss of ductility

A decrease in the ability of a material to resist fracture. This 

phenomenon results from thermal aging embrittlement, radiation 

embrittlement, or hydrogen embrittlement.

Loss of material The destructive removal of material due to general corrosion, 

pitting corrosion, crevice corrosion, galvanic corrosion, 

microbiologically influenced corrosion, or aggressive chemical 

attack. In concrete structures, loss of material can result from 

local flaking, spalling, or peeling away of the near-surface portion 

of hardened concrete.

Loss of preload A reduction in the clamping force in a mechanically loaded joint.

Loss of shielding A diminishment of the capability of a material to shield radiation. 

Loss of strength A decrease in the ability of a material to support a mechanical 

load. In metals, loss of strength may be due to thermal aging or 

annealing. In concrete structures, loss of strength can also be 

caused by the leaching of calcium hydroxide or reaction with 

aggregates.

None A term used in the AMR tables for certain material and 

environment combinations that may not be subject to credible

aging mechanisms; thus, there are no relevant aging effects that 

require management.
2-10

Table 2-4 Use of terms for aging effects

Term Usage in This Document

Precursor to SCC A material condition that initiates SCC. Both pitting and crevice 

corrosion are known to be precursors to SCC and, as such, can 

lead to cracking of stainless steel canisters.

Reduction of concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

A decrease in the alkalinity of concrete. If the pH of concrete in 

which steel is embedded is reduced below 11.5 by intrusion of 

aggressive ions (e.g., chlorides > 500 ppm) in the presence of 

oxygen, embedded steel may corrode. A reduction in pH can be 

caused by carbonation.

1 
3-1

1 3 EVALUATION OF AGING MECHANISMS

2 3.1 Introduction

3 This chapter evaluates known aging degradation mechanisms to determine which of those 

4 could adversely affect an important-to-safety function in the 20- to 60-year period of extended 

5 operation. These evaluations provide the technical bases for the recommendations in the aging 

6 management review (AMR) tables and aging management programs (AMPs) in Chapters 4 

7 and 6, respectively. This chapter is first divided into major component areas (e.g., casks and 

8 internals, concrete overpacks), which in turn are subdivided into discussions of the aging 

9 mechanisms for each of the materials of construction (e.g., steel, aluminum). 

10 Each evaluation in this chapter concludes with a determination of whether the aging mechanism 

11 is considered “credible” in the period of extended operation. A credible aging mechanism is one 

12 that could affect an important-to-safety function if the mechanism were not addressed by an 

13 aging management activity. The AMR tables in Chapter 4 recommend an AMP, time-limited 

14 aging analysis (TLAA), or other analysis to address the effects of aging.

15 Table 3-2 through Table 3-6 summarize the conclusions in this chapter. For each material, the 

16 tables show in which environments the aging mechanisms were determined to be credible and

17 noncredible. Not all combinations of materials, environments, and aging mechanisms were 

18 evaluated in each major component area. This occurs because some material-environment 

19 combinations do not exist in every major component area or, in some instances, aging 

20 mechanisms were not considered to be reasonably plausible, and thus an evaluation was not 

21 performed. The reviewer should note that these conclusions are based only on a review of the 

22 specific storage system designs described in Section 1.2 and Chapter 4, and thus the reviewer 

23 should consider the credibility of aging mechanisms for other systems on a case-by-case basis.

24 Table 3-1 provides the environment abbreviations used in the summary tables. 

Table 3-1 Environment abbreviations

Outdoor air OD

Demineralized water DW

Embedded in concrete E-C

Embedded in metal E-M

Embedded in neutron shielding E-NS

Fully encased or lined FE

Helium HE

Groundwater/soil GW

Sheltered SH

25

26
3-2 

Table 3-2 Casks and internals aging mechanism evaluations

Material Aging Mechanism Credible 

Environments

Noncredible 

Environments Section 

Steel

General corrosion OD, SH, DW, GW, E-C E-M, E-NS, HE 3.2.1.1

Pitting and crevice 

corrosion

OD, SH, DW, GW, E-C E-M, E-NS, HE 3.2.1.2

Galvanic corrosion* OD, SH 3.2.1.3

Microbiologically influenced 

corrosion (MIC)

GW, E-C OD, SH, DW, E-M, 

E-NS, HE 3.2.1.4

Stress corrosion cracking 

(SCC)

OD, SH 3.2.1.5

Creep OD, SH, DW, GW, 

E-M, E-NS, HE 3.2.1.6

Fatigue Evaluate design code TLAA, if applicable 3.2.1.7

Thermal aging OD, SH, DW, GW, 

E-M, E-NS, HE 3.2.1.8

Radiation embrittlement OD, SH, DW, GW, 

E-M, E-NS, HE

3.2.1.9

Stress relaxation SH OD 3.2.1.10

Wear OD 3.2.1.11

Stainless Steel

General corrosion OD, SH, DW, E-M, 

E-NS, HE 3.2.2.1

Pitting and crevice 

corrosion†

OD, SH DW, E-M, E-NS, HE 3.2.2.2

Galvanic corrosion* OD, SH 3.2.2.3

MIC OD, SH, DW, E-M, 

E-NS, HE 3.2.2.4

SCC‡ OD, SH DW, E-M, E-NS, HE 3.2.2.5

Creep OD, SH, DW, E-M, 

E-NS, HE 3.2.2.6

Fatigue Evaluate design code TLAA, if applicable 3.2.2.7

Thermal aging HE§ OD, SH, DW, E-M, 

E-NS 3.2.2.8

Radiation embrittlement OD, SH, DW, E-M, 

E-NS, HE 3.2.2.9

Stress relaxation OD, SH 3.2.2.10

Wear OD 3.2.2.11

*where dissimilar material galvanic couples exist

†as a precursor to SCC

‡SCC is credible at welds and other regions where sufficient stress exists; transfer cask components exposed to 

indoor/outdoor air are not considered to be susceptible to SCC because their surfaces are periodically rinsed with 

demineralized water.

§thermal aging is credible only for precipitation-hardened martensitic stainless steels

1 
3-3

Table 3-2 Casks and internals aging mechanism evaluations (continued)

Material Aging Mechanism Credible 

Environments

Noncredible 

Environments Section

Aluminum Alloys

General corrosion* SH E-M, E-NS, HE 3.2.3.1

Pitting and crevice corrosion SH E-M, E-NS, HE 3.2.3.2

Galvanic corrosion† SH HE 3.2.3.3

MIC SH, E-M, E-NS, HE 3.2.3.4

Creep analyses required‡ 3.2.3.5

Fatigue Evaluate design code TLAA, if applicable 3.2.3.6

Thermal aging analyses required‡ 3.2.3.7

Radiation embrittlement SH, E-M, E-NS, HE 3.2.3.8

Nickel Alloys

General corrosion OD 3.2.4.1

Pitting and crevice corrosion OD 3.2.4.2

MIC OD 3.2.4.3

SCC OD 3.2.4.4

Fatigue Evaluate design code TLAA, if applicable 3.2.4.5

Radiation embrittlement OD 3.2.4.6

Stress relaxation OD 3.2.4.7

Wear OD 3.2.4.8

Copper Alloys

General corrosion OD 3.2.5.1

Pitting and crevice corrosion OD 3.2.5.2

MIC OD 3.2.5.3

Radiation embrittlement OD 3.2.5.4

Lead All E-M 3.2.6

Depleted 

Uranium All E-M 3.2.7

Coatings Radiation embrittlement analyses required 3.2.8

*general corrosion is not considered to be credible for anodized aluminum

†where dissimilar metal couples exist

‡creep and thermal aging are relevant only for load-bearing components.

1

2
3-4 

Table 3-3 Neutron shielding materials aging mechanism evaluations

Material Aging Mechanism Credible 

Environments

Noncredible 

Environments Section 

Neutron 

Shielding

Boron depletion analyses required 3.3.1.1

Thermal aging FE* 3.3.1.2

Radiation embrittlement FE* 3.3.1.3

*thermal aging and radiation embrittlement are credible only for polymer-based neutron-shielding materials.

1 

2 

Table 3-4 Neutron poison materials aging mechanism evaluations

Material Aging Mechanism Credible 

Environments

Noncredible 

Environments Section 

Borated 

Stainless Steels

General corrosion HE 3.4.1

Galvanic corrosion HE 3.4.1

Wet corrosion and blistering HE 3.4.1

Boron depletion HE* 3.4.1.1

Creep HE 3.4.1.2

Thermal aging HE 3.4.1.3

Radiation embrittlement HE 3.4.1.4

Borated 

Aluminum and 

Aluminum-based 

Composites 

General corrosion HE 3.4.2.1

Galvanic corrosion HE 3.4.2.2

Wet corrosion and blistering HE 3.4.2.3

Boron depletion HE* 3.4.2.4

Creep HE† 3.4.2.5

Thermal aging HE† 3.4.2.6

Radiation embrittlement HE 3.4.2.7

*when a boron depletion analysis is included in the design basis, applicants must provide a TLAA to demonstrate

that depletion will not challenge noncriticality in the period of extended operation

†creep and thermal aging are relevant only for load-bearing aluminum components.

3 

4 
3-5 

Table 3-5 Concrete overpacks, support pads, and ceramic fiber insulation aging

 mechanism evaluations 

Material Aging Mechanism Credible 

Environments

Noncredible 

Environments Section 

Concrete

Freeze and thaw OD, GW 

(above freeze line)

SH, FE, GW 

(below freeze line) 3.5.1.1

Creep all 3.5.1.2

Reaction with aggregates all* 3.5.1.3

Differential settlement OD, SH, GW 3.5.1.4

Aggressive chemical attack OD, GW SH, FE 3.5.1.5

Corrosion of reinforcing 

steel

OD, GW SH, FE 3.5.1.6

Shrinkage OD, SH, GW, FE 3.5.1.7

Leaching of calcium 

hydroxide

OD, SH, GW FE 3.5.1.8

Radiation damage OD, SH, GW, FE 3.5.1.9

Fatigue OD, SH, GW, FE 3.5.1.10

Dehydration at high 

temperature

OD, SH, GW, FE 3.5.1.11

Microbiological degradation GW OD, SH, FE 3.5.1.12

Delayed ettringite formation OD, SH, GW, FE 3.5.1.13

Salt scaling OD, GW 

(above freeze line)

SH, FE, GW 

(below freeze line) 3.5.1.14

Ceramic Fiber 

Insulation

Radiation damage analysis required 3.5.2.1

Moisture absorption 3.5.2.2 3.5.2.2

*where moisture is available

1 
3-6 

Table 3-6 Spent fuel assembly aging mechanism evaluations

Material Aging Mechanism Credible 

Environments

Noncredible 

Environments Section 

Cladding 

Materials

(Zirconiumbased Alloys)

Hydride reorientation† HE* 3.6.1.1

Delayed hydride cracking† HE 3.6.1.2

Thermal creep† HE 3.6.1.3

Low-temperature creep† HE 3.6.1.4

Mechanical overload† HE 3.6.1.5

Oxidation HE 3.6.1.6

Pitting corrosion HE 3.6.1.7

Galvanic corrosion HE 3.6.1.8

SCC HE 3.6.1.9

Radiation embrittlement HE 3.6.1.10

Fatigue HE 3.6.1.11

Assembly 

Hardware 

Materials 

(Zirconiumbased, Inconel, 

and Stainless 

Steel Alloys)

Creep HE 3.6.2.1

Hydriding HE 3.6.2.2

General corrosion HE 3.6.2.3

SCC HE 3.6.2.4

Radiation embrittlement HE 3.6.2.5

Fatigue HE 3.6.2.6

*Although hydride reorientation and hydride-induced embrittlement of high-burnup cladding is credible, these

mechanisms are only expected to potentially compromise intended functions under pinch-type loads. Such loads 

are not expected to be present during storage.

†applicable to high-burnup fuel

1 
3-7

1 3.2 Casks and internals

2 “Casks and internals” include various metallic subcomponents of the storage casks or canisters, 

3 the fuel baskets and other internal subcomponents (other than spent fuel assemblies), the 

4 storage modules or overpacks, and the transfer casks. These subcomponents are exposed to 

5 several environments within and outside the dry storage systems (DSSs), such as sheltered 

6 environments, indoor air, outdoor air, demineralized water, groundwater or soil, helium, and 

7 embedded environments. The spent nuclear fuel (SNF) also exposes subcomponents to 

8 elevated temperatures and radiation, with heat exposure and dose depending on the 

9 subcomponent location and the SNF characteristics (e.g., burnup and age of spent fuel). The 

10 materials of construction for these subcomponents include steel, stainless steel, aluminum 

11 alloys, nickel alloys, copper alloys, and lead.

12 A set of known aging mechanisms for metallic cask and internal subcomponents was 

13 established by first broadly identifying all potential mechanisms through a review of gap 

14 assessments for DSSs, technical literature, and operating experience from nuclear and 

15 nonnuclear applications (NRC, 2014, 2010a; Chopra et al., 2014; Hanson et al., 2012; 

16 Sindelar et al., 2011; NWTRB, 2010). The known environmental, thermal, mechanical, and 

17 irradiation-induced aging mechanisms are as follows:

18 • general corrosion

19 • pitting and crevice corrosion

20 • galvanic corrosion

21 • MIC

22 • SCC (including hydrogen embrittlement)

23 • creep

24 • fatigue

25 • thermal aging

26 • radiation embrittlement

27 • stress relaxation

28 • wear

29 Not all of these mechanisms are considered to be credible for each structure, system, and 

30 component (SSC). For example, temperatures are not considered sufficiently high to cause 

31 creep of steel and stainless steel subcomponents. Also, general corrosion is not considered to 

32 be a credible aging mechanism for subcomponents fabricated from stainless steels, because 

33 these materials exhibit passive behavior and negligible general corrosion rates. Detailed 

34 discussions regarding potential aging mechanisms for each material and the technical bases for 

35 those requiring aging management follow.

36 3.2.1 Steel (carbon, low-alloy, high-strength low-alloy)

37 In DSSs, steel subcomponents are commonly used and are exposed to sheltered environments, 

38 outdoor air, helium, demineralized water, and groundwater or soil, and also may be embedded 

39 in concrete or neutron-shielding materials. The exterior surfaces of some steel subcomponents 

40 are coated with epoxy or inorganic zinc to mitigate corrosion; however, these coatings can 

41 degrade, resulting in exposure of steel to the atmosphere. Steels used to construct transfer

42 casks are predominately exposed to an indoor environment, except for short periods of outdoor

43 exposure during transfer operations. For such air-indoor/outdoor environment exposure, aging 
3-8 

1 effects from aqueous corrosion processes are expected to be bounded by the outdoor 

2 environment. As such, the indoor air environment is not discussed separately. 

3 3.2.1.1 General corrosion

4 General corrosion, also known as uniform corrosion, proceeds at approximately the same rate 

5 over a metal surface (Phull, 2003b). Freely exposed steel surfaces in contact with moist air or 

6 water are subject to general corrosion. The corrosion rate depends on solution composition, 

7 pH, and temperature. The iron Pourbaix diagram shows that iron undergoes active corrosion 

forming Fe2+ or Fe3+ 8 ions at pH values lower than 8.5 to 9 (Kodama, 2005). At higher values of 

9 pH, iron can be passive, leading to a very low corrosion rate. 

10 Steel subcomponents exposed to outdoor and sheltered environments

11 If steel is placed in a completely dry atmosphere, oxide film growth is so small that the corrosion 

12 rate is virtually negligible. However, in outdoor conditions, rain, fog, snow, and dew 

13 condensation can generate moisture layers on the steel surface that cause general corrosion. 

14 Atmospheric corrosion rates can vary from 0 to 0.2 millimeters/year (mm/yr) [0 to 7.9 mils/yr] 

15 depending on relative humidity, temperature, and levels of chloride and pollutants in the 

16 atmosphere (NACE, 2002). Rates can be more significant in industrial and marine 

17 environments (McCuen and Albrecht, 1994). 

18 In a sheltered environment, deliquescence of airborne salts below the dew point also could 

19 generate an aqueous electrolyte initiating general corrosion. These salts may be chloride rich 

20 and originate from marine environments, deicing salts, and condensed water from cooling 

21 towers, as well as a range of other nonchloride-rich species originating from industrial, 

22 agricultural, and commercial activities. Studies have shown that MgCl2, a component of sea salt 

23 with a low deliquescence relative humidity, would deliquesce below 52 degrees C 

24 [126 degrees F] under realistic absolute humidities in nature (He et al., 2014). The heat 

25 generated by the radioactive decay of spent fuel decreases over time. Time-temperature 

26 profiles calculated for the stainless steel canister shell suggest that, while initial temperatures 

27 are high, the threshold temperature for deliquescence of some salts on the external surface of 

28 the shell could be reached during the 60-year timeframe (EPRI, 2006; Meyer et al., 2013). 

29 Because steel subcomponents exposed to sheltered environments are usually located farther 

30 away from the fuel compared to the stainless steel canister shell, they are expected to reach 

31 these threshold temperatures for deliquescence at an earlier time. As such, the potential for 

32 general corrosion of steel subcomponents exposed to a sheltered environment is present.

33 Because aqueous electrolytes initiating general corrosion of steels exposed to outdoor and 

34 sheltered environments are potentially present, and corrosion rates may be sufficient to affect 

35 component intended functions, general corrosion is considered to be credible, and therefore, 

36 aging management is required during the 60-year timeframe.

37 Steel subcomponents exposed to demineralized water

38 Demineralized water is used in the steel water jacket of some transfer casks for radiation 

39 shielding. In some cases, 25-percent ethylene glycol is added to the water to decrease the 

40 freezing point, and this is expected to decrease the corrosivity of water (van Bodegom et al., 

41 1987). The iron Pourbaix diagram shows that iron undergoes active corrosion at neutral pH, as 

42 long as water is present (Kodama, 2005). The corrosion rate for iron is approximately 
3-9

1 0.1 mm/yr [3.9 mils/yr] in stagnant fresh water at atmospheric temperatures (Kodama, 2005). In 

2 60 years of continuous exposure in such water, the material thinning is expected to be 

3 approximately 6 mm [0.2 in]. This is a conservative estimate of the corrosion of steel water 

4 jackets, as the jackets are not necessarily filled when the transfer cask is not in use. However, 

5 general corrosion of steels exposed to demineralized water is nonetheless considered to be 

6 credible, and therefore, aging management is required during the 60-year timeframe. 

7 Steel subcomponents exposed to groundwater or soil

8 The corrosion rate of steel in groundwater or soil depends on many factors, such as the oxygen 

9 level; resistivity; pH, buffer capacity; redox potential; and the presence of chlorides, sulfides, 

10 neutral salts, and sulfates. Soils may be acidic, neutral, or alkaline, with pH values typically 

11 ranging from 4.5–8.5 (Kodama, 2005), which is in the range of active corrosion discussed

12 previously. Corrosion rate data for iron artifacts buried in soil show that most corrosion rates 

13 are 0.1 to 10 micrometers (μm)/yr [0.004 to 0.4 mils/yr], despite the variety of artifacts in terms 

14 of origin and environmental conditions (David et al, 2002). In 60 years of continuous soil 

15 exposure, the material thinning is expected to be approximately 0.006 to 3.6 mm [0.2 to 

16 142 mils]. As such, general corrosion of steels exposed to groundwater or soil is considered to 

17 be credible, and therefore, aging management is required during the 60-year timeframe.

18 Steel subcomponents exposed to an embedded (concrete) environment

19 In overpacks, some steel subcomponents are embedded in concrete. The concrete is in contact 

20 with air or soil. When the concrete is intact, the alkaline concrete solution passivates the steel.

21 As concrete degrades with time, embedded steel can be exposed to water containing dissolved 

22 carbonates and chlorides, and general corrosion can be significant, as discussed previously. As 

23 such, general corrosion of steels exposed to an embedded (concrete) environment is 

24 considered to be credible, and therefore, aging management is required during the 60-year 

25 timeframe.

26 Steel subcomponents exposed to an embedded (neutron-shielding) environment

27 In DSSs, some polymer-based, neutron-shielding materials are poured into a steel structure, 

28 leaving one side of the steel embedded. The neutron-shielding materials include HoltiteTM and 

29 BISCO NS-3. Because the embedded side of the steel has limited exposure to water and 

30 oxygen, general corrosion is not considered to be credible, and therefore, aging management is 

31 not required during the 60-year timeframe. 

32 Steel subcomponents exposed to helium

33 As mentioned previously, the iron Pourbaix diagram shows that iron undergoes active corrosion

34 at neutral pH as long as water is present (Kodama, 2005). However, there is very little residual

35 water in internal environments following drying and refilling with inert gas, and thus the corrosion 

36 reaction with steel will be limited. Jung et al. (2013) show that the relative humidity inside the 

37 system after drying is no more than 5 percent at the beginning of storage and is less than 

38 0.5 percent in 60 years. Furthermore, some steel subcomponents are coated by aluminum or 

39 electroless nickel, which are more corrosion resistant than steel. As such, general corrosion of 

40 steel exposed to helium is not considered to be credible, and therefore, aging management is 

41 not required during the 60-year timeframe.
3-10

1 3.2.1.2 Pitting and crevice corrosion

2 Pitting corrosion is a localized form of corrosion that is confined to a point or small area of a 

3 metal surface (Frankel, 2003). It takes the form of cavities called pits. Crevice corrosion is 

4 another localized form of corrosion that occurs in a wetted environment when a crevice exists 

5 (Kelly, 2003). It occurs more frequently in connections, lap joints, splice plates, bolt threads, 

6 under bolt heads, or at points of contact between metals and nonmetals. Crevice corrosion is 

7 associated with stagnant or low-flow solutions. As discussed previously, the common form of 

8 corrosion for steel is general corrosion. However, steel is also known to be susceptible to pitting 

9 and crevice corrosion in an oxidizing and alkaline environment, especially in the presence of 

10 chlorides. The exterior surfaces of some subcomponents are coated with epoxy or inorganic 

11 zinc to mitigate corrosion (e.g., the outer shell of the bolted cask system). Depending on the 

12 quality and chemical composition of the coating, water and corrosive agents can permeate 

13 coating defects, initiating pitting. After initiation of a coating defect, the coating could function as 

14 a crevice former and initiate crevice corrosion. 

15 Steel subcomponents exposed to outdoor and sheltered environments, demineralized water, 

16 groundwater or soil, and embedded (concrete) environments

17 As discussed in Section 3.2.1.1, the potential to form aqueous electrolytes on surfaces exposed 

18 to outdoor and sheltered environments is present, either via direct exposure to precipitation or 

19 through deliquescence of deposited salts. These electrolytes, demineralized water, and 

20 groundwater or soil could be conducive to pitting and crevice corrosion of steel. For steel 

21 embedded in concrete, as concrete degrades with time, steel can be exposed to water 

22 containing dissolved carbonates and chlorides, which could be conducive to pitting and crevice 

23 corrosion as well. 

24 Localized corrosion of steels is attributed to the presence of macro-galvanic cells, where local 

25 differences in electrochemical potential are created by conditions such as chemical composition 

26 differences within the steel matrix, discontinuous surface films (e.g., mill scale), and differences 

27 in oxygen supply (Revie, 2000). 

28 Because steel subcomponents exposed to outdoor and sheltered environments are likely to 

29 come into contact with aqueous electrolytes, and the localized corrosion in these environments 

30 is possible, loss of material due to pitting and crevice corrosion is considered to be credible. 

31 Therefore, aging management is required during the 60-year timeframe.

32 Steel subcomponents exposed to embedded (neutron-shielding materials) environments

33 Because of the limited water and oxygen in embedded environments, pitting and crevice 

34 corrosion are not considered to be credible, and therefore, aging management is not required 

35 during the 60-year timeframe. 

36 Steel subcomponents exposed to helium

37 Inside DSSs, there is very little residual water following drying, and thus the corrosion reaction 

38 with steel will be limited. Jung et al. (2013) show that the relative humidity inside the system is 

39 no more than 5 percent at the beginning of storage and is less than 0.5 percent in 60 years. 

40 Furthermore, some steel subcomponents are coated by aluminum or electroless nickel, which 

41 are more corrosion resistant than steel. As such, localized corrosion of steel exposed to helium 
3-11

1 is considered to be insignificant, and therefore, aging management is not required during the 

2 60-year timeframe, regardless of the coating. 

3 3.2.1.3 Galvanic corrosion

4 Galvanic corrosion occurs when two dissimilar metals or conductive materials are in physical 

5 contact in the presence of a conducting solution (Baboian, 2003; Hack, 1993). Under these 

6 conditions, an electrolytic cell is formed, transmitting an electrical current between an anode 

7 (i.e., less noble material) and a cathode (i.e., more noble material). Oxidation occurs at the 

8 anode, and reduction occurs at the cathode. The relative nobility of different materials has been 

9 most commonly constructed from measurements in seawater (Baboian, 2003). With certain 

10 exceptions, it is broadly applicable to other natural waters and in uncontaminated atmospheres. 

11 It is used here to infer the relative nobility of the canister materials during extended storage 

12 (e.g., steel is less noble than stainless steel, graphite, nickel, and brass). The extent of galvanic 

13 corrosion depends on potential differences between the two metals, surface area ratio of the 

14 anode and cathode, environment, reaction kinetics, corrosion products, and other factors 

15 (Baboian, 2003). In DSSs, galvanic coupling exists between steel and other more noble 

16 materials such as stainless steel, graphite, nickel, and brass. These galvanic couples can be 

17 exposed to sheltered and outdoor air environments.

18 Steel subcomponents exposed to outdoor and sheltered environments 

19 Aqueous electrolytes for subcomponents exposed to outdoor and sheltered environments are

20 present during the 60-year timeframe. Because these electrolytes could initiate steel corrosion,

21 and corrosion of steel is expected to be enhanced under galvanic coupling, loss of material due 

22 to galvanic corrosion of steel is considered to be credible in dissimilar metal couples, and 

23 therefore, aging management is required during the 60-year timeframe.

24 3.2.1.4 Microbiologically influenced corrosion

25 MIC is corrosion caused or promoted by the metabolic activity of microorganisms 

26 (Dexter, 2003). Active microbial metabolism requires water in the form of water vapor, 

27 condensation, or deliquescence, and available nutrients to support microbial activity (Horn and 

28 Meike, 1995). Biofilms can form even under radiation environments (Bruhn et al., 2009). 

29 Bacteria resistant to radiation include Micrococcus radiodurans, which can tolerate 10 kilograys 

(kGy) [106 30 rads] of irradiation. MIC is limited where relative humidity is below 90 percent and 

31 negligible for relative humidity below 60 percent (King, 2009). MIC has been found to be 

32 operable within a temperature range of −5 degrees C to 110 degrees C [23 to 230 degrees F]. 

33 Several types of microbes can exist within a biofilm. For instance, sulfate-reducing bacteria are 

34 of primary concern in wet, cool, and anoxic environments (Little and Wagner, 1996). Another 

35 type of microbe is the acid-producing bacteria, which can promote depassivation of oxide films 

36 on metals. Other types of bacteria are created by ammonia production, metal deposition, and 

37 hydrogen production (Walch and Mitchell, 1983; Little and Wagner, 1996). Although most of the 

38 evidence of MIC for metallic components is from conditions under which the metal surface is 

39 kept continuously wet, microorganisms can live in many environments, such as water, soil, and 

40 air, where aerobic bacteria (e.g., iron-manganese oxidizing bacteria, sulfur/sulfide oxidizing 

41 bacteria, methane producers, organic acid-producing bacteria), fungi, and algae can develop. 

42 This is borne out by research studies on MIC in soils (Jack et al., 1996) and in tropical 

43 environments (Caprio et al., 1995). 
3-12

1 Steel subcomponents exposed to groundwater/soil and embedded (concrete) environments

2 For soils, MIC rates for steel and iron have been correlated with the pH, oxidation reduction 

3 potential, resistivity, and water content of the soil, as well as with the type of soil. Moist, aerobic 

4 soils, where oxygen can readily reach exposed steel, show MIC rates typically in the range of 

5 0.04 to 0.2 mm/yr [2 to 8 mils/yr] (Jack et al., 1996). Anaerobic soil environments show 

6 intermediate MIC rates of steel on the order of 0.002 to 0.01 mm/yr [0.08 to 0.3 mils/yr]. Typical 

7 MIC rates of metal loss for unprotected line pipe steel in a sulfate-reducing bacteria/FeS 

8 environment are 0.2 mm/yr [8 mils/yr] for general corrosion and 0.7 mm/yr [28 mils/yr] for pitting 

9 corrosion. When steel is embedded in concrete, it can be exposed to groundwater or soil, as 

10 concrete degrades with time, which could be conducive to MIC as well. As such, MIC of steel in 

11 soil and concrete environments is considered to be credible, and therefore, aging management 

12 is required during the 60-year timeframe.

13 Steel subcomponents exposed to sheltered and outdoor environments

14 As discussed in Section 3.2.1.1, the potential to form aqueous electrolytes for subcomponents 

15 exposed to outdoor and sheltered environments is present, either from direct exposure to 

16 precipitation or by deliquescence of deposited salts. These electrolytes have the potential to 

17 support microbial activity.

18 A limited number of research studies have shown that MIC may occur on steel surfaces 

19 exposed to tropical and polluted atmospheric conditions (Caprio et al., 1995; Parra et al., 1996; 

20 Maruthamuthu et al., 2008). However, there is no operating experience of MIC degradation of 

21 steel engineering components that are exposed to environments similar to those of dry cask 

22 storage systems, where continuous exposure to a relative humidity above 90 percent is not 

23 expected. The operating experience of MIC for metallic components is largely from instances in 

24 which the metal surface was kept continuously wet. Because there is no applicable operating 

25 experience of MIC damage of steel under relevant atmospheric conditions, MIC is not 

26 considered to be credible, and therefore, aging management is not required during the 60-year 

27 timeframe.

28 Steel subcomponents exposed to demineralized water

29 The transfer cask water jackets are filled with demineralized water and drained during each 

30 loading campaign. If any bacteria are introduced during these operations, the concentration is 

31 expected to be insignificant. Microbial metabolism and growth depends upon adequate supplies 

32 of essential macro and micro nutrients. Critical nutrients, such as carbon, nitrogen, and 

33 phosphorous, must be present in appropriate concentrations (Dragun, 1988). It is expected that 

34 the concentrations of these species in demineralized water are well below the critical values. As 

35 such, MIC of steel in this environment is considered to be insignificant, and therefore, aging 

36 management is not required during the 60-year timeframe.

37 Steel subcomponents exposed to helium and embedded (neutron shielding) environments

38 Because of the limited amount of water and nutrients in the helium environments within casks 

39 and canisters, and the limited water in embedded environments, MIC of steel is not credible for 

40 the 60-year timeframe, and therefore, aging management is not required.
3-13

1 3.2.1.5 Stress corrosion cracking

2 SCC is the cracking of a metal produced by the combined action of corrosion and tensile stress 

3 (applied or residual) (Jones, 1992). SCC is highly chemical specific in that certain alloys are 

4 likely to undergo SCC only when exposed to a small number of chemical environments. SCC is 

5 the result of a combination of three factors: (1) a susceptible material, (2) exposure to a 

6 corrosive environment, and (3) tensile stresses. High-strength steels with yield strengths 

7 greater than or equal to 150,000 pounds per square inch [150 ksi] have been found to be 

8 susceptible to SCC under exposure to aqueous electrolytes, particularly when containing H2S)

9 (Jones, 2003; McMahon, 2001; EPRI, 2007). 

10 Steel subcomponents exposed to sheltered and outdoor environments

11 In DSSs, some steels with moderately high strength are used as bolting material, such as the lid 

12 bolts for the direct-load bolted cask systems. These steel subcomponents are exposed to 

13 sheltered and outdoor environments, and thus an aqueous electrolyte necessary to support 

14 SCC could be present. 

15 SCC also requires the presence of a sufficient tensile stress. Calculations using the approach 

16 proposed by Baggerly (1999) show that the stress threshold to initiate SCC of steel bolts is 

17 usually larger than 70 percent of the bolting material’s minimum yield strength, while the Electric 

18 Power Research Institute (EPRI, 2007) states that stresses near the yield strength are required 

19 to initiate SCC. The high-strength steel bolting in DSSs is expected to be loaded to stresses 

20 much lower than these SCC thresholds. For example, under normal conditions, the stress 

21 experienced by the lid bolts of bolted cask systems is primarily from the bolt preload applied to 

22 seat, or engage, the lid gaskets, and these preloads are well below the bolting material’s yield 

23 strength. Also, in the Standardized NUHOMS system, the high-strength structural bolts in the 

24 horizontal storage module (HSM) are installed “snug tight” and are not loaded close to critical 

25 stresses. 

26 Because of the low applied stresses, SCC of steel bolts exposed to sheltered and outdoor 

27 environments is not considered to be credible, and therefore, aging management is not required 

28 during the 60-year timeframe.

29 3.2.1.6 Creep

30 Creep is the time-dependent inelastic deformation that takes place at an elevated temperature 

31 and a constant stress (Gibeling, 2000). Because the deformation processes that produce creep 

32 are thermally activated, the rate of this time-dependent deformation is a strong function of the 

33 temperature. The creep rate also depends on the applied stress but does not generally vary 

34 with the environment. As a general rule of thumb, at temperatures below 0.4Tm, where Tm is the 

35 melting point of the metal in Kelvin (K), thermal activation is insufficient to produce significant 

36 creep (Cadek, 1988). With a melting point of 1,789 K (1,516 degrees C [2,760 degrees F]), 

37 temperatures of at least 716 K (443 degrees C [829 degrees F]) are required to initiate creep in 

38 steels. However, the 0.4Tm rule of thumb underestimates the minimum creep temperature for 

39 steels, as temperatures above 500 degrees C [932 degrees F] have been found to be required 

40 for creep in steels (Samuels, 1988). 
3-14

1 Steel subcomponents exposed to helium

2 The highest temperatures within the DSSs are at locations close to the fuel rods. The maximum 

3 expected temperature of fuel cladding has been estimated to be 400 degrees C [752 degrees F]

4 at the beginning of storage (Jung et. al., 2013). This cladding temperature is expected to 

5 decrease to around 266 degrees C [510 degrees F] after 20 years and to approximately 

6 127 degrees C [261 degrees F] after 60 years. These estimates depend on many factors, such 

7 as the initial heat load of the SNF. Because the fuel rods are the only heat source within the 

8 system, these temperatures provide upper temperature limits for all subcomponents. It is 

9 apparent from these temperatures that internal subcomponents will not approach the minimum 

10 500 degrees C [932 degrees F] temperature that has been found to be required for significant 

11 creep to occur in steels. Hence, creep of steel internals exposed to helium is not expected to be 

12 credible, and therefore, aging management is not required during the 60-year timeframe. 

13 Steel subcomponents exposed to sheltered, outdoor air, demineralized water, groundwater or 

14 soil, and embedded (all) environments

15 Because steel subcomponents exposed to sheltered, outdoor air, demineralized water, 

16 groundwater or soil, and embedded environments experience significantly lower temperatures 

17 than those experienced by the internal subcomponents, creep of these steel subcomponents is 

18 not considered to be credible, and therefore, aging management is not required during the 

19 60-year timeframe. 

20 3.2.1.7 Fatigue

21 Fatigue is the progressive structural damage that occurs when a metal is subjected to cyclic 

22 loading (Hoeppner, 1996). Because spent fuel storage is a static application, cyclic loading by a 

23 purely mechanical means is largely limited to transfer cask lifting trunnions, which are loaded 

24 each time a canister is moved from the spent fuel pool to the dry storage pad. Other 

25 subcomponents, however, could experience cyclic loads due to thermal effects. 

26 The reviewer should review all fatigue analyses contained in the applicant’s design basis 

27 documents to determine whether the renewal application adequately addresses the implications

28 of extending the operating period to 60 years. This re-examination of the original fatigue 

29 analyses are defined as TLAAs. 

30 As described in greater detail in Chapter 5 of this report, the reviewer should review the design 

31 codes and standards to identify any required fatigue analyses and ensure that the applicant 

32 addresses those analyses with a TLAA. For example, components that were designed in 

33 accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel 

34 Code (ASME Code) Section III, Division 1, Subsections NB or NC (ASME, 2007a) were

35 evaluated for the effects of cyclic loading per subparagraphs NB-3222.4 and NC-3219.2, 

36 respectively. Also, the designs of some steel support structures may be performed in 

37 accordance with the American Institute of Steel Construction (AISC) Standard 360, 

38 “Specifications for Structural Steel Buildings” (AISC, 2010). Appendix 3 of AISC 360, “Design 

39 for Fatigue,” provides criteria for the evaluation of cyclic loading. 

40 The staff’s guidance for the review of TLAAs is provided in NUREG–1927, Revision 1, and 

41 Chapter 5 of this report. In its evaluation of a TLAA, an applicant may conclude that an analysis 

42 can no longer support a determination that aging will not adversely affect an important-to-safety 
3-15

1 function in the 60-year timeframe of the period of extended operation. In that case, the 

2 applicant may manage the aging of the associated SCC with an AMP.

3 The AMR tables in Chapter 4 recommend that applicants address any applicable TLAAs

4 associated with components with a structural function. If no fatigue analysis was performed in 

5 support of the component design, no action is required of the applicant. 

6 3.2.1.8 Thermal aging

7 The microstructures of most steels will change, given sufficient time at temperature, and this 

8 can affect mechanical properties. This process is commonly called thermal aging. The effect of 

9 thermal aging will depend on the time at temperature and the microstructure and carbon content 

10 of the steel subcomponents. 

11 Steel subcomponents exposed to helium

12 The maximum expected temperature of fuel cladding has been estimated to be 400 degrees C 

13 [752 degrees F] at the beginning of storage (Jung et. al., 2013). This upper-bound cladding 

14 temperature is expected to decrease to around 266 degrees C [510 degrees F] after 20 years 

15 and to approximately 127 degrees C [261 degrees F] after 60 years. Although the temperature 

16 of steel components within the cask internal environment will be lower than that of the fuel 

17 cladding, consideration of the cladding temperatures provides a conservative estimate of the 

18 effects of thermal aging. 

19 Carbon steels in the normalized condition (ferrite/pearlite microstructures) are commonly used 

20 in the petroleum and chemical industry with exposure temperatures similar to those in DSS

21 internal environments, approximately 400 degrees C [752 degrees F] and lower

22 (ASM International, 1998). ASME Code Section II, Part D, provides allowable operating 

23 stresses for carbon steels at these temperatures (ASME, 2007b). 

24 The ASME Code also provides for the use of hardened (quenched and tempered) alloy steels at 

25 temperatures typically expected within storage systems during the 20- to 60-year period of 

26 extended operation. For example, ASME type SA-537 Grade 2 alloy steel receives a tempering 

27 heat treatment of at least 595 degrees C [1,100 degrees F] following quenching, and the 

28 ASME Code provides allowable operating stresses up to 371 degrees C [700 degrees F]. This 

29 compares to the estimated upper-bound 266 degrees C [510 degrees F] temperature during the 

30 period of extended operation. Some hardened alloy steels can experience reductions in 

31 fracture toughness when tempered at temperatures greater than 200 degrees C 

32 [392 degrees F]. The degree of the reduction in toughness depends on the carbon content and 

33 the tempering conditions that were employed during processing (Krauss, 2005). 

34 The effects of elevated storage temperatures on material properties are evaluated during the 

35 initial license application (typically first 20 years of storage). Although the temperatures inside 

36 the canister after 20 years may still have the capacity to alter mechanical properties, it is likely 

37 that the steel tempering that occurs during manufacture and the higher temperatures present 

38 during the initial storage period would dominate any effects of tempering at the lower 

39 temperatures during the period of extended operation. 

40 It can thus be concluded that thermal aging generally is not expected to produce degradation of 

41 the mechanical properties of steels in the period of extended operation, and therefore, aging 
3-16

1 management is not required during the 60-year timeframe. Nevertheless, the reviewer should 

2 verify this conclusion on a case-by-case basis.

3 Steel subcomponents exposed to sheltered, outdoor air, demineralized water, groundwater or 

4 soil, and embedded (all) environments

5 As stated above, undesired material property changes due to tempering of hardened steels

6 could occur at temperatures greater than 200 degrees C [392 degrees F]. The temperatures of 

7 steel subcomponents exposed to sheltered, outdoor air, demineralized water, groundwater or 

8 soil, and embedded environments are bounded by the stainless steel canister shell temperature, 

9 because these subcomponents are located farther away from the fuel. Time-temperature 

10 profiles calculated for the stainless steel canister shell estimate that the peak temperature is 

11 below 200 degrees C [392 degrees F] (EPRI, 2006; Meyer et al., 2013). Because the peak 

12 temperatures for steel subcomponents exposed to sheltered, outdoor air, demineralized water, 

13 and embedded environments are below the temperature required to cause reductions in 

14 toughness, thermal aging is not considered to be credible for these subcomponents, and 

15 therefore, aging management is not required during the 60-year timeframe. 

16 3.2.1.9 Radiation embrittlement

17 Embrittlement of metals may occur under exposure to neutron radiation. Depending on the 

18 neutron fluence, radiation can cause changes in mechanical properties, such as loss of ductility, 

19 reduced fracture toughness, and decreased resistance to cracking.

20 Neutron irradiation has the potential to increase the tensile and yield strength and decrease the 

21 toughness of carbon and alloy steels (Nikolaev et al., 2002). Neutron fluence levels greater 

than 1019 neutrons/square centimeter (n/cm2

) [6.5 × 1019 n/in2 22 ] are required to produce a 

23 measureable degradation of the mechanical properties (Nikolaev et al., 2002; Odette and 

24 Lucas, 2001). 

For dry cask storage, a neutron flux of 104

–106 n/cm2-s [6.5 × 104–6.5 × 106 n/in2 25 -s] is typical 

26 (Sindelar et al., 2011). At these flux levels, the accumulated neutron fluence after 60 years is 

about 1013–1015 n/cm2 [6.5 × 1013–6.5 × 1015 n/in2 27 ]. To verify the conservatism of this estimate, 

28 the NRC staff performed an independent calculation of the maximum potential accumulated 

29 neutron fluence on DSS components. The staff considered components most directly exposed 

30 to the radiation source (middle of the fuel basket) and assumed fuel is loaded immediately after 

31 it is removed from the reactor vessel and stored for 100 years. To further provide a bounding 

32 estimate, the staff assumed a cask design that uses 40 Westinghouse 17×17 PWR fuel 

33 assembles with an average burnup of 70 GWd/MTU and 4.0 fuel enrichment. The staff 

34 calculated the neutron source term for neutrons with energy at or greater than 1 MeV using the 

35 Origen/Arp computer code of the SCALE 6.1 computer code system. At this location, the total 

accumulated neutron fluence after 100 years of storage was calculated to be 2.63 × 1016 n/cm2 36

[1.70 × 1017 n/in2 37 ]. This worst-case estimate is greater than that calculated using the flux levels 

38 reported in Sindelar, however, the NRC estimated fluence level is still three orders of magnitude 

39 below the levels reported to degrade the fracture resistance of carbon and alloy steels.

40 Thus, radiation embrittlement of steel exposed to any environment is not a credible aging 

41 mechanism, and therefore, aging management is not required during the 60 year timeframe.
3-17

1 3.2.1.10 Stress relaxation

2 Stress relaxation of bolting or other tightening subcomponents is the steady loss of elastic 

3 stress in a loaded part due to atomic movement at elevated temperature (Earthman, 2000). It 

4 results in a loss of clamping forces or preload in a heavily loaded joint. In the stress relaxation 

5 process, the total strain is constant and the stress reduction at constant temperature occurs as 

6 an elastic strain is converted to an inelastic strain. Stress relaxation is a strong function of 

7 temperature and bolt material. It also depends on geometry of the bolt and thread quality 

8 (Sachs and Evans, 1973). It decreases with time, as the tensile stress in the bolt decreases 

9 (Kulak et al., 2001). Steel bolting is used in several DSS applications in sheltered and outdoor 

10 environments, such as in the NUHOMS canister support structure and the HI-STORM overpack 

11 lid. 

12 Steel subcomponents exposed to sheltered environments

13 Bickford (2008) demonstrated that the residual stress of carbon steel bolts due to relaxation is 

14 about 85 percent of the initial applied stress at temperatures greater than about 100 degrees C 

15 [212 degrees F]. Meyer et al. (2013) show that the external surface temperature of storage 

16 canisters can be greater than 200 degrees C [392 degrees F] at the beginning of the storage 

17 period. Thus, stress relaxation of steel bolting exposed to sheltered environments adjacent to 

18 the canister is considered to be credible, and therefore, aging management is required during 

19 the 60-year timeframe. 

20 Steel subcomponents exposed to outdoor environments

21 Bolting in outdoor environments is not considered to be exposed to sufficiently high 

22 temperatures to cause stress relaxation. Similarly, transfer cask bolting in indoor/outdoor 

23 environments is not considered to be exposed to high temperatures for a sufficient amount of 

24 time to cause stress relaxation. Thus, for steel bolting exposed to outdoor environments, aging 

25 management is not required during the 60-year timeframe.

26 3.2.1.11 Wear

27 Rolling contact wear results from the repeated mechanical stressing of the surface of a body 

28 rolling on another body (Blau, 1992). For the HI-TRAC transfer cask exposed to indoor and 

29 outdoor air, ASME SA36 steel is used to construct the transfer lid wheel track, which could 

30 experience rolling contact during SNF loading and unloading operations. Thus, wear of these 

31 steel subcomponents is considered to be credible, and therefore, aging management is required 

32 during the 60-year timeframe.

33 3.2.2 Stainless steel

34 Austenitic, ferritic, martensitic, duplex, and precipitation-hardened stainless steels are used in 

35 constructing DSS subcomponents. They are exposed to outdoor, sheltered, embedded, helium, 

36 and demineralized water environments. Some stainless steels are used to construct the 

37 transfer cask, which is predominately exposed to an indoor environment or otherwise encased 

38 without direct air ingress, except for short periods of air exposure during transfer operations. 

39 For such air-indoor/outdoor environments, the aging mechanisms from aqueous corrosion 

40 processes are expected to be bound by the outdoor environment, because it is more corrosive.

41 As such, the indoor air environment is only discussed separately for the evaluation of SCC, 
3-18

1 where periodic rinsing of the transfer cask external surfaces is expected to minimize halide 

2 deposition. 

3 3.2.2.1 General corrosion

4 Stainless steels exhibit passive behavior in all DSS environments, resulting in negligible general 

5 corrosion rates (Grubb, 2005). As such, general corrosion of stainless steel exposed to all 

6 environments is not considered to be credible, and therefore, aging management is not required 

7 during the 60-year timeframe. 

8 3.2.2.2 Pitting and crevice corrosion

9 As discussed in Section 3.2.1.2, pitting corrosion is a localized form of corrosion that is confined 

10 to a point or small area of a metal surface (Frankel, 2003), and crevice corrosion occurs in a 

11 wetted environment when a crevice exists that allows a corrosive environment to develop in a 

12 component (Kelly, 2003). In DSSs, crevice corrosion may occur (i) where the canister contacts 

13 the support rails for horizontal canister designs and (ii) between canister and guide rails or the 

14 support pedestal in some vertical designs. Stainless steels are susceptible to pitting and 

15 crevice corrosion, with chloride being the most common agent for initiation (Grubb et al., 2005). 

16 Other halides, notably bromides, and hypochlorites are also initiation agents (EPRI, 2007). 

17 Stainless steel subcomponents exposed to outdoor and sheltered environments

18 As discussed in Section 3.2.1.1, the potential to form aqueous electrolytes for subcomponents 

19 exposed to outdoor and sheltered environments is present, either via direct exposure to 

20 precipitation or by deliquescence of deposited salts. These electrolytes could be conducive to 

21 pitting and crevice corrosion of stainless steel. Atmospheric corrosion of stainless steels 

22 typically proceeds in the form of localized corrosion (Cook et al., 2010; Shirai et al., 2011; 

23 Tani et al., 2009). However, experimentally measured penetration rates for pitting and crevice 

24 corrosion are quite low. Stainless steel exposed to a saturated NaCl steam mist at 

25 60 degrees C [140 degrees F] and 95 percent relative humidity (NWTRB, 2010) yielded 

26 maximum penetration rates of 0.02 mm/yr [8 mils/yr] for pitting and 0.03 mm/yr [11 mils/yr] for 

27 crevice corrosion. These maximum rates suggest that penetration of a 15-mm [0.59-in]-thick 

28 canister wall by pitting or crevice corrosion would require 750 years and 495 years, respectively. 

29 Davison et al. (1987) reported pitting penetration of 0.028 mm [1.1 mils] after 15 years, which 

30 yields a penetration rate of 0.0019 mm/yr [0.075 mils/yr]. Using the penetration depth versus 

31 time equation in Eq. (3.2-1) from NRC (2014): 

d = At-n and n= 0.33 to 0.5, (3.2-1)

32 the penetration rate in Davison et al. (1987), and n = 0.5 yields a penetration time for a 15-mm 

33 [0.59-in]-thick canister wall of 19,000 years. Based on these penetration rates, the canister wall 

34 would not be penetrated in the 60-year timeframe. The rate of pit propagation can be much 

35 higher in aggressive environments. Morrison (1972) reported pit penetrations exceeding 

36 0.5 mm [20 mils] in 304 and 316 stainless steels after a 28-month exposure at the Kennedy 

37 Space Center, Florida. However, the pitting rates measured under aggressive marine 

38 environments would require more than 250 years to penetrate 12.7-mm [0.5-in]-thick stainless 

39 steel. Hence, neither pitting nor crevice corrosion itself is expected to produce damage to the 

40 stainless steel subcomponents in the 60-year timeframe. 
3-19

1 However, both pitting and crevice corrosion are known to be precursors to SCC. He et al. 

2 (2014) observed that all the SCC cracks started at the bottom of the pits. Therefore, pitting and 

3 crevice corrosion are also considered to be credible during the 60-year timeframe, due to their 

4 role as precursors to atmospheric SCC, and aging management is required accordingly.

5 Stainless steel subcomponents exposed to helium, demineralized water, and embedded (all)

6 environments

7 Stainless steel exposed to helium and demineralized water is not susceptible to pitting and 

8 crevice corrosion due to the lack of halides. Because of limited water and oxygen, stainless 

9 steel is also not susceptible to pitting and crevice corrosion in embedded environments. As 

10 such, pitting and crevice corrosion of stainless steel exposed to helium, demineralized water, 

11 and embedded environments are not considered to be credible, and therefore, aging 

12 management is not required during the 60-year timeframe. 

13 3.2.2.3 Galvanic corrosion

14 As discussed in Section 3.2.1.3, galvanic corrosion occurs when two dissimilar metals or 

15 conductive materials are in physical contact in the presence of a conducting solution 

16 (Baboian, 2003; Hack, 1993). In DSSs, graphite is used to lubricate stainless steel 

17 subcomponents such as the stainless steel upper trunnion for the TN-68 bolted cask and the 

18 interface between the NUHOMS canister shell and support structure, resulting in galvanic 

19 contact between stainless steel and graphite. Because graphite is strongly cathodic and the 

20 contact is close, the galvanic coupling effect is expected to be strong. These galvanic couples 

21 are exposed to sheltered and outdoor environments.

22 Because these electrolytes conducive to galvanic corrosion exist in both sheltered and outdoor 

23 environments, galvanic corrosion of stainless steel in contact with graphite lubricants is 

24 considered to be credible, and therefore, aging management is required during the 60-year 

25 timeframe.

26 3.2.2.4 Microbiologically influenced corrosion

27 As discussed in Section 3.2.1.4, MIC is caused or promoted by the metabolic activity of 

28 microorganisms (Dexter, 2003). Microorganisms can live in many environments, such as water, 

29 soil, and air, where aerobic bacteria (e.g., iron-manganese oxidizing bacteria, sulfur/sulfide 

30 oxidizing bacteria, methane producers, organic acid-producing bacteria), fungi, and algae 

31 can develop.

32 Stainless steel subcomponents exposed to sheltered and outdoor environments

33 As discussed in Section 3.2.1.1, the potential to form aqueous electrolytes for subcomponents 

34 exposed to outdoor and sheltered environments is present during the 60-year timeframe, either 

35 from direct exposure to precipitation or by deliquescence of deposited salts. These electrolytes 

36 could support microbial activity; however, there has not yet been any operating experience of

37 MIC in atmospheric environments where stainless steel surfaces are only intermittently wetted. 

38 Due to the absence of any operating experience of MIC damage of stainless steel under 

39 atmospheric conditions, MIC is not considered to be credible, and therefore, aging management 

40 is not required during the 60-year timeframe.

41
3-20

1 Stainless steel subcomponents exposed to demineralized water

2 The transfer cask water jackets are filled with demineralized water and drained during each 

3 loading campaign. If any bacteria are introduced during these operations, the concentration is 

4 expected to be insignificant. Microbial metabolism and growth depends upon adequate supplies 

5 of essential macro and micro nutrients. Critical nutrients such as carbon, nitrogen, and 

6 phosphorous must be present in appropriate concentrations (Dragun, 1988). It is expected that 

7 the concentrations of these species in demineralized water are well below the critical values. As 

8 such, MIC of stainless steel exposed to demineralized water is not considered to be credible, 

9 and therefore, aging management is not required during the 60-year timeframe.

10 Stainless steel subcomponents exposed to helium and embedded (all) environments

11 Because of the limited amount of water and nutrients in the helium environments within casks 

12 and canisters, and the limited water in embedded environments, MIC of stainless steel is not 

13 credible for the 60-year timeframe, and therefore, aging management is not required.

14 3.2.2.5 Stress corrosion cracking 

15 SCC is the cracking of a metal produced by the combined action of corrosion and tensile stress 

16 and is highly chemical specific (Jones, 1992, 2003). Most ferritic and duplex stainless steels are 

17 either immune or highly resistant to SCC; however, all austenitic grades, especially Types 304, 

18 304L, 304LN, 316, 316L, and 316LN, have long been reported in the literature to be susceptible 

19 to chloride-induced SCC in the normal wrought condition (Grubb et al., 2005; Morgan, 1980; 

20 Kain, 1990). This susceptibility increases when the material is sensitized (He et al., 2014). In 

21 the welded condition, the heat-affected zone, which is a thin band located adjacent to the weld, 

22 can be sensitized by the precipitation of carbides that extract chromium out of the metal matrix. 

23 The Electric Power Research Institute (EPRI, 2005, 2006) and the Nuclear Decommissioning 

24 Authority in the United Kingdom (Nuclear Decommissioning Authority, 2007) published review

25 reports on SCC of stainless steel. More recently, the NRC released Information Notice 

26 (IN) 2012-20, “Potential for Chloride-Induced Stress Corrosion Cracking of Austenitic Stainless 

27 Steel and Maintenance of Dry Cask Storage Systems” (NRC, 2012). The IN describes several 

28 incidents in commercial nuclear power plants where SCC of austenitic stainless steel 

29 components was attributed to atmospheric chloride exposure (NRC, 1999, 2010c; FPL, 2005; 

30 Alexander et al., 2010). These events involved components such as emergency core cooling 

31 system piping, SNF pool cooling lines, and outdoor tanks. The IN notes that chlorides may be 

32 present in the atmosphere, not only in marine environments but also near cooling towers, salted 

33 roads, or other locations. The susceptibility of austenitic stainless steels to SCC tends to 

34 increase as the chloride concentration in the solution increases, but the level of chlorides 

35 required to produce SCC is very low and is dependent on the type of chloride salts present. 

36 The material is more resistant to SCC in NaCl solutions but cracks readily in MgCl2 solutions 

37 (Grubb et al., 2005). Increased temperature and the presence of oxygen tend to aggravate 

38 chloride-induced SCC.

39 Stainless steel subcomponents exposed to outdoor and sheltered environments

40 As discussed in Section 3.2.1.1, the potential to form electrolytes for subcomponents exposed 

41 to outdoor and sheltered environments is present, either via direct exposure to precipitation or 

42 by deliquescence of deposited salts. These electrolytes could be conducive to SCC of stainless 

43 steel. SCC also requires the presence of a tensile stress, which commonly exists at welds 
3-21

1 originating from fabrication processes, contacts between components, and bolted structures. 

2 Fuhr et al. (2013) stated that stresses well below yield can cause SCC and the required stress

3 for SCC initiation decreases as chloride concentration and temperature increase. SCC tests 

4 were performed with Type 304L C-ring specimens strained to 0.4 or 1.5 percent (He et al., 

5 2014). At the strain of 0.4 percent, the stress on the C-ring specimen was approximately equal 

6 to the material yield stress. SCC initiation was observed on specimens deposited with 1 or 

10 grams/square meter (g/m2

) [0.003 or 0.03 ounces/square foot (oz/ft2 7 )] of simulated sea salt at 

8 both strain levels. Constant load tensile tests were performed on Type 304 between 0.5 and 

9 1.75 times the material yield stress (Mayuzumi et al., 2008). Surface chloride concentration was 

estimated to exceed 10 g/m2 [0.03 oz/ft2 10 ], while test conditions were 80 degrees C 

11 [176 degrees F] at 35 percent relative humidity. Specimens failed at the stress level of 

12 0.5 times the yield stress. 

13 For DSS subcomponents, the stainless steel canister shell is welded. Welds also exist in other 

14 subcomponents, such as the cover plates for the vent and drain ports, grapple ring and grapple 

15 support, and the Nitronic 60 support rail plate of the NUHOMS system used to support the 

16 canister. Fuhr et al. (2013) concluded that the driving stress for SCC of the welded canister is 

17 expected to be weld residual stress, considering that the applied stresses are low and residual 

18 compressive stresses are believed to be present on the shell outer diameter due to rolling. 

19 Their calculations indicate that residual stresses parallel to the weld are tensile through-wall and 

20 significantly above the original yield strength of the base metal, while those transverse to the 

21 weld are either compressive along the outer canister surface or slightly tensile on the outer 

22 diameter but compressive along the midwall. Based on these calculated residual weld stresses, 

23 it was concluded that through-wall SCC is most likely to occur transverse to the weld direction. 

24 Weld residual stress modeling conducted by the NRC (2013) also indicates that through-wall 

25 tensile stresses of sufficient magnitude to support SCC are likely to exist in the weld 

26 heat-affected zone.

27 Because sufficient weld residual stresses and more susceptible material conditions are present 

28 near the welds, and aqueous electrolytes conducive to SCC are present in sheltered and 

29 outdoor environments, the potential for SCC of the welds in the canister shell and other 

30 stainless steel subcomponents is present in the 60-year timeframe. Additionally, the SCC 

31 initiation times are relatively short (NWTRB, 2010) with reported crack growth rates of austenitic 

32 stainless steels at the weld heat-affected zones ranging from 0.1 mm/yr [3.9 mils/yr]

33 (Hosler, 2010) to 0.67 mm/yr [26.1 mils/yr] (Basson and Wicker, 2002). As a result, 

34 through-wall penetration could occur during the 60-year timeframe. This is consistent with the

35 observation of outer-diameter-initiated through-wall SCC in stainless steel piping after 20 to 

36 30 years of exposure in marine environments (Fuhr et al., 2013). As such, atmospheric SCC of 

37 stainless steel subcomponents with welds exposed to sheltered and outdoor air is considered to 

38 be credible, and therefore, aging management is required during the 60-year timeframe. 

39 For weld-free austenitic stainless steel subcomponents or regions away from welds, such as the 

40 canister body, atmospheric SCC is a likely aging mechanism if sufficient stress exists. Its 

41 significance and corresponding aging management requirement will need to be assessed case

42 by case, based on applied and residual stresses, operating temperatures, and the presence of 

43 chlorides in the environment.

44

45
3-22

1 Stainless steel subcomponents exposed to indoor/outdoor environments and 

2 demineralized water

3 Stainless steel transfer casks are exposed to indoor environments during their storage between 

4 cask loading campaigns, and thus an aqueous electrolyte is not likely to be present on the 

5 transfer cask external surfaces for extended periods. Also, the transfer cask external surfaces 

6 are periodically rinsed with demineralized water as they are removed from the spent fuel pool, 

7 which would be expected to remove any halides present. As a result, SCC is not considered to 

8 be a credible degradation mechanism. In the demineralized water environments of transfer 

9 cask neutron shields, SCC is also not considered to be a credible degradation mechanism

10 because of the lack of halides. Therefore, aging management of stainless steel subcomponents 

11 exposed to an indoor environment and demineralized water is not required during the 60-year 

12 timeframe.

13 Stainless steel subcomponents exposed to helium and embedded (all) environments

14 Because of the lack of halides and the small amount of water in helium and embedded 

15 environments, SCC of stainless steel is not considered to be credible. Therefore, aging 

16 management of stainless steel subcomponents exposed to helium and embedded environments

17 is not required during the 60-year timeframe.

18 3.2.2.6 Creep

19 As discussed in Section 3.2.1.6, as a general rule of thumb, thermal activation is insufficient to 

20 produce significant creep at temperatures below 0.4Tm, where Tm is the melting point of the 

21 metal in Kelvin (Cadek, 1988). The term “stainless steel” covers a wide range of compositions 

22 and microstructures, including austenitic, ferritic, martensitic, duplex, and precipitation 

23 hardening stainless steels. This discussion will focus on the austenitic or 300 series stainless 

24 steels, because they are most commonly used in DSSs and have the lowest melting point and 

25 minimum creep temperature. With a melting point of 1,698 K [1,425 degrees C; 

26 2,597 degrees F], temperatures of at least 679 K [406 degrees C; 763 degrees F] are required 

27 to initiate creep in the austenitic stainless steels. 

28 Stainless steel eubcomponents Exposed to helium

29 The highest temperatures within the DSSs are at locations close to the fuel rods where the 

30 environment is helium. The maximum expected temperature of fuel cladding has been 

31 estimated to be 400 degrees C [752 degrees F] at the beginning of storage (Jung et. al., 2013). 

32 This cladding temperature is expected to decrease to around 266 degrees C [510 degrees F] 

33 after 20 years and to approximately 127 degrees C [261 degrees F] after 60 years. These 

34 estimates depend on many factors, such as the initial heat load of the SNF. Because the fuel 

35 rods are the only heat source within the canister, these temperatures provide upper temperature 

36 limits for all subcomponents within the canister. It is apparent from these temperatures that 

37 subcomponents within the canister will not reach the 406 degrees C [763 degrees F] minimum 

38 temperature that is required for significant creep to occur in austenitic stainless steels. 

39 Similarly, significant creep would also not be expected to occur in the other classes of stainless 

40 steel, which all have higher minimum creep temperatures. Hence, creep of stainless steel 

41 internals exposed to helium is not credible, and therefore, aging management is not required 

42 during the 60-year timeframe. 
3-23

1 Stainless steel subcomponents exposed to sheltered, outdoor air, demineralized water, and 

2 embedded (all) environments

3 Because stainless steel subcomponents exposed to sheltered, outdoor air, demineralized water, 

4 and embedded environments experience significantly lower temperatures than those 

5 experienced by the internal subcomponents, creep of these stainless steel subcomponents is 

6 not considered to be credible, and therefore, aging management is not required during the 

7 60-year timeframe.

8 3.2.2.7 Fatigue

9 As discussed previously in Section 3.2.1.7, because spent fuel storage is a static application, 

10 cyclic loading by a purely mechanical means is largely limited to transfer cask lifting trunnions, 

11 which are loaded each time a canister is moved from the spent fuel pool to the dry storage pad. 

12 Other subcomponents, however, could experience cyclic loads due to thermal effects, such as 

13 those caused by daily and seasonal fluctuations in the temperature of the external environment.

14 The NRC reviewer should review the fatigue analyses contained in the applicant’s original 

15 design-basis documents to determine whether the renewal application adequately addresses 

16 the implications of extending the operating period to 60 years. This reexamination of the 

17 original fatigue analyses would typically be defined as TLAAs. 

18 As described in greater detail in Chapter 5 of this report, the reviewer should review the design 

19 codes and standards to identify any required fatigue analyses and ensure that the applicant 

20 addresses those analyses with a TLAA. For example, components that were designed in 

21 accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel 

22 Code (ASME Code) Section III, Division 1, Subsections NB or NC (ASME, 2007a) were 

23 evaluated for the effects of cyclic loading per subparagraphs NB 3222.4 and NC 3219.2, 

24 respectively.

25 The staff’s guidance for the review of TLAAs is provided in NUREG–1927, Revision 1, and 

26 Chapter 5 of this report. In its evaluation of a TLAA, an applicant may conclude that an analysis 

27 can no longer support a determination that aging will not adversely affect an important-to-safety 

28 function in the 60-year timeframe of the period of extended operation. In that case, the 

29 applicant may manage the aging of the associated SCC with an AMP.

30 The AMR tables in Chapter 4 recommend that applicants address any applicable TLAAs 

31 associated with components with a structural function. If no fatigue analysis was performed in 

32 support of the component design, no action is required of the applicant.

33 3.2.2.8 Thermal aging

34 The microstructures of most stainless steels will change, given sufficient time at temperature, 

35 and these changes may alter the material’s strength and fracture toughness. This process is 

36 commonly called thermal aging. For stainless steel subcomponents, the thermal aging process 

37 differs for welded and nonwelded subcomponents.

38 Welded austenitic stainless steel subcomponents exposed to helium

39 The ferrite present in austenitic stainless steel welds can transform by spinodal decomposition 

40 to form Fe-rich alpha and Cr-rich alpha prime phases, and further aging can produce an 
3-24

1 intermetallic G-phase. The spinodal decomposition and the formation of the intermetallic 

2 G-phase takes place during extended exposure to temperatures between 300 and 

3 400 degrees C [572 and 752 degrees F] (Alexander and Nanstad, 1995; Chandra et al., 2012). 

4 The maximum expected temperature of fuel cladding has been estimated to be 400 degrees C 

5 [752 degrees F] at the beginning of storage (Jung et. al., 2013). This cladding temperature is 

6 expected to decrease to around 266 degrees C [510 degrees F] after 20 years and to 

7 approximately 127 degrees C [261 degrees F] after 60 years. Based on these temperature 

8 estimates, subcomponents located inside the canister and near the fuel could be above the 

9 300 degrees C [572 degrees F] minimum temperature required for these phase changes. 

10 Because the phase transformations take place only within the ferrite phase, they increase the 

11 hardness and reduce the toughness of the ferrite phase but do not alter the mechanical 

12 properties of the austenite phase. Hence, the degree of embrittlement of a weld will depend on 

13 a number of factors, including the amount and distribution of ferrite present in the weld and the 

14 time spent within the 300 to 400 degrees C [572 and 752 degrees F] temperature range. 

15 Based on Charpy impact toughness testing of cast duplex stainless steels, Kim and Kim (1998) 

16 concluded that ferrite levels above 15 percent are required for significant embrittlement, 

17 because ferrite resides in discrete islands below this level and does not provide a continuous 

18 low-toughness fracture path. Because most welds contain around 4 to 15 percent ferrite 

19 (Gavendra et al., 1996), substantial embrittlement of austenitic stainless steel welds is not 

20 expected. Gavendra et al. (1996) in NUREG/CR-6428, “Effects of Thermal Aging on Fracture 

21 Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds,” concluded that thermal 

22 aging produced moderate decreases (no more than 25 percent) in the upper shelf Charpy 

23 impact energy and relatively small decreases in the fracture toughness of a wide range of 

24 austenitic welds. Although the phase changes associated with thermal embrittlement of 

25 austenitic stainless steel welds could take place in subcomponents near the fuel within the 

26 60-year timeframe, the minor reductions in fracture toughness that would be produced in the 

27 weld indicate that this is not a credible aging mechanism for subcomponents in proximity to the 

28 fuel rods, and therefore, aging management is not required.

29 Subcomponents near the internal wall of a canister or cask would experience temperatures 

30 lower than those close to the fuel rods. Time-temperature profiles calculated for a canister 

31 surface (EPRI, 2006; Meyer et al., 2013) suggest that maximum canister temperatures would be 

32 well below the 300 degrees C [572 degrees F] minimum temperature required for the embrittling 

33 phase changes. Hence, thermal aging would not produce any degradation in these 

34 subcomponents constructed from austenitic stainless steel, and therefore, aging management is 

35 not required during the 60-year timeframe.

36 Nonwelded austenitic stainless steel subcomponents exposed to helium

37 Because the phase changes described previously occur only within the ferrite-containing, 

38 heat-affected zone of a weld, embrittlement will not occur in austenitic stainless steel 

39 subcomponents that do not contain a weld. The only significant thermal aging possible in 

40 nonwelded austenitic stainless steels would be a decrease in strength due to a decrease in 

41 dislocation density, recrystallization, and an increase in grain size. These processes occur 

42 during annealing at temperatures above 1,000 degrees C [1,832 degrees F]. The temperatures 

43 of less than 400 degrees C [752 degrees F] that will be experienced by cask internal 

44 subcomponents will not degrade nonwelded austenitic stainless steels. Thus, thermal aging of 

45 nonwelded austenitic stainless steel is not credible, and therefore, aging management is not 

46 required during the 60-year timeframe.
3-25

1 Precipitation-hardened martensitic stainless steel subcomponents exposed to helium

2 Type 17-4 precipitation-hardened (PH) martensitic stainless steel with Cu and Nb additions is 

3 used to construct some fuel basket subcomponents. Operating experience has shown that this 

4 material is susceptible to thermal embrittlement, in both welded and nonwelded conditions, at 

5 temperatures above 243 degrees C [470 degrees F] (Andresen et al., 2007; Olender et al., 

6 2015; NRC, 2007). The embrittlement mechanism arises from an intra-granular decomposition 

7 of the martensitic matrix into two phases, α and α’, which are rich in iron and chromium, 

8 respectively, and formation of copper rich ε-phase upon further aging. This process leads to an 

9 increase in hardness, but decrease in fracture toughness. Olender et al. (2015) reviewed 

10 reactor operating experience with 17-4 PH stainless steels. Susceptibility to thermal 

11 embrittlement is dependent on several factors including the alloy composition within the 

12 allowable specifications, the initial heat treatment and the operating temperature. For operating 

13 temperatures between 243 and 316 degrees C [470 to 600 degrees F] Olender et al (2015) 

14 recommends an evaluation of conditions on a per-component basis considering operating 

15 temperature, exposure time, operating environment, stress levels, and material composition. 

16 Above 316 degrees C [600 degrees F] the use of 17-4 PH stainless steel in any condition is not 

17 recommended. Subcomponents located inside the canister and near the fuel could be above 

18 the temperature threshold for thermal aging. As such, thermal aging of Type 17-4 PH stainless 

19 steel is considered to be credible.

20 Although the above generic evaluation identifies thermal aging of Type 17-4 PH stainless steel 

21 as a credible aging mechanism, the degree of embrittlement of a specific SSC will depend on 

22 the service temperature and time duration, as well as the initial heat treatment condition of the 

23 SSC. As such, a review of the thermal aging effects should be performed on a case-by-case 

24 basis for all subcomponents constructed from Type 17-4 PH stainless steel. The reviewer 

25 should ensure that the application provides a bounding analysis to show that reduction in 

26 mechanical properties due to thermal aging is not expected to compromise the SSC’s intended 

27 function during the period of extended operation.

28 Stainless steel subcomponents exposed to sheltered, outdoor, demineralized water, and 

29 embedded (all) environments

30 Because the peak temperatures for stainless steel subcomponents exposed to sheltered, 

31 outdoor air, demineralized water, and embedded environments are below the temperature 

32 required for the phase changes associated with thermal embrittlement of stainless steels, 

33 thermal aging is not considered to be credible for these subcomponents, and therefore, aging 

34 management is not required during the 60-year timeframe.

35 3.2.2.9 Radiation embrittlement

36 Embrittlement of metals may occur under exposure to neutron radiation. Depending on the 

37 neutron fluence, radiation can cause changes in stainless steel mechanical properties, such as 

38 loss of ductility, fracture toughness, and resistance to cracking (Was et al., 2006).

39 Cracking has been observed in boiling-water reactor oxygenated water at fluences above 

2 to 5 × 1020 n/cm2 [1.3 to 3.2 × 1021 n/in2 40 ] (Was et al., 2006). Gamble (2006) found that neutron 

fluence levels greater than 1 × 1020 n/cm2 [6.5 × 1020 n/in2 41 ] are required to produce 

42 measureable degradation of the mechanical properties. Caskey et al. (1990) also indicates that 

neutron fluence levels of up to 2 × 1021 n/cm2 [1.3 × 1022 n/in2 43 ] were not found to enhance 

44 SCC susceptibility. 
3-26

1 As discussed in Section 3.2.1.9 of this report, the maximum potential accumulated neutron 

fluence on DSS components after 100 years was calculated to be 2.63 × 1016 n/cm2 2

[1.70 × 1017 n/in2 3 ]. This fluence level is four orders of magnitude below the level that would 

4 degrade the mechanical properties of stainless steels. As such, radiation embrittlement of 

5 stainless steel exposed to any environment is not credible.

6 3.2.2.10 Stress relaxation

7 In DSSs, some stainless steel bolts or screws are used in applications exposed to sheltered and 

8 outdoor environments. Section 3.2.1.10 explained that stress relaxation of bolting is the steady 

9 loss of stress due to atomic movement at elevated temperature in a loaded part with dimensions 

10 that are fixed (Earthman, 2000). The loss of initial applied stress in austenitic stainless steel 

11 bolting due to stress relaxation is negligible at temperatures below 300 degrees C 

12 [572 degrees F] (Bickford, 2008). This temperature is significantly below those expected in 

13 sheltered and outdoor environments. Thus, stress relaxation of stainless steel subcomponents 

14 exposed to sheltered and outdoor environments is not considered to be credible, and therefore,

15 aging management is not required during the 60-year timeframe.

16 3.2.2.11 Wear

17 Adhesive wear occurs when two metallic components slide against each other under an applied 

18 load where no abrasives are present (Magee, 1992). For the NUHOMS transfer cask exposed 

to indoor and outdoor air, Nitronic® 19 60 stainless steel (UNS S21800) is used to construct the 

20 rails in the cask cavity. The additions of silicon and manganese make this alloy best known for 

21 its wear and galling resistance, even in the annealed condition (Magee, 1992). The rails could 

22 experience repeated sliding contact over multiple canister transfer operations. Thus, wear of 

23 these stainless steel rails is considered to be credible, and therefore, aging management is 

24 required during the 60-year timeframe.

25 3.2.3 Aluminum alloys

26 In DSSs, aluminum and its 6000 series alloys are commonly used in canister internals to 

27 transfer heat because of aluminum’s good thermal conductivity. For example, in the NUHOMS 

28 HSM, anodized Al 1100 is used to construct part of the heat shield assemblies, which are 

29 exposed to a sheltered environment. In the TN-32 and 68 systems, the lid seal is a double 

30 metallic O-ring exposed to a sheltered environment, where the outer jacket of the O-ring is 

31 aluminum. Also, Al 6063-T5 is used in the TN systems to hold the radial neutron shield 

32 material, in which one side of the aluminum is embedded in borated polyester resin and the 

33 other side is in contact with steel. 

34 3.2.3.1 General corrosion

35 General corrosion, also known as uniform corrosion, proceeds at approximately the same rate 

36 over a metal surface (Phull, 2003b). Freely exposed aluminum surfaces in contact with moist 

37 air or water are subject to general corrosion. The corrosion rate depends on solution 

38 composition, pH, and temperature. The corrosion rate of aluminum is normally controlled by the 

39 formation of a passive film of Al2O3 at the metal and water interface. The Pourbaix diagram for 

40 aluminum shows that aluminum is passive in the pH range of approximately 4 to 8.5 at 

41 25 degrees C [77 degrees F] (Kaufman, 1999). However, the aluminum passive film is reported 

42 to be more porous than the chromium oxide film that passivates stainless steel materials 

43 (Bass, 1956). 
3-27

1 Aluminum subcomponents exposed to helium

2 Above a temperature of about 230 degrees C [446 degrees F], an aluminum protective film no 

3 longer develops in the presence of water or steam (Ghali 2010; 2011). As such, general 

4 corrosion of aluminum is possible if exposed to moisture, because initial temperatures near the 

5 spent fuel are above 200 degrees C [392 degrees F]. However, there is very little residual water 

6 in the cask internal environment following drying. Assuming a residual water content of 1 liter 

7 (L) [0.26 gallon (gal)], Jung et al. (2013) calculated that oxidation of all aluminum in the basket 

8 assembly is limited to just 0.54 g [0.019 oz], which is equivalent to a 20- or 2-μm 

[0.79- or 0.079-mils]-thick layer of aluminum over a surface area of 100 or 1,000 cm2 9

[15.5 or 155 in2 10 ]. This suggests that material thinning from oxidation is a very small fraction of 

11 the millimeter-thick [tens of mils-thick] aluminum materials used inside the system. As a result, 

12 sufficient general corrosion to challenge SSC functions is not credible, and therefore, aging 

13 management is not required during the 60-year timeframe in helium environments.

14 Aluminum subcomponents exposed to sheltered and embedded (all) environments

15 Section 3.2.1.1 discussed how an aqueous electrolyte can be developed under a sheltered 

16 environment through deliquescence of deposited salts. The deliquescent brine can be 

17 concentrated and acidic, initiating general corrosion. Therefore, general corrosion of aluminum 

18 lid seals exposed to a sheltered environment is considered to be credible, and aging 

19 management is required during the 60-year timeframe.

20 Anodized aluminum, in which a surface oxide film is deliberately formed in an electrochemical 

21 process, can increase the resistance to corrosion (Vargel, 2004). The successful formation of a

22 protective oxide during manufacture depends on the anodizing solution, applied voltages, and 

23 sealing operations. Because of its anodized film and the relatively low temperatures present, 

24 general corrosion of the NUHOMS aluminum heat shield is not considered to be credible. 

25 However, if defects develop in the anodized film, deep pitting in the underlying metal could 

26 occur, and this is discussed below in Section 3.2.3.2. In the embedded environment, because it 

27 is moisture free, general corrosion is also not considered to be credible. Therefore, aging 

28 management is not required during the 60-year timeframe for anodized aluminum exposed to a 

29 sheltered environment and standard aluminum exposed to embedded environments.

30 3.2.3.2 Pitting and crevice corrosion

31 As discussed in Section 3.2.1.2, pitting corrosion is a localized form of corrosion that is confined 

32 to a point or small area of a metal surface (Frankel, 2003), and crevice corrosion occurs in a 

33 wetted environment when a crevice exists that allows a corrosive environment to develop in a 

34 component (Kelly, 2003). Aluminum and its alloys form a passive film on the surface. Localized 

35 corrosion in the form of pitting or crevice corrosion could occur for these passive aluminum

36 materials, especially in the presence of halides.

37 Aluminum subcomponents exposed to sheltered environments

38 Section 3.2.1.1 discussed how an aqueous electrolyte can be developed on a stainless steel 

39 canister surface in a sheltered environment through deliquescence of deposited salts. The 

40 aluminum heat shield would be expected to be cooler than the canister surface, because it is 

41 farther away from the fuel, and thus the time to reach the critical temperatures for the 

42 development of an aqueous electrolyte in sheltered environments is much lower. 
3-28

1 The protection of aluminum against corrosion, especially the anodized material, depends on the 

2 stability of the passivating oxide films. In chloride-rich environments, the passive layer breaks 

3 down and pitting corrosion becomes the predominant corrosion mode (Foley, 1986; Nguyen and 

4 Foley, 1979). Analyses of surface deposits demonstrate that aluminum exposed to sheltered 

5 environments accumulates adherent particles containing large concentrations of chloride and 

6 sulfate ions (Munier, 1982). Pitting corrosion rates on the order of 25 μm/yr [0.98 mils/yr] have 

7 been reported in seawater (Summerson et al., 1957). In 1 molar NaCl solution, crevice 

8 corrosion rates of aluminum can be as large as 1.3 mm/yr [51 mils/yr] (Baumgattner and 

9 Kaesche, 1988).

10 Because temperatures of aluminum heat-shield surfaces are expected to drop below the 

11 deliquescence threshold for airborne salts during the 60-year timeframe, and the corrosion rate 

12 is not negligible, pitting and crevice corrosion of aluminum in sheltered environments is 

13 considered to be credible, and therefore, aging management is required. 

14 Aluminum subcomponents exposed to helium and embedded environments

15 Pitting and crevice corrosion of aluminum is not considered to be credible in helium and 

16 embedded environments because of (i) the lack of moisture and halides in helium environments 

17 within the cask or canister and (ii) low moisture and oxygen in the embedded environment. 

18 Therefore, aging management of pitting and crevice corrosion is not required for aluminum 

19 exposed to helium and embedded environments during the 60-year timeframe.

20 3.2.3.3 Galvanic corrosion

21 As discussed in Section 3.2.1.3, galvanic corrosion occurs when two dissimilar metals or 

22 conductive materials are in physical contact in the presence of a conducting solution 

23 (Baboian, 2003; Hack, 1993). In DSSs, galvanic coupling exists between aluminum and steel, 

24 stainless steel, and nickel (where aluminum is less noble in each case). For example, the 

25 aluminum lid seal is in contact with stainless steel in the TN-32 and TN-68 systems and an 

26 aluminum plate is in contact with the stainless steel fuel compartment within the TN-32

27 bolted cask.

28 Aluminum subcomponents exposed to sheltered environments

29 Section 3.2.1.1 discussed how an aqueous electrolyte conducive to corrosion can be developed 

30 in sheltered environments through deliquescence of deposited salts. Caseres (2007) reported 

31 corrosion rates of aluminum coupled to carbon steel of about 0.2 mm/yr [8 mils/yr] in solutions 

32 containing chloride ions. The galvanic corrosion rate of aluminum coupling to stainless steel is 

33 expected to be larger, because the corrosion potential difference between stainless steel and 

34 aluminum is larger than carbon steel and aluminum. Because an aqueous electrolyte conducive 

35 to corrosion may be present and corrosion of aluminum is expected to be enhanced under 

36 galvanic coupling, loss of material due to galvanic corrosion of aluminum is considered to be 

37 credible, and therefore, aging management is required during the 60-year timeframe.

38 Aluminum subcomponents exposed to helium

39 There is very little residual water within a cask or canister following drying. Assuming a residual 

40 water content of 1 L [0.26 gal], Jung et al. (2013) calculated that oxidation of all aluminum in 

41 the basket assembly is limited to 0.54 g [0.019 oz], which is equivalent to a 20 or 2-μm 

[0.79- or 0.079-mils]-thick layer of aluminum over a surface area of 100 or 1,000 cm2 42
3-29

[15.5 or 155 in2 1 ]. This suggests that material thinning from oxidation is a very small fraction of 

2 the aluminum materials used inside the system. In conclusion, loss of material due to galvanic 

3 corrosion in helium environments is not considered to be credible, and therefore, aging 

4 management is not required during the 60-year timeframe.

5 3.2.3.4 Microbiologically influenced corrosion

6 As discussed in Section 3.2.1.4, MIC is corrosion caused or promoted by the metabolic activity 

7 of microorganisms (Dexter, 2003). Microorganisms can live in many environments, such as 

8 water, soil, and air, where aerobic bacteria (e.g., iron-manganese oxidizing bacteria, 

9 sulfur/sulfide oxidizing bacteria, methane producers, organic acid-producing bacteria), fungi, 

10 and algae can develop.

11 Aluminum subcomponents exposed to sheltered environments

12 Section 3.2.1.1 discussed how an aqueous electrolyte conducive to corrosion can be developed 

13 in sheltered environments through deliquescence of deposits. This electrolyte also has the 

14 potential to support microbial activity. 

15 A single research study found MIC on an aluminum compact disc exposed to tropical 

16 atmospheres (Garcia-Guinea et al., 2001). However, there is no operating experience of MIC 

17 degradation of aluminum engineering components that operate in environments similar to those 

18 of dry cask storage systems. All of the operating experience of MIC for metallic components is 

19 from conditions in which the metal surface is kept continuously wet. Due to the absence of any 

20 applicable experience of MIC damage of aluminum components under atmospheric conditions, 

21 MIC is not considered to be significant in sheltered environments, and therefore, aging 

22 management is not required during the 60-year timeframe.

23 Aluminum subcomponents exposed to helium and embedded (all) environments

24 Because of the limited amount of water and nutrients in the helium environments within casks 

25 and canisters, and because of the limited water in embedded environments, MIC of aluminum is 

26 not credible for the 60-year timeframe, and therefore, aging management is not required.

27 3.2.3.5 Creep

28 Section 3.2.1.6 explained that, as a general rule of thumb, thermal activation is insufficient to 

29 produce significant creep at temperatures below 0.4Tm, where Tm is the melting point of the 

30 metal in Kelvin (Cadek, 1988). With melting points of 911 to 930 K [638 to 657 degrees C ; 

31 1,180 to 1,215 degrees F], temperatures of at least 364 to 372 K [91 to 99 degrees C

32 ; 196 to 210 degrees F]] are required to initiate significant creep in aluminum. These 

33 temperatures are consistent with Sindelar et al. (2011), which indicates that creep in aluminum 

34 is possible at temperatures greater than 100 degrees C [212 degrees F]. Microstructure also 

35 plays a significant role in a metal’s resistance to creep. Hence, while this 100 degrees C 

36 [212 degrees F] minimum temperature for creep is representative for pure aluminum, creep in 

37 precipitation hardened aluminum alloys does not become significant until about 200 degrees C 

38 [392 degrees F] (Samuels, 1988). Additionally, at temperatures near these threshold values, 

39 high stresses are required to produce creep.
3-30

1 Aluminum subcomponents exposed to helium

2 The highest temperatures within the DSSs are at locations close to the fuel rods, where the 

3 environment is helium. The maximum expected temperature of fuel cladding has been 

4 estimated to be 400 degrees C [752 degrees F] at the beginning of storage (Jung et. al., 2013). 

5 This cladding temperature is expected to decrease to around 266 degrees C [510 degrees F] 

6 after 20 years and to approximately 127 degrees C [261 degrees F] after 60 years. These 

7 estimates depend on many factors, such as the initial heat load of the SNF. Because the fuel 

8 rods are the only heat source within the cask or canister, these temperatures provide upper 

9 temperature limits for all subcomponents. It is apparent from these temperatures that 

10 subcomponents within the cask or canister could be exposed to temperatures above the 

11 minimum creep temperatures for aluminum during at least the first 40 years.

12 Because the minimum creep temperature will be exceeded during a portion of the 60-year 

13 period, it is necessary to consider the load applied to the subcomponent to determine whether 

14 creep deformation will occur and whether the creep affects safety. Subcomponents that do not 

15 serve a structural function are not expected to be under loads other than their own weight, and 

16 in many instances, their weight is also supported by adjacent structures. Due to the minimal 

17 applied loads, creep of nonstructural subcomponents will not produce significant damage during 

18 the 60-year timeframe. Conversely, aluminum subcomponents that serve a structural function 

19 may experience loads that are high enough to produce sufficient creep deformation to affect the 

20 subcomponents’ safety functions.

21 Aluminum subcomponents exposed to sheltered and embedded (all) environments

22 Aluminum subcomponents exposed to sheltered and embedded environments experience lower 

23 temperatures than those experienced by the internal subcomponents. Time-temperature 

24 profiles calculated for the canister surface (EPRI, 2006; Meyer et al., 2013) suggest that 

25 temperatures in excess of 200 degrees C [392 degrees F] could initially be present on portions 

26 of the canister surface and temperatures above 100 degrees C [212 degrees F] could persist for 

27 30 years. Based on these temperatures, creep is credible during the 60-year timeframe but only 

28 on aluminum subcomponents that are attached directly to the canister shell or cask wall and 

29 have a structural function. 

30 The NRC reviewer should review the creep analyses for aluminum structural components that 

31 are exposed to the elevated temperatures discussed above, as contained in the applicant’s 

32 original design-bases documents, to determine whether the renewal application adequately 

33 addresses the implications of extending the operating period to 60 years. This reexamination of 

34 the original analyses would typically be defined as TLAAs in the renewal application. The staff’s 

35 guidance for the review of TLAAs is provided in NUREG–1927, Revision 1. If the original design 

36 basis does not include the pertinent analyses, the reviewer nevertheless should ensure that the 

37 application addresses this potential aging mechanism.

38 If the TLAA or other supplemental analyses demonstrate that creep does not have the potential 

39 to challenge an important-to-safety function, aging management is not required during the 

40 60-year timeframe.

41 Conversely, an applicant may conclude that an analysis cannot support a determination that 

42 creep damage will not challenge an important-to-safety function in the 60-year timeframe of the 

43 period of extended operation. In that case, the applicant may manage the aging of the 

44 associated SSC with an AMP.
3-31

1 3.2.3.6 Fatigue

2 As discussed previously in Section 3.2.1.7, because spent fuel storage is a static application, 

3 cyclic loading by a purely mechanical means is largely limited to transfer cask lifting trunnions. 

4 Some aluminum subcomponents, however, could experience cyclic loads due to thermal 

5 effects, such as those caused by daily and seasonal fluctuations in the temperature of the 

6 external environment.

7 The NRC reviewer should review the fatigue analyses contained in the applicant’s original 

8 design-basis documents to determine whether the renewal application adequately addresses 

9 the implications of extending the operating period to 60 years. This reexamination of the 

10 original fatigue analyses would typically be defined as TLAAs. 

11 As described in greater detail in Chapter 5 of this report, the reviewer should review the design 

12 codes and standards to identify any required fatigue analyses and ensure that the applicant 

13 addresses those analyses with a TLAA. For example, components that were designed in 

14 accordance with the American Society of Mechanical Engineers Boiler and Pressure 

15 Vessel Code (ASME Code) Section III, Division 1, Subsections NB or NC (ASME, 2007a) 

16 were evaluated for the effects of cyclic loading per subparagraphs NB 3222.4 and 

17 NC 3219.2, respectively.

18 The staff’s guidance for the review of TLAAs is provided in NUREG–1927, Revision 1, and 

19 Chapter 5 of this report. In its evaluation of a TLAA, an applicant may conclude that an analysis 

20 can no longer support a determination that aging will not adversely affect an important-to-safety 

21 function in the 60-year timeframe of the period of extended operation. In that case, the 

22 applicant may manage the aging of the associated SCC with an AMP.

23 The AMR tables in Chapter 4 recommend that applicants address any applicable TLAAs 

24 associated with components with a structural function. If no fatigue analysis was performed in 

25 support of the component design, no action is required of the applicant.

26 3.2.3.7 Thermal aging

27 The microstructures of many aluminum alloys will change, given sufficient time at temperature. 

28 This process is commonly called thermal aging. The effect of the thermal aging on mechanical 

29 properties will depend on the time at temperature and the microstructure and chemical 

30 composition of the aluminum components. In some DSSs, Al 1100 and its 6000 series alloys 

31 are used inside and outside the system to transfer heat because of their good thermal 

32 conductivity.

33 Aluminum subcomponents exposed to helium, sheltered, and embedded (all) environments

34 The 6000 series aluminum alloys, such as 6061 and 6063 used in the system internals, are 

35 precipitation-hardened alloys. The precipitation treatment is performed between 163 and 

36 204 degrees C [325 and 399 degrees F] (ASM International, 1991). Prolonged elevated 

37 temperature exposure is known to significantly reduce the strength of these alloys due to 

38 microstructural changes. For example, Farrell (1995) shows that, when alloy 6061-T6 is held at 

39 200 degrees C (392 degrees F), its yield strength drops from approximately 18 ksi at 10,000 

40 hours (1.14 years) to approximately 11.5 ksi at 100,000 hours (11.4 years). Because of this 

41 sensitivity to exposure time, ASME B&PV Code Section II requires that time-dependent 

42 properties be used for exposures above 177 degrees C (350 degrees F) for this alloy.
3-32

1 The maximum expected temperature of fuel cladding has been estimated to be 400 degrees C 

2 [752 degrees F] at the beginning of storage (Jung et. al., 2013). This cladding temperature is 

3 expected to decrease to around 266 degrees C [510 degrees F] after 20 years and to 

4 approximately 127 degrees C [261 degrees F] after 60 years. It is apparent from these 

5 temperatures that the 6061 and 6063 aluminum alloys may experience significant overaging at 

6 a higher temperature than that for precipitation treatment, leading to loss of strength. This loss 

7 of strength could be an issue for any subcomponents that perform a structural function. 

8 Because AI 1100 aluminum is not a precipitation-hardened alloy, it will not experience any 

9 overaging. However, if it is used in the cold worked state, it will anneal at temperatures above 

10 300 degrees C [572 degrees F] (ASM International, 1991). This annealing will reduce strength, 

11 which could be significant for subcomponents that serve a structural function.

12 Aluminum subcomponents exposed to sheltered and embedded environments experience lower 

13 temperatures than the internal subcomponents. Time-temperature profiles calculated for the 

14 canister surface (EPRI, 2006; Meyer et al., 2013) suggest that temperatures in excess of 

15 200 degrees C [392 degrees F] could initially be present on portions of the canister surface and 

16 temperatures above 100 degrees C [212 degrees F] could persist for 30 years. Based on these 

17 temperatures, thermal aging could occur on aluminum subcomponents that have a structural 

18 function and are attached directly to the canister shell or cask wall.

19 Because thermal aging of aluminum is a possible aging mechanism, the NRC reviewer should 

20 review any aging analyses for aluminum structural components that are exposed to the elevated 

21 temperatures discussed above, as contained in the applicant’s original design-bases 

22 documents, to determine whether the renewal application adequately addresses the 

23 implications of extending the operating period to 60 years. This reexamination of the original 

24 analyses would typically be defined as TLAAs in the renewal application. The staff’s guidance 

25 for the review of TLAAs is provided in NUREG–1927, Revision 1. If the original design basis 

26 does not include the pertinent analyses, the reviewer nevertheless should ensure that the 

27 application addresses the potential for thermal aging to adversely affect the structural function of 

28 aluminum components.

29 3.2.3.8 Radiation embrittlement

30 Embrittlement of metals may occur under exposure to neutron radiation. Depending on the 

31 neutron fluence, radiation can cause changes in mechanical properties, such as loss of ductility, 

32 fracture toughness, and resistance to cracking.

33 Farrell and King (1973) showed that pure aluminum had increased strength but decreased 

ductility after being irradiated to fast fluences in the range of 1 to 3 × 1022 n/cm2 34

[6.5 to 19.4 × 1022 n/in2 35 ] from a research reactor for 8 years. Alexander (1999) showed that 

irradiation at 1022 n/cm2 [6.5 × 1022 n/in2 36 ] simulating reactor conditions affected the mechanical 

37 properties of aluminum alloy 6061-T651. 

38 Some results from radiation testing of aluminum-based neutron poisons are reported in the 

39 literature (EPRI, 2009a). Gamma, thermal neutron, and fast neutron radiation testing of an 

aluminum-based laminate composite in water for 9 years and exposed to up to 7 × 1011 40 rad 

gamma, 3.6 × 1018 n/cm2 [2.2 × 1019 n/in2

] fast neutron fluence, and 2.7 × 1019 n/cm2 41

[1.7 × 1020 n/in2 42 ] thermal neutron fluence showed no change in ultimate strength and no other 

43 signs of physical deterioration except for severe oxidation because of the presence of water. 

44 Also, radiation testing of an aluminum-based, sintered composite subjected to up to 

1.5 × 1020 n/cm2 [9.7 × 1020 n/in2

] fast neutron fluence and a maximum of 3.8 × 1011 45 rad gamma 
3-33

1 exposure showed little change in the yield strength and ultimate strength (EPRI, 2009a). 

Finally, neutron radiation of borated aluminum to fluences of 1017 n/cm2 [6.5 × 1017 n/in2 2 ] 

3 showed no dimensional change or radiation damage (EPRI, 2009a). These test conditions are 

4 expected to be more severe than those experienced by aluminum alloys in the extended 

5 storage application (EPRI, 2009a). 

6 As discussed in Section 3.2.1.9 of this report, the maximum potential accumulated neutron 

fluence on DSS components after 100 years was calculated to be 2.63 × 1016 n/cm2 7

[1.70 × 1017 n/in2 8 ]. This fluence is well below the levels that have been found degrade the 

9 mechanical properties of aluminum alloys. Thus, radiation embrittlement of aluminum 

10 subcomponents exposed to any environment is expected to be insignificant, and therefore, 

11 aging management is not required during the 60-year timeframe.

12 3.2.4 Nickel alloys

13 Nickel alloys are used in only a few DSS applications. In the HI-STAR overpack, nickel 

14 alloy 718 (ASME SB637) is used to construct closure plate bolts and trunnion bolts, and nickel 

15 alloy X750 is used to construct seals. These components are exposed to an outdoor 

16 environment. Nickel alloy 718 (ASME SB637) is also used to construct the trunnion for the 

17 HI-TRAC transfer cask, which is predominantly exposed to an indoor environment or otherwise 

18 encased without direct air ingress except for short periods of air exposure during transfer 

19 operations. For such air-indoor/outdoor environments, the aging effects from aqueous corrosion 

20 processes are expected to be bounded by those from the outdoor environment. Both nickel 

21 alloys 718 and X750 are precipitation-hardened alloys that contain chromium to form a passive 

22 oxide film on the surface (Crook, 2005). 

23 3.2.4.1 General corrosion

24 The high chromium contents of alloys 718 and X750 (greater than 17 and 14 weight percent,

25 respectively), make these alloys very resistant to general corrosion, even in such reducing acids 

26 as hydrochloric acid (Crook, 2005). Because of its passive behavior and high corrosion 

27 resistance, general corrosion of nickel alloys exposed to outdoor environments is not 

28 considered to be credible, and therefore, aging management is not required during the 60-year 

29 timeframe. 

30 3.2.4.2 Pitting and crevice corrosion

31 As discussed in Section 3.2.1.2, pitting corrosion is a localized form of corrosion that is confined 

32 to a point or small area of a metal surface (Frankel, 2003) and crevice corrosion occurs in a 

33 wetted environment when a crevice exists that allows a corrosive environment to develop in a 

34 component (Kelly, 2003). 

35 Section 3.2.1.1 discussed how an aqueous electrolyte can be developed in outdoor air. This 

36 electrolyte could contain chemical species such as halides and sulfides. Localized corrosion in 

37 the form of pitting and/or crevice corrosion may occur for some passive nickel alloys, but overall, 

38 nickel alloys are more resistant to localized corrosion than stainless steels (Crook, 2005). 

39 Nickel alloy 718 is used in sea water applications, where the chloride concentration is much 

40 higher than that from outdoor air. Furthermore, for many nickel alloys in different environmental

41 systems, localized corrosion growth is often observed to slow down or stop, which is referred to 

42 as the stifling and arrest phenomena (He and Dunn, 2007). Because of the high corrosion 

43 resistance, pitting or crevice corrosion of nickel subcomponents exposed to outdoor air is not 
3-34

1 considered to be credible, and therefore, aging management is not required during the 60-year 

2 timeframe.

3 3.2.4.3 Microbiologically influenced corrosion

4 As discussed in Section 3.2.1.4, MIC is corrosion caused or promoted by the metabolic activity 

5 of microorganisms (Dexter, 2003). Microorganisms can live in many environments, such as 

6 water, soil, and air, where aerobic bacteria (e.g., iron-manganese oxidizing bacteria, 

7 sulfur/sulfide oxidizing bacteria, methane producers, organic acid-producing bacteria), fungi, 

8 and algae can develop. 

9 Although the moisture necessary to support microbial activity may be present on surfaces 

10 exposed to outdoor environments, all of the operating experience of MIC of metallic components 

11 is from conditions where the surface is continuously wet. Furthermore, there is no operational 

12 or experimental evidence of MIC degradation of nickel-chromium alloys similar to 718 and X750 

13 (Little and Lee, 2009). Due to the absence of any operating experience of MIC damage to

14 nickel alloys under atmospheric conditions, MIC of nickel subcomponents exposed to outdoor 

15 air is not considered to be credible, and therefore, aging management is not required during the 

16 60-year timeframe.

17 3.2.4.4 Stress corrosion cracking

18 As discussed in Section 3.2.1.5, SCC is the cracking of a metal produced by the combined 

19 action of corrosion and tensile stress (applied or residual) (Jones, 1992, 2003). SCC of nickel 

20 alloys has been experienced in high-temperature water and hot caustic solutions (Phull, 2003). 

21 These conditions do not exist in the outdoor air environment of DSSs. Although 

22 chloride-containing electrolytes could develop in outdoor air, as discussed in Section 3.2.1.1, 

23 nickel-based alloys are known to be highly resistant to the chloride-induced SCC that affects 

24 stainless steels. In indoor air, the probability of developing a corrosive aqueous electrolyte is 

25 negligible. Because alloys 718 and X750 are not susceptible to the dry storage outdoor air 

26 environments, SCC is not expected to be credible. Therefore, aging management is not 

27 required during the 

28 60-year timeframe.

29 3.2.4.5 Fatigue

30 As discussed previously in Section 3.2.1.7, because spent fuel storage is a static application, 

31 cyclic loading by a purely mechanical means is largely limited to transfer cask lifting trunnions, 

32 which are loaded each time a canister is moved from the spent fuel pool to the dry storage pad. 

33 Other subcomponents, however, could experience cyclic loads due to thermal effects, such as 

34 those caused by daily and seasonal fluctuations in the temperature of the external 

35 environment.The NRC reviewer should review the fatigue analyses contained in the applicant’s 

36 original design-basis documents to determine whether the renewal application adequately 

37 addresses the implications of extending the operating period to 60 years. This reexamination of 

38 the original fatigue analyses would typically be defined as TLAAs. 

39 In some cases, fatigue analyses may have been performed to support the original design but 

40 are not explicitly discussed in the design basis documentation. As a result, the reviewer should 

41 review the design codes and standards to identify any required fatigue analyses and ensure that 

42 the applicant addresses those analyses with a TLAA. For example, components that were 

43 designed in accordance with the American Society of Mechanical Engineers Boiler and 
3-35

1 Pressure Vessel Code (ASME Code) Section III, Division 1, Subsections NB or NC 

2 (ASME, 2007a) were evaluated for the effects of cyclic loading per subparagraphs NB 3222.4 

3 and NC 3219.2, respectively.

4 The staff’s guidance for the review of TLAAs is provided in NUREG–1927, Revision 1, and 

5 summarized in Chapter 5 of this report. In its evaluation of a TLAA, an applicant may conclude 

6 that an analysis can no longer support a determination that aging will not adversely affect an 

7 important-to-safety function in the 60-year timeframe of the period of extended operation. In 

8 that case, the applicant may manage the aging of the associated SCC with an AMP.

9 The AMR tables in Chapter 4 recommend that applicants address any applicable TLAAs 

10 associated with components with a structural function. If no fatigue analysis was performed in 

11 support of the component design, no action is required of the applicant. 

12 3.2.4.6 Radiation embrittlement

13 Depending on the neutron fluence, radiation can cause changes in mechanical properties such 

14 as loss of ductility, fracture toughness, and resistance to cracking. Nickel-based alloys 

15 experienced significant reductions in tensile ductility during neutron irradiation at elevated 

16 temperatures of 400–600 degrees C [752–1,112 degrees F] for neutron doses approaching

17 10–15 displacements per atom (dpa), which corresponds to a neutron fluence of about 

1021–1022 n/cm2 [6.5 × 1021–6.5 × 1022 n/in2 18 ] (Was et al., 2006; Rowcliffe, 2009). Nickel 

19 alloy X-750 cracking has been observed extensively in nuclear power plant applications after 

attaining an end-of-life fluence of 1 to 10 × 1021 n/cm2 [6.5 to 65 × 1021 n/in2 20 ] (Was et al., 2006). 

21 As discussed in Section 3.2.1.9 of this report, the maximum potential accumulated neutron 

fluence on DSS basket components after 100 years was calculated to be 2.63 × 1016 n/cm2 22

[1.70 × 1017 n/in2 23 ]. This fluence is five to six orders of magnitude below the level at which the 

24 mechanical properties of nickel have been observed to be degraded. In addition, for the nickel 

25 overpack and transfer cask subcomponents, the neutron exposure is significantly lower than the 

26 calculated exposure for the basket components in Section 3.2.1.9. Thus, radiation 

27 embrittlement of nickel alloys is expected to be insignificant, and therefore, aging management 

28 is not required during the 60-year timeframe.

29 3.2.4.7 Stress relaxation

30 Section 3.2.1.10 explained that stress relaxation of bolting is the steady loss of stress due to 

31 atomic movement at elevated temperature in a loaded part where dimensions are fixed 

32 (Earthman, 2000). The service temperature limit for nickel alloy 718 is 649 degrees C 

33 [1,200 degrees F] (Bickford, 2008), which is much higher than the external temperature of the 

34 HI-STAR overpack in which nickel bolts are used. Below the service temperature limit, the bolts 

35 are expected to maintain their original clamping force. Thus, stress relaxation of nickel alloy 

36 subcomponents exposed to the outdoor environment is not considered to be credible, and 

37 therefore, aging management is not required during the 60-year timeframe.

38 3.2.4.8 Wear

39 Fretting wear is the repeated cyclical rubbing between two surfaces. For the HI-TRAC transfer 

40 cask exposed to air-indoor/outdoor environments, the nickel alloy used to construct the lifting 

41 trunnion may experience cyclic rubbing during loading and unloading. Thus, wear of the nickel 
3-36

1 alloy is considered to be credible, and therefore, aging management is required during the 

2 60-year timeframe.

3 3.2.5 Copper alloys

4 Copper alloys are used in only a few DSS applications. In the HI-STAR overpack, brass, which 

5 is a copper-zinc alloy containing more than 50 percent copper, is used as the rupture disk 

6 material. In the NUHOMS HSM, copper is used to construct the lightning protection system. 

7 Both subcomponents are exposed to outdoor air.

8 3.2.5.1 General corrosion

9 General corrosion, also known as uniform corrosion, proceeds at approximately the same rate 

10 over a metal surface (Phull, 2003b). Freely exposed copper surfaces in contact with moist air or 

11 water are subject to general corrosion. The corrosion rate depends on solution composition, 

12 pH, and temperature. The copper Pourbaix diagram (Pourbaix, 1974) indicates that copper and 

13 copper alloys are reactive with water in the presence of oxygen, but the low corrosion rate has 

14 allowed their wide use in industrial, marine, and rural atmospheres (Cohen, 2005). General 

15 corrosion of copper and its alloys is the predominant corrosion mode, because they do not form 

16 a truly passive oxide film on the surface. 

17 Atmospheric corrosion of copper has been observed and studied extensively (Leidheiser, 1974; 

18 Rozenfeld, 1972). The corrosion rate of copper is strongly dependent on relative humidity and 

19 the concentration of pollutants in the air (e.g., chlorides, sulfur dioxide, hydrogen sulfide). The 

20 presence of NaCl in a marine environment has a strong corrosive effect toward copper under 

21 thin electrolyte layers and in alternating wet and dry cyclic conditions. Copper corrosion rates 

22 usually decrease with time, following an exponential decay law (Feliu et al., 1993). Typical 

23 corrosion rates of copper exposed to marine and industrial environments are 0.6–2.5 μm/yr 

24 [0.024–0.098 mils/yr] and 1.3 μm/yr [0.051 mils/yr], respectively (Tracy, 1955; Herman and 

25 Castillo, 1974). Fonseca et al. (2004) recorded copper corrosion in marine environments as 

26 high as 7.8 μm/yr [0.31 mils/yr]. In atmospheric marine environments, copper corrosion is on 

27 the order of 16 μm/yr [0.62 mils/yr] (Farro et al., 2009). Assuming a corrosion rate of 10 μm/yr 

28 [0.39 mils/yr], the metal loss could be 0.6 mm [23.6 mils] over 60 years. As such, general 

29 corrosion of copper alloys exposed to an outdoor air environment is considered to be credible, 

30 and therefore, aging management is required during the 60-year timeframe.

31 3.2.5.2 Pitting and crevice corrosion

32 As discussed in Section 3.2.1.2, pitting corrosion is a localized form of corrosion that is confined 

33 to a point or small area of a metal surface (Frankel, 2003), and crevice corrosion occurs in a 

34 wetted environment when a crevice exists that allows a corrosive environment to develop in a 

35 component (Kelly, 2003). 

36 The common form of atmospheric corrosion for copper exposed to outdoor air is general 

37 corrosion, because copper alloys do not have a true protective film (Cohen, 2005). In an 

38 oxidizing environment, copper could experience surface roughening, initially appearing like 

39 localized corrosion; however, localized corrosion tends to converge with general corrosion 

40 (i.e., the penetration front of localized corrosion merges with that of general corrosion). 

41 Long-term tests of copper alloys show that the average pit depth does not continually increase 

42 with extended times of exposure (Cohen, 2005). Copper has been commonly used for 

43 architectural components exposed to outdoor air for many years, such as when used for roofing, 
3-37

1 building fronts, and statues, where localized corrosion is not shown to be evident. Because 

2 localized corrosion is not a primary corrosion mechanism for copper alloys exposed to outdoor 

3 air, and it tends to converge with general corrosion, it is not considered to be credible, and 

4 therefore, aging management is not required during the 60-year timeframe.

5 3.2.5.3 Microbiologically influenced corrosion

6 As discussed in Section 3.2.1.4, MIC is corrosion caused or promoted by the metabolic activity 

7 of microorganisms (Dexter, 2003). Although the moisture necessary to support microbial 

8 activity may be present on surfaces exposed to the outdoor environment, all of the operating 

9 experience of MIC of metallic materials is from conditions under which the surface is 

10 continuously wet, and it is unclear whether these rates could be sustained if the conditions to 

11 support MIC are only present on an intermittent basis. Furthermore, there is no experimental 

12 evidence of MIC degradation of copper alloys under atmospheric conditions. Due to the 

13 absence of any operating experience of MIC damage of copper alloys under atmospheric 

14 conditions, MIC is not considered to be significant, and therefore, aging management is not 

15 required during the 60-year timeframe.

16 3.2.5.4 Radiation embrittlement

17 Depending on the neutron fluence, radiation can cause changes in mechanical properties, such 

18 as loss of ductility, fracture toughness, and resistance to cracking. Radiation hardening and 

19 embrittlement of pure copper and copper-based alloys have been observed at temperatures in 

the range of 60–90 degrees C [140–194 degrees F] in the dose range of 10−3

–10−1 20 dpa 

21 (Fabritsiev et al., 2004). Blewitt et al. (1957) observed yield drop on stress–strain curves, 

22 hardening, and a decrease in uniform and total elongation upon irradiation of pure copper at 

60 degrees C [140 degrees F] to doses of 1019 n/cm2 [6.5 ×1019 n/in2 23 ]. 

24 As discussed in Section 3.2.1.9 of this report, the maximum potential accumulated neutron 

fluence on DSS basket components after 100 years was calculated to be 2.63 × 1016 n/cm2 25

[1.70 × 1017 n/in2 26 ]. This fluence is at least three orders of magnitude below the level at which 

27 the mechanical properties of copper alloys have been reported to be degraded. In addition, for 

28 locations outside of the overpack where copper alloys are used, the accumulated dose is much 

29 lower than the level calcluated in Section 3.2.1.9. Thus, radiation embrittlement of copper alloys 

30 exposed to outdoor air is expected to be insignificant, and therefore, aging management is not 

31 required during the 60-year timeframe. 

32 3.2.6 Lead

33 Lead is used as gamma radiation shielding in the NUHOMS and Holtec transfer casks, as well 

34 as some NUHOMS dry shielded canister designs. In each case, the lead is encased in steel or 

35 stainless steel and thus is not exposed to water or atmospheric contaminants. Lead is well

36 known to be very resistant to corrosion in a variety of environments (Alhasan, 2005). Because 

37 there are no credible aging mechanisms that could challenge the ability of lead to perform its 

38 shielding (and, in some cases, heat transfer) functions, aging management of this material is not 

39 required during the 60-year timeframe.

40 3.2.7 Depleted uranium

41 Depleted uranium is used as a shield plug in the FuelSolutions canister. The material is 

42 encased in steel or stainless steel and thus is not exposed to water or atmospheric 
3-38

1 contaminants. Uranium is known to be resistant to corrosion in a variety of environments 

2 (Lillard and Hanrahan, 2005). Because there are no credible aging mechanisms that could 

3 challenge the ability of depleted uranium to perform its shielding functions, aging management 

4 of this material is not required during the 60-year timeframe. 

5 3.2.8 Coatings

6 Coatings in DSSs are used primarily for corrosion mitigation, to facilitate decontamination, and 

7 to improve heat-rejection capability by increasing the emissivity of cask internal components. A 

8 wide array of coating materials is used to fulfill these functions, such as organic epoxy, inorganic 

9 zinc-rich coatings, galvanized zinc, aluminum, nickel, and cadmium. However, coatings are 

10 often present for operational purposes and may not be credited as supporting an important-to11 safety function. Thus, the reviewer should examine the DSS design-basis documentation to 

12 verify that the renewal applicant appropriately identified the coatings that meet the renewal 

13 scoping criteria in NUREG–1927, Revision 1.

14 Coatings are exposed to outdoor air, indoor/outdoor air (transfer cask), and sheltered 

15 environments, which are characterized by elevated temperature and radiation exposure. As 

16 discussed in greater detail for neutron shielding materials in Section 3.3.1, polymeric materials 

17 may be susceptible to heat- and radiation-induced molecular scission (breaking) and cross18 linking that can cause embrittlement and cracking.

19 The variety of coatings and the proprietary nature of many coating systems make a generic 

20 evaluation of specific degradation mechanisms impractical. Nevertheless, the NRC recognizes 

21 that coatings may degrade, either through aging or inappropriate application methods, and 

22 recommends in-service condition assessments of coatings to ensure that they continue to 

23 support their important-to-safety functions. NRC Regulatory Guide 1.54, Revision 2, “Service 

24 Level I, II, and III Protective Coatings Applied to Nuclear Power Plants,” references American 

25 Society for Testing and Materials (ASTM) standards that are considered appropriate guidance 

26 for coating maintenance in nuclear power plants (NRC, 2010d). The ASTM standards typically 

27 recommend periodic visual inspections for blistering, cracking, flaking/peeling, and rusting, 

28 which may be followed by physical tests when degradation is identified. Regulatory Guide 1.54 

29 also notes that the Electric Power Research Institute (EPRI) Report 1019157, “Guideline on 

30 Nuclear Safety-Related Coatings,” Revision 2, provides additional information on the 

31 maintenance of coatings (EPRI, 2009b). 

32 The reviewer should verify that the renewal applicant has an existing coating maintenance 

33 program or proposes a new aging management program for coatings that are credited with 

34 performing an important-to-safety function or protecting an important-to-safety component. The 

35 AMR tables identify thermal and radiation effects as most likely to degrade coatings, and a 

36 site-specific AMP consistent with ASTM guidelines is recommended to manage aging.

37 3.2.9 References

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42 pp. 1,027–1,044. 1999.
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22 Chopra, O., D. Diercks, R. Fabian, Z. Han, and Y. Liu. “Managing Aging Effects on Dry Cask 

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29 Jones, R.H. Stress corrosion Cracking. Materials Park, Ohio: ASM International: 1992.

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35 Kaufman, J.G. Properties of Aluminum Alloys: Tensile, Creep, and Fatigue Data at High and 

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12 NACE. Corrosion Engineer's Reference Book. Third Edition. Edited by R. Baboian. 

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20 U.S. Nuclear Regulatory Commission. ADAMS Accession No. ML16179A148. 2016.

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31 _____. NUREG−1801, “Generic Aging Lessons Learned (GALL) Report.” Rev. 2. Washington, 

32 DC: U.S. Nuclear Regulatory Commission. ADAMS Accession No. ML103490041. 2010a.

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36 _____. “Outside Diameter Initiated Stress Corrosion Cracking Revised Final White Paper.” 

37 PA–MSC–0474.” Letter (October 14) to NRC From M.L. Arey, Jr. (PWROG Owners Group). 
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36 John Wiley and Sons. 2000.
3-47

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16 2011.

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1 West, G.A. and C.D. Watson. “Gamma Radiation Damage and Decontamination Evaluation of 

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3 ORNL-3589. Oak Ridge, Tennessee: Oak Ridge National Laboratory. 1965.
3-49

1 3.3 Neutron shielding materials

2 Neutron shielding typically is provided by either borated or nonborated polymeric or 

3 cementitious materials. Hydrogen and oxygen reduce the energy of the neutrons such that the 

4 neutrons are more effectively absorbed by the boron. The degradation and possible relocation 

5 of shielding materials may be mitigated by encasing or reinforcing materials. For example, 

6 shielding is often cast within a metal liner, which prevents ingress of water and contaminants. 

7 Also, some shielding materials include reinforcements (e.g., fiberglass) for stability. 

8 A set of known aging mechanisms with the potential to affect the performance of shielding 

9 materials was identified from reviews of a range of information; sources of the information

10 include gap assessments for DSSs, relevant technical literature, and operating experience from 

11 nuclear applications (NRC, 2014a, 2010; Chopra et al., 2014; Hanson et al., 2012; 

12 Sindelar et al., 2011; NWTRB, 2010; EPRI, 2011). These mechanisms, which are induced by 

13 thermal and irradiation conditions, include boron depletion, thermal aging, and radiation 

14 embrittlement. Detailed discussions regarding each of these aging mechanisms follow. 

15 3.3.1 Neutron-shielding materials

16 Polymer based

17 The TN-32 and TN-68 systems use both a borated polyester resin and polypropylene for 

shielding, while Holtec’s HI-STAR overpack and HI-TRAC transfer cask use Holtite-A.TM 18 

Holtite-ATM 19 is a composite material consisting of an epoxy polymer, boron carbide powder, and 

20 aluminum hydroxide. 

21 Cement based

22 The cementitious BISCO NS-3 material is used in one of the NUHOMS transfer cask designs for 

23 neutron shielding. The structural concrete used to construct overpacks also serves as neutron 

24 and gamma shielding; the degradation of such concrete is discussed separately in Section 3.5.

25 3.3.1.1 Boron depletion (borated materials)

26 The boron concentration in the neutron shields decreases as boron atoms in the borated 

27 materials absorb neutrons. Boron-10 nuclei capture neutrons, yielding excited Boron-11 nuclei, 

28 which in turn decay into high-energy alpha particles and Lithium-7 nuclei. The neutron shielding 

29 material will lose one boron-10 atom per such a reaction. Significant depletion of boron-10

30 atoms may occur over time, if the shielding material is exposed to sufficient neutron fluence.

31 The NRC reviewer should ensure that the applicant provides a bounding analysis to show that 

32 boron-10 depletion is not a credible aging mechanism for its specific DSS design. The reviewer 

33 should review any boron depletion analyses contained in the applicant’s original design-bases 

34 documents, if present, to determine whether the design-basis analysis or license renewal 

35 application adequately addresses the implications of extending the operating period to 60 years. 

36 This reexamination of the original analyses would typically be defined as TLAAs in the renewal 

37 application. The staff’s guidance for the review of TLAAs is provided in NUREG–1927, 

38 Revision 1 (NRC, 2016). If the original design basis does not include an analysis for loss of 

39 boron-10, the reviewer nevertheless should ensure that the renewal application adequately 

40 addresses this aging mechanism. 
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1 Rather than demonstrating performance through an analysis, an applicant may choose to 

2 manage loss of neutron shielding, such as through radiation monitoring, to confirm the 

3 shielding’s continued effectiveness. In that case, the reviewer should refer to NRC guidance on 

4 the review of AMPs in NUREG–1927, Revision 1. 

5 3.3.1.2 Thermal aging

6 Polymers may be susceptible to heat-induced changes to material properties and configuration 

7 due to a number of mechanisms. At elevated temperatures, the long chain backbone of a 

8 polymer can undergo molecular scission (breaking) and cross linking. Also, gaseous products 

9 may be formed, including H2, CH4, and CO2. These reactions may cause embrittlement, 

10 shrinkage, decomposition, and changes in physical configuration (e.g., loss of hydrogen or 

11 water) (EPRI, 2002; McManus and Chamis, 1996). Shrinkage and embrittlement can locally 

12 displace shielding material and potentially diminish shielding effectiveness, although this may be 

13 mitigated in part by reinforcement materials within the polymer matrix and the support provided

14 by the encasing metal. Because many polymers are known to degrade at elevated 

15 temperatures, thermal aging for polymer-based neutron-shielding materials is a credible aging 

16 mechanism. Therefore, either a supporting analysis for the material’s continued use or an AMP

17 is required during the 60-year timeframe.

18 The cementitious BISCO NS-3 shielding material used in one of the NUHOMS transfer cask

19 designs may experience some loss of hydrogen (neutron moderator) when exposed to elevated 

20 temperatures. However, the material is subjected to elevated temperatures only during 

21 relatively brief periods when the storage canister is being transported from the spent fuel pool to 

22 the storage pad. Thus, the time of thermal exposure in the transfer cask is minimal compared to 

23 the continuous thermal exposure NS-3 experiences in other NRC-approved applications 

24 (e.g., the MC-10 metal storage cask) (NRC, 2005). As a result, thermal aging of the NS-3 

25 shielding material is not considered to be a credible aging mechanism in the transfer cask, and 

26 therefore, aging management is not required during the 60-year timeframe. 

27 3.3.1.3 Radiation embrittlement

28 Similar to the thermal aging mechanism discussed above, radiation can alter polymer structures 

29 by molecular scission and cross linking to reduce ductility, fracture toughness, and resistance to 

30 cracking (Fu, et al., 1988; Cota, et al., 2007). For example, the threshold for radiation 

embrittlement has been found to be about 106 31 rad for polyethylene and significantly lower for 

32 other polymers, such as polytetrafluoroethylene (EPRI, 1998). Depending on the DSS design 

33 and the specific SNF, this dose can be reached in 10–100 years (EPRI, 1998). Embrittlement 

34 can locally displace shielding material and potentially reduce shielding effectiveness, although 

35 this may be mitigated, in part, by reinforcement materials within the polymer matrix and the 

36 support provided by the encasing metal. As a result, radiation embrittlement of polymer-based 

37 neutron-shielding materials is a credible aging mechanism, and therefore, either a supporting 

38 analysis for the material’s continued use or an AMP is required during the 60-year timeframe. 

39 An acceptable AMP may include monitoring and trending of radiation dose to confirm the 

40 absence of an decreasing trend in shielding effectiveness.

41 An analysis of the effects of radiation on the shielding properties of BISCO NS-3 has shown that 

42 both the gamma and neutron radiation dose the shielding material receives over 60 years in the 

43 NUHOMS transfer cask are several orders of magnitude below the material’s exposure limit 

44 (BISCO, 1986; NRC, 2014b). As a result, radiation embrittlement of the NS-3 shielding material 
3-51

1 is not considered to be a credible aging mechanism, and therefore, aging management is not 

2 required during the 60-year timeframe.

3 3.3.2 References

4 BISCO Products, Inc. “NS-3 Specification Sheet.” (ADAMS ML110730731), June 23, 1986. 

5 Chopra, O., D. Diercks, R. Fabian, Z. Han, and Y. Liu. “Managing Aging Effects on Dry Cask 

6 Storage Systems for Extended Long-Term Storage and Transportation of Used Fuel.” 

7 FCRD–UFD–2014–000476. ANL–13/15, Rev. 2. Washington, DC: U.S. Department of 

8 Energy. 2014.

9 Cota, S.S., V. Vasconcelos, M. Senne, Jr., L.OL. Carvalho, D.B. Rezende, and R.F. Cõrrea. 

10 “Changes in Mechanical Properties Due to Gamma Irradiation of High-Density 

11 Polyethylene.”Brazilian Journal of Chemical Engineering. Volume 24, No. 02. pp. 259–265. 

12 2007.

13 EPRI. “Extended Storage Collaboration Program (ESCP) Progress Report and Review of Gap 

14 Analyses.” Report 1022914. Palo Alto, California: Electric Power Research Institute. 2011.

15 _____. “Technical Bases for Extended Dry Storage of Spent Nuclear Fuel.” Report 1003416. 

16 Palo Alto, California: Electric Power Research Institute. 2002.

17 _____. “Data Needs for Long-Term Dry Storage of LWR Fuel.” Report TR-108757. 

18 Palo Alto, California: Electric Power Research Institute. 1998.

19 Fu, L., R.A. Fouracre, and H.M. Banford. “An Investigation of Radiation Damage in Cured 

20 Epoxy Resin System Using Regression Experiment Design, Electrical Insulation and Dielectric

21 Phenomena.” 1988 Annual Report, Conference on Electrical Insulation and Dielectric

22 Phenomena. IEEE Dielectrics and Electrical Insulation Society. 1988.

23 Hanson, B., H. Alsaed, C. Stockman, D. Enos, R. Meyer, and K. Sorenson. “Used Fuel 

24 Disposition Campaign: Gap Analysis to Support Extended Storage of Used Nuclear Fuel, 

25 Rev. 0.” Richland, Washington: Pacific Northwest National Laboratory. 2012.

26 McManus, H.L. and C.C. Chamis. “Stress and Damage in Polymer Matrix Composite Materials

27 Due to Material Degradation at High Temperatures.” NASA Technical Memorandum 4682.

28 Cambridge, Massachusetts: Massachusetts Institute of Technology. 1996.

29 NRC. NUREG–1927, “Standard Review Plan for Renewal of Specific Licenses and Certificates 

30 of Compliance for Dry Storage of Spent Nuclear Fuel.” Revision 1. ADAMS Accession 

31 No. ML16179A148. Washington, DC: U.S. Nuclear Regulatory Commission. 2016. 

32 _____. “Identification and Prioritization of the Technical Information Needs Affecting Potential 

33 Regulation of Extended Storage and Transportation of Spent Nuclear Fuel.” Washington, DC: 

34 U.S. Nuclear Regulatory Commission. 2014a.

35 _____. “Safety Evaluation Report for License Renewal: Calvert Cliffs Nuclear Power Plant 

36 Independent Spent Fuel Storage Installation.” Washington, DC: U.S. Nuclear Regulatory 

37 Commission. 2014b.
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1 _____. NUREG−1801, “Generic Aging Lessons Learned (GALL) Report.” Rev. 2. 

2 Washington, DC: U.S. Nuclear Regulatory Commission. 2010.

3 _____. “Safety Evaluation Report for License Renewal: Surry Independent Spent Fuel Storage 

4 Installation.” Washington, DC: U.S. Nuclear Regulatory Commission. ADAMS Accession No. 

5 ML050590266. 2005.

6 NWTRB. “Evaluation of the Technical Basis for Extended Dry Storage and Transportation of 

7 Used Nuclear Fuel.” Washington, DC: Nuclear Waste Technical Review Board. 2010.

8 Sindelar, R.L., A.J. Duncan, M.E. Dupont, P.-S. Lam, M.R. Louthan, Jr., and T.E. Skidmore. 

9 NUREG/CR-7116, “Materials Aging Issues and Aging Management for Extended Storage and 

10 Transportation of Spent Nuclear Fuel.” Washington, DC: U.S. Nuclear Regulatory Commission. 

11 2011. 
3-53

1 3.4 Neutron poison materials

2 Subcriticality of the SNF in DSSs may be maintained, in part, by the placement of neutron 

3 absorbers, or poison plates, around the fuel assemblies. Commonly used neutron poisons 

4 include borated stainless steel, borated aluminum alloys, aluminum metal-matrix composites 

such as MetamicTM and Boralyn® 5 , and aluminum-boron carbide laminate composites, commonly 

referred to as cermets, such as Boral® 6 . These materials are exposed to helium environments, 

7 where temperature and radiation levels are expected to be high because of their proximity to the 

8 fuel assemblies. This environment also could include small amounts of water left after the 

9 drying operations.

10 A list of known aging mechanisms that have the potential to affect the performance of neutron 

11 poison plates was identified from reviews of a range of information sources, including gap 

12 assessments for DSSs, relevant technical literature, and operating experience from nuclear and

13 nonnuclear applications (NRC, 2014, 2010; Chopra et al., 2014; Hanson et al., 2012; 

14 Sindelar et al., 2011; NWTRB, 2010). These mechanisms, which are induced by various 

15 physicochemical, thermal-mechanical, and irradiation conditions, include general corrosion, 

16 galvanic corrosion, wet corrosion and blistering, creep, thermal aging, radiation embrittlement, 

17 and boron depletion.

18 3.4.1 Borated stainless steel

19 The Type 304 borated stainless steels used as neutron poison plates are similar in composition 

20 to standard Type 304 stainless steels used in other engineering applications, except that the 

21 borated steels contain boron, which has a much higher thermal neutron absorption cross

22 section. ASTM A887–89 defines eight types of borated stainless steels (304B and 304B1–

23 304B7) with natural boron concentrations (including both B-10 and B-11 isotopes) ranging from

24 0.2 to 2.25 weight percent (ASTM International, 2009). Boron is essentially insoluble in 

25 stainless steel, and thus it is present as iron and chromium borides (Fe2B, Cr2B) in a steel 

26 matrix. 

27 Of the identified aging mechanisms for neutron poison plates discussed in Section 3.4 above, 

28 the following were removed from consideration for aging effects of borated stainless steels,

29 because they were determined not to be reasonably credible: (i) general corrosion, (ii) galvanic 

30 corrosion and (iii) wet corrosion and blistering. The technical justifications for the decisions to 

31 eliminate these aging mechanisms follow.

32 • General corrosion: Similar to other austenitic stainless steel alloys, borated stainless 

33 steel exhibits passive behavior in helium environments, and thus, general corrosion 

34 rates are expected to be negligible. 

35 • Galvanic corrosion: Borated stainless steel could be coupled to steel, aluminum, or 

36 other stainless steel alloys. The galvanic corrosion behavior of stainless steel is 

37 complicated by the fact that its relative nobility with respect to other materials may 

38 depend on whether a passivating oxide film is present. Nevertheless, both passivated 

39 and nonpassivated stainless steels are generally more noble than steel and aluminum 

40 (Baboian, 2003). In addition, there is no aqueous electrolyte inside the cask or canister

41 to support galvanic corrosion in the helium environment. 

42
3-54

1 • Wet corrosion and blistering: Because borated stainless steel is solid without porosity,

2 no water can be trapped inside the material. Thus, wet corrosion and blistering are not

3 considered to be credible.

4 More detailed discussions regarding the other aforementioned potential aging mechanisms for 

5 borated stainless steel are provided below.

6 3.4.1.1 Boron depletion

7 Boron depletion in boron-based neutron poison plates refers to the loss of boron and hence the 

8 loss of the neutron-absorbing capacity of a material when it is exposed to neutron fluence. For 

9 example, under a neutron fluence, boron-10 nuclei capture neutrons, yielding excited Boron-11 

10 nuclei, which in turn decay into alpha particles and Lithium-7 nuclei. In this nuclear reaction, 

11 one neutron absorption reaction results in the loss of one boron-10 atom. Significant depletion 

12 of boron-10 atoms may occur if the poison material is exposed to sufficient neutron fluence. 

Borated stainless steel typically has an areal density of 1019 to 1021 boron-10 atoms/cm2 13

[6.5 × 1019 to 1021 boron-10 atoms/in2 14 ] (EPRI, 2009). The boron areal density can reach this 

15 level by adjusting the thickness of the poison plate, by adjusting the weight fraction of added 

16 boron, and through the use of enriched boron (i.e., boron-10) (EPRI, 2009). A neutron flux of 

104

–106 n/cm2-s [6.5 × 104–6.5 × 106 n/in2 17 -s] is typical for dry cask storage (Sindelar et al., 

18 2011). At a typical neutron flux and boron-10 concentration, the neutron poison plates would 

19 deplete at most 0.0002 percent of the available boron-10 atoms after 60 years of storage. 

20 Using the highest expected neutron flux and the lowest boron-10 concentration as a most 

21 conservative scenario, only 0.02 percent of the available boron-10 atoms would be depleted 

22 after 60 years, an amount too small to decrease the criticality control function of the neutron23 absorbing materials. As such, boron depletion is not considered to be credible, and therefore, 

24 aging management is not required during the 60-year timeframe.

25 Although boron depletion in borated stainless steel is not generally considered to be a credible 

26 aging mechanism, the reviewer nevertheless should ensure that the renewal application 

27 addresses any depletion analyses that exist in the orginal design basis to consider the 

28 implication of extending the operating period to 60 years. Staff guidance for the review of such 

29 TLAAs is provided in NUREG–1927. 

30 3.4.1.2 Creep

31 As discussed in Section 3.2.1.6, as a general rule of thumb, significant creep can occur at 

32 temperatures above 0.4Tm, where Tm is the melting point of the metal in Kelvin (Cadek, 1988). 

33 At these temperatures, plastic deformation or distortion can occur over long times, even under 

34 stresses that normally would not be considered sufficient to cause yielding of the material. 

35 Robino and Cieslak (1997) show that borated stainless steel has a melting range of 

36 1,250−1,340 degrees C [2,282−2,444 degrees F], corresponding to the melting of borides and 

37 the austenitic structure, which is slightly lower than standard nonborated stainless steel. 

38 Applying the 0.4Tm rule, a temperature range of 336−372 degrees C [637−702 degrees F] is 

39 required to initiate significant creep in borated stainless steels, which is below the estimated 

40 peak fuel cladding temperature of 400 degrees C [752 degrees F] at the beginning of the 

41 storage period (Jung et al, 2013). The maximum cladding temperature is estimated to drop 

42 below the creep range (336 degrees C [637 degrees F]) in fewer than 9 years, well before the 

43 period of extended operation. Also, the borated stainless steel poison plates, which are used in 

44 the verticle DSSs, are not expected to be under loads other than their own weight, and in many 
3-55

1 instances, their weight is also supported by adjacent structures. As such, creep of borated 

2 stainless steel is not considered to be credible, and therefore, aging management is not 

3 required during the 60-year timeframe.

4 3.4.1.3 Thermal aging

5 As previously discussed in Section 3.2.2.8, the microstructures of most stainless steels will 

6 change, given sufficient time at elevated temperatures, and this can affect its mechanical 

7 properties. The thermal aging resistance highly depends on material chemical composition and 

8 microstructure. Borated stainless steel alloys consist of (Fe,Cr)2B precipitates dispersed in an 

9 austenite stainless steel matrix. Robino and Cieslak (1997) showed that the estimated peak fuel 

10 cladding temperature of 400 degrees C [752 degrees F] in storage (Jung et al, 2013) is well 

11 below the temperatures that are needed to cause a change in the boride precipitates. Also, as 

12 discussed in Section 3.2.2.8, the austenite matrix is not expected to be susceptible to 

13 microstructure changes until temperatures exceed 1,000 degrees C [1,832 degrees F]. As 

14 such, thermal aging of borated stainless steel is not considered to be credible, and therefore,

15 aging management is not required during the 60-year timeframe.

16 3.4.1.4 Radiation embrittlement

17 Embrittlement of metals occurs when radiation displaces atoms in metal crystal structures, 

18 creating defects. Neutron radiation (rather than gamma radiation) has the greatest potential to 

19 cause this phenomenon. Depending on the neutron fluence, radiation can cause changes in 

20 mechanical properties such as loss of ductility, fracture toughness, and resistance to cracking.

21 Neutron embrittlement effects on the mechanical properties and the microstructures of borated 

22 stainless steel were studied by irradiating borated stainless steel to different radiation levels, 

from 1013 to 1017 n/cm2 [6.5 × 1013 to 1017 n/in2 23 ] (Soliman et al., 1991). Tests included samples 

24 manufactured by both powder metallurgical and conventional wrought processes. The energy 

25 of the neutron source was such that approximately 20 percent of the neutron flux had an energy 

26 above 0.1 megaelectron-volt (MeV), meaning that a significant portion of the flux contained the 

27 most damaging intermediate or fast neutrons. The investigators reported that there was almost 

no change in mechanical properties with the fluence level up to 1017 n/cm2 [6.5 × 1017 n/in2 28 ]. As 

29 discussed in Section 3.2.1.9, for dry cask storage, the maximum potential accumulated neutron 

fluence on DSS basket components after 100 years was calculated to be 2.63 × 1016 n/cm2 30

[1.70 × 1017 n/in2 31 ], which is about one order of magnitude below the level of that used in the 

32 tests by Soliman et al. (1991). In addition, neutron flux decreases with time during storage, 

33 which will limit the radiation effects. As such, radiation embrittlement of borated stainless steel 

34 is not considered to be credible, and therefore, aging management is not required during the 

35 60-year timeframe.

36 3.4.2 Borated aluminum alloys and aluminum-based composites

37 As in stainless steels, boron is essentially insoluble in aluminum. In borated aluminum, boron is 

38 present in the form of aluminum or titanium boride precipitates (AlB2, TiB2) that reside in an 

39 aluminum matrix. In aluminum metal-matrix composites, boron is in the form of boron carbides

(B4C) in an aluminum matrix. The laminate composites (e.g., Boral® 40 ) consist of (i) a core of 

41 uniformly distributed boron carbide and aluminum alloy particles and (ii) a surface cladding of 

42 aluminum alloy on both sides of the core. 
3-56

1 Of the identified potential aging mechanisms for neutron poison plates listed in Section 3.4 

above, wet corrosion and blistering are considered to be credible only for Boral® 2 , because only 

3 this material has porosity that can trap water and initiate this mechanism. Detailed discussions 

4 of all aging mechanisms for aluminum-based poison materials are provided below.

5 3.4.2.1 General corrosion

6 Because aluminum is present as a continuous matrix (borated aluminum and aluminum 

metal-matrix composites) or used as an outer cladding (Boral®) 7 ), the degree of general corrosion 

8 of each of the neutron poison plate materials is considered to be largely governed by the 

9 corrosion of aluminum. As discussed in Section 3.2.3.1 for other aluminum components, 

10 aluminum forms a protective oxide film at temperatures below approximately 230 degrees C 

11 [446 degrees F]. Above this temperature, the protective film no longer forms if water or steam is 

12 present. As such, general corrosion of aluminum is possible if aluminum were exposed to 

13 moisture in the internal helium environment. However, there is very little residual water in the 

14 cask or canister internal environment following drying. Assuming a residual water content of 1 L

15 [0.26 gal], Jung et al. (2013) calculated that oxidation of all aluminum in the basket assembly is 

16 limited to 0.54 g [0.019 oz], which is equivalent to a 2-μm [0.079-mils]-thick layer of aluminum 

over a surface area of 1,000 cm2 [155 in2 17 ]. Thus, the potential for material thinning from 

18 oxidation is a very small fraction of the aluminum poison materials used inside the system. As a 

19 result, general corrosion is not considered to be credible, and therefore, aging management is 

20 not required during the 60-year timeframe.

21 3.4.2.2 Galvanic corrosion

22 Galvanic corrosion occurs when two dissimilar metals or conductive materials are in physical 

23 contact in the presence of a conducting solution (Baboian, 2003; Hack, 1993). The 

24 aluminum-based neutron poison materials used inside DSSs can be in galvanic contact with 

25 stainless steel, where aluminum is less noble.

26 As discussed above in the evaluation of general corrosion, there is very little residual water 

27 within a cask or canister following drying. Thus, there is a limited potential for the presence of a 

28 conducting solution that can support galvanic corrosion. As a result, loss of material due to 

29 galvanic corrosion is not considered to be credible, and therefore, aging management is not 

30 required during the 60-year timeframe.

31 3.4.2.3 Wet corrosion and blistering

32 The core of aluminum-boron carbide laminate composites is not fully sintered and, as a result, 

33 can have a porosity of 1 to 8 percent with varying degrees of interconnectivity among pores. 

34 This may allow water ingress into the core, where the water can react with the aluminum to form 

35 aluminum oxide and hydrogen gas (EPRI, 2009; 2012). Blistering has been observed in the 

Boral® 36 cladding in wet and dry storage applications. Tests simulating the wetting and vacuum 

drying cycles during canister closure operations show that Boral® 37 can form blisters in the 

38 aluminum cladding because of water ingress through its exposed edges (EPRI, 2004). The 

39 blisters are characterized by a local area where the aluminum cladding separates from the 

40 underlying boron carbide-aluminum core, and the cladding is physically deformed outward. 

41 Although wet corrosion and blistering may occur, this aging mechanism has not been observed 

to reduce the neutron absorbing capability of Boral® 42 in spent fuel pool surveillance coupons 

43 (EPRI, 2009). It is important to note that, because only a trace amount of water will be left in a 
3-57

1 dry storage cask after dehydradation and helium backfill, the occurrence of wet corrosion and 

2 blistering will be minimal in a dry cask environment during the period of extended operation. 

3 Therefore, wet corrosion and blistering are not considered to be an aging mechanism requiring 

aging management and aging management is not required for Boral® 4 in the DSSs with respect 

5 to criticality safety during the 60-year timeframe.

6 3.4.2.4 Boron depletion

7 Boron depletion refers to the loss of the capability of a material to absorb neutrons when the 

8 neutron fluence significantly consumes boron-10 atoms. Neutron poison plates typically contain 

1019 to 1021 boron-10 atoms/cm2 [6.5 × 1019 to 1021 boron-10 atoms/in2 9 ] (EPRI, 2009). A neutron 

flux of 104

–106 n/cm2-s [6.5 × 104–6.5 × 106 n/in2 10 -s] is typical for dry cask storage (Sindelar et 

11 al., 2011). Under a neutron flux, boron-10 nuclei capture neutrons, yielding excited boron-11 

12 nuclei, which, in turn, decay into high-energy alpha particles and lithium-7 nuclei. In this nuclear 

13 reaction, one neutron would deplete one boron-10 atom. At typical levels of neutron flux and 

14 boron-10 concentration, the neutron dose after 60 years would deplete at most 0.0002 percent

15 of the available boron-10 atoms. Using the highest expected neutron flux and the lowest boron16 10 concentration as a worst case scenario, only 0.02 percent of the available boron-10 atoms 

17 would be depleted after 60 years, which is too small to challenge the criticality control function of 

18 the neutron poisons. As such, boron depletion for borated aluminum alloys, aluminum metal 

matrix composites, and Boral® 19 is not expected to result in significant changes in the criticality 

20 control function. As such, boron depletion is not considered to be credible, and therefore, aging 

21 management is not required during the 60-year timeframe.

22 Although the above generic evaluation does not identify boron depletion as a significant aging 

23 mechanism, the reviewer nevertheless should ensure that the renewal application addresses 

24 any depletion analyses that exist in the orginal design basis to consider the implication of 

25 extending the operating period to 60 years. Staff guidance for the review of such TLAAs is 

26 provided in NUREG–1927.

27 3.4.2.5 Creep

28 As discussed in Section 3.2.1.6, as a general rule of thumb, significant creep occurs at 

29 temperatures above 0.4Tm, where Tm is the melting point of the metal in Kelvin (Cadek, 1988). 

30 At these temperatures, plastic deformation or distortion can occur over long times, even under 

31 stresses that normally would not be considered sufficient to cause yielding of the material. 

32 Because aluminum is present as a continuous matrix and as an external cladding in the neutron 

33 poison plates, and aluminum has a lower melting point than the other portions of the material

34 microstructures (e.g., aluminum and titanium borides, boron carbides), the creep behavior of 

35 poison materials is considered to be governed by the behavior of aluminum. Applying the 0.4Tm

36 rule, the critical creep temperature for aluminum is 100 degrees C [212 degrees F]. 

37 The highest temperatures within DSSs are at locations close to the fuel rods. For example, the 

38 maximum expected temperature of the cladding on the fuel rods has been estimated to be

39 400 degrees C [752 degrees F] at the beginning of the storage period, and the cladding 

40 temperatures are expected to decrease to approximately 266 degrees C [510 degrees F] after 

41 20 years and 127 degrees C [261 degrees F] after 60 years (Jung et al., 2013). These 

42 estimates depend on many factors, such as the initial heat load of the SNF. It is apparent from 

43 these temperatures that subcomponents within the cask or canister could be exposed to 

44 temperatures above the minimum creep temperatures for aluminum during at least the first 

45 40 years.
3-58

1 Because temperatures within DSSs have the potential to exceed the minimum creep 

2 temperature of aluminum, it is necessary to consider the load applied to the subcomponent to 

3 determine whether significant creep deformation will occur, as well as the specific application to 

4 determine whether the creep affects safety. Typically, neutron poison plates do not serve a 

5 structural function and are thus not expected to be under loads other than their own weight. 

6 Also, in many instances, their weight is also supported by adjacent structures. For example, the 

7 neutron poison plates in the Holtec HI-STORM 100 system are completely enclosed in stainless 

8 steel sheating (Holtec International, 2014). Due to the minimal applied loads and presence of 

9 adjacent supporting structures, the impact of creep on the criticality control function of the 

10 neutron poisons is not considered to be credible, and therefore, aging management is not 

11 required during the 60-year timeframe. 

12 3.4.2.6 Thermal aging

13 Prolonged exposure to elevated temperatures can lead to a loss of fracture toughness and 

14 ductility in some materials as a result of changes to their microstructure. Testing of 

15 aluminum-based neutron poison plates, however, has shown that these materials typically

16 increase in ductility when they are aged at high temperatures. For example, a series of 

17 elevated temperature tensile tests on an aluminum metal-matrix composite (METAMICTM) found 

18 an increase in elongation to break (a measure of ductility) when the material was aged at 

19 399 degrees C [750 degrees F] for 8,523 hours (EPRI, 2009). These and other material 

20 qualification tests performed on neutron poisons demonstrate that microstructural changes 

21 induced by aging typically make the aluminum softer and more ductile as it is annealed, while 

22 the boride and carbide particulates are thermally stable at cask internal temperatures. 

23 Also, as discussed above for the creep mechanism, decreases in strength due to thermal aging 

24 are not expected to affect the criticality control function of the poison plates, because they

25 typically do not serve a structural function and may be supported by adjacent structures. 

26 Consequently, thermal aging of neutron poison materials is not considered to be credible, and 

27 therefore, aging management is not required over the 60-year timeframe.

28 3.4.2.7 Radiation embrittlement

29 As discussed in Section 3.4.1.4 above, embrittlement of metals may occur under exposure to 

30 radiation. Neutron radiation (rather than gamma radiation) has the greatest potential to cause 

31 this phenomenon. 

32 Depending on the neutron fluence, radiation can cause changes in mechanical properties such 

33 as loss of ductility, fracture toughness, and resistance to cracking. Farrell and King (1973) 

34 showed that pure aluminum had increased strength but decreased ductility after being irradiated 

to fast neutron fluences (energy greater than 0.1 MeV) in the range of 1 to 3 × 1022 n/cm2 35

[6.5 to 19.4 × 1022 n/in2 36 ] from a research reactor for 8 years. However, these radiation levels 

37 are six orders of magnitude higher than the estimated fluence after dry storage for 100 years as 

38 discussed in Section 3.2.1.9. 

39 Some results from radiation testing of aluminum-based neutron poisons are reported in the 

literature (EPRI, 2009). Gamma, thermal neutron, and fast neutron radiation testing of Boral® 40 in 

water was performed for 9 years. With exposures of to up to 7 × 1011 41 rad of gamma, 

3.6 × 1018 n/cm2 [2.3 × 1019 n/in2

] fast neutron fluence, and 2.7 × 1019 n/cm2 [1.7 × 1020 n/in2 42 ] 

43 thermal neutron fluence, the specimen showed no change in ultimate strength and no other 

44 signs of physical deterioration, except for severe oxidation because of the presence of water. 
3-59

Also, radiation testing of a sintered composite subjected to up to 1.5 × 1020 n/cm2 1

[9.7 × 1020 n/in2

] fast neutron fluence and a maximum of 3.8 × 1011 2 rad gamma exposure 

3 showed little change in the yield strength and ultimate strength (EPRI, 2009). Finally, neutron 

radiation of borated aluminum to fluences of 1017 n/cm2 [6.5 × 1017 n/in2 4 ] showed no dimensional

5 change or radiation damage (EPRI, 2009). These test conditions are expected to be more 

6 severe than those experienced by the aluminum-based neutron poison materials in the 

7 extended storage application (EPRI, 2009). Therefore, radiation embrittlement of borated 

aluminum alloys, aluminum metal-matrix composites, and Boral® 8 is not expected to be credible.

9 Consequently, aging management is not required during the 60-year timeframe.

10 3.4.3 References

11 Alexander, D.J. “Effects of Irradiation on the Mechanical Properties of 6061-T651 

12 Aluminum Base Metal and Weldments.” ASTM Special Technical Publication. Vol. 1325. 

13 pp. 1,027–1,044. 1999.

14 ASTM International. “Standard Specification for Borated Stainless Steel Plate, 

15 Sheet, and Strip for Nuclear Application.” ASTM A887–89. West Conshohocken, 

16 Pennsylvania: ASTM International. 2009.

17 Baboian, R. “Galvanic Corrosion.” In ASM Handbook, Vol. 13A, Corrosion: Fundamentals, 

18 Testing, and Protection. Materials Park, Ohio: ASM International. pp. 210–213. 2003.

19 Cadek, J. Creep in Metallic Materials. Elsevier Science Publishing Company, Inc. 1988.

20 Chopra, O., D. Diercks, R. Fabian, Z. Han, and Y. Liu. “Managing Aging Effects on Dry Cask 

21 Storage Systems for Extended Long-Term Storage and Transportation of Used Fuel.” 

22 FCRD–UFD–2014–000476. ANL–13/15, Rev. 2. Washington, DC: U.S. Department of 

23 Energy. 2014.

EPRI. “Strategy for Managing the Long-Term Use of BORAL® 24 in Spent Fuel Storage Pools.” 

25 Report 1025204. Palo Alto, California: Electric Power Research Institute. 2012.

26 _____. “Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Transportation and 

27 Storage Applications.” Report 1019110. Palo Alto, California: Electric Power Research 

28 Institute. 2009.

_____. “BORAL® 29 Behavior Under Simulated Cask Vacuum Drying. Part 2 Test Results.” 

30 Report 1009696. Palo Alto, California: Electric Power Research Institute. 2004.

31 Farrell, K. and R.T. King. “Radiation-Induced Strengthening and Embrittlement in Aluminum.” 

32 Metallurgical Transactions A. Physical Metallurgy and Materials Science. Vol. 4. 

33 pp. 1,223–1,231. 1973.

34 Gibeling, J.C. “Creep Deformation of Metals, Polymers, Ceramics, and Composites.” In 

35 ASM Handbook. Vol. 8. Mechanical Testing and Evaluation. Materials Park, Ohio: 

36 ASM International. pp. 363–368. 2000.

37 Hack, H.P. Galvanic Corrosion Test Methods. Houston, Texas: NACE International. 1993.
3-60

1 Hanson, B., H. Alsaed, C. Stockman, D. Enos, R. Meyer, and K. Sorenson. “Gap Analysis to 

2 Support Extended Storage of Used Nuclear Fuel.” Rev. 0. FCRD–USED–2011–000136. 

3 PNNL–20509. Washington, DC: U.S. Department of Energy. 2012.

4 Holtec International. “Final Safety Analysis Report for the HI-STORM 100 Cask System, 

5 Revision 12.” Holtec Report No. HI-2002444. USNRC Docket No. 72-1014. pp. 1.2-18. 

6 ADAMS Accession No. ML14086A410. 2014

7 Jung, H., P. Shukla, T. Ahn, L. Tipton, K. Das, X. He, and D. Basu. “Extended Storage and 

8 Transportation: Evaluation of Drying Adequacy.” ADAMS Accession No. ML13169A039. 

9 San Antonio, Texas: Center for Nuclear Waste Regulatory Analyses. 2013.

10 Kaufman, J.G. Properties of Aluminum Alloys—Tensile, Creep, and Fatigue Data at High and 

11 Low Temperatures. Materials Park, Ohio: ASM International. 1999.

12 NRC. “Identification and Prioritization of the Technical Information Needs Affecting Potential 

13 Regulation of Extended Storage and Transportation of Spent Nuclear Fuel.” Washington, DC: 

14 U.S. Nuclear Regulatory Commission. May 2014.

15 _____. NUREG−1801, “Generic Aging Lessons Learned (GALL) Report.” Rev. 2. 

16 Washington, DC: U.S. Nuclear Regulatory Commission. 2010.

17 NWTRB. “Evaluation of the Technical Basis for Extended Dry Storage and Transportation of 

18 Used Nuclear Fuel.” Washington, DC: Nuclear Waste Technical Review Board. 2010.

19 Robino, C.V. and M.J. Cieslak. “Fusion Welding of a Modern Borated Stainless Steel.” Welding 

20 Journal. Vol. 76, No. 1. pp. 11-s – 23-s. 1997.

21 Sindelar, R.L., A.J. Duncan, M.E. Dupont, P.-S. Lam, M.R. Louthan, Jr., and T.E. Skidmore. 

22 NUREG/CR–7116, “Materials Aging Issues and Aging Management for Extended Storage and 

23 Transportation of Spent Nuclear Fuel.” Washington, DC: U.S. Nuclear Regulatory Commission. 

24 2011.

25 Soliman, S.E., D.L. Youchison, A.J. Baratta, and T.A. Ballrelt. “Neutron Effects on Borated 

26 Stainless Steel.” Nuclear Technology. Vol. 96. pp. 346–352. 1991.

27
3-61

1 3.5 Concrete overpacks, support pads, and ceramic fiber insulation

2 Concrete overpacks and support pads include various structural subcomponents constructed of 

3 concrete and reinforcing steel, as well as pad-supporting materials constructed of engineered 

4 fill, natural soil, or treated soil. These subcomponents may be exposed to several 

5 environments, such as outdoor air, groundwater or soil, and flowing water, or they may be 

6 sheltered or embedded in concrete or steel. The environment may also include elevated 

7 temperatures due to heat released by the SNF and radiation, with dose rates depending on the 

8 SNF characteristics (e.g., burnup and age of fuel), exposure time, and location of the 

9 subcomponent. 

10 Potential aging mechanisms for the concrete overpack and pad subcomponents were identified 

11 from reviews of gap assessments of DSSs, relevant technical literature, American Concrete 

12 Institute (ACI) guides and reports, and operating experience from nuclear and nonnuclear 

13 applications (NRC, 2014, 2011a, 2010a; Chopra et al., 2014; Hanson et al., 2011; 

14 NWTRB, 2010). Additional mechanisms were identified during a recent NRC concrete expert 

15 panel workshop (NRC, 2015). Thermal, mechanical, chemical, and irradiation-induced 

16 degradation mechanisms were identified as follows:

17 • freeze and thaw

18 • creep

19 • reaction with aggregates

20 • aggressive chemical attack

21 • corrosion of reinforcing steel

22 • differential settlement

23 • shrinkage

24 • leaching of calcium hydroxide

25 • radiation damage

26 • fatigue

27 • dehydration at high temperature

28 • microbiological degradation

29 • delayed ettringite formation

30 • salt scaling 

31 In addition, a review of known degradation modes for ceramic fiber insulation was performed, 

32 which resulted in consideration of the following:

33 • radiation damage

34 • moisture absorption

35 Potential mechanisms were refined by considering the thermal, mechanical, chemical, and 

36 irradiation conditions specific to each subcomponent. This process eliminated several 

37 mechanisms from consideration for some subcomponents in the AMR tables in Chapter 4. 

38 Detailed discussions regarding potential aging mechanisms for each material and the technical 

39 bases for those requiring aging management are included in the following sections. 

40 These discussions do not consider potential synergistic effects, if any, due to coupled 

41 degradation mechanisms. Coupled degradation mechanisms in concrete refer to degradation 

42 modes that can interact, affecting their relative times for initiation and progression

43 (e.g., freeze-thaw cracking that leads to water ingress and subsequent leaching of calcium 
3-62

1 hydroxide). Few in-depth studies have been published on the effects of concrete damage 

2 caused by these potential coupled degradation mechanisms. However, the staff expects that an 

3 AMP is an adequate approach for addressing potential synergistic effects due to coupled 

4 degradation mechanisms. The example of an AMP for concrete structures in Chapter 6 relies 

5 on the licensee’s corrective action program to ensure that conditions that may lead to a loss of 

6 intended function will be reviewed and dispositioned by trained personnel. If a particular aging 

7 effect is detected, part of the licensee’s corrective action may include a root-cause evaluation to 

8 determine the cause of the aging effect. If the root-cause evaluation determines that the rate of 

9 degradation is being accelerated by the effects of coupled degradation modes, followup 

10 corrective actions may include a review of the inspection or monitoring procedures to ensure 

11 that aging management activities remain adequate for the remaining period of extended 

12 operation. 

13 3.5.1 Concrete

14 3.5.1.1 Freeze and thaw

15 Concretes exposed to outdoor and groundwater/soil (below-grade) environments above the 

16 freeze line

17 Concretes that are nearly or fully saturated with water can be damaged by repeated freezing 

18 and thawing cycles in environments with weathering indexes (i.e., the product of the average 

19 annual number of freezing cycle days and the average annual winter rainfall in inches) on the 

20 order of 100 day-in/yr or greater (NRC, 2010a). For environments with weathering indexes less 

21 than 100 day-in/yr, freeze and thaw degradation is not considered to be significant. The 

22 weathering index for the continental United States and adequate data sourcing for determining 

23 the weathering for any locality can be found in ASTM C216 (ASTM, 2016). For below-grade 

24 concrete structures above the freeze line, water that resides in soil can also be subject to 

25 freezing conditions, potentially promoting freeze and thaw damage.

26 Freeze and thaw damage has been observed in outdoor concrete structures in nuclear power 

27 plants (NRC, 1995, 2012). Because water expands when freezing, fully or mostly saturated 

28 concrete will experience internal stresses from the expanding ice, which can cause concrete 

29 cracking or scaling when pressures exceed the concrete tensile strength (ACI, 2008c; Pigeon, 

30 1994; Marchand et al., 1994; Sawan, 1987; Fagerlund,1977).

31 The degradation mode would initiate at the outer concrete surface of the DSS exposed to 

32 outdoor environments, primarily at horizontal surfaces where water ponding can occur. 

33 Operating experience has identified freeze and thaw damage in the roofs of the concrete 

34 storage modules at the Three Mile Island Unit 2 (TMI-2) and the Millstone independent spent 

35 fuel storage installation (ISFSI) (NRC, 2012).

36 Therefore, freeze and thaw damage is considered credible in concrete exposed to outdoor and 

37 groundwater or soil (below-grade) environments above the freeze line, and aging management 

38 is required during the 60-year timeframe.

39

40
3-63

1 Concretes exposed to sheltered environments, fully encased (lined) in steel, and exposed to 

2 groundwater/soil (below-grade) environments under the freeze line

3 Freeze and thaw degradation of concrete exposed to sheltered environments with low water 

4 availability is not considered credible; the heat load from the fuel in the DSS is expected to aid 

5 in drying the interior concrete surfaces of the overpacks, preventing freeze and thaw damage. 

6 Freeze and thaw degradation of concrete exposed to groundwater or soil (below-grade)

7 environments at temperatures above freezing is not considered credible. 

8 Freeze and thaw damage also is not considered credible for concrete fully encased in metallic 

9 liners (not in direct contact with outdoor environments or proven to be protected from water 

10 ingress); the lack of water transfer from the outside environment into the concrete prevents the 

11 degradation mechanism.

12 Therefore, aging management of concrete for freeze and thaw degradation in these 

13 environments is not required.

14 3.5.1.2 Creep

15 Creep in concrete is the time-dependent deformation resulting from sustained loads (Wang and 

16 Salmon, 1998). Cement paste in concrete exhibits creep due to its porous structure and a large 

17 internal surface area that is sensitive to water movements. Creep manifests as cracking on the 

18 concrete outer surfaces and causes redistributions of internal forces. Factors affecting creep 

19 are concrete constituents (composition and fineness of the cement; admixtures; and size, 

20 grading, and mineral content of aggregates), water content and water-cement ratio, curing 

21 temperature, relative humidity, concrete age at loading, duration and magnitude of loading, 

22 surface-volume ratio, and slump (Wang and Salmon, 1998; Neville and Dilger, 1970). However, 

23 the most important parameter controlling creep is concrete sustained loading. Creep increases 

24 with increasing load and temperature (McDonald, 1972). However, the creep rate decreases 

25 exponentially with time (Branson, 1977; NRC, 2014; Wang and Salmon, 1998). In summary, in 

26 the case of a given concrete mix design, concrete creep is generally understood to be a 

27 phenomenon that would affect concrete structures early in the service life under sustained 

28 loading. Thus, the age of concrete and the magnitude and duration of sustained loading are the 

29 primary factors that determine the magnitude of the creep of concrete (Neville and Dilger, 1970). 

30 For example, if a sustained load is applied on 2-year-old and 40-year-old concrete, the 

31 2-year-old concrete will have significantly more creep. Also, the creep in concrete could largely 

32 be mitigated by proper design practices, in accordance with ACI 318-05 (ACI, 2005) or 

33 ACI 349-06 (ACI, 2007). Furthermore, creep-induced concrete cracks are not generally large 

34 enough to reduce the compressive strength of concrete, cause deterioration of concrete, or 

35 cause exposure of reinforcing steel to the environment. In a DSS, the initial sustained load is 

36 normally low, and no significant change of load is expected during the 40-year timeframe 

37 beyond initial licensing. Thus, creep is not considered credible for any environment, and aging 

38 management is not required during the 60-year timeframe.

39 3.5.1.3 Reaction with aggregates

40 The two most common alkali-aggregate reactions are alkali-silica reaction (ASR) and 

41 alkali-carbonate reaction, with ASR being the most common and damaging. ASR is a chemical 

42 reaction between hydroxyl ions (present in the alkaline cement pore solution) and reactive forms 

43 of silica present in some aggregates (e.g., opal, chert, chalcedony, tridymide, cristabolite, 
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1 strained quartz). An aggregate that presents a large surface area for reaction (i.e., amorphous, 

2 glassy) is susceptible to ASR (Poole, 1992). The resulting chemical reaction produces an alkali3 silica gel that swells with the absorption of moisture, exerting expansive pressures within the 

4 concrete (Figg, 1987). ASR damage in the concrete manifests as a characteristic map cracking 

5 on the concrete surface (ACI, 2008a). The internal damage results in the degradation of 

6 concrete mechanical properties, and in severe cases, the expansion can result in undesirable 

7 dimensional changes and popouts. In reinforced concrete, cracks tend to align parallel to the 

8 direction of maximum restraint and rarely progress below the level of the reinforcement. In 

9 general, ASR is a slow degradation mechanism that can cause serviceability issues and may 

10 exacerbate other deterioration mechanisms. 

11 The requisite conditions for initiation and propagation of ASR include (i) a sufficiently high alkali 

12 content of the cement (or alkali from other sources, such as deicing salts, seawater, and 

13 groundwater), (ii) a reactive aggregate, and (iii) available moisture, generally accepted to be 

14 relative humidity greater than 80 percent (Pedneault, 1996; Stark, 1991). A study by the 

15 California Department of Transportation (Glauz et al., 1996) revealed that ASR increases 

16 proportionally to the cement content, alkali content greater than 0.6 percent can accelerate 

17 ASR, high calcium oxide content can promote ASR, and the use of various types of admixtures 

18 in certain doses can mitigate ASR (ACI, 2008a; ASTM, 1998). At higher concentrations of alkali 

19 hydroxides, even the more stable forms of silica are susceptible to ASR attack (Xu, 1987). 

20 Repeated cycles of wetting and drying can accelerate ASR (ACI, 1998). As a result, it is 

21 desirable to minimize both available moisture and wet-dry cycles by providing good drainage. 

22 Moreover, concretes exposed to warm environments are more susceptible to ASR than those 

23 exposed to colder environments (Perenchio et al., 1991).

24 As mentioned earlier, ASR is generally a slow degradation mechanism. ASR may take from 

25 3 to more than 25 years to develop in concrete structures, depending on the nature 

26 (reactivity level) of the aggregates, the moisture and temperature conditions to which the 

27 structures are exposed, and the concrete alkali content (Thomas et al., 2013). The delay in 

28 exhibiting deterioration indicates that there may be less reactive forms of silica that can 

29 eventually cause deterioration (Mindess and Young, 1981). Recent operating experience has 

30 revealed degradation of the concrete in the Seabrook reactor containment as a result of ASR 

31 (NRC, 2011b). The concrete used at the Seabrook plant passed all industry standard ASR 

32 screening tests (ASTM, 2007, 2012) at the time of construction. However, ASR-induced 

33 degradation was identified in August 2010. In addition, ASR screening tests are not conducted 

34 on each aggregate source but rather in select batches, which increases the risk for use of 

35 aggregates of different reactivities when procured from different sources. Due to the 

36 uncertainties in screening tests that can effectively be used to eliminate the potential for ASR 

37 and previous ASR operating experience at a nuclear facility, the aging mechanism is considered 

38 credible in concrete exposed to any environment with available moisture, and therefore, aging 

39 management is required during the 60-year timeframe. 

40 3.5.1.4 Differential settlement

41 Differential settlement is a result of the uneven deformation of the supporting foundation soil 

42 (Das, 1999; NAVFAC, 1986). The factors affecting structural settlement include the type of 

43 foundation soil (e.g., clayey soil, sandy soil) and its physical properties, thickness of soil layers, 

44 water-table level, depth of foundation mat below the ground surface, liquefaction during seismic 

45 events, and load. Differential settlement, which causes distortion (loss of form) and damage 

46 (cracking) to concrete structures, is a function of the uniformity of the soil, stiffness of the 
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1 structure, stiffness of the soil, and distribution of loads within the structure (U.S. Department of 

2 the Army, 1990; NAVFAC, 1996). 

3 The settlement of saturated cohesive soil consists of three components: (1) immediate 

4 settlement occurring due to the applied load, (2) consolidation settlement occurring gradually 

5 due to dissipation of the excess pore pressures generated by the applied load, and 

6 (3) secondary compression that depends on the composition and structure of the soil skeleton 

7 (NAVFAC, 1986). The settlement of course-grained granular soils subject to applied load 

8 occurs immediately, primarily from the compression of the soil skeleton due to rearrangement 

9 of particles. However, most settlement issues involving a combination of immediate 

10 settlement and progressing long-term settlement are typically discovered in less than 1 year 

11 of construction. 

12 Differential settlement is addressed during the design-basis calculations. The analyses 

13 generally include calculations to predict differential settlement based on the sequential DSS 

14 placement; the analyses are used to determine an optimum DSS placement sequence to limit 

15 differential settlement of the ISFSI support pad. However, operating experience has shown that 

16 it can occur; periodic walkdowns ensure these limited occurrences are evaluated on a 

17 case-by-case basis. NUREG–1522, “Assessment of In-service Conditions of Safety-Related 

18 Nuclear Plant Structures” (NRC, 1995), stated that foundation settlement of concrete structures 

19 at Point Beach and Beaver Valley, inspected during walkdowns, experienced appreciable 

20 differential settlement. In addition, the loads on the concrete pad are expected to increase over 

21 time as more loaded DSSs are placed on the pad. Therefore, differential settlement of 

22 concretes exposed to sheltered, outdoor, and groundwater or soil (below-grade) environments 

23 is considered credible, and aging management is required during the 60-year timeframe.

24 3.5.1.5 Aggressive chemical attack

25 The intrusion of aggressive ions or acids into the pore network of the concrete can cause 

26 various degradation phenomena. The aggressive chemical attack typically originates from an 

27 external source of sulfate or magnesium ions as well as acidic environmental conditions. 

28 Depending on the type of aggressive chemical, the degradation of concrete can manifest in the 

29 form of cracking, loss of strength, concrete spalling and scaling, and reduction in concrete pH.

30 Concretes exposed to outdoor and groundwater/soil (below-grade) environments

31 External sulfate attack

32 External sulfate attack is a process whereby ions in species such as K2SO4, Na2SO4, CaSO4, 

33 and MgSO4, which are present in groundwater, seawater, and rainwater, penetrate the concrete 

34 and chemically react with alkali and calcium ions to form a precipitate of calcium sulfate in 

35 addition to other forms of calcium and sulfate-based compounds (e.g., ettringite). The 

36 manifestation of sulfate attack is cracking, increase in concrete porosity and permeability, loss 

37 of strength, and surface scaling generated by the expansion associated with the formation of 

38 ettringite within the concrete and the pressure generated by the precipitated calcium and 

39 sulfate-base compounds inside the concrete pore network (Poe, 1998; NWTRB, 2010). Unlike 

40 the alkali sulfates, no decalcification of the calcium silicate hydrate phase occurs in the CaSO4

41 attack. On the other hand, the MgSO4 attack is significantly faster and more thorough than the 

42 attack by the other sulfate compounds because of the limited solubility of magnesium hydroxide 

43 (Mg(OH)2) in the high pH of concrete (Drimalas et al., 2010). In addition, magnesium ions 
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1 present in deicing salts can react with calcium silicate hydrate, gradually converting it to 

2 magnesium silicate hydrate, which is not cementitious in nature.

3 A service life model for sulfate attack in concrete was developed by Atkinson and Hearne 

4 (1990). Cases of sulfate attack in the field are fairly uncommon, mainly because most 

5 transportation regulatory agencies have adopted specifications aimed at preventing this damage

6 mode 

7 (Weiss et al., 2009; Van Dam and Peshkin, 2009). In particular, degradation due to external 

8 sulfate attack has not been reported in nuclear applications. Atkinson and Hearne (1990) 

9 developed a concrete service life model to assess degradation due to sulfate attack. Using 

10 aggressive soil and groundwater conditions (sulfate concentration of 1,500 ppm as specified in 

11 ASME Code Section XI, Subsection IWL (ASME, 1995)) and typical concrete properties 

12 (i.e., elastic modulus, roughness factor, Poisson’s ratio, and concrete porosity), the model 

13 predicts that sulfate damage can occur within 60 years of exposure (Berntz et al., 2001). 

14 Magnesium attack

15 Magnesium ions can rapidly replace calcium ions in the silica hydrate compounds. In 

16 groundwater, magnesium ions are commonly found in the form of MgSO4. The magnesium ion 

17 attack is more commonly observed in arid western U.S. areas and in below-grade structures. At 

18 present, there is no stipulation on the threshold concentration of magnesium ions needed to 

19 promote damage to concrete structures for nuclear and nonnuclear applications. Because 

20 magnesium attack could be part of the sulfate attack, the timeframe implications and exposure 

21 conditions are expected to be comparable to those of sulfate attack.

22 Acid attack

23 Acids with a pH less than 3 can dissolve both hydrated and unhydrated cement compounds 

24 (e.g., calcium hydroxide, calcium silicate hydrates, and calcium aluminate hydrates) as well as 

25 calcareous aggregate in concrete without any significant expansion reaction (Gutt and Harrison, 

26 1997; Mehta, 1986). In most cases, the chemical reaction forms water-soluble calcium 

27 compounds, which are then leached away by aqueous solutions. The dissolution of concrete 

28 commences at the surface and propagates inward as the concrete degrades. The signs of 

29 acidic attack are loss of alkalinity (also disturbing of electrochemical passive conditions for the 

30 embedded steel reinforcement), loss of material (i.e., concrete cover), and loss of strength.

31 The extent and rate of concrete degradation depends on the type, concentration and pH of the 

32 acidic solution, concrete permeability, calcium content in the cement, the water-to-cement ratio, 

33 and the type of cement and mineral admixtures (Pavlik and Uncik, 1997). Sulfuric acid is 

34 particularly aggressive to concrete, because the calcium sulfate formed from the acid reaction 

35 will also deteriorate concrete via sulfate attack (Pavlik, 1994). Even slightly acidic solutions that 

36 are lime deficient can attack concrete by dissolving calcium from the paste, leaving behind a 

37 deteriorated paste consisting primarily of silica gel.

38 Acids can come from groundwater as well as from acid rain containing SO2, NOx, and HCl from 

39 polluted regions, which can compromise the durability of concrete (Webster and Kukacka, 

40 2009). Ueda et al. (2001) proposed a model for acid rain deterioration, which is dependent on 

41 the amount of acid absorption into the concrete, type of acid, mix proportion, and contact time or 

42 interval of rainfalls. The model can predict the depth of concrete damage as a function of 

43 environmental pH. A study by Manjeeth and Rama (2015) found that the compressive strength 

44 and mass loss of concrete samples decreased after 28 days of exposure to sulfuric acid 
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1 solutions with pH ranging from 1 to 7. As such, this degradation mode is expected to affect the 

2 concrete shortly after the concrete surface is in contact with the acid solution.

3 In summary, aggressive chemical attack of concretes exposed to outdoor and groundwater or 

4 soil (below-grade) environments is considered to be credible, and therefore, aging management 

5 is required during the 60-year timeframe.

6 Concretes exposed to sheltered and fully encased (lined) in steel environments

7 With regard to concrete in sheltered environments and fully encased (lined) in steel, external 

8 sources of sulfate, magnesium, and acid entering concrete are considered to be insignificant. In 

9 addition, the heat load from the fuel in the DSS is expected to aid in drying the interior concrete 

10 surfaces, thus decreasing water availability at the concrete surface, which is necessary to 

11 promote this degradation mode. Thus, aggressive chemical attack of sheltered and fully

12 encased (lined) concrete is not considered credible, and therefore, aging management is not 

13 required during the 60-year timeframe.

14 3.5.1.6 Corrosion of reinforcing steel

15 Concretes exposed to outdoor and groundwater/soil (below-grade) environments

16 Corrosion of the reinforcing steel embedded in the concrete is mainly caused by the presence of 

17 chloride ions in the concrete pore solution and carbonation of the concrete. Chloride attack of 

18 concrete structures is well established in the literature (Cheung et al., 2009). The highly alkaline 

19 environment provided by the concrete (normally with pore water pH>13.0) results in the 

20 formation of a metal-adherent oxide film on the reinforcement steel bar surface, which 

21 passivates the steel (Page, 1982). However, chloride ions may penetrate the concrete matrix 

22 and break down the steel passive layer, once the chloride concentration at the reinforcing steel 

23 surface exceeds a threshold value, triggering corrosion of the reinforcing steel and shortening 

24 the service life of a concrete structure. For instance, chlorides may already exist at low levels 

25 within the base mix constituents. In most practical situations, chloride ions penetrate from the 

26 outside environment, such as when using deicing salts, from groundwater, and in marine 

27 environments (Tang and Sandberg, 1996). The presence of corrosion products at the steel 

28 surface can generate internal stresses within the concrete matrix, causing cracks and spalling of 

29 the concrete cover with consequent structural damage.

30 The threshold chloride concentration in concrete required to promote corrosion of the reinforcing 

31 steel depends on the pH of the concrete pore solution. The onset of corrosion can be enhanced 

when acid attack or concrete carbonation1 32 reduces the concrete pH at the steel surface. Thus, 

33 the chloride-to-hydroxide ratio is an important parameter in evaluating the steel corrosion. The 

34 present literature does not provide a clear agreement on the value of the critical chloride ion 

35 concentration required for corrosion initiation. Glass and Buenfeld (1997) have reviewed the 

36 chloride threshold values reported for steel embedded in concrete structures. From this 

37 investigation, it was concluded that a universal, well-defined chloride threshold value does not 

38 exist. The lowest limit of chloride threshold value in concrete ranged from 0.2 to 2.5 percent

39 (by weight of cement). Factors such as the chemical composition of the rebar, as well as its 

 

1Carbonation results from the chemical reaction between the hydrated cement material and atmospheric carbon 

dioxide, which lowers the pH of the concrete and reduces the passivation effect of calcium hydroxide in preventing 

the corrosion of reinforcing steel. The carbonation rate depends on the external CO2 concentration, concrete type, 

temperature, time of wetness of the concrete surface, and degree of moisture (Bertolini et al., 2004).
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1 surface roughness, can influence the chloride threshold (Szklarska-Smialowska, 1986). 

2 Groundwater aggressiveness is defined based on the chloride threshold concentration of 

3 500 ppm [milligram (mg)/kilogram (kg)] with a pH less than 5.5 (ASME, 1995, Section XI, 

4 Subsection IWL). This value is consistent with those recommended in ACI 201.2R-08

5 (ACI, 2008c). 

6 Concrete durability is directly related to the quality of the concrete, the external concentration of 

7 chlorides on the concrete surface, and the reinforcement material. The service life of concretes 

8 exposed to chloride attack depends on the concrete cover, the surface chloride concentration, 

9 the chloride diffusion coefficient, the type of cementitious material, and the reinforcing steel 

10 material. Several service life models have been proposed to determine the durability of 

11 concrete subject to chloride-induced corrosion (Schiessl et al., 2006; DuraCrete, 2000; 

12 Berntz et al., 2001). For example, for a constant surface chloride concentration of 0.05 percent 

13 by weight of concrete (i.e., the maximum chloride concentration in soil and groundwater per 

14 ASME Code Section XI, Subsection IWL ASME, (ASME, 1995)), a 2.54-cm [1-in] concrete 

15 cover, and a chloride threshold of 0.03 percent by weight of concrete, the onset of 

16 chloride-induced corrosion in concrete occurs in about 6, 20, and 120 years for constant 

chloride diffusion coefficients of 6.45 × 10-7 cm2

/second (sec) [10−7 in2 17 /sec] (poor concrete 

quality), 6.45 × 10−8 cm2

/sec [10−8 in2/sec] (moderate concrete quality), and 6.45 × 10−9 cm2 18 /sec 

[10−9 in2 19 /sec] (good concrete quality), respectively (Berntz et al., 2001). 

20 Although no cases of corrosion-induced damage have been reported, the results of the 

21 durability model presented by Berntz et al. (2001) show that corrosion of the reinforcing steel in 

22 concrete can potentially initiate and propagate within the 60-year timeframe for concretes of 

23 moderate to low quality. Thus, corrosion of reinforcing steel in concrete exposed to outdoor and 

24 groundwater or soil (below-grade) environments is considered to be credible, and therefore, 

25 aging management is required during the 60-year timeframe.

26 Concretes exposed to sheltered environments and fully encased (lined) in steel

27 Chloride ingress is expected to be insignificant for steel reinforcement embedded in concrete in 

28 sheltered environments with limited exposure to water. In addition, the heat load from the fuel in 

29 the DSS is expected to aid in drying the interior concrete surfaces, thus decreasing water 

30 availability at the concrete surface, which is necessary to promote this degradation mode. 

31 Chloride ingress will also be impeded in concrete fully encased (lined) in steel. Thus, corrosion 

32 of reinforcing steel is not considered credible for concrete in these environments, and therefore, 

33 aging management is not required during the 60-year timeframe. 

34 3.5.1.7 Shrinkage

35 Shrinkage occurs when hardened concrete dries from a saturated condition to a state of 

36 equilibrium in about 50 percent relative humidity (NRC, 2012). As excess concrete water 

37 evaporates, tensile stresses are induced in the concrete due to internal pressure from the 

38 capillary action of water movement, which results in cracking. The factors affecting shrinkage 

39 are cement content, water-to-cement ratio, degree of hydration, elastic modulus of aggregates, 

40 amount and characteristics of concrete admixtures, temperature and humidity during curing, and 

41 size and shape of concrete (NRC, 2014; Branson, 1977; Mindess and Young, 1981).

The maximum shrinkage is in the range of 400 × 10−6 to 780 × 10−6 cm/cm [400 × 10−6 42 to 

780 × 10−6 43 in/in] (NRC, 2014; Branson, 1977) and decreases exponentially with time 

44 (Branson, 1977). Shrinkage of concrete occurs initially during curing, which can be controlled 
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1 through concrete formulation and the density and distribution of internal reinforcement 

2 (ACI, 2005, 2007). According to ACI 209R-92 (ACI, 2008b), over 90 percent of the shrinkage 

3 occurs during the first year, reaching 98 percent by the end of the first 5 years. Thus, shrinkage 

4 is not expected to influence concrete performance after the initial storage or licensing period,

5 because most of the shrinkage will take place early on in the life of the concrete. As a result, 

6 shrinkage of concretes exposed to sheltered, outdoor, groundwater or soil (below grade), and 

7 fully encased environments is not considered to be credible, and therefore, aging management 

8 is not required during the 60-year timeframe.

9 3.5.1.8 Leaching of calcium hydroxide

10 Concretes exposed to outdoor, sheltered, and groundwater/soil (below-grade) environments

11 A constant or intermittent flux of water through a concrete surface can result in the removal or 

12 leaching of calcium hydroxide (Hanson et al., 2011). Calcium hydroxide leaching is observed in 

13 the form of white leachate deposits (calcium carbonate) on the concrete surface. Calcium 

14 hydroxide leaching causes loss of concrete strength, converting the cement into gels that have 

15 no strength. Leaching also increases the concrete porosity and permeability, making it more 

16 susceptible to other forms of aggressive attack. In addition, leaching of calcium hydroxide in 

17 concrete lowers the concrete pH, affecting the integrity of the protective oxide film of the 

18 reinforcing steel (EPRI, 2007).

19 The extent of the leaching depends on the environmental salt content and temperature 

20 (NRC, 2011a), and it can take place above and below ground. However, the leaching rate is 

21 generally slow and controlled by diffusion (Berner, 1992). For example, interior inspections 

22 conducted at the Calvert Cliffs ISFSI revealed the presence of white-colored stalactite debris in 

23 the gap between the heat shield and the concrete ceiling of two sheltered DSS concrete 

24 structures after 15–20 years in service. Stalactites are formed when water leaches calcium 

25 hydroxide out of the concrete, which precipitates as calcium carbonate on contact with carbon 

26 dioxide in the air. The licensee concluded that water entering the outlet vent stack promoted 

27 calcium hydroxide leaching (Gellrich, 2012). Other exterior inspections conducted at the TMI-2 

28 ISFSI revealed efflorescence growth on multiple DSS concrete structures exposed to an 

29 outdoor environment. The licensee concluded that the efflorescence deposits were formed by

30 water entering freeze and thaw cracks in the anchor blockout holes on the roof of the HSMs. 

31 The licensee conducted core sample testing to verify concrete compressive strength. 

32 Therefore, operating experience indicates that leaching of calcium hydroxide is a mechanism 

33 that can be exacerbated by other degradation mechanisms or designs that do not adequately 

34 prevent ingress of precipitation into the sheltered structure. As such, leaching of calcium 

35 hydroxide in concrete exposed to outdoor, sheltered, and groundwater or soil (below-grade)

36 environments is considered to be credible, and therefore, aging management is required during 

37 the 60-year timeframe.

38 Concretes fully encased (lined) in steel

39 Leaching of calcium hydroxide is not considered a credible mechanism for concrete fully

40 encased (lined) in steel because of the lack of water ingress, and therefore, aging management 

41 is not required during the 60-year timeframe. 
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1 3.5.1.9 Radiation damage

2 Radiation effects on concrete properties will depend on the gamma and neutron radiation 

3 doses, temperature, and exposure period. Gamma radiation can decompose and evaporate 

4 water in concrete (Bouniol and Aspart, 1998). Because most of the water is contained in the 

5 cement paste, the effect of gamma radiation on cement paste is more significant than on the 

6 aggregates. Gamma radiation can also decompose the SiO bond within calcium silicate hydrate 

7 (Kontani et al., 2010). Neutron radiation deteriorates concrete by reducing stiffness, forming 

8 cracks by swelling, and changing the microstructure of the aggregates. This consequently 

9 reduces concrete strength (Kontani et al., 2010). The changes in aggregate microstructure also 

10 can lead to higher reactivity of aggregates to certain aggressive chemicals. 

11 NUREG/CR-7171, “A Review of the Effects of Radiation on Microstructure and Properties of 

12 Concretes Used in Nuclear Power Plants,” provides a comprehensive review of the effects of 

13 gamma and neutron radiation on the microstructure and properties of concrete used in nuclear 

14 power plants (NRC, 2013). Concrete structures have been regarded as being sound as long as 

15 the cumulative radiation does not exceed critical levels over the life of the structure. In general, 

16 the critical radiation levels to reduce concrete strength and elastic modulus are considered to be 

approximately 1 × 1019 n/cm2 [6.5 × 1019 n/in2 17 ] for fast neutrons (neutron energy >1 MeV) and 

1-2 × 1010 rad [1-2 × 108 18 grays] for gamma rays (Hilsdorf et al., 1978; EPRI, 2012; IAEA, 1998; 

19 ASME, 2007).

20 As discussed in Section 3.2.1.9, the maximum potential accumulated neutron fluence on DSS 

basket components after 100 years was calculated to be 2.63 × 1016 n/cm2 [1.70 × 1017 n/in2 21 ], 

22 which is three orders of magnitude below the level that would lead to a reduction of concrete 

23 strength and elastic modulus. The gamma dose is also expected to be several orders of 

24 magnitude less than the limits defined in the above references, per the specific DSS design 

25 bases. Thus, radiation damage is not considered credible for concrete, and therefore, aging 

26 management is not required during the 60-year timeframe. 

27 3.5.1.10 Fatigue

28 Concrete fatigue strength is defined as the maximum stress that the concrete can sustain 

29 without failure under a given number of stress cycles (NRC, 2014). Because dry storage is a 

30 static application, mechanical cyclic loading is not expected. However, restraint of the concrete 

31 from expanding and contracting as it is exposed to rapid changes in temperature will lead to 

32 internal stresses in the structure. If the changes in temperature are severe and the resulting 

33 strains are sufficient, local plastic deformation can occur. Repeated application of this thermal 

34 loading can lead to crack initiation and propagation in low-cycle fatigue. 

35 Concrete fatigue in the DSS reinforced concrete may be caused by diurnal and seasonal 

36 temperature gradients through the wall of the DSS assembly. The inside surface of the 

37 concrete wall is hotter than the outside surface of the concrete wall, which causes compressive 

38 stresses in the DSS concrete near the inside of the concrete wall and tensile stresses in the 

39 rebar near the outside of the concrete wall.

40 Extreme seasonal temperature variations are expected to be significantly higher than diurnal 

41 variations; these would be capable of producing higher cyclic stress amplitudes. Assuming 

42 ambient temperatures of -40 degrees C [−40 degrees F] (winter) and 52 degrees C 

43 [125 degrees F] (summer), the maximum thermal gradient across the DSS concrete is expected 
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1 to be less than 16 degrees C [60 degrees F]. The number of extreme seasonal temperature 

2 cycles, conservatively postulated to occur 10 times per year, is 600 over 60 years.

3 Diurnal temperature fluctuations in ambient air temperatures are assumed to occur once per 

4 day. For conservatism, it is assumed that the diurnal temperature fluctuations are 25 degrees C 

5 (the largest mean daily change of temperature in the United States). Therefore, the total 

6 number of thermal cycles due to diurnal temperature variations in ambient temperatures over 

7 60 years is 21,900 thermal cycles. Thus, the total number of thermal cycles due to seasonal 

8 and daily variations over 60 years is 22,500 cycles. The thermally induced stress, σ, defined in 

9 Section 3.2.1.7, can be used to determine the stress in the concrete during each 

temperature cycle. Using a thermal expansion coefficient of 1.1 × 10-5 10 cm/cm/degrees C

[6.5 × 10-6 in/in/degrees F] and an elastic modulus of 2.764 × 10 11 4 megapascals (MPa)

[4.035 × 103 12 ksi], which are typical for concretes, the computed values of σ are 7.53 MPa 

13 [1.09 ksi] and 9.99 MPa [1.45 ksi] for the diurnal and seasonal temperature 

14 fluctuations, respectively.

15 The seasonal change in stress is assumed bounding for the cumulative number of cycles of 

16 both diurnal and seasonal temperature fluctuations. Assuming that these cyclic stresses are the 

17 only cyclic mechanical loading experienced by the DSS (an adequate assumption for a passive 

18 system), the ratio of the concrete compressive stress to its design strength is less than 0.29 

19 (i.e., 1.45 ksi/5 ksi). This calculated ratio at 22,500 cycles is lower than the lowest 

20 stress/cycles-to-failure (S-N) curve for concrete reported in ACI 215R (ACI, 1997). Thus, 

21 fatigue of concrete exposed to sheltered, outdoor, groundwater or soil (below-grade), and fully

22 encased environments is not considered to be credible, and therefore, aging management is not 

23 required during the 60-year timeframe.

24 Notwithstanding the conclusion above, the NRC reviewer must review any fatigue analyses for 

25 concrete structures contained in the applicant’s original design-bases documents to determine 

26 whether the renewal application adequately addresses the implications of extending the 

27 operating period to 60 years. This reexamination of the original fatigue analyses should be 

28 defined as TLAAs in the renewal application. The staff’s guidance for the review of TLAAs is 

29 provided in NUREG–1927, Revision 1, and is summarized in Chapter 5 of this report.

30 3.5.1.11 Dehydration at high temperature

31 Exposure of concrete to elevated temperatures can affect its mechanical and physical 

32 properties (Phan and Carino, 2000). It is well known that concretes can degrade at high 

33 temperatures due to dehydration of the hydrated cement paste, thermal incompatibility between 

34 the cement and aggregates, and physicochemical deterioration of the aggregates (NRC, 2006). 

35 As the temperature increases to about 105 degrees C [221 degrees F], all evaporable water is 

36 removed from the concrete. At temperatures above 105 degrees C [221 degrees F], the 

37 strongly absorbed and chemically combined water are gradually lost, with the dehydration 

38 essentially complete at 850 degrees C [1,562 degrees F] (Harmathy, 1970). High-temperature 

39 degradation in concrete manifests as a change in compressive strength and stiffness, as well as 

40 an increase in concrete shrinkage and transient creep, resulting in the formation of cracks 

41 (Naus, 1981, 1988; Schneider et al., 1981). The effect of the elevated temperature is 

42 most significant on the concrete’s modulus of elasticity, which can decrease up to 40 percent 

43 (Freskakis, 1979). Concretes in the temperature range of 20 to 200 degrees C [68 to 

44 392 degrees F] show small changes in compressive strength. Beyond 350 degrees C 

45 [662 degrees F], concrete compressive strength decreases rapidly (NRC, 2006).
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1 NUREG–1536, “Standard Review Plan for Spent Fuel Dry Storage Systems at a General 

2 License Facility” (NRC, 2010b), provides staff guidance for acceptable temperature limits during 

3 operation of DSS concrete structures. By design, general or local concrete temperatures should

4 be kept below 93 degrees C [200 degrees F] to avoid mechanical deterioration. For DSS 

5 concrete designs that satisfy additional acceptance criteria, the maximum temperature during 

6 operation can exceed 93 degrees C [200 degrees F] but should remain less than 149 degrees C 

7 [300 degrees F]. Therefore, the effects of thermal dehydration are addressed during the initial 

8 ISFSI licensing or DSS approval. Because the fuel temperature decreases over time, the 

9 design temperature considerations in NUREG–1536 are expected to continue to be adequate. 

10 Thus, dehydration of concrete at high temperature is not considered to be credible in sheltered, 

11 outdoor, groundwater or soil (below-grade), and fully encased (lined) environments, and 

12 therefore, aging management is not required during the 60-year timeframe.

13 3.5.1.12 Microbiological degradation

14 Concretes exposed to groundwater/soil (below-grade) environments

15 Biodeterioration is caused by colonization of microbes and microorganisms that grow on 

16 concrete surfaces that offer favorable environmental conditions (e.g., available moisture, near 

17 neutral pH, presence of nutrients). Conducive environments may have elevated relative 

18 humidity (i.e., greater than about 60 percent), long cycles of humidification and drying, freezing 

19 and thawing, high carbon dioxide concentrations, high concentrations of chloride ions or other 

20 salts, or high concentrations of sulfates and small amounts of acids (Wei et al., 2013). 

21 According to Sanchez-Silva and Rosowsky (2008), biodeterioration may lead to reduction of the 

22 protective cover depth and increase both concrete porosity and the transport of aggressive 

23 chemicals. In addition, this degradation mode can promote a reduction in concrete pH, loss of 

24 concrete strength, and spalling/scaling.

25 Evidence shows that a wide variety of organisms can cause concrete deterioration in polluted 

26 soils and groundwater. The biodeterioration of concrete typically is confined to the surface. The 

27 rate of deterioration is slow, but the degradation mode has been observed within 40 years of 

28 exposure (Hu et al., 2011). Recent observations in Texas, Alabama, Georgia, and Mississippi 

29 have identified several sites where microorganisms have caused deterioration of the columns of 

30 concrete bridges embedded in soil (Trejo et al., 2008). Giannantonio et al. (2009), 

31 Magniont et al. (2011), Vollertsen et al. (2008), and Ghafoori and Mathis (1997) provide a list of 

32 microorganisms that can promote degradation in concrete in soils and waters. According to 

33 Bastidas-Arteaga et al. (2008), biodeterioration of concrete is mainly caused by bacteria, fungi, 

34 algae and lichens, and mussels (Perez et al., 2003). Once the pH of the surface of the concrete 

35 drops below 9 in the presence of sufficient nutrients, moisture, and oxygen, some species of 

36 sulfur bacteria, such as Thiobacillus sp., can attach to the concrete surface and reproduce 

37 (Mori et al., 1992). As the pH continues to fall to moderate or weakly acidophilic conditions, 

38 T. novellus, T. neapolitanus, and T. intermedius establish on the surface of concrete 

39 (Milde et al., 1983). The type of bacteria is strongly dependent on the concrete pH and 

40 environmental conditions (Okabe et al., 2007). 

41 Although no cases of microbiological degradation of concrete have been reported in nuclear 

42 applications, the degradation mode is considered credible, as below-grade environments may 

43 be conducive to microbe and bacteria growth. Thus, microbiological degradation of concrete 

44 structures exposed to groundwater or soil (below-grade) environments is considered credible, 

45 and therefore, aging management is required during the 60-year timeframe.
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1 Concretes exposed to outdoor, sheltered, and fully encased (lined) environments

2 The outdoor and sheltered environments may provide favorable conditions for microbiological 

3 degradation mechanisms because of the potential presence of moisture. However, the 

4 conditions may be intermittent, and there is no evidence that actual concrete subcomponents in 

5 the DSS environment microbiologically degrade. In addition, fully encased concrete is 

6 considered to be largely protected from moisture intrusion. Thus, microbiological degradation of 

7 concretes exposed to outdoor, sheltered, and fully encased (lined) environments is not 

8 considered credible, and therefore, aging management is not required during the 60-year 

9 timeframe.

10 3.5.1.13 Delayed ettringite formation

At the initial stage of fresh concrete curing, ettringite,1 11 commonly referred to as “naturally 

12 occurring ettringite,” is formed by the reaction of tricalcium aluminate and gypsum in the 

13 presence of water. The formation of naturally occurring ettringite in fresh concrete is not 

14 detrimental to the overall concrete performance. At the still-early stage of concrete curing, the 

15 naturally occurring ettringite may convert to monosulfoaluminate if curing temperatures are 

16 greater than about 70 degrees C [158 degrees F] (Fu, 1996). After concrete hardens, if the 

17 temperature decreases below this value, the monosulfoaluminate becomes unstable and, in the 

18 presence of sulfates released by the C-S-H gel, ettringite will reform. This mechanism is called 

19 “delayed ettringite formation” (DEF), which results in volume expansion and increased internal 

20 pressures in the concrete (Fu, 1996). Because the concrete has hardened at this stage, the 

21 volume expansion leads to cracking and spalling, with greatest severity commonly observed in 

22 below-ground structures with elevated temperatures from curing and heat of hydration 

23 (Shayan and Quick, 1992; Hobbs, 1999). DEF has been reported in precast concrete railroad 

24 ties in Sweden (Sahu and Thaulow, 2004), cast-in-place concrete structures in the southern 

25 United States after 10 years in service (Thomas et al., 2008), and mass concretes with high 

26 cement contents in the United Kingdom (Hobbs, 1999; Johansen and Thaulow, 1999). 

27 However, to date, no operating experiences exist of DEF degradation for concrete structures at 

28 nuclear power plants.

29 The conditions necessary for the occurrence of DEF are excessive temperatures during 

30 concrete placement and curing, the presence of internal sulfates, and a moist environment. 

31 ACI 318-05 (ACI, 2005) indicates that inspection reports shall document concrete temperature 

32 and protection during placement when the ambient temperature is above 35 degrees C 

33 [95 degrees F]. Protection measures during concrete placement include lowering the 

34 temperature of the batch water, cement, and aggregates as referenced in ACI 305R-10 

35 (ACI, 2010). As such, following the ACI 318-05, ACI 305R-10, and ACI 308R-01 (ACI, 2008d) 

36 guidelines during concrete placement and curing can effectively limit the concrete temperature 

37 to below 70 degrees C [158 degrees F], therefore preventing the development of DEF. 

38 NUREG–1536 (NRC, 2010b) cites ACI 349 (ACI, 2007) and ACI 318 as applicable codes for the 

39 design and construction of concrete structures of the DSSs. In addition to the adequate 

40 placement and curing standards, no occurrences of DEF-related degradation of concrete have 

41 been reported in nuclear applications. Thus, DEF of concrete is not considered credible in 

 

1Ettringite (3CaO·Al2O3·3CaSO4·32H2O) is the product of the reaction of gypsum and other sulfate compounds with 

calcium aluminate in the cement within the first few hours after mixing with water.
3-74

1 outdoor, sheltered, groundwater or soil (below-grade), and fully encased (lined) environments, 

2 and therefore, aging management is not required during the 60-year timeframe.

3 3.5.1.14 Salt scaling

4 Concretes exposed to outdoor environments and grounwater/soil (below-grade) environments 

5 above the freeze line

6 Salt scaling is defined as superficial damage caused by freezing a saline solution on the surface 

7 of a concrete body. The damage is progressive and consists of the removal of small chips or 

8 flakes of material. Similar to freeze and thaw damage, salt scaling takes place when concrete is 

9 exposed to freezing temperatures, moisture, and dissolved salts. The degradation is maximized 

10 at a moderate concentration of salt (e.g., from deicing salts), called the pessimum concentration 

11 (Marchand et al., 1999). Verbeck and Klieger (1957) reported that the pessimum concentration 

12 is independent of the types of salt species and is about 3 to 4 percent of the solute by weight. 

13 The most common deicing salts are sodium chloride and calcium chloride. Other deicing 

14 chemicals include magnesium chloride, urea, potassium chloride, ammonium sulfate, and 

15 ammonium nitrate.

16 Salt scaling of concrete roadways, pavements, sidewalks, driveways, decks, and other slabs is 

17 a common problem in locations exposed to cyclic freezing and thawing and deicing salts. For 

18 vertical surfaces, this damage mechanism is not expected to be operative unless the DSS 

19 concrete structure is surrounded by standing water containing salts. Therefore, this degradation 

20 mode is only expected to initiate and manifest in horizontal structures exposed to outdoor 

21 environments where water ponding can occur. Because salt scaling is closely related to freeze 

22 and thaw damage, the timeframe associated with the initiation of salt scaling of concrete could 

23 be relevant for both short- and long-term exposures. Thus, salt scaling damage is considered 

24 credible within the 60-year timeframe for DSS concrete structures exposed to outdoor and 

25 groundwater or soil (below-grade) environments above the freeze line, and therefore, aging 

26 management is required during the 60-year timeframe.

27 Concretes exposed to sheltered environments, fully encased (lined) in steel, and exposed to 

28 groundwater/soil (below-grade) environments under the freeze line

29 Concretes exposed to sheltered environments with low water availability or below-grade 

30 concrete maintained above freezing temperatures are not susceptible to salt scaling 

31 degradation. The heat load from the emplaced fuel in DSSs is expected to aid in drying the 

32 internal concrete surface, preventing the development of salt scaling inside the DSSs’ concrete 

33 structure. Salt scaling damage is also expected to be insignificant for concretes fully encased 

34 by liners (e.g., metallic compartments)—even under freezing conditions—due to the lack of 

35 water and salt transfer between the concrete and the outside environment. Thus, interior DSS 

36 concrete surfaces, below-grade concretes maintained under the freeze line, and fully encased 

37 (lined) concrete not in direct contact with outdoor environments are not expected to undergo salt 

38 scaling damage within the 60-year timeframe, and therefore, aging management is not required.

39 3.5.2 Ceramic fiber insulation 

40 The HI-STORM 100U underground system uses a divider shell to separate the intake cooling air 

41 from the heated air that streams up around the canister. This shell is insulated to minimize the 

42 preheating of the intake cooling air, with Kaowool® ceramic fiber insulation being a preferred 

43 insulation material in this DSS design.
3-75

1 3.5.2.1 Radiation damage

2 Neutron radiation has been shown to affect the strength and thermal diffusivity of ceramic fiber 

3 insulation. The effects will generally depend on the radiation dose, moisture content, 

4 temperature, and exposure period. 

5 Snead et al. (1992) provide an example of the effects of neutron irradiation on ceramic-fiber 

6 interfacial strength. Results comparing unirradiated and 1-dpa neutron-irradiated ceramic fiber 

7 insulation samples (SiC/C/Nicalon) exhibited a marked decrease in both interfacial shear 

8 strength and frictional resistance to sliding. The decrease in interfacial shear strength resulted 

9 in the decrease of the ultimate strength of the ceramic fibers by about 25 percent. Similarly, the 

10 decrease in frictional resistance resulted in increased fiber toughness. The changes in the 

11 mechanical properties were attributed to the fiber shrinkage that causes a partial debonding of 

12 the fiber and matrix interface.

13 Other research provides examples of the effects of neutron irradiation on the thermal diffusivity 

14 of ceramic fiber insulation (Akiyoshi and Yano, 2008; Snead et al., 2000; Akiyoshi, 2009; 

15 Akiyoshi et al., 2006; Yano et al., 2000; and Snead et al., 2005). For example, Akiyoshi and 

16 Yano (2008) showed a degradation of thermal diffusivity in neutron-irradiated specimens by 

17 studying the macroscopic property changes in as-irradiated and annealed specimens under 

18 different temperatures from 373 to 766 degrees C [703 to 1,411 degrees F] and different 

neutron doses from 0.4 to 8.0 × 1022 n/cm2 [2.6 to 51.6 × 1022 n/in2 19 ]. The thermal diffusivity of 

20 as-irradiated specimens showed dependence on the neutron-irradiation dose and the irradiation 

21 temperature. Snead et al. (2000) have also demonstrated that the thermal conductivity of most 

22 ceramic fiber insulation materials undergoes a rapid reduction with irradiation when subjected to 

a fast-neutron fluence up to about 3.4 × 1021 n/cm2 [ 2.2 × 1022 n/in2 23 ] and irradiation temperature 

24 of about 

25 200–700 degrees C [392–1,292 degrees F]. Gamma irradiation also results in a permanent 

26 decrease in the volume and surface resistivity of ceramic fibers at gamma values of around 

1 × 109 rads [1 × 107 27 grays] (Davies, 1966). In general, the reduction of thermal diffusivity of 

28 ceramic fiber insulation should result in improved thermal insulation performance.

29 While the reduction of strength of ceramic fiber insulation due to radiation is not expected to 

30 compromise the SSC’s intended function, a review of the radiation effects should be performed 

31 on a case-by-case basis.

32 The NRC reviewer should review the analyses contained in the applicant’s original 

33 design-bases documents to determine whether the renewal application adequately addresses 

34 radiation damage of ceramic fiber insulation for an extended operating period of 60 years. This 

35 reexamination of the original analyses would typically be defined as TLAAs in the renewal 

36 application. The staff’s guidance for the review of TLAAs is provided in NUREG–1927, 

37 Revision 1. If the original design basis does not include an analysis for an SSC that could 

38 reasonably be expected to be subject to radiation damage in the 60-year timeframe, the 

39 reviewer nevertheless should ensure that the application addresses this potential aging effect.

40 An applicant may conclude that an analysis cannot support a determination that fatigue will not 

41 challenge an important-to-safety function in the 60-year timeframe of the period of extended 

42 operation. In that case, the applicant may manage the aging of the associated SSC with 

43 an AMP.
3-76

1 3.5.2.2 Moisture absorption

2 Ceramic fiber insulation materials are generally porous (either open- or closed-pore network) 

3 and filled with atmospheric air in the dry condition. In nonencased SSCs, moisture transport 

4 through the insulation can be realized by diffusion and/or capillary suction. Vafai and Sarkar 

5 (1986) first modeled the transient heat and moisture transfer with condensation. The effect of 

6 condensates on the effective thermal conductivity and radiative heat transfer have also been 

7 considered in a transient model in porous media (Fan et al., 2000). This model suggests that 

8 the initial water content, service temperature, and insulation thickness are key factors 

9 influencing the insulation performance. Other parameters, such as the water vapor resistance, 

10 the thermal conductivity, and the insulation porosity were found to have smaller effects. The 

11 presence of moisture can significantly increase the insulation thermal conductivity 

12 (Cai et al., 2012).

13 The ceramic fiber insulation is foil faced or jacketed and therefore encased and protected from 

14 moisture. The high zinc content in the coating of the adjacent divider shell in the 

15 HI-STORM 100U system provides protection for the foil/jacket from galvanic corrosion. In 

16 addition, SCC of the foil/jacket is not a credible aging mechanism due to low stresses derived 

17 from the dead weight of the foil or jacket. Therefore, the integrity of the foil or jacket is not 

18 expected to be compromised, which will prevent moisture entering the ceramic fiber insulation. 

19 As such, moisture absorption of ceramic fiber insulation is not considered to be credible, and 

20 therefore, aging management is not required during the 60-year timeframe.

21 3.5.3 References

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31 _____. ACI 308R-01, “Guide to Curing Concrete.” Farmington Hills, Michigan: 

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33 _____. ACI 349-06, “Evaluation of Existing Nuclear Safety-Related Concrete Structures.” 

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35 _____. ACI 318-05, “Building Code Requirements for Structural Concrete and Commentary.” 

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37 _____. ACI 221.1R-98, “State-of-the-Art Report on Alkali-Aggregate Reactivity.” 

38 Farmington Hills, Michigan: American Concrete Institute. 1998.
3-77

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33 Concrete Structures.” BE95-1347/R17. CUR, Gouda, The Netherlands. The European 

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21 Civil Engineering. Vol. 9. pp. 35–40. 1997.

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27 Glauz, D.L., D. Roberts, V. Jain, H. Moussavi, R. Llewellen, and B. Lenz. “Evaluate the Use of 

28 Mineral Admixtures in Concrete to Mitigate Alkali-Silica Reactivity.” Report 

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4 Hobbs, D.W. “Expansion and Cracking in Concrete Associated with Delayed Ettringite 

5 Formation.” Ettringite, the Sometimes Host of Destruction. B. Erlin, ed. SP177 

6 Farmington Hills, Michigan: American Concrete Institute International. pp. 159–181. 1999.

7 Hu, J., D. Hahn, W. Rudzinski, Z. Wang, and L. Estrada. “Evaluation, Presentation and Repair 

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9 Texas Department of Transportation Research and Technology Implementation Office. 2011.

10 IAEA. “Assessment and Management of Ageing of Major Nuclear Power Plant Components 

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12 1998.

13 Johansen, V. and N. Thaulow. “Heat Curing and Late Formation of Ettringite.” ACI SP-177. 

14 Bernard Erlin, ed. Farmington Hills, Michigan: American Concrete Institute. pp. 199–206. 

15 1999.

16 Kontani, O., Y. Ichikawa, A. Ishizawa, M. Takizawa, and O. Sato. “Irradiation Effects on 

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1 3.6 Spent fuel assemblies

2 The SNF assembly components evaluated in this section include the zirconium-based cladding 

3 and fuel assembly hardware that provide structural support to ensure that the spent fuel is 

4 maintained in a known geometric configuration. The safety analyses for the ISFSI or DSS rely 

5 on the fuel assembly having a specific configuration (e.g., geometric form, a certain number of 

6 fuel rods or solid replacement filler rods in the assembly lattice). Although the spent fuel 

7 assembly is not an SSC of the ISFSI or DSS, depending on the particular design bases, the 

8 spent fuel must remain in its analyzed configuration during the period of extended operation for 

9 continuation of the approved design bases. Therefore, for these ISFSIs and DSSs, the 

10 condition of the SNF assembly and cladding are within the scope of renewal and are reviewed 

11 for aging mechanisms and effects that may lead to a change in the analyzed fuel configuration.

12 The experimental confirmatory basis that low-burnup fuel (≤45 gigawatt days per metric ton of 

13 uranium (GWd/MTU)) will remain in its analyzed configuration during the period of extended 

14 operation was provided in NUREG/CR-6745, “Dry Cask Storage Characterization Project—

15 Phase 1; CASTOR V/21 Cask Opening and Examination” (Bare and Torgerson, 2001), and 

16 NUREG/CR-6831, “Examination of Spent PWR Fuel Rods after 15 Years in Dry Storage” 

17 (Einziger et al., 2003). This research demonstrated that low-burnup fuel cladding and other 

18 cask internals had no deleterious effects after 15 years of storage and confirmed the basis for 

19 the guidance on creep deformation and radial hydride reorientation in Interim Staff Guidance 

20 (ISG)-11, “Cladding Considerations for the Transportation and Storage of Spent Fuel, 

21 Revision 3” (NRC, 2003). In ISG-11, Revision 3, the NRC staff indicated that the spent fuel 

22 configuration is expected to be maintained as analyzed in the safety analyses for the ISFSI or 

23 DSS, provided certain acceptance criteria (regarding maximum fuel clad temperature and 

24 thermal cycling) are met and the fuel is stored in a dry inert atmosphere. The research results 

25 in NUREG/CR-6745 and NUREG/CR-6831 support a determination that degradation of 

26 low-burnup fuel cladding and assembly hardware should not result in changes to the approved 

27 design bases during the first period of extended operation, provided that the cask/canister 

28 internal environment is maintained. The U.S. Department of Energy (DOE) is expected to

29 gather similar experimental confirmatory data to support the technical basis for storage of 

30 high-burnup (HBU) fuel during the first period of extended operation (EPRI, 2014).

31 The staff reviewed gap assessments for DSS, relevant technical literature, and operating 

32 experience from nuclear applications (NRC, 2014a; Chopra et al., 2014; Hanson et al., 2012; 

33 Sindelar et al., 2011; NWTRB, 2010) to identify potential degradation mechanisms in 

34 consideration of the materials and condition of the SNF at loading and the environment in dry 

35 storage. The SNF cladding materials are zirconium-based alloys. The primary components of 

36 the fuel assembly hardware are spacer grids, end fittings, guide tubes (PWR only), and 

37 assembly channels (BWR only). The materials of construction for these components include 

38 zirconium-based alloys, nickel alloys, and stainless steel. The staff’s assessement of the 

39 condition of the SNF assembly at loading considered changes to the fuel pellets and the 

40 zirconium-based cladding during reactor service, including hydrogen absorption by the cladding, 

41 swelling of the fuel pellets, increased rod pressurization due to helium and fission gas release, 

42 and pellet-cladding interactions. The storage environment is helium or an alternative cover gas 

43 in high radiation and temperature. A minimal amount of water (about 0.43 gram mole) is also 

44 considered to be retained inside the cask/canister (NRC, 2010). This moisture content is based 

45 on a design-basis drying process that evacuates the cask/canister to less than or equal to 3 torr 

46 [0.06 psi] and maintains a constant pressure for 30 minutes before closure.
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1 The aging mechanisms considered for high burnup zirconium-based cladding (i.e., average 

2 assembly burnups exceeding 45 GWd/MTU) include hydride reorientation, delayed hydride 

3 cracking, thermal and athermal (low-temperature) creep and localized mechanical overload. 

4 These mechanisms are primarily driven by cladding hoop stresses, which are lower in low

5 burnup fuel. The technical bases for these mechanisms (Sections 3.6.1.1–3.6.1.5) considered 

6 cladding hoop stresses for high burnup fuel, therefore these discussions are considered 

7 bounding to low burnup fuel. In addition, the demonstration program discussed in 

8 NUREG/CR-6745 (Bare and Torgerson, 2001) and NUREG/CR-6831 (Einziger et al., 2003)

9 provided confirmation that hydride reorientation and creep will not compromise the configuration

10 of low burnup fuel during the renewal period. Other aging mechanisms considered for both low 

11 and high burnup zirconium-based cladding include radiation embrittlement, fatigue, oxidation, 

12 pitting corrosion, galvanic corrosion, and SCC and MIC. Of these potential mechanisms, MIC 

13 was not considered to be applicable, as the aging mechanism is not expected to be operable 

14 under the inert atmosphere of dry storage. Detailed discussions regarding each of the potential

15 aging mechanisms for zirconium-based cladding are provided in Section 3.6.1.

16 The degradation mechanisms considered for the assembly hardware include creep, fatigue, 

17 hydriding, general corrosion, SCC, and radiation embrittlement. Detailed discussions regarding

18 each of these applicable aging mechanisms for assembly hardware are provided in 

19 Section 3.6.2.

20 3.6.1 Cladding materials

21 3.6.1.1 Hydride reorientation (high burnup fuel)

22 In reactor service, the zirconium-based fuel cladding absorbs hydrogen, which leads to the 

23 precipitation of hydride platelets as the dissolved hydrogen exceeds the solubility limit of the 

24 cladding. The primary source of the hydrogen is water-side corrosion (oxidation) of the cladding 

25 (Hanson et al., 2012; IAEA, 1993). The total concentration of hydrogen absorbed by the 

26 cladding (i.e., dissolved in the zirconium matrix and in precipitated hydrides) increases with 

27 burnup and varies axially across the fuel rods. For burnups above 45 GWd/MTU and up to 

28 62 GWd/MTU (the current NRC licensing limit), the total hydrogen content for Zircaloy-2 is 

29 expected to be in the range of 260–300 weight parts per million [wppm] (NRC, 2015a; 

30 Geelhood and Luscher, 2014), 200–1,200 wppm for Zircaloy-4 (Mardon et al., 2010; 

31 Thomazet et al., 2005; King et al., 2002; Bossis et al., 2007; Hanson, 2016), ≤ 100 wppm for 

M5® 32 (King et al., 2002; Bossis et al., 2007; Mardon et al., 2010; Thomazet et al., 2005, 

33 Billone, 2013, Hanson, 2016), and up to 550± 300 wppm for ZIRLOTM (Billone et al., 2013, 

34 Billone et al., 2015). When discharged from the reactor and during wet storage, the hydride 

35 platelets are mostly oriented in the circumferential-axial direction, with a smaller fraction 

36 oriented in the radial-axial direction.

37 Once the SNF assemblies are removed from wet storage and loaded into a DSS, the 

38 cask/canister cavity is vacuum dried and backfilled with an inert gas. During vacuum drying, the 

39 temperature of the SNF assemblies and the temperature-dependent solubility limit of hydrogen 

40 in the cladding will also increase. As a result, some of the hydrides present in the cladding will 

41 redissolve as hydrogen. The amount of dissolved hydrogen will depend on the peak cladding 

42 temperature during the vacuum drying operations, which, per ISG-11, Revision 3 (NRC, 2003),

43 is not to exceed 400 degrees C [752 degrees F] for HBU fuel. For example, the maximum 

44 dissolved hydrogen at 400 degrees C [752 degrees F] is approximately 200 wppm based on 

45 representative solubility correlations (Kammenzind et al., 1996; Kearns et al., 1967). Once the 

46 loaded cask/canister is dried and backfilled, the cladding temperature will decrease over time, 
3-87

1 and upon a sufficient temperature drop (~65 degrees C [117 degrees F]), some of the hydrogen 

2 in solution will reprecipitate as new hydrides. During this process, the orientation of these 

3 precipitated hydrides may change from the circumferential-axial to the radial-axial direction. 

4 The degree of reorientation is primarily driven by the metallurgical microstructure of the cladding 

5 alloy and the cladding hoop stresses during drying operations and subsequent cooling, which 

6 are determined by the rod internal pressure at a given gas temperature.

7 Cladding with a high concentration of radial hydrides (determined by the DSS drying conditions) 

8 has been shown to have reduced ductility under pinch-load stresses at sufficiently low 

9 temperatures, thereby affecting the ability to retrieve the HBU fuel (Billone at al., 2013; Aomi et 

10 al., 2008). The degradation of the mechanical properties at a particular temperature (described 

11 as the “ductile-to-brittle transition temperature” or DBTT) depends on the interconnectivity and 

12 number density of radial hydrides (as determined by their length, distribution, and orientation), 

13 and the thickness of the outer-surface hydride rim. This phenomenon has led the staff to 

14 express concern about potential cladding failures when subjected to pinch-load stresses higher 

15 than the fuel’s mechanical limit, if the cladding temperature decreases below the corresponding 

16 DBTT (NRC, 2015b). Therefore, as the cladding cools down during the 60-year timeframe, the 

17 extent of radial hydride reorientation and the DBTT are important for evaluating the cladding 

18 performance and ensuring that the HBU fuel remains in the analyzed configuration.

19 The primary driving force for radial hydride reorientation is the cladding hoop stresses, which 

20 are determined by the peak cladding temperature during drying operations. A review indicates 

21 that there is no consensus in the literature on minimum level or threshold hoop stresses needed 

22 to reorient hydrides for a given cladding alloy and temperature, as discussed in the following 

23 references:

24 • Zircaloy-4: Data from Chung (2004), Daum et al. (2006), and Chu et al. (2008) suggest 

25 that the threshold hoop stress for hydride reorientation in Zircaloy-4 is about 90 MPa 

26 [13 ksi] for peak temperatures at or near 400 degrees C [752 degrees F] for both 

27 irradiated and unirradiated rods. Other data obtained from irradiated cladding (Einziger 

28 and Kohli, 1984; Cappelaere, et al., 2001; and Goll, et al., 2001) suggest that hoop 

29 stresses greater than 120 MPa [17 ksi] may be required. Most recently, Kim et al 

30 (2015a) showed threshold stresses for hydride reorientation in unirradiated Zircaloy-4 of

31 60 ± 5 MPa 

32 [8.7 ± 0.7 ksi] at 400 degrees C [752 degrees F], 68 ± 5 MPa [9.8 ± 0.7 ksi] at 

33 335 degrees C [635 degrees F], 75 ± 6 MPa [10.9 ± 0.9 ksi] at 300 degrees C 

34 [572 degrees F], and 90 ± 6 MPa [13.0 ± 0.9 ksi] at 235 degrees C [455 degrees F]. 

35 Kamimura (2010) also reported a threshold stress for Zircaloy-4 of about 100 MPa 

36 [16 ksi] at 275 degrees C [527 degrees F] for a nominal burnup of 48 GWd/MTU. 

37 • Zircaloy-2: Kamimura (2010) reported a threshold hoop stress of 70 MPa [10 ksi] for 

38 Zircaloy-2 (no zirconium liner) of nominal burnup of 40 GWd/MTU at 200 degrees C 

39 [392 degrees F], and 70 MPa [10 ksi] for Zircaloy-2 (with zirconium liner) of nominal 

40 50 GWd/MTU and 55 GWd/MTU burnups at 300 degrees C [572 degrees F].

41 • Advanced alloys: Kamimura (2010) reported a threshold stress of 90 MPa [13 ksi] for 

42 ZIRLOTM at 250 degrees C [482 degrees F] for a nominal burnup of 55 GWd/MTU. 

43 Billone et al. (2013) reported reorientation of M5® cladding at their lowest studied hoop 

44 stress of 90 MPa [16 ksi] for a peak cladding temperature of 400 degrees C

45 [752 degrees F] and nominal burnup of 68 GWd/MTU.
3-88

1 These threshold hoop stresses for hydride reorientation were compared to estimated hoop 

2 stresses for representative BWR and PWR fuel assemblies. Raynaud and Einziger (2015) 

3 estimated the hoop stresses for 10 × 10 BWR and 17 × 17 PWR fuel assemblies as a function 

4 of decay gas release and fuel pellet swelling, which accounted for decay gas released to the 

5 pellet-clad gap. The maximum calculated hoop stress during drying operations for the BWR 

6 cladding was approximately 40 MPa [5.8 ksi] at a peak cladding temperature close to 

7 400 degrees C [752 degrees F]. Similarly, the maximum calculated hoop stress during drying 

8 operations for PWR cladding was approximately 100 MPa [14.5 ksi] at 400 degrees C 

9 [752 degrees F], which rapidly decays and falls well below 50 MPa after a few decades in dry 

10 storage. These calculations did not account for ZIRLOTM-clad integral fuel burnable absorber 

11 (IFBA) rods with hollow and solid blanket pellets; however, these rods are expected to 

12 experience higher maximum hoop stresses (Bratton et. al, 2015). Since the calculated hoop 

13 stresses exceed the experimental values in the literature for when radial hydride reorientation 

14 was observed, the staff considers that the radial hydride precipitation is credible in both in BWR 

15 and PWR fuel claddings in dry storage. 

16 The cladding alloy and corresponding fabrication process are also important factors for defining

17 the extent of hydride reorientation. Two predominant cladding microstructures are produced 

18 during fabrication: (1) recrystallized annealed (RXA) and (2) cold worked stress relieved 

(CWSR) annealed. Zircaloy-4 (PWR) and ZIRLOTM 19 (PWR) are generally CWSR, whereas 

Zircaloy-2 and M5® 20 are RXA. Because hydrides tend to precipitate in the grain boundaries, 

21 RXA claddings are more susceptible to hydride reorientation, since these cladding types have a 

22 larger fraction of grain boundaries in the radial direction (equiaxed grains) relative to CWSR 

23 claddings (which have more elongated grains).

24 The staff also considered the effect of the cladding cooling rate on the degree of hydride 

25 reorientation. The cooling rate post-drying and under dry storage is expected to be in the range 

of 10−3 to 10−5 degrees C/hr [1.8 × 10−3 to 1.8 × 10−5 26 degrees F/hr]. Most of the experimental 

27 studies reported in the literature have used cooling rates in the range of 0.6–30 degrees C/hr 

28 [1.08–54 degrees F/hr] (Aomi et al., 2008). However, an analysis of ductility data collected at 

29 different cooling rates in Aomi et al. does not show a clear trend. Chan (1996) also developed a 

30 micromechanical model to determine the effect of slow cooling rates on hydride reorientation

31 and morphology, including volume fraction of both radial and circumferential hydrides and 

32 continuity of the hydride network. Using experimental data to validate the model, Chan 

33 concluded that the cooling rate exerts no direct influence on radial hydride precipitation; instead, 

34 hydride orientation is dictated by the cladding stresses during hydride precipitation, regardless 

35 of the cooling rate. Therefore, the staff concludes that the slow cooling rates experienced post36 drying and during dry storage are not expected to inhibit the precipitation of radial hydrides.

37 Available DBTT data on HBU fuel cladding samples with radial hydrides have been obtained 

38 under conservative conditions and acceptance criteria (e.g. testing was performed on defueled 

39 samples, which do not account for the composite pellet-clad mechanical behavior) (Fuketa et 

40 al., 2003; Billone et al., 2013; Aomi et al., 2008). For example, Billone et al. showed that 

Zircaloy-4, ZIRLOTM, and M5® 41 cladding samples subjected to a radial hydride reorientation 

42 treatment exhibited lower ductility under pinch-load stresses at low relative temperatures 

43 (less than 200 degrees C [392 degrees F]). The radial hydride treatment was designed to 

44 simulate drying and storage conditions (i.e., peak cladding temperature of 400 degrees C 

45 [752 degrees F] and peak hoop stresses of ~110 MPa [16.0 ksi] and ~140 MPa [20.3 ksi]). 

46 General conclusions from Billone et al. were that: (1) the DBTT generally increases with 

47 increasing hoop stresses (i.e., the degradation of cladding mechanical properties shifts to higher 

48 cladding temperature), (2) both the susceptibility to radial hydride precipitation and degradation 
3-89

1 of mechanical properties depend on cladding type and initial hydrogen content, and 

2 (3) depending on the cladding and test conditions, the DBTT can occur at temperatures in the 

3 range of approximately 20 degrees C to 185 degrees C [68 to 328 degrees F]. The results for 

4 as-irradiated Zircaloy-4 are consistent with studies by Wisner and Adamson (1998) and Bai et 

5 al. (1994).

6 Considering the hydrogen content, peak drying temperatures, and corresponding hoop stresses, 

7 the staff concludes that hydride reorientation in zirconium-based HBU cladding is credible

8 during the 60-year timeframe. Further, depending on the specific fuel contents, it is possible for 

9 some of the cladding to reach temperatures near or below the DBTTs reported in the literature. 

10 Therefore, degradation of mechanical properties during pinch-type stresses due to hydride

11 reorientation is considered a credible aging mechanism for HBU fuel claddings.

12 The degradation of mechanical properties due to hydride reorientation is only expected to 

13 potentially compromise the ability to maintain the analyzed fuel configuration during pinch-type 

14 loads. These loads are only expected during fuel retrieval operations, if the design bases of the 

15 DSS or ISFSI rely on retrievabilty of the HBU fuel on a single-assembly basis. Pinch-type loads 

16 are not expected to be present during normal, off-normal, and accident conditions of storage. 

17 More specifically, the tensile stress field associated with potential inertial rod bending during 

18 storage is expected to be parallel to both radial and circumferential hydrides and not expected 

19 to compromise the structural integrity of the cladding. The NRC is sponsoring confirmatory 

20 research to this effect at Oak Ridge National Laboratory, and the results will be 

21 publicly-available soon (see Wang and Wang (2015) for details on the experimental protocol).

22 Until the Oak Ridge data is publicly-available, the staff has proposed two alternatives for 

23 demonstrating that the safety analyses pertaining to the analyzed spent fuel configuration will 

24 not be compromised by the effects of hydride reorientation. The first approach relies on the 

25 applicant performing a defense-in-depth analysis, assuming credible reconfiguration based on 

26 1 percent fuel failure for normal conditions of storage, 10 percent failure for off-normal 

27 conditions of storage, and 100 percent or other justifiable value for accident conditions. The 

28 staff has issued a generic consequence analysis for both vertical and horizontal storage 

29 configurations in NUREG/CR-7203 (Scaglione et al., 2015), which can be used by applicants in 

30 the development of their defense-in-depth analysis. A second approach relies on the evaluation 

31 of data from a demonstration (surrogate) program consistent with the guidance in Appendix D of 

32 NUREG–1927, Revision 1 (NRC, 2016). For example, destructive examination from the 

33 DOE/EPRI cask demonstration project (EPRI, 2014) may be used as confirmation that hydride 

34 reorientation has not compromised the ability to retrieve the spent fuel on a single-assembly 

35 basis. An example AMP consistent with the guidance in Appendix D of NUREG–1927, 

36 Revision 1 is provided in Chapter 5.

37 3.6.1.2 Delayed hydride cracking (high burnup fuel)

38 Delayed hydride cracking (DHC) is a time-dependent mechanism traditionally thought to occur 

39 by the diffusion of hydrogen to an incipient crack tip (notch, flaw) in the cladding, followed by 

40 nucleation, growth, and subsequent fracture of the precipitated hydrides at the crack tip 

41 (Hanson et al., 2012). Hydrogen dissolved in the cladding (see Section 3.6.1.1) can diffuse up a 

42 stress gradient in the crystalline lattice, or into the stress field at the core of an edge dislocation 

43 (Cox, 1997). The concentration gradient established by the stress gradient may lead to 

44 hydrogen supersaturation (i.e., solubility limit being exceeded) leading to the precipitation of 

45 hydrides at the crack tip. The precipitated hydride will continue to grow by the dissolution of 

46 hydrides in the low-stress regions of the material and by the continued diffusion of hydrogen up 
3-90

1 the stress gradient. Once the hydride reaches a critical size, it will crack and propagate to the 

2 end of the hydride, where it will blunt. The cycle could then repeat, until the crack propagates 

3 through the thickness of the material. DHC of spent fuel cladding has been studied under 

4 thermal transients representative of reactor operation (Kubo, 2012; Kim, 2009b) and 

5 representative of dry storage (Sasahara and Matsumura, 2008; EPRI, 2002). 

6 Requisite conditions for DHC are the presence of: (i) hydrides, (ii) existing crack tips 

7 (notch, flaws) that act as initiating sites, and (iii) sufficient cladding hoop stresses. Regarding 

8 requisite hydrides, a threshold for crack initiation cannot be readily defined. Simpson and Ells 

9 (1974) observed DHC with hydrogen concentration as little as 10 ppm in Zr-2.5 percent Nb 

10 cladding, although testing was performed at room temperature (i.e., a much lower temperature 

11 than those expected during the renewal period). Similarly, Coleman et al. (2009) were able to 

12 induce DHC in Zircaloy-4 at 200 wppm of hydrogen. Regarding requisite existing (incipient) 

13 crack tips, EPRI (2002) estimated the maximum initial depth of existing crack tips to be 140 μm 

14 [5.5 mils] or approximately 28 percent of the remaining wall of a typical 17 × 17 PWR cladding 

15 with 600 μm [23.6 mils] of original cladding thickness, and 100 μm [4 mils] of oxidation during its 

16 exposure in the reactor. Conversely, Raynaud and Einziger (2015) estimated the maximum 

17 initial depth of existing crack tips to be 120 μm [4.7 mils] for a cladding oxide thickness of 

18 100 μm [4 mils]. Regarding requisite hoop stresses for crack initiation, the mechanism requires 

19 that the stress intensity factor at the crack tip exceed a threshold value, denoted as KIH.

20 Most DHC studies have been performed under thermal transients representative of reactor 

21 operation, primarily on CANDU pressure tubes (Zr–2.5 percent Nb) and Zircaloy-2 cladding. 

22 Chan (2013) conducted an extensive literature review of experimentally determined KIH values 

23 for DHC crack initiation. In that review, KIH values for Zircaloy-2 are in the range of 5–

24 14 MPa√m [4.55–12.74 ksi√in] at 25–300 degrees C [77–572 degrees F], and in the range of 5–

25 10 MPa√m [4.55–9.10 ksi√in] for Zr-2.5 percent Nb cladding at 75–300 degrees C [167–572 

26 degrees F] (Chan, 2013, Figures 2 and 3). Kubo et al. (2012) also compiled KIH values for 

27 Zircaloy-2 in the range of 3–13 MPa√m [2.73–11.8 ksi√in]. Kim (2009a) also measured a KIH

28 value of 2.5 MPa√m [2.28 ksi√in] for Zr-2.5 Nb cladding at 160 degrees C [320 degrees F]. 

29 Based on the available data, the staff considered a reference KIH value of 5.0 MPa√m 

30 [2.73 ksi√in] to be reasonable for determining the potential for DHC initiation.

31 Raynaud and Einziger (2015) estimated the cladding hoop stresses while conservatively 

32 accounting for release of fission gases and decay gases during storage, including stresses due 

33 to radiation-induced pellet swelling during storage. Raynaud and Einziger concluded that DHC 

34 cannot occur for a KIH of 5 MPa√m [4.55 ksi√in], because the flaw size needed to induce DHC is 

35 much larger than the initial depth of potential existing cracks (120 μm [4.7 mils]). The estimated 

36 critical flaw size needed to initiate DHC in BWR fuel cladding is larger than 50 percent of the 

37 cladding thickness for 300 years of dry storage. For PWR cladding, the critical flaw size is 

38 larger than 30 percent of the cladding thickness for the first 5 years of the dry storage and larger 

39 than 50 percent of the cladding thickness beyond the first 5 years up to 300 years of dry 

40 storage. The calculations in Raynaud and Einziger did not account for the hoop stresses in 

41 ZIRLOTM-clad IFBA rods with hollow and solid blanket pellets, which are expected to be higher 

42 than standard rods (Bratton et al., 2015). Therefore, the staff performed similar calculations to 

43 those in Raynaud and Einziger for IFBA rods, assuming a KIH value of 5 MPa√m [2.73 ksi√in] 

44 and a conservative IFBA-rod hoop stress of 130 MPa [21.75 ksi]. These calculations show that 

45 the critical flaw size for the PWR cladding is still larger than 30 percent of the cladding thickness 

46 for the first 5 years of dry storage and larger than approximately 45 percent of the cladding 

47 thickness beyond the first 5 years up to 300 years of dry storage. Therefore, the staff concludes 

48 that the critical flaw size needed to induce DHC, in both standard and IFBA rods, is much larger 
3-91

1 than the initial depth of potentially existing cracks (120 μm [4.7 mils]). The staff considers that 

2 the hoop stress value assumed for IFBA rods is adequately conservative for this calculation,

3 since a limited (less than 1 percent) population of the rods is expected to experience these 

4 pressures (Bratton et al., 2015). In addition, most design-bases peak cladding temperatures are 

5 well below the limit defined in ISG-11, Revision 3 (i.e., 400 degrees C [752 degrees F]), which 

6 would considerably decrease the cladding hoop stresses. Therefore, the assumptions and 

7 analyses discussed above are considered reasonably bounding and indicate that DHC is not a 

8 credible aging mechanism during the 60-year timeframe.

9 The staff also considered a DHC model proposed by Kim et al. (2008, 2009b), which evaluated 

10 cladding absent thermal cycling, where multiple parameters including creep deformation, 

11 cladding burnup, solvus hysteresis, and the δ-to-γ hydride phase transition were analyzed. This 

12 model, still under review by the international DHC research community (NRC, 2014a), suggests 

13 that KIH may be reduced (i) upon cooling below 180 degrees C [356 degrees F] (due to a 

14 hydride phase transformation from the γ to δ phase) and (ii) if there are sufficient stresses and 

15 stress risers in the rod (e.g., residual stresses at the end cap weld region, incipient cracks due 

16 to fuel-cladding interaction). Thermal gradients may also affect the kinetics of hydride 

17 precipitation. The staff reviewed this study, in light of the assumptions made in the previous 

18 discussion. However, Kim does not quantify KIH values; therefore, adequate conclusions cannot 

19 be made with respect to threshold stresses. The NRC (2014a) and Hanson et al, (2012) 

20 summarized Kim’s work and proposed additional research for confirmation.

21 Finally, the staff considered the contribution of cladding stresses due to pellet-clad bonding and 

22 its potential to facilitate DHC initiation (Wang, 2014a,b). The previously-discussed Raynaud 

23 and Einziger (2015) study did not account for potential stress concentration effects due to 

24 pellet-pellet interfaces and pellet fragment-to-fragment friction forces that could result in more 

25 severe pellet-to-cladding mechanical interaction (PCMI) than for a perfectly cylindrical pellet 

26 (as assumed in the paper). Recently, Ahn et al. (2013) estimated stress concentrations from 

27 pellet-clad mechanical stresses due to the radiation-induced pellet swelling up to 100 years, 

28 independent of hoop stresses due to continued fission and decay gas release. The work 

29 estimated that, for HBU fuel, the average pellet-swelling-induced PCMI stress concentration 

was on the order of 200 MPa [29 ksi] locally.1 30 Literature indicates that radiation-induced pellet 

31 swelling is expected reach its maximum value beyond the 60-year timeframe (Rondinella et al., 

32 2010a,b; 2012). Therefore, the staff does not have evidence that the potential for high PCMI 

33 stress concentrations due to radiation-induced pellet swelling would facilitate DHC crack 

34 initiation until past the first renewal period.

35 Based on the above analyses and discussion, the staff concludes that delayed hydride cracking 

36 of the zirconium-based cladding is not credible during the 60-year timeframe and therefore,

37 aging management is not required.

38 3.6.1.3 Thermal creep (high burnup fuel)

39 Creep is the time-dependent deformation of a material under stress. Creep in zirconium-based 

40 cladding is caused by the hoop stresses from the rod internal pressure at a given fuel 

41 temperature. Therefore, the mechanism is expected to be self-limiting, due to the decreasing 

42 temperatures and creep-induced volume expansion, which results in lower internal rod

 

1For low-burnup fuel, pellet expansion stresses will be minimal, because the gap between the cladding and the pellet 

will accommodate the swelling.
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1 pressures with time. Excessive creep of the cladding during dry storage could lead to thinning, 

2 hairline cracks, or gross ruptures (Hanson et al., 2012), which may affect the ability to safely 

3 retrieve the HBU fuel on a single-assembly basis (if required by the design bases). 

4 The main driving force for cladding creep at a given temperature is the hoop stress caused by 

5 internal rod pressure, which accounts for the fission and decay gases released to the interspace 

6 between the fuel and cladding. Fuel pellet swelling also may result in localized stresses due to 

7 the mechanical interaction between the cladding and the fuel. Pellet swelling may occur due to 

8 (i) the incorporation of soluble and insoluble solid fission products in the fuel matrix, (ii) the 

9 formation of intra- and intergranular fission gas bubbles, particularly in the hot interior region of 

10 a fuel pellet, and (iii) the formation of a large number of small gas bubbles in the fine-grained 

11 ceramic structure that builds inward from the outer pellet surface for HBU fuel.

12 Raynaud and Einziger (2015) estimated the transient cladding hoop stresses during dry storage 

13 for typical 10 × 10 BWR and 17 × 17 PWR fuel assemblies. These estimates accounted for a 

14 credible release of fission and decay gases to the fuel-cladding interspace, pellet swelling, and 

15 fuel and cladding temperature decay with time. The study reported peak cladding hoop 

16 stresses less than 50 MPa [7.25 ksi] for BWR and less than 100 MPa [14.5 ksi] for PWR fuel 

17 assemblies. Raynaud and Einziger used these hoop stress estimates to calculate cumulative 

18 cladding strains for the representative assemblies over a 60-year period of dry storage. The 

19 authors reported a maximum cladding strain of 0.54 percent for the representative 10 × 10 BWR 

20 fuel cladding and 1.04 percent for the representative 17 × 17 PWR fuel cladding. However, 

21 these calculations did not account for the hoop stresses in ZIRLOTM-clad IFBA rods with hollow 

22 and solid blanket pellets, which are expected to be higher than those for standard rods (Bratton

23 et al., 2015). Therefore, the staff performed calculations to estimate the cladding strain for IFBA 

24 rods using the Raynaud and Einziger approach. Using a conservatively bounding hoop stress 

25 of 150 MPa [21.75 ksi], the maximum cladding strain was estimated to be near 2.1 percent. The 

26 elastic strain limit for various zirconium-based cladding alloys with circumferential hydrides is 

27 less than 1 percent (Geelhood et al., 2008) and is expected to be lower for cladding containing 

28 both circumferential and radial hydrides. Therefore, the staff concludes that the cladding in both 

29 standard and IFBA fuel rods is expected to undergo creep strains during the 60-year timeframe. 

30 The staff has discussed the potential for creep deformation in ISG-11, Revision 3 (NRC, 2003), 

31 which includes acceptance criteria (regarding maximum fuel clad temperature during dry 

32 storage operations and adequate thermal cycling limits) to provide reasonable assurance that 

33 the spent fuel assemblies will remain in the configuration analyzed in the approved design 

34 bases. The references cited in ISG-11, Revision 3, provide experimental evidence that cladding 

35 failures are not expected for creep strains below 2 percent. These references provide support 

36 that gross ruptures of the cladding are unlikely due to creep during dry storage, because the 

37 creep-induced strain is expected to be near or less than 2 percent for the majority of the 

38 cladding alloys and close to 2 percent for the ZIRLOTM-clad IFBA rods. For example, no failures 

39 were observed for creep strains below 2 percent strain for in-creep tests at temperatures 

40 between 250 and 400 degrees C [482 and 752 degrees F] for Zircaloy cladding irradiated up to 

41 burnup of 64 GWd/MTU (Spilker et al., 1997; Goll et al., 2001; EPRI, 2002). In addition, 

42 Bouffioux and Rupa (1998) conducted various cladding creep tests with unirradiated, 

43 prehydrided, stress-relief annealed low-Sn Zircaloy-4 PWR cladding tubes, with hydrogen levels 

44 in the range of 100–1,100 wppm. The authors observed gross ruptures of the cladding only 

45 after creep strains exceeding 8 percent. Tsai and Billone (2003) also tested irradiated 

46 stress-relief annealed Zircaloy-4 with varying levels of hydrogen levels at various temperature 

47 and hoop stresses, which did not reveal cladding failures at a strain of 5.83 percent. More 
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1 recent data on optimized ZIRLOTM by Pan et al. (2013) also indicate a plastic strain range in the 

2 same range as Zircaloy.

3 The staff concludes that thermal creep of zirconium-based cladding is credible during the 

4 60-year timeframe. However, due to the high creep capacity of zirconium-based alloys, thermal 

5 creep is not expected to result in cladding failures and reconfiguration of the fuel, if the 

6 approved design bases are consistent with the acceptance criteria in ISG-11, Revision 3. The 

7 staff recognizes that the experimental evidence used in support of ISG-11, Revision 3, is based 

8 on short-term testing. Therefore, the staff issued guidance in Appendix D of NUREG–1927, 

9 Revision 1 (NRC, 2016) for the use of a demonstration program to confirm these expected fuel 

10 conditions after a substantial storage period (~10 years). The program would provide 

11 confirmation for accelerated cladding creep testing used as basis for the guidance 

12 recommendation for the maximum temperature in ISG-11 (NRC, 2003), and that sufficient creep 

13 capacity exists for the renewal period. For example, non-destructive and destructive 

14 examination from the DOE/EPRI cask demonstration project (EPRI, 2014) may be used as 

15 confirmation that the design-basis fuel remains in the analyzed configuration and that sufficient 

16 creep margin exists for the first renewal period. An example AMP consistent with the guidance 

17 in Appendix D of NUREG–1927, Revision 1 is provided in Chapter 5.

18 3.6.1.4 Low-temperature creep (high burnup fuel)

19 Low-temperature creep (also called “athermal creep”) may occur when sustained hoop stresses 

20 operate on the cladding material at or near ambient temperature (NRC, 2014a). Various 

21 athermal creep mechanisms have been proposed at low stresses (e.g., Nabarro-Herring, Coble, 

22 and Harper-Dorn creep mechanisms) (Murty, 2000), although there is no evidence or literature 

23 information to support that these will be operational on zirconium-based alloys. However, the 

24 literature shows that low-temperature creep has been shown to occur in titanium and its alloys, 

25 which leads to deformation twinning (Jaworski and Ankem, 2006). Since both titanium and 

26 zirconium have the same crystalline structure (hexagonal close packed crystalline), the 

27 zirconium-based cladding was reviewed for its susceptibility to low-temperature creep.

28 In materials such as α and α-β titanium alloys, which are comparable to the zirconium-based 

29 alloys used for fuel cladding, low-temperature creep has been observed when tensile stresses 

30 exceed 25 percent of the yield strength (Ankem and Wilt, 2006). For example, Ankem and Wilt 

31 reported a threshold stress in the range of 25–50 percent of the yield stress for Ti Grade 7, and 

32 35–60 percent of the yield stress for Ti Grade 24. The yield strength of the irradiated 

33 zirconium-based cladding at low temperatures (550–1,000 MPa [79.8–145 ksi]; Geelhood et al, 

34 2008; Forgeaud, et al., 2009; Cazalis et al., 2005) is expected to be close to the yield strength of 

35 Ti Grade 24 (825 MPa [119.6 ksi]) and well above the yield strength of Ti Grade 7 (275 MPa 

36 [39.9 ksi]) (Ibarra et al., 2007). Therefore, the staff considered the results in Ankem and Wilt to 

37 provide reasonable acceptance criteria for determining if low-temperature creep is a credible 

38 aging mechanism in the 60-year time frame.

39 The main sources of sustained hoop stresses at low temperatures are expected to be the rod 

40 internal pressure and pellet-cladding mechanical interaction. Raynaud and Einziger (2015) 

41 estimated the cladding hoop stresses after 300 years of storage to be approximately 25 MPa 

42 [3.62 ksi] and 35 MPa [5.07 ksi] for representative BWR and PWR fuel cladding, respectively. 

43 These estimates accounted for a credible release of fission and decay gases to the 

44 fuel-cladding interspace, pellet swelling, and fuel and cladding temperature. The hoop stresses 

45 for IFBA rods are conservatively expected to be around or less than 75 MPa [10.87 ksi] 

46 (Bratton et al., 2015). These hoop stress estimates are all less than 25 percent of the yield 
3-94

1 strength of zirconium-based cladding (i.e., below the expected range of 550–1,000 MPa [79.8–

2 145 ksi] near ambient temperature for cladding with circumferential hydrides only

3 (Geelhood et al., 2008; Fourgeaud, et al. 2015; Cazalis et al., 2005)). Further, more recent data

4 (Kim et al., 2015a, 2015b) suggest that, even with the potential decrease in yield strength due to 

5 radial hydrides (which conservatively does not account for a potential increase in yield strength 

6 due to irradiation), the hoop stresses in the cladding are still maintained below 25 percent of the 

7 yield strength of irradiated cladding with both circumferential and radial hydrides. 

8 Raynaud and Einziger acknowledged that the low-temperature creep models are not 

9 programmed into FRAPCON-DATING, which the authors used to predict the elevated 

10 temperature cladding creep (see Section 3.6.1.3). The authors noted that extrapolations of the 

11 high-temperature cladding creep model results in immeasurably small values of cladding strains 

12 at low temperature. However, the lack of cladding creep beyond 50 years (corresponding to 

13 temperatures below approximately 200 degrees C [ 392 degrees F]) results in smaller strains 

14 being predicted in these calculations. Therefore, the calculated cladding hoop stresses are 

15 conservative when compared to the 25-percent criteria, as athermal creep-induced strains 

16 would reduce these stresses.

17 The staff further considered the contribution of cladding stresses due to pellet-clad bonding and 

18 its potential to facilitate athermal creep. The previously discussed Raynaud and Einziger study 

19 did not account for potential stress concentration effects due to pellet-pellet interfaces and pellet 

20 fragment-to-fragment friction forces that could result in more severe PCMI than for a perfectly 

21 cylindrical pellet (as assumed in the paper). Recently, Ahn et al. (2013) estimated stress 

22 concentrations from pellet-clad mechanical stresses caused by the radiation-induced pellet 

23 swelling up to 100 years, independent of hoop stresses due to fission and decay gas release. 

24 The work estimated that, for HBU fuel, the average pellet-swelling-induced PCMI stress 

25 concentration was on the order of 200 MPa [29 ksi] locally. Literature indicates that 

26 radiation-induced pellet swelling is expected to reach its maximum value beyond the 60-year 

27 timeframe (Rondinella et al., 2010a,b; 2012). Therefore, PCMI stress concentrations due to 

28 radiation-induced pellet swelling are not expected to exceed a threshold stress of 25 percent of 

29 the yield stress (similar to the titanium data in Ankem and Wilt, 2006) during the 60-year 

30 timeframe. 

31 In summary, literature on the creep strain and creep rate of the zirconium-based cladding 

32 materials at room temperature per the hoop stresses expected during extended storage is not 

33 available. Therefore, it is not possible to directly assess the low-temperature creep of the 

34 zirconium-based cladding materials. However, the staff has reviewed the threshold levels of 

35 tensile stresses for low-temperature creep in the similar crystalline-structured (hexagonal close 

36 packed crystalline) materials, which indicate that cladding hoop stresses on the cladding must 

37 exceed approximately 25 percent of yield strength for athermal creep to be credible. The room 

38 temperature hoop stresses on the zirconium-based cladding are expected to be less than 

39 25 percent of the yield strength. Therefore, the low-temperature (athermal) creep mechanism 

40 is not considered credible, even for the unlikely scenario where fuel reaches room

41 temperature during the 60-year timeframe. Therefore aging management is not required during 

42 the 60-year timeframe.

43 3.6.1.5 Mechanical overload (high burnup fuel)

44 Mechanical overload is generally associated with PCMI, which could compromise the cladding 

45 integrity during storage. PCMI is likely during reactor operations when the reactivity transient 

46 during a reactivity-initiated accident (RIA) results in a rapid increase in a fuel rod power, leading 
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1 to a nearly adiabatic heating of the fuel pellets and potential failure of the fuel cladding. In either 

2 commercial BWRs or PWRs, cladding failures have not been attributed to PCMI. However, data 

3 generated in experimental reactors conducting ramp testing of heavily hydrided fuel claddings 

4 indicate that hydride rims with large hydride number density at the cladding outer surface may 

5 lead to crack initiation (Adamson et al., 2006). The cracks could propagate from the outside 

6 toward the inner cladding surface, potentially resulting in failures.

7 During dry storage, PCMI stresses could develop due to pellet swelling and release of fission 

8 gases to the gap between the fuel and cladding. PCMI could lead to the opening of existing 

9 flaws in the cladding, potentially resulting in the release of fission gases and other fission 

10 products into the cask environment. The existing flaws in undamaged fuel are likely to be of any 

11 of the following: (i) surface (nonthrough-wall) cracks on the inner or outer wall, (ii) hairline 

12 cracks, (iii) wall thinning due to oxide spallation on the outer surface, or (iv) wall thinning due to 

13 fretting wear on the outer surface (NRC, 2014a).

14 Jernkvist et al. (2004) developed a criterion to determine the likelihood of PCMI during RIA, 

15 which relies on estimating a threshold strain as a function of temperature, strain rate, hydrogen 

16 concentration in cladding, and neutron fluence. However, this criterion is only applicable when

17 the cladding temperature is increasing, making it inapplicable to dry storage, where

18 temperatures decrease with time, barring any fluctuations from changes in ambient

19 temperature.

20 A method previously used to characterize PCMI failures in the cladding involves measuring the 

21 creep strain capacity at a given creep strain rate (Jernkvist et al., 2004). More specifically, 

22 PCMI-induced failures are observed when the cladding strain at a given strain rate exceeds a 

23 threshold (Jernkvist et al., 2004; Fuketa et al., 2003). The threshold strain is a function of 

24 cladding temperature, irradiation, and hydrogen concentration. PCMI-induced failures have 

25 been reported at cladding strains exceeding 1 percent for strain rates in the range of 

10−5 to 10−3 s−1 26 at room temperature for various levels of hydrogen concentration (Jernkvist et 

27 al., 2004). At higher temperatures, the strain at failure is above 6 percent between 523 and 

673 K [482 to 752 degrees F] for strain rates in the range of 10−5 to 10−3 s−1 28 (Jernkvist et al., 

29 2004). This threshold strain at higher temperature is applicable for cladding hydrogen content 

30 up to 1,200 wppm. These results are consistent with those by Fuketa et al. (2003), which 

31 exhibited similar threshold strains between 373 and 573 K [212 to 572 degrees F] with hydrogen 

32 concentrations up to 1,450 wppm. These results can be compared with data discussed in 

Section 3.6.1.3, which show that, for comparable strain rates in the order of 10−4 s−1 to 10−5 s−1 33 , 

34 no failures were observed for creep strains below 2 percent for in-creep tests at temperatures 

35 between 150 and 400 degrees C [423 and 752 degrees F] for Zircaloy cladding irradiated up to 

36 burnup of 64 GWd/MtU (Spilker et al., 1997; Goll et al., 2001; EPRI, 2002).

37 The staff reviewed the aforementioned creep strain and strain rate threshold criteria against the 

38 results in Raynaud and Einziger (2015), which estimated the temperature-dependent hoop 

39 stresses on the cladding while accounting for credible release of fission and decay gases and 

40 pellet swelling. The authors estimated maximum cladding strains of 0.54 percent for the 

41 10 × 10 BWR fuel cladding and 1.04 percent for the 17 × 17 PWR fuel cladding at a strain rate 

of 10−10 s−1 42 expected during dry storage. The authors stated that all of the cladding strain is 

43 expected to occur during the first 50 years of storage. These calculations did not account for 

44 the hoop stresses in ZIRLOTM-clad IFBA rods with hollow and solid blanket pellets, which are 

45 expected to be higher than standard rods (Bratton et al., 2015). The staff performed 

46 calculations to estimate the cladding strain for IFBA rods using the Raynaud and Einziger 

47 approach. Using a conservatively bounding hoop stress of 150 MPa [21.75 ksi], the maximum 
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1 cladding strain was estimated to be near 2.1 percent for IFBA rods. These values indicate 

2 sufficient strain capacity per the previously discussed creep strain and strain rate threshold 

3 criteria (Jernkvist et al., 2004; Fuketa et al., 2003), which is considered conservatively bounding 

4 as the strain rates in dry storage are expected to be approximately five to seven orders of 

magnitude lower than 10−5 to 10−3 s−1 5 . Therefore, the staff concludes that cladding failures due 

6 to PCMI-induced mechanical overload are not considered credible during the 60-year 

7 timeframe, and aging management is not required. 

8 3.6.1.6 Oxidation

9 In the presence of residual amounts of water and high enough temperature, zirconium-based 

10 cladding can be oxidized according to the following chemical reaction: Zr + 2H2O = ZrO2 + 2H2

11 (Jung et al., 2013; Cox, 1976, 1988; Rothman, 1984).

12 Jung et al. (2013) conducted various scoping calculations to determine the extent of cladding 

13 oxidation during dry storage in the presence of up to 1 L [0.26 gal] (equivalent to 55.5 moles) of 

14 residual water. The amount of residual water considered is significantly higher than the residual 

15 water amount of 0.43 moles expected after vacuum drying, as per NUREG–1536 (NRC, 2010). 

16 The scoping calculations were based on a representative storage system loaded with the 

17 equivalent of 21 Babcock & Wilcox SNF assemblies, each containing 208 fuel rods in a storage 

18 canister. Jung et al. discussed temperature-dependent cladding oxidation kinetics for both 

19 Zircaloy-2 and Zircaloy-4, concluding that the maximum cladding thickness loss due to oxidation 

20 is not expected to exceed 10 μm [0.4 mils], even with complete consumption of the assumed 1 L 

21 [0.26 gal] of residual water. The loss of cladding thickness due to oxidation represents less than 

22 2 percent of the original cladding thickness. Therefore, cladding oxidation is considered to be 

23 insignificant, and aging management is not required during the 60-year timeframe.

24 3.6.1.7 Pitting corrosion

25 Pitting corrosion initiates and propagates when (i) there is an aggressive chemical environment 

26 that results in corrosion potential being greater than the repassivation potential and (ii) there is 

27 enough cathodic capacity to sustain the propagation of the pitting corrosion (Shukla et al., 

28 2008). Zirconium is a passive material and is protected by a ZrO2 surface film (Palit and 

29 Gadiyar, 1987). The surface oxide readily reforms if broken, but zirconium is not completely 

30 immune to pitting. Halides (i.e., anions of fluorine, chlorine, bromine, and iodine) in aqueous or 

31 gaseous forms could initiate pitting. For example, pitting of zirconium has been shown to occur 

in hydrochloric acid solutions containing ferric (Fe3+) or cupric (Cu2+ 32 ) ions (Palit and Gadiyar, 

33 1987).

34 Inside the cask’s or canister’s internal environment, a limited amount of residual water is 

35 expected to be retained following drying, which will be in the liquid state once temperatures are 

36 near or below 100 degrees C [212 degrees F]. The residual water amount is expected to be 

37 less than 1 mole per NUREG–1536 (NRC, 2010). During storage, most residual water is 

38 expected to decompose into hydrogen and oxidizing species, such as oxygen and hydrogen 

39 peroxide, with time (Jung et al., 2013). It is possible for trace amounts of water to remain in the 

40 vapor phase but is not expected to be in the liquid phase during dry storage, due to the low 

41 relative humidity in the cask or canister cavity. For example, the relative humidity inside a 

cavity volume of 2.1 m3 42 [554.8 gal], assuming a residual water content of 0.43 mole 

43 [per NUREG–1536] at 25 degrees C [77 degrees F], is estimated to be approximately 

44 15 percent using a backfill pressure of 1 atmosphere (atm) [14.7 psi], or 6 percent, using a 

45 backfill pressure of 5 atm [73.5 psi]. Further, any residual water in the vapor phase is expected 
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1 to be spread throughout the cavity and is not expected to be sufficient to provide enough 

2 cathodic capacity to initiate and propagate pitting corrosion of the cladding. Confirmation of this 

3 expectation is provided in Einziger et al. (2003), which did not observe any evidence of pitting 

4 corrosion in cladding after 15 years of dry storage. Therefore, pitting corrosion of the cladding is 

5 not considered credible, and aging management is not required during the 60-year timeframe.

6 3.6.1.8 Galvanic corrosion

7 Galvanic corrosion can occur due to a mismatch in corrosion potentials between two metals in 

8 an aqueous solution. In fuel assemblies, the mismatch can occur when the cladding is in 

9 contact with other metallic components, which could result in the formation in a galvanic cell,

10 provided there is an aqueous solution between the two subcomponents. For example, some of 

11 the PWR and BWR fuel assemblies contain spacer grids that are made of Inconel alloys, such 

12 as Inconel 718 and Inconel 625. The dominant constituents of these Inconel alloys include 

13 nickel, chromium, molybdenum, iron, niobium, and tantalum. A galvanic cell could form if 

14 residual water condenses in the gap between the rod and a spacer grid, simultaneously 

15 contacting both materials. The cladding could also be covered with a crud layer deposit during 

16 reactor operations, which could further facilitate formation of the contact.

17 The standard electrode potential for zirconium and ZrO2 in aqueous solution at 25 degrees C 

18 [77 degrees F] is approximately in the range of −1.5 to −1.6 VSHE, where the subscript “SHE” 

19 stands for standard hydrogen electrode (Haynes et al., 2013). The standard electrode 

20 potentials for chromium, nickel, molybdenum, and iron are approximately equal to −0.74, −0.20, 

21 −0.26, and −0.44 VSHE, respectively, at 25 degrees C [77 degrees F] (Bard and Faulkner, 1980; 

22 Haynes et al, 2013). The standard electrode potential data indicate that zirconium would be 

23 oxidized to zirconium ions during the galvanic reaction, and oxidizing species, such as oxygen 

24 and hydrogen peroxide in aqueous solution, would be reduced at the Inconel alloy. The extent 

25 of loss of cladding material would depend on the amount of oxidants present in the condensed 

26 water. For example, per the stoichiometry of the oxidation and reduction reactions (Jung et al, 

27 2013), reduction of 1 mole of hydrogen peroxide would result in oxidation of 0.5 mole of 

28 zirconium. Similarly, reduction of 1 mole of oxygen would result in oxidation of 1.0 mole of 

29 zirconium. Jung et al. reported scoping calculations to determine the extent of zirconium 

30 oxidation with 1 mole of a 5 weight percent H2O2 aqueous solution saturated with oxygen at 

31 25 degrees C [77 degrees F] and 1 atm [14.7 psi]. Jung et al. concluded that the extent of 

32 oxidation would depend on the spread of the condensed water over the large surface area. 

33 Therefore, the effect of galvanic corrosion is not expected to be localized.

34 The amount of residual water inside the cask or canister following drying is expected to be less 

35 than 1 mole after vacuum drying, as per guidance in NUREG–1536 (NRC, 2010). Most residual 

36 water is expected to decompose over time into hydrogen and oxidizing species, such as oxygen 

37 and hydrogen peroxide (Jung et al., 2013). It is possible for some trace amount of water to 

38 remain in the vapor phase inside the canister after the first renewal period but is not expected to 

39 condense into liquid phase during dry storage due to the low relative humidity of the 

40 containment cavity. For example, the relative humidity inside a canister with a cavity volume of 

2.1 m3 41 [554.8 gal], assuming a residual water content of 0.43 mole (per NUREG–1536) and at 

42 25 degrees C [77 degrees F] is estimated to be approximately 15 percent with a backfill 

43 pressure of 1 atm, or 6 percent with backfill pressure of 5 atm [73.5 psi]. Further, any residual 

44 water in the vapor phase is expected to be spread throughout the containment cavity and is not 

45 expected to be sufficient to form a corrosion cell between the cladding and the spacer grids 

46 made of Inconel alloys. Therefore, galvanic corrosion of the zirconium-based cladding alloys is 

47 not considered credible, and aging management is not required during the 60-year timeframe.
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1 3.6.1.9 Stress corrosion cracking 

2 SCC occurs as a result of a synergistic combination of a susceptible material, an aggressive 

3 environment, and sufficiently high tensile stress. The corrosive environment associated with 

4 SCC of fuel rods has been attributed to specific fission products, such as iodine, cesium, and 

5 cadmium, generated during reactor irradiation (Wisner and Adamson, 1982; Sidky, 1998). SCC 

6 of the cladding can occur at the rod’s inner surface where the fuel pellet and cladding 

7 mechanically interact and is related to PCMI hoop stresses on the cladding. SCC of 

8 zirconium-based cladding has been observed in BWRs during power ramp-up (NRC, 1985; 

9 Adamson, 2006). PWR cladding is unlikely to undergo similar SCC because of the more 

10 gradual power ramp-up. Fuel pellets in PWR cladding are unlikely to undergo sudden 

11 expansion and induce high stresses, as in BWR cladding. No cladding failures from SCC are 

12 known to have occurred either during pool storage or under dry storage conditions.

13 Prescatore and Cowgill (EPRI, 1997) compiled SCC failure data from Yagee et al. (1979, 1980), 

14 Mattas et al.(1982), Shimada and Nagai (1983), Kreyns et al. (1976), and Crescimanno (1984) 

15 for the following irradiated cladding materials: recrystallized Zircaloy-2, stress-relieved 

16 Zircaloy-2, recrystallized Zircaloy-4, and stress-relived Zircaloy-4. For Zircaloy-2, the reported 

17 data’s temperature and tensile stress ranges were 325 to 350degrees C [617 to 662 degrees F], 

18 and 119 to 513 MPa [17.3 to 74.4 ksi], respectively. Similarly for Zircaloy-4, the reported SCC 

19 data’s temperature and tensile stress ranges were 316 to 350 degrees C [601 to 

20 662 degrees F], and 164 to 414 MPa [23.8 to 60 ksi], respectively. In the listed data, the 

21 SCC-induced failure was reported at 157 MPa [22.8 ksi] and 325 degrees C [617 degrees F] for 

22 Zircaloy-2, and at 205 MPa [29.7 ksi] and 360 degrees C [680 degrees F] for Zircaloy-4 (Yagee, 

23 1979). Regarding these two failure data points (157 MPa [22.8 ksi] and 325 degrees C 

24 [617 degrees F] for Zircaloy-2 and 205-MPa [29.7-ksi] and 360 degrees C [680 degrees F] for 

25 Zircaloy-4), Prescatore and Cowgill (EPRI, 1997) argued that failures were misclassified as 

26 SCC-induced failures and were more akin to nondetrimental pinhole breaches. Prescatore and 

27 Cowgill stated that gross rupture, in the form of axial splitting, was noted in many instances 

28 when the stress was greater than about 270 MPa [39.2 ksi], but at lower stresses, pinhole 

29 leakage was by far the more common failure mode. If the 157 MPa [22.8 ksi] and 

30 325 degrees C [617 degrees F] data point is excluded from the listed data for Zircaloy-2, as 

31 argued by Prescatore and Cowgill, the next incident of the SCC-induced failure is noted at 

32 247 MPa [35.8 ksi] at 325 degrees C [617 degrees F] for Zircaloy-2. Similarly, if the 205 MPa 

33 [29.7 ksi] at 360 degrees C [680 degrees F] data point is excluded for Zircaloy-4, as argued by 

34 Prescatore and Cowgill, the next incident of the SCC-induced failure is noted at 273 MPa 

35 [39.6 ksi] at 360 degrees C [680 degrees F]. This analysis indicates that at least 240 MPa 

36 [34.8 ksi] of hoop stresses are needed to induce SCC for both Zircaloy-2 and Zircaloy-4.

37 Recent work by Raynaud and Einziger (2015) shows that hoop stresses are expected to be 

38 below 100 MPa [14.5 ksi], with the most realistic estimate of release of the decay and fission 

39 gases from fuel pellets and with the best estimate of fuel swelling during a 300-year dry storage 

40 period. However, hoop stresses in ZIRLOTM-clad IFBA rods with hollow and solid blanket 

41 pellets could be considerably higher. The Raynaud and Einziger study did not account for 

42 potential stress concentration effects due to pellet-pellet interfaces and pellet 

43 fragment-to-fragment friction forces that could result in more severe PCMI than for a perfectly 

44 cylindrical pellet (as assumed in Raynaud and Einziger). Recently, Ahn et al. (2013) estimated 

45 stress concentrations from pellet-clad mechanical stresses due to the radiation-induced pellet 

46 swelling up to 100 years, independent of hoop stresses due to fission and decay gas release. 

47 The work estimated that, for HBU fuel, the average pellet-swelling-induced PCMI stress 

48 concentration was on the order of 200 MPa [29 ksi] locally. For low-burnup fuel, pellet 
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1 expansion stresses will be minimal, because the gap between the cladding and the pellet will 

2 accommodate the swelling. Literature indicates that radiation-induced pellet swelling is 

3 expected to reach its maximum beyond the first renewal period (Rondinella et al., 2010a,b; 

4 2012). Even with the PCMI-induced hoop stresses, the cladding stresses will remain well below 

5 the 240 MPa [34.8 ksi] criterion for inducing SCC. Therefore, SCC of the cladding is not 

6 considered credible, and aging management is not required during the 60-year timeframe.

7 3.6.1.10 Radiation embrittlement

8 Radiation embrittlement of cladding can result in degradation of the mechanical properties of the 

9 cladding, such as ductility and strength (PNNL, 2012; NRC, 2014a). This can lead to the 

10 reduction in the maximum load that the cladding can withstand, potentially leaving the cladding 

11 vulnerable to failure under external loads.

12 Radiation embrittlement of the cladding is mostly observed during reactor operation due to 

cumulative fast neutron fluence on the order of 1022 n/cm2 [6.5 × 1022 n/in2 13 ] (Hermann et al., 

14 2001) for recrystallized annealed Zircaloy-2 and cold-worked stress-relieved Zircaloy-4 (Morize 

15 et al., 1987). During normal operation in the reactor, the cladding material is bombarded with 

16 fast neutrons that cause atomic displacement cascades, resulting in the formation of point 

17 defects (PNNL, 2012; NRC, 2014a; NWTRB, 2010). This leads to the reduction in the 

18 mechanical properties of the cladding material.

19 In dry storage, the cumulative neutron fluence is expected be five orders of magnitude less than 

20 in reactor service (Jung et al., 2013). In addition, annealing of irradiation hardening could occur 

21 during storage, which would help recover some ductility. It has been shown in literature 

22 (Masafumi et al., 2007; Torimaru, et al., 1996) that a post-irradiation heat treatment performed 

23 at a temperature above the irradiation temperature can lead to the recovery of the 

24 radiation-induced hardening and increased ductility of the cladding. Ito et al. (2004) further 

25 showed that hardness also recovers at temperatures lower than an irradiation temperature of 

26 360 degrees C [680 degrees F]. More specifically, Ito et al. (2004) showed that hardness 

27 continued to recover, albeit quite slowly, at temperatures as low as 330 degrees C 

28 [626 degrees F] for 8,000 hours (0.9 year), and nearly 50 percent recovery was observed 

29 compared to the annealing over the same time at 360 degrees C [680 degrees F]. Thus, over 

30 many years of extended storage, it is possible that thermal annealing could increase cladding 

31 ductility, thereby reducing the effects of radiation embrittlement.

32 Because radiation embrittlement is associated with a cumulative fluence of on the order of 

1022 n/cm2 [6.5 × 1022 n/in2 33 ], which is not expected during storage, radiation embrittlement of 

34 cladding is not considered credible, and therefore, aging management is not required during the 

35 60-year timeframe.

36 3.6.1.11 Fatigue

37 Fatigue occurs when a material is subjected to repeated loading and unloading stresses. If the 

38 loads are above a certain threshold, microscopic cracks will begin to form at stress 

39 concentrators at the surface, persistent slip bands, and grain interfaces. As a crack reaches a 

40 critical size, it will propagate until fracture. Because dry storage is a passive application, purely 

41 mechanical cyclic loading is not expected. However, the cladding will experience thermal cycles 

42 due to daily and seasonal fluctuations in ambient temperature, as well as extreme weather 

43 events within a larger seasonal pattern. These thermal cycles will induce cyclic stresses on the 

44 cladding due to either (i) changes in fission and decay gas pressure, as governed by gas laws, 
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1 which would result in fluctuations in cladding hoop stresses, and (ii) partial restraint on cladding 

2 thermal expansion and contraction due to top and bottom nozzles, hold-down springs, and 

3 spacer grids. These thermally induced stresses and corresponding strains can produce fatigue 

4 damage in the same manner as purely mechanical cyclic loading. 

5 Devoe and Robb (2015) conducted steady-state analyses to show that the change in peak 

6 cladding temperature is directly proportional to the change in external air temperature of the 

7 canister. Although the large thermal mass of the DSS is likely to reduce the amplitude and 

8 frequency of the thermal cycles on fuel and cladding temperature, Devoe and Robb assumed a 

9 correlation coefficient of unity between the peak cladding and external air temperature. Thus, a 

10 1 degree C [1.8 degree F] change in air temperature would result in approximately 1 degree C 

11 [1.8 degree F] change in cladding temperature. When evaluating daily temperature fluctuations, 

12 the analysis assumed a conservative 25 degrees C maximum daily change [equivalent to 

13 45 degrees F change], which is the mean daily temperature change in the United States. The 

14 model further assumes a total of 21,900 thermal cycles, corresponding to steady-state 

15 temperature cycle every day for 60 years. The staff assumed these conditions to determine if

16 the resulting changes in cladding hoop stresses could lead to fatigue-induced failure of the 

17 cladding.

18 Raynaud and Einziger (2015) estimated the cladding hoop stresses while accounting for release 

19 of fission gases and decay gases during storage, including pellet swelling stresses due to 

20 radiation damage during storage. Raynaud and Einziger estimates included the effect of fuel 

21 temperature on cladding hoop stresses. As per the Raynaud and Einziger estimates, a 

22 25 degree C variation [45 degree F variation] in cladding temperature will cause up to 10 and 

23 30 MPa [1.45 and 4.35 ksi] fluctuations in hoop stress of the BWR and PWR claddings, 

24 respectively. Lin and Haicheng (1998) conducted experimental studies to determine fatigue 

25 properties of zirconium and Zircaloy-4. Lin and Haicheng (1998) provided a fatigue lifetime 

26 curve for zirconium and Zircaloy-4 under reversal bending as a function of the cyclic stress. As 

27 per the fatigue lifetime curve in Lin and Haicheng, a cyclic stress amplitude of more than 

260 MPa [37.7 ksi] is needed for fatigue-induced failure in Zircaloy-4 in 107 28 cycles. The curve 

29 also bounds the data for zirconium, and hence, is also assumed to be applicable for other 

zirconium-based cladding materials, such as Zircaloy-2, ZIRLOTM, and M5® 30 . Therefore, using 

31 the fatigue lifetime curve in Lin and Haicheng, these fluctuations in hoop stresses (per the 

32 assumed conditions in Devoe and Robb, 2015) are not sufficient for fatigue-induced failure in 

33 the cladding.

34 The staff also evaluated the effects of extreme seasonal temperature variations, as these are 

35 expected to be significantly higher than daily variations and could result in higher cyclic stress 

36 amplitudes. Using the off-normal DSS operating conditions of −40 degrees C [−40 degrees F] 

37 (winter) and 103 degrees C [217 degrees F] (summer) yields a maximum seasonal temperature 

38 variation of 143 degrees C [variation of 257 degrees F]. Similar to the previous analysis, per the

39 Raynaud and Einziger (2015) estimates, a 143 degree C variation [257.4 degree F variation] in 

40 cladding temperature will cause up to 10 and 55 MPa [1.45 and 7.8 ksi] fluctuations in hoop 

41 stress of the BWR and PWR claddings, respectively. Using the fatigue lifetime curve in Lin and 

42 Haicheng (1998), these fluctuations in hoop stresses (per the assumed conditions in Devoe and 

43 Robb, 2015) are also not sufficient for fatigue-induced failure in the cladding.

44 As discussed in Section 3.2.1.7, the cyclic stress, σσ, induced by the thermal variations also 

45 depends on the material’s coefficient of thermal expansion (α0) and Young’s modulus of 

46 elasticity (E), the actual change in temperature (∆TT), and the degree of constraint on the 

47 component. Since the degree of constraint for the cladding is not readily available for cladding, 
3-101

1 a conservative approach is employed to estimate the cyclic stresses and associated potential 

2 impact of thermal fatigue. The coefficient of thermal expansion is estimated to be approximately 

4.16 × 10−6 3 /K, based on the data in Luscher and Geelhood (2010). The Young’s modulus of 

4 elasticity of various zirconium-based cladding materials ranges between 32 and 100 GPa 

5 [4,641 and 14,504 ksi] (Luscher and Geelhood, 2010); a value of 100 GPa [14,504 ksi] is 

6 conservatively used. The assumed values of α0 and E result in a thermally induced cyclic stress 

7 of 10.4 MPa [1.5 ksi] and 59.5 MPa [8.6 ksi] for ∆TT equal to 25 and 143 degrees C [45 and 

8 257 degrees F] , respectively. As per the fatigue lifetime curve in Lin and Haicheng (1998), 

9 these fluctuations in hoop stresses are also not sufficient for fatigue-induced failure in 

10 the cladding.

11 The staff further considered the cumulative cyclic stresses for all cases described above, which 

12 results in stresses ranging from 20 to 70 MPa [2.9 and 10.2 ksi] for BWR and from 65 to

13 115 MPa [9.4 and 16.7 ksi] for PWR claddings. Even the combined conservative values are 

14 well below the threshold of 260 MPa [37.7 ksi] needed for fatigue-induced failure in the cladding, 

15 per Lin and Haicheng (1998). Therefore, the staff concludes that fatigue-induced failure of the 

16 cladding is not credible during the 60-year timeframe, and aging management is not required.

17 3.6.2 Assembly hardware materials

18 The assembly hardware considered here includes guide tubes, spacer grids, and lower and 

19 upper end fittings. The guide tubes are fabricated using zirconium-based alloys. The other 

20 components are fabricated using one of the following materials: zirconium-based alloys, 

21 Inconel 718, Inconel 625, Inconel X-750, and stainless steel 304L. These subcomponents are 

22 not expected to experience sustained external loads during passive dry storage except for their 

23 own weight.

24 3.6.2.1 Creep

25 Creep is defined as the time-dependent deformation that takes place at an elevated 

26 temperature and constant stress. Because the deformation processes that produce creep are 

27 thermally activated, the rate of this time-dependent deformation (i.e., the creep rate) is a strong 

28 function of the temperature. The creep rate also depends on the applied stress but does not 

29 generally vary with the environment. As a general rule of thumb, at temperatures below 0.4Tm, 

30 where Tm is the melting point of the metal in Kelvin, thermal activation is insufficient to produce 

31 significant creep (Cadek, 1988). The melting temperature of various zirconium alloys is above 

32 1,800 degrees C [3,272 degrees F]. Similarly, the melting temperature of various Inconel alloys 

33 is above 1,260 degrees C [2,300 degrees F]. In addition, the melting temperature of 304L 

34 stainless steels is close to 1,400 degrees C [2,552 degrees F].

35 Regarding the zirconium alloys, the 0.4Tm criterion yields a creep threshold of 556 degrees C 

36 [1,033 degrees F]. The maximum expected temperature of fuel cladding has been estimated to 

37 be 400 degrees C [752 degrees F] at the beginning of storage (Jung et. al., 2013). This 

38 cladding temperature is expected to decrease to around 266 degrees C [510 degrees F] after 

39 20 years and to approximately 127 degrees C [261 degrees F] after 60 years. This indicates 

40 that creep of the zirconium alloys is unlikely during the renewal period.

41 Regarding Inconel alloys, the 0.4Tm criterion yields a creep threshold of 340 degrees C 

42 [644 degrees F]. As stated previously, the peak temperature inside the storage canister is 

43 expected to be below 266 degrees C [510 degrees F] after 20 years of storage. This indicates 

44 that creep of various Inconel alloys is unlikely during the renewal period.
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1 Regarding 304L stainless steel, the 0.4Tm criterion yields a creep threshold of 396 degrees C 

2 [755 degrees F]. As stated previously, the peak temperature inside the storage canister is 

3 expected to be below 300 degrees C [572 degrees F] after 20 years of storage. Further, the 

4 0.4Tm rule of thumb underestimates the minimum creep temperature for steels, because 

5 temperatures above 500 degrees C [932 degrees F] are required for significant creep in steels 

6 (Samuels, 1988). This indicates that creep of 304L stainless steel is unlikely during the 

7 renewal period. 

8 Therefore, creep of the assembly hardware is not considered credible, and aging management 

9 is not required during the 60-year timeframe.

10 3.6.2.2 Hydriding

11 Assembly hardware such as guide tubes and spacer grid materials made from zirconium alloys 

12 could potentially be subjected to hydriding effects that could reduce the material’s ductility and

13 fracture toughness, particularly at lower temperatures (less than 200 degrees C 

14 [392 degrees F]), once the fuel has cooled (PNNL, 2012).

15 Hydriding may occur in zirconium alloys that experience hydrogen pickup in reactor service 

16 (NRC, 2014a). As the temperature of the assembly hardware decreases, zirconium hydrides

17 precipitate due to the decreasing hydrogen solubility in the zirconium matrix. The hydride 

18 precipitation will occur when the hardware cools in the spent fuel pools after reactor discharge. 

19 Some of the hydride will dissolve during the drying process and will reprecipitate due to 

20 subsequent cooling during storage. Unlike fuel rods with cladding, there is no hoop stress for 

21 the zirconium-based assembly hardware to cause hydride reorientation. Any load on the 

22 assembly hardware is predominantly expected due to its own weight, which is not sufficient to 

23 be equivalent to hoop stresses to cause hydride reorientation. In addition, any additional 

24 hydriding of the assembly hardware during extended storage is expected to be negligible 

25 (Jung et al., 2013). 

26 In summary, the impact of hydriding effects on assembly hardware, especially guide tubes, is far 

27 less severe than for cladding with fuel (EPRI, 2011; PNNL, 2012; Hanson et al., 2012). 

28 Because there is limited load during storage on assembly hardware, it is unlikely that hydriding 

29 will affect the ability of the assembly hardware to ensure that the spent fuel remains in the 

30 as-analyzed configuration. Confirmation of this expectation is provided by Einziger et al. (2003), 

31 which did not observe any hydriding effects on assembly hardware after 15 years of dry storage. 

32 Therefore, hydriding of assembly hardware components is not considered to be significant, and 

33 aging management is not required during the 60-year timeframe.

34 3.6.2.3 General corrosion

35 Various assembly hardware components made of stainless steel or Inconel may be subjected to 

36 general corrosion in the presence of humid air or an aqueous solution. General corrosion of 

37 assembly hardware made of zirconium alloys is not considered here; it is excluded per the 

38 technical basis discussed in Section 3.6.1.6. The amount of residual water in the canister 

39 during the extended storage is expected to be less than 1 mole per the guidance in NUREG–

40 1536 (NRC, 2010). Most residual water is expected to decompose into hydrogen and oxidizing 

41 species, such as oxygen and hydrogen peroxide, with time (Jung et al., 2013). However, it is 

42 possible for trace amounts of water to remain in the vapor phase in the canister’s internal 

43 environment for the extended period.
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1 The general corrosion rate of the nickel-based Inconel alloys due to humid air is expected to be 

on the order of 25 nm/yr [10 2 ‒3 mils/yr] (Van Rooyen and Copson, 1968). The general 

3 corrosion rate of 304 stainless steel in the presence of humid air has been reported to be 

4 negligible (INCO, 1970), and the low-carbon grade 304L is expected to behave similarly. 

5 Further, as corrosion proceeds, the residual water would deplete with time. Considering the low 

6 general corrosion rate of the Inconel alloy, the negligible corrosion rate of 304 stainless steel 

7 under humid air conditions, and the radiolysis of the residual water, it is concluded that the 

8 effect of general corrosion in the presence of trace amounts of water is insignificant on 

9 assembly hardware components during the renewal period. As such, general corrosion of 

10 assembly hardware is considered to be insignificant, and therefore, aging management is not 

11 required during the 60-year timeframe.

12 3.6.2.4 Stress corrosion cracking

13 Various stainless steel and Inconel assembly hardware components could be susceptible to 

14 SCC in the presence of an aggressive environment and sufficient residual tensile stresses. 

15 SCC of the structural components may lead to cracking, which can compromise the structural 

16 integrity of the component. SCC of assembly hardware made of zirconium alloys is not 

17 considered here; it is excluded per the technical basis discussed in Section 3.6.1.9.

18 Residual tensile stresses are expected to be present in the assembly hardware, primarily in 

19 welded areas. Regarding the chemical environment, various types of stainless steels are prone 

20 to SCC, even in high-purity demineralized water at the temperatures of the BWRs, typically 

21 290 degrees C [554 degrees F] (Kain, 2011). This observation is attributed to the presence of 

22 dissolved oxygen and other oxidizing species in the primary coolant water (Kain, 2011) of a 

23 BWR. Various types of nickel-based alloys, including Inconel, are susceptible to SCC in the 

24 presence of hot water, hot caustic solution, hot wet hydrofluoric acid solution, or aqueous 

25 solution containing a sufficient amount of chloride at high temperatures (Rebak, 2011).

26 In the canister environment, the water could exist in the liquid state only when the temperature 

27 is near or below 100 degrees C [212 degrees F]. The residual water content inside the canister 

28 is expected to be less than 1 mole during dry storage, as per guidance in NUREG–1536 

29 (NRC, 2010). During storage, most residual water would decompose into hydrogen and 

30 oxidizing species, such as oxygen and hydrogen peroxide, due to radiolysis (Jung et al., 2013). 

31 However, it is possible for a trace amount of residual water to persist in the vapor phase of the 

32 containment cavity. The trace amount of water is unlikely to condense into the liquid phase 

33 during dry storage because the relative humidity of the DSS internal environment cannot reach 

34 100 percent when the residual amount of water is less than 1 mole. For example, the relative

humidity inside a containment cavity volume of 2.1 m3 35 [554.8 gal] at 25 degrees C 

36 [77 degrees F], assuming a residual water amount of 0.43 mole [expected after vacuum drying 

37 as per NUREG–1536], is estimated to be approximately 15 percent, using a backfill pressure of 

38 1 atm [14.7 psi], or 6 percent using a backfill pressure of 5 atm [73.5 psi] (Green and Perry, 

39 2007). Further, SCC of stainless steel and Inconel has not been reported in a nonchloride 

40 humid air environment. 

41 Because of the lack of halides and the small amount of water in helium and embedded 

42 environments, SCC of stainless steel is not considered to be credible. Therefore, aging 

43 management of SCC of stainless steel subcomponents exposed to helium is not required during 

44 the 60-year timeframe.
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1 3.6.2.5 Radiation embrittlement

2 Radiation embrittlement of assembly hardware such as guide tubes and spacer grid materials 

3 made from zirconium alloys is excluded using the basis provided in Section 3.6.1.10. Similarly, 

4 radiation embrittlement of assembly hardware made of stainless steel or Inconel is not 

5 considered credible per the technical bases provided in Sections 3.2.1.9, 0, and 0. Therefore, 

6 aging management of radiation embrittlement of assembly hardware subcomponents exposed 

7 to helium and embedded environments is not required during the 60-year timeframe. 

8 3.6.2.6 Fatigue

9 Fatigue of assembly hardware such as guide tubes and spacer grid materials made from 

10 zirconium alloys is excluded using the basis provided in Section 3.6.1.11. Similarly, fatigue of 

11 assembly hardware made of stainless steel or Inconel is not considered credible per the 

12 technical bases provided in Sections 3.2.1.7, 3.2.2.7, and 3.2.4.5. Therefore, aging 

13 management of fatigue of assembly hardware subcomponents exposed to helium is not 

14 required during the 60-year timeframe.

15 3.6.3 References

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2 Cladding in Spent-Fuel-Storage and Accident Situations.” 2004 International Meeting on LWR 

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4-1 

1 4 ANALYSIS OF DRY STORAGE SYSTEMS AND

2 SPENT FUEL ASSEMBLIES 

3 4.1 Introduction

4 This chapter provides (1) a brief description of selected storage system designs and (2) aging 

5 management tables for each design that identify the aging mechanisms and effects that must be 

6 managed to ensure that the functions of structures, systems, and components (SSCs) are 

7 maintained in the period of extended operation. The analyses in Chapter 3 provide the 

8 technical bases for those determinations. The aging management tables also identify the use of 

9 either a time-limited aging analysis (TLAA), aging management program (AMP), or other 

10 analysis to address the aging effects that require management.

11 The following system descriptions are for general information only. In the review of a renewal 

12 application, the technical reviewer should refer to the application, safety analysis report, and 

13 drawings to identify the SSCs within the scope of renewal and their functions, materials of 

14 construction, and operating environment. Table 4-1 describes the storage system designs that 

15 are discussed below and evaluated in the aging management tables. 

Table 4-1 Evaluated storage system designs

MAPS

Section No. Name

NRC Docket 

No.

Amendments 

Evaluated

4.2

Standardized NUHOMS®* 72-1004 1–11 and 13

Standardized Advanced NUHOMS® 72-1029 1 and 3

4.3

HI-STORM 100 72-1014 1–10

HI-STAR 100 72-1008 1–2

4.4

TN-32 72-1021 1

TN-68 72-1027 1

4.5

NAC-UMS 72-1015 1–5

NAC-MPC 72-1025 1–6

MAGNASTOR 72-1031 1–6

4.6 FuelSolutions 72-1026 1–4

4.7 Concrete Pad (generic) — —

4.8 Spent Fuel Assemblies (generic) — —

*The staff’s review of the Calvert Cliffs specific license renewal application (NRC, 2014) informed the evaluation of

the NUHOMS system, and thus the aging management tables for this system may include some unique elements 

of this site.

16

4-3

4.2 NUHOMS® 1 systems: Standardized and Standardized Advanced

2 4.2.1 System description

3 The NUHOMS family of modular storage systems provide for the horizontal storage of spent 

4 nuclear fuel (SNF) in a dry shielded canister (DSC) that is placed in a concrete horizontal 

5 storage module (HSM). Each NUHOMS system model type is designated by NUHOMS-XXY. 

6 The two digits (XX) refer to the number of fuel assemblies stored in the DSC, and the character 

7 (Y) designates the type of fuel being stored—P for pressurized-water reactor (PWR) or B for 

8 boiling-water reactor (BWR). For some systems, a fourth character (T) is added to designate 

9 that the DSC is also intended for transportation in packages approved under Title 10 of the 

10 Code of Federal Regulations (10 CFR) Part 71,“Packaging and Transportation of Radioactive 

11 Material.” Also, two additional characters, HB, are added for systems that are used to store 

12 high-burnup fuels (e.g., NUHOMS-24PHB).

13 The Standardized NUHOMS design is presently licensed for use in the United States under 

14 NRC Docket 72-1004, in combination with the 24P, 24PT2, 24PHB, 24PTH, 32PT, 32PTH1, 

15 37PTH, 52B, 61BT, 61BTH, and 69BTH DSCs, while the Standardized Advanced NUHOMS 

16 design is licensed for use under NRC Docket 72-1029, with the 24PT1, 24PT4, and 32PTH2 

17 DSCs. The principal components of the NUHOMS system include (i) a stainless steel DSC with 

18 an internal basket to hold SNF assemblies, (ii) a structural steel assemblage that supports the 

19 DSC, and (iii) an HSM that is constructed of reinforced concrete (see Figure 4-1). Additional 

20 components include an onsite transfer cask (TC) and other fuel transfer and auxiliary equipment 

21 used to support DSC loading and transfer operations. 

22 The Standardized Advanced system differs from the Standardized system in that it includes 

23 modifications to accommodate sites with high seismic levels, limited space, and needs for 

24 enhanced radiation shielding performance. To accomplish this, a modified version of the HSM 

25 was created, designated as the Advanced Horizontal Storage Modulue (AHSM). A brief 

26 summary of the components of the NUHOMS storage systems are provided below.

27 4.2.2 Dry shielded canister

28 The NUHOMS DSC is a welded stainless steel canister that uses redundant multipass closure 

29 welds. After fuel loading, draining and drying, the canister is backfilled with helium to provide an 

30 inert environment. Figure 4-2 and Figure 4-3 show the components of two DSC configurations, 

31 which comprise the shell assembly and the internal basket assembly.

32 Shell assembly

33 The DSC shell assembly consists of a stainless steel cylindrical shell that is joined to top and 

34 bottom end assemblies with double, redundant seal welds to form the confinement boundary. 

35 The bottom end assembly welds are made during fabrication of the DSC, while the top end 

36 assembly welds are made after fuel loading. The shell assembly also includes two shielding 

37 plugs at both ends for biological shielding. Siphon and vent ports penetrate the top shield plug 

38 and are sealed after DSC drying operations are complete. Figure 4-4 shows the pressure and 

39 confinement boundaries for the NUHOMS-32PT DSC.

40

41
4-4 

Figure 4-1 NUHOMS dry storage system (Pacific Nuclear Fuel Services, Inc., 1991)

1 

Figure 4-2 NUHOMS-24PT2 DSC assembly–spacer disk design (Transnuclear, 2004)

2 

3 
4-5 

Figure 4-3 NUHOMS-32PT DSC assembly–tube or plate design (Transnuclear, 2004)

1 

Figure 4-4 Pressure and confinement boundaries for NUHOMS-32PT DSC 

(Transnuclear, 2004)

2 
4-6 

1 Internal Basket Assembly

2 The internal basket assembly contains a storage position for each fuel assembly. The basket 

3 assembly may consist of an assemblage of spacer disc plates supported on vertical rods that 

4 extend the length of the DSC cavity (spacer disc design) or individual tubes or plates welded to 

5 form a grid-like structure (tube or plate design). 

6 The 24P, 24PT1, 24PT2, 24PT4, 24PHB, and 52B DSCs use the spacer disc basket design, as 

7 shown in Figure 4-2. Subcriticality is maintained through the geometric separation of the fuel 

8 assemblies by the DSC basket assembly and the neutron absorbing capability of the DSC 

9 materials of construction. The 52B DSC contains fixed neutron poison material for additional 

10 criticality control. 

11 The 61BT, 32PT, 24PTH, 61BTH, 32PTH1, 32PTH2, 69BTH, and 37PTH DSCs use the tube or 

12 plate grid basket design, as shown in Figure 4-3. Fixed neutron poison material provides the 

13 necessary criticality control. Aluminum sheets or plates are used to provide the heat conduction 

14 paths from the fuel assemblies to the canister shell. Transition rails, consisting of welded 

15 stainless steel plates or aluminum parts, form the transition between the box-like fuel 

16 compartment structure and the cylindrical DSC shell.

17 Table 4-2 and Table 4-3 evaluate potential aging mechanisms and effects requiring 

18 management for specific components of the Standardized and Standardized Advanced 

19 NUHOMS DSC shell and basket designs. The tables also identify AMPs that provide an 

20 acceptable approach to managing the aging effects.

21 4.2.3 Horizontal storage module

22 Both the HSM and AHSM storage modulus are low-profile structures constructed from 

23 reinforced concrete and structural steel that provides a means for passive removal of spent fuel 

24 decay heat, structural support and environmental protection of the DSC, and radiation shielding. 

25 The AHSM design is similar to the HSM; however, the AHSM contains improved shielding and 

26 resistance to high seismic events. The AHSM consists a base storage unit and a top shield 

27 block that is tied to the base unit by steel rods in the vertical direction and interlocking concrete 

28 keys in the horizontal direction.

29 Heat removal is achieved by a combination of radiation, conduction, and convection. As shown 

30 in Figure 4-5 and Figure 4-6, ambient air enters the HSMs through ventilation inlet openings 

31 located in the lower region of the front or side walls and circulates around the DSC. Air exits 

32 through outlet openings in the top regions of the HSM walls. Thermal monitoring or visual 

33 inspections are used to provide indication of HSM performance or a blocked vent condition. 

34 Environmental protection and radiation shielding are provided by the thick side walls and roof of 

35 the HSM, supplemented by thick wall units attached at the ends of the array and at the rear 

36 walls of the HSM if the array is of single row configuration. Each HSM has an access opening 

37 or docking flange in the front wall to accommodate transfer of DSCs from and into the shielded 

38 TC. The access opening is covered by a thick shielded access door.

39

40
4-7 

Figure 4-5 Air flow diagram for a typical standardized HSM design 

(Pacific Nuclear Fuel Services, Inc., 1991)

1 

Figure 4-6 Advanced NUHOMS horizontal storage module (AHSM) 

(Transnuclear, 2003)

2 

3 
4-8 

1 Structural support of the loaded DSC is provided by a structural steel frame structure 

2 (HSM model 80 and model 102) anchored to the floor slab and walls of the HSM, or a structural 

3 steel rail assembly (HSM models HSM-H, -152, -202, HSM-HS, AHSM, and AHSM-HS). 

4 Figure 4-7 shows drawings of the side elevation and end view of the DSC rail assembly. 

5 Stainless steel cover plates coated with a dry film lubricant are attached to the rails to provide a 

6 sliding surface for DSC insertion and retrieval. In some designs, Nitronic 60 plates are welded 

7 to the cover plates because of this material’s good high-temperature properties and resistance 

8 to oxidation, wear, and galling. Seismic restraints using steel plates or tubes are welded to the 

9 rear and front of the rails for retaining the DSC in place during seismic events.

10 Table 4-4 and Table 4-5 provide a generic evaluation of potential aging mechanisms and effects 

11 requiring management for specific components of the Standardized NUHOMS HSM and 

12 Standardized Advanced NUHOMS AHSM. The tables also identify the AMPs that provide an 

13 acceptable approach to managing the effects of aging.

Figure 4-7 Side elevation and end view of the DSC support structure 

(Transnuclear, 2004)

14 4.2.4 Transfer cask

15 The NUHOMS TC is a cylindrical vessel with a bolted top cover plate and a welded bottom end 

16 assembly (Transnuclear, 2014). There are five alternate configurations of the cask. 

17 • The basic configuration, where the TC is provided with a solid neutron shield, is denoted 

18 as the standardized onsite cask. 

19 • A second configuration includes the OS197 and OS197H (H: modified for increased 

20 strength), in which water is used to provide neutron shielding. 

21 • The third configuration, designated as OS197FC, OS197HFC OS197FC-B, or 

22 OS197HFC-B TC, is equipped with a modified top lid to allow air circulation through the 

23 annulus between the DSC and the TC. 

24 • The fourth configuration, designated as OS197L TC and shown in Figure 4-8, is a 

25 reduced weight version of the OS197 TC. 
4-9 

Figure 4-8 OS197L transfer cask (Transnuclear, 2008)

1 • The fifth configuration is designated as OS200 or OS200FC TC and has a larger 

2 diameter to accommodate the larger diameter DSCs with 32PTH1, 37PTH, or 69BTH 

3 SNF assemblies.

4 For all the configurations except the OS197L TC, the TCs are constructed from two concentric 

5 cylindrical shells: a stainless steel inner shell and a structural shell made of stainless steel or 

6 carbon steel. The annulus formed by these two shells is filled with cast lead to provide gamma 

7 shielding. The TC also includes an outer jacket made of stainless steel or carbon steel, which is 

8 filled with BISCO NS-3 material or water for neutron shielding. The inner and structural shells 

9 are welded to heavy forged ring assemblies at the top and bottom ends. The bottom end plate 

10 has a removable stainless steel ram access penetration ring. A stainless steel bottom cover 

11 plate is provided to seal the hydraulic ram access penetration of the cask during fuel loading. 

12 Rails fabricated from a nongalling, wear-resistant stainless steel coated with a high contact 

13 pressure dry film lubricant are provided to facilitate DSC transfer.

14 The OS197L TC is constructed from a single, thicker stainless steel structural shell. To 

15 compensate for the lack of lead shielding, the OS197L TC relies on the use of supplemental 

16 shielding in conjunction with remote operations during handling in the fuel or reactor building, 

17 transfer to the ISFSI, and insertion into the HSM operations. The cask support skid 

18 supplemental shielding consists of a thick carbon steel upper shielding bell and a lower 

19 shielding sleeve that enclose the TC in the decontamination area, and thick carbon steel plates 

20 and covers that enclose the TC while on the transfer trailer.

21 The NUHOMS TCs have four trunnions made of stainless steel or nickel alloy that are welded to 

22 the structural shell. Two upper lifting trunnions are located near the top of the cask for lifting the 

23 cask in the SNF pool building. The lower trunnions, located near the base of the cask, serve as 

24 the axis of rotation and as supports during transport to the HSM.

25 Table 4-6 provides a generic evaluation of potential aging mechanisms and effects requiring 

26 management for specific components of the NUHOMS transfer casks. The table also identifies 

27 the AMPs that provide an acceptable approach to managing the aging effects.
4-10

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Guide sleeves 

(DSC basket)

CR, SR, 

TH*

Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Oversleeves 

(DSC basket)

CR, SR, TH Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Aluminum plate or 

sheet, basket 

plate, compartment 

plate (DSC basket)

CR, SH, TH Aluminum Helium Thermal aging Loss of strength No 

Creep Change in

dimensions

No 

General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Spacer disks 

(DSC basket)

CR, SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

Dimensions

No 

Radiation

embrittlement

Cracking No 

Steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-11

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Spacer disks 

(DSC basket)

CR, SR Steel Helium General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Support rods 

(DSC basket)

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Stainless steel 

(welded 17-4 

PH)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

TLAA/AMP or a 

supporting analys

required

Stainless steel 

(17-4 PH)

Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Spacer sleeves

(DSC basket)

CR, SR Stainless steel 

(welded 17-4 

PH)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

TLAA/AMP or a 

supporting analys

required

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-12

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Basket rails 

(DSC basket)

CR, SH, 

SR, TH

Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Basket rail inserts 

and shims 

(DSC basket)

SR, TH Aluminum Helium Thermal aging Loss of strength TLAA/AMP or a 

supporting analys

required

Creep Change in

dimensions

TLAA/AMP or a 

supporting analys

required

General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Basket assembly 

plates (DSC 

basket)

CR, SH, 

SR, TH

Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Grid assembly 

(DSC basket)

SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-13

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Holddown ring 

assembly and 

plates (DSC 

basket)

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Fuel compartment 

tubes, wraps, 

inserts (DSC 

basket)

CR, SH, 

SR, TH

Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Transition rails 

(DSC basket)

CR, SH, 

SR, TH

Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Aluminum Helium Thermal aging Loss of strength TLAA/AMP or a 

supporting analys

required

Creep Change in

dimensions

TLAA/AMP or a 

supporting analys

required

General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 
4-14

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Neutron absorbing 

plates, poison 

plates (DSC 

basket)

CR, TH Borated 

stainless steel

Helium Boron depletion Loss of 

criticality control

No; a TLAA may b

required

Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Loss of fracture

toughness and 

loss of ductility

No 

Neutron absorbing 

plates or sheets, 

poison plates, 

chevron neutron 

absorbers (DSC 

basket)

CR, SH, TH Boralyn®, 

MetamicTM

Helium Boron depletion Loss of 

criticality control

No; a TLAA may b

required

Thermal aging Loss of strength No 

General 

corrosion

Loss of material No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Loss of fracture

toughness and 

loss of ductility

No 

Boral® Helium Boron depletion Loss of 

criticality control

No; a TLAA may b

required

Thermal aging Loss of strength No 

Wet corrosion 

and blistering

Change in

dimensions

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Loss of fracture

toughness and 

loss of ductility

No 
4-15

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Neutron absorbing 

plates or sheets, 

poison plates, 

chevron neutron 

absorbers (DSC 

basket)

CR, SH, TH Borated 

aluminum

Helium Boron depletion Loss of 

criticality control

No; a TLAA may b

required

Thermal aging Loss of strength No 

General 

corrosion

Loss of material No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Loss of fracture

toughness and 

loss of ductility

No 

Support bars (DSC 

basket)

SR Steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Fastener 

components

SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

General 

corrosion

Loss of material No 
4-16

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Fastener 

components

SR Steel Helium Radiation

embrittlement

Cracking No 

Tool socket and 

closure plate

SR Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Components for 

damaged fuel

CO, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Cover plates 

(inner)

CO, SH, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

Dimensions

No 

Radiation

embrittlement

Cracking No 

Shield plug (top) CO, SH, 

SR, TH

Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-17

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Shield plug (top) CO, SH, 

SR, TH

Steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Shield plug 

(bottom)

CO, SH, 

SR, TH

Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

General 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Lead shielding SH Lead Embedded 

(steel, stainless 

steel)

None identified None identified No 
4-18

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Siphon and vent 

block

CO, SH, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Siphon and vent 

port cover plate

CO, SH, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Test port plug CO Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design c

TLAA, if applicabl

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Key, shear key SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-19

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Pin, anti-rotation 

pin

SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

DSC support ring SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

Dimensions

No 

Radiation

embrittlement

Cracking No 

Lifting lugs SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

DSC shell CO, SH, 

SR, TH

Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corrosi

and Stress corros

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of material

(precursor to 

stress 

corrosion 

cracking)

Localized Corrosi

and Stress corros

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP
4-20

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

DSC shell CO, SH, 

SR, TH

Stainless steel Sheltered Galvanic 

corrosion

Loss of material Localized Corrosi

and Stress corros

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP

Microbiologically 

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate design c

TLAA, if applicabl

Radiation 

embrittlement

Cracking No 

Helium Radiation 

embrittlement

Cracking No 

Cover plates 

(outer)

CO, SH, 

SR, TH

Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corrosi

and Stress corros

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of material

(precursor to 

stress 

corrosion 

cracking)

Localized Corrosi

and Stress corros

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate design c

TLAA, if applicabl

Radiation

embrittlement

Cracking No 
4-21

Table 4-2 Standardized NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Grapple ring and 

grapple support

SR Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corrosi

and Stress corros

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of material

(precursor to 

stress 

corrosion 

cracking)

Localized Corrosi

and Stress corros

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-22

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Guide sleeves 

(DSC basket)

CR, SR, 

TH*

Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Oversleeves, stop 

plates (DSC 

basket)

CR, SR, TH Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Spacer disks (DSC 

basket)

CR, SR Steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

General 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Support rods (DSC 

basket)

CR, SR Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in

dimensions

No 

Radiation 

embrittlement

Cracking No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-23

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Support rods (DSC

basket)

CR, SR Stainless steel 

(welded 17-4 

PH)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

TLAA/AMP or a 

supporting analysi

required

Stainless steel 

(17-4 PH)

Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Spacer sleeves 

(DSC basket)

CR, SR Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Stainless 

steel (welded 

17-4 PH)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

TLAA/AMP or a 

supporting analysi

required

Stainless 

steel (17-4 

PH)

Helium Creep Change in

dimensions

No 

Radiation 

embrittlement

Cracking No 

Shims (DSC 

basket)

SR, TH Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Aluminum Helium Thermal aging Loss of 

strength

TLAA/AMP or a 

supporting analysi

required
4-24

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Shims (DSC 

basket)

SR, TH Aluminum Helium Creep Change in 

dimensions

TLAA/AMP or a 

supporting analysi

required

General 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Basket assembly 

plates (DSC 

basket)

CR, SH, 

SR, TH

Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Aluminum Helium Thermal aging Loss of 

strength

TLAA/AMP or a 

supporting analysi

required

Creep Change in 

dimensions

TLAA/AMP or a 

supporting analysi

required

General 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Fuel compartment 

(DSC basket)

CR, SH, 

SR, TH

Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 
4-25

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Transition rails 

(DSC basket)

CR, SH, 

SR, TH

Aluminum Helium Thermal aging Loss of 

strength

TLAA/AMP or a

supporting analysi

required

Creep Change in 

dimensions

TLAA/AMP or a

supporting analysi

required

General 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Neutron absorbing 

poison plate (DSC 

basket)

CR, SH, TH Boron 

carbide/ 

aluminum 

metal-matrix 

it

Helium Boron depletion Loss of 

criticality 

control

No; a TLAA may b

required

Thermal aging Loss of 

strength

No 

General 

corrosion

Loss of 

material

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Loss of fracture 

toughness and 

loss of ductility

No 

Neutron absorbing 

sheets (DSC 

basket)

CR, SH, TH Boral® Helium Boron depletion Loss of 

criticality 

control

No; a TLAA may b

required

Thermal aging Loss of 

strength

No 

Wet corrosion 

and blistering

Change in 

dimensions

No 
4-26

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Neutron absorbing 

sheets (DSC 

basket)

CR, SH, TH Boral® Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Loss of fracture 

toughness and 

loss of ductility

No 

Fastener 

components

SR Steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

General 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Components for 

damaged fuel

CO, SR Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Cover plates 

(inner)

CO, SH, SR Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 
4-27

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Cover plates 

(inner)

CO, SH, SR Stainless 

steel

Helium Radiation 

embrittlement

Cracking No 

Shield plug (top) CO, SH, 

SR, TH

Stainless 

steel

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 

General 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Shield plug casing 

(bottom)

CO, SH, 

SR, TH

Stainless 

steel (welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corrosio

and Stress corrosi

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP
4-28

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Shield plug casing 

(bottom)

CO, SH, 

SR, TH

Stainless 

steel

Sheltered Pitting and 

crevice corrosion

Loss of 

material 

(Precursor to 

stress 

corrosion 

cracking)

Localized Corrosio

and Stress corrosi

Cracking of Welde

Stainless Steel Dr

Storage Canisters 

AMP

Microbiologically 

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design co

TLAA, if applicable

Radiation 

embrittlement

Cracking No 

Embedded 

(lead)

Radiation 

embrittlement

Cracking No 

Shield plug 

(bottom)

CO, SH, 

SR, TH

Steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 

General 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Lead shielding SH Lead Embedded 

(stainless steel)

None identified None identified No 

Siphon and vent 

block

CO, SH, SR Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 
4-29

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Siphon and vent 

block

CO, SH, SR Stainless 

steel

Helium Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Siphon and vent 

port cover plate

CO, SH, SR Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Test port plug CO Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Fatigue Cracking Evaluate design co

TLAA, if applicable

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Pin SR Stainless 

steel

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 
4-30

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

DSC support ring SR Stainless 

steel (welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless 

steel

Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

DSC shell CO, SH, 

SR, TH

Stainless 

steel (welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corrosion 

and Stress Corrosio

Cracking of Welded 

Stainless Steel Dry 

Storage Canisters A

Stainless 

steel

Sheltered Pitting and 

crevice corrosion

Loss of 

material 

(Precursor to 

stress 

corrosion 

cracking)

Localized Corrosion 

and Stress Corrosio

Cracking of Welded 

Stainless Steel Dry 

Storage Canisters A

Galvanic 

corrosion

Loss of 

material

Localized Corrosion 

and Stress Corrosio

Cracking of Welded 

Stainless Steel Dry 

Storage Canisters A

Microbiologically 

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design co

TLAA, if applicable

Radiation 

embrittlement

Cracking No 

Helium Radiation 

embrittlement

Cracking No 
4-31

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Cover plates 

(outer)

CO, SH, 

SR, TH

Stainless 

steel (welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corrosion 

and Stress corrosio

Cracking of Welded 

Stainless Steel Dry 

Storage Canisters A

Stainless 

steel

Sheltered Pitting and 

crevice corrosion

Loss of 

material 

(Precursor to 

stress 

corrosion 

cracking)

Localized Corrosion 

and Stress corrosio

Cracking of Welded 

Stainless Steel Dry 

Storage Canisters A

Microbiologically 

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design co

TLAA, if applicable

Radiation 

embrittlement

Cracking No 

Grapple ring and 

grapple support

SR Stainless 

steel (welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corrosion 

and Stress corrosio

Cracking of Welded 

Stainless Steel Dry 

Storage Canisters A

Stainless 

steel

Sheltered Pitting and 

crevice corrosion

Loss of 

material 

(Precursor to 

stress 

corrosion 

cracking)

Localized Corrosion 

and Stress corrosio

Cracking of Welded 

Stainless Steel Dry 

Storage Canisters A

Microbiologically 

influenced 

corrosion

Loss of 

material

No 
4-32

Table 4-3 Standardized Advanced NUHOMS dry shielded canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Managem

Grapple ring and 

grapple support

SR Stainless 

steel

Sheltered Radiation 

embrittlement

Cracking No 
4-33

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Reinforced concrete: 

base walls, floor slab, 

roof; basemat; end and 

rear shield walls, corner 

shield wall; shielded 

ventilation air inlet 

plenum; inlet/outlet vents

SH, SR, 

TH*

Concrete Air—outdoor Aggressive 

chemical attack

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Loss of material

(spalling, scaling)

Reinforce

Structures 

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Creep Cracking No 

Dehydration at 

high temperature

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Structures 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforce

Structures 

Loss of material

(spalling, scaling)

Reinforce

Structures 

Radiation damage Cracking No 

Loss of strength No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-34

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Reinforced concrete: 

base walls, floor slab, 

roof; basemat; end and 

rear shield walls, corner 

shield wall; shielded 

ventilation air inlet 

plenum; inlet/outlet vents

SH, SR, 

TH

Concrete Air—outdoor Reaction with

aggregates

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Salt scaling Loss of material

(spalling, scaling)

Reinforce

Structures 

Shrinkage Cracking No 

Leaching of 

calcium

hydroxide

Loss of strength Reinforce

Structures 

Increase in porosity

and permeability

Reinforce

Structures 

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Sheltered Aggressive 

chemical attack

Loss of strength No 

Cracking No 

Loss of material

(spalling, scaling)

No 

Creep Cracking No 

Dehydration at 

high temperature

Cracking No 

Loss of strength No 

Delayed ettringite 

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Structures 

Fatigue Cracking No 
4-35

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Reinforced concrete: 

base walls, floor slab, 

roof; basemat; end and 

rear shield walls, corner 

shield wall; shielded 

ventilation air inlet 

plenum; inlet/outlet vents

SH, SR, 

TH

Concrete Sheltered Freeze and thaw Cracking No 

Loss of material

(spalling, scaling)

No 

Radiation damage Cracking No 

Loss of strength No 

Reaction with

aggregates

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Salt scaling Loss of material

(spalling, scaling)

No 

Shrinkage Cracking No 

Groundwater/soil Aggressive 

chemical attack

Cracking Reinforce

Structures 

Loss of material

(spalling, scaling)

Reinforce

Structures 

Loss of strength Reinforce

Structures 

Reduction of

concrete pH 

(reducing 

corrosion 

resistance of steel 

embedments)

Reinforce

Structures 
4-36

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Reinforced concrete: 

base walls, floor slab, 

roof; basemat; end and 

rear shield walls, corner 

shield wall; shielded 

ventilation air inlet 

plenum; inlet/outlet vents

SH, SR, 

TH

Concrete Groundwater/soil Creep Cracking No 

Dehydration at 

high temperature

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Structures 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforce

Structures 

Loss of material

(spalling, scaling)

Reinforce

Structures 

Microbiological

degradation

Loss of strength Reinforce

Structures 

Loss of material

(spalling, scaling)

Reinforce

Structures 

Increase in porosity

and permeability

Reinforce

Structures 

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Radiation damage Cracking No 

Loss of strength No 
4-37

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Reinforced concrete: 

base walls, floor slab, 

roof; basemat; end and 

rear shield walls, corner 

shield wall; shielded 

ventilation air inlet 

plenum; inlet/outlet vents

SH, SR, 

TH

Concrete Groundwater/soil Reaction with

aggregates

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Salt scaling Loss of material

(spalling, scaling)

Reinforce

Structures 

Shrinkage Cracking No 

Leaching of 

calcium

hydroxide

Loss of strength Reinforce

Structures 

Increase in porosity

and permeability

Reinforce

Structures 

Leaching of 

calcium

hydroxide

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Reinforcing 

steel

Air—outdoor; 

groundwater

Corrosion of

reinforcing steel

Loss of

concrete/steel bond

Reinforce

Structures 

Loss of material

(spalling, scaling)

Reinforce

Structures 

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

DSC support structure

assembly hardware, 

base unit assembly 

hardware, module 

accessories

SR Steel Sheltered Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitoring 

Compone
4-38

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

DSC support structure

assembly hardware, 

base unit assembly 

hardware, module 

accessories

SR Steel Sheltered Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic C

AMP

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload External S

Monitorin

Metallic

Compone

DSC support structure 

assembly: support rail, 

rail extension plate and 

rail baseplate, plates, 

crossbeam, DSC stop 

plate extension

SR, TH Stainless 

steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking External S

Monitorin

Metallic

Compone

Stainless 

steel

Sheltered Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material

(precursor to stress 

corrosion cracking)

External S

Monitorin

Metallic C

AMP

Radiation

embrittlement

Cracking No 
4-39

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

DSC support structure 

assembly: support rail 

beams, support structure 

miscellaneous steel, 

plates, attachment/ 

installation hardware, 

DSC stop plate 

assembly, rail extension 

embedment, tube steel 

leg column

SR, TH Steel Sheltered Galvanic corrosion Loss of material External S

Monitorin

Metallic

Compone

General corrosion Loss of material External S

Monitorin

Metallic

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic

Compone

Radiation

embrittlement

Cracking No 

DSC support structure

assembly: support 

rail plate

SR, TH Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking External S

Monitorin

Metallic

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material

(precursor to 

stress corrosion

cracking)

External S

Monitorin

Metallic

Compone

Radiation

embrittlement

Cracking No 
4-40

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

DSC axial retainer

assembly: axial retainer, 

plate

SR Steel Sheltered General corrosion Loss of material External S

Monitorin

Metallic

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic

Compone

Radiation

embrittlement

Cracking No 

Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically

Influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Cask restrain assembly: 

embedment assembly 

(rods, hex nuts, sleeve 

nuts), cask restraint 

embedment

SR Steel Sheltered Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitorin

Metallic

Compone

Microbiologically

influenced 

corrosion

Loss of material No 
4-41

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Cask restrain assembly: 

embedment assembly 

(rods, hex nuts, sleeve 

nuts), cask restraint 

embedment

SR Steel Sheltered Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload External S

Monitorin

Metallic

Compone

Heat shield assemblies: 

attachment hardware

SR Steel Sheltered Stress corrosion 

cracking

Cracking No 

Galvanic corrosion Loss of material External S

Monitorin

Metallic 

Compone

General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload External S

Monitorin

Metallic 

Compone
4-42

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Heat shield assemblies: 

support structure, 

Z bracket, screw

SR Stainless 

Steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Heat shield assemblies: 

roof and side wall

mounted heat shields, Z 

bracket

TH Steel 

(galvanized)

Sheltered Galvanic corrosion Loss of material External S

Monitorin

Metallic 

Compone

General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 
4-43

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Heat shield assemblies: 

roof and side wall 

mounted heat shields/Z 

bracket, side heat shield 

fins, backing sheet, top 

louvered heat shield

TH Aluminum Sheltered Galvanic corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Heat shield assemblies: 

side heat shield, top heat 

shield

TH Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Shielded door assembly: 

door attachment 

hardware

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 
4-44

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Shielded door assembly: 

door attachment 

hardware

SR Steel Air—outdoor Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Shielded door 

assembly: steel plates

SH, SR Steel Air—outdoor General corrosion Loss of material External S

Monitorin

Metallic

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic

Compone

Radiation

embrittlement

Cracking No 

Sheltered General corrosion Loss of material External S

Monitorin

Metallic

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic

Compone

Radiation

embrittlement

Cracking No 
4-45

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Shielded door assembly: 

concrete core

SH, SR Reinforced 

concrete, 

nonshrink 

grout or pea 

gravel or 

mortar mix

Fully encased 

(steel)

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Cracking No 

Loss of strength No 

Radiation damage Cracking TLAA/AM

supportin

is require

Radiation damage Loss of strength TLAA/AM

supportin

is require

Reaction with

aggregates

Cracking Reinforce

Concrete 

AMP

Loss of strength Reinforce

Concrete 

AMP

Inlet/outlet vents: outlet 

vent attachments

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-46

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Inlet/outlet vents: liner 

plates

SH, TH Steel Air—outdoor General corrosion Loss of material External S

Monitorin

Metallic C

AMP

Inlet/outlet vents: liner 

plates

SH, TH Steel Air—outdoor Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic C

AMP

Radiation

embrittlement

Cracking No 

Shielded ventilation air 

inlet plenum

TH Stainless 

steel 

(welded)

Air—outdoor Stress corrosion 

cracking

Cracking External S

Monitorin

Metallic C

AMP

Stainless 

steel

Air—outdoor Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material

(precursor to stress 

corrosion cracking)

External S

Monitorin

Metallic C

AMP

Radiation

embrittlement

Cracking No 

Ventilation air outlet 

shielding blocks

TH Stainless 

steel 

(welded)

Air—outdoor Stress corrosion 

cracking

Cracking External S

Monitorin

Metallic C

AMP

Stainless 

Steel

Air—outdoor Microbiologically

influenced 

corrosion

Loss of material No 
4-47

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Ventilation air outlet 

shielding blocks

TH Stainless 

Steel

Air—outdoor Pitting and crevice

corrosion

Loss of material

(precursor to 

stress corrosion 

cracking)

External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Roof attachment 

assembly: angles, 

plates, dowel bar splicer

SR Steel Sheltered General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Roof attachment 

assembly: roof 

attachment hardware

SR Steel Sheltered Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic

Compone

Radiation

embrittlement

Cracking No 
4-48

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Roof attachment 

assembly: roof 

attachment hardware

SR Steel Sheltered Stress relaxation Loss of preload External S

Monitorin

Metallic

Compone

End and rear shield walls 

attachment hardware

SR Steel Sheltered Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload External S

Monitorin

Metallic 

Compone

HSM-to-HSM spacer

channels

SR Steel Air—outdoor General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone
4-49

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

HSM-to-HSM spacer

channels

SR Steel Air—outdoor Radiation

embrittlement

Cracking No 

Dose reduction 

hardware: dose 

reduction assembly

SH Steel Air—outdoor Galvanic corrosion Loss of material External S

Monitorin

Metallic 

Compone

General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Stainless 

steel

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Module-to-module

connections

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitorin

Metallic 

Compone
4-50

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Module-to-module

connections

SR Steel Air—outdoor Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Lightning protection 

system

SR Copper Air—outdoor General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Threaded fasteners and

expansion anchors

SH, TH Stainless 

Steel 

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 
4-51

Table 4-4 Standardized NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Mana

Threaded fasteners and

expansion anchors

SH, TH Stainless 

Steel 

Air—outdoor Stress relaxation Loss of preload No 

Handrail SR Steel Air—outdoor General corrosion Loss of material External S

Monitorin

Metallic 

Compone

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External S

Monitorin

Metallic 

Compone

Radiation

embrittlement

Cracking No 
4-52

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Reinforced concrete: 

base unit walls, floor 

slab, roof, top shield 

block, basemat, end and 

rear shield walls, corner 

shield wall, inlet/outlet 

vents

SH, SR, 

TH*

Concrete Air—outdoor Aggressive 

chemical attack

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Loss of material 

(spalling, scaling)

Reinforce

Structures 

Reduction of 

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Creep Cracking No 

Dehydration at 

high temperatures

Cracking No 

Loss of strength No 

Delayed ettringite 

formation

Loss of material 

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Structures 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforce

Structures 

Loss of material 

(spalling, scaling)

Reinforce

Structures 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-53

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Reinforced concrete: 

base unit walls, floor 

slab, roof, top shield 

block, basemat, end and 

rear shield walls, corner 

shield wall, inlet/outlet 

vents

SH, SR, TH Concrete Air—outdoor Radiation damage Cracking No 

Loss of strength No 

Reaction with 

aggregates

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Salt scaling Loss of material 

(spalling, scaling)

Reinforce

Structures 

Shrinkage Cracking No 

Leaching of 

calcium hydroxide

Loss of strength Reinforce

Structures 

Increase in porosity 

and permeability

Reinforce

Structures 

Reduction of 

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Sheltered Aggressive 

chemical attack

Loss of strength No 

Cracking No 

Loss of material 

(spalling, scaling)

No 

Creep Cracking No 
4-54

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Reinforced concrete: 

base unit walls, floor 

slab, roof, top shield 

block, basemat, end and 

rear shield walls, corner 

shield wall, inlet/outlet 

vents

SH, SR, TH Concrete Sheltered Dehydration at 

high temperatures

Cracking No 

Loss of strength No 

Delayed ettringite 

formation

Loss of material 

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Structures 

Fatigue Cracking No 

Freeze and thaw Cracking No 

Loss of material 

(spalling, scaling)

No 

Radiation damage Cracking No 

Loss of strength No 

Reaction with 

aggregates

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Salt scaling Loss of material 

(spalling, scaling)

No 

Shrinkage Cracking No 
4-55

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Reinforced concrete: 

base unit walls, floor 

slab, roof, top shield 

block, basemat, end and 

rear shield walls, corner 

shield wall, inlet/outlet 

vents

SH, SR, TH Concrete Groundwater/soil Aggressive 

chemical attack

Cracking Reinforce

Structures 

Loss of material 

(spalling, scaling)

Reinforce

Structures 

Loss of strength Reinforce

Structures 

Reduction of 

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Creep Cracking No 

Dehydration at 

high temperatures

Cracking No 

Loss of strength No 

Delayed ettringite 

formation

Loss of material 

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Structures 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforce

Structures 

Loss of material 

(spalling, scaling)

Reinforce

Structures 
4-56

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Reinforced concrete: 

base unit walls, floor 

slab, roof, top shield 

block, basemat, end and 

rear shield walls, corner 

shield wall, inlet/outlet 

vents

SH, SR, TH Concrete Groundwater/

soil

Microbiological 

degradation

Loss of strength Reinforce

Structures 

Loss of material 

(spalling, scaling)

Reinforce

Structures 

Increase in porosity 

and permeability

Reinforce

Structures 

Reduction of 

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Radiation damage Cracking No 

Loss of strength No 

Reaction with 

aggregates

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Salt scaling Loss of material 

(spalling, scaling)

Reinforce

Structures 

Shrinkage Cracking No 

Leaching of 

calcium hydroxide

Loss of strength Reinforce

Structures 

Increase in porosity 

and permeability

Reinforce

Structures 
4-57

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Reinforced concrete: 

base unit walls, floor 

slab, roof, top shield 

block, basemat, end and 

rear shield walls, corner 

shield wall, inlet/outlet 

vents

SH, SR, TH Reinforcing 

steel

Air—outdoor, 

groundwater

Leaching of 

calcium hydroxide

Reduction of 

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforce

Structures 

Corrosion of 

reinforcing steel

Loss of 

concrete/steel bond

Reinforce

Structures 

Loss of material 

(spalling, scaling)

Reinforce

Structures 

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

DSC support structure 

assembly hardware

SR Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-58

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

DSC support structure 

assembly: support rail, 

rail extension plate, rail 

baseplate, stiffener plate, 

gusset plate, crossbeam, 

DSC stop plate 

extension

SR, TH Stainless 

steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking External S

Monitoring 

Metallic 

Compone

Stainless 

steel

Sheltered Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material 

(precursor to stress 

corrosion cracking)

External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

DSC axial retainer 

assembly: axial retainer

SR Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

DSC axial retainer 

assembly: axial retainer, 

plates

SR Steel Sheltered General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 
4-59

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

DSC axial retainer 

assembly: axial retainer, 

plates

SR Steel Sheltered Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Heat shield assemblies: 

attachment hardware

SR Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Steel Sheltered Stress corrosion 

cracking

Cracking No 

Galvanic corrosion Loss of material External S

Monitoring 

Metallic 

Compone

General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 
4-60

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Heat shield assemblies: 

attachment hardware

SR Steel Sheltered Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload External S

Monitoring 

Metallic 

Compone

Heat shield assemblies: 

side heat shield, top heat 

shield

TH Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Shielded door assembly: 

door attachment 

hardware

SR Stainless 

steel

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-61

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Shielded door assembly: 

door attachment 

hardware

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Shielded door assembly: 

backing plates

SR Stainless 

steel

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 
4-62

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Shielded door assembly: 

backing plates

SR Stainless 

steel

Sheltered Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Shielded door assembly: 

plates

SH, SR Steel Air—outdoor General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Sheltered General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 
4-63

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Shielded door assembly: 

concrete core

SH, SR Reinforced 

concrete, 

non-shrink 

grout or pea 

gravel or 

mortar mix

Embedded

(steel or 

stainless steel)

Delayed ettringite 

formation

Loss of material 

(spalling, scaling)

No 

Cracking No 

Loss of strength No 

Radiation damage Cracking TLAA/AM

supporting 

is required

Loss of strength TLAA/AM

supporting 

is required

Reaction with 

aggregates

Cracking Reinforce

Structures 

Loss of strength Reinforce

Structures 

Inlet/outlet vents: outlet 

vent attachment 

hardware

SR Stainless 

steel

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-64

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Inlet/outlet vents: outlet 

vent attachment 

hardware

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitoring 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Compone

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Inlet/outlet vents: liner 

plates

SH, TH Stainless 

steel

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material

(precursor to stress 

corrosion cracking)

No 

Radiation 

embrittlement

Cracking No 

Roof attachment 

assembly: angles, plates, 

dowel bar splicer

SR Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 
4-65

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Roof attachment 

assembly: angles, plates, 

dowel bar splicer

SR Stainless 

steel

Sheltered Pitting and crevice 

corrosion

Loss of material

(precursor to stress 

corrosion cracking)

No 

Radiation 

embrittlement

Cracking No 

Steel Sheltered General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Roof attachment 

assembly: roof 

attachment hardware

SR Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-66

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Roof attachment 

assembly: roof 

attachment hardware

SR Stainless 

steel

Sheltered Stress relaxation Loss of preload No 

Steel Sheltered Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload External S

Monitoring 

Metallic 

Compone

End and rear shield walls 

attachment hardware

SR Stainless 

steel

Sheltered Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material No 
4-67

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

End and rear shield walls 

attachment hardware

SR Stainless 

steel

Sheltered Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Steel Sheltered Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External S

Monitoring 

Metallic 

Compone

Microbiologically 

influenced 

corrosion

Loss of material No 

Pitting and crevice 

corrosion

Loss of material External S

Monitoring 

Metallic 

Compone

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload External S

Monitoring 

Metallic 

Compone

Module-to-module 

connection hardware

SR Stainless 

steel

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically 

influenced 

corrosion

Loss of material No 
4-68

Table 4-5 Standardized Advanced NUHOMS horizontal storage module

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Ma

Module-to-module 

connection hardware

SR Stainless 

steel

Air—outdoor Pitting and crevice 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-69

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Structural shell (Cask 

body)

SH, SR, TH* Steel Embedded 

(neutron shielding)

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Stainless steel 

(welded)

Demineralized 

water

Stress corrosion

cracking

Cracking No 

Air —

indoor/outdoor

Stress corrosion

cracking

Cracking No 

Stainless steel Embedded 

(neutron shielding)

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-70

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Structural shell (Cask 

body)

SH, SR, TH Stainless steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Inner shell (Cask body) SH, SR, TH Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Embedded (Lead) Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Top flange (Cask body) SH, SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-71

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Top flange (Cask body) SH, SR Stainless steel Air—

indoor/outdoor

Stress corrosion

cracking

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Embedded 

(neutron shielding)

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Bottom support ring and

bottom end forging (Cask 

body)

SH, SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 
4-72

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Bottom support ring and

bottom end forging (Cask 

body)

SH, SR Stainless steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Embedded (lead, 

neutron shielding)

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Bottom end plate (Cask 

body)

SH, SR Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 
4-73

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Bottom end plate (Cask 

body)

SH, SR Stainless steel Embedded 

(stainless steel)

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Lead gamma shielding

(Cask body)

SH, TH Lead Embedded (steel, 

stainless steel)

None identified None 

identified

No 

Rails (Cask attachments) SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Ca

AMP

Screw thread insert 

(Cask attachments)

SH, SR Stainless steel Embedded 

(stainless steel)

Fatigue Cracking Evaluate de

code TLAA

applicable

Radiation

embrittlement

Cracking No 
4-74

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Upper trunnions 

(Cask attachments)

SH, SR Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion

cracking

Cracking No 

Demineralized 

water

Stress corrosion

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Upper trunnion sleeves

(Cask attachments)

SH, SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of 

material

Transfer Ca

AMP
4-75

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Upper trunnion sleeves

(Cask attachments)

SH, SR Steel Air—

indoor/outdoor

Pitting and 

crevice 

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP

Demineralized 

water

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP
4-76

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Upper trunnion sleeves

(Cask attachments)

SH, SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP
4-77

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Upper trunnion cover 

plate and pad (Cask 

attachments)

SH, SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Upper trunnion nickel 

alloy (Cask attachments)

SR Inconel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Upper and lower trunnion

neutron shielding (Cask 

attachments)

SH, TH Bisco NS-3 Embedded (steel, 

stainless steel)

Thermal aging Loss of 

fracture

toughness 

and loss of 

ductility

No 
4-78

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Upper and lower trunnion

neutron shielding (Cask 

attachments)

SH, TH Bisco NS-3 Embedded (steel, 

stainless steel)

Radiation

embrittlement

Cracking No 

Boron depletion Loss of 

shielding

TLAA/AMP 

supporting 

analysis is 

required

Lower trunnions (Cask

attachments)

SH, SR Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Lower trunnions sleeves

(Cask attachments)

SH, SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 
4-79

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Lower trunnions sleeves

(Cask attachments)

SH, SR Steel Air—

indoor/outdoor

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP

Demineralized 

water

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 
4-80

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Lower trunnions sleeves

(Cask attachments)

SH, SR Stainless steel Air—

indoor/outdoor

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Wear Loss of 

material

Transfer Ca

AMP

Lower trunnion sleeve

nickel alloy weld overlay

(Cask attachments)

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 
4-81

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Lower trunnion sleeve

nickel alloy weld overlay

(Cask attachments)

SR Stainless steel Air—

indoor/outdoor

Fatigue Cracking Evaluate de

code TLAA

applicable

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Internal sleeve 

components 

(Cask attachments)

SR Aluminum Embedded 

(stainless steel)

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Bottom head cap screw 

for internal sleeve 

(Cask attachments)

SR Steel Embedded 

(stainless steel, 

aluminum)

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Stress relaxation Loss of 

preload

No 
4-82

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Washer for internal 

sleeve (Cask 

attachments)

SR Stainless steel Embedded (steel, 

stainless steel)

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Spacer assembly 

(Cask attachments)

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Ram access penetration

ring (Cask penetration)

SH, SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable
4-83

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Upper and lower rings,

outer shell relief valve 

support plates (Cask 

neutron shield)

SH, SR, TH Stainless steel Air—

indoor/outdoor

Pitting and 

crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Embedded 

(neutron shielding)

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable
4-84

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Neutron shield panel

support angles (Cask 

neutron shield)

SH, SR, TH Stainless steel Embedded 

(neutron shielding)

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Neutron shield panels 

and plates (Cask neutron

shield)

SH, SR, TH Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress.corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No. 
4-85

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Neutron shield panels 

and plates (Cask neutron 

shield)

SH, SR, TH Stainless steel Air—

indoor/outdoor

Fatigue Cracking Evaluate de

code TLAA

applicable

Embedded 

(neutron shielding)

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Demineralized 

water

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Castable neutron 

shielding material (Cask 

neutron shield)

SH, TH Bisco NS-3 Embedded (steel, 

stainless steel)

Thermal aging Loss of 

fracture

toughness 

and loss of 

ductility

No 

Radiation

embrittlement

Cracking No 
4-86

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Castable neutron 

shielding material (Cask 

neutron shield)

SH, TH Bisco NS-3 Embedded (steel, 

stainless steel)

Boron depletion Loss of 

shielding

TLAA/AMP 

supporting 

analysis is 

required

Inner, outer, and side top

cover plates (Cask cover 

assembly)

SH, SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable
4-87

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Bottom cover plate (Cask

cover assembly)

SH, SR, TH Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Top and bottom cover

neutron shielding (Cask 

cover assembly)

SH Bisco NS-3 Embedded 

(stainless steel)

Thermal aging Loss of 

fracture

toughness 

and loss of 

ductility

No 

Radiation

embrittlement

Cracking No 

Boron depletion Loss of 

shielding

TLAA/AMP 

supporting 

analysis is 

required

Bolts, screws, and

washers for top and 

bottom cover plates 

(Cask cover assembly)

SH, SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice

corrosion

Loss of 

material

Transfer Ca

AMP
4-88

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Bolts, screws, and

washers for top and 

bottom cover plates 

(Cask cover assembly)

SH, SR Steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Stress relaxation Loss of 

preload

No 

Socket head cap screws

for bottom cover plate

(Cask cover assembly)

SH, SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Stress relaxation Loss of 

preload

No 

Airflow wedge plates

(Cask cover assembly)

SH, SR, TH Stainless steel Air—

indoor/outdoor

Pitting and 

crevice

corrosion

Loss of 

material

No 
4-89

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Airflow wedge plates

(Cask cover assembly)

SH, SR, TH Stainless steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Support skid 

supplemental

shielding

SH, SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice 

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Bolts and washers for

support skid 

supplemental shielding

SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice 

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-90

Table 4-6 NUHOMS transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Agin

Manage

Bolts and washers for

support skid 

supplemental shielding

SR Steel Air—

indoor/outdoor

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable

Stress relaxation Loss of 

preload

No 

Upper and lower decon

area cask shielding

SH, SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of 

material

Transfer Ca

AMP

Pitting and 

crevice

corrosion

Loss of 

material

Transfer Ca

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Fatigue Cracking Evaluate de

code TLAA

applicable
4-91

1 4.3 HI-STORM 100 and HI-STAR 100 systems

2 4.3.1 System description

3 Holtec International developed the HI-STORM (Holtec International–Storage and Transfer 

4 Operation Reinforced Module) 100 system and the HI-STAR (Holtec International–Storage, 

5 Transport, and Repository) 100 system. The HI-STORM 100 system consists of a metallic 

6 multipurpose canister (MPC) that contains the SNF assemblies, a vertical concrete storage 

7 overpack that contains the MPC during storage, and a HI-TRAC (Holtec International–Transfer 

8 Cask) TC that contains the MPC during loading, unloading, and transfer operations. The 

9 HI-STAR 100 system consists of an MPC and a vertical metal overpack, which is used to load, 

10 unload, transfer, and store the SNF assemblies contained in the MPC. The HI-STORM 100 

11 system is certified only for storage, while the HI-STAR 100 system (including its metal overpack) 

12 is certified for both storage and transportation. Figure 4-9 presents schematics of the 

13 HI-STORM 100 and HI-STAR 100 systems.

14 The HI-STORM design is presently licensed for use in the United States under NRC 

15 Docket 72-1014, in combination with the MPC-24, MPC-32, and MPC-68 canisters, while the 

16 HI-STAR design is licensed for use under NRC Docket 72-1008, with the MPC-24 and MPC-68 

17 canisters. As in the case for the NUHOMS DSCs, the names of the Holtec MPCs reflect the 

18 number of fuel assemblies each MPC can hold. In addition, a variant design of the HI-STAR 

19 overpack, designated HI-STAR HB, is being used in conjunction with the MPC-HB canister 

20 under a site-specific license at the Humboldt Bay ISFSI. The details of the components of the 

21 two storage systems are provided below.

22 4.3.2 Multipurpose canister

23 The MPCs are welded cylindrical structures with an identical outer diameter, so that any MPC

24 will fit into either the HI-STORM or HI-STAR overpacks. However, only certain MPC and 

25 overpack combinations are currently licensed for use. Each MPC is an assembly consisting of a 

26 honeycombed fuel basket, baseplate, canister shell, lid, and closure ring. A cross sectional 

27 elevation view of a fuel basket for the MPC-68 series is shown in Figure 4-10. The number of 

28 spent fuel storage locations in each of the MPCs depends on the SNF assembly characteristics. 

29 Ten MPC models, distinguished by the type and number of SNF assemblies authorized for 

30 loading, are presently certified by the NRC for use in the United States. These are the MPC-24 

31 series (including the MPC-24E and MPC-24EF), the MPC-32 series (including the MPC-32F), 

32 and the MPC-68 series (including the MPC-68F, MPC-68FF, MPC-68M, and MPC-HB), shown 

33 in cross sectional views in Figure 4-11. The numerical suffix for each canister series denotes 

34 the maximum number of fuel elements that it can accommodate. Those canisters with “E” and 

35 “F” designations are designed for the storage of damaged fuel rods and fuel debris. The MPC36 68M design contains a fuel basket constructed of Metamic-HTTM, a neutron absorbing material 

37 that also has a structural function.

38 The fuel storage cells in the MPC-24 series are physically separated from one another by a 

39 water gap, also called a flux trap, for criticality control. Flux traps are not used in the MPC-32 

40 and MPC-68 series. The MPC-32 design includes credit for soluble boron in the MPC water 

41 during wet fuel loading and unloading operations for criticality control. The MPC fuel basket is 

42 positioned and supported within the MPC shell by a set of basket supports welded to the inside 

43 of the MPC shell. In the early-vintage MPCs that were loaded under the original HI-TORM 100 

44 design, optional heat conduction elements (fabricated from thin aluminum Alloy 1100) may have 
4-92

Figure 4-9 HI-STORM 100 (left) (Holtec International, 2013) and HI-STAR 100 (right) 

(Holtec International, 2001) systems

Figure 4-10 Cross section elevation view of MPC (Holtec International, 2013)
4-93

(a) MPC-24/24E/24EF (b) MPC-32/32F

(c) MPC-68/68F/68FF

Figure 4-11 Cross sectional views of different MPC designs

(Holtec International, 2013)

1 been installed between the periphery of the basket, the MPC shell, and the basket supports. 

2 For shorter SNF assemblies, upper and lower fuel spacers, as appropriate, maintain the axial 

3 position of the SNF assembly within the MPC basket. 

4 All structural components in MPCs are made of a material designated by the manufacturer as 

5 Alloy X. Candidate Alloy X materials include Types 304, 304LN, 316, and 316LN austenitic 

6 stainless steels. All MPC components that are likely to come in contact with spent fuel pool 

7 water or the ambient environment are constructed from stainless steel, with the exception of 

8 neutron poison plates, aluminum seals on vent and drain port caps, and optional aluminum heat 

9 conduction elements.

10 Lifting lugs attached to the inside surface of the MPC canister shell (shown in Figure 4-10) 

11 permit placement of the empty MPC into the HI-TRAC transfer cask and also serve to axially 

12 locate the MPC lid before welding. They are not used to handle a loaded MPC, because the 

13 MPC lid is installed before any handling of a loaded canister. 

14 The top end of the MPC incorporates a redundant closure system. The MPC lid is a circular 

15 plate (fabricated from one piece or two pieces—split top and bottom) that is welded to the MPC 
4-94

1 outer shell. In the two-piece lid design, only the top piece comprises a part of the enclosure 

2 vessel’s pressure boundary; the bottom piece is attached to the top piece with a nonstructural, 

3 nonpressure-retaining weld and acts as a radiation shield. The lid is equipped with vent and 

4 drain ports that are used to remove moisture and air from the MPC and backfill the MPC with 

5 helium. The vent and drain ports are covered and seal-welded before the closure ring is 

6 installed. The closure ring is a circular ring edge-welded to the MPC shell and lid. The MPC lid 

7 provides sufficient rigidity to allow the entire MPC, loaded with spent nuclear fuel, to be lifted by 

8 the threaded holes in the MPC lid.

Boral® 9 and METAMICTM neutron poison panels are used in the basket and are enclosed in 

10 Alloy X stainless steel sheathing that is stitch-welded to the MPC basket cell walls along their 

11 entire periphery. Thus, the neutron poison material is contained in a tight, welded pocket 

12 enclosure. The shear strength of the pocket weld joint, which is an order of magnitude greater 

13 than the weight of a fuel assembly, ensures that the neutron poison and its enveloping 

14 sheathing pocket will maintain their as-installed position under all loading, storage, and transport 

15 conditions. In addition, the pocket joint detail ensures that fuel assembly insertion or withdrawal 

16 into or out of the MPC basket will not lead to a disconnection of the sheathing from the cell wall.

17 The MPC does not require any valves, gaskets or mechanical seals for confinement. 

18 Figure 4-12 shows the MPC confinement boundary. All components of the confinement 

19 boundary are safety significant and are fabricated entirely of stainless steel. The primary 

20 confinement boundary is defined by the outline formed by the sealed, cylindrical enclosure of 

21 the MPC shell (including any associated axial or circumferential welds) welded to the baseplate 

22 at the bottom, the MPC lid welded around the top circumference to the shell wall, and the port 

23 cover plates welded to the lid. A shield lid is bolted to the top of the MPC lid and provides 

24 radiation shielding.

25 The helium backfill gas plays an important role in the MPC thermal performance. It fills all the 

26 spaces between solid components and provides an improved conduction medium relative to air 

27 for dissipating decay heat in the MPC. Furthermore, the pressurized helium environment within 

28 the MPC sustains a closed-loop thermo-siphon action, removing SNF decay heat by upward 

29 flow of helium through the storage cells.

30 Table 4-7 provides a generic evaluation of potential aging mechanisms and effects requiring 

31 management for specific components of the MPC. The AMPs that provide an acceptable 

32 approach to managing the aging effects are also identified in the table.

33 4.3.3 HI-STORM concrete overpack

34 The HI-STORM overpacks are thick-walled concrete cylindrical vessels enclosed in a steel 

35 shell. There are three base HI-STORM overpack designs: (i) HI-STORM 100, 

36 (ii) HI-STORM 100S, and (iii) HI-STORM 100S Version B. The significant differences among 

37 the three are overpack height, MPC pedestal height, location of the air outlet ducts, and vertical 

38 alignment of the inlet and outlet air ducts. The HI-STORM 100S Version B overpack design 

39 does not include a concrete-filled pedestal to support the MPC. Instead, the MPC rests upon a 

40 steel plate that maintains the MPC sufficiently above the inlet air ducts to prevent direct 

41 radiation shine through the ducts. Figure 4-13 shows cross sectional views of the 

42 HI-STORM 100 and 100S overpacks. The HI-STORM 100A and 100SA overpack designs are 

43 the anchored variant of the HI-STORM 100 and 100S designs.

44
4-95

Figure 4-12 MPC confinement boundary (Holtec International, 2013) 

Figure 4-13 Cross sectional views of the HI-STORM 100 and 100S overpacks with an 

MPC inserted (Holtec International, 2013)

1 A base HI-STORM overpack design is capable of storing each type of MPC. The overpack 

2 inner shell is provided with channels distributed around the inner cavity that provide guidance 

3 for MPC insertion and removal, and a flexible medium to absorb some of the impact during a 

4 tipover. They also allow the flow of cooling air through the overpack. The main structural 

5 function of the HI-STORM overpack is provided by carbon steel, and the main shielding function 

6 is provided by concrete. The concrete, enclosed by cylindrical inner and outer steel shells, a 

7 thick baseplate, and a top plate, is specified to provide the necessary shielding properties and 

8 compressive strength. The overpack lid has appropriate concrete shielding to provide neutron 

9 and gamma attenuation in the vertical direction.
4-96

1 The HI-STORM overpack has air ducts to allow for passive natural convection cooling of the 

2 contained MPC. A minimum of four air inlets and four air outlets are located at the lower and 

3 upper extremities of the storage system, respectively. The vertical annulus between the MPC 

4 and the inner shell of the overpack facilitates an upward flow of air by buoyancy forces, drawing 

5 ambient air from the inlet vents and releasing it from the outlet vents at the top of the HI-STORM 

6 storage system. The annulus ventilation flow cools the hot MPC surfaces and transfers decay 

7 heat to the outside environment. 

8 The principal function of the concrete is to provide shielding against gamma and neutron 

9 radiation. However, it also imparts a large thermal inertia to the HI-STORM overpack, allowing 

10 it to moderate the rise in temperature of the system under hypothetical conditions when all 

11 ventilation passages are assumed to be blocked. The high thermal inertia characteristics of the 

12 HI-STORM concrete also control the temperature of the MPC in the event of a postulated fire 

13 accident at the ISFSI. Although the annular concrete mass in the overpack shell is not a 

14 structural member, it does act as an elastic/plastic filler of the intershell space.

15 Four threaded anchor blocks, located at 90-degree intervals around the circumference of the top 

16 of the overpack lid, are provided for lifting. The anchor blocks are integrally welded to the radial 

17 plates, which in turn are full-length welded to the overpack inner shell, outer shell, and 

18 baseplate (HI-STORM 100) or the inlet air duct horizontal plates (HI-STORM 100S).

19 The HI-STORM 100S Version B overpack design incorporates partial-length radial plates at the 

20 top of the overpack to secure the anchor blocks and uses both gussets and partial-length radial 

21 plates at the bottom of the overpack for structural stability. The overpack may also be lifted 

22 from the bottom using specially designed lifting transport devices, including hydraulic jacks, air 

23 pads, Hillman rollers, or other designs based on site-specific needs and capabilities.

24 For anchoring, the HI-STORM 100A overpack baseplate is extended to allow it to be attached to 

25 the reinforced concrete structure of the ISFSI. Sector lugs are bolted to the ISFSI pad using 

26 anchor studs. The lateral load-bearing capacity of the HI-STORM/pad interface is many times 

27 greater than the horizontal sliding force exerted on the cask under the postulated design-basis 

28 earthquake seismic event. Thus, the potential for lateral sliding of the HI-STORM 100A system 

29 during a seismic event is precluded, as is the potential for any bending action on the 

30 anchor studs. 

31 The HI-STORM 100 system also includes a variant 100U underground module design. The 

32 HI-STORM 100U design provides storage of an MPC inside a cylindrical cavity located entirely 

33 below the top of the grade of the ISFSI. HI-STORM 100U comprises the cavity enclosure 

34 container, consisting of the container shell welded to the bottom plate and the container flange, 

35 and the closure lid, divider shell, insulation, and bearing pads, as well as the interfacing and 

36 proximate structures, such as interface pad, support foundation pad, subgrade surrounding the 

37 module, top surface pad, and retaining wall.

38 Table 4-8 provides a generic evaluation of potential aging mechanisms and effects requiring 

39 management for specific components of the HI-STORM overpack, respectively. The AMPs that 

40 provide an acceptable approach to managing the aging effects are also identified in the tables.

41 4.3.4 HI-STAR metal overpack

42 The HI-STAR 100 overpack is a sealed, thick-walled carbon and low-alloy steel cylindrical 

43 vessel. The overpack containment boundary is formed by an inner shell welded at the bottom to 
4-97

1 a cylindrical main flange and bolted to a top closure plate. The HI-STAR 100 overpack with the 

2 MPC partially inserted is shown in Figure 4-9. The overpack consists of one inner shell, five 

3 intermediate shells, and one enclosure shell, which form the body of the overpack. Figure 4-14

4 and Figure 4-15 provide an elevation and cross section view, respectively, of the overpack. 

5 Two concentric grooves are machined into the closure plate to accept the metallic seals. The 

6 bolted closure plate is recessed into the top flange, and the bolted joint is configured to provide 

7 maximum protection to the closure bolts and seals in the event of a drop accident. The closure 

8 plate has test and vent ports, which are sealed by a threaded port plug with a metallic seal. The 

9 bottom plate has a drain port that is sealed by a threaded port plug with a metallic seal. The 

10 inner surfaces of the HI-STAR overpack form an internal cylindrical cavity for housing the MPC.

11 The outer surface of the overpack inner shell is buttressed with the five layers of intermediate 

12 shells of gamma shielding in the form of layers of carbon steel plate installed so as to ensure a 

13 permanent state of contact between adjacent layers. Besides serving as an effective gamma 

14 shield, these intermediate layers provide additional strength to the overpack to resist potential 

15 punctures or penetrations from external missiles. Radial channels are vertically welded to the 

16 outside surface of the outermost intermediate shell at equal intervals around the circumference 

17 (see Figure 4-15). The radial channels act as fins for improved heat conduction to the overpack 

18 outer enclosure shell surface and as cavities for retaining and protecting the Holtite-ATM neutron 

19 shield described below.

20 The outer enclosure shell is formed by welding enclosure shell panels between each pair of 

21 radial channels to form the neutron shielding cavities, as shown in Figure 4-15. At the top of the

22 outer enclosure shell, rupture disks are positioned in a recessed area. These rupture disks

23 relieve internal pressure that may develop as a result of a fire accident and subsequent off 

24 gassing of the neutron shield material. Within each radial channel, a layer of silicone sponge is 

25 positioned to act as a thermal expansion foam to compress as the neutron shield expands. 

26 The exposed steel surfaces of the overpack are painted to prevent corrosion. Lifting trunnions 

27 are attached to the overpack top flange forging for lifting and for rotating the cask body between 

28 vertical and horizontal positions. The lifting trunnions are located 180 degrees apart in the sides 

29 of the top flange. Pocket trunnions are welded to the lower side of the overpack to provide a 

30 pivoting axis for rotation. The lifting trunnions do not protrude beyond the cylindrical envelope of 

31 the overpack enclosure shell. This feature reduces the potential for a direct impact on a 

32 trunnion in the event of an overpack side impact. The overpack is provided with aluminum 

33 honeycomb impact limiters, one at each end, to ensure that the impact loadings during accident 

34 conditions are maintained below the design levels. The neutron shielding material used in the 

35 HI-STAR 100 overpack is Holtite-ATM, a poured-in-place solid borated synthetic 

36 neutron-absorbing polymer.

37 Table 4-9 provides a generic evaluation of potential aging mechanisms and effects requiring 

38 management for specific components of the HI-STAR overpack. The AMPs that provide an 

39 acceptable approach to managing the aging effects are also identified in the table.
4-98

Figure 4-14 HI-STAR 100 overpack elevation view (Holtec International, 2001)

Figure 4-15 HI-STAR 100 overpack cross sectional view (Holtec International, 2001)

1 
4-99

1 4.3.5 Transfer cask

2 The HI-TRAC TC is a heavy-walled carbon steel cylindrical vessel composed of an inner shell 

3 and an outer shell with lead in between to provide gamma shielding (Holtec International, 2013). 

4 The TC also includes an exterior carbon steel water jacket for neutron shielding. There are four 

5 basic HI-TRAC TC designs: two standard designs, which are a 125-ton HI-TRAC 125 and a 

6 100-ton HI-TRAC 100, and two optional designs with a dual-purpose lid for loading and transfer 

7 operations, which are the 125-ton HI-TRAC 125D and the 100-ton HI-TRAC 100D. Figure 4-16

8 shows the cross section of a standard HI-TRAC 125 TC with both a pool lid and a transfer lid 

9 attached. Since all the MPCs have the same outer diameter, the inner diameter of all HI-TRAC 

10 TCs is the same. However, the external dimensions of the HI-TRAC TCs are different, because 

11 the 100-ton TCs have a reduced thickness of lead and water shielding. 

12 The main structural function of the HI-TRAC TCs is provided by carbon steel, and the main 

13 neutron and gamma shielding functions are provided by water and lead, respectively. The top 

14 lid of the HI-TRAC 125 and HI-TRAC 125D TCs contains additional Holtite-ATM neutron 

15 shielding material. The MPC access hole through the HI-TRAC top lid allows the lowering or 

16 raising of the MPC between the TC and the overpack. 

17 The standard design HI-TRAC TCs (including HI-TRAC 100 and HI-TRAC 125) include two 

18 bottom lids (pool lid and transfer lid). The pool lid is bolted to the bottom flange of the HI-TRAC 

19 and is used during MPC fuel loading and sealing operations. In addition to providing shielding 

20 in the axial direction, the pool lid incorporates two gasket seals, one between the pool lid top 

21 and the bottom flange and the other between the MPC outer wall and the TC inner wall close to 

22 the top lid of the TC. These seals provide a barrier from contamination of the exterior of the 

23 MPC by the spent fuel pool water. After the MPC has been drained, dried, and sealed, the pool 

24 lid is removed and the transfer lid is attached. The transfer lid incorporates two sliding doors 

25 that allow the opening of the HI-TRAC bottom for the MPC to be raised or lowered. Unlike the 

26 standard designs, the HI-TRAC 100D and HI-TRAC 125D TCs do not require swapping the pool 

27 lid for a transfer lid to facilitate transfer of the MPC. The HI-STORM mating device is used to 

28 remove the pool lid during MPC transfer operations.

29 In the standard designs, the HI-TRAC TC is equipped with two lifting trunnions located below 

30 the top flange and two pocket trunnions located above the bottom flange. The lifting trunnions, 

31 made of nickel alloy or stainless steel, are used to vertically handle the HI-TRAC TC. The 

32 carbon steel pocket trunnions provide a pivot point for the rotation of the HI-TRAC TC for 

33 downending or upending the HI-TRAC TC with a loaded MPC. The HI-TRAC 100D and 

34 HI-TRAC 125D TCs are equipped with only lifting trunnions.

35 Table 4-10 provides a generic evaluation of potential aging mechanisms and effects requiring 

36 management for specific components of the HI-TRAC TC. The AMPs that provide an 

37 acceptable approach to managing the aging effects are also identified in the table.
4-100

Figure 4-16 Cross sectional views of the HI-TRAC 125 transfer cask with pool lid (left) 

and transfer lid (right) (Holtec International, 2013)

1 

2 
4-101

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Shell CO, SH, SR, 

TH*

Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of 

material

(precursor to 

stress 

corrosion

cracking)

Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Baseplate CO, SH, SR, 

TH

Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-102

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Baseplate CO, SH, SR, 

TH

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of 

material

(precursor to 

stress 

corrosion

cracking)

Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Lid CO, SH, SR, 

TH

Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of 

material

(precursor to 

stress 

corrosion

cracking)

Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-103

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lid CO, SH, SR, 

TH

Stainless steel Sheltered Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Closure ring CO Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of 

material

(precursor to 

stress 

corrosion

cracking)

Localized Corr

and Stress cor

Cracking of W

Stainless Stee

Storage Canis

AMP

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Port cover plates CO Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 
4-104

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Port cover plates CO Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Basket cell plates CR, SH, SR, 

TH

Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

BWR fuel basket CR, SH, SR, 

TH

Metamic-HTTM Helium Boron depletion Loss of 

criticality

control

No; a TLAA m

required

Thermal aging Loss of 

strength

No 

General 

corrosion

Loss of 

material

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Loss of 

fracture

toughness and 

loss of ductility

No 

Neutron absorber CR, SH, TH Boral® Helium Boron depletion Loss of 

criticality

control

No; a TLAA m

required

Thermal aging Loss of 

strength

No 
4-105

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Neutron absorber CR, SH, TH Boral® Helium Wet corrosion 

and blistering

Change in

dimensions

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Loss of 

fracture

toughness and 

loss of ductility

No 

MetamicTM Helium Boron depletion Loss of 

criticality

control

No; a TLAA m

required

Thermal aging Loss of 

strength

No 

General 

corrosion

Loss of 

material

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Loss of 

fracture

toughness and 

loss of ductility

No 

Drain and vent 

shield blocks

SH Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Bottom portion of 

two-piece lid

SH Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 
4-106

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Bottom portion of 

two-piece lid

SH Stainless steel Helium Radiation

embrittlement

Cracking No 

Steel coated with 

stainless

steel

Helium Thermal aging Loss of 

fracture

toughness and

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

General 

corrosion

Loss of 

material

No 

Sheathing SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Basket supports SR, CR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-107

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lifting lugs SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Lifting lug base 

plate

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Upper fuel spacer 

column

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Upper fuel spacer 

end plate

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-108

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lower fuel spacer 

column

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Lower fuel spacer 

end plate

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Vent shield block 

spacer

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Vent and drain 

tubes

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-109

Table 4-7 HI-STORM / HI-STAR multipurpose canister

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Damaged fuel 

container

SR, CO Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Threaded disc, 

plug adjustment

CO Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Vent and drain 

plugs

CO Aluminum Helium Thermal aging Loss of 

strength

No 

Creep Change in

dimensions

No 

General 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 
4-110

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Concrete shield: 

radial shield, 

shield block, 

pedestal shield, lid 

shield

SH* Concrete Fully encased

(steel)

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Cracking No 

Loss of strength No 

Radiation 

damage

Cracking No 

Loss of strength No 

Reaction with 

aggregates

Cracking TLAA/AMP o

supporting an

required

Loss of strength TLAA/AMP o

supporting an

required

Shield block 

(base, ring, shell)

SH Steel Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-111

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lid inner ring SR Steel Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Lid outer ring SH Steel Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Shield shell SH Steel Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Gamma shield 

cross plates

SH Stainless 

steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking External Surf

Monitoring of 

Components 

Air—outdoor Stress corrosion 

cracking

Cracking External Surf

Monitoring of 

Components 

Stainless 

steel

Sheltered Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material

(precursor to stress 

corrosion cracking)

External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Air—outdoor Microbiologically

influenced 

corrosion

Loss of material No 
4-112

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Gamma shield 

cross plates

SH Stainless 

steel

Air—outdoor Pitting and 

crevice

corrosion

Loss of material

(precursor to stress 

corrosion cracking)

External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Baseplate, base 

spacer block

SR Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Outer shell SR Steel Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 
4-113

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Outer shell SR Steel Air—outdoor Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Inner shell, lid 

bottom plate, and 

lid shell

SR Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Pedestal shell SR Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

Radiation

embrittlement

Cracking No 
4-114

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Pedestal platform, 

MPC support

SH Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Inlet/outlet vent, 

vertical and 

horizontal plates, 

top plate, lid top 

plate, shear ring

SR Steel Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Heat shield, 

heat/lid shield ring

TH Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 
4-115

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Heat shield, 

heat/lid shield ring

TH Steel Sheltered Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Radial plate, 

radial gusset

SR Steel Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Lid stud and nut, 

lid closure bolt

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Lid stud SR Stainless 

steel

Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-116

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lid washer SR Stainless 

steel

Air—outdoor Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Bolt anchor block SR Steel Embedded 

(concrete)

Radiation

embrittlement

Cracking No 

Channel SR Steel 

(galvanized)

Sheltered Galvanic 

corrosion

Loss of material External Surf

Monitoring of 

Components 

General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Stainless 

steel 

(welded)

Sheltered Atmospheric 

stress

corrosion 

cracking

Cracking External Surf

Monitoring of 

Components 

Stainless 

steel

Sheltered Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material

(precursor to stress 

corrosion cracking)

External Surf

Monitoring of 

Components 
4-117

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Channel SR Stainless 

steel

Sheltered Radiation

embrittlement

Cracking No 

Channel mounts SR Steel Sheltered Galvanic 

corrosion

Loss of material External Surf

Monitoring of 

Components 

General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Lid lift block SR Steel Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Lug support ring, 

gusset

SR Steel Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 
4-118

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lug support ring, 

gusset

SR Steel Air—outdoor Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Stud with nut SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Closure lid 

concrete 

(HI-STORM 

100U)

SH Concrete Fully-encased 

(steel)

Dehydration at 

high 

temperatures

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Radiation 

damage

Cracking TLAA/AMP o

supporting an

required

Loss of strength TLAA/AMP o

supporting an

required
4-119

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Closure lid 

concrete 

(HI-STORM 

100U)

SH Concrete Fully-encased 

(steel)

Reaction with

aggregates

Cracking TLAA/AMP o

supporting an

required

Reaction with

aggregates

Loss of strength TLAA/AMP o

supporting an

required

Closure lid steel 

(HI-STORM 

100U)

SH Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

Radiation

embrittlement

Cracking No 
4-120

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Container shell, 

bottom plate 

(HI-STORM 100U)

SR Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Groundwater/soil General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material External Surf

Monitoring of 

Components 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Embedded 

(concrete)

General corrosion Loss of material TLAA/AMP o

supporting an

required

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material TLAA/AMP o

supporting an

required

Radiation

embrittlement

Cracking No 

Embedded (steel) Radiation

embrittlement

Cracking No 
4-121

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Container flange 

(HI-STORM 100U)

SR Steel Air—outdoor General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Divider shell and 

divider shell 

restraints 

(HI-STORM 100U)

TH Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Upper and lower 

MPC guides 

(HI-STORM 100U)

SR Steel Sheltered General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 
4-122

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

MPC bearing pads

(HI-STORM 100U)

SR Steel (with 

stainless 

steel liners)

Sheltered Galvanic 

corrosion

Loss of material External Surf

Monitoring of 

Components 

General corrosion Loss of material External Surf

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Pitting and 

crevice

corrosion

Loss of material External Surf

Monitoring of 

Components 

Radiation

embrittlement

Cracking No 

Insulation 

(HI-STORM 100U)

TH Kaowool 

(ceramic 

fiber) or 

equivalent

Fully encased 

(steel)

Moisture 

absorption

Loss of insulation

efficiency 

(increasing thermal 

conductivity)

No 

Radiation

embrittlement

Cracking TLAA/AMP o

supporting an

required

Reinforced 

concrete: VVM 

interface pad, top 

surface pad 

(HI-STORM 100U)

SR, SH Concrete Air—outdoor Aggressive 

chemical attack

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM

Creep Cracking No 
4-123

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Reinforced 

concrete: VVM 

interface pad, top 

surface pad 

(HI-STORM 100U)

SR, SH Concrete Air—outdoor Dehydration at 

high 

temperatures

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential

settlement

Cracking Reinforced C

Structures AM

Fatigue Cracking No 

Freeze and thaw Cracking Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Radiation 

damage

Cracking TLAA/AMP o

supporting an

required

Loss of strength TLAA/AMP o

supporting an

required

Reaction with

aggregates

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM

Salt scaling Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Shrinkage Cracking No 
4-124

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Reinforced 

concrete: VVM 

interface pad, top 

surface pad 

(HI-STORM 100U)

SR, SH Concrete Air—outdoor Leaching of 

calcium 

hydroxide

Loss of strength Reinforced C

Structures AM

Increase in porosity

and permeability

Reinforced C

Structures AM

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM

Groundwater/soil Aggressive 

chemical attack

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM

Creep Cracking No 

Dehydration at 

high 

temperatures

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 
4-125

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Reinforced 

concrete: VVM 

interface pad, top 

surface pad 

(HI-STORM 100U)

SR, SH Concrete Groundwater/soil Differential 

settlement

Cracking Reinforced C

Structures AM

Fatigue Cracking No 

Freeze and thaw Cracking Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Microbiological

degradation

Loss of strength Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Increase in porosity

and permeability

Reinforced C

Structures AM

Microbiological

degradation

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM

Radiation 

damage

Cracking TLAA/AMP o

supporting an

required

Loss of strength TLAA/AMP o

supporting an

required

Reaction with

aggregates

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM

Salt scaling Loss of material

(spalling, scaling)

Reinforced C

Structures AM
4-126

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Reinforced 

concrete: VVM 

interface pad, top 

surface pad 

(HI-STORM 100U

SR, SH Concrete Groundwater/soil Shrinkage Cracking No 

Leaching of 

calcium

hydroxide

Loss of strength Reinforced C

Structures AM

Increase in porosity

and permeability

Reinforced C

Structures AM

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM

Reinforcing 

steel

Air - outdoor; 

groundwater

Corrosion of

reinforcing steel

Loss of

concrete/steel bond

Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM

Retaining wall, 

support foundation 

pad (HI-STORM 

100U)

SR, SH Concrete Groundwater/soil Aggressive 

chemical attack

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM
4-127

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Retaining wall, 

support foundation 

pad (HI-STORM 

100U)

SR, SH Concrete Groundwater/soil Creep Cracking No 

Dehydration at 

high 

temperatures

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential

settlement

Cracking Reinforced C

Structures AM

Fatigue Cracking No 

Freeze and thaw Cracking No 

Loss of material

(spalling, scaling)

No 

Microbiological

degradation

Loss of strength Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Microbiological

degradation

Increase in porosity

and permeability

Reinforced C

Structures AM

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM

Radiation 

damage

Cracking TLAA/AMP o

supporting an

required

Loss of strength TLAA/AMP o

supporting an

required
4-128

Table 4-8 HI-STORM 100 overpack

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Retaining wall, 

support foundation 

pad (HI-STORM 

100U)

SR, SH Concrete Groundwater/soil Reaction with

aggregates

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM

Salt scaling Loss of material

(spalling, scaling)

No 

Leaching of 

calcium

hydroxide

Loss of strength Reinforced C

Structures AM

Increase in porosity

and permeability

Reinforced C

Structures AM

Reduction of

concrete pH 

(reducing corrosion 

resistance of steel 

embedments)

Reinforced C

Structures AM

Reinforcing 

steel

Air - outdoor; 

groundwater

Corrosion of

reinforcing steel

Loss of

concrete/steel bond

Reinforced C

Structures AM

Loss of material

(spalling, scaling)

Reinforced C

Structures AM

Cracking Reinforced C

Structures AM

Loss of strength Reinforced C

Structures AM
4-129

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Inner shell CO, SH* Steel Helium Fatigue Cracking Evaluate d

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Bottom plate CO, SH, SR Steel Air—outdoor Fatigue Cracking Evaluate d

code TLAA

applicable

Galvanic corrosion Loss of material External Su

Monitoring 

Metallic

Componen

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Helium Fatigue Cracking Evaluate d

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Top flange CO, SH, SR Steel Air—outdoor Fatigue Cracking Evaluate d

code TLAA

applicable

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-130

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Top flange CO, SH, SR Steel Air—outdoor Galvanic corrosion Loss of material External Su

Monitoring 

Metallic

Componen

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Helium Fatigue Cracking Evaluate d

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Closure plate CO, SH, SR Steel Air—outdoor Fatigue Cracking Evaluate d

code TLAA

applicable

Galvanic corrosion Loss of material External Su

Monitoring 

Metallic

Componen

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 
4-131

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Closure plate CO, SH, SR Steel Air—outdoor Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Helium Fatigue Cracking Evaluate d

code TLAA

applicable

Radiation

embrittlement

Cracking No 

Closure plate 

bolts

CO, SR Nickel alloy Air—outdoor Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate d

code TLAA

applicable

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Port plug CO Stainless steel Air—outdoor Stress corrosion 

cracking

Cracking No 

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-132

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Port plug seal and 

port cover seal

CO Nickel alloy Air—outdoor Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Closure plate 

seals

CO Nickel alloy Air—outdoor Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Intermediate 

shells

SH, SR Steel Embedded 

(Holtite-ATM)

Radiation

embrittlement

Cracking No 

Neutron shield SH Holtite-ATM Embedded 

(steel)

Radiation

embrittlement

Cracking TLAA/AMP 

supporting 

is required

Thermal aging Loss of fracture

toughness and 

loss of ductility

TLAA/AMP 

supporting 

is required

Boron depletion Loss of shielding TLAA/AMP 

supporting 

is required

Removable shear 

ring

SH Steel Air—outdoor General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 
4-133

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Pocket trunnion 

plug plate

SH Stainless steel 

(welded)

Air—outdoor Stress corrosion 

cracking

Cracking External Su

Monitoring 

Metallic

Componen

Stainless steel Air—outdoor Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material

(precursor to 

stress corrosion 

cracking)

External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Radial channels SR, TH Steel Air—outdoor General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Embedded 

(Holtite-ATM)

Radiation

embrittlement

Cracking No 

Pocket trunnion SH Stainless steel 

(welded)

Air—outdoor Stress corrosion 

cracking

Cracking External Su

Monitoring 

Metallic

Componen

Stainless steel Air—outdoor Microbiologically

influenced corrosion

Loss of material No 
4-134

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Pocket trunnion SH Stainless steel Air—outdoor Pitting and crevice

corrosion

Loss of material

(precursor to 

stress corrosion 

cracking)

External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Lifting trunnion SR Nickel alloy Air—outdoor Atmospheric stresscorrosion crackingCracking No 

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Rupture disk SR Brass Air—outdoor General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Rupture disk 

plate

SR Steel Air—outdoor Galvanic corrosion Loss of material External Su

Monitoring 

Metallic

Componen

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen
4-135

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Rupture disk 

plate

SR Steel Air—outdoor Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Removable shear 

ring bolt, pocket 

trunnion plug 

screw, and 

alignment pin

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Enclosure shell 

panels and 

enclosure shell 

return

SR Steel Air—outdoor General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 
4-136

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Enclosure shell 

panels and 

enclosure shell 

return

SR Steel Air—outdoor Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Embedded 

(Holtite-ATM) 

Radiation

embrittlement

Cracking No 

Port cover SR Steel Air—outdoor Galvanic corrosion Loss of material External Su

Monitoring 

Metallic

Componen

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Port cover bolt SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 
4-137

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Port cover bolt SR Steel Air—outdoor Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Trunnion locking 

pad and end cap 

bolts

SR Steel Air—outdoor Stress corrosion 

cracking

Cracking No 

Galvanic corrosion Loss of material External Su

Monitoring 

Metallic

Componen

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen

Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Lifting trunnion 

end cap and 

locking pad

SR Steel Air—outdoor Galvanic corrosion Loss of material External Su

Monitoring 

Metallic

Componen

General corrosion Loss of material External Su

Monitoring 

Metallic

Componen
4-138

Table 4-9 HI-STAR 100 overpack

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect

Agin

Manage

Lifting trunnion 

end cap and 

locking pad

SR Steel Air—outdoor Microbiologically

influenced corrosion

Loss of material No 

Pitting and crevice

corrosion

Loss of material External Su

Monitoring 

Metallic

Componen

Radiation

embrittlement

Cracking No 
4-139

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Outer shell SH, SR, TH* Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice 

corrosion

Loss of 

material

Transfer Cas

Microbiologically

Influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Embedded 

(lead)

Radiation

embrittlement

Cracking No 

Demineralized 

water or 25% 

ethylene glycol 

solution

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice 

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Inner shell SH, SR, TH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice 

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Embedded 

(lead)

Radiation

embrittlement

Cracking No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-140

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Radial ribs SH, SR, TH Steel Demineralized 

water or 25% 

ethylene glycol 

solution

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Radial lead shield SH, TH Lead—ASTM 

B29

Embedded 

(steel)

None identified None 

identified

No 

Water jacket 

enclosure shell 

panels

SH, SR, TH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Demineralized 

water or 25% 

ethylene glycol 

solution

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Lower water jacket 

shell

SH, SR, TH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas
4-141

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lower water jacket 

shell

SH, SR, TH Steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Demineralized 

water or 25% 

ethylene glycol 

solution

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Water jacket end 

plate, short rib

SH, SR, TH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Demineralized 

water or 25% 

ethylene glycol 

solution

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

Corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Pool lid outer ring SH, SR, TH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas
4-142

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Pool lid outer ring SH, SR, TH Steel Air—

indoor/outdoor

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Pool lid top and 

bottom plates

SH, SR, TH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Pool lid bolt SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Pool lid lead shield SH, TH Lead Embedded 

(steel)

None identified None 

identified

No 

Top flange SR, SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas
4-143

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Top flange SR, SH Steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Top lid outer and 

inner rings, top and 

bottom plates, 

lifting block

SR, SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Embedded 

(Holtite-ATM)

Radiation

embrittlement

Cracking No 

Top lid stud or bolt SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Top lid nut SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-144

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Top lid nut SR Steel Air—

indoor/outdoor

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Top lid shielding SH, TH Holtite-ATM Embedded 

(steel)

Radiation

embrittlement

Cracking TLAA/AMP o

supporting an

required

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

TLAA/AMP o

supporting an

required

Boron depletion Loss of 

shielding

TLAA/AMP o

supporting an

required

Fill port plugs SR, SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Embedded 

(lead)

Radiation

embrittlement

Cracking No 

Lifting trunnion 

block

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-145

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lifting trunnion 

block

SR Steel Air—

indoor/outdoor

Radiation

embrittlement

Cracking No 

Lifting trunnion SR Nickel alloy Air—

indoor/outdoor

Pitting and crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Cas

Stainless steel Air—

indoor/outdoor

Pitting and crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Cas

Lifting trunnion end 

cap

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-146

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lifting trunnion end 

cap

SR Steel Air—

indoor/outdoor

Radiation

embrittlement

Cracking No 

Pocket trunion, 

removable pocket 

trunion, pocket 

trunnion base

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Dowel pins, pocket

trunnion bolts

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Bottom flange SR, SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

Embrittlement

Cracking No 
4-147

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Transfer lid top, 

bottom, 

intermediate, cover, 

and cover side 

plates

SR, SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Transfer lid door 

top, middle, bottom, 

interface, side, and 

end plates

SR, SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Transfer lid door 

top, middle, and 

side plates

SR, SH Steel Embedded 

(Holtite-ATM)

Radiation

embrittlement

Cracking No 

Embedded 

(lead)

Radiation

embrittlement

Cracking No 

Transfer lid door 

wheel housing

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 
4-148

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Transfer lid door 

wheel housing

SR Steel Embedded 

(lead)

Radiation

embrittlement

Cracking No 

Embedded 

(Holtite-ATM) 

Radiation

embrittlement

Cracking No 

Transfer lid wheel 

shaft

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Transfer lid housing

stiffener

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Transfer lid door 

lock bolt

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 
4-149

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Transfer lid door 

lock bolt

SR Steel Air—

indoor/outdoor

Stress relaxation Loss of 

preload

No 

Transfer lid lifting 

lug, lug pad

SR Steel Air—

indoor/outdoor

General corrosion Loss of

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Transfer lid wheel 

track

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Cas

Transfer lid door 

stop block

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

Embrittlement

Cracking No 
4-150

Table 4-10 HI-TRAC transfer cask

Structure, System, 

or Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Transfer lid door 

stop block bolt

SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Cas

Pitting and crevice

corrosion

Loss of 

material

Transfer Cas

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Transfer lid door 

shielding

SH, TH Holtite-ATM Embedded 

(steel)

Radiation

embrittlement

Cracking TLAA/AMP o

supporting an

required

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

TLAA/AMP o

supporting an

required

Boron depletion Loss of 

shielding

TLAA/AMP o

supporting an

required

Transfer lid door 

lead shield

SH, TH Lead Embedded 

(steel)

None identified None 

identified

No 

Transfer lid side 

lead shield

SH, TH Lead Embedded 

(steel)

None identified None 

identified

No 
4-151

1 4.4 TN-32 and TN-68 systems

2 4.4.1 System description

3 The Transnuclear Inc. (TN) spent-fuel storage cask is a vertical metal cask with a bolted lid 

4 closure and two metallic O-rings forming the seal. As a storage cask, it provides confinement, 

5 shielding, criticality control, and passive heat removal. There are three types of TN metal 

6 storage casks: TN-32, TN-40 (TN-40HT), and TN-68. Only the TN-32 (NRC Docket 72-1021) 

7 and TN-68 (NRC Docket 72-1027) casks are evaluated here. The TN-32 cask accommodates 

8 32 PWR fuel assemblies. The TN-68 cask accommodates up to 68 BWR fuel assemblies and is 

9 also licensed for transportation. Damaged fuel that can be handled by normal means may be 

10 stored in eight peripheral compartments of the TN-68 cask that are fitted with damaged-fuel end 

11 caps designed to retain gross fragments of fuel.

12 4.4.2 Bolted metal cask

13 The TN-32 and TN-68 cask body is a right circular cylinder composed of the following 

14 components: (i) confinement vessel with bolted lid closure, (ii) basket for fuel assemblies, 

15 (iii) gamma and neutron shield, (iv) pressure/leak-tightness monitoring system, (v) weather 

16 cover, and (vi) and trunnions. Figure 4-17 shows the components of the TN-32 cask, and 

17 Figure 4-18 shows the confinement-boundary components of the TN-68 cask. The details of the 

18 components of the TN-32 cask are provided below as an example of both TN metal casks.

19 Confinement boundary, closure lid, and pressure-monitoring system

20 The TN-32 cask confinement boundary consists of a welded cylindrical low-alloy steel inner 

21 shell with an integrally welded low-alloy steel bottom closure. A flange forging is welded to the 

22 top of the inner shell to accommodate a bolted low-alloy steel lid closure. The inner shell has a 

23 sprayed metallic aluminum coating for corrosion protection. The confinement vessel is 

24 surrounded by a carbon steel gamma shield wall and bottom. The cask is sealed with a carbon 

25 steel closure lid, which is secured to the top flange of the containment vessel by 48 bolts.

26 The closure lid uses a double-barrier seal system with two metallic O-rings (Helicoflex seals) 

27 forming the seal. The annular space between the metallic O-rings is connected to a pressure 

28 monitoring system placed between the lid and the protective cover, also called the weather 

29 cover, shown in Figure 4-19. Pressure in the tank of the pressure-monitoring system is 

30 maintained above the pressure in the cask cavity to prevent either flow of fission gases out of or 

31 air into the cask cavity, which, under normal storage conditions, is pressurized above 

32 atmospheric pressure with helium. The transducers/switches monitor the pressure in the 

33 annular space between the metallic O-rings to provide an indication of seal failure before any 

34 release is possible. Two identical transducers/switches are provided to ensure a functional 

35 system through redundancy.

36 The TN-32 cask body has four carbon steel trunnions that are welded to the gamma shield. 

37 Two of these are located near the top of the cylindrical steel forging, spaced 180 degrees apart, 

38 and are used for lifting the cask. The remaining two trunnions are 180 degrees apart and 

39 located near the bottom of the cask. The lower trunnions are used to rotate the unloaded cask 

40 between vertical and horizontal positions. The lifting trunnions are hollow to permit installation 

41 of neutron-shielding material and eliminate a path for neutron streaming. The TN-68 design 

42 differs from the TN-32 design in that its two top trunnions are bolted to the gamma shield.
4-152

Figure 4-17 Components of the TN-32 storage cask (NRC, 1996)

Figure 4-18 TN-68 cask confinement boundary components (Transnuclear Inc., 2005)
4-153 

Figure 4-19 TN-32 cask seal pressure-monitoring system (NRC, 1996) 

1 The TN-32 cask lid has three confinement access ports—a drain port, a vent port, and an 

2 overpressure system port. The drain and vent ports are covered by a bolted stainless steel 

3 closure plate having a double-barrier seal system with two metallic O-rings forming the seal, 

4 similar to the one used for the lid closure. The overpressure port is also covered by a bolted 

5 stainless steel closure plate but has a single metallic O-ring forming the seal. The closure lid 

6 has drilled interseal passageways connecting the annular space between the seals at each port 

7 to the annular space between the closure-lid seals, as shown in Figure 4-19. The cavity drain 

8 line penetrates the closure lid and terminates in the bottom of the cask cavity. This line is used 

9 to drain water from the cask cavity after underwater fuel loading. It is also used during the 

10 drying and helium backfilling of the cask cavity. 

11 The all-metal Helicoflex seal used in the TN metal casks has a central helical energizing spring 

12 with inner and outer liners. Sealing is accomplished by plastic flow of the outer liner against the 

13 mating sealing surfaces. The helical spring aids in keeping a sufficient load against the outer 

14 liner to follow temperature fluctuations and small deformations. 

15 The TN-32 confinement vessel has a cylindrical cavity that holds a fuel basket with 

16 32 compartments to locate and support the PWR fuel assemblies. The basket assembly also 

17 transfers heat from the fuel assembly to the cask body wall and provides neutron absorption to 

18 satisfy nuclear criticality requirements, especially during loading and unloading operations that 

19 occur underwater. During storage, with the cavity dry, filled with inert gas, and sealed from the 

20 environment, criticality control measures within the cask are not necessary because of the low 

21 reactivity of the fuel in the dry cask and the assurance that no water can enter the cask during 

22 storage. 

23 Fuel basket assemblies and shielding 

The fuel cavities in the basket are formed by a sandwich of aluminum plates, Boral® 24 plates, and 

25 stainless steel boxes. The stainless steel fuel-compartment box sections are attached by a 
4-154

1 series of stainless steel plugs that pass through the aluminum plates and the poison plates and 

2 are fusion-welded to both adjacent stainless steel box sections. The aluminum provides the 

3 heat-conduction paths from the fuel assemblies to the cask cavity wall. The poison material 

4 provides the necessary criticality control. The basket is held in place by aluminum rails that run 

5 the axial length of the cask body, as shown in Figure 4-20. 

6 Surrounding the outside of the confinement vessel wall is a steel gamma shield, as shown in 

7 Figure 4-21. The bolted closure lid provides the gamma shielding at the upper end of the cask 

8 body. Neutron emissions from the stored fuel are attenuated by a neutron shield, consisting of 

9 a borated polyester resin compound, enclosed in long aluminum boxes that surround the 

10 gamma shield. These aluminum containers are held in place by a steel shell. Neutron 

11 emissions from the top of the cask are attenuated by a polypropylene disc, encased in a steel 

12 shell and placed on the top of the closure lid. There is no neutron shielding provided on the 

13 bottom of the cask.

14 The inside surfaces of the inner shell and bottom have a sprayed metallic coating of aluminum

15 for corrosion protection. The external surfaces of the cask are metal-sprayed with aluminum 

16 and/or painted for ease of decontamination and corrosion protection. The neutron shield, 

17 pressure-monitoring system, and shield cap are placed on top of the cask after fuel is loaded 

18 into the cask. A stainless steel overlay is applied to the O-ring seating surfaces on the body for 

19 corrosion protection. A protective cover is bolted to the top of the cask body to provide weather 

20 protection for the lid penetrations and other components.

21 Table 4-11 provides a generic evaluation of potential aging mechanisms and effects requiring 

22 management for specific components of the TN-32 and TN-68 casks. The AMPs that provide 

23 an acceptable approach to managing the aging effects are also identified in the table.

Figure 4-20 Radial cross section of TN-32 cask showing basket, basket rails, and

 gamma and neutron shields (NRC, 1996) 
4-155

Figure 4-21 TN-32 cask shielding configuration (NRC, 1996)

1
4-156

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Outer shell SH, SR, 

TH* 

Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Radial neutron 

shield

SH, TH Borated 

polyester resin

Embedded 

(aluminum)

Radiation

embrittlement

Cracking TLAA/AMP or a 

supporting ana

required

Thermal 

aging

Loss of fracture

toughness and 

loss of ductility

TLAA/AMP or a 

supporting ana

required

Boron 

depletion

Loss of shielding TLAA/AMP or a 

supporting ana

required

Radial neutron 

shield box

TH Aluminum Embedded 

(borated 

polyester 

resin)

Thermal aging Loss of strength No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Gamma shield SH, SR, 

TH

Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-157

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Gamma shield SH, SR, 

TH

Steel Air—outdoor Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Cask body 

bottom

SH, SR, 

TH

Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Upper and lower 

trunnions

SR Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Wear Loss of material External Surfac

Monitoring of M

Components A

Upper trunnion SR Stainless steel Air—outdoor Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 
4-158

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Upper trunnion SR Stainless steel Air—outdoor Galvanic corrosion Loss of material External Surfac

Monitoring of M

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Wear Loss of material External Surfac

Monitoring of M

Components A

Trunnion bolts SR Steel Air—outdoor Stress relaxation Loss of preload No 

Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material External Surfac

Monitoring of M

Components A

Galvanic corrosion Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Trunnion cover 

screw

SH, SR Stainless steel Air—outdoor Stress relaxation Loss of preload No 

Stress corrosion 

cracking

Cracking No 
4-159

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Trunnion cover 

screw

SH, SR Stainless steel Air—outdoor Pitting and 

crevice corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Top neutron 

shield

SH, TH Polypropylene Embedded 

(steel)

Thermal aging Loss of fracture 

toughness and 

loss of ductility

TLAA/AMP or a 

supporting ana

required

Radiation 

embrittlement

Cracking No 

Top neutron 

shield bolt, 

vent & drain 

port cover 

bolts

SR Stainless steel Sheltered Stress relaxation Loss of preload No 

Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Steel Sheltered Stress relaxation Loss of preload Bolted Cask Se

Leakage Monit

AMP

Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP
4-160

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Top neutron 

shield bolt, vent 

& drain port 

cover bolts

SR Steel Sheltered Galvanic corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Pitting and 

crevice corrosion

Loss of material Bolted Cask Se

Leakage Monit

AMP

Microbiologically 

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Top neutron 

shield flat 

washer

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lid CO, SH, 

SR, TH

Steel Sheltered General corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Galvanic corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Pitting and 

crevice corrosion

Loss of material Bolted Cask Se

Leakage Monit

AMP
4-161

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Lid CO, SH, 

SR, TH

Steel Sheltered Microbiologically 

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Helium General corrosion Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Lid assembly 

shim

SH, SR, 

TH

Steel Embedded 

(steel)

Radiation 

embrittlement

Cracking No 

Flange CO, SH, 

SR, TH

Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components A

Galvanic corrosion Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Sheltered General corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP
4-162

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Flange CO, SH, 

SR, TH

Steel Sheltered Galvanic corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Pitting and 

crevice corrosion

Loss of material Bolted Cask Se

Leakage Monit

AMP

Microbiologically 

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Helium General corrosion Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Lid bolts CO, SH, 

SR, TH

Steel Sheltered Stress relaxation Loss of preload Bolted Cask Se

Leakage Monit

AMP

Stress corrosion 

cracking

Cracking No 

General corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Galvanic corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Pitting and 

crevice corrosion

Loss of material Bolted Cask Se

Leakage Monit

AMP
4-163

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Lid bolts CO, SH, 

SR, TH

Steel Sheltered Microbiologically 

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Lid threaded 

insert

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lid seal, vent & 

drain port cover 

seal

CO, SH, 

SR, TH

Aluminum Sheltered General corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Galvanic corrosion Loss of material Bolted Cask Se

Leakage Monit

AMP

Pitting and 

crevice corrosion

Loss of material Bolted Cask Se

Leakage Monit

AMP

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Helium General corrosion Loss of material No 
4-164

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Lid seal, vent & 

drain port cover 

seal

CO, SH, 

SR, TH

Aluminum Helium Thermal 

aging

Loss of strength Bolted Cask Se

Leakage Monit

AMP

Creep Change in 

dimensions

Bolted Cask Se

Leakage Monit

AMP

Radiation 

embrittlement

Cracking No 

Drain port 

cover, vent port 

cover

CO, SH, 

SR, TH

Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Lid shield plate SH, SR, 

TH

Steel Helium General corrosion Loss of material No 

Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 
4-165

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Inner 

confinement 

shell, bottom 

confinement 

plate

CO, SH, 

SR, TH

Steel Helium General corrosion Loss of material No 

Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Basket rails SR, TH Aluminum Helium General corrosion Loss of material No 

Thermal aging Loss of strength TLAA/AMP or a 

supporting ana

required

Creep Change in 

dimensions

TLAA/AMP or a 

supporting ana

required

Radiation 

embrittlement

Cracking No 

Basket rail shim TH Aluminum Helium General corrosion Loss of material No 

Creep Change in 

dimensions

TLAA/AMP or a 

supporting ana

required

Radiation 

embrittlement

Cracking No 

Basket shim SR, TH Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Aluminum plate TH Aluminum Helium General corrosion Loss of material No 

Galvanic corrosion Loss of material No 
4-166

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Aluminum plate TH Aluminum Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Poison plate CR, TH Borated 

aluminum

Helium General corrosion Loss of material No 

Galvanic corrosion Loss of material No 

Thermal aging Loss of 

strength

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Boron depletion Loss of

criticality control

No; a TLAA ma

required

Boron carbide/ 

aluminum 

metal-matrix 

composite

Helium General corrosion Loss of material No 

Galvanic corrosion Loss of material No 

Thermal aging Loss of strength No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Boron depletion Loss of 

criticality 

control

No; a TLAA ma

required.

Boral® Helium General corrosion Loss of material No 

Galvanic corrosion Loss of material No 

Thermal aging Loss of 

strength

No 

Wet corrosion 

and blistering

Change in 

dimensions

No 

Creep Change in 

dimensions

No 
4-167

Table 4-11 TN bolted metal casks

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment Aging Mechanism Aging Effect Aging Manag

Poison plate CR, TH Boral® Helium Radiation 

embrittlement

Cracking No 

Boron depletion Loss of 

criticality 

control

No; a TLAA ma

required

Fuel 

compartment, 

structural 

plates, basket 

hold down

CR, SR, 

TH

Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Basket shear 

key

SR Steel Helium General corrosion Loss of material No 

Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

4-169

1 4.5 NAC International Systems

2 4.5.1 System description

3 NAC International Inc. (NAC) has three dry storage systems (DSSs) approved for use under a 

4 general license: the Universal Storage System (NAC-UMS), the Multi-Purpose Canister 

5 (NAC-MPC) system, and the Modular Advanced Generation Nuclear All-Purpose Storage 

6 (MAGNASTOR) system. These systems are canister-plus-overpack designs that use a vertical 

7 concrete storage cask to house a stainless steel storage canister with a welded closure. The 

8 sections below describe the details of the three storage systems.

9 4.5.2 NAC-UMS

10 The principal components of NAC-UMS are the transportable storage canister (TSC), vertical 

11 concrete cask (VCC), and transfer cask (TC), as shown in Figure 4-22 (NAC International, 

12 2004). There are five TSC configurations of different lengths for storage of different classes of 

13 PWR and BWR fuel assemblies. The TSC assembly consists of a right circular cylindrical shell 

14 with a welded bottom plate, fuel basket, shield lid, two penetration port covers, and a structural 

15 lid. The cylindrical shell, bottom plate, and lids constitute the confinement boundary. 

16 Figure 4-23 shows the various components of the NAC-UMS TSC for PWR fuel. All TSC 

17 components are made of stainless steel, with the exception of neutron poison plates, heat 

18 transfer disks, and support disks (BWR TSC fuel baskets only). The fuel basket is designed to 

19 accommodate up to 24 PWR or 56 BWR fuel assemblies. The fuel tubes are laterally supported 

20 by a series of stainless steel support disks in the PWR basket or carbon steel support disks in 

21 the BWR basket, which are retained by spacers on radially located tie rods. The carbon steel 

22 support disks are coated with electroless nickel. The square fuel tubes in the PWR basket 

include stainless-steel encased Boral® 23 sheets on all four sides for criticality control. The square 

fuel tubes in the BWR basket may include stainless-steel encased Boral® 24 sheets on up to two 

25 sides for criticality control. Aluminum heat transfer disks are spaced midway between the 

26 support disks and are the primary path for conducting heat from the fuel assemblies to the TSC 

27 wall.

28 The VCC is the storage overpack for the TSC and provides structural support, shielding, 

29 protection from environmental conditions, and natural convection cooling of the TSC during 

30 storage. Five concrete casks of different lengths are designed to accommodate different TSC 

31 configurations. The VCC side walls consist of reinforced concrete and a carbon steel inner 

32 liner. The VCC has an annular air passage to allow the natural circulation of air around the 

33 canister to remove the decay heat from the spent fuel stored in the TSC. The steel-lined air 

34 inlet and outlet vents take nonplanar paths to the cask cavity to minimize radiation streaming. 

35 The base plate assembly contains the air inlets and the pedestal that supports the TSC inside 

36 the VCC. The top of the VCC is closed by a shield plug consisting of a carbon steel plate for 

37 gamma shielding and solid neutron shielding of Bisco NS-3 or NS-4-FR, and a carbon steel lid. 

38 The carbon steel lid is installed above the shield plug and is bolted in place. The VCC is lifted 

39 from the bottom using an air-pad system. In an alternative design, a set of four carbon steel 

40 lifting lugs at the top of the VCC allows for lifting the cask with a loaded TSC from the top end.

41 The transfer cask is used for the vertical transfer of the TSC between workstations and the 

42 VCC. Five TCs of different lengths are designed to handle the five TSC configurations. The TC

43 incorporates a multiwall (steel/lead/neutron shield/steel) design and a top retaining ring that is 

44 bolted in place to prevent a loaded canister from being inadvertently removed through the top of 

45 the TC. All transfer cask structural components are fabricated with high-strength, low-alloy 
4-170

1 steel, with the exception of stainless steel retaining-ring bolts and shield door lock bolts. The 

2 TC contains retractable bottom shield doors for transfer of the TSC from the transfer cask into 

3 the VCC, as shown in Figure 4-24. Shield door rails are welded to the bottom plate of the TC to 

4 facilitate TSC transfer. The TC has two trunnions near the top of the cask. The trunnions are 

5 welded to the inner and outer shells for vertical cask-handling operations. All of the exposed 

6 surfaces of the TC, other than the load-bearing surfaces of the trunnions and the bottom door 

7 rails, are coated with an epoxy enamel coating to protect the carbon steel and to provide a 

8 smooth surface to facilitate decontamination.

9 Table 4-12 through Table 4-14 provide a generic evaluation of potential aging mechanisms and 

10 effects requiring management for specific components of the NAC-UMS. The tables also 

11 identify AMPs that provide an acceptable approach to managing the aging effects.

12 4.5.3 NAC-MPC

13 The NAC-MPC system is similar to the NAC-UMS but is designed for fuel from specific, older 

14 power plants. Like the UMS, the principal components of the NAC-MPC include a TSC, VCC, 

15 and a TC (NAC International, 2000). 

16 The TSC contains a fuel basket that is designed to accommodate up to 36 PWR spent fuel 

17 assemblies and reconfigured fuel assemblies with up to 4 damaged fuel cans from the Yankee 

18 Rowe Nuclear Power Station, 24 or 26 PWR spent fuel assemblies and reconfigured fuel 

19 assemblies with up to 4 damaged fuel cans from the Connecticut Yankee Nuclear Power Plant, 

20 or 68 BWR spent fuel assemblies and reconfigured fuel assemblies with up to 32 damaged fuel 

21 cans from the LaCrosse Nuclear Generating Station. 

22 The canister assembly for the Yankee Rowe and Connecticut Yankee configurations consists of 

23 a right circular cylindrical shell with a welded bottom plate, a fuel basket, a shield lid, two 

24 penetration port covers, and a structural lid. The cylindrical shell, bottom plate, and lids 

25 constitute the confinement boundary. The canister assembly for the La Crosse configuration

26 consists of a right circular cylindrical shell with a welded bottom plate, a fuel basket, a closure 

27 lid, a closure ring, and two sets of redundant penetration port covers. The cylindrical shell, 

28 bottom plate, closure lid, and inner port covers constitute the confinement boundary.

29 All TSC components are made of stainless steel, with the exception of neutron poison plates, 

30 aluminum heat-transfer disks, and an aluminum spacer plate attached to the underside of the 

31 closure lid (La Crosse BWR-MPC only). The fuel tubes are laterally supported by a series of 

32 support disks that are retained by spacers on radially located tie rods. Aluminum heat-transfer 

33 disks are spaced midway between the support disks and are the primary path for conducting 

34 heat from the fuel assemblies in the TSC wall. The fuel assemblies are contained in square 

stainless steel fuel tubes. The fuel tubes are covered with stainless steel-encased Boral® 35

36 sheets on all four sides for criticality control. An alternative fuel basket design has enlarged fuel 

tubes in the four corner locations. In this alternative configuration, the Boral® 37 sheet and 

38 stainless steel cover are removed from each side of the fuel tube in the four corner locations.

39 The VCC serves as the storage overpack for the TSC and provides structural support, shielding, 

40 protection from environmental conditions, and natural convection cooling of the TSC during 

41 storage. The VCC is fabricated from reinforced concrete with a carbon steel liner and base. 

42 The VCC has an annular air passage to allow the natural circulation of air around the TSC. The 

43 air inlet and outlet vents are steel-lined penetrations that take nonplanar paths to the cask cavity 

44 to minimize radiation streaming. The base-plate assembly contains the air inlets and the 
4-171

1 pedestal that supports the TSC inside the VCC. The top of the VCCs for the Yankee Rowe and 

2 Connecticut Yankee configurations is closed by a shield plug and a carbon steel lid bolted in 

3 place. The shield plug incorporates a carbon steel plate for gamma shielding and Bisco NS-3 or 

4 NS-4-FR for neutron shielding. For the La Crosse configuration, the top of the VCC is closed by 

5 a carbon steel and concrete lid bolted in place. The steel-enclosed concrete extends into the 

6 cask cavity from the bottom surface of the carbon steel lid. 

7 The TC is similar in design and construction to that of the UMS described in Section 4.5.2. It is

8 a multiwall (steel/lead/neutron shield/steel) design with retractable bottom shield doors to allow 

9 the TSC to be lowered into the VCC.

10 Table 4-15 through Table 4-17 provide a generic evaluation of potential aging mechanisms and 

11 effects requiring management for specific components of the NAC-MPC system. The tables 

12 also identify AMPs that provide an acceptable approach to managing the effects of aging.

13 4.5.4 MAGNASTOR

14 NAC developed the MAGNASTOR system to improve upon its previous designs in terms of 

15 storage capacity, thermal performance, and operations. The principal components of the 

16 MAGNASTOR system include a TSC with a welded closure, a concrete cask, and a TC

17 (NAC International, 2015). Figure 4-25 presents schematics of the MAGNASTOR system.

18 The TSC consists of a right circular cylindrical shell with a welded bottom plate, a fuel basket, a 

19 closure lid, a closure ring, and two sets of redundant penetration port covers. There are two 

20 TSC lengths to accommodate fuel of different lengths. The cylindrical shell plus the bottom 

21 plate, closure lid, and welded inner port covers are constructed of stainless steel and constitute 

22 the confinement boundary. The closure ring and the outer redundant port covers are also 

23 stainless steel and provide the required redundant closure for a welded canister system. There 

24 is an alternative closure lid design with a two-piece composite lid assembly that consists of a 

25 stainless steel closure lid and a carbon steel shield plate. The shield plate is coated with 

26 electroless nickel and bolted to the closure lid. The fuel basket, fabricated from carbon steel 

27 and coated with electroless nickel, is designed to accommodate up to 87 BWR fuel assemblies 

28 or 37 PWR fuel assemblies, including up to four damaged fuel can locations. The fuel basket 

29 design is an arrangement of square fuel tubes held in a right circular cylinder configuration using 

30 support weldments that are bolted to the outer fuel tubes. The fuel assembly cells in the fuel 

31 baskets include neutron absorber panels on up to four sides for criticality control. The materials 

of construction for the neutron absorber panels include Boral® 32 , borated aluminum, and borated 

33 metal matrix composite. Each neutron absorber panel is covered by a stainless steel sheet to 

34 protect the material during fuel loading and unloading and to maintain it in position. The neutron 

35 absorber and stainless steel cover are secured to the fuel tube using weld posts located across 

36 the width and along the length of the fuel tube.

37 The concrete cask is the storage overpack for the TSC and provides structural support, 

38 shielding, protection from environmental conditions, and natural convection cooling of the TSC 

39 during storage. The concrete cask is a reinforced concrete structure with a carbon steel inner 

40 liner and base, shown in Figure 4-26 (NAC International, 2014). There are four concrete cask 

41 configurations of different lengths and design variations. The concrete cask provides an annular 

42 air passage to allow the natural circulation of air around the TSC to remove the decay heat from 

43 the stored spent fuel. The lower air inlets and upper air outlets are steel-lined penetrations in 

44 the concrete cask body. The base plate assembly contains the air inlets and the pedestal that 

45 supports the TSC. Carbon steel channels that are attached to the inner liner assist in centering 
4-172

1 the TSC in the overpack. The top of the concrete cask is closed by a carbon steel and concrete 

2 lid bolted in place. The concrete cask is lifted by means of carbon steel lift anchor and lug 

3 assemblies embedded in the top of the concrete cask body. Alternatively, the concrete cask 

4 can be lifted and moved by means of air pads inserted into the four lower inlets.

5 The transfer cask provides shielding during TSC movements between workstations and the 

6 concrete cask. The TC is provided in two different configurations (referred to as MTC1 and 

7 MTC2) that differ primarily in the structural materials and overall length. The materials of 

8 construction for the TC structural components are carbon steel for the MTC1 configuration and 

9 stainless steel for the MTC2 configuration. The TC is a multiwall (steel/lead/NS-4-FR/steel) 

10 design. It incorporates stainless steel retaining blocks or a bolted retaining ring to prevent a 

11 loaded TSC from being inadvertently removed through the top of the TC. The TC contains 

12 retractable bottom shield doors that are used during TSC loading and unloading operations. 

13 Shield door rails are welded to the bottom ring of the TC to facilitate TSC transfer. The TC has 

14 two trunnions near the top of the cask that are welded to the top ring for vertical cask-handling 

15 operations. The exposed carbon steel surfaces of the MTC1 transfer cask, except for the wear 

16 surfaces of the shield doors and rails, are coated with an epoxy enamel coating to protect the 

17 components from corrosion and adverse interactions with the operating environments.

18 Table 4-18 through Table 4-20 provide a generic evaluation of potential aging mechanisms and 

19 effects requiring management for specific components of the MAGNASTOR system. The tables 

20 also identify AMPs that provide an acceptable approach to managing the aging effects. 

Figure 4-22 NAC-UMS (NAC International, 2004)

21
4-173

Figure 4-23 NAC-UMS transportable storage canister for PWR fuel 

(NAC International, 2004)

1 

Figure 4-24 NAC-UMS VCC and transfer cask arrangement (NAC International, 2004)

2 

3 
4-174

Figure 4-25 NAC MAGNASTOR TSC and concrete cask (NAC International, 2015)

1 

Figure 4-26 NAC MAGNASTOR concrete cask (NAC International, 2014)

2 
4-175

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Shell CO, SR* Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress corrosion 

cracking)

Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 

Bottom CO, SR Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress corrosion 

cracking)

Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-176

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Bottom CO, SR Stainless steel Sheltered Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 

Strucural lid SR Stainless steel 

(welded) 

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress corrosion 

cracking)

Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 
4-177

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Spacer ring SR Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress corrosion 

cracking)

Localized Corr

and Stress cor

Cracking of We

Stainless Stee

Storage Canist

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Shield lid, support 

ring

SH, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Port cover CO Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 
4-178

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Port cover CO Stainless steel Helium Radiation

embrittlement

Cracking No 

Fuel tube, cladding, 

flange

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Neutron absorber CR Boral Helium General corrosion Loss of material No 

Galvanic 

corrosion

Loss of material No 

Thermal aging Loss of strength No 

Wet corrosion 

and blistering

Change in

dimensions

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Boron depletion Loss of criticality

control

No; a TLAA ma

required.
4-179

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Fuel basket 

bottom weldment

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Fuel basket top 

weldment

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Fuel basket tie rod, 

spacer, washer

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-180

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Fuel basket support 

disk

SR Stainless steel 

(17-4 PH)

Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Steel Helium General corrosion Loss of material No 

Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Fuel basket top nut SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 
4-181

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Fuel basket top nut SR Stainless steel Helium Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Fuel basket heat 

transfer disk

TH Aluminum Helium General corrosion Loss of material No 

Thermal aging Loss of strength No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Maine Yankee fuel 

can tube body

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Maine Yankee fuel 

can bottom and 

side plates

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic
4-182

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Maine Yankee fuel 

can bottom and 

side plates

CR, SR Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Maine Yankee fuel 

can lid plate, lid 

bottom

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Maine Yankee fuel 

can lift tee, support 

ring

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Maine Yankee fuel 

can lid collar, 

screen

CO Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 
4-183

Table 4-12 NAC-UMS transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Maine Yankee fuel 

can lid collar, 

screen

CO Stainless steel Helium Radiation

embrittlement

Cracking No 
4-184

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Concrete shell SH, SR* Concrete Air—outdoor Aggressive 

chemical attack

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM

Loss of material

(spalling, scaling)

Reinforced Co

Structures AM

Creep Cracking No 

Dehydration at 

high 

temperatures

Cracking No 

Loss of strength No 

Delayed ettringite 

formation

Loss of material

(spalling, scaling)

No

 

Loss of strength No 

Cracking No 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforced Co

Structures AM

Loss of material

(spalling, scaling)

Reinforced Co

Structures AM

Radiation 

damage

Cracking No 

Loss of strength No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-185

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Concrete shell SH, SR Concrete Air—outdoor Reaction with

aggregates

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM

Salt scaling Loss of material

(spalling, scaling)

Reinforced Co

Structures AM

Shrinkage Cracking No 

Leaching of 

calcium 

hydroxide

Loss of strength Reinforced Co

Structures AM

Increase in 

porosity and 

permeability

Reinforced Co

Structures AM

Reduction of 

concrete pH 

(reducing 

corrosion 

resistance of 

steel 

embedments)

Reinforced Co

Structures AM

Reinforcing steel Air—outdoor,

groundwater

Corrosion of

reinforcing steel

Loss of 

concrete/steel 

bond

Reinforced Co

Structures AM

Loss of material

(spalling, scaling)

Reinforced Co

Structures AM

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM
4-186

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Inner shell SH, SR Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Embedded

(concrete)

Radiation 

embrittlement

Cracking No 

Top flange, support 

ring

SR Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Pedestal plate SR Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Galvanic 

corrosion

Loss of material External Surfac

Monitoring of M

Components A
4-187

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Pedestal plate SR Steel Sheltered Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Pedestal cover SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Base plate 

assembly

SH, SR, TH Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-188

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Base plate nelson 

studs

SR Steel Embedded

(concrete)

General 

corrosion

Loss of material TLAA/AMP or 

supporting ana

required

Pitting and 

crevice corrosion

Loss of material TLAA/AMP or 

supporting ana

required

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Outlet vent 

assembly

SH, SR, TH Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Galvanic 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Inlet and outlet vent 

hardware

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 
4-189

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Inlet and outlet vent 

hardware

SR Stainless steel Sheltered Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Inlet vent 

supplemental

shielding assembly

SH Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lid SR Steel Air—outdoor General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Galvanic 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 
4-190

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lid SR Steel Air—outdoor Radiation 

embrittlement

Cracking No 

Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lid hardware SR Stainless steel Air—outdoor Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Shield plug 

assembly

SH, SR Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A
4-191

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Shield plug 

assembly

SH, SR Steel Sheltered Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Neutron shield 

(Shield plug)

SH NS-4-FR Embedded 

(steel)

Thermal aging Loss of fracture 

toughness and 

loss of ductility

TLAA/AMP or 

supporting ana

required

Radiation 

embrittlement

Cracking TLAA/AMP or 

supporting ana

required

Boron depletion Loss of shielding TLAA/AMP or 

supporting ana

required

NS-3 Embedded 

(steel)

Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Radiation 

embrittlement

Cracking No 

Boron depletion Loss of shielding TLAA/AMP or 

supporting ana

required

Lift anchor SR Steel Air—outdoor General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A
4-192

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lift anchor SR Steel Air—outdoor Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Embedded

(concrete)

General 

corrosion

Loss of material TLAA/AMP or 

supporting ana

required

Pitting and 

crevice corrosion

Loss of material TLAA/AMP or 

supporting ana

required

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lift lug SR Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lift anchor 

hardware

SR Steel Sheltered General 

corrosion

Loss of material External Surfac

Monitoring of M

Components A
4-193

Table 4-13 NAC-UMS vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lift anchor 

hardware

SR Steel Sheltered Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload External Surfac

Monitoring of M

Components A
4-194

Table 4-14 NAC-UMS transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Outer shell SR* Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 

Galvanic 

corrosion

Loss of material Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Embedded 

(NS-4-FR)

Radiation 

embrittlement

Cracking No 

Inner shell SR Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Embedded 

(lead)

Radiation 

embrittlement

Cracking No 

Gamma shield 

(Cask body)

SH Lead Embedded 

(steel, NS-4-FR)

None identified None identified No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-195

Table 4-14 NAC-UMS transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Neutron shield 

(Cask body)

SH NS-4-FR Embedded 

(steel, lead)

Thermal aging Loss of fracture 

toughness and 

loss of ductility

TLAA/AMP or a

supporting analy

required

Radiation

embrittlement

Cracking TLAA/AMP or a

supporting anal

required

Boron depletion Loss of 

shielding

TLAA/AMP or a

supporting anal

required

Top plate SR Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Bottom plate SR Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-196

Table 4-14 NAC-UMS transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Retaining ring SR Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 

Galvanic 

corrosion

Loss of material Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Retaining ring bolts SR Stainless steel

(ferritic)

Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Trunnion SR Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 
4-197

Table 4-14 NAC-UMS transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Trunnion SR Steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Wear Loss of material Transfer Casks 

Shield door plates SH, SR Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Shield door rails SH, SR Steel Air—

indoor/outdoor

General corrosion Loss of material Transfer Casks 

Pitting and 

crevice corrosion

Loss of material Transfer Casks 

Galvanic 

corrosion

Loss of material Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Wear Loss of material Transfer Casks 
4-198

Table 4-14 NAC-UMS transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Shield door lock 

bolts

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Transfer adapter SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks 

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Wear strip SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 
4-199

Table 4-14 NAC-UMS transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Wear strip SR Stainless steel Air—

indoor/outdoor

Radiation 

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Casks 
4-200

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Shell CO, SR* Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

Stainless steel Sheltered Pitting and crevice

corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking TLAA/AMP or a

supporting analy

required

Bottom CO, SR Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

Stainless steel Sheltered Pitting and crevice

corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-201

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Bottom CO, SR Stainless steel Sheltered Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 

Closure lid; 

structural lid

SR Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

Stainless steel Sheltered Pitting and crevice

corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 
4-202

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Closure ring; 

spacer ring

SR Stainless steel 

(welded)

Sheltered Atmospheric stress

corrosion cracking

Cracking Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

Stainless steel Sheltered Pitting and crevice

corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Wel

Stainless Steel D

Storage Caniste

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Closure lid 

assembly bolt,

washer

SR Stainless steel Helium Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Closure lid 

assembly spacer

SR Aluminum Helium General corrosion Loss of 

material

No 

Galvanic corrosion Loss of 

material

No 

Thermal aging Loss of 

strength

TLAA/AMP or a

supporting analy

required

Creep Change in 

dimensions

TLAA/AMP or a

supporting analy

required
4-203

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Closure lid 

assembly spacer

SR Aluminum Helium Radiation 

embrittlement

Cracking No 

Shield lid, support 

ring

SH, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Port cover CO Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Sheltered Atmospheric stress

corrosion cracking

Cracking Localized Corros

Stress corrosion 

Cracking of Wel

Stainless Steel D

Storage Caniste

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Sheltered Pitting and crevice 

corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion 

Cracking of Wel

Stainless Steel D

Storage Caniste
4-204

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Port cover CO Stainless steel Sheltered Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Fuel tube, 

cladding, flange

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Neutron absorber CR Boral Helium General corrosion Loss of 

material

No 

Galvanic corrosion Loss of 

material

No 

Thermal aging Loss of 

strength

No 

Wet corrosion and 

blistering

Change in 

dimensions

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 
4-205

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Neutron absorber CR Boral Helium Boron depletion Loss of 

criticality 

control

No; a TLAA may 

required

Plate in lieu of 

neutron absorber

TH Aluminum Helium General corrosion Loss of 

material

No 

Galvanic corrosion Loss of 

material

No 

Thermal aging Loss of 

strength

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Fuel basket bottom 

weldment

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Fuel basket top 

weldment

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab
4-206

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Fuel basket top 

weldment

SR Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Fuel basket tie rod, 

spacer, washer

SR Stainless steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Fuel basket 

support disk

SR Stainless steel

(17-4 PH)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

TLAA/AMP or a 

supporting analy

required

Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Fuel basket top nut SR Stainless steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 
4-207

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Fuel basket top nut SR Stainless steel Helium Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Fuel basket heat 

transfer disk

TH Aluminum Helium General corrosion Loss of 

material

No 

Thermal aging Loss of 

strength

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Damaged fuel can 

tube body

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Damaged fuel can 

bottom and side 

plates, screen 

cover plate

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 
4-208

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Damaged fuel can 

bottom and side 

plates, screen 

cover plate

CR, SR Stainless steel Helium Radiation 

embrittlement

Cracking No 

Damaged fuel can 

lid plate, lid bottom

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Damaged fuel can

lift tee, support ring

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Damaged fuel can 

screen

CO Stainless steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 
4-209

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Damaged fuel can 

screen

CO Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Removable fuel 

rod retainer 

assembly 

(Yankee-MPC)

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Reconfigured fuel 

assembly shell 

casing, top ring 

(Yankee-MPC)

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Reconfigured fuel 

assembly top end 

fitting assembly 

(Yankee-MPC)

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in 

dimensions

No 
4-210

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Reconfigured fuel 

assembly top end 

fitting assembly 

(Yankee-MPC)

CR, SR Stainless steel Helium Radiation 

embrittlement

Cracking No 

Reconfigured fuel 

assembly bottom 

end fitting 

assembly 

(Yankee-MPC)

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Reconfigured fuel 

assembly top 

nozzle bolt

(Yankee-MPC)

CR, SR Stainless steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Reconfigured fuel 

assembly fuel 

basket corner 

angle, tie plate 

(Yankee-MPC)

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab
4-211

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Reconfigured fuel 

assembly fuel 

basket corner 

angle, tie plate 

(Yankee-MPC)

CR, SR Stainless steel Helium Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Reconfigured fuel 

assembly fuel 

tube, top and 

bottom caps 

(Yankee-MPC)

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Reconfigured fuel 

assembly fuel 

tube, corner 

angle, support 

grid (CY-MPC)

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 
4-212

Table 4-15 NAC-MPC transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Reconfigured fuel 

assembly screens 

(CY-MPC)

CO Stainless steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Reconfigured fuel 

assembly top and 

bottom housing, 

retaining plate 

and ring, guide 

plate, rod 

retaining plate, 

screen ring and

housing 

(CY-MPC)

SR Stainless steel Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 
4-213

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Concrete shell SH, SR* Concrete Air—outdoor Aggressive 

chemical attack

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM

Loss of material

(spalling, scaling)

Reinforced Co

Structures AM

Creep Cracking No 

Dehydration at 

high temperature

Cracking No 

Loss of strength No 

Delayed ettringite 

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforced Co

Structures AM

Loss of material

(spalling, scaling)

Reinforced Co

Structures AM

Radiation damage Cracking No 

Loss of strength No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-214

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Concrete shell SH, SR Concrete Air—outdoor Reaction with 

aggregates

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM

Salt scaling Loss of material

(spalling, scaling)

Reinforced Co

Structures AM

Shrinkage Cracking No 

Leaching of 

calcium hydroxide

Loss of strength Reinforced Co

Structures AM

Increase in 

porosity and 

permeability

Reinforced Co

Structures AM

Reduction of 

concrete pH 

(reducing 

corrosion 

resistance of steel 

embedments)

Reinforced Co

Structures AM

Reinforcing steel Air—outdoor,

groundwater

Corrosion of 

reinforcing steel

Loss of 

concrete/steel 

bond

Reinforced Co

Structures AM

Loss of material

(spalling, caling)

Reinforced Co

Structures AM

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM
4-215

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Inner shell SR Steel Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Embedded

(concrete)

Radiation 

embrittlement

Cracking No 

Top flange, 

support ring

SR Steel Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-216

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Pedestal plate SR Steel Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 

Galvanic 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Pedestal cover SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and crevice 

corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-217

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Base plate 

assembly

SH, SR, TH Steel Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Base plate nelson 

studs

SR Steel Embedded

(concrete)

General corrosion Loss of material TLAA/AMP or 

supporting an

required

Pitting and crevice 

corrosion

Loss of material TLAA/AMP or 

supporting an

required

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Outlet vent 

assembly

SH, SR, TH Steel Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 
4-218

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Outlet vent 

assembly

SH, SR, TH Steel Sheltered Galvanic 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Inlet and outlet 

vent hardware

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and crevice 

corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Inlet vent 

supplemental

shielding assembly

SH Steel Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 
4-219

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Inlet vent 

supplemental

shielding assembly

SH Steel Sheltered Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lid assembly SH, SR, TH Steel Air—outdoor General corrosion Loss of material External Surfa

Monitoring of 

Components 

Galvanic 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 
4-220

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lid assembly SH, SR, TH Steel Sheltered Radiation 

embrittlement

Cracking No 

Concrete Embedded 

(steel)

Delayed ettringite 

formation

Loss of material

(spalling, scaling)

No 

Cracking No 

Loss of strength No 

Radiation damage Cracking TLAA/AMP or 

supporting an

required

Loss of strength TLAA/AMP or 

supporting an

required

Reaction with 

aggregates

Cracking TLAA/AMP or 

supporting an

required

Loss of strength TLAA/AMP or 

supporting an

required

Lid center support,

nelson studs

SR Steel Embedded

(concrete)

General corrosion Loss of material TLAA/AMP or 

supporting an

required

Pitting and crevice 

corrosion

Loss of material TLAA/AMP or 

supporting an

required

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-221

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Lid hardware SR Stainless steel Air—outdoor Stress corrosion 

cracking

Cracking No 

Pitting and crevice 

corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Shield plug 

assembly

SH Steel Sheltered General corrosion Loss of material External Surfa

Monitoring of 

Components 

Pitting and crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Neutron shield 

(Shield plug)

SH NS-4-FR Embedded 

(steel)

Thermal aging Loss of fracture 

toughness and 

loss of ductility

TLAA/AMP or 

supporting an

required
4-222

Table 4-16 NAC-MPC vertical concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Neutron shield 

(Shield plug)

SH NS-4-FR Embedded 

(steel)

Radiation 

embrittlement

Cracking TLAA/AMP or 

supporting an

required

Boron depletion Loss of shielding TLAA/AMP or 

supporting an

required

NS-3 Embedded 

(steel)

Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Radiation 

embrittlement

Cracking No 

Boron depletion Loss of shielding TLAA/AMP or 

supporting an

required
4-223

Table 4-17 NAC-MPC transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Outer shell SR* Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Galvanic 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Embedded 

(NS-4-FR)

Radiation 

embrittlement

Cracking No 

Inner shell SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Embedded 

(lead)

Radiation 

embrittlement

Cracking No 

Gamma shield

(Cask body)

SH Lead Embedded 

(steel,

NS-4-FR)

None identified None identified No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-224

Table 4-17 NAC-MPC transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Neutron shield

(Cask body)

SH NS-4-FR Embedded 

(steel, lead)

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

TLAA/AMP or a

supporting analy

required

Radiation 

embrittlement

Cracking TLAA/AMP or a

supporting analy

required

Boron depletion Loss of 

shielding

TLAA/AMP or a

supporting analy

required

Top plate SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Bottom plate SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 
4-225

Table 4-17 NAC-MPC transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Retaining ring SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Galvanic 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Retaining ring 

bolts

SR Stainless steel Air—

indoor/outdoor

Pitting and

crevice corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Trunnion SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-226

Table 4-17 NAC-MPC transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Trunnion SR Steel Air—

indoor/outdoor

Radiation 

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Casks A

Shield door plates SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Shield door rails SH, SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and 

crevice corrosion

Loss of 

material

Transfer Casks A

Galvanic 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Casks A

Shield door lock 

bolts

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of 

material

No 
4-227

Table 4-17 NAC-MPC transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Shield door lock 

bolts

SR Stainless steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 
4-228

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Shell CO, SR* Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 

Bottom CO, SR Stainless steel 

(welded) 

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Pitting and 

crevice corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-229

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Bottom CO, SR Stainless steel Sheltered Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 

Closure lid SR Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 
4-230

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Closure ring SR Stainless steel

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Stainless steel Sheltered Pitting and 

crevice

corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Closure lid 

assembly bolt,

washer

SR Stainless steel Helium Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 
4-231

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Closure lid shield 

plate

SH, SR Steel Helium General corrosion Loss of 

material

No 

Galvanic 

corrosion

Loss of 

material

No 

Pitting and

crevice corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 

Port cover CO Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of 

material

(Precursor to 

stress 

corrosion 

cracking)

Localized Corros

Stress corrosion

Cracking of Weld

Stainless Steel D

Storage Canister

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Radiation

embrittlement

Cracking No 
4-232

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Port cover CO Stainless steel Helium Radiation

embrittlement

Cracking No 

Lifting lug, antirotation lugSR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Fuel tube CR, SR Steel Helium General corrosion Loss of 

material

No 

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Neutron absorber CR Borated aluminum Helium General corrosion Loss of 

material

No 

Galvanic 

corrosion

Loss of 

material

No 

Thermal aging Loss of 

strength

No 

Creep Change in

dimensions

No 
4-233

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Neutron absorber CR Borated aluminum Helium Radiation

embrittlement

Cracking No 

Boron depletion Loss of 

criticality

control

No; a TLAA may 

required

Boral Helium General corrosion Loss of 

material

No 

Galvanic 

corrosion

Loss of 

material

No 

Thermal aging Loss of 

strength

No 

Wet corrosion 

and blistering

Change in

dimensions

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Boron depletion Loss of 

criticality

control

No; a TLAA may 

required

Borated metal 

matrix composite

Helium General corrosion Loss of 

material

No 

Galvanic 

corrosion

Loss of 

material

No 

Thermal aging Loss of 

strength

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-234

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Neutron absorber CR Borated metal 

matrix composite

Helium Boron depletion Loss of 

criticality

control

No; a TLAA may 

required

Plate in lieu of 

neutron absorber

TH Aluminum Helium General corrosion Loss of 

material

No 

Galvanic 

corrosion

Loss of 

material

No 

Thermal aging Loss of 

strength

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Neutron absorber 

retainer

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Weld post SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 
4-235

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Weld post SR Stainless steel Helium Radiation

embrittlement

Cracking No 

Fuel basket 

support plates & 

gussets, 

connector pin 

washer

SR Steel Helium General corrosion Loss of 

material

No 

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Neutron absorber 

retainer clip, fuel 

basket support 

tube, vent and 

drain tube 

restrictor plate, 

fuel basket pins, 

spacer

SR Steel Helium General corrosion Loss of 

material

No 

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-236

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Fuel basket corner

support bar, 

support pin, fuel 

tube pin, 

connector pin

SR Steel Helium General corrosion Loss of 

material

No 

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Basket restraining 

block

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Basket support 

mounting bolt

SR Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 
4-237

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Basket support 

mounting bolt

SR Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Basket support 

washer, blocking 

strap

SR Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Basket shim SR Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Damaged fuel can 

tube body

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 
4-238

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Damaged fuel can 

tube body

CR, SR Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Damaged fuel can

bottom, side plates

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Damaged fuel can 

lid plate, lid guide, 

lid bottom

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-239

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Damaged fuel can 

collar, lift tee, 

support ring, tid 

tab

SR Stainless steel 

(welded)

Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Damaged fuel can

screens

CO Stainless steel Helium Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Damaged fuel 

spacer plate

SR Stainless steel Helium Fatigue Cracking Evaluate design 

TLAA, if applicab

Thermal aging Loss of 

fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-240

Table 4-18 MAGNASTOR transportable storage canister

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Damaged fuel 

closure lid shield 

plate

SH Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation

embrittlement

Cracking No 
4-241

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Concrete shell SH, SR* Concrete Air—outdoor Aggressive 

chemical attack

Cracking Reinforced Con

Structures AMP

Loss of strength Reinforced Con

Structures AMP

Loss of material

(spalling, 

scaling)

Reinforced Con

Structures AMP

Creep Cracking No 

Dehydration at 

high temperatures

Cracking No 

Loss of strength No 

Delayed ettringite 

formation

Loss of material

(spalling, 

scaling)

No 

Loss of strength No 

Cracking No 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforced Con

Structures AMP

Loss of material

(spalling, 

scaling)

Reinforced Con

Structures AMP

Radiation 

damage

Cracking No 

Loss of strength No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-242

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Concrete shell SH, SR Concrete Air—outdoor Reaction with

aggregates

Cracking Reinforced Con

Structures AMP

Loss of strength Reinforced Con

Structures AMP

Salt scaling Loss of material

(spalling, 

scaling)

Reinforced Con

Structures AMP

Shrinkage Cracking No 

Leaching of 

calcium hydroxide

Loss of strength Reinforced Con

Structures AMP

Increase in 

porosity and 

permeability

Reinforced Con

Structures AMP

Reduction of 

concrete pH 

(reducing 

corrosion 

resistance of 

steel 

embedements)

Reinforced Con

Structures AMP

Reinforcing steel Air—outdoor,

groundwater

Corrosion of

reinforcing steel

Loss of 

concrete/steel

bond

Reinforced Con

Structures AMP

Loss of material

(spalling, 

scaling)

Reinforced Con

Structures AMP

Cracking Reinforced Con

Structures AMP

Loss of strength Reinforced Con

Structures AMP

Inner shell SR, SH, TH Steel Sheltered General corrosion Loss of material External Surfac

Monitoring of M

Components AM
4-243

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Inner shell SR, SH, TH Steel Sheltered Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Embedded

(concrete)

Radiation

embrittlement

Cracking No 

Top flange SR Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Sheltered General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-244

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Pedestal plate SR Steel Sheltered General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Galvanic 

corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Pedestal cover SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Base plate 

assembly

(including nelson 

studs)

SH, SR, TH Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 
4-245

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Base plate 

assembly 

(including nelson 

studs)

SH, SR, TH Steel Air—outdoor Radiation 

embrittlement

Cracking No 

Sheltered General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Channels SR Steel Sheltered General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lid assembly SH, SR, TH Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Galvanic 

corrosion

Loss of material External Surfac

Monitoring of M

Components AM
4-246

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lid assembly SH, SR, TH Steel Air—outdoor Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Sheltered General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Concrete Embedded 

(steel)

Delayed ettringite 

formation

Loss of material

(spalling, 

scaling)

No 

Cracking No 

Loss of strength No 

Radiation 

damage

Cracking TLAA/AMP or a

supporting anal

required

Loss of strength TLAA/AMP or a

supporting anal

required

Reaction with 

aggregates

Cracking TLAA/AMP or a

supporting anal

required
4-247

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lid assembly SH, SR, TH Concrete Embedded 

(steel)

Reaction with 

aggregates

Loss of strength TLAA/AMP or a

supporting anal

required

Lid hardware SR Stainless steel Air—outdoor Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Lift anchor 

(standard 

configuration)

SR Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lift anchor 

(standard and 

alternative 

configurations)

SR Steel Embedded

(concrete)

General corrosion Loss of material TLAA/AMP or a

supporting anal

required

Pitting and 

crevice corrosion

Loss of material TLAA/AMP or a

supporting anal

required

Microbiologically

influenced 

corrosion

Loss of material No 
4-248

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lift anchor 

(standard and 

alternative 

configurations)

SR Steel Embedded

(concrete)

Radiation 

embrittlement

Cracking No 

Lift lug SR Steel Sheltered General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Lift lug bolt SR Nickel alloy Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice

corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Lift lug washer, 

base

plate dowel pin

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 
4-249

Table 4-19 MAGNASTOR concrete cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Lift lug washer, 

base

plate dowel pin

SR Stainless steel Sheltered Radiation 

embrittlement

Cracking No 

Cover plate SR Steel Air—outdoor General corrosion Loss of material External Surfac

Monitoring of M

Components AM

Galvanic 

corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Pitting and 

crevice corrosion

Loss of material External Surfac

Monitoring of M

Components AM

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Cover plate 

hardware

SR Stainless steel Air—outdoor Stress-corrosion 

cracking

Cracking No 

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 
4-250

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Outer shell SR* Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Embedded 

(NS-4-FR)

Radiation 

embrittlement

Cracking No 

Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion 

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Embedded 

(NS-4-FR)

Radiation 

embrittlement

Cracking No 

Inner shell SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-251

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Inner shell SR Steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Embedded 

(lead)

Radiation 

embrittlement

Cracking No 

Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion 

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Embedded 

(lead)

Radiation 

embrittlement

Cracking No 

Gamma shield

(Cask body)

SH Lead Embedded 

(steel,

NS-4-FR)

None identified None 

identified

No 

Embedded

(stainless steel, 

NS-4-FR)

None identified None 

identified

No 

Neutron shield

(Cask body)

SH NS-4-FR Embedded 

(steel, lead)

Thermal aging Loss of 

fracture

toughness 

and loss of 

ductility

TLAA/AMP or a

supporting analy

required

Radiation 

embrittlement

Cracking TLAA/AMP or a

supporting analy

required
4-252

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Neutron shield

(Cask body)

SH NS-4-FR Embedded 

(steel, lead)

Boron depletion Loss of 

shielding

TLAA/AMP or a

supporting analy

required

Embedded

(stainless steel, 

lead)

Thermal aging Loss of 

fracture

toughness 

and loss of 

ductility

TLAA/AMP or a

supporting analy

required

Radiation 

embrittlement

Cracking TLAA/AMP or a

supporting analy

required

Boron depletion Loss of 

shielding

TLAA/AMP or a

supporting analy

required

Top ring SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 
4-253

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Bottom ring SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Trunnion SR Steel Air—

indooroutdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 
4-254

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Trunnion SR Stainless steel Air—

indoor/outdoor

Radiation 

embrittlement

Cracking No 

Trunnion bushing, 

rotating bushing

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Casks A

Shield door plates SH Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 
4-255

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Shield door rails SH, SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Radiation 

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Casks A

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Casks A

Shield door tab SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of

material

No 

Radiation 

embrittlement

Cracking No 
4-256

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Shield door tab SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Shield door lock 

pin

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Retaining block 

and ring

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Retaining pin SR Steel Air—

indoor/outdoor

General corrosion Loss of 

material

Transfer Casks A

Pitting and crevice 

corrosion

Loss of 

material

Transfer Casks A

Microbiologically

influenced 

corrosion

Loss of 

material

No 
4-257

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Retaining pin SR Steel Air—

indoor/outdoor

Radiation 

embrittlement

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Retainer assembly 

spring plunger, 

guide bolt, handle 

bolt

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Retainer assembly 

handle

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 
4-258

Table 4-20 MAGNASTOR transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Manag

Retaining ring bolt 

and screw thread

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of 

preload

No 

Wear strip

(cask attachment)

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 

Wear Loss of 

material

Transfer Casks A

Fill/drain assembly 

tube and bar

SR Stainless steel Air—

indoor/outdoor

Pitting and crevice 

corrosion

Loss of 

material

No 

Microbiologically

influenced 

corrosion

Loss of 

material

No 

Stress-corrosion 

cracking

Cracking No 

Radiation 

embrittlement

Cracking No 
4-259

4.6 FuelSolutionsTM 1 storage system

2 4.6.1 System description

The FuelSolutionsTM 3 storage system uses a stainless steel storage canister stored within a

4 vertical cylindrical concrete storage cask. The principal components of the storage system are 

5 the W21 and W74 canisters, the W150 concrete storage cask, and the W100 transfer cask. The 

6 W21 canister is designed to accommodate nearly all domestic commercial spent nuclear fuel 

7 with a capacity of up to 21 PWR fuel assemblies. The W74 canister is designed to 

8 accommodate the three assembly types used at the Big Rock Point Nuclear Plant, including 

9 mixed oxide, partial, and damaged fuel assemblies, with a capacity of up to 64 BWR fuel 

10 assemblies. The W150 concrete storage cask provides radiation shielding and contains internal 

11 air flowpaths that allow decay heat from the canister spent fuel contents to be removed by 

12 natural air circulation around the canister wall. The W100 TC is used to move the loaded 

13 canisters to and from the storage cask. The sections below summarize the components of the 

FuelSolutionsTM 14 storage system.

15 4.6.2 W21 and W74 canisters

16 The W21 and W74 canisters, shown in Figure 4-27, each have several design configurations

17 consisting of different materials of construction and dimensions. A typical W21 or W74 canister 

18 consists of a shell assembly, top and bottom inner closure plates, vent and drain port covers, 

19 internal basket assembly, top and bottom shield plugs, and top and bottom outer closure plates.

20 All structural components of the canister are constructed of high-strength carbon or stainless 

21 steel. Any carbon steel used in the canister is coated with electroless nickel for corrosion 

22 protection. The canister shell, top and bottom inner closure plates, and vent and drain port 

23 covers form the confinement boundary and are fabricated from stainless steel.

24 W21 basket assembly

25 The W21 PWR fuel basket assembly is a right circular cylinder configuration with 21 stainless 

26 steel guide tubes for PWR contents (FuelSolutions, 2007a). The guide tubes are laterally 

27 supported by a series of spacer plates, held in position by support rods that run through support 

rod sleeves placed between the spacer plates. The square guide tubes include Boral® 28 neutron 

29 poison sheets on all four sides for criticality control. There are two classes of canister for the 

30 W21 canister based on different materials of construction: W21M and W21T. Each class of 

31 canister has four different types, which differ in dimension (exterior canister length and internal 

32 cavity length) and the material used for end plug shielding (steel, lead, or depleted uranium

33 (W21M only)). 

34 W74 basket assembly

35 The W74 BWR fuel basket assembly consists of two right circular cylindrical baskets, with a 

36 total of 56 guide tubes and a capacity of up to 64 assemblies (FuelSolutions, 2007b). The guide 

37 tubes are supported by a series of spacer plates held in position by support rods that run 

38 through support rod sleeves placed between the spacer plates. The square guide tubes include 

39 neutron poison sheets made of borated stainless steel, either on one side or on two opposite 

40 sides, in an arrangement within the basket that assures that there is a poison sheet between all 

41 of the assemblies. There are two classes of canister for the W74 canister based on different 

42 materials of construction (W74M and W74T). Unlike the W21 design, each W74 canister class 
4-260

1 has only one canister length and one cavity size, and carbon steel is used for end plug shielding 

2 material.

3 Table 4-21 provides a generic evaluation of potential aging mechanisms and effects requiring 

4 management for specific components of the W21 and W74 canisters. The table also identifies 

5 AMPs that provide an acceptable approach to managing the aging effects.

(a) (b)

Figure 4-27 Typical FuelSolutionsTM (a) W21 and (b) W74 canisters 

(FuelSoutions, 2007a,b)

6 4.6.3 W150 storage cask

7 The W150 storage cask, shown in Figure 4-28, is the overpack for storing both the long and 

8 short versions of the W21 and W74 canisters by varying the length of the middle concrete 

9 segment (FuelSolutions, 2007c). The overpack consists of a standard reinforced concrete 

10 structure with three precast segments (top, middle, and bottom) and a top cover made of steel 

11 and concrete. Stainless steel tie rods are used to tie the concrete segments together. A shear 

12 key between each two concrete segments provides positive lateral engagement and alignment 

13 and serves to minimize radiation streaming. Grout is installed between the keyed joints of the 
4-261

1 concrete segments to provide a weather barrier. The exterior surfaces of the concrete are 

2 exposed to the outdoor environment. 

Figure 4-28 FuelSolutionsTM W150 storage cask (FuelSolutions, 2007c)

3 The top cover of the overpack is bolted to the overpack top end segment shielding ring and is 

4 sealed with weather sealant. Inside the cavity of the overpack are a steel liner, an aluminum 

5 thermal shield, steel support and guide rails, and stainless steel canister support tubes. Guide 

6 rails are welded to the steel liner for centering the canister radially in the cavity. Canister 

7 support tubes are welded to the bottom plate of the steel liner plate to provide vertical support of 

8 the canister and to limit the g-load on the canister in a postulated accident. All steel 

9 components, such as the liner, top cover, and support and guide rails, are coated with 

10 temperature- and radiation-resistant coatings.

11 The overpack concrete bottom segment includes four inlet vents that converge into a single 

12 cylindrical inlet duct at the bottom center of the cask cavity. The center inlet duct also provides 

13 hydraulic ram access during horizontal canister transfer operations. The inlet and outlet vents 

14 have protective screens to prevent debris or wildlife from entering the ventilation ducts.

15 Table 4-22 provides a generic evaluation of potential aging mechanisms and effects requiring 

16 management for specific components of the W150 storage cask. The table also identifies AMPs 

17 that provide an acceptable approach to managing the aging effects.
4-262

1 4.6.4 W100 transfer cask

2 The W100 TC, shown in Figure 4-29, is a multiwall, stainless steel cylindrical vessel with covers 

3 on both ends (FuelSolutions, 2007c). The TC is composed of a structural shell and a stainless 

4 steel inner liner, with lead in the annular space to provide gamma shielding. The TC also 

5 includes an outer stainless steel jacket filled with demineralized water for neutron shielding. 

6 The penetrations in the neutron shield cavity consist of two quick-connect fittings that are used 

7 to drain and fill the neutron shield cavity and to prevent intrusion of contaminated spent fuel pool 

8 water. A pressure relief device is used to provide over-pressure protection for the neutron 

9 shield.

10 The structural shell and inner liner are welded to stainless steel flanges at the top and bottom 

11 ends. Both the top and bottom covers are made of stainless steel plates and an encased solid 

12 neutron shielding of RX-277 or BISCO NS-3. The top cover includes a secondary central cover 

13 for ram access during horizontal loading and unloading operations. The bottom cover has 

14 O-rings to prevent spent fuel pool water from entering the cask during loading operations. 

15 Nitronic 60 stainless steel guide rails are welded to the inner shell cavity to facilitate horizontal 

16 canister transfer.

17 The W100 TC has four stainless steel trunnions. Two upper lifting trunnions located near the 

18 top of the cask for vertical cask handling operations are welded to the structural shell and inner 

19 liner. The lower trunnions used for upending and downending the TC are welded to the 

20 structural shell. Heat removal from the TC is primarily by conduction through the cask wall. A 

21 high emissivity, low absorptivity coating is applied to the exterior of the liquid neutron shield 

22 jacket to facilitate radiative heat transfer to the environment.

23 Table 4-23 provides a generic evaluation of potential aging mechanisms and effects requiring 

24 management for specific components of the W100 TC. The table also identifies AMPs that 

25 provide an acceptable approach to managing the aging effects.

26

27
4-263

Figure 4-29 FuelSolutionsTM W100 transfer cask (FuelSolutions, 2007c)

1
4-264

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Shell CO, SR* Stainless steel 

(welded)

Sheltered Stress-corrosion 

cracking

Cracking Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress corrosion 

cracking)

Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 

Bottom closure 

plate

CO, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Thermal/Hea
4-265

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Bottom closure 

plate

CO, SR Stainless steel Helium Radiation

embrittlement

Cracking No 

Embedded 

(steel, depleted 

uranium)

Radiation

embrittlement

Cracking No 

Bottom end plate SR Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress-corrosion 

cracking)

Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Embedded 

(steel, depleted 

uranium)

Radiation

embrittlement

Cracking No 

Top outer closure 

plate

CO, SR* Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist
4-266

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Top outer closure 

plate

CO, SR Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress corrosion 

cracking)

Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Helium Radiation

embrittlement

Cracking No 

Top inner closure 

plate

CO, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Alignment bar, 

adapter

SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-267

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Shield plug SH Steel Helium General corrosion Loss of material No 

Galvanic 

corrosion

Loss of material No 

Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Embedded 

(stainless steel)

Radiation

embrittlement

Cracking No 

Embedded 

(stainless steel, 

depleted 

uranium)

Radiation

embrittlement

Cracking No 

Lead Embedded 

(stainless steel)

None identified None identified No 

Depleted uranium Embedded 

(stainless steel)

None identified None identified No 

Embedded 

(stainless steel,

steel)

None identified None identified No 

Shield plug support 

assembly

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-268

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Leak test port cover CO Stainless steel 

(welded)

Sheltered Stress corrosion 

cracking

Cracking Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist

Stainless steel Sheltered Pitting and 

crevice corrosion

Loss of material

(Precursor to 

stress corrosion 

cracking)

Localized Corr

Stress corrosio

Cracking of We

Stainless Stee

Storage Canist

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Helium Radiation 

embrittlement

Cracking No 

Instrument port 

cover, vent/drain 

port cover

CO Stainless steel 

(welded)

Helium Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 

Vent and drain port SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Creep Change in

dimensions

No 
4-269

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Vent and drain port SR Stainless steel Helium Radiation

embrittlement

Cracking No 

Guide tube 

assembly

CR, SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Neutron absorber CR Boral Helium General corrosion Loss of material No 

Galvanic 

corrosion

Loss of material No 

Thermal aging Loss of strength No 

Wet corrosion 

and blistering

Change in

dimensions

No 

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Boron depletion Loss of criticality

control

No; a TLAA ma

required

Borated stainless 

steel

Helium Boron depletion Loss of criticality 

control

No; a TLAA ma

required

Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Loss of fracture 

toughness and 

loss of ductility

No 
4-270

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Fuel basket support 

rod, support sleeve

SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Fuel basket support 

rod

SR Stainless steel 

(17-4 PH)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

TLAA/AMP or 

supporting ana

required

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Fuel basket support 

sleeve

SR Steel Helium General corrosion Loss of material No 

Galvanic 

corrosion

Loss of material No 

Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 
4-271

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Fuel basket bolt SR Steel Helium General corrosion Loss of material No 

Galvanic 

corrosion

Loss of material No 

Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload TLAA/AMP or 

supporting ana

required

Fuel basket spacer

assembly

SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Steel Helium General corrosion Loss of material No 

Galvanic 

corrosion

Loss of material No 

Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic
4-272

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Fuel basket spacer 

assembly

SR Steel Helium Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Damaged fuel can 

top lid assembly 

(W74 Canister)

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Damaged fuel can 

top lid assembly 

hardware 

(W74 Canister)

SR Stainless steel Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 

Radiation

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Damaged fuel can 

guide tube 

assembly 

(W74 Canister)

SR Stainless steel 

(welded)

Helium Thermal aging Loss of fracture

toughness and 

loss of ductility

No 

Stainless steel Helium Fatigue Cracking Evaluate desig

TLAA, if applic

Creep Change in

dimensions

No 
4-273

Table 4-21 FuelSolutionsTM canister

Structure, 

System, or 

Component

Intended 

Safety 

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Damaged fuel can 

guide tube 

assembly 

(W74 Canister)

SR Stainless steel Helium Radiation

embrittlement

Cracking No 

Damaged fuel can 

neutron absorber 

(W74 Canister)

CR Borated stainless 

steel

Helium Boron depletion Loss of criticality 

control

No; a TLAA ma

required

Thermal aging Loss of fracture 

toughness and 

loss of ductility

No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Loss of fracture 

toughness and 

loss of ductility

No 
4-274

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Concrete shell, 

shear key

SH, SR* Concrete, grout Air—outdoor Aggressive 

chemical attack

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM

Loss of material

(spalling, 

scaling)

Reinforced Co

Structures AM

Creep Cracking No 

Dehydration at 

high 

temperature

Cracking No 

Loss of strength No 

Delayed 

ettringite 

formation

Loss of material

(spalling,

scaling)

No 

Loss of strength No 

Cracking No 

Fatigue Cracking No 

Freeze and thaw Cracking Reinforced Co

Structures AM

Loss of material

(spalling, 

scaling)

Reinforced Co

Structures AM

Radiation 

damage

Cracking TLAA/AMP or 

supporting ana

required

Loss of strength TLAA/AMP or 

supporting ana

required

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-275

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Concrete shell, 

shear key

SH, SR Concrete Air—outdoor Reaction with

aggregates

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM

Salt scaling Loss of material

(spalling, 

scaling)

Reinforced Co

Structures AM

Shrinkage Cracking No 

Leaching of 

calcium 

hydroxide

Loss of strength Reinforced Co

Structures AM

Increase in 

porosity and 

permeability

Reinforced Co

Structures AM

Reduction of 

concrete pH 

(reducing 

corrosion 

resistance of 

steel 

embedments)

Reinforced Co

Structures AM

Concrete shell SH, SR Reinforcing steel Air—outdoor,

groundwater

Corrosion of

reinforcing steel

Loss of 

concrete/steel 

bond

Reinforced Co

Structures AM

Loss of material

(spalling, 

scaling)

Reinforced Co

Structures AM

Cracking Reinforced Co

Structures AM

Loss of strength Reinforced Co

Structures AM
4-276

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Steel liner, shield 

ring

SH, SR Steel Sheltered General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Galvanic 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Embedded

(concrete)

Radiation 

embrittlement

Cracking No 

Thermal shield TH Aluminum Sheltered General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Galvanic 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Thermal aging Loss of strength No 

Creep Change in 

dimensions

No 

Radiation 

embrittlement

Cracking No 
4-277

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Shear lug, thermal 

shield support lug

SR Steel Sheltered General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Support rail, guide 

rail

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice 

corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Wear Loss of material External Surfa

Monitoring of 

Components A

Canister support 

tube

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice 

corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-278

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Tie rod, tie rod 

plate

SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice 

corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Tie rod hardware SR Stainless steel Sheltered Stress corrosion 

cracking

Cracking No 

Pitting and 

crevice 

corrosion

Loss of material No 

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Steel Sheltered General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Galvanic 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 
4-279

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Tie rod hardware SR Steel Sheltered Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload External Surfa

Monitoring of 

Components A

Ram anchor SR Steel Sheltered General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Top cover 

assembly

SR Steel Air—outdoor General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation

embrittlement

Cracking No 

Sheltered General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A
4-280

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Top cover 

assembly

SR Steel Sheltered Microbiologically 

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

Concrete Embedded 

(steel)

Delayed 

ettringite 

formation

Loss of material

(spalling, 

scaling)

No 

Cracking No 

Loss of strength No 

Radiation 

damage

Cracking TLAA/AMP or 

supporting ana

required

Loss of strength TLAA/AMP or 

supporting ana

required

Reaction with 

aggregates

Cracking TLAA/AMP or 

supporting ana

required

Loss of strength TLAA/AMP or 

supporting ana

required

Top cover bolt SR Steel Sheltered General 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Pitting and 

crevice 

corrosion

Loss of material External Surfa

Monitoring of 

Components A

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 
4-281

Table 4-22 FuelSolutionsTM storage cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Top cover bolt SR Steel Sheltered Stress relaxation Loss of preload External Surfa

Monitoring of 

Components A

Coating on carbon 

steel components

SR Coating Air—outdoor Radiation 

embrittlement

Coating 

degradation

TLAA/AMP or 

supporting ana

required

Thermal aging Coating 

degradation

TLAA/AMP or 

supporting ana

required

Sheltered Radiation 

embrittlement

Coating 

degradation

TLAA/AMP or 

supporting ana

required

Thermal aging Coating 

degradation

TLAA/AMP or 

supporting ana

required
4-282

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Structural shell SR* Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion 

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Inner liner SR Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion 

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Neutron shield 

jacket, trunnion 

support plate, 

thermowell

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-283

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Neutron shield 

jacket, trunnion 

support plate, 

thermowell

SR Stainless steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Neutron shield 

jacket support rib

SR Stainless steel Demineralized 

water

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Gamma shield SH Lead Embedded

(stainless steel)

None identified None identified No 

Guide rail SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Wear Loss of material Transfer Cask
4-284

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Top flange, bottom 

flange

SR Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion 

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Screw thread 

insert

SR Stainless steel Embedded

(stainless steel)

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation

embrittlement

Cracking No 

Block SR Stainless steel Embedded

(stainless steel, 

lead)

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Swagelok quick 

connect body, 

coupling, fitting, 

cap

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 
4-285

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Upper trunnion, 

lower trunnion

SR Stainless steel 

(welded)

Air—

indoor/outdoor

Stress corrosion 

cracking

Cracking No 

Demineralized 

water

Stress corrosion 

cracking

Cracking No 

Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Demineralized 

water

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Trunnion retainer, 

trunnion sleeve

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Wear Loss of material Transfer Cask
4-286

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Trunnion retainer, 

trunnion sleeve

SR Stainless steel Demineralized 

water

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Wear Loss of material Transfer Cask

Trunnion bolt SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Bolt for top cover, 

bottom cover, ram 

access cover

SR Steel Air—

indoor/outdoor

General 

corrosion

Loss of material Transfer Cask

Pitting and 

crevice corrosion

Loss of material Transfer Cask

Galvanic 

corrosion

Loss of material Transfer Cask
4-287

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Bolt for top cover, 

bottom cover, ram 

access cover

SR Steel Air—

indoor/outdoor

Microbiologically

influenced 

corrosion

Loss of material No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Stress relaxation Loss of preload No 

Washer for 

trunnion, top 

cover, bottom 

cover, ram access 

cover

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Top cover, ram 

access cover

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 
4-288

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Bottom cover SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Top lifting insert, 

bottom support 

ring

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 

Neutron shield 

plate

SR Stainless steel Air—

indoor/outdoor

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Stress corrosion 

cracking

Cracking No 

Fatigue Cracking Evaluate desig

TLAA, if applic

Radiation 

embrittlement

Cracking No 
4-289

Table 4-23 FuelSolutionsTM transfer cask

Structure, 

System, 

or Component

Intended

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Mana

Neutron shield 

plate

SR Stainless steel Embedded

(RX-277, NS-3)

Radiation 

embrittlement

Cracking No 

Neutron shield SH RX-277, NS-3 Embedded

(stainless steel)

Thermal aging Loss of fracture 

Toughness and 

loss of ductility

No 

Radiation 

embrittlement

Cracking No 

Boron depletion Loss of 

shielding

TLAA/AMP or 

supporting ana

required

Coating on 

neutron

shield jacket

TH Coating Air—

indoor/outdoor

Radiation 

embrittlement

Coating 

degradation

TLAA/AMP or 

supporting ana

required

Thermal aging Coating 

degradation

TLAA/AMP or 

supporting ana

required

Pressure relief 

device

SR Brass Air—

indoor/outdoor

General 

corrosion

Loss of material Transfer Cask

Pitting and 

crevice corrosion

Loss of material No 

Microbiologically

influenced 

corrosion

Loss of material No 

Radiation 

embrittlement

Cracking No 

4-291

1 4.7 Concrete pad

2 The support pad of an ISFSI is a reinforced concrete structure that provides a stable foundation 

3 for the DSSs and transfer equipment. Depending on the design basis of the system or site, the 

4 pad may be within the scope of renewal as an important-to-safety component or as a not5 important-to-safety component, the failure of which could prevent the fulfillment of a fuction that 

6 is important-to-safety. Typically, the concrete pad is exposed to outdoor air and groundwater or 

7 soil environments and is designed and constructed in accordance with ACI codes and 

8 standards.

9 Table 4-24 provides a generic evaluation of potential aging mechanisms and effects requiring 

10 management for the concrete pad. The AMPs that provide an acceptable approach to 

11 managing the aging effects are also identified in the table.
4-292

Table 4-24 Concrete pad

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Manag

Reinforced 

concrete: ISFSI 

pad

SR* Concrete Air—outdoor Aggressive 

chemical attack

Cracking Reinforce

Concrete 

Structure

Loss of strength Reinforce

Concrete 

Structure

Loss of material

(spalling, scaling)

Reinforce

Concrete 

Structure

Reduction of

concrete pH 

(reducing 

corrosion 

resistance of steel 

embedments)

Reinforce

Concrete 

Structure

Creep Cracking No 

Dehydration at 

high temperatures

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Concrete 

Structure

Fatigue Cracking No 

Freeze and thaw Cracking Reinforce

Concrete 

Structure

Loss of material

(spalling, scaling)

Reinforce

Concrete

Structure

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-293

Table 4-24 Concrete pad

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Manag

Reinforced 

concrete: ISFSI 

pad

SR Concrete Air—outdoor Radiation damage Cracking No 

Loss of strength No 

Reaction with

aggregates

Cracking Reinforce

Concrete 

Structure

Loss of strength Reinforce

Concrete 

Structure

Salt scaling Loss of material

(spalling, scaling)

Reinforce

Concrete 

Structure

Shrinkage Cracking No 

Leaching of 

calcium

hydroxide

Loss of strength Reinforce

Concrete 

Structure

Increase in 

porosity

and permeability

Reinforce

Concrete 

Structure

Reduction of

concrete pH 

(reducing 

corrosion 

resistance of steel 

embedments)

Reinforce

Concrete 

Structure

Groundwater/soil Aggressive 

chemical attack

Cracking Reinforce

Concrete 

Structure

Loss of strength Reinforce

Concrete 

Structure
4-294

Table 4-24 Concrete pad

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Manag

Reinforced 

concrete: ISFSI 

pad

SR Concrete Groundwater/soil Aggressive 

chemical attack

Loss of material

(spalling, scaling)

Reinforce

Concrete 

Structure

Reduction of

concrete pH 

(reducing 

corrosion 

resistance of steel 

embedments)

Reinforce

Concrete 

Structure

Creep Cracking No 

Dehydration at 

high temperatures

Cracking No 

Loss of strength No 

Delayed ettringite

formation

Loss of material

(spalling, scaling)

No 

Loss of strength No 

Cracking No 

Differential 

settlement

Cracking Reinforce

Concrete 

Structure

Fatigue Cracking No 

Freeze and thaw Cracking Reinforce

Concrete 

Structure

Loss of material

(spalling, scaling)

Reinforce

Concrete 

Structure

Microbiological

degradation

Loss of strength Reinforce

Concrete 

Structure

Loss of material

(spalling, scaling)

Reinforce

Concrete 

Structure
4-295

Table 4-24 Concrete pad

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Manag

Reinforced 

concrete: ISFSI 

pad

SR Concrete Groundwater/soil Microbiological

degradation

Increase in 

porosity and 

permeability

Reinforce

Concrete 

Structure

Reduction of

concrete pH 

(reducing 

corrosion 

resistance of steel 

embedments)

Reinforce

Concrete 

Structure

Radiation damage Cracking No 

Loss of strength No 

Reaction with

aggregates

Cracking Reinforce

Concrete 

Structure

Loss of strength Reinforce

Concrete 

Structure

Salt scaling Loss of material

(spalling, scaling)

Reinforce

Concrete 

Structure

Shrinkage Cracking No 

Leaching of 

calcium

hydroxide

Loss of strength Reinforce

Concrete 

Structure

Increase in 

porosity

and permeability

Reinforce

Concrete 

Structure
4-296

Table 4-24 Concrete pad

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect

Ag

Manag

Reinforced 

concrete: ISFSI 

pad

SR Concrete Groundwater/soil Leaching of 

calcium

hydroxide

Reduction of

concrete pH 

(reducing 

corrosion 

resistance of steel 

embedments)

Reinforce

Concrete 

Structure

Reinforcing 

steel

Air—outdoor; 

groundwater

Corrosion of

reinforcing steel

Loss of

concrete/steel 

bond

Reinforce

Concrete 

Structure

Loss of material

(spalling, scaling)

Reinforce

Concrete 

Structure

Cracking Reinforce

Concrete 

Structure

Loss of strength Reinforce

Concrete 

Structure
4-297

1 4.8 Spent fuel assemblies

2 4.8.1 Spent fuel assembly description

3 Dry storage systems are designed to store a wide range of SNF assemblies in a dried and 

4 inerted (helium) atmosphere. This section provides a general description of the PWR and BWR 

5 spent fuel assembly components.

6 4.8.2 Fuel cladding and assembly hardware

7 Pressurized-water reactor fuel assemblies

8 While there are a number of fuel assembly design variants for PWRs, the assemblies mainly 

9 consist of the top nozzle, fuel rods, spacer grids, guide thimble tubes, and bottom nozzle. The 

10 various components of a typical 17 × 17 PWR fuel assembly are shown in Figure 4-30. Each 

11 fuel rod consists of enriched uranium dioxide pellets inserted into a cladding tube. The cladding 

12 tube is then capped with Zircaloy end plugs and seal welded at both ends to confine the fuel 

13 pellets and fission gases. The fuel cladding, fabricated from zirconium-based alloys, including 

Zircaloy-4, ZIRLOTM, and M5® 14 , provides a confinement barrier.

15 The structural support of the fuel assembly is provided by the top and bottom nozzles, the 

16 spacer grid assemblies, and the guide thimbles. Guide tubes, fabricated from zirconium-based 

17 alloys, are the main structural members of the fuel assembly. They also provide channels for 

18 neutron absorber rods and burnable poison rods. The bottom of the guide tube is fitted with an 

19 end plug with a flow port, which is then fastened into the bottom nozzle. Both the top and 

20 bottom nozzles are made of either stainless steel or Inconel, which also serve as structural 

21 members of the fuel assembly. The spacer grid assemblies provide support for the fuel 

22 cladding tubes. Two types of grid assemblies, fabricated from zirconium-based alloys or 

23 Inconel, are used in the fuel assemblies. 

24 Boiling-water reactor fuel assemblies

25 Similar to the case for PWRs, there are a number of fuel assembly design variants for BWRs. 

26 The main components include the (i) upper tie plate, (ii) fuel rods, (iii) spacer grids, (iv) water 

27 rods, (v) channel, and (vi) lower tie plate, as shown in Figure 4-31 for the GE14 BWR fuel 

28 assembly in a 10 × 10 fuel rod array. Two types of fuel rods are used in the GE14 fuel bundles: 

29 standard rods and tie rods. The fuel rods are hollow cladding tubes fabricated from Zircaloy-2. 

30 Zircaloy end plugs are welded into place to seal the ends of the fuel rods. The tie rods differ 

31 from the standard fuel rods in that the end plugs are threaded into the tie plates. They hold the 

32 fuel bundle together and support the weight of the fuel bundle during fuel handling operations. 

33 In the BWR fuel assembly, fuel bundles are enclosed in open-ended, square tubes (also called 

34 channels) and are supported at the ends of the fuel bundles by the upper and lower tie plates. 

35 The channels made of zirconium-based alloys are ducts for coolant flow that prevent lateral flow 

36 of coolant among the fuel assemblies operating at different power levels. Both the upper tie and 

37 lower tie plates are fabricated from stainless steel. The upper tie plate provides alignment and 

38 support for the fuel rods at the top of the fuel bundle, while the lower tie plate positions the fuel 

39 rods laterally. The spacer grids, fabricated from zirconium-based alloys or Inconel, hold the fuel 

40 rods in the proper location so that optimum fuel spacing is maintained.

41
4-298

Figure 4-30 Typical pressurized-water reactor fuel assembly (NRC, 2002)

1 
4-299

Figure 4-31 Boiling-water reactor GE14 fuel assembly (GNF, 2005)

1 Table 4-25 provides a generic evaluation of potential aging mechanisms and effects requiring 

2 management for specific components of the SNF assemblies. The AMPs that provide an 

3 acceptable approach to managing the aging effects are also identified in the table.

4
4-300

Table 4-25 Spent fuel assemblies

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Man

Fuel rod cladding CO, CR, 

RE, SH, 

SR, TH* 

Zirconium-based 

alloy (Zircaloy-2, 

Zircaloy-4, 

ZIRLOTM, or M5®) 

Helium Oxidation Loss of load 

bearing 

capacity

No 

Pitting corrosion Loss of material No 

Galvanic 

corrosion

Loss of material No 

Stress corrosion

cracking

Cracking No 

Hydride-induced 

embrittlement

Loss of ductility No 

Delayed hydride 

cracking

Cracking No 

Thermal Creep Changes in 

dimensions

High-Burnu

Monitoring a

Assessmen

Low-temperature 

creep

Changes in 

dimensions

No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 

Mechanical 

overload

Cracking No 

Guide tubes 

(PWR) or water 

channels (BWR)

RE, SR Zirconium-based 

alloy

Helium Creep Changes in 

dimensions

No 

Hydriding Changes in 

dimensions

No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 

Spacer grids CR, RE, 

SR, TH

Zirconium-based 

alloy

Helium Creep Changes in 

dimensions

No 

*Safety Functions: Confinement (CO), Subcriticality (CR), Retrievability (RE), Radiation Shielding (SH), Structural Integrity (SR), Therm
4-301

Table 4-25 Spent fuel assemblies

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Man

Spacer grids CR, RE, 

SR, TH

Zirconium-based 

alloy

Helium Hydriding Changes in 

dimensions

No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 

Inconel Helium Creep Change in 

dimensions

No 

General 

corrosion

Loss of material No 

Stress corrosion

cracking

Cracking No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 

Lower and upper 

end fittings

CR, RE, 

SR

Stainless steel Helium Creep Change in 

dimensions

No 

General 

corrosion

Loss of material No 

Stress corrosion

cracking

Cracking No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 

Inconel Helium Creep Change in 

dimensions

No 

General 

corrosion

Loss of material No 

Stress corrosion

cracking

Cracking No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 
4-302

Table 4-25 Spent fuel assemblies

Structure, 

System, or 

Component

Intended 

Safety

Function Material Environment

Aging 

Mechanism Aging Effect Aging Man

Fuel channel 

(BWR)

CR, TH Zirconium-based 

alloy

Helium Creep Change in 

dimensions

No 

Hydriding Change in 

dimensions

No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 

Poison rod 

assemblies (PWR)

CR Stainless steel Helium Creep Change in 

dimensions

No 

General 

corrosion

Loss of material No 

Stress corrosion

cracking

Cracking No 

Radiation 

embrittlement

Loss of 

strength

No 

Fatigue Cracking No 
4-303

1 4.9 References

2 EPRI. “Industry Spent Fuel Storage Handbook.” Report 1021048. Palo Alto, California: 

3 Electric Power Research Institute. July 2010.

4 FuelSoutions. “FuelSoutions W21 Canister Storage Final Safety Analysis Report.” 

5 Docket No. 72-1026, Revision 5. ADAMS Accession No. ML071510213. April 2007a.

6 _____. “FuelSoutions W74 Canister Storage Final Safety Analysis Report.” 

7 Docket No. 72-1026, Revision 6. ADAMS Accession No. ML071510213. April 2007b.

8 _____. “FuelSoutions Storage System Final Safety Analysis Report.” Docket No. 72-1026. 

9 Revision 4. ADAMS Accession No. ML071510207. April 2007c.

10 GNF. “GE14 Fuel Assembly Mechanical Design Report.” NEDC–33236. 

11 Wilmington, North Carolina: Global Nuclear Fuel. ADAMS Accession No. ML053540338. 

12 2005. 

13 Holtec International. “Final Safety Analysis Report for the HI‐STORM 100 Cask System.” 

14 Non-Proprietary. Docket No. 72-1014. HI-20024444. Rev. 11. ADAMS Accession 

15 No. ML13246A042. August 1, 2013.

16 _____. “Final Safety Analysis Report for the Holtec International Storage, Transport, and 

17 Repository Cask System (HI-STAR 100 Cask System).” HI-2012610, Rev. 0, Vol I of II. 

18 Holtec International. ADAMS Accession No. ML072410190. March 2001.

19 NAC International. “MAGNASTOR, Final Safety Analysis Report,” Non-Proprietary Version. 

20 Docket No. 72-1025. Rev. 15A. ADAMS Accession No. ML15225A469. July 2015.

21 _____. Retrieved from http://www.nacintl.com/magnastor. 2014.

22 _____. “Final Safety Analysis Report for the UMS Universal Storage System.” 

23 Docket No. 72-1015, Rev 3. ADAMS Accession No. ML051290397. March 2004.

24 _____. “Final Safety Analysis Report for the NAC Multi-Purpose Canister System.” 

25 Non-Proprietary. Docket No. 72-1025. Rev. 0. ADAMS Accession No. ML073551117. 

26 April 2000.

27 NRC. “Safety Evaluation Report for License Renewal: Calvert Cliffs Nuclear Power Plant 

28 Independent Spent Fuel Storage Installation.” ADAMS Accession No. ML14274A038 

29 Washington, DC: U.S. Nuclear Regulatory Commission. October, 2014.

30 _____. “Westinghouse Systems Course R–304P.” ADAMS Accession No. ML023030412. 

31 Washington, DC: U.S. Nuclear Regulatory Commission. 2002.

32 _____. “Safety Evaluation Report for the TN‐32 Dry Storage Cask.” Washington, DC: 

33 U.S. Nuclear Regulatory Commission. November 1996.

34
4-304

1 Pacific Nuclear Fuel Services, Inc. “Topical Report for the NUTECH Horizontal Modular 

2 Storage System for Irradiated Nuclear Fuel NUHOMS‐24P.” NUH–002.0103, Rev. 2A, 

3 Volume I. San Jose, California: Pacific Nuclear Fuel Services, Inc. ADAM Accession No. 

4 ML110730769. April 1991.

5 Transnuclear Inc. “Certificate of Compliance Renewal Application for the Standardized 

6 NUHOMS System.” Rev. 0. ADAMS Accession No. ML14309A343. Columbia, Maryland: 

7 Transnuclear, Inc. November 4, 2014. 

8 _____. “Updated Final Safety Analysis Report for the Standardized NUHOMS Modular Storage 

9 System for Irradiated Nuclear Fuel.” NUH–003.0103, Rev. 10, Volume 1 and 3 (Appendix M). 

10 Columbia, Maryland: Transnuclear, Inc. February 2008.

11 _____. “Final Safety Analysis Report for the TN‐68 Dry Storage Cask.” Amendment 1, Rev. 0. 

12 Hawthorne, New York: Transnuclear, Inc. January 2005.

13 _____. “Final Safety Analysis Report for the Standardized NUHOMS Modular Storage System 

14 for Irradiated Nuclear Fuel.” NUH–003.0103, Rev. 8, Volume 1 and 3 (Appendix M). ADAMS 

15 Accession Nos. ML042110421 and ML051040570. Hawthorne, New York: Transnuclear, Inc. 

16 June 2004.

17 _____. “Final Safety Analysis Report for the Standardized Advanced NUHOMS Modular 

18 Storage System for Irradiated Nuclear Fuel.” ANUH–01.0150, Rev. 0, Volume 1. ADAMS 

19 Accession No. ML050410252 (Cover through Section 3.6.3). Fremont, CA: Transnuclear, Inc. 

20 February 2003. 
5-1

1 5 TIME-LIMITED AGING ANALYSES

2 5.1 Introduction

3 Renewal applicants are required to reevaluate all aging-related calculations or analyses 

4 involving time-limited assumptions that were contained in the original design basis (e.g., fatigue 

5 analyses, corrosion wastage calculations). These evaluations are designated as time-limited 

6 aging analyses (TLAAs), and Title 10 of the Code of Federal Regulations (10 CFR) 72.3, 

7 “Definitions,” defines them as those calculations and analyses meeting all six of the following 

8 criteria:

9 (1) Involve SSCs important to safety within the scope of the specific-license renewal, as 

10 delineated in Subpart F of 10 CFR Part 72, or within the scope of the spent fuel storage 

11 CoC renewal, as delineated in Subpart L of 10 CFR Part 72, respectively.

12 (2) Consider the effects of aging.

13 (3) Involve time-limited assumptions defined by the current operating term.

14 (4) Were determined to be relevant by the specific licensee or certificate holder in making a 

15 safety determination.

16 (5) Involve conclusions or provide the basis of conclusions related to the capability of SSCs 

17 to perform their intended safety functions.

18 (6) Are contained or incorporated by reference in the design bases.

19 5.2 Review

20 NUREG-1927, Revision 1, “Standard Review Plan for Renewal of Specific Licenses and 

21 Certificates of Compliance for Dry Storage of Spent Nuclear Fuel” (NRC, 2016), provides 

22 detailed staff guidance for the review of TLAAs.

23 The NRC reviewer should use the final safety analysis report (FSAR) and other documents that 

24 detail the design bases and confirm that the renewal applicant did not omit any TLAAs 

25 submitted as part of the approved design bases. In some cases, the original analyses may 

26 have been performed as part of a code design but not explicitly discussed in the FSAR. Thus, 

27 the reviewer must identify and review any design codes and standards associated with a 

28 storage system to ensure that any required analyses are captured in the applicant’s TLAAs. 

29 Table 5-1 identifies some examples of fatigue analyses that are incorporated into the design 

30 codes for the dry storage system designs this report evaluates. 

31 The reviewer also should ensure that the applicant addresses any design basis calculations that 

32 use materials properties that may be time dependent. For example, aluminum alloys used in 

33 some fuel baskets can lose strength over time at elevated temperatures (see Section 3.2.3.7),

34 and this may affect the performance of the fuel basket in a cask tipover analysis. If the original 

35 design basis calculations did not adequately account for such material property changes

36 through the period of extended operation, the analyses should be updated.

37 The reviewer should ensure that the applicant has appropriately dispositioned an identified 

38 TLAA by using one of the following methods:
5-2 

1 • Demonstrate that the existing analysis remains valid for the period of extended

2 operation, has already considered the requested period of extended operation, and

3 concludes that the structure, system, or component (SSC) will continue to perform its

4 intended function through the end of the requested period of extended operation.

5 • Revise or update the existing analysis to demonstrate that it has been projected to the

6 end of the requested period of extended operation and concludes that the SSC will

7 continue to perform its intended function through the end of the requested period of

8 extended operation.

9 • Manage the effects of aging on the SSC for the requested period of extended operation

10 through an aging management program.

Table 5-1 Examples of fatigue analyses contained within storage system design bases

System SSC

Fatigue Evaluation Criteria

(ASME Code Section III, Division 1 

(ASME, 2007))

Standardized and 

Advanced NUHOMS®*

DSC Confinement NB-3222.4

Transfer Cask NC-3219

HI-STORM 100,

HI-STAR 100

MPC Confinement NB-3222.4

Fuel Basket NG-3222.4

HI-STAR Overpack Helium 

Boundary

NB-3222.4

TN-32 & 68

Confinement Boundary NB-3222.4

Fuel Basket NB-3222.4

NAC-UMS, MPC, and

MAGNASTOR

Canister Confinement NB-3222.4

Fuel Basket NG-3222.4

FuelSolutionsTM

Canister Confinement NB-3222.4

Fuel Basket NG-3222.4

Transfer Cask NC-3219

11

12 5.3 References

13 NRC. NUREG–1927, “Standard Review Plan for Renewal of Specific Licenses and Certificates 

14 of Compliance for Dry Storage of Spent Nuclear Fuel.” Revision 1. Washington, DC: 

15 U.S. Nuclear Regulatory Commission. ADAMS Accession No. ML16179A148. 2016.

16 ASME. Boiler and Pressure Vessel Code, Section III, Division 1, “Rules for Construction of 

17 Nuclear Facility Components,” Division 1, Subsection NB, “Class 1 Components,” Subsection 

18 NC, “Class 2 Components,” and Subsection NG, “Core Support Structures”; American Society 

19 of Mechanical Engineers. 2007.
6-1

1 6 EXAMPLE AGING MANAGEMENT PROGRAMS

2 6.1 Introduction

3 The example aging management programs (AMPs) presented in this chapter and listed in

4 Table 6-1 below describe a generically acceptable approach to managing the credible aging 

5 effects that were identified in the technical bases discussions in Chapter 3 and the aging 

6 management review tables in Chapter 4. AMPs monitor and control the degradation of 

7 structures, systems, and components (SSCs) within the scope of renewal, so that aging effects 

8 will not result in a loss of intended functions during the period of extended operation. An AMP 

9 includes all activities that are credited for managing aging mechanisms or effects for specific 

10 SSCs. An effective AMP prevents, mitigates, or detects the aging effects and provides for the 

11 prediction of the extent of the effects of aging and timely corrective actions before there is a loss 

12 of intended function. 

13 If an applicant credits these generic AMPs in the renewal application, the NRC staff should 

14 ensure that the applicant demonstrates that the design features, environmental conditions, and 

15 operating experience for the subject independent spent fuel storage installation (ISFSI) or dry 

16 storage system (DSS) are bounded by those evaluated in this report. Otherwise, the staff 

17 should ensure that the applicant augments the AMPs as appropriate to address the impact of 

18 unique design or operating parameters.

Table 6-1 Example aging management programs

Section AMP

6.5 Localized Corrosion and Stress Corrosion Cracking of Welded Stainless Steel 

Dry Storage Canisters

6.6 Reinforced Concrete Structures

6.7 External Surfaces Monitoring of Metallic Components

6.8 Ventilation Systems

6.9 Bolted Cask Seal Leakage Monitoring

6.10 Transfer Casks

6.11 High-Burnup Fuel Monitoring and Assessment

19 6.2 Alternative approaches

20 An applicant may propose alternative approaches to manage the effects of aging. In its review 

21 of alternative AMPs, the staff should use the guidance in NUREG–1927, Revision 1, “Standard 

22 Review Plan for Renewal of Specific Licenses and Certificates of Compliance for Dry Storage of 

23 Spent Nuclear Fuel” (NRC, 2016). As described in greater detail in NUREG–1927, an AMP 

24 generally should contain the following 10 elements:

25 (1) Scope of program: the specific SSCs and subcomponents covered by the AMP and the 

26 intended functions to be maintained, in addition to stating the specific materials, 

27 environments, and aging mechanisms and effects to be managed
6-2 

1 (2) Preventive actions: actions to prevent aging or mitigate the rates of aging for SSCs

2 (3) Parameters monitored or inspected: the specific parameters that will be monitored or 

3 inspected and a description of how those parameters will be capable of identifying 

4 degradation before a loss of intended function

5 (4) Detection of aging effects: the inspection and monitoring details, including method or 

6 technique, frequency, sample size, data collection, and timing of inspections

7 (5) Monitoring and trending: how data will be evaluated and trended to ensure timely 

8 corrective actions

9 (6) Acceptance criteria: the criteria against which the need for corrective action will 

10 be evaluated

11 (7) Corrective actions: The measures to be taken when the acceptance criteria are not met, 

12 including root cause determination and prevention of recurrence, as appropriate

13 (8) Confirmation process: processes in place to verify that preventive actions are adequate 

14 and that appropriate corrective actions have been completed and are effective

15 (9) Administrative controls: processes in place that provide a formal review and approval 

16 process for activities related to the AMP (e.g., inspector requirements, instrument 

17 calibration) 

18 (10) Operating experience: a review of operational experience that supports the 

19 determination that the AMP is capable of maintaining SSC functions in the period of 

20 extended operation 

21 The reviewer should examine the applicant’s proposed 10 elements to verify that the program is 

22 capable of managing the specific aging mechanisms and effects identified by the aging 

23 management review (AMR). The reviewer should recognize that an applicant may develop 

24 AMPs following a different format or style. For such reviews, the NRC staff should ensure that 

25 sufficient detail (i.e., supporting technical bases) is provided in the alternative format in 

26 comparison with the 10 AMP elements of this guidance.

27 An applicant may credit existing site maintenance and inspection activities to manage the 

28 effects of aging. In such cases, the reviewer should ensure that the design basis 

29 documentation describes those activities with sufficient detail ensure that the 10 AMP elements 

30 are fully addressed.

31 6.3 Learning aging management

32 As described in NUREG–1927, the reviewer should ensure that the application includes 

33 provisions to conduct periodic future reviews of operating experience to confirm the 

34 effectiveness of the AMPs or identify a need to enhance or modify an AMP. The reviewer also 

35 should verify that the applicant: (1) references a specific system to be used to obtain, 

36 aggregate, and enter site-specific, design-specific, and industrywide operating experience, and 

37 (2) discusses how it intends to provide timely reporting of operating experience to this system.
6-3 

1 If an applicant follows this approach, the reviewer should ensure that the description of the 

2 periodic assessments includes specific performance criteria (e.g., program-specific performance 

3 indicators for each of the 10 AMP elements) and proposed actions based on the assessment 

4 findings. The reviewer should also ensure that the timing of the assessments appropriately 

5 considers the rate of aging degradation and the anticipated availability of data from industry 

6 initiatives. 

7 Nuclear Energy Institute (NEI) 14-03, “Format, Content, and Implementation Guidance for Dry 

8 Cask Storage Operations-Based Aging Management,” Revision 2, provides a proposed 

9 framework for learning AMPs through the use of “tollgates” (NEI, 2016). NEI 14-03 defines 

10 “tollgates” as periodic points within the period of extended operation when licensees would be 

11 required to evaluate aggregate feedback and perform and document a safety assessment that 

12 confirms the safe storage of spent fuel. At the time of publication of this report, the NRC staff 

13 was continuing its review of NEI 14-03, Revision 2, for proposed NRC endorsement. However, 

14 until a time when NEI 14-03 may be endorsed by the NRC, Section 3.6.1.10 of NUREG–1927, 

15 Revision 1, provides guidance to reviewers on ensuring AMP effectiveness. 

16 6.4 References 

17 NEI. “Format, Content and Implementation Guidance for Dry Cask Storage Operations-Based 

18 Aging Management for Dry Cask Storage.” NEI 14-03, Rev. 2. ADAMS Accession 

19 No. ML16356A210. Nuclear Energy Institute. 2016. 

20 NRC. NUREG–1927, “Standard Review Plan for Renewal of Specific Licenses and Certificates 

21 of Compliance for Dry Storage of Spent Nuclear Fuel.” Revision 1. Washington, DC: 

22 U.S. Nuclear Regulatory Commission. 2016. 

6-5

1 6.5 Localized Corrosion and Stress Corrosion Cracking of Welded Stainless 

2 Steel Dry Storage Canisters

3 Welded stainless steel canisters are used in the majority of the DSSs in the United States for 

4 spent nuclear fuel (SNF) from commercial power reactors at both specific-licensed and 

5 general-licensed ISFSIs. The welded stainless steel canisters are the primary confinement 

6 boundary during storage. While there are no known operational occurrences of aging or 

7 localized corrosion of welded stainless steel canisters, operational experience with nuclear 

8 reactors that were located close to an open ocean or bay has shown that pitting corrosion, 

9 crevice corrosion, and chloride-induced stress corrosion cracking (CISCC) can occur in welded 

10 stainless steel components as a result of atmospheric deposition and deliquescence of 

11 chloride-containing salts. Laboratory and natural exposure tests suggest that CISCC can occur 

12 with sufficient surface chloride concentrations and that, with those concentrations of chloride, 

13 crack propagation rates can be of engineering significance for welded stainless steel canisters 

14 during the period of extended operation.

15 Based on reactor operating experience, as well as laboratory and field testing, localized 

16 corrosion and CISCC are potential aging mechanisms for welded stainless steel canisters. 

17 Environments where chloride-containing salts may be deposited on welded stainless steel 

18 canisters include coastal locations near salt water and locations that are close to cooling towers 

19 or roads that are salted. The Electric Power Research Institute (EPRI) has developed aging 

20 management guidance to address the potential for CISCC of welded stainless steel canisters 

21 (Fuhr et al., 2017). In addition, the American Society of Mechanical Engineers Boiler and 

22 Pressure Vessel Code (ASME Code), Section XI, has formed a task group to develop a code 

23 case to establish the requirements for inservice inspection and acceptance criteria for DSS

24 canisters (Code Case N-860) that may follow the recommendations of the EPRI aging 

25 management guidance. However, the development of a consensus-based code case for 

26 inservice inspection of DSS canisters may take several years to complete. To address potential 

27 aging effects as a result of localized corrosion cracking and stress corrosion cracking (SCC) in 

28 the absence of an acceptable code case, the NRC has provided an example AMP for welded 

29 stainless steel canisters used in DSSs that relies on guidance from consensus codes for 

30 inservice inspection of nuclear power plant components. Elements of an NRC staff-developed 

31 example AMP are described in Table 6-2.

32

33
6-6 

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

1. Scope of

Program

Inspection of welded stainless steel dry storage canister confinement 

boundary external surfaces for atmospheric deposits, localized 

corrosion, and SCC. 

Examinations should be focused on areas with the following 

attributes:

• canister fabrication welds and weld heat affected zones

• closure welds and weld heat affected zones

• areas of the canister to which temporary supports or

attachments were attached by welding and subsequently

removed

• locations where a crevice is formed on the canister surface

• horizontal (±30°) surfaces where deposit accumulation may

accumulate at a faster rate compared to vertical surfaces

• canister surfaces that are cold relative to the average surface

temperature

• canister surfaces with higher amounts of atmospheric deposits

Effort should be made to identify and prioritize examinations of areas 

on canisters that have two or more of the above attributes 

(e.g., canister surface that is cold relative to average surface 

temperature and also has a weld or weld heat affected zone). 

2. Preventive

Actions

None; AMP is for condition monitoring. However, DSS canister 

designs may include preventive actions such as fabrication 

procedures and surface modification methods to impart compressive 

residual stresses on the canister welds and weld heat-affected zones 

to reduce the potential for SCC. Preventive actions may also include 

the use of DSS canister confinement boundary materials that are 

resistant to localized corrosion and SCC. For such cases the 

preventive actions described should be supported with an analysis 

and data demonstrating the preventive actions are effective.

3. Parameters

Monitored/

Inspected

Parameters monitored or inspected should include:

• visual evidence of discontinuities and imperfections such as

localized corrosion, including pitting corrosion, crevice corrosion

and SCC of the canister welds and weld heat-affected zones

• size and location of localized corrosion and SCC

• appearance and location of deposits on the canister surfaces

4. Detection of

Aging Effects

Visually examine deposits on the canister surfaces and identify 

corrosion products that may be indicators of localized corrosion and 

SCC in the welds and weld heat-affected zones. Visual examination 

instrumentation with demonstrated sizing and depth measurement 
6-7 

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

capability may be useful in the determination of the size and depth of 

pits open to the surface. Visual examination may also detect the 

presence of cracks originating from pits. However, the ability to 

detect cracks on clean metal surfaces using visual examination 

methods is dependent on several factors and can be difficult for tight 

crack opening displacements (Cumblidge et al., 2004, 2007). The

presence of significant corrosion product accumulation may also

interfere with the identification of SCC using visual examination

methods. 

Volumetric examination is necessary to characterize SCC. 

Volumetric examination of pits and areas immediately adjacent to pits 

is necessary when pits are located within 25 mm [1 in] of a through 

thickness weld or within 25 mm [1 in] of an area where an temporary 

attachment was known to be located.

Visual Examination

Pitting and crevice corrosion that is open to the surface can 

potentially be detected by visual testing (ASME Code Section V, 

Table A-110). Because of the high neutron and gamma radiation 

fields near the surface of the stainless steel dry storage canisters, 

direct visual examination is not possible. Procedures for remote 

visual examination should be performance demonstrated; procedure 

attributes, for example, equipment resolution and lighting 

requirements, should reference applicable standards, such as ASME

Code Section XI, Article IWA-2200 for VT-1 and VT-3 examinations 

(ASME, 2007) and BWRVIP-03 (Selby, 2005) for EVT-1 

examinations.

Volumetric Examination

Additional assessment is necessary for suspected areas of localized 

corrosion and SCC. In these cases, the severity of degradation must 

be assessed, including the dimensions of the affected area and the 

depth of penetration with respect to the thickness of the canister. For 

accessible areas where adequate cleaning can be performed, remote 

visual examination meeting the requirements for VT-1 Examination 

(ASME Code Section XI, IWA-2211) may be used to determine the 

type of degradation present (e.g., pitting corrosion or SCC) and the 

location of degradation. Examinations to characterize the extent and 

severity of localized corrosion and SCC should be conducted using 

surface or volumetric examination methods consistent with the 

requirements of ASME Code Section XI, IWB-2500, for category B-J 

components (ASME, 2007).
6-8 

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

Sample Size

For sites where inspections are necessary, there should be a 

minimum of one canister at each site. Preference should be given to 

the canisters with the greatest susceptibility for localized corrosion or 

SCC. Factors to be considered include older and colder canisters 

with the greatest potential for the accumulation and deliquescence of 

deposited salts that may promote localized corrosion and SCC, types 

of systems used at the site, canister location with respect to potential 

sources of atmospheric deposits, system design, and operational 

experience. Industry guidance on evaluating susceptibility has been 

published by the EPRI (Fuhr et al., 2015). 

Justification for not conducting inspections for localized corrosion or 

SCC should be provided on a case-by-case basis for each ISFSI site 

where welded stainless steel canisters are in use. Acceptable 

justification may be based on a comparison of susceptibility for the 

ISFSI location versus at least two other ISFSI sites determined to 

have greater susceptibility but that showed no evidence of localized 

corrosion or SCC in inspections completed within 5 years of the time 

of the assessment. The justification must consider the full range of 

available ISFSI susceptibility assessments and welded stainless steel 

canister examination results. 

Data Collection

Canister Examination: documentation of the examination of the 

canister, location, and appearance of deposits and an assessment of 

the suspect areas where corrosion products were observed as 

described in corrective actions

Bounding Analysis: a complete listing of other sites considered, 

susceptibility assessments for those sites, and results of 

examinations conducted at those sites, as well as a justification for 

not including other sites where examinations showed evidence of 

localized corrosion or SCC 

Frequency

The frequency of inspection should be determined based on the 

localized corrosion and SCC susceptibility of both the site and the 

canisters in service, aggregated operational experience of similar 

storage system canisters and previous site specific examination 

results 
6-9 

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

Timing of Inspections

The timing of the inspections includes the preapplication inspection or 

general-licensee baseline inspection, performed per Sections 3.4.1.2 

and 3.6.1.10 of NUREG–1927, Revision 1, and at the frequency 

specified by the AMP.

Alternative detection methods or techniques may be provided. For

these cases:

• The method or technique should be adequate and proven to be

capable of evaluating the condition of the external surface of the

canister against the acceptance criteria for the detection of

localized corrosion and SCC.

• The proposed intervals for inspection or monitoring are

consistent with applicable site-specific, design-specific, or

industrywide operating experience and should have sufficient

frequency to ensure that the confinement function will be

maintained until the next scheduled inspection.

• The data collection methods should be sufficient for evaluating

localized corrosion and SCC and should reference specific

methods to be used for data acquisition, including any

applicable consensus codes and standards.

5. Monitoring and

Trending

Monitoring and trending methods are in accordance with ASME Code 

Section XI evaluation criteria. 

Monitoring and trending methods reference plans/procedures are

used to do the following: 

• Establish a baseline before or at the beginning of the period of

extended operation

• Track trending of parameters or effects not corrected following a

previous inspection including

— the locations and size of any areas of localized corrosion 

or SCC

— the disposition of canisters with identified aging effects 

and the results of supplemental canister inspections

Monitoring and trending should also include:

• the appearance of the canister, particularly at welds and in

crevice locations, documented with images and video that will

allow comparison in subsequent examinations
6-10

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

• changes to the size and number of any rust-colored stains as a

result of iron contamination of the surface in subsequent

inspections

6. Acceptance

Criteria

No indications of localized corrosion pits, etching, crevice corrosion, 

SCC, red-orange-colored corrosion products emanating from crevice 

locations, or red-orange-colored corrosion products in the vicinity of 

canister fabrication welds, closure welds, and welds associated with 

temporary attachments during canister fabrication.

Flaws identified must be assessed in accordance with the acceptance 

standards identified in ASME B&PV Code Section XI, IWB-3514. 

Indications Requiring Additional Evaluation

Confirmed or suspected areas of crevice corrosion, pitting corrosion, 

and SCC must be assessed in accordance with the acceptance 

criteria identified in ASME B&PV Code Section XI, IWB-3640.

Although shop and handling procedures include controls to prevent 

iron contamination of the stainless steel surfaces, contamination does 

occur and is usually identified by rust-colored surface deposits. Iron 

contamination can exacerbate CISCC in stainless steels. In 

accessible locations, removal of the deposits and rust stains that 

reveal undamaged welds (i.e., absence of pits, crack, localized 

attack, or etching) and the original machining/grinding marks on the 

stainless steel base metal, including weld heat-affected zones, may 

be used to confirm that localized corrosion or SCC has not been 

initiated. 

Indications of interest that are subject to additional examination and 

disposition include: 

• localized corrosion pits, crevice corrosion, SCC, and etching

(note that these indications may be covered by obstructions

(i.e., crevices)); deposits; or corrosion products

• discrete red-orange-colored corrosion products that are 1 mm

[0.039 in] in diameter or larger, in SCC susceptible locations on

the canister surface that includes areas adjacent to fabrication

welds, closure welds, locations where temporary attachments

may have been welded to and subsequently removed from the

stainless steel dry storage canister, and the weld heat-affected

zones

• linear appearance of any color of corrosion products of any size

parallel to or traversing fabrication welds, closure welds,

locations where temporary attachments may have been welded

to and subsequently removed from the stainless steel dry
6-11

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

storage canister, and the weld heat-affected zones of these 

areas 

• red-orange-colored corrosion products greater than 1 mm

[0.039 in] in diameter or red-orange-colored corrosion tubercles 

of any size combined with deposit accumulations in SCC 

susceptible locations on the stainless steel canister 

• red-orange-corrosion products present at the mouth of a crevice

that includes a portion of the canister surface

Alternative acceptance criteria may be provided. For such cases, the 

acceptance criteria should:

• include a quantitative basis (justifiable by operating experience,

engineering analysis, consensus codes/standards)

• avoid the use of nonquantifiable phrases (e.g., significant,

moderate, minor, little, slight, few)

• be achievable and clearly actionable

7. Corrective

Actions

Results that do not meet the acceptance criteria are addressed as 

conditions adverse to quality or significant conditions adverse to 

quality under those specific portions of the specific- or generallicensee quality assurance (QA) program approved under 

10 CFR Part 72, Subpart G, or 10 CFR Part 50, Appendix B, 

respectively. The QA program ensures that corrective actions are 

completed within the specific- or general-licensee’s Corrective Action 

Program (CAP), and include provisions to

• perform functionality assessments

• perform apparent cause evaluations and root cause evaluations

• address the extent of condition

• determine actions to prevent recurrence for significant

conditions adverse to quality; ensure justifications for nonrepairs

• trend conditions

• identify operating experience actions, including modification to

the existing AMP (e.g., increased frequency)

• determine if the condition is reportable to the NRC per

10 CFR 72.75

Extent of Condition

Confirmation of localized corrosion or SCC may warrant inspection of 

additional canisters at the same ISFSI location to determine the 

extent of condition. Priority for additional inspections should be to 

canisters with similar time in service and initial loading. Canisters 

with confirmed localized corrosion or SCC must be evaluated for 

continued service. Canisters with localized corrosion or SCC that do 
6-12

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

not meet the prescribed evaluation criteria are not permitted to 

remain in service without an engineering analysis or mitigation 

actions. 

8. Confirmation

Process

The confirmation process will be commensurate with the specific or 

general licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B. The QA program 

ensures that the confirmation process includes provisions to preclude 

repetition of significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine followup actions to verify effective implementation of

corrective actions

• monitor for adverse trends due to recurring or repetitive findings

or observations

9. Administrative

Controls

The administrative controls are in accordance with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that administrative controls include provisions that 

define: 

• instrument calibration and maintenance

• inspector requirements

• record retention requirements

• document control

The administrative controls describe or reference: 

• methods for reporting results to NRC per 10 CFR 72.75

• frequency for updating an AMP based on site-specific,

design-specific, and industrywide operational experience

10. Operating

Experience

The AMP references and evaluates applicable operating experience,

before renewal, and will continue to do so as new operating 

experience is developed and made available after renewal, including: 

• internal and industrywide condition reports

• internal and industrywide corrective action reports

• vendor-issued safety bulletins

• NRC generic communications

• applicable U.S. Department of Energy (DOE) or industry

initiatives (e.g., EPRI- or DOE-sponsored inspections)

The AMP clearly identifies any degradation in the referenced 

operating experience as either age related or event driven, with 
6-13

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

proper justification for that assessment. Past operating experience 

supports the adequacy of the proposed AMP, including the 

method/technique, acceptance criteria, and frequency of inspection.

The AMP references the methods for capturing operating experience 

from other ISFSIs with similar in-scope SSCs.

CISCC of austenitic stainless steels is a known degradation 

mechanism for aqueous environments; however, operational 

experience in aqueous environments is not directly applicable in 

assessing the potential for atmospheric CISCC for austenitic stainless 

steel dry storage canisters. Atmospheric CISCC of austenitic 

stainless steels has been reported in a range of industries, including 

welded stainless steel components and piping in operating nuclear 

power plants.

Spent Fuel Storage

Inspections of dry storage canisters after 20 years in service have 

been conducted at a few ISFSI sites. Details of the inspection 

conducted at nuclear power plant ISFSIs are documented in EPRI 

and Sandia National Laboratories reports (Waldrop et al., 2016; 2014; 

Bryan and Enos, 2014). No evidence of localized corrosion was 

identified but some amount of chloride-containing salts were 

determined to be present and corrosion products believed to be 

related to iron contamination were identified at the Calvert Cliffs 

ISFSI. 

Operating Power Reactors

NRC Information Notice 2012-20 (NRC, 2012) documents previous 

cases of atmospheric CISCC of welded stainless steel piping systems 

and tanks at operating reactor locations. Atmospheric CISCC growth 

rates determined from operational experience at both domestic and 

foreign nuclear power plants, including events at San Onofre, Turkey 

Point, St. Lucie, and Koeberg (South Africa), range from 3.6 × 

10−12 m/sec to 2.9 × 10−11 m/sec for components at ambient 

temperatures.

Relevant Literature and Testing

EPRI has recently conducted a literature review of CISCC that 

summarizes the results of many previous laboratory investigations 

(Gorman et al., 2014). 

The NRC has recently published the results of a completed 

investigation of CISCC testing of type 304, 304L, and 316L stainless 
6-14

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

steel and welds (He et al., 2014). This study indicates that SCC was 

initiated at stresses just above the yield strength in tests conducted 

using 304 stainless steel C-ring specimens. Testing with U-bend 

specimens showed that CISCC was observed with the lowest 

simulated sea salt concentrations tested (100 mg salt/m2 or ~55 mg 

chloride/m2

) at temperatures of 52 degrees C [125.6 degrees F] using 

a maximum absolute humidity of 30 g/m3

, which is generally accepted 

as being near the maximum absolute humidity in a natural 

environment. 

Both laboratory and field investigations have been conducted by 

CRIEPI and TEPCO. This includes the early work by Tokiwai et al. 

(1985), who reported the critical surface chloride concentrations of 

8 mg/m2 for CISCC on sensitized 304 stainless steel. Kosaki (2008) 

reported crack growth rates of 9.6 × 10−12 m/sec obtained in natural 

exposure tests on Miyakojima Island with type 304 base metals and 

welds, type 304L welds, and type 316LN welds. Hayashibara et al. 

(2008) reported activation energy for crack growth in type 304 

stainless steel of 5.6 to 9.4 kcal/mol [23 to 39 kJ/mol], based on 

testing conducted at temperatures of 50 to 80 degrees C [122 to 

176 degrees F].

References ASME. “Boiler and Pressure Vessel Code Section XI—Rules for 

Inservice Inspection of Nuclear Power Plant Components.” 

New York, New York: American Society of Mechanical Engineers. 

2007.

Bryan, C.R. and D.G. Enos. SAND2014-16383, “Analysis of Dust 

Samples Collected From Spent Nuclear Fuel Interim Storage 

Containers at Hope Creek, Delaware, and Diablo Canyon, California.” 

Albuquerque, New Mexico: Sandia National Laboratories. July 2014.

Cumblidge, S.E., M.T. Anderson, and S.R. Doctor. NUREG/CR6860, “An Assessment of Visual Testing.” ADAMS Accession 

No. ML043630040. Richland, Washington. Pacific Northwest 

National Laboratory. 2004.

Cumblidge, S.E., M.T. Anderson, S.R. Doctor, F.A. Simonen, and 

A.J. Elliot. NUREG/CR-6943, “A Study of Remote Visual Methods to 

Detect Cracking in Reactor Components.” ADAMS Accession 

No. ML073110060. Richland, Washington. Pacific Northwest 

National Laboratory. 2007.

Fuhr, K., J. Broussard, and G. White. “Susceptibility Assessment 

Criteria for Chloride-Induced Stress Corrosion Cracking (CISCC) of 

Welded Stainless Steel Canisters for Dry Cask Storage Systems.” 
6-15

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

EPRI-3002005371. Palo Alto, California: Electric Power Research 

Institute. 2015.

Fuhr, K., J. Broussard, and G. White. “Aging Management Guidance 

to Address Potential Chloride-Induced Stress Corrosion Cracking of 

Welded Stainless Steel Canisters,” EPRI-3002008193. Palo Alto, 

California: Electric Power Research Institute. 2017.

Gorman, J., K. Fuhr, and J. Broussard. “Literature Review of 

Environmental Conditions and Chloride-Induced Degradation 

Relevant to Stainless Steel Canisters in Dry Cask Storage Systems.” 

EPRI-3002002528. Palo Alto, California: Electric Power Research 

Institute. 2014.

Hayashibara, H., M. Mayuzumi, Y. Mizutani, and J. Tani. “Effect of 

Temperature and Humidity on Atmospheric Stress Corrosion 

Cracking of Stainless Steel.” Corrosion 2008. Paper 08492, 

Houston, Texas: NACE International. 2008. 

He, X., T.S. Mintz, R. Pabalan, L. Miller, and G. Oberson. 

“Assessment of Stress Corrosion Cracking Susceptibility for 

Austenitic Stainless Steels Exposed to Atmospheric Chloride and 

Non-Chloride Salts.” NUREG/CR-7170. ADAMS Accession 

No. ML14051A417. Washington, DC. U.S. Nuclear Regulatory 

Commission, February 2014, 

Kosaki, A. “Evaluation Method of Corrosion Lifetime of Conventional 

Stainless Steel Canister Under Oceanic Air Environment.” 

Nuclear Engineering and Design. Vol. 238. pp.1,233–1,240. 2008.

NRC. “Information Notice 2012-20: “Potential Chloride-Induced 

Stress Corrosion Cracking of Austenitic Stainless Steel and 

Maintenance of Dry Cask Storage System Canisters.” ADAMS 

Accession No. ML12319A440. Washington, DC: U.S. Nuclear 

Regulatory Commission. 2012.

Selby, G. “BWR Vessel and Internals Project, Reactor Pressure 

Vessel and Internals Examination Guidelines.” EPRI 1011689, TR105696-R8 (BWRVIP-03) Rev. 8. Palo Alto, California: Electric 

Power Research Institute. 2005.

Tokiwai, M., H. Kimura, and H. Kusanagi. “The Amount of Chlorine 

Contamination for Prevention of Stress Corrosion Cracking in 

Sensitized Type 304 Stainless Steel.” Corrosion Science. Vol. 25, 

Issue 8–9. pp. 837–844. 1985.
6-16

Table 6-2 Example aging management program for Localized Corrosion And Stress

 Corrosion Cracking Of Welded Stainless Steel Dry Storage Canisters 

Element Description

Waldrop, K., C. Bryan, D. Enos, “Diablo Canyon Stainless Steel Dry 

Storage Canister Inspection,” EPRI-3002002822, Palo Alto, CA: 

EPRI, 2016.

Waldrop, K., W. Bracey, K. Morris, C. Bryan, and D. Enos. “Calvert 

Cliffs Stainless Steel Dry Storage Canister Inspection.” EPRI1025209. Palo Alto, California: Electric Power Research Institute. 

2014.

1 
6-17

1 6.6 Reinforced Concrete Structures

2 An example AMP for reinforced concrete structures is provided below. The AMP consists of 

3 condition monitoring, performance monitoring, and mitigation and prevention activities. The 

4 program includes periodic visual inspections by personnel qualified to monitor reinforced 

5 concrete for applicable aging effects, such as those described in the American Concrete 

6 Institute (ACI) guides ACI 349.3R-02, ACI 201.1R-08, and American National Standards 

7 Institute/American Society of Civil Engineers guidelines (ANSI/ASCE) 11-99. Identified aging 

8 effects are evaluated against acceptance criteria derived from the design bases or industry 

9 guides and standards, including ACI 349, ACI 318, ACI 349.3R-02 and ASME Code Section XI, 

10 Subsection IWL.

11 The program also includes periodic sampling and testing of groundwater and the need to 

12 assess the impact of any changes in its chemistry on below-grade concrete structures. 

13 Additional activities include radiation surveys to ensure the shielding functions of the concrete 

14 structure are maintained and daily inspections to ensure the air convection vents are not 

15 blocked (per the requirements of the approved design bases). The program also includes 

16 provisions where modifications may be appropriate.

17
6-18

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

1. Scope of Program The scope of the program includes the following aging management

activities:

1. visual inspection of above-grade (readily accessible, normally

inaccessible) and below-grade (underground) concrete areas

(see Element 4 for sample size and justification of areas to be

inspected)

2. groundwater chemistry monitoring program to identify

conditions conducive to the following below-grade

(underground) aging mechanisms:

• corrosion of embedded steel

• chemical attack (chloride- and sulfate-induced

degradation)

3. radiation surveys1 to:

• ensure compliance with 10 CFR 72.104 (i.e., dose

equivalent requirements beyond the controlled area

during normal and off-normal conditions of storage)

• monitor performance of the concrete as a

neutron/gamma shield at near-system locations as an

indicator of concrete degradation

The program provides means to adequately identify the following 

aging effects, as described in ACI 349.3R-02 (ACI, 2010) and 

SEI/ASCE 11-99 (SEI/ASCE, 2000):

• cracking or loss of material (spalling, scaling) due to Freeze

and thaw degradation

• cracking, loss of material (spalling, scaling), loss of strength

and reduction of concrete pH (corrosion resistance of steel

reinforcement) due to aggressive chemical attack

• cracking and loss of strength due to reaction with aggregates

• cracking, loss of material, and loss of strength due to corrosion

of embedded steel

• increase in porosity/permeability, loss of strength, and

reduction in concrete pH due to leaching of calcium hydroxide

• cracking due to differential settlement

• loss of material (spalling, scaling) due to salt scaling

• loss of material (spalling, scaling), loss of strength, increased

porosity and permeability, and reduction in concrete pH

1The NRC reviewer should consider the design features of the DSS when determining if radiation surveys can be 

excluded from the scope of this AMP on a case-by-case basis. For example, internal surfaces of a concrete 

overpack may be permanently blocked by a steel liner, which may prevent assessing the condition of those concrete 

surfaces by remote visual inspection. The NRC reviewer should evaluate any engineering justification and/or 

operating experience to determine if visual inspections of readily accessible and normally inaccessible (i.e., not 

permanently blocked) surfaces can adequately characterize the condition of the structure and provide reasonable 

assurance that the intended functions are maintained during the period of extended operation.
6-19

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

(corrosion resistance of steel reinforcement) due to 

microbioligical degradation

Additional site-specific AMPs may be required for the following 

scenarios:

• A dewatering system is used to prevent long-term settlement.

• The design bases include embedded aluminum 

subcomponents without a protective insulating coating.

• Protective coatings are relied upon to manage the effects of 

aging for a subcomponent.

2. Preventive Actions Preventive actions include continuance of inspections to ensure that 

air inlet/outlet vents are not blocked and/or temperature monitoring, 

if applicable, to ensure design temperature limits are not exceeded

(see Section 6.8, AMP on Ventilation Systems). These inspections 

would be part of the approved design bases and be continued for 

the sample size and inspection frequency identified in the 

respective technical specification (TS).

Additional preventive actions are not required for structures 

designed and fabricated in accordance with ACI 318 (ACI, 2011) or 

ACI 349 (ACI, 2007a), as specified in the design bases. Otherwise, 

a site-specific AMP may be required.

3. Parameters 

Monitored or 

Inspected

For visual inspections, the parameters monitored or inspected 

quantify the following aging effects:

• cracking

• loss of material (spalling, scaling)

• loss of bond

• increased porosity/permeability

AMP procedures reference the following parameters for 

characterizing the above aging effects, as appropriate, per the 

acceptance criteria:

2

• affected surface area

• geometry/depth of defect

• cracking, crazing, delaminations, drummy areas

• curling, settlements or deflections

• honeycombing, bug holes

• popouts and voids

• exposure of embedded steel

• staining/ evidence of corrosion

 

2The terminology is consistent with ACI standard CT-13 (ACI, 2013b).
6-20

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

• dusting, efflorescence of any color

The parameters evaluated consider any surface geometries that 

may support water ponding and potentially increase the rate of 

degradation.

For the groundwater chemistry program, the parameters monitored

or inspected include:

• water pH

• concentration of chlorides and sulfates in the water

For radiation surveys, the parameters monitored or inspected 

include gamma dose rate and neutron fluence rate.

4. Detection of Aging

Effects

Method or technique

Visual inspections of readily accessible areas are performed with 

feeler gauges, crack comparators, or other suitable visual 

quantification methods per the acceptance criteria in ACI 349.3R-02 

(ACI, 2010).

Visual inspections of normally inaccessible areas are performed 

using a remote inspection system that has been qualified for the 

specific DSS and site-specific characteristics. Procedures for 

remote visual inspections should be demonstrated to ensure the 

acceptance criteria in ACI 349.3R-02 (ACI, 2010) are achievable; 

procedure attributes should include, for example, equipment 

resolution and lighting requirements and should reference 

applicable standards when possible.

Groundwater chemistry is characterized using a chemical analysis 

method with a valid measurement range and adequate resolution 

and sensitivity. Procedures for groundwater chemistry analyses 

should be demonstrated to ensure the acceptance criteria in ASME 

Code Section XI, Subsection IWL, are achievable

Radiation surveys are performed using calibrated neutron and 

gamma detectors with valid energy ranges, per the acceptance 

criteria (see Element 6).

Procedure attributes for all inspection and monitoring activities 

within the scope of this program should be commensurate with 

10 CFR 72.164 and 10 CFR Part 50, Appendix B, as appropriate.
6-21

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

Frequency of Inspection

The schedule for visual inspections is commensurate with 

ACI 349.3R-02 (ACI, 2010). Alternative inspection frequencies 

must be adequately justified by a valid technical basis (engineering 

justification, operational experience data).

Inspections of above-grade (both readily accessible and normally

inaccessible) areas are conducted at least once every 5 years. The 

inspections of below-grade (underground) areas are opportunistic; 

inspections are performed when excavations occur for any reason.

The frequency for monitoring groundwater chemistry is justified 

(e.g., quarterly, semiannually), per an adequate technical basis 

(site-specific operating experience, engineering justification).

The frequency for radiation surveys is justified (e.g., quarterly), per 

an adequate technical basis (engineering justification, operating 

experience).

Sample size

Visual inspections cover 100 percent of readily accessible surfaces 

(or a justified coverage) of all concrete structures within the scope 

of renewal (e.g., all normally accessible exterior surfaces of all 

loaded overpacks), and 100 percent of normally inaccessible 

surfaces (or a justified coverage) for a justified subset of the 

reinforced concrete structures within the scope of renewal 

(e.g., interior surfaces of two overpacks, including the overpack 

earliest loaded and the overpack loaded with the highest heat-load 

canister). The extent of inspection coverage should be specified 

and demonstrated to sufficiently characterize the condition of the 

structure.

For the groundwater chemistry program and radiation surveys, the 

sample size identifies and justifies specific locations where 

inspection or monitoring will be conducted to sufficiently 

characterize the condition of the structure (e.g., periodic dose rate 

measurements will be performed at the same locations specified in 

the TS for dose rate measurements at loading).

Data collection:

Data collection for visual inspections is commensurate with 

consensus standards and guides (see ACI 224.1R (ACI, 2007b) for 

quantitative analysis (crack width, extent), ACI 562, (ACI, 2013a), 

ACI 364.1R (ACI, 2007c)).
6-22

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

Data from all inspection and monitoring activities, including 

evidence of degradation and its extent and location, shall be 

documented on a checklist or inspection form. The results of the 

inspection shall be documented, including descriptions of observed 

aging effects and supporting sketches, photographs, or video. 

Corrective actions from AMP activities shall also be documented.

An adequate clearinghouse is used for documenting inspection and 

monitoring operating experience.

Timing

Initial inspections and monitoring activities are completed before

entering the period of extended operation; the activities may be part 

of a preapplication inspection or a general-licensee baseline 

inspection (see NUREG–1927, Rev. 1 (NRC, 2016)).

5. Monitoring and

Trending

Monitoring and trending methods are commensurate with 

concensus defect evaluation guides and standards (see

ACI 201.1R (ACI, 2008a), ACI 207.3R (ACI, 2008b), ACI 364.1R 

(ACI, 2007c), ACI 562 (ACI, 2013a), or ACI 224.1R (ACI, 2007b) 

for crack evaluation).

Inspection and monitoring results are compared to those obtained 

during previous inspections, so that the progression of degradation 

can be evaluated and predicted.

Monitoring and trending methods reference plans and procedures 

used to:

• establish a baseline before or at the beginning of the period of

extended operation

• track trending of parameters or effects not corrected in a

previous inspection, for example

— crack growth/extent

— pore/void density and affected areas

— dose rates

6. Acceptance Criteria The acceptance criteria for visual inspections are commensurate

with the 3-tier quantitative criteria in ACI 349.3R-02:

• Tier 1: acceptance without further evaluation

• Tier 2: acceptance after review

• Tier 3: acceptance requiring further evaluation

All conditions not meeting the Tier 2 acceptance criteria are 

evaluated in the Corrective Action Program (CAP) to reasonably 
6-23

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

ensure that the intended functions of the structure will be 

adequately maintained until a followup inspection, at a minimum.

The acceptance criteria for the groundwater chemistry program are 

commensurate with ASME Code Section XI, Subsection IWL, which 

states that an aggressive below-grade environment is defined as 

pH < 5.5, chlorides > 500 ppm, or sulfates > 1500 ppm.

The acceptance criteria for radiation surveys are justified and 

sufficient to ensure compliance with 10 CFR 72.104 and identify 

dose rates that statistically exceed calculated or expected dose 

rates at predetermined measurement locations. The adequacy of 

the acceptance criteria considers measured dose rates versus 

calculated or expected dose rates for a DSS, given the DSS 

contents and accounting for the decay of the source term since the 

DSS loading. Measurement locations should be consistent with 

those specified in the license or Certificate of Compliance (CoC) 

conditions or TS (if any) and locations where dose rates were 

calculated in the final safety analysis report (FSAR) and likely 

measured at the time of loading.

Alternative acceptance criteria should be reviewed on a 

case-by-case basis. For such cases, the acceptance criteria shall:

• include a quantitative basis (justifiable by operating 

experience, engineering analysis, consensus codes and 

standards)

• avoid use of nonquantifiable phrases (e.g., significant, 

moderate, minor, little, slight, few)

• be achievable and clearly actionable

7. Corrective Actions Results that do not meet the acceptance criteria are addressed as 

conditions adverse to quality or significant conditions adverse to 

quality under those specific portions of the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that corrective actions are completed within the 

specific- or general-licensee’s Corrective Action Program (CAP), 

and include provisions to

• perform functionality assessments

• perform apparent cause evaluations, and root cause

evaluations

• address the extent of condition

• determine actions to prevent recurrence for significant 

conditions adverse to quality; ensure justifications for 

nonrepairs
6-24

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

• trend conditions

• identify operating experience actions, including modifications

to the existing AMP (e.g., increased frequency)

• determine if the condition is reportable to the NRC per

10 CFR 72.75

Corrective actions shall be consistent with applicable consensus 

rehabilitation guides or standards, unless an engineering 

justification is provided (e.g., for cracking: ACI 224.1R, ACI 562, 

ACI 364.1R, and ACI RAP Bulletins; for spalling/scaling: ACI 562, 

ACI 364.1R, ACI 506R, and ACI RAP Bulletins).

8. Confirmation

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the confirmation process includes provisions 

to preclude repetition of significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine followup actions to verify effective implementation

of corrective actions

• monitor for adverse trends due to recurring or repetitive

findings or observations.

9. Administrative

Controls

The administrative controls are in accordance with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the administrative controls include provisions 

that define:

• instrument calibration and maintenance

• inspector requirements (commensurate with ACI 349.3R-02)

• record retention requirements

• document control

The administrative controls describe or reference: 

• methods for reporting results to the NRC per 10 CFR 72.75

• frequency for updating the AMP based on industrywide

operational experience

10. Operating

Experience

Structures monitoring programs using the acceptance criteria in 

ACI 349.3R-02 (ACI, 2010) have proven effective for aging 

management of concrete structures in nuclear power plants during 

their period of extended operation (NRC, 2010b). NUREG–1522 

documents the results of a survey sponsored in 1992 by the Office 

of Nuclear Reactor Regulation to obtain information on the types of 

distress in the concrete and steel structures and components, the 
6-25

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

type of repairs performed, and the durability of the repairs. 

Licensees who responded to the survey reported cracking, scaling, 

and leaching of concrete structures. The degradation was 

attributed to drying shrinkage, Freeze and thaw, and abrasion. The 

NUREG also describes the results of NRC staff inspections at six 

plants. The staff observed concrete degradation, corrosion of 

component support members and anchor bolts, cracks and other 

deterioration of masonry walls, and groundwater leakage and 

seepage into underground structures. The observed and reported 

degradations were more severe at coastal plants than those 

observed in inland plants, as a result of brackish and sea water. 

Previous reactor license renewal applicants reported similar 

degradation and corrective actions taken through their structures 

monitoring program.

NRC Information Notice 2011-20 (NRC, 2011) documents the 

occurrence of alkali-silica reaction (ASR)-induced concrete 

degradation of a seismic Category 1 below-grade structure at the 

Seabrook Station power plant. The concrete used in the structure 

passed all industry standard ASR screening tests (ASTM, 2007, 

2012) at the time of construction; however, ASR-induced 

degradation was identified in August 2010. The licensee completed 

a prompt operability determination that concluded margins to the 

design limits remained such that the structural integrity of the 

building continued to be demonstrated.

NRC Information Notice 2013-07 documents the occurrence of 

Freeze and thaw cracking near the anchor blockout holes on the 

roof of horizontal storage modules (HSMs) at an ISFSI in Idaho. 

The cracking led to water migration into the concrete, resulting in 

efflorescence of calcium carbonate deposits. The degradation of 

the roofslabs was not related to age-related degradation but to a 

design feature leading to water accumulation. More extensive 

visual inspections of the HSMs also revealed map cracking on the 

vertical wall surfaces, random and radial cracking at the door edges 

in base units, and spalling at the bottom edge of shield walls. The 

licensee conducted nondestructive and destructive examination, 

which revealed adequate concrete quality and compressive 

strength.

Additional visual inspections of concrete structures in DSSs have 

been conducted at the Calvert Cliffs ISFSI (Gellrich, 2012) and the 

Palisades ISFSI. Remote visual inspections of two HSMs at the 

Calvert Cliffs ISFSI revealed efflorescence of the concrete and the 

formation of calcium carbonate stalactites in the 2-inch gap 

between the heat shield and the concrete ceiling. These stalactites 

were attributed to water ingress through the outlet vent stack. A 

condition report was issued that did not identify an operability issue. 
6-26

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

Inspections of the exterior surfaces of a ventilated concrete cask 

(VCC) and the concrete support pad at the Palisades ISFSI 

revealed bugholes exceeding preestablished acceptance criteria 

and requiring grout repair, and a void at the interface between the 

VCC bottom plate and the vertical VCC concrete wall. No 

conditions were identified to compromise the intended functions of 

the VCC.

Walkdowns and visual inspections of readily accesible surfaces of 

concrete overpacks and HSMs are generally conducted during the 

initial storage period, although the acceptance criteria may vary 

from those in ACI 349.3R.02 (ACI, 2010). The NRC reviewer 

should evaluate relevant inspection results included in the renewal 

application, based on design and environmental similarities, and 

evaluate if activities in this generic AMP should be augmented as a 

result of those inspections.

References ACI. ACI 506R-05, “Guide to Shortcrete.” American Concrete 

Institute. 2005.

_____. ACI 349-06, “Code Requirements for Nuclear SafetyRelated Concrete Structures.” American Concrete Institute. 2007a.

_____. ACI 224.1R-07, “Causes, Evaluation, and Repair of Cracks 

in Concrete Structures.” American Concrete Institute. 2007b.

_____. ACI 364.1R-07, “Guide for Evaluation of Concrete 

Structures before Rehabilitation.” American Concrete Institute. 

2007c.

_____. ACI 201.1R-08, “Guide for Conducting a Visual Inspection 

of Concrete in Service.” American Concrete Institute. 2008a.

_____. ACI 207.3R-94, “Practices for Evaluation of Concrete in 

Existing Massive Structures for Service Conditions.” 

American Concrete Institute. 2008b.

_____. ACI 349.3R-02, “Evaluation of Existing Nuclear SafetyRelated Concrete Structures.” American Concrete Institute. 2010.

_____. ACI 318-11, “Building Code Requirements for Structural 

Concrete.” American Concrete Institute. 2011.

_____. ACI 562-13, “Code Requirements for Evaluation, Repair, 

and Rehabilitation of Concrete Buildings.” American Concrete 

Institute. 2013a.

_____. ACI CT-13, “ACI Concrete Terminology.” 

American Concrete Institute. 2013b.
6-27

Table 6-3 Example aging management program for Reinforced Concrete Structures

Element Description

ASME Boiler and Pressure Vessel Code, Section XI, Subsection 

IWL (2013), “Requirements for Class CC Concrete Components of 

Light-Water-Cooled Plants”

ASTM International. ASTM C289, “Standard Test Method for 

Potential Alkali-Silica Reactivity of Aggregates (Chemical Method).” 

West Conshohocken, Pennsylvania: American Society for Testing 

and Materials. 2007.

_____. ASTM C295, “Standard Guide for Petrographic 

Examination of Aggregates for Concrete.” West Conshohocken, 

Pennsylvania: American Society for Testing and Materials. 2012.

Gellrich, G. “Calvert Cliffs Nuclear Power Plant.” Letter to 

U.S. Nuclear Regulatory Commission, Response to Request for 

Supplemental Information. RE: Calvert Cliffs Independent Spent 

Fuel Storage Installation License Renewal Application (TAC No. 

L24475). ADAMS Accession No. ML12212A216. 2012.

NRC. “Standard Review Plan for Spent Fuel Dry Storage 

Facilities.” NUREG–1567, Rev. 0. Washington, DC. ADAMS 

Accession No. ML003686776. 2000. 

_____. “Standard Review Plan for Spent Fuel Dry Storage 

Systems at a General License Facility.” NUREG–1536, Rev. 1. 

ADAMS Accession No. ML091060180.Washington, DC. 2010a.

_____. NUREG–1801, “Generic Aging Lessons Learned (GALL) 

Report.” Rev. 2, Washington DC. ADAMS Accession No. 

ML103490041. 2010b.

_____. “Information Notice 2011-20, Concrete Degradation by 

Alkali-Silica Reaction.” Washington, DC: U.S. Nuclear Regulatory 

Commission. ADAMS Accession No. ML112241029. 2011. 

_____. NUREG–1927, “Standard Review Plan for Renewal of 

Specific Licenses and Certificates of Compliance for Dry Storage of 

Spent Nuclear Fuel.” Revision 1. Washington, DC: U.S. Nuclear 

Regulatory Commission. ADAMS Accession No. ML16179A148.

2016.

SEI/ASCE 11-99, “Guideline for Structural Condition Assessment of 

Existing Buildings.” 2000.

1

6-29

1 6.7 External Surfaces Monitoring Of Metallic Components

2 An example AMP for external surfaces monitoring of metallic components is provided below. 

3 The AMP manages all metallic surfaces that are directly exposed to outdoor air, concrete, or are 

4 sheltered within DSS overpacks, except for stainless steel storage canisters and transfer casks, 

5 which are addressed by other AMPs. The AMP is a condition monitoring program that consists 

6 of periodic visual inspections to monitor for corrosion, wear, cracking, and loss of preload 

7 (bolting). 

8
6-30

Table 6-4 Example aging management program for External Surfaces Monitoring Of

 Metallic Components 

Element Description

1. Scope of

Program

This program manages the effects of aging for the external surfaces of 

steel and stainless steel components that are directly exposed to 

outdoor air or are sheltered within DSS overpacks (e.g., NUHOMS 

HSM, HI-STORM). The scope of the program includes metallic 

overpack exterior surfaces, dry storage canister support structures, 

access doors, vents, heat shields, embedments and anchorages, 

bolting, and other components important to safety. 

The scope of this program does not include stainless steel dry storage 

canisters housed within overpacks, transfer casks, or the top closure 

(confinement) boundary of bolted casks. The Localized Corrosion and 

Stress corrosion Cracking of Welded Stainless Steel Dry Storage 

Canisters AMP manages the effects of aging for stainless steel 

canisters. The Transfer Casks AMP manages the effects of aging of 

all transfer cask components. The Bolted Cask Seal Leakage 

Monitoring AMP manages the effects of aging on the integrity of the 

top confinement boundary of bolted spent fuel storage casks. 

Periodic visual inspections monitor for general and localized corrosion, 

wear, cracking, and loss of preload (bolting). 

2. Preventive

Actions

This program is a condition monitoring program to detect evidence of 

degradation. It does not provide guidance for the prevention of aging.

3. Parameters

Monitored/

Inspected

This program monitors the condition of external metallic surfaces to 

identify general corrosion, localized corrosion, wear, and loss of 

preload of bolted connections. Localized corrosion of stainless steels 

may be a precursor to SCC. 

Parameters monitored or inspected for external metallic surfaces 

include:

• visual evidence of discontinuities, imperfections, and rust

staining indicative of corrosion, SCC, and wear

• visual evidence of loose or missing bolts, physical displacement,

and other conditions indicative of loss of preload

• visual evidence of coating degradation (e.g., blisters, cracking,

flaking, delamination) indicative of corrosion of the base metal

4. Detection of

Aging Effects

Readily Accessible Surfaces

Visual inspections are performed in accordance with ASME Code

Section XI, Article IWA-2213, for VT-3 examinations. The inspections 

cover 100 percent of normally accessible surfaces, including the

external surfaces of metallic overpacks, bolting, lightning protection

system components, access doors, vents, and other metallic 

components. 
6-31

Table 6-4 Example aging management program for External Surfaces Monitoring Of

 Metallic Components 

Element Description

Normally Inaccessible Surfaces

Opportunistic visual inspections are performed with remote inspection 

techniques on metallic surfaces within overpacks that are accessed 

during inspections of dry storage canisters, including heat shields, 

canister support structures, and other metallic components. 

The condition of metallic surfaces in contact with concrete 

(i.e., overpack/cask bottoms) are assessed with visual inspections on 

a justified frequency.

Procedures for visual inspections should be demonstrated; procedure 

attributes should include, for example, equipment resolution and

lighting requirements and should reference applicable standards 

(e.g., ASTM Code Section XI, Article IWA-2200, for VT-3 

examinations). The extent of inspection coverage should be specified 

and demonstrated to sufficiently characterize the condition of the 

metallic components. 

Sample Size

The readily accessible exterior metallic surfaces of all casks and 

overpacks are inspected. The inspections of normally inaccessible 

surfaces within overpacks is opportunistic; inspections are performed 

whenever the overpacks are accessed for dry storage canister 

inspections. Overpack and cask bottoms are inspected on a justified 

sample size.

Frequency

Inspections of readily accessible surfaces are conducted at least once 

every 5 years. Normally inaccessible surfaces within overpacks are 

inspected when those surfaces are accessed during remote 

inspections of dry storage canisters. Overpack and cask bottoms are 

inspected on a justified frequency.

Data Collection

Data from the examination, including evidence of degradation and its 

extent and location, shall be documented on a checklist or inspection 

form. The results of the inspection shall be documented, including 

descriptions of observed aging effects and supporting sketches, 

photographs, or video. Corrective actions resulting from each AMP 

inspection shall also be documented.

Timing

Initial inspections are completed before entering the period of 

extended operation.
6-32

Table 6-4 Example aging management program for External Surfaces Monitoring Of

 Metallic Components 

Element Description

5. Monitoring and

Trending

Inspection results are compared to those obtained during previous 

inspections, so that the progression of degradation can be evaluated 

and predicted. 

Monitoring and trending methods reference plans and procedures 

used to:

• establish a baseline before or at the beginning of the period of

extended operation

• track trending of parameters or effects not corrected following a

previous inspection, including

— locations and size of any areas of corrosion, wear, or 

cracking

— disposition of components with identified aging effects 

and the results of supplemental inspections

6. Acceptance

Criteria

The acceptance criteria for the visual inspections are:

• no detectable loss of material from the base metal, including

uniform wall thinning, localized corrosion pits, and crevice

corrosion

• no red-orange-colored corrosion products on the base metal,

coatings, or concrete

• no coating defects (e.g., blisters, cracking, flaking, delamination)

• no indications of loose bolts or hardware, displaced parts

If evidence of corrosion, wear, or coating degradation is identified, then 

the severity of the degradation must be determined using approved 

site-specific procedures. These may include additional visual, surface 

or volumetric nondestructive examination (NDE) methods to determine 

the loss of material and, for welded stainless steels, the presence of 

cracking. 

Alternative acceptance criteria are developed from system-specific 

design standards, industry codes or standards, or engineering 

evaluation. Where possible, acceptance criteria are quantitative 

(e.g., minimum wall thickness). Where qualitative acceptance criteria 

are used, the criteria are sufficiently clear to reasonably ensure that a 

singular decision is derived based on the observed condition, avoiding 

the use of ambiguous phrases (e.g., significant, moderate). 

EPRI technical reports, Technical Report (TR)-1007933, “Aging 

Assessment Field Guide” (EPRI, 2003), and TR-1009743, “Aging 

Identification and Assessment Checklist: Mechanical Components” 

(EPRI, 2004), provide general guidance for the evaluation of materials 

and the development of criteria for their acceptance when performing 

visual inspections.
6-33

Table 6-4 Example aging management program for External Surfaces Monitoring Of

 Metallic Components 

Element Description

7. Corrective

Actions

Results that do not meet the acceptance criteria are addressed as 

conditions adverse to quality or significant conditions adverse to 

quality under those specific portions of the specific- or generallicensee QA program approved under 10 CFR Part 72, Subpart G, or 

10 CFR Part 50, Appendix B, respectively. The QA program ensures 

that corrective actions are completed within the specific- or generallicensee’s Corrective Action Program (CAP), and include provisions to

• perform functionality assessments

• perform apparent cause evaluations and root cause evaluations

• address the extent of condition

• determine actions to prevent recurrence for significant conditions

adverse to quality; ensure justifications for nonrepairs

• trend conditions

• identify operating experience actions, including modification to

the existing AMP (e.g., increased frequency)

• determine if the condition is reportable to the NRC per

10 CFR 72.75

8. Confirmation

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the confirmation process includes provisions to 

preclude repetition of significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine followup actions to verify effective implementation of

corrective actions

• monitor for adverse trends due to recurring or repetitive findings

or observations.

9. Administrative

Controls

The administrative controls are addressed through those portions of 

the specific- or general-licensee QA program that are used to meet 

10 CFR Part 72, Subpart G, or 10 CFR Part 50, Appendix B.

10. Operating

Experience

External surface inspections through system inspections and 

walkdowns in support of the Maintenance Rule (10 CFR Part 50.65) 

have proven effective in maintaining the material condition of nuclear 

power plant systems. 

NRC Information Notice 2012-20 (NRC, 2012) documents cases of 

atmospheric CISCC of welded stainless steel piping systems and 

tanks at operating reactor locations. Atmospheric CISCC growth rates 

determined from operational experience at both domestic and foreign 

nuclear power plants, include events at San Onofre, Turkey Point, 
6-34

Table 6-4 Example aging management program for External Surfaces Monitoring Of

 Metallic Components 

Element Description

St. Lucie, and Koeberg (South Africa), range from 3.6 × 10−12 to 

2.9 × 10−11 m/sec for components at ambient temperatures. 

References EPRI. EPRI Technical Report 1007933, “Aging Assessment Field 

Guide.” Palo Alto, California: Electric Power Research Institute. 

December 2003.

_____. EPRI Technical Report 1009743, “Aging Identification and 

Assessment Checklist–Mechanical Components.” Palo Alto, 

California: Electric Power Research Institute. August 27, 2004.

NRC. NRC Information Notice 2012-20, “Potential Chloride-Induced 

Stress Corrosion Cracking of Austenitic Stainless Steel and 

Maintenance of Dry Cask Storage System Containers.” 

Washington, DC: U.S. Nuclear Regulatory Commission. 

November 14, 2012.

1 

2 
6-35

1 6.8 Ventilation Systems

2 An example AMP for ventilation systems is provided below. The AMP manages all inlet and 

3 outlet vents and conduits providing convective cooling in DSSs. This is a condition monitoring 

4 program that performs periodic visual inspection of vents as defined in the approved design 

5 bases, with additional focused inspections to address normally unobservable vent areas, as well 

6 as evidence of degradation that could result in obstructions. Temperature monitoring may be 

7 used in lieu of the periodic visual surveillances to verify cooling performance. 

8
6-36

1 

Table 6-5 Example aging management program for Ventilation Systems

Element Description

1. Scope of

Program

This program manages potential loss of cooling capabilities due to 

blockage of the ventilation system (air inlet/outlets, convection 

conduits) in DSSs. Surveillance/monitoring and focused inspections 

of the ventilation system (i) ensure that blockage does not result in 

design temperature limits being exceeded and (ii) prevent 

unanticipated adverse degradation of components of the DSS 

(e.g., high-temperature dehydration of the concrete1

, hydride 

reorientation due to fuel cladding temperatures exceeding 

design-bases limits2

).

Visual Surveillances of Inlet and Outlet Vents

The scope of the program includes continuance of surveillances 

(periodic walkdowns) of air inlet/outlet vents, as defined in the 

approved design bases (FSAR, license/CoC TS). The program 

provides for additional focused inspections if (i) the normally 

unobservable vent area exceeds the allowable blockage, and 

(ii) there is evidence of degradation of other components (e.g., loss 

of coatings, spalling or leaching of the concrete overpack) that could 

result in obstructions.3

Temperature Monitoring

The scope of the program includes temperature monitoring of DSS 

components in lieu of visual surveillances, as specified in the 

approved design bases (FSAR, license/CoC TS). Continuance of 

temperature monitoring provides a means to detect anomalous 

temperature changes in the DSS. The program further provides for 

focused visual inspections of the ventilation system (inlet/outlet 

vents, conduits) in the event that anomalous temperature changes 

are measured. Focused visual inspections allow for detection of 

degradation of other components that could result in obstructions 

(e.g., loss of coatings, inner spalling or leaching of the 

concrete overpack).

The scope of the program does not include inspection and/or 

maintenance activities for aging of bird screens used to prevent vent 

blockage (see the External Surfaces Monitoring of Metallic 

Components AMP).

1See NUREG–1536/NUREG–1567 (NRC, 2010, 2002) for design criteria on maximum concrete temperatures.

2

See ISG-11, Revision 3 (NRC, 2003), for cladding considerations for the transportation and storage of spent fuel.

3The approved design bases have adequately addressed the occurrence of extreme natural phenomena, such as 

heavy snowstorm or flooding. The QA program ensures that corrective actions are completed within the specific- or 

general-licensee’s CAP in the event of extreme natural phenomena.
6-37

Table 6-5 Example aging management program for Ventilation Systems

Element Description

2. Preventive 

Actions

This program is a condition monitoring program to detect obstruction 

or blockages of the ventilation system that could result in 

design-bases temperature limits being exceeded. It does not provide 

guidance for the prevention of aging of components.

3. Parameters 

Monitored/ 

Inspected

Visual Surveillances of Inlet and Outlet Vents

Parameters monitored or inspected include blockage or obstruction 

in the air inlet and outlet vents.

Temperature Monitoring

Parameters monitored or inspected include temperature 

measurements of the DSS, which could be based on (i) direct 

measurements of the overpack temperatures, (ii) direct measurement 

of the canister temperatures, (iii) a comparison of the inlet and outlet 

temperature difference to predicted temperature differences for each 

individual overpack, or (iv) other means that would identify and allow 

for the correction of off-normal thermal conditions that could lead to 

exceeding design-bases temperature limits for the concrete and/or 

fuel cladding.

Focused Inspections

Parameters monitored or inspected include (i) blockage or 

obstructions of the air inlets/outlets and (ii) degradation of other 

components (e.g., loss of coatings, inner spalling or leaching of the 

concrete overpack) that could result in obstructions of inaccessible 

convective conduits.

4. Detection of 

Aging Effects

Method/Technique

Visual surveillances of the air inlet/outlet vents are performed during 

periodic walkdowns, without the need of remote equipment. 

Surveilling personnel should have an unobstructed view of vent 

areas that allows confirmation that the maximum allowable blockage 

is not exceeded (up to the boundary of the vent bird screen, at a 

minimum). The maximum allowable blockage is defined in the 

approved design bases (FSAR, license/CoC TS).

Temperature monitoring is performed with qualified and calibrated 

measurement devices or sensors that are maintained in accordance 

with the site QA program.

Focused inspections are performed with remote inspection 

techniques. Procedures for remote visual inspections should be 

demonstrated; procedure attributes should include, for example,

equipment resolution and lighting requirements, in consideration of 

the ventilation system design.
6-38

Table 6-5 Example aging management program for Ventilation Systems

Element Description

Frequency of Inspection/Monitoring

Visual surveillances and temperature monitoring are conducted at a 

frequency consistent with the approved design bases (i.e., as defined 

in the FSAR, or the relevant license/CoC TS). Generally, visual 

surveillances are conducted daily (not exceeding a 48-hour interval) 

and temperature monitoring is performed continuously.

The frequency of focused inspections for vent areas should be 

justified based on the design (percentage of normally unobservable 

vent area relative to allowable blockage) and operable degradation 

modes of the storage system components that could lead to 

blockage. The frequency of focused inspections provides reasonable 

assurance that blockages in the normally inaccessible convective 

conduits will be identified before a loss of function by considering 

conduit-free volume relative to postulated obstructions (e.g., upon 

consideration of potential coating loss or concrete spalling relative to 

conduit-free volume). Previous operating experience may be used to 

justify the use of opportunistic inspections. When continuous 

temperature monitoring is used to verify ventilation performance, 

focused inspections are performed whenever anomalous 

temperatures are measured.

Sample Size

Visual surveillances include all directly observable areas of the inlet 

and outlet vents. Visual surveillances are performed on all loaded 

systems, or as justified by the approved design bases (i.e., as 

defined in the FSAR, or the relevant license/CoC TS).

Temperature monitoring is performed in all loaded systems, or as 

justified by the approved design bases (i.e., as defined in the FSAR, 

or the relevant license/CoC TS).

For focused inspections, the extent of inspection coverage should be 

specified and demonstrated to sufficiently characterize the condition 

of the ventilation system. Focused inspections include all normally 

unobservable vent areas exceeding the allowable blockage. The 

extent of inaccessible conduit inspection is justified based on the 

ventilation system design (conduit-free volume, accessibility) and 

consideration of operable degradation modes of the storage system 

materials. The use of continuous temperature monitoring may be 

used in lieu of focused inspections if anomalous temperatures are 

not measured.

Data Collection

Data collection should be consistent with site procedures in 

compliance with the specific- or general-licensee’s QA program.
6-39

Table 6-5 Example aging management program for Ventilation Systems

Element Description

Timing

A baseline focused inspection is conducted on a sample DSS upon 

entering the period of extended operation to identify any operable 

degradation modes that may result in an obstruction or blockage 

difficult to observe during a visual surveillance. The baselinefocused inspection includes 100 percent of the vents and 

inaccessible convective conduits of the sample DSS, or a justified 

volume based on design considerations (e.g., accessibility, dose 

rates). A baseline-focused inspection on a sample system is not 

necessary if temperature monitoring is used in lieu of visual 

surveillances.

5. Monitoring and 

Trending

Results from visual surveillances and temperature monitoring are 

trended to identify conditions (materials/environmental) leading to 

obstructions or blockages.

Results from focused inspections are compared with prior 

inspections to monitor and trend operable degradation modes of the 

storage system materials that have resulted in partial blockage.

6. Acceptance 

Criteria

The acceptance criteria are defined to ensure that the need for 

corrective actions will be identified before (i) blockage results in 

design temperature limits being exceeded and (ii) unanticipated 

adverse degradation of components of the DSS results in a loss of 

intended function. Where possible, acceptance criteria are 

quantitative (e.g., 50-percent areal blockage or a specific allowed 

temperature range). Where qualitative acceptance criteria are used, 

the criteria are sufficiently clear to reasonably ensure that a singular 

decision is derived based on the observed condition, avoiding the 

use of ambiguous phrases (e.g., significant, moderate).

The acceptance criteria for visual surveillances and focused 

inspections are justified based on the ventilation system design, 

thermal performance criteria, and consideration of operable 

degradation modes of the storage system materials. The acceptance 

criteria may be further justified by parallel maintenance activities 

under a separate AMP.

The acceptance criteria for temperature monitoring are justified and 

conservative to the short-term temperature limits for a blocked vent 

condition, as defined in the approved design bases (i.e., as defined in 

the FSAR, or the relevant license/CoC TS).

7. Corrective 

Actions

Results that do not meet the acceptance criteria are addressed as 

conditions adverse to quality or significant conditions adverse to 

quality under those specific portions of the specific- or generallicensee QA program approved under 10 CFR Part 72, Subpart G, or 
6-40

Table 6-5 Example aging management program for Ventilation Systems

Element Description

10 CFR Part 50, Appendix B, respectively. The QA program ensures 

that corrective actions are completed within the specific- or generallicensee’s Corrective Action Program (CAP), and include provisions 

to:

• perform functionality assessments

• perform apparent cause evaluations and root cause evaluations

• address the extent of condition

• determine actions to prevent recurrence for significant

conditions adverse to quality; ensure justifications for

nonrepairs

• trend conditions

• identify operating experience actions, including modification to

the existing AMP (e.g., increased frequency)

• determine if the condition is reportable to the NRC per

10 CFR 72.75

8. Confirmation

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the confirmation process includes provisions to 

preclude repetition of significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine followup actions to verify effective implementation of

corrective actions

• monitor for adverse trends due to recurring or repetitive findings

or observations.

9. Administrative

Controls

The administrative controls are addressed through those portions of 

the specific- or general-licensee QA program that are used to meet 

10 CFR Part 72, Subpart G, or 10 CFR Part 50, Appendix B.

To ensure the temperature monitoring devices will remain accurate 

during the period of extended operation, the electronic circuitry 

associated with the temperature monitoring devices should be 

periodically calibrated in accordance with the licensee’s QA 

requirement in 10 CFR 72.164 and specific-license requirement in 

10 CFR 72.44(c)(3)(ii). In addition, the calibration data are 

periodically evaluated to identify anomalous trends that could 

indicate degraded instrumentation or degradation in the ventilation 

system. All external components in the temperature measurement 

devices should be periodically inspected and calibrated to ensure 

that no degradation due to corrosion, wear, or cracking has occurred.
6-41

Table 6-5 Example aging management program for Ventilation Systems

Element Description

10. Operating 

Experience

Visual surveillance of the exterior of the air inlets and outlets, 

inspections for ventilation blockage and temperature monitoring have 

been in effect at ISFSIs and have been proven effective in 

maintaining the convective cooling capabilities of DSSs during the 

initial license or certification period.

Degradation of inner overpack materials has been observed in the 

field. NRC Information Notice 2013-07 (NRC, 2013) documents 

experience at the Three Mile Island, Unit 2, ISFSI, where water 

entered anchor bolt blockout holes on the roof of HSM concrete 

overpacks. Subsequent freeze and thaw cycles resulted in crack 

formation, crack growth, and efflorescence of the concrete. 

Inspections of two HSMs at the Calvert Cliffs ISFSI have also shown 

efflorescence of the concrete and the formation of calcium carbonate 

stalactites in the 2-inch gap between the heat shield and the concrete 

ceiling. These stalactites were observed near the outlet vent stack. 

A condition report was issued that did not identify an operability issue

(CENG, 2012).

Partial blockage of air inlet duct screens from snowfall has been 

identified. Decay heat from the spent fuel and/or stored heat in the 

overpack material (e.g., concrete) quickly melts any partial snow 

buildups after the snowfall has ceased. The existing activities 

(surveillance, monitoring, inspection) have proved adequate for 

natural phenomena during the initial license or certification period.

References CENG. Letter to NRC, “Calvert Cliffs Nuclear Power Plant 

Independent Spent Fuel Storage Installation Material License No. 

SNM-2505, Docket No. 72-8, Response to Request for Supplemental 

Information, RE: Calvert Cliffs.” Independent Spent Fuel Storage 

Installation License Renewal Application, Calvert Cliffs Nuclear 

Power Plant, LLC. ADAMS Accession No. ML12212A216. July 27, 

2012.

NRC. NUREG–1567, “Standard Review Plan for Spent Fuel Dry 

Storage Facilities.” Washington DC: U.S. Nuclear Regulatory 

Commission. March 2000.

_____. Interim Staff Guidance (ISG)–11, “Cladding Considerations 

for the Transportation and Storage of Spent Fuel.” Rev. 3. 

Washington DC: U.S. Nuclear Regulatory Commission. 

November 2003.

_____. NUREG–1536, “Standard Review Plan for Spent Fuel Dry 

Storage Systems at a General License Facility.” Washington DC: 

U.S. Nuclear Regulatory Commission. July 2010.

_____. Information Notice 2013-11, “Premature Degradation of 

Spent Fuel Storage Cask Structures and Components from 
6-42

Table 6-5 Example aging management program for Ventilation Systems

Element Description

Environmental Moisture.” Washington, DC: U.S. Nuclear Regulatory 

Commission. April 16, 2013.

1 
6-43

1 6.9 Bolted Cask Seal Leakage Monitoring

2 This AMP manages the aging of all bolted casks that employ leakage monitoring to verify the 

3 integrity of the top confinement boundary. The program relies on existing pressure monitoring 

4 systems to assess the integrity of cask closure seals. The program also performs periodic 

5 visual inspections of normally inaccessible components under the cask protective cover to 

6 monitor for corrosion, coating degradation, loose bolts, and evidence of water intrusion. 

7
6-44

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

1. Scope of

Program

This program is used to manage the aging effects on the integrity of 

the top confinement boundary of bolted spent fuel storage casks to 

ensure that timely and appropriate corrective actions can be taken to 

maintain the safe storage conditions of the casks. The aging effects 

include loss of material as a result of corrosion of the sealing 

surfaces, O-rings, and bolts; loss of strength due to thermal aging 

and change in dimension due to creep of the metallic O-rings that 

results in loss of sealing forces; and loss of preload of the closure 

bolts. 

The specific components and systems that are typically managed by 

this program include the shield lid, primary lid, closure lid, protective 

covers, O-ring assemblies, and associated bolts and welds. The 

types of bolted cask designs covered by the program include TN‐24, 

‐32, ‐40, and ‐68; NAC‐S/T (I26), -C28 S/T, -I28 S/T, and ‐STC, 

CASTOR V/21 and X/33; and Westinghouse MC-10 bolted casks.

The program relies on continuous pressure-leakage monitoring to 

verify the integrity of the confinement boundary. In addition, the 

program relies on periodic visual inspections for evidence of aging 

that may affect the intended function of the identified SSCs and 

subcomponents.

2. Preventive

Actions

Preventive actions include compliance with the NRC’s ISG on the 

materials selection for fabrication, design, and testing of casks, as 

described in NRC ISG-5, “Confinement Evaluation” (NRC, 1999); 

ISG-15, “Materials Evaluation” (NRC, 2001); and ISG-25, “Pressure 

and Helium Leakage Testing of the Confinement Boundary of Spent 

Fuel Dry Storage Systems” (NRC, 2010).

3. Parameters

Monitored/

Inspected

The program relies on existing pressure-monitoring systems to 

assess the integrity of the cask closure seals. To verify the integrity 

of the seal assemblies in the bolted casks, these systems 

continuously monitor pressure: 

• between the metallic seal assemblies in the TN‐24, ‐32, ‐40,

and ‐68; NAC‐S/T (I26), -C28 S/T, -I28 S/T, and ‐STC,

CASTOR V/21 and X/33 casks, and

• inside the cask cavity in the MC-10 casks.

Parameters monitored/inspected for closure seal components 

include:

• visual evidence of loss of material from general, localized, and

galvanic corrosion

• visual evidence of coating degradation that could indicate

corrosion of the base metal
6-45

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

• visual evidence of clearances and physical displacements in

bolted joints indicative of loss of preload or failed or missing

components

• visual evidence of water intrusion under the protective cover

4. Detection of

Aging Effects

Aging effects may be revealed by:

• overpressure and pressure loss (leakage)

• water intrusion under protective covers

• physical displacement, surface discontinuities, and

imperfections indicative of loss of preload and corrosion.

Method or Technique

The program credits the pressure-monitoring system, which 

continuously monitors the pressure between the seal assemblies in 

the TN‐24, ‐32, ‐40, and ‐68; NAC‐S/T (I26), -C28 S/T, -I28 S/T, and 

‐STC, CASTOR V/21 and X/33 metal casks and inside the cask 

cavity of the MC-10 casks. Continuous monitoring with a pressure 

alarm provides a means for early detection of aging effects on the 

seal assemblies.

Direct or remote VT-3 visual examination, as described in ASME

Code Section XI, Article IWA-2213 (ASME, 2007), shall be performed 

and evaluated by personnel qualified in accordance with the 

requirements of IWE-2330. 

Frequency

Pressure-monitoring systems provide continuous monitoring of the 

bolted cask seal integrity. Checks of system operation shall be 

conducted, in accordance with the established requirements for these 

systems. Inspection and calibration of the components of the 

overpressure leakage-monitoring systems shall be performed in 

accordance with manufacturer specifications. Opportunistic 

inspections of the overpressure leakage monitoring systems shall be 

conducted when the protective cover plate is removed for other 

inspection or maintenance actions. 

Visual VT-3 inspection of the normally inaccessible top sealing 

components in the confinement boundary, after removing the 

protective cover, shall be conducted on a justified frequency. This 

includes the condition of externally accessible surfaces of the bolts, 

protective covers, and protective coatings. In addition, opportunistic 

inspections of the top confinement boundary subcomponents shall be 

conducted when the protective cover is removed for other inspection 

or maintenance actions.
6-46

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

Sample Size 

• pressure-monitoring system: all casks

• visual inspection of normally inaccessible surfaces: as justified

Data Collection

Data from the examination, including the condition of the coating, 

locations and areas of coating degradation, and corrosion of any 

exposed steel surfaces shall be collected and documented on a 

checklist or visual inspection form. The results of the inspection shall 

be documented and include descriptions of observed aging effects 

and accompanied with sketches and/or photographs. Video 

coverage may also be used to document the inspection. Corrective 

actions resulting from each AMP inspection shall also be 

documented.

Timing of Inspections

Initial visual inspection of normally inaccessible surfaces and 

subcomponents shall be completed before entering the period of 

extended operation. Licensees may credit inspections conducted 

within the 5 years before the period of extended operation. 

5. Monitoring and

Trending

The pressure-monitoring data are trended to provide early detection 

of aging effects that result in leakage and to indicate when corrective 

action needs to be taken to maintain safe storage conditions. 

The results of visual inspections are documented, including evidence 

of corrosion of subcomponents, failure of protective coatings, and 

physical displacement of subcomponents of the cask-sealing system. 

Locations of all areas of degradation are documented to allow a 

direct comparison to prior inspection results. The inspection results 

will be documented and trended to identify aging-related degradation, 

the need for supplemental inspections, mitigation actions, and repair 

or replacement of subcomponents affected by aging. 

Corrective actions will be recorded and trended to evaluate the 

effectiveness of the actions taken. 

6. Acceptance

Criteria

Pressure readings should be within the range stated by the 

Certificate of Compliance (CoC) holder’s, general licensee’s, or sitespecific licensee’s TS. Casks with pressure-monitoring systems in 

the alarmed condition do not meet the acceptance criteria. The CoC 

holder’s, general licensee’s, or site-specific licensee’s TS contain 

pressure-monitoring alarm response procedures that include criteria 

and specifications for corrective actions and response.
6-47

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

For the cask-sealing subcomponents, the acceptance criteria for 

visual inspections are the absence of:

• coating degradation, including blistering, peeling or flaking

• visual indication of corrosion on steel surfaces normally

protected by a coating

• loose or missing hardware

• displaced subcomponents or parts

If coating degradation and corrosion are identified, then the severity 

of corrosion must be determined using approved site-specific or 

general licensee procedures. These may include additional visual, 

surface, or volumetric NDE methods to determine the loss of 

material. Corrosion that results in a loss of material that does not 

meet the design specifications is not acceptable for continued service 

and must be repaired or replaced. 

7. Corrective

Actions

Results that do not meet the acceptance criteria are addressed as 

conditions adverse to quality or significant conditions adverse to 

quality under those specific portions of the specific- or generallicensee QA program approved under 10 CFR Part 72, Subpart G, or 

10 CFR Part 50, Appendix B, respectively. The QA program ensures 

that corrective actions are completed within the specific- or generallicensee’s Corrective Action Program (CAP), and include provisions 

to:

• perform functionality assessments

• perform apparent cause evaluations and root cause evaluations

• address the extent of condition

• determine actions to prevent recurrence for significant

conditions adverse to quality; ensure justifications for

nonrepairs

• trend conditions

• identify operating experience actions, including modification to

the existing AMP (e.g., increased frequency)

• determine if the condition is reportable to the NRC per

10 CFR 72.75

Once the low‐pressure alarm is triggered, troubleshooting of the 

pressure leakage should be performed and, if necessary, an 

engineering evaluation conducted to determine whether the 

degradation of the seal assemblies requires immediate correction. 

8. Confirmation

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 
6-48

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the confirmation process includes provisions to 

preclude repetition of significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine followup actions to verify effective implementation of

corrective actions

• monitor for adverse trends due to recurring or repetitive findings

or observations.

9. Administrative

Controls

The pressure-leakage monitoring system is periodically checked to 

ensure the system is functioning properly. Maintenance, calibration, 

and replacement of pressure transducers are performed in 

accordance with manufacturer requirements.

The administrative controls will be commensurate with the specific or 

general licensee QA program and consistent with 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B. The QA program 

ensures that inspections, evaluations, and corrective actions are 

completed in accordance with the specific or general licensee’s CAP. 

The requirements of 10 CFR Part 72, Appendix G, and 

10 CFR Part 50, Appendix B, are acceptable to address the 

corrective actions, confirmation process, and administrative controls.

10. Operating

Experience

Existing Operational Experience

Helium leakage in two of the TN-68 bolted casks at Peach Bottom 

was detected in October 2010 (NRC, 2013). The root cause 

analyses indicated that the leakage in one cask was caused by a 

material defect in the weld plug that provides sealing of the drilled 

inter‐seal passageway associated with the drain port penetration of 

the cask lid. The defective welds were repaired in accordance with 

the ASME Code and cask design requirements. In the other cask, 

leakage existed in the cask main lid outer closure seal. The seal 

leakage was caused by galvanic corrosion at the interface between 

the aluminum‐clad cask lid seal and the stainless steel clad cask 

body sealing surface of the outer portion of the cask lid seal. The 

corrosion resulted from water infiltration through the access plate in 

the protective cover. The primary corrective actions involved 

improving the access plate design and developing a method for 

verifying protective cover seal integrity. Additional corrective actions 

included a change to the torqueing process for the lid bolts and 

ensuring that the access plate gaskets and O-rings were inspected at 

installation. Corrosion of the TN-32 lid bolts and outer metallic lid 

seals has been observed in the Surry ISFSI owing to external water 
6-49

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

intrusion near the lid bolts and outer metallic seals, resulting in five 

seal replacements. One seal on a CASTOR X/33 cask has also 

been replaced at Surry (Virginia Electric and Power Company, 2002).

An inspection was carried out in 2011 on the lead cask TN-40 01 at 

Prairie Island in conjunction with the license renewal application for 

the ISFSI (Schimmel, 2012). The components inspected included 

the carbon steel cask bottom and underlying concrete pad; the cask 

shell, lid, lid bolts, and trunnions; and the top neutron shield 

enclosure and shield bolts. In addition, the cask protective cover was 

removed to permit visual inspection of the protective cover, bolts, and 

seal; the access cover and bolts; and the overpressure tank, isolation 

valve and tubing, port cover, and port cover bolts. The only 

significant degradation observed was disbondment of approximately 

25 percent of the protective coating on the bottom of the cask, minor 

uniform general corrosion at the upper trunnions, and a very minor 

rust coating on the stainless steel portions of the containment flange. 

In addition, the protective cover was found to have thin uniform 

corrosion on the flange sealing surface on the outer side of the O-ring 

and minor corrosion at the cover bolt holes, and the cask access 

cover had minor rust spots on the outside at the bolt holes. The 

protective cover Viton O-ring was in good condition and was not 

replaced, and the access cover gasket was also in good condition 

but was replaced. The protective cover on TN-40 cask number 13 

was also removed to permit a visual inspection. Here, all 

components were found to be in good condition, and the only 

degradation noted was minor rust stains on the protective coating 

directly below the access cover from corrosion products dripping off 

the access cover.

An inspection of an MC-10 cask was performed after about 20 years 

in service at Surry (Virginia Electric and Power Company, 2006). 

Twelve knurled nuts, which fasten the closure cover to the cask, were 

removed for inspection. While there was some oxidation of the outer 

O-ring edge, the O-ring seal surface and the areas underneath the 

closure cover had no cracks or indications of degradation.

Stress relaxation and leakage tests on Helicoflex metallic seals, 

which are used in the CASTOR and TN cask designs, have been 

conducted in Germany at temperatures from room temperature to 

150 degrees C [302 degrees F]. These tests found that the pressure 

force on the seal and its elastic recovery (or usable resilience) 

decrease approximately linearly when plotted against the logarithm of 

time, but usable lives beyond 40 years with acceptable leak rates are 

extrapolated. Corrosion tests were also initiated on this same seal 

design in 2001 with borated (2,400 ppm) water or a NaCl solution 

(10−3 mol) between the inner and outer jackets of the seal, and no 
6-50

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

increase in leakage rate has been detected to date (Völzke et al., 

2012; Völzke et al., 2013). In addition, the behavior of elastomer 

seals at low temperature (below room temperature) has been studied 

to determine the minimum temperature at which these materials can 

function in DSS applications (Wolff et al., 2013).

References ASME. “Boiler and Pressure Vessel Code Section XI—Rules for 

Inservice Inspection of Nuclear Power Plant Components.” 

New York, New York: American Society of Mechanical Engineers. 

2007.

Code of Federal Regulations. Title 10, Energy, Part 72, “Licensing 

Requirements for the Independent Storage of Spent Nuclear Fuel, 

High-Level Radioactive Waste, and Reactor-Related Greater than 

Class C Waste.” Washington, DC: Office of the Federal Register. 

2015a

Code of Federal Regulations. Title 10, Energy, Part 50, “Domestic 

Licensing of Production and Utilization Facilities.” Washington, DC: 

Office of the Federal Register. 2015b.

NRC. “Confinement Evaluation.” Interim Staff Guidance-5. Rev. 1. 

Washington, DC: U.S. Nuclear Regulatory Commission. 1999.

_____. “Materials Evaluation. ” Interim Staff Guidance-15. Rev. 1. 

Washington, DC: U.S Nuclear Regulatory Commission. 2001.

_____. “Pressure and Helium Leakage Testing of the Confinement 

Boundary of Spent Fuel Dry Storage Systems.” Interim Staff 

Guidance-25. Washington, DC: U.S. Nuclear Regulatory 

Commission. 2010.

_____. “Premature Degradation of Spent Fuel Storage Cask 

Structures and Components from Environmental Moisture.” 

Information Notice 2013-07. Washington, DC: U.S. Nuclear 

Regulatory Commission. 2013.

Schimmel, M. “Prairie Island Independent Spent Fuel Storage 

Installation, Attachment 1 to Letter to U.S. Nuclear Regulatory 

Commission, Responses to Requests for Supplemental Information, 

RE: Prairie Island Independent Spent Fuel Storage Installation 

License Renewal Application.” (TAC No. L24592). ADAMS 

Accession No. ML12065A073. 2012.

Virginia Electric and Power Company. “Surry Independent Spent 

Fuel Storage Installation License Renewal Application.” Docket No. 

72–2. Richmond, Virginia: Virginia Electric and Power Company. 

April 29, 2002.
6-51

Table 6-6 Example aging management program for Bolted Cask Seal Leakage

 Monitoring 

Element Description

_____. “Surry Independent Spent Fuel Storage Installation 

Completion of License Renewal Inspection Requirement.” Docket 

No. 72–2, License Number SNM‐2501. Richmond, Virginia: 

Virginia Electric and Power Company. August 22, 2006.

Völzke, H. and D. Wolff. “Spent Fuel Storage in Dual Purpose Casks 

Beyond the Original Design Basis.” Proceedings of the International 

High‐Level Radioactive Waste Management Conference (IHLRWMC) 

April 28–May 2, 2013. La Grange Park, IL: American Nuclear 

Society. 2013.

Völzke, H., U. Probst, D. Wolff, S. Nagelschmidt, and S. Schultz. 

“Seal and Closure Performance in Long Term Storage.” Proceedings 

of the PSAM11 & ESREL 2012 Conference, Helsinki, Finland. 2012.

Wolff, D., M. Jaunich, and W. Stark. “Investigating the Performance 

of Rubber Seals at Low Temperatures.” Proceedings of the 

International High‐Level Radioactive Waste Management Conference 

(IHLRWMC) April 28–May 2, 2013. La Grange Park, IL: American 

Nuclear Society. 2013.

1 

2 

6-53

1 6.10 Transfer Casks

2 An example AMP for transfer casks is provided below. The AMP manages all transfer cask 

3 subcomponents. This is a condition monitoring program that performs periodic visual 

4 inspections of accessible cask internal and external surfaces to monitor for corrosion, wear, and 

5 loss of preload (bolting). Steel neutron shield water jackets are monitored for wall thickness or 

6 inspected for through-wall leakage. 
6-54

Table 6-7 Example aging management program for Transfer Casks

Element Description

1. Scope of

Program

This program manages loss of material due to corrosion and wear to 

ensure that this aging effect does not challenge the capability of the 

transfer casks to fulfill structural support, radiation shielding, and heat 

transfer functions. 

Visual inspections are performed on the accessible internal and 

external surfaces of steel transfer cask subcomponents that are 

exposed to indoor and outdoor air environments. Inaccessible steel 

surfaces in contact with water neutron shielding are evaluated with 

volumetric wall thickness measurements or inspections for 

through-wall leakage. 

If not addressed with a fatigue analysis, this AMP also includes 

inspections of trunnions for cracking. 

An additional site-specific AMP may be required to manage 

protective coatings that are credited in the design basis for 

preventing corrosion of the base metal.

2. Preventive

Actions

This program is a condition-monitoring program to detect evidence of 

degradation. It does not provide guidance for prevention of aging.

3. Parameters

Monitored/

Inspected

This program monitors the condition of internal and external steel 

surfaces to identify general, pitting, crevice, and galvanic corrosion, 

and wear. The condition of inaccessible steel internal surfaces that 

are continuously or intermittently exposed to a liquid neutron shield 

are monitored from the external side of the shield shell. 

Parameters monitored or inspected for accessible surfaces include:

• visual evidence of surface discontinuities and imperfections

indicative of corrosion

• visual evidence of coating degradation (e.g., blisters, cracking,

flaking, delamination) indicative of corrosion of the base metal

Parameters monitored or inspected to evaluate inaccessible steel 

surfaces exposed to a liquid neutron shield include either:

• wall thickness

• visual evidence of leakage on external surfaces

If trunnions are not addressed with a fatigue analysis, trunnion 

surfaces are monitored for the presence of cracks.

4. Detection of

Aging Effects

Normally Accessible Surfaces

Visual inspections are performed in accordance with the ASME Code 

Section XI, Article IWA-2213, for VT-3 examinations. The 

inspections cover 100 percent of the normally accessible steel cask 
6-55

Table 6-7 Example aging management program for Transfer Casks

Element Description

surfaces, including the cask exterior, cask interior cavity, lid surfaces, 

and the cask bottom (during lifting or down ending).

Normally Inaccessible Internal Surfaces (liquid neutron shield)

Wall thicknesses of steel liquid neutron shield subcomponents are 

measured with ultrasonic thickness techniques. Alternatively, the 

condition of internal surfaces of the neutron shield shell is monitored 

by inspections for leakage when the shield is filled with water, 

following ASME Code Section XI, Article IWA-2212, VT-2 (visual) 

inspection requirements. 

Trunnions

If the fatigue of trunnions is not addressed with an analysis, surface 

or volumetric inspection techniques are performed on 100 percent of 

trunnion surfaces to identify the presence of fatigue cracks.

Sample Size

All transfer casks are inspected.

Frequency

Inspections are conducted at least once every 5 years. If a transfer 

cask is used less frequently than once every 5 years, inspections are 

conducted before its use in each loading campaign. 

Data Collection

Data from the examination, including evidence of degradation and its 

extent and location, shall be documented on a checklist or inspection 

form. The results of the inspection shall be documented, including 

descriptions of observed aging effects and supporting sketches, 

photographs, or video. Corrective actions resulting from each AMP 

inspection shall also be documented.

Timing

Initial inspections are completed before the use of the transfer casks 

in the first loading campaign in the period of extended operation.

5. Monitoring and 

Trending

Inspection results are compared to those obtained during previous 

inspections, so that the progression of degradation can be evaluated 

and predicted. 

Monitoring and trending methods reference plans/procedures used 

to:

• establish a baseline before or at the beginning of the period of 

extended operation

• track trending of parameters or effects not corrected following a 

previous inspection
6-56

Table 6-7 Example aging management program for Transfer Casks

Element Description

— the locations, size, and depth of any areas of corrosion

— the disposition of components with identified aging 

effects and the results of supplemental inspections

6. Acceptance

Criteria

For accessible surfaces, including trunnions, acceptance criteria are:

• no detectable loss of material from the base metal, including

uniform wall thinning, localized corrosion pits, crevice

corrosion, and wear scratches/gouges

• no red-orange-colored corrosion products on the base metal or

coatings

• no coating defects (e.g., blisters, cracking, flaking,

delamination)

For inaccessible internal surfaces, the acceptance criteria are no 

evidence of leakage of the water neutron shield or loss of wall 

thickness beyond a predetermined limit established by systemspecific design standards or industry codes and standards.

If evidence of corrosion, wear, or coating degradation are identified, 

then the severity of the degradation of the base metal must be 

determined using approved site-specific procedures. These may 

include additional visual, surface, or volumetric NDE methods to 

determine the loss of material. 

Alternative acceptance criteria are developed from system-specific 

design standards, industry codes or standards, or engineering 

evaluation. Where possible, acceptance criteria are quantitative 

(e.g., minimum wall thickness). Where qualitative acceptance criteria 

are used, the criteria are sufficiently clear to reasonably ensure that a 

singular decision is derived based on the observed condition, 

avoiding the use of ambiguous phrases (e.g., significant, moderate). 

EPRI Technical Reports, TR-1007933, “Aging Assessment Field 

Guide” (EPRI, 2003), and TR-1009743, “Aging Identification and 

Assessment Checklist: Mechanical Components” (EPRI, 2004), 

provide general guidance for the evaluation of materials and the 

development of criteria for their acceptance when performing visual 

inspections.

7. Corrective

Actions

Results that do not meet the acceptance criteria are addressed as 

conditions adverse to quality or significant conditions adverse to 

quality under those specific portions of the specific- or generallicensee QA program approved under 10 CFR Part 72, Subpart G, or 

10 CFR Part 50, Appendix B, respectively. The QA program ensures 

that corrective actions are completed within the specific- or generallicensee’s CAP, and include provisions to:
6-57

Table 6-7 Example aging management program for Transfer Casks

Element Description

• perform functionality assessments

• perform apparent cause evaluations and root cause evaluations

• address the extent of condition

• determine actions to prevent recurrence for significant 

conditions adverse to quality; ensure justifications for 

nonrepairs

• trend conditions

• identify operating experience actions, including modification to 

the existing AMP (e.g., increased frequency)

• determine if the condition is reportable to the NRC per 

10 CFR 72.75

8. Confirmation 

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the confirmation process includes provisions to 

preclude repetition of significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine followup actions to verify effective implementation of 

corrective actions

• monitor for adverse trends due to recurring or repetitive findings 

or observations.

9. Administrative 

Controls

The administrative controls are addressed through those portions of 

the specific or general licensee’s QA program that are used to meet 

10 CFR Part 72, Subpart G, or 10 CFR Part 50, Appendix B.

10. Operating 

Experience

External surface inspections through system inspections and 

walkdowns in support of the Maintenance Rule (10 CFR Part 50.65) 

have proven effective in maintaining the material condition of nuclear 

power plant systems. 

References EPRI. “Aging Assessment Field Guide.” Technical Report 1007933. 

Palo Alto, California: Electric Power Research Institute. December 

2003.

_____. “Aging Identification and Assessment Checklist–Mechanical 

Components.” Technical Report 1009743. Palo Alto, California: 

Electric Power Research Institute. August 27, 2004.

1

2

6-59

1 6.11 High-Burnup Fuel Monitoring And Assessment

An example of a High Burnup (HBU) Fuel 2 1 Monitoring and Assessment Program is provided 

3 below. This is a licensee program that monitors and assesses data and other information 

4 regarding HBU fuel performance, to confirm that the design-bases HBU fuel configuration is 

5 maintained during the period of extended operation. This example HBU Fuel Monitoring and 

6 Assessment Program relies on a surrogate demonstration program to provide data on HBU fuel 

7 performance. Guidance for determining if a surrogate demonstration program can provide the 

8 data to support a licensee’s HBU Fuel Monitoring and Assessment Program is given in 

9 Appendix D of NUREG–1927, Revision 1 (NRC, 2016). Although this example focuses on the 

10 use of a surrogate demonstration program, a licensee may use alternative approaches that are 

11 appropriately justified, including the use of test or research results and safety analyses for the 

12 fuel, to demonstrate that the DSS’s intended functions continue to be met during the period of 

13 extended operation. 

14 The aging management review is not expected to identify any aging effects that could lead to 

15 fuel reconfiguration, as long as the HBU fuel is stored in a dry inert environment, temperature 

16 limits are maintained, and thermal cycling is limited. Short-term testing (i.e., laboratory scale 

17 testing up to a few months) and scientific analyses examining the performance of HBU fuel have 

18 provided a foundation for the technical basis that storage of HBU fuel in the period of extended 

19 operation may be performed safely and in compliance with regulations. However, there has 

20 been relatively little operating experience, to date, with dry storage of HBU fuel. 

21 Therefore, the purpose of a HBU Fuel Monitoring and Assessment Program is to monitor and 

22 assess data and other information regarding HBU fuel performance to confirm there is no 

23 degradation of HBU fuel that would result in an unanalyzed configuration during the period of 

24 extended operation. The following description of an example HBU Fuel Monitoring and 

25 Assessment Program presents the applicable information in a format using each element of an 

26 effective AMP, to provide a framework for such a monitoring and assessment program. 

27

 

1These are fuel assemblies with discharge burnup greater than 45 gigawatt-days per metric ton of uranium 

(GWd/MTU).
6-60

Table 6-8 Example aging management program for High-Burnup Fuel Monitoring and

 Assessment 

Element Description

1. Scope of the

Program

The scope of the program provides a description of (i) the design 

bases characteristics of the HBU fuel, (ii) the surrogate 

demonstration program that will be used to provide data on the 

applicable design-bases HBU fuel performance, and (iii) how the 

parameters of the surrogate demonstration program are applicable 

to the design-bases HBU fuel.

Aging effects will be determined for material/environment 

combinations per an alternative surrogate demonstration program 

meeting the guidance in Appendix D of NUREG–1927, Revision 1 

(NRC, 2016).

Example language to address this “scope of the program” element 

follows: Fuel stored in a [define cask/canister model] is limited to an 

assembly average burnup of [define design-bases limit] GWd/MTU. 

The cladding materials for the HBU fuel are [define types of 

cladding], and the fuel is stored in a dry helium environment. HBU 

fuel was first placed into dry storage in a [define cask/canister 

model] on [start date of storage term of first storage of HBU fuel].

The program relies on the joint EPRI and DOE HBU Dry Storage 

Cask Research and Development Project (HDRP) (EPRI, 2014), 

conducted in accordance with the guidance in Appendix D of 

NUREG–1927, Revision 1, as a surrogate demonstration program 

that monitors the performance of HBU fuel in dry storage.

The HDRP is a program designed to collect data from an SNF 

storage system containing HBU fuel in a dry helium environment. 

The program entails loading and storing an AREVA TN-32 bolted lid 

cask (the “Research Project Cask”) at Dominion Virginia Power’s 

North Anna Power Station with intact HBU fuel (of nominal burnups 

ranging between 53 GWd/MTU and 58 GWd/MTU). The fuel to be 

used in the program includes four kinds of cladding (Zircaloy-4, 

low-tin Zircaloy-4, ZIRLOTM, and M5TM). The Research Project Cask 

is to be licensed to the temperature limits contained in ISG-11, 

Rev. 3 (NRC, 2003), and loaded such that the fuel cladding 

temperature is as close to the limit as practicable. [If an alternative 

surrogate demonstration program is used, provide a description of 

the program.]

The parameters of the surrogate demonstration program are 

applicable to the design-bases HBU fuel, as the (i) maximum burnup 

of the design-bases HBU fuel [define value] is less than the burnup 

of the fuel in the surrogate demonstration program [define value], 

(ii) the cladding type of the design-bases HBU fuel [define type] is 

the same as the surrogate demonstration program [define type], and

(iii) the temperatures in the surrogate demonstration program [define 
6-61

Table 6-8 Example aging management program for High-Burnup Fuel Monitoring and

 Assessment 

Element Description

values] bound the design bases temperature/heat load of the loaded 

systems [define values]. 

2. Preventive Actions There are no specific preventive actions associated with this HBU

Fuel Monitoring and Assessment Program. However, the applicant 

should discuss the design-bases characteristics of the 

licensed/certified DSS, in terms of initial cask loading operations, to 

show the HBU fuel is stored in a dry inert environment.

Example language follows:

During the initial loading operations of the cask/canister, the design 

and ISFSI TS require that the fuel be stored in a dry inert 

environment. TS [name and number] demonstrates that the 

cask/canister cavity is dry by maintaining a cavity absolute pressure 

less than or equal to [value] for a [time period] with the cask/canister 

isolated from the vacuum pump. TS [name and number], requires 

that the cask/canister then be backfilled with helium. These two TS 

requirements ensure that the HBU fuel is stored in an inert 

environment, thus preventing cladding degradation due to oxidation 

mechanisms. TS [name and number] also requires that the helium 

environment be established within [time] hours of commencing 

cask/canister draining. The cask/canister is loaded in accordance 

with the criteria of ISG-11, Revision 3 (NRC, 2003).

3. Parameters

Monitored or

Inspected

The applicant identifies the parameters monitored and inspected in a 

surrogate demonstration program that are applicable to its particular 

design-bases HBU fuel and describes how this meets the guidance 

of Appendix D of NUREG–1927, Revision 1.

4. Detection of Aging

Effects

The applicant identifies the detection of aging effects in a surrogate 

demonstration program that are applicable to its particular 

design-bases HBU fuel and describes how this meets the guidance 

of Appendix D of NUREG–1927, Revision 1.

5. Monitoring and

Trending

As information/data from a surrogate demonstration program or from 

other sources (such as testing or research results and scientific 

analyses) become available, the licensee will monitor, evaluate, and 

trend the information via its operating experience program and/or the 

CAP to determine what actions should be taken. 

The licensee will evaluate the information/data from a surrogate 

demonstration program or from other sources to determine whether 

the acceptance criteria in Element 6 are met. 

• If all of the acceptance criteria are met, no further assessment

is needed.

• If any of the acceptance criteria are not met, the licensee must

conduct additional assessments and implement appropriate

corrective actions (see Element 7).
6-62

Table 6-8 Example aging management program for High-Burnup Fuel Monitoring and

Assessment 

Element Description

Formal evaluations of the aggregate information from a surrogate 

demonstration program and other available domestic or international 

operating experience (including data from monitoring and inspection 

programs, NRC-generated communications, and other information) 

will be performed at specific points in time during the period of 

extended operation, as delineated in Table B-4 of NUREG–1927, 

Revision 1. 

6. Acceptance

Criteria

The HBU Fuel Monitoring and Assessment Program acceptance 

criteria are:

• hydrogen content–Maximum hydrogen content of the cover gas

over the approved storage period should be extrapolated from

the gas measurements to be less than the design-bases limit

for hydrogen content.

• moisture content–The moisture content in the cask/canister,

accounting for measurement uncertainty, should be less than

the expected upper-bound moisture content per the

design-bases drying process1

.

• fuel condition/performance2

–nondestructive examination

(e.g., fission gas analysis) and destructive examination (e.g., to

obtain data on creep, fission gas release, hydride reorientation,

cladding oxidation, and cladding mechanical properties) should

confirm the design-bases fuel condition (i.e., no changes to the

analyzed fuel configuration considered in the safety analyses of

the approved design bases).

The applicant should provide information on the design-bases 

characteristics of the DSS, with regard to these criteria. The 

applicant should reference the source of specific values, or explain 

any assumptions made, for defining design-bases characteristics of 

the fuel condition/performance. 

7. Corrective Actions The corrective actions are in accordance with the specific or general

licensee QA program and consistent with 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. 

Corrective actions should be implemented if data from a surrogate 

demonstration program or other sources of information indicate that 

1The applicant will need to provide the expected upper-bound moisture content based on its design-bases drying 

process. If the design-bases drying process involves a vacuum drying method of evacuating a cask/canister to less 

than or equal to 3 torr and maintaining a constant pressure for 30 minutes after the cask/canister is isolated from the 

vacuum pump, the expected water content is about 0.43 gram-mole. (See NRC, 2010.) 

2While it is not a fuel performance criterion, the spatial distribution and time history of the temperature must be known 

to evaluate the relationship between the performance of the rods in a surrogate demonstration program and the HBU

fuel rod behavior expected in the cask.
6-63

Table 6-8 Example aging management program for High-Burnup Fuel Monitoring and

 Assessment 

Element Description

any of the HBU Fuel Monitoring and Assessment Program 

acceptance criteria (in Element 6) are not met.

If any of the acceptance criteria are not met, the licensee will: 

(i) assess fuel performance (impacts on fuel and changes to fuel 

configuration) 

(ii) assess the design-bases safety analyses, considering 

degraded fuel performance (and any changes to fuel 

configuration), to determine the ability of the DSS to continue to 

perform its intended functions under normal, off-normal, and 

accident conditions. 

The licensee will determine what corrective actions should be taken 

to: 

(i) manage fuel performance, if any

(ii) manage impacts related to degraded fuel performance to 

ensure that all intended functions for the DSS are met. 

In addition, the licensee will obtain the necessary NRC approval in 

the appropriate licensing/certification process for modification of the 

design bases to address any conditions outside of the approved 

design bases.

8. Confirmation

Process

The confirmation process is commensurate with the specific- or 

general-licensee QA program approved under 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the confirmation process includes provisions 

to preclude repetition of significant conditions adverse to quality.

The confirmation process describes or references procedures to:

• determine followup actions to verify effective implementation of

corrective actions

• monitor for adverse trends due to recurring or repetitive

findings or observations.

9. Administrative

Controls

The administrative controls are in accordance with the specific or 

general licensee QA program and consistent with 10 CFR Part 72, 

Subpart G, or 10 CFR Part 50, Appendix B, respectively. The QA 

program ensures that the administrative controls include provisions 

that define:

• formal review and approval processes

• record retention requirements

• document control
6-64

Table 6-8 Example aging management program for High-Burnup Fuel Monitoring and

Assessment 

Element Description

10. Operating

Experience

The program references and evaluates applicable operating 

experience, including:

• internal and industrywide condition reports

• internal and industrywide corrective action reports

• vendor-issued safety bulletins

• NRC Information Notices

• applicable DOE or industry initiatives (e.g., HDRP)

• applicable research (e.g., Oak Ridge National Laboratory

studies on bending responses of the fuel, Argonne National

Laboratory and Central Research Institute of Electric Power

Industry studies on hydride reorientation effects)

The review of operating experience clearly identifies any HBU fuel 

degradation as either age related or event driven, with proper 

justification for that assessment. Past operating experience 

supports the adequacy of the HBU Fuel Monitoring and Assessment 

Program.

Surrogate demonstration programs with storage conditions and fuel 

types similar to those in the licensed/certified DSS that meet the 

guidance in Appendix D of NUREG–1927, Revision 1, are a viable 

method to obtain operating experience. 

New data/research on fuel performance from both domestic and 

international sources that are relevant to the licensed/certified HBU 

fuel in the DSS should be evaluated on a periodic basis.

References EPRI. “HBU Dry Storage Cask Research and Development Project 

Final Test Plan.” DOE Contract No.: DE-NE-0000593. 

Palo Alto, California: Electric Power Research Institute. 2014.

NRC. “NRC Interim Staff Guidance 11, “Cladding Considerations for 

the Transportation and Storage of Spent Fuel.” Rev. 3. 

ADAMS Accession No. ML033230335. Washington, DC: 

U.S. Nuclear Regulatory Commission. November 17, 2003. 

_____. NUREG–1536, “Standard Review Plan for Spent Fuel Dry 

Storage Systems at a General License Facility.” Rev. 1. 

ADAMS Accession No. ML101040620. Washington, DC. U.S. 

Nuclear Regulatory Commission. 2010.

_____. NUREG–1927, “Standard Review Plan for Renewal of 

Specific Licenses and Certificates of Compliance for Dry Storage of 

Spent Nuclear Fuel.” Revision 1. ADAMS Accession 

No. ML16179A148. Washington, DC: U.S. Nuclear Regulatory 

Commission. 2016. .

1 




NUREG

-2214

Draft

Managing Aging Processes In Storage (MAPS) Report