NUREG-1617

Standard Review Plan

for Transportation Packages for Spent Nuclear Fuel

Final Report

U.S. Nuclear Regulatory Commission

Office of Nuclear Material Safety and Safeguards

NUREG-1617

Standard Review Plan

for Transportation Packages for Spent Nuclear Fuel

Final Report

Manuscript Completed: January 2000

Date Published: March 2000

Spent Fuel Project Office

Office of Nuclear Material Safety and Safeguards

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001
NUREG-1617 ii
iii NUREG-1617

ABSTRACT

The Standard Review Plan (SRP) for Transportation Packages for Spent Nuclear Fuel provides guidance

for the review and approval of applications for packages used to transport spent nuclear fuel under

10 CFR Part 71.

This document is intended for use by the U.S. Nuclear Regulatory Commission (NRC) staff. Its

objectives are to (1) summarize 10 CFR Part 71 requirements for spent fuel transport package approval,

(2) describe the procedures by which NRC staff determines that these requirements have been satisfied,

and (3) document the practices used by the staff in reviews of package applications.

This NUREG is expected to be updated on a periodic basis. As issues arise between updates, the Spent

Fuel Project Office will issue interim staff guidance (ISG) where the SRP guidance needs revising. ISG’s

are placed in the NRC public document room and on the NRC WEB for public information. ISG’s, as

issued, replace specific portions of the SRP. Comments regarding errors or omissions, as well as

suggestions for improvement of this NUREG and subsequent ISG, should be sent to the Director, Spent

Fuel Project Office, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
NUREG-1617 iv
v NUREG-1617

CONTENTS

ABSTRACT ................................................................. iii

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

GLOSSARY ................................................................. xv

INTRODUCTION .............................................................. 1

1 GENERAL INFORMATION REVIEW .......................................... 1-1

1.1 REVIEW OBJECTIVE .................................................. 1-1

1.2 AREAS OF REVIEW ................................................... 1-1

1.2.1 General SAR Format ................................................ 1-1

1.2.2 Package Design Information .......................................... 1-1

1.2.3 Package Description ................................................ 1-1

1.2.4 Compliance with 10 CFR Part 71 ....................................... 1-1

1.2.5 Appendix ........................................................ 1-1

1.3 REGULATORY REQUIREMENTS ........................................ 1-1

1.3.1 General SAR Format ................................................ 1-1

1.3.2 Package Design Information .......................................... 1-2

1.3.3 Package Description ................................................ 1-2

1.3.4 Compliance with 10 CFR Part 71 ....................................... 1-2

1.4 ACCEPTANCE CRITERIA .............................................. 1-3

1.4.1 General SAR Format ................................................ 1-3

1.4.2 Package Design Information .......................................... 1-3

1.4.3 Package Description ................................................ 1-3

1.4.4 Compliance with 10 CFR Part 71 ....................................... 1-3

1.5 REVIEW PROCEDURES ................................................ 1-3

1.5.1 General SAR Format ................................................ 1-4

1.5.2 Package Design Information .......................................... 1-4

1.5.3 Package Description ................................................ 1-9

1.5.4 Compliance with 10 CFR Part 71 ...................................... 1-12

1.5.5 Appendix ....................................................... 1-13

1.6 EVALUATION FINDINGS ............................................. 1-13

1.7 REFERENCES ....................................................... 1-13

2 STRUCTURAL REVIEW .................................................... 2-1

2.1 REVIEW OBJECTIVE .................................................. 2-1

2.2 AREAS OF REVIEW ................................................... 2-1

2.2.1 Description of Structural Design ........................................ 2-1

2.2.2 Material Properties ................................................. 2-1

2.2.3 Lifting and Tie-down Standards for All Packages ........................... 2-1

2.2.4 General Considerations for Structural Evaluation of Packaging .................. 2-1

2.2.5 Normal Conditions of Transport ........................................ 2-1

2.2.6 Hypothetical Accident Conditions ....................................... 2-1
NUREG-1617 vi

2.2.7 Special Requirement for Irradiated Nuclear Fuel Shipments .................... 2-2

2.2.8 Internal Pressure Test ............................................... 2-2

2.2.9 Appendix ........................................................ 2-2

2.3 REGULATORY REQUIREMENTS ........................................ 2-2

2.3.1 Description of Structural Design ........................................ 2-2

2.3.2 Material Properties ................................................. 2-2

2.3.3 Lifting and Tie-down Standards for All Packages ........................... 2-2

2.3.4 General Considerations for Structural Evaluation of Packaging .................. 2-2

2.3.5 Normal Conditions of Transport ........................................ 2-3

2.3.6 Hypothetical Accident Conditions ....................................... 2-3

2.3.7 Special Requirement for Irradiated Nuclear Fuel Shipments .................... 2-3

2.3.8 Internal Pressure Test ............................................... 2-3

2.4 ACCEPTANCE CRITERIA .............................................. 2-3

2.4.1 Description of Structural Design ........................................ 2-3

2.4.2 Material Properties ................................................. 2-3

2.4.3 Lifting and Tie-Down Standards for All Packages ........................... 2-4

2.4.4 General Considerations for Structural Evaluation of Packaging .................. 2-4

2.4.5 Normal Conditions of Transport ........................................ 2-4

2.4.6 Hypothetical Accident Conditions ....................................... 2-4

2.4.7 Special Requirement for Irradiated Nuclear Fuel Shipments .................... 2-4

2.4.8 Internal Pressure Test ............................................... 2-4

2.5 REVIEW PROCEDURES ................................................ 2-4

2.5.1 Description of Structural Design ........................................ 2-4

2.5.2 Material Properties ................................................. 2-7

2.5.3 Lifting and Tie-Down Standards for All Packages ........................... 2-8

2.5.4 General Considerations for Structural Evaluation of Packaging .................. 2-9

2.5.5 Normal Conditions of Transport ....................................... 2-11

2.5.6 Hypothetical Accident Conditions ...................................... 2-13

2.5.7 Special Requirement for Irradiated Nuclear Fuel Shipments ................... 2-15

2.5.8 Internal Pressure Test .............................................. 2-15

2.5.9 Appendix ....................................................... 2-15

2.6 EVALUATION FINDINGS ............................................. 2-15

2.6.1 Description of Structural Design ....................................... 2-15

2.6.2 Material Properties ................................................ 2-16

2.6.3 Lifting and Tie-down Standards for All Packages .......................... 2-16

2.6.4 General Considerations for Structural Evaluation of Packaging ................. 2-16

2.6.5 Normal Conditions of Transport ....................................... 2-16

2.6.6 Hypothetical Accident Conditions ...................................... 2-16

2.6.7 Special Requirement for Irradiated Nuclear Fuel Shipments ................... 2-16

2.6.8 Internal Pressure Test .............................................. 2-16

2.7 REFERENCES ....................................................... 2-16

3 THERMAL REVIEW ....................................................... 3-1

3.1 REVIEW OBJECTIVE .................................................. 3-1

3.2 AREAS OF REVIEW ................................................... 3-1

3.2.1 Description of the Thermal Design ...................................... 3-1
vii NUREG-1617

3.2.2 Material Properties and Component Specifications ........................... 3-1

3.2.3 General Considerations for Thermal Evaluations ............................ 3-1

3.2.4 Evaluation of Accessible Surface Temperatures ............................ 3-1

3.2.5 Evaluation under Normal Conditions of Transport ........................... 3-1

3.2.6 Evaluation under Hypothetical Accident Conditions .......................... 3-1

3.2.7 Appendix ........................................................ 3-1

3.3 REGULATORY REQUIREMENTS ........................................ 3-2

3.3.1 Description of the Thermal Design ...................................... 3-2

3.3.2 Material Properties and Component Specifications ........................... 3-2

3.3.3 General Considerations for Thermal Evaluations ............................ 3-2

3.3.4 Evaluation of Accessible Surface Temperatures ............................ 3-2

3.3.5 Thermal Evaluation under Normal Conditions of Transport ..................... 3-2

3.3.6 Thermal Evaluation under Hypothetical Accident Conditions .................... 3-3

3.4 ACCEPTANCE CRITERIA .............................................. 3-3

3.4.1 Description of the Thermal Design ...................................... 3-3

3.4.2 Material Properties and Component Specifications ........................... 3-3

3.4.3 General Considerations for Thermal Evaluations ............................ 3-3

3.4.4 Evaluation of Accessible Surface Temperature ............................. 3-3

3.4.5 Thermal Evaluation under Normal Conditions of Transport ..................... 3-3

3.4.6 Thermal Evaluation under Hypothetical Accident Conditions .................... 3-3

3.5 REVIEW PROCEDURES ................................................ 3-4

3.5.1 Description of the Thermal Design ...................................... 3-4

3.5.2 Material Properties and Component Specifications ........................... 3-6

3.5.3 General Considerations for Thermal Evaluations ............................ 3-8

3.5.4 Evaluation of Accessible Surface Temperatures ........................... 3-11

3.5.5 Thermal Evaluation under Normal Conditions of Transport .................... 3-12

3.5.6 Thermal Evaluation under Hypothetical Accident Conditions ................... 3-14

3.5.7 Appendix ....................................................... 3-15

3.6 EVALUATION FINDINGS ............................................. 3-18

3.6.1 Description of the Thermal Design ..................................... 3-18

3.6.2 Material Properties and Component Specifications .......................... 3-18

3.6.3 General Considerations for Thermal Evaluations ........................... 3-18

3.6.4 Evaluation of Accessible Surface Temperature ............................ 3-18

3.6.5 Evaluation under Normal Conditions of Transport .......................... 3-18

3.6.6 Evaluation under Hypothetical Accident Conditions ......................... 3-18

3.7 REFERENCES ....................................................... 3-18

4 CONTAINMENT REVIEW ................................................... 4-1

4.1 REVIEW OBJECTIVE .................................................. 4-1

4.2 AREAS OF REVIEW ................................................... 4-1

4.2.1 Description of Containment System ..................................... 4-1

4.2.2 Containment under Normal Conditions of Transport .......................... 4-1

4.2.3 Containment under Hypothetical Accident Conditions ......................... 4-1

4.2.4 Appendix ........................................................ 4-1

4.3 REGULATORY REQUIREMENTS ........................................ 4-1

4.3.1 Description of Containment System ..................................... 4-1
NUREG-1617 viii

4.3.2 Containment under Normal Conditions of Transport ......................... 4-2

4.3.3 Containment under Hypothetical Accident Conditions ......................... 4-2

4.4 ACCEPTANCE CRITERIA .............................................. 4-2

4.4.1 Description of Containment System ..................................... 4-2

4.4.2 Containment under Normal Conditions of Transport .......................... 4-3

4.4.3 Containment under Hypothetical Accident Conditions ......................... 4-3

4.5 REVIEW PROCEDURES ................................................ 4-3

4.5.1 Description of the Containment System ................................... 4-3

4.5.2 Containment under Normal Conditions of Transport .......................... 4-6

4.5.3 Containment under Hypothetical Accident Conditions ......................... 4-8

4.5.4 Appendix ........................................................ 4-9

4.6 EVALUATION FINDINGS .............................................. 4-9

4.6.1 Description of Containment System ..................................... 4-9

4.6.2 Containment under Normal Conditions of Transport .......................... 4-9

4.6.3 Containment under Hypothetical Accident Conditions ......................... 4-9

4.7 REFERENCES ....................................................... 4-10

5 SHIELDING REVIEW ....................................................... 5-1

5.1 REVIEW OBJECTIVE .................................................. 5-1

5.2 AREAS OF REVIEW ................................................... 5-1

5.2.1 Description of the Shielding Design ...................................... 5-1

5.2.2 Source Specification ................................................ 5-1

5.2.3 Model Specification ................................................. 5-1

5.2.4 Evaluation ........................................................ 5-1

5.2.5 Appendix ........................................................ 5-1

5.3 REGULATORY REQUIREMENTS ........................................ 5-1

5.3.1 Description of the Shielding Design ...................................... 5-1

5.3.2 Source Specification ................................................ 5-2

5.3.3 Model Specification ................................................. 5-2

5.3.4 Evaluation ........................................................ 5-2

5.4 ACCEPTANCE CRITERIA .............................................. 5-2

5.4.1 Description of the Shielding Design ...................................... 5-2

5.4.2 Source Specification ................................................ 5-2

5.4.3 Model Specification ................................................. 5-2

5.4.4 Evaluation ........................................................ 5-3

5.5 REVIEW PROCEDURES ................................................ 5-4

5.5.1 Description of the Shielding Design ...................................... 5-5

5.5.2 Source Specification ................................................ 5-5

5.5.3 Model Specification ................................................. 5-7

5.5.4 Evaluation ........................................................ 5-8

5.5.5 Appendix ....................................................... 5-10

5.6 EVALUATION FINDINGS ............................................. 5-10

5.6.1 Description of the Shielding Design ..................................... 5-10

5.6.2 Source Specification ............................................... 5-10

5.6.3 Model Specification ................................................ 5-11

5.6.4 Evaluation ....................................................... 5-11
ix NUREG-1617

5.7 REFERENCES ....................................................... 5-11

6 CRITICALITY REVIEW ..................................................... 6-1

6.1 REVIEW OBJECTIVE .................................................. 6-1

6.2 AREAS OF REVIEW ................................................... 6-1

6.2.1 Description of Criticality Design ........................................ 6-1

6.2.2 Spent Nuclear Fuel Contents .......................................... 6-1

6.2.3 General Considerations for Criticality Evaluations ........................... 6-1

6.2.4 Single Package Evaluation ............................................ 6-1

6.2.5 Evaluation of Package Arrays under Normal Conditions of Transport ............. 6-1

6.2.6 Evaluation of Package Arrays under Hypothetical Accident Conditions ............ 6-1

6.2.7 Benchmark Evaluations .............................................. 6-1

6.2.8 Burnup Credit ..................................................... 6-1

6.2.9 Appendix ........................................................ 6-2

6.3 REGULATORY REQUIREMENTS ........................................ 6-3

6.3.1 Description of Criticality Design ........................................ 6-3

6.3.2 Spent Nuclear Fuel Contents .......................................... 6-3

6.3.3 General Considerations for Criticality Evaluations ........................... 6-3

6.3.4 Single Package Evaluation ............................................ 6-3

6.3.5 Evaluation of Package Arrays under Normal Conditions of Transport ............. 6-4

6.3.6 Evaluation of Package Arrays under Hypothetical Accident Conditions ............ 6-4

6.3.7 Benchmark Evaluations .............................................. 6-4

6.3.8 Burnup Credit ..................................................... 6-4

6.4 ACCEPTANCE CRITERIA .............................................. 6-4

6.4.1 Description of Criticality Design ........................................ 6-4

6.4.2 Spent Nuclear Fuel Contents .......................................... 6-4

6.4.3 General Considerations for Criticality Evaluations ........................... 6-4

6.4.4 Single Package Evaluation ............................................ 6-5

6.4.5 Evaluation of Package Arrays under Normal Conditions of Transport ............. 6-5

6.4.6 Evaluation of Package Arrays under Hypothetical Accident Conditions ............ 6-5

6.4.7 Benchmark Evaluations .............................................. 6-5

6.4.8 Burnup Credit Evaluation ............................................. 6-5

6.5 REVIEW PROCEDURES ................................................ 6-7

6.5.1 Description of the Criticality Design ..................................... 6-7

6.5.2 Spent Nuclear Fuel Contents ......................................... 6-10

6.5.3 General Considerations for Criticality Evaluations .......................... 6-11

6.5.4 Single Package Evaluation ........................................... 6-14

6.5.5 Evaluation of Package Arrays under Normal Conditions of Transport ............ 6-14

6.5.6 Evaluation of Package Arrays under Hypothetical Accident Conditions ........... 6-15

6.5.7 Benchmark Evaluations ............................................. 6-15

6.5.8 Burnup Credit .................................................... 6-17

6.5.9 Appendix ....................................................... 6-17

6.6 EVALUATION FINDINGS ............................................. 6-17

6.6.1 Description of Criticality Design ....................................... 6-17

6.6.2 Spent Nuclear Fuel Contents ......................................... 6-17

6.6.3 General Considerations for Criticality Evaluations .......................... 6-17
NUREG-1617 x

6.6.4 Single Package Evaluation ........................................... 6-17

6.6.5 Evaluation of Package Arrays under Normal Conditions of Transport ............ 6-18

6.6.6 Evaluation of Package Arrays under Hypothetical Accident Conditions ........... 6-18

6.6.7 Benchmark Evaluations ............................................. 6-18

6.6.8 Burnup Credit .................................................... 6-18

6.7 REFERENCES ....................................................... 6-18

7 OPERATING PROCEDURES REVIEW ......................................... 7-1

7.1 REVIEW OBJECTIVE .................................................. 7-1

7.2 AREAS OF REVIEW ................................................... 7-1

7.2.1 Package Loading ................................................... 7-1

7.2.2 Package Unloading ................................................. 7-1

7.2.3 Preparation of Empty Package for Transport ............................... 7-1

7.2.4 Other Procedures .................................................. 7-1

7.2.5 Appendix ........................................................ 7-1

7.3 REGULATORY REQUIREMENTS ........................................ 7-1

7.3.1 Package Loading ................................................... 7-1

7.3.2 Package Unloading ................................................. 7-3

7.3.3 Preparation of Empty Package for Transport ............................... 7-3

7.3.4 Other Procedures .................................................. 7-3

7.4 ACCEPTANCE CRITERIA .............................................. 7-3

7.4.1 Package Loading ................................................... 7-3

7.4.2 Package Unloading ................................................. 7-3

7.4.3 Preparation of Empty Package for Transport ............................... 7-3

7.4.4 Other Procedures .................................................. 7-3

7.5 REVIEW PROCEDURES ................................................ 7-4

7.5.1 Package Loading ................................................... 7-5

7.5.2 Package Unloading ................................................. 7-6

7.5.3 Preparation of Empty Package for Transport ............................... 7-7

7.5.4 Other Procedures .................................................. 7-8

7.5.5 Appendix ........................................................ 7-8

7.6 EVALUATION FINDINGS .............................................. 7-8

7.6.1 Package Loading ................................................... 7-8

7.6.2 Package Unloading ................................................. 7-8

7.6.3 Preparation of Empty Package for Transport ............................... 7-8

7.6.4 Other Procedures .................................................. 7-9

7.7 REFERENCES ........................................................ 7-9

8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW .................. 8-1

8.1 REVIEW OBJECTIVE .................................................. 8-1

8.2 ACCEPTANCE TESTS .................................................. 8-1

8.2.1 Areas of Review ................................................... 8-1

8.2.2 Regulatory Requirements ............................................. 8-1

8.2.3 Acceptance Criteria ................................................. 8-2

8.2.4 Review Procedures ................................................. 8-2

8.2.5 Evaluation Findings ................................................. 8-4
xi NUREG-1617

8.3 MAINTENANCE PROGRAM ............................................ 8-5

8.3.1 Areas of Review ................................................... 8-5

8.3.2 Regulatory Requirements ............................................. 8-5

8.3.3 Acceptance Criteria ................................................. 8-5

8.3.4 Review Procedures ................................................. 8-5

8.3.5 Evaluation Findings ................................................. 8-7

8.4 REFERENCES ........................................................ 8-7

APPENDIX A S STANDARD REVIEW PLAN CORRELATION WITH 10 CFR PART 71 AND

REGULATORY GUIDE 7.9 ................................................. A-1

APPENDIX B S TABLE OF EXTERNAL DOSE RATES FOR EXCLUSIVE-USE SHIPMENTS B-1
NUREG-1617 xii

FIGURES

Figure 1-1 SAR Information Flow for the General Information Review ....................... 1-5

Figure 2-1 SAR Information Flow for Structural Review ................................. 2-5

Figure 3-1 SAR Information Flow for the Thermal Review ............................... 3-5

Figure 4-1 SAR Information Flow for the Containment Review. ............................ 4-4

Figure 5-1 SAR Information Flow for the Shielding Review. .............................. 5-4

Figure 6-1 SAR Information Flow for the Criticality Review. .............................. 6-9

Figure 7-1 SAR Information Flow for the Operating Procedures Review. ..................... 7-4

Figure 8-1 SAR Information Flow for the Acceptance Tests Review. ....................... 8-3

Figure 8-2 SAR Information Flow for the Maintenance Program Review. .................... 8-6

TABLES

Table 1-1 Fabrication, Examination, and Testing Criteria for SNF Transportation Packages based on the

B&PV Code ............................................................. 1-7

Table 1-2 Welding Criteria for SNF Transportation Packages ............................. 1-9

Table 4-1 Release Fractions and Specific Activities for the Contributors to the Releasable Source Term

for Packages Designed to Transport Irradiated Fuel Rods ............................. 4-7

Table 5-1 External Radiation Level Limits for Exclusive-Use Shipments ..................... 5-3

Table A-1 Standard Review Plan Correlation with 10 CFR Part 71 and Regulatory Guide 7.9 ...... A-1

Table B-1 External Dose Rates for Packages (Exclusive-Use Shipment) ..................... B-1
xiii NUREG-1617

ACRONYMS AND ABBREVIATIONS

ALARA as low as is reasonably achievable (radiation exposure)

ANS American Nuclear Society

ANSI American National Standards Institute

ASME American Society of Mechanical Engineers

ASTM American Society for Testing and Materials

B&PV Boiler and Pressure Vessel (ASME Code)

Bq Becquerel

BWR boiling-water reactor

EC degrees Celsius

Ci Curie

CFR U.S. Code of Federal Regulations

DOT U.S. Department of Transportation

EF degrees Fahrenheit

g gravitational unit

keff “k” effective-neutron multiplication factor

LSA low specific activity

MIL military

MNOP maximum normal operating pressure

mrem millirem

mSv millisievert (1 mSv = 100 mrem)

NMSS NRC Office or Nuclear Material Safety and Safeguards

NRC U.S. Nuclear Regulatory Commission
NUREG-1617 xiv

PWR pressurized-water reactor

RG regulatory guide (NRC)

SAR safety analysis report

SCO surface contaminated object

SER safety evaluation report

SI International System of Units

SFPO Spent Fuel Project Office (NRC NMSS)

SNF spent nuclear fuel

SRP standard review plan

SSCs structures, systems, and components

Sv Sievert
xv NUREG-1617

GLOSSARY

The following terms are defined here by the staff for the purpose of this SRP. Many of the terms are

taken from 10 CFR 20.1004, 10 CFR 71.4, or 49 CFR 173.403. The definitions from these CFR sections

have not been changed in the list below, but are repeated for convenience. Standards are expressed in

the International System of Units (SI). The U. S. standard or customary unit equivalents presented in

parentheses are for information only. 

A1the maximum activity of special form radioactive material permitted in a

Type A package.

A2the maximum activity of radioactive material, other than special form,

LSA and SCO material, permitted in a Type A package.

As low as is reasonably making every reasonable effort to maintain exposures to radiation as far

achievable (ALARA) below the dose limits in this part as is practical consistent with the

purpose for which the licensed activity is undertaken, taking into account

the state of technology, the economics of improvements in relation to

benefits to the public health and safety, and other societal and

socioeconomic considerations, and in relation to utilization of nuclear

energy and licensed materials in the public interest.

Becquerel (Bq) a unit, in the SI, of measurement of radioactivity equal to one

transformation per second.

Benchmarking validation of the accuracy of a computer code by comparison of obtained

results with those of previously determined experimental values. 

Bias For criticality calculations, ANSI/ANS-8.1 defines bias as a measure of

systematic differences between calculations and experimental data and

subsequently defines uncertainty in the bias. See NUREG/CR-6361 for

further discussion of bias. The determination of bias must adequately

consider the variation in the differences between the calculations and

experimental data.

Carrier a person engaged in the transportation of passengers or property by land

or water as a common, contract, or private carrier, or by civil aircraft. 

Certificate holder a person who has been issued a certificate of compliance or other

package approval by NRC.

Certificate of compliance a certificate issued by NRC which authorizes the use of a specific

packaging, for a specified time, and for a specified scope of activity.

Close reflection by water immediate contact by water of sufficient thickness for maximum

reflection of neutrons.
NUREG-1617 xvi

Closed transport vehicle a transport vehicle or conveyance equipped with a securely attached

exterior enclosure that during normal transportation restricts the access

of unauthorized persons to the cargo space containing the Class 7

(radioactive) materials. The enclosure may be either temporary or

permanent, and in the case of packaged materials may be of the “seethrough” type, and must limit access from the top, sides, and bottom.

Code used generically to refer to national or “consensus” codes, standards, and

specifications, or specifically to refer to the ASME Boiler and Pressure

Vessel Code or may be used to describe computer models.

Confirmatory calculations calculations made by the reviewer to determine whether the package

design and specifications meet the regulations. These calculations do not

replace the design calculations and are intended to assess and confirm

the basis and conclusions of the applicant’s calculations.

Containment system the assembly of components of the packaging intended to retain the

radioactive material during transport. 

Conveyance for transport by public highway or rail, any transport vehicle or large

freight container; for transport by water, any vessel, or any hold,

compartment, or defined deck area of a vessel, including any transport

vehicle on board the vessel; and for transport by aircraft, any aircraft. 

Curie (Ci) the basic unit to describe the intensity of radioactivity in a sample

material. A curie is equal to 37 billion disintegrations per second.

Damaged spent nuclear spent nuclear fuel with known or suspected cladding defects greater than 

fuel a hairline crack or pinhole leak.

Depleted uranium uranium containing less uranium-235 than the naturally occurring

distribution of uranium isotopes.

Docketed formal submissions made to NRC by an applicant, and officially filed by

NRC in the Agency’s records for the application. NRC assigns a docket

number to the transportation package, which is used for the application

and subsequent submissions and other correspondence regarding the

package. Except when NRC concurs in a request that material be

protected as being “proprietary data,” docketed material, in accordance

with 10 CFR 2.790, becomes available for public copying.

Enriched uranium uranium containing more uranium-235 than the naturally occurring

distribution of uranium isotopes.

Exclusive use the sole use by a single consignor of a conveyance for which all initial,

intermediate, and final loading and unloading are carried out in
xvii NUREG-1617

accordance with the direction of the consignor or consignee. The

consignor and the carrier must ensure that any loading or unloading is

performed by personnel having radiological training and resources

appropriate for safe handling of the consignment. The consignor must

issue specific instructions, in writing, for maintenance of exclusive use

shipment controls, and include them with the shipping paper information

provided to the carrier by the consignor.

Fissile material plutonium-238, plutonium-239, plutonium-241, uranium-233, uranium-235,

or any combination of these radionuclides. Unirradiated natural uranium

and depleted uranium, and natural uranium or depleted uranium that has

been irradiated in thermal reactors only, are not included in this definition. 

Certain exclusions from fissile material controls are provided in

10 CFR 71.53. 

Fissile material package a fissile material packaging together with its fissile material contents.

g gravitational unit. (1 g = force exerted on a mass vertically by gravity)

Independent calculation calculations separate from the applicant’s. Input data should be taken from

primary sources such as the package drawings and manufacturer’s

specifications. Models should be developed separately by the reviewer. To the

extent possible, different techniques, codes, and cross section sets or other

derived data sets should be used.

“k” effective the ratio of the number of neutrons resulting from fission in one

generation to the number of neutrons resulting from fission in the

preceding generation.

Low specific activity radioactive material with limited specific activity that satisfies the

(LSA) material descriptions and limits set forth below. Shielding materials surrounding

the LSA material may not be considered in determining the estimated

average specific activity of the package contents. LSA material must be

in one of three groups: 

(1) LSA-I. 

(i) Ores containing only naturally occurring radionuclides (e.g.,

uranium, thorium) and uranium or thorium concentrates of such ores;

or 

(ii) Solid unirradiated natural uranium or depleted uranium or natural

thorium or their solid or liquid compounds or mixtures; or 

(iii) Radioactive material, other than fissile material, for which the A2

value is unlimited; or 

(iv) Mill tailings, contaminated earth, concrete, rubble, other debris,

and activated material in which the radioactive material is essentially

uniformly distributed, and the average specific activity does not

exceed 10-6 A2/g. 
NUREG-1617 xviii

(2) LSA-II. 

(i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/liter);

or 

(ii) Material in which the radioactive material is essentially uniformly

distributed, and the average specific activity does not exceed 10-4

A2/g for solids and gases, and 10-5 A2/g for liquids. 

(3) LSA-III. Solids (e.g., consolidated wastes, activated materials) in 

which: 

(i) The radioactive material is essentially uniformly distributed

throughout a solid or a collection of solid objects, or is essentially

uniformly distributed in a solid compact binding agent (such as

concrete, bitumen, ceramic, etc.); 

(ii) The radioactive material is relatively insoluble, or it is intrinsically

contained in a relatively insoluble material, so that, even under loss of

packaging, the loss of radioactive material per package by leaching,

when placed in water for 7 days, would not exceed 0.1 A2; and 

(iii) The average specific activity of the solid does not exceed

2 × 10-3 A2/g. 

Low toxicity alpha natural uranium, depleted uranium, natural thorium; uranium-235,

emitters uranium-238, thorium-232, thorium-228 or thorium-230 when contained in

ores or physical or chemical concentrates or tailings; or alpha emitters

with a half-life of less than 10 days. 

Maximum normal operating the maximum gauge pressure that would develop in the containment

pressure (MNOP) system in a period of 1 year under the heat condition specified in

10 CFR 71.71(c)(1), in the absence of venting, external cooling by an

ancillary system, or operational controls during transport.

Natural thorium thorium with the naturally occurring distribution of thorium isotopes

(essentially 100 weight percent thorium-232). 

Natural uranium uranium with the naturally occurring distribution of uranium isotopes

(approximately 0.711 weight percent uranium-235, and the remainder by

weight essentially uranium-238).

Normal form radioactive radioactive material that has not been demonstrated to qualify as “special 

material form radioactive material.” 

Optimum interspersed the presence of hydrogenous material between packages to such an

extent

hydrogenous moderation that the maximum nuclear reactivity results. 

Package the packaging together with its radioactive contents as presented for

transport.
xix NUREG-1617

Packaging the assembly of components necessary to ensure compliance with the

packaging requirements of 10 CFR Part 71. It may consist of one or

more receptacles, absorbent materials, spacing structures, thermal

insulation, radiation shielding, and devices for cooling or absorbing

mechanical shocks. The vehicle, tie-down system, and auxiliary

equipment may be designated as part of the packaging. 

Radiation level the radiation dose-equivalent rate expressed in millisievert(s) per hour or

mSv/h (millirem(s) per hour or mrem/h). Neutron flux densities may be

converted into radiation levels according to Table 1, 49 CFR 173.403.

Radioactive contents a Class 7 (radioactive) material, together with any contaminated liquids

or gases within the package.

Radioactive material any material having a specific activity greater than 70 Bq per gram

(0.002 microcurie per gram).

Rem the special unit of any of the quantities expressed as dose equivalent. 

The dose equivalent in rems is equal to the absorbed dose in rads

multiplied by the quality factor (1 rem = 0.01 sievert). 

Rule unless used generically, a requirement stated in the Code of Federal

Regulations.

Sievert (Sv) the SI unit of any of the quantities expressed as dose equivalent. The

dose equivalent in sieverts is equal to the absorbed dose in grays

multiplied by the quality factor (1 Sv = 100 rems).

Safety analysis report in the context of this SRP, the report submitted by the applicant in

(SAR) compliance with 10 CFR Part 71, Subpart D. The fundamental contents

of the report are described in 10 CFR 71.31. Guidance on format and

content of the report is provided by Regulatory Guide 7.9, “Standard

Format and Content of Part 71 Applications for Approval of Packaging

for Radioactive Material.” The SAR is considered to be the submitted

application, along with any supplemental data and responses submitted to

NRC staff to resolve questions arising during the staff's review. Only

docketed material is considered to form part of the submission. The

effective SAR is that submitted, as amplified and/or modified by the

supplemental and later submissions.

Safety evaluation report in the context of this SRP, the report prepared by NRC staff to

(SER) document the acceptability of the applicants SAR and other required

submissions. The SER also identifies NRC staff’s conclusions and the

conditions of approval that will be included in the certificate of

compliance.
NUREG-1617 xx

Special form radioactive radioactive material that satisfies the following conditions:

material (1) It is either a single solid piece or is contained in a sealed capsule that

can be opened only by destroying the capsule; 

(2) The piece or capsule has at least one dimension not less than 5 mm

(0.2 in); and 

(3) It satisfies the requirements of 10 CFR 71.75. A special form

encapsulation designed in accordance with the requirements of

10 CFR 71.4 in effect on June 30, 1983 (see 10 CFR Part 71, revised as

of January 1, 1983), and constructed before July 1, 1985, and a special

form encapsulation designed in accordance with the requirements of

10 CFR 71.4 in effect on March 31, 1996, (see 10 CFR Part 71, revised

as of January 1, 1983), and constructed before April 1, 1998, may

continue to be used. Any other special form encapsulation must meet the

specifications of this definition.

Specific activity the radioactivity of the radionuclide per unit mass of that nuclide.

of a radionuclide The specific activity of a material in which the radionuclide is essentially

uniformly distributed is the radioactivity per unit mass of the material.

Spent nuclear fuel fuel that has been withdrawn from a nuclear reactor following

(SNF) irradiation, the constituent elements of which have not been separated by

reprocessing.

Surface contaminated object a solid object that is not itself classed as radioactive material, but which

(SCO) has radioactive material distributed on any of its surfaces. SCO must be 

in one of two groups with surface activity not exceeding the following

limits:

(1) SCO-I: A solid object on which: 

(i) The non-fixed contamination on the accessible surface averaged

over 300 cm2 (or the area of the surface if less than 300 cm2) does

not exceed 4 Bq/cm2 (10-4 microcurie/cm2) for beta and gamma and

low toxicity alpha emitters, or 0.4 Bq/cm2 (10-5 microcurie/cm2) for

all other alpha emitters; 

(ii) The fixed contamination on the accessible surface averaged over

300 cm2 (or the area of the surface if less than 300 cm2) does not

exceed 4 × 104 Bq/cm2 (1.0 microcurie/cm2) for beta and gamma

and low toxicity alpha emitters, or 4 × 103 Bq/cm2

(0.1 microcurie/cm2) for all other alpha emitters; and 

(iii) The non-fixed contamination plus the fixed contamination on the

inaccessible surface averaged over 300 cm2 (or the area of the

surface if less than 300 cm2) does not exceed 4 × 104 Bq/cm2

(1 microcurie/cm2) for beta and gamma and low toxicity alpha

emitters, or 4 × 103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha

emitters. 

(2) SCO-II: A solid object on which the limits for SCO-I are exceeded

and on which:
xxi NUREG-1617

(i) The non-fixed contamination on the accessible surface averaged

over 300 cm2 (or the area of the surface if less than 300 cm2) does

not exceed 400 Bq/cm2 (10-2 microcurie/cm2) for beta and gamma

and low toxicity alpha emitters or 40 Bq/cm2 (10-3 microcurie/cm2)

for all other alpha emitters;

(ii) The fixed contamination on the accessible surface averaged over

300 cm2 (or the area of the surface if less than 300 cm2) does not

exceed 8 × 105 Bq/cm2 (20 microcuries/cm2) for beta and gamma

and low toxicity alpha emitters, or 8 × 104 Bq/cm2

(2 microcuries/cm2) for all other alpha emitters; and

(iii) The non-fixed contamination plus the fixed contamination on the

inaccessible surface averaged over 300 cm2 (or the area of the

surface if less than 300 cm2) does not exceed 8 × 105 Bq/cm2

(20 microcuries/cm2) for beta and gamma and low toxicity alpha

emitters, or 8 × 104 Bq/cm2 (2 microcuries/cm2) for all other alpha

emitters.

Transport index the dimensionless number (rounded up to the next tenth) placed on the

label of a package, to designate the degree of control to be exercised by

the carrier during transportation. The transport index is determined as

follows:

(1) For non-fissile material packages, the number determined by

multiplying the maximum radiation level in millisievert (mSv) per hour at

one meter (3.3 ft) from the external surface of the package by 100

(equivalent to the maximum radiation level in millirem per hour at one

meter (3.3 ft)); or

(2) For fissile material packages, the number determined by multiplying

the maximum radiation level in millisievert per hour at one meter (3.3 ft)

from the external surface of the package by 100 (equivalent to the

maximum radiation level in millirem per hour at one meter (3.3 ft)), or,

for criticality control purposes, the number obtained as described in

10 CFR 71.59, whichever is larger.

Type A quantity a quantity of radioactive material, the aggregate radioactivity of which

does not exceed A1 for special form radioactive material, or A2, for

normal form radioactive material, where A1 and A2 are given in Table

A-1 of 10 CFR Part 71, or may be determined by procedures described

in Appendix A of 10 CFR Part 71.

Type B package a Type B packaging together with its radioactive contents. On approval,

a Type B package design is designated by NRC as B(U) unless the

package has a maximum normal operating pressure of more than

700 kPa (100 lb/in2) gauge or a pressure relief device that would allow

the release of radioactive material to the environment under the tests

specified in 10 CFR 71.73 (hypothetical accident conditions), in which

case it will receive a designation B(M). B(U) refers to the need for
NUREG-1617 xxii

unilateral approval of international shipments; B(M) refers to the need

for multilateral approval of international shipments. There is no

distinction made in how packages with these designations may be used in

domestic transportation. To determine their distinction for international

transportation, see DOT regulations in 49 CFR Part 173. A Type B

package approved before September 6, 1983, was designated only as

Type B. Limitations on its use are specified in 10 CFR 71.13. 

Type B quantity a quantity of radioactive material greater than a Type A quantity.

U.S. Code of Federal organized by titles (e.g., Title 10, “Energy”), chapters (e.g., Chapter I,

Regulations (CFR) “U.S. Nuclear Regulatory Commission”), parts (e.g., Part 71, “Packaging

and Transportation of Radioactive Material”), subparts (e.g., Subpart D,

“Application for Package Approval”), and sections (e.g., 10 CFR 71.31). 

See also Title 49, “Transportation.”
1 NUREG-1617

INTRODUCTION

The Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, referred to here as the

Standard Review Plan (SRP), provides guidance for the U.S. Nuclear Regulatory Commission (NRC)

safety reviews of packages used in the transport of spent nuclear fuel (SNF) under Title 10 of the U.S.

Code of Federal Regulations (CFR), Part 71 (10 CFR Part 71). It is not intended as an interpretation of

NRC regulations. This SRP supplements NRC Regulatory Guide (RG) 7.9, “Standard Format and

Content of Part 71 Applications for Approval of Packaging for Radioactive Material,” for review of

package applications. Nothing contained in this plan may be construed as having the force and effect of

NRC regulations (except where the regulations are cited), or as indicating that applications supported by

safety analyses and prepared in accordance with RG 7.9 will necessarily be approved, or as relieving any

person from the requirements of 10 CFR Parts 20, 30, 40, 60, 70, or 71 or any other pertinent regulations. 

The principal purpose of the SRP is to ensure the quality and uniformity of staff reviews. It is also the

intent of this plan to make information about regulatory matters widely available and improve

communications between NRC, interested members of the public, and the nuclear industry, thereby

increasing the understanding of NRC staff review process. In particular, this guidance assists potential

applicants by indicating one or more acceptable means of demonstrating compliance with the applicable

regulations.

The SRP is intended for use by NRC staff reviewers of package applications, amendments, and

renewals. The SRP provides specific guidance for the staff's preparation of NRC safety evaluation

report (SER). The SRP provides guidance relating to compliance with 10 CFR Part 71, and portions of

other CFR titles and parts incorporated by reference in or applicable to 10 CFR Part 71.

The SRP is organized to correlate with the recommended content for a safety analysis report (SAR) as

detailed in RG 7.9. The individual sections address the matters that are reviewed, the basis for the

review, how the review is accomplished, the conclusions that are sought, and follow a common outline of

subsections, illustrated below.

Appendix A provides a correlation of the SRP with 10 CFR Part 71 and RG 7.9.

Current packages for shipment of SNF are generally intended to be shipped only on an exclusive-use

vehicle. NRC staff anticipates that future transport of SNF will also be made primarily by exclusive-use

vehicle. Therefore, this SRP addresses only the regulatory requirements and acceptance criteria for

exclusive-use shipment of SNF.

Subsection 1. Review Objective

This subsection states the purpose and scope of the review of the SAR section in question.

Subsection 2. Areas of Review

This subsection provides the general outline used for subsections 3, 4, 5 and 6 (see below). This

subsection identifies the systems, components, analyses, data or other information that are

reviewed as part of the particular SAR section in question.
NUREG-1617 2

Subsection 3. Regulatory Requirements

This subsection summarizes the requirements of 10 CFR Part 71 that relate to the SAR section in

question. The requirements are organized in accordance with the major areas of review identified

in Subsection 2 above.

Subsection 4. Acceptance Criteria

This subsection includes the regulatory requirements by reference and identifies other criteria that

are acceptable practice for demonstrating that the package design meets the regulatory

requirements. The criteria are organized in accordance with the major areas of review identified

in Subsection 2 above.

This subsection typically identifies minimum acceptance criteria that are acceptable to the staff in

dealing with a specific safety or design issue. These acceptance criteria are identified in the SRP

so that staff reviewers can take uniform and well-understood positions as similar safety issues

arise in future cases. Like RGs, these acceptance criteria are acceptable to the staff, but they

are not considered as the only possible means of demonstrating compliance with applicable

regulations.

Subsection 5. Review Procedures

This subsection provides guidance specifically developed for the reviewer in preparation of the

SER. The review is organized in accordance with the areas of review identified in Subsection 2

above. Subsection 5 addresses procedures that the reviewer is to follow to provide verification

that the applicable safety criteria have been met. In addition, it supplements the general

requirement for review of all submitted documentation with guidance based on prior staff reviews,

and NRC experience gained from the regulation of existing transportation packages.

To assist the reviewer, a chart is provided for the SAR section in question depicting the flow of

pertinent information into, within, and from the review process.

Subsection 6. Evaluation Findings

This subsection provides examples of review conclusions appropriate for the SER. The findings

are organized in accordance with the major areas of review identified in Subsection 2 above.

Subsection 7. References

This subsection identifies references used in review of the SAR section in question.

The Director of the Spent Fuel Project Office will direct and approve revisions, including clarifications,

corrections, and modifications, as necessary.
3 NUREG-1617

Suggested revisions and other comments will be considered and should be sent to the Director, Spent Fuel

Project Office, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,

Washington, D.C. 20555-0001.
NUREG-1617 4
1-1 NUREG-1617

1 GENERAL INFORMATION REVIEW

1.1 REVIEW OBJECTIVE

The objective of this review is to establish (1) that the application includes an overview of relevant

package information including intended use; and (2) a summary description of the packaging, operational

features, and contents that provide reasonable assurance that the package can meet the regulations and

operating objectives.

1.2 AREAS OF REVIEW

The SAR should be reviewed for adequacy of the package description and drawings of the packaging. 

Areas of review include the following:

1.2.1 General SAR Format

1.2.2 Package Design Information

1.2.2.1 Purpose of Application

1.2.2.2 Quality Assurance Program

1.2.2.3 Proposed Use/General Contents

1.2.2.4 Package Type and Model Number

1.2.2.5 Package Category and Maximum Activity

1.2.2.6 Materials Specifications, Fabrication, and Welding Criteria

1.2.2.7 Transport Index and Maximum Number of Packages

1.2.3 Package Description

1.2.3.1 Packaging

1.2.3.2 Operational Features

1.2.3.3 Contents of Packaging

1.2.4 Compliance with 10 CFR Part 71

1.2.4.1 General Requirements of 10 CFR 71.43

1.2.4.2 Condition of Package after Tests in 10 CFR 71.71 and 10 CFR 71.73

1.2.4.3 Structural, Thermal, Containment, Shielding, Criticality

1.2.4.4 Operational Procedures, Acceptance Tests and Maintenance

1.2.5 Appendix

1.3 REGULATORY REQUIREMENTS

Regulatory requirements of 10 CFR Part 71 applicable to the general information review are as follows:

1.3.1 General SAR Format

There are no specific regulatory requirements on the format of the SAR. SAR format provisions are

given in RG 7.9.
NUREG-1617 1-2

1.3.2 Package Design Information

The application for package approval must: (1) include the classification and model number

[10 CFR 71.31(a)(1), 10 CFR 71.33(a)(1), and 10 CFR 71.33(a)(3)]; (2) include a quality assurance

program description or a reference to a previously approved quality assurance program applicable to the

package [10 CFR 71.31(a)(3) and 10 CFR 71.37]; (3) identify applicable codes and standards used in

package design, fabrication, assembly, testing, maintenance, and use [10 CFR 71.31(c)]; and (4) include

the transport index for nuclear criticality control. [10 CFR 71.31(a)(2), 10 CFR 71.35(b), and

10 CFR 71.59]

An application for renewal of a previously approved package design must be submitted to NRC no later

than 30 days prior to the expiration date of the approval to assure continued use and is subject to the

provisions of 10 CFR 71.13. [10 CFR 71.38]

All changes in the conditions specified in the package approval must be approved by NRC. An

application for modification of a previously approved package design may be subject to the provisions of

10 CFR 71.13(c) and 10 CFR 71.31(b). [10 CFR 71.107(c)]

1.3.3 Package Description

The description of the packaging must include a containment system, materials of construction, weights,

dimensions, methods of fabrication, and coolant receptacle volumes in sufficient detail to provide an

adequate basis for their evaluation. [10 CFR 71.31(a)(1), 10 CFR 71.33(a)(2), 10 CFR 71.33(a)(4),

10 CFR 71.33(a)(5), and 10 CFR 71.33(a)(6)]

The SAR must identify, with respect to the contents of the package, the maximum radioactive and fissile

constituents, physical and chemical form, neutron absorbers or moderators, extent of reflection,

moderator-to-fissile ratio, maximum normal operating pressure, maximum weight, maximum decay heat,

and any coolant volumes. [10 CFR 71.31(a)(1) and 10 CFR 71.33(b)]

The outside of the package must incorporate a feature that, while intact, would be evidence that the

package has not been opened by unauthorized persons. [10 CFR 71.43(b)]

Spent fuel, with plutonium in excess of 0.74 TBq (20 Ci) per package, in the form of debris, particles,

loose pellets, or fragmented rods or assemblies must be packaged in a separate inner container (second

containment system) in accordance with 10 CFR 71.63(b). [10 CFR 71.63]

1.3.4 Compliance with 10 CFR Part 71

The package must be evaluated to demonstrate compliance with the requirements specified in 10 CFR

Part 71, Subpart E, under the conditions and tests of Subpart F. [10 CFR 71.31(a)(2), 10 CFR 71.35(a),

and 10 CFR 71.41(a)]
1-3 NUREG-1617

1.4 ACCEPTANCE CRITERIA

1.4.1 General SAR Format

The application should be prepared in accordance with the general format provisions of RG 7.9.

1.4.2 Package Design Information

The regulatory requirements in Section 1.3.2 identify the acceptance criteria.

1.4.3 Package Description

In addition to the regulatory requirements identified in Section 1.3.3, a discussion of the operation of the

package should be provided. [RG 7.9 (1.2.2)]

In addition to the regulatory requirements identified in Section 1.3.3, spent nuclear fuel with known or

suspected cladding defects greater than a hairline crack or a pinhole leak should be canned. Canning of

damaged fuel is to facilitate handling and to confine gross fuel particles to a known subcritical volume

under normal conditions of transport and hypothetical accident conditions.

In addition to the regulatory requirements identified in Section 1.3.3, engineering drawings of the package

should be provided. [RG 7.9 (1.3)]

1.4.4 Compliance with 10 CFR Part 71

In addition to the regulatory requirements identified in Section 1.3.4, a concise statement by the applicant,

that the package complies with the requirements in 10 CFR Part 71 for a Type B(U)F package, should be

provided in the General Information section of the SAR. This summary statement should provide a

reference to the sections of the SAR that are used to specifically address compliance with the

requirements of Subparts E and F of 10 CFR Part 71.

1.5 REVIEW PROCEDURES

The review should ensure that the General Information section of the SAR provides an adequate

description of the spent nuclear fuel (SNF) transportation package so that its design and operation can be

evaluated in subsequent sections. Although the General Information section of the SAR will not contain

enough information by itself to perform a technical review of the package, the General Information

section serves as a vehicle to facilitate consistency and reduce repetition between the various review

disciplines (e.g., structural and shielding reviews), and presents summary information for the non-technical

reviewers. The following procedures are generally applicable to the general information review of all

SNF transportation packages.

Packages for shipment of SNF are generally intended to be shipped only on an exclusive-use vehicle. 

NRC staff anticipates that future transport of SNF will also be made primarily by exclusive-use vehicle. 
NUREG-1617 1-4

Therefore, this SRP addresses only the regulatory requirements and acceptance criteria for exclusive-use

shipment of SNF.

The general information review is based in part on the descriptions and evaluations presented in the

General Information section of the SAR and follows the sequence established to evaluate the packaging

against applicable 10 CFR Part 71 requirements. Similarly, results of the general information review are

considered in the review of the SAR sections on Structural Evaluation, Thermal Evaluation, Containment

Evaluation, Shielding Evaluation, Criticality Evaluation, Operating Procedures, and Acceptance Tests and

Maintenance Program. Examples of SAR information flow within and from the general information

review are shown in Figure 1-1.

1.5.1 General SAR Format

Verify that the SAR has been prepared in accordance with the use of standard format, style and

composition, revisions, and physical specifications described in RG 7.9 (i.e., paper size, paper stock, ink,

page margins, printing, binding, page numbering, separators, and number of copies).

1.5.2 Package Design Information

1.5.2.1 Purpose of Application

The purpose of the application should be clearly stated. The application may be for approval of a new

design, for amendment, or for renewal of an existing approval (i.e., certificate of compliance). 

Applications for approval of a new design should be whole and complete and should contain the

information identified in Subpart D of 10 CFR Part 71. If the application is for modification of an

approved design, verify that the changes being requested are clearly identified. Modifications may include

design changes, changes in authorized contents, or changes in conditions of the approval. Design changes

should be clearly identified and should be included in revised packaging drawings. Packagings that do not

conform to the drawings referenced in the NRC approval are not authorized for use under the general

license in 10 CFR 71.12. Likewise, only contents specified in the approval may be transported. Package

operating procedures, acceptance tests, and a maintenance program may also be specified as conditions

of the approval.

Applications for modifications to an approved design should include an assessment of the requested

changes and justification that these changes do not affect the ability of the package to meet the

requirements of 10 CFR Part 71. Applications for modifications may be subject to the provisions of

10 CFR 71.13(c) and 10 CFR 71.31(b), as applicable. When the modification is submitted under the

provision of 10 CFR 71.13(c)(1) or 10 CFR 71.13(c)(2), the application should justify that the requested

change is not significant.

Applications for renewal of an existing approval should be made not less than 30 days before the

expiration of the approval to assure continued use. Applications for renewal are subject to the provisions

of 10 CFR 71.38.

1.5.2.2 Quality Assurance Program
1-5 NUREG-1617

Verify that the applicant has obtained NRC approval of its quality assurance program, or has identified by

reference a quality assurance program that has been previously approved under the requirements of

10 CFR 71.12, 10 CFR 71.37 and 10 CFR Part 71, Subpart H.
NUREG-1617 1-6

Figure 1-1 SAR Information Flow for the General Information Review
1-7 NUREG-1617

1.5.2.3 Proposed Use/General Contents

Verify that the description for the proposed use of the packaging and the general contents of the package

are sufficient to allow the reviewer to understand exactly how the packaging is to be used and what is to

be transported. Since packages for shipment of SNF are generally intended to be shipped by exclusive

use, only exclusive-use shipments are assumed in the following SRP review procedures. Verify that the

package is to be shipped by exclusive use and ensure that any restrictions regarding the use or type of

conveyance are designated.

1.5.2.4 Package Type and Model Number

Confirm that the type and model number of the package are designated. A new SNF transportation

package will be designated B(U)F-85 unless it has a maximum normal operating pressure (MNOP)

greater than 700 kPa (100 lb/in2) or a pressure relief device that would allow the release of radioactive

material under the tests specified in 10 CFR 71.73 (hypothetical accident conditions). In those cases, the

package will be designated B(M)-85. Verify that a model number is designated for the package and that

it is specified on the appropriate drawings.

1.5.2.5 Package Category and Maximum Activity

Category I is assigned to a package whose content activity exceeds either 1.11 × 1015 Bq (30,000 Ci),

3000 A1, or 3000 A2 whichever is less. (SNF transportation packages are assumed to be Category I in

the following SRP review procedures.) Verify that the package is designated Category I and that the

maximum activity of the package contents is specified.

1.5.2.6 Materials Specifications, Fabrication, and Welding Criteria

ASME has published Section III, Division 3, ASME Boiler and Pressure Vessel (B&PV Division 3) Code

for the design and construction of the containment system of SNF transport packagings. NRC staff

expects full compliance with the B&PV Division 3 Code for the containment system, including the

services of an Authorized Inspection Agency. However, the SAR may justify alternatives as appropriate. 

The code used for the design should also be used for materials specifications, fabrication, and welding

criteria.

Criteria acceptable for other components of SNF transportation packages are also based on the ASME

Boiler and Pressure Vessel (B&PV) Code. Table 1-1 summarizes the appropriate B&PV Code sections

for the fabrication, examination, and testing of the Containment, Criticality, and Other Safety component

groups. Table 1-2 summarizes the appropriate B&PV Code sections for welding of the key elements of

the Containment, Criticality, and Other Safety component groups.

Verify that the fabrication, welding, and examination criteria for the package are specified for each major

component and that they are appropriate for a SNF package. Verify that materials specifications and

standards have been specified for all major components and that they are consistent for the product form

to be fabricated. Verify that a reference is provided to the sections of the SAR where a discussion of

any fabrication, welding, and examination of package components may be found.
NUREG-1617 1-8

Table 1-1 Fabrication, Examination, and Testing Criteria for SNF Transportation Packages based on the B&PV Code.1

Container

contents

Component safety group

Containment Criticality Other Safety

Primary vessel,

bolts, piping,

fittings, valves,

closure

Primary

seal

Support

structures/

neutron

absorber2

Gamma

shielding3

Structural

shell, bolts,

 and

closure

Secondary

seal

Neutron shielding,

piping, fittings,

valves, relief

 device, and tanks4

Lifting

lugs

Impact

limiters5

Tie

down

devices

Heat

transfer

devices

B&PV Code

section

Sec. III, Division 3 Sec. III,

Subsection NG

Sec. VIII, Div. 16

or Sec. III, Subsection NF

Materials 7 WB-2000

8

NG-2000

8

Forming, fitting

 and aligning

WB-42003 NG-4200

Heat treatment WB-4600 NG-4600

Examination WB-5000 NG-5000

Acceptance testing WB-60009

9 10 11

1 These criteria should be referenced in the associated SAR. Criteria for special processes used, but not included in this table, should be documented in the

SAR. Fabrication criteria for welding and brazing are recommended in Table 1-2. Quality assurance criteria are provided in 10 CFR Part 71 and RG 7.10. 

Referenced supporting portions of Section II; Section III, Subsection NCA; Section V: and Section IX of the B&PV Code are part of the recommended

criteria.

2 The designer may specify a neutron absorber material by a commercial trade name or as a mixture of elements or common compounds. When appropriate,

qualification data should be included in the SAR to demonstrate that the material functions as specified. When special absorber materials are used to

control criticality, an acceptance test should be performed for each container to ensure that the absorber material has been properly installed. Structural

criteria do not apply to neutron absorbers unless they are used for structural support.

3 The installation of shielding may involve processes such as lead pouring around the primary vessel or shrink fitting of uranium castings onto the primary

vessel which could affect the vessels structural integrity. In such cases, the fabrication criteria for the specific process and an engineering evaluation of any

associated structural effects should be included in the SAR. Acceptance testing of the gamma shield should be performed to ensure its effectiveness.
1-9 NUREG-1617

4 The designer may specify a neutron shielding material by a commercial trade name or as a mixture of elements or common compounds. When appropriate,

qualification data should be included in the SAR to demonstrate that the material functions as specified. Acceptance testing may be required to

demonstrate the effectiveness of the neutron shielding.

5

Impact limiters may use special materials such as wood or honeycomb metals to provide the specified crushing characteristics. Any special processes,

physical properties, or other information needed to install the impact limiter or qualify its proper function should be included in the SAR.

6 Specific articles in Section VIII, Division 1 have not been listed since the fabrication process is dependent on the fabrication method and materials used. 

Once the method of fabrication and materials of construction have been specified, the appropriate fabrication criteria can be found in Subsection A,

General Requirements; Subsection B, Methods of Fabrication; and Subsection C, Classes of Materials. Criteria from equivalent ASTM materials and

standards, DOT specifications or articles in B&PV Code Section III, Subsection NF may also be substituted, all or in part.

7 The B&PV Code was written for pressure vessel and nuclear component fabrication and does not include many of the materials used in the shipping

container industry. The designer may specify the material to be used by either a commercial trade name or an applicable ASTM specification. For each

material used, information or references should be included in the SAR to permit an evaluation of the materials properties and the intended use. For

thicknesses up to four inches, ferritic materials should satisfy the fracture toughness criteria recommended in RG 7.11 for the relevant container category

instead of the fracture toughness criteria specified in the B&PV Code. Fracture toughness criteria for ferritic steel thicknesses greater than four inches are in

RG 7.12.

8 The B&PV Code does not have specifications for either gasket or seal materials. The designer may specify the material and configuration by a commercial

trade name. Information which demonstrates the qualification of the seal or gasket (including those used for valves and relief devices) should be included

in the SAR.

9 Leak testing of the primary containment, including seals, should be performed in accordance with ANSI N14.5.

10 Shipping containers involved in critical lifts in nuclear facilities should have their lifting lugs fabricated and tested to the criteria specified in NUREG0612 and ANSI N14.6.

11 Heat transfer devices required to contain pressure should be hydrostatically tested to Section VIII, Subsection UG-99. Acceptance testing for each shipping

container may be necessary to ensure that the specified heat transfer rate is obtained.
NUREG-1617 1-10

Table 1-2 Welding Criteria for SNF Transportation Packages.

Key welding elements

Weld type

Containment-related Criticality-related Other safety-related

B&PV Code section Sec. III, Division 3 Sec. III, Subsection NG Sec. VIII, Div. 1

 or Sec. III, Subsection NF

 Base materials WB-2000, WB-4100,

 and applicable Code cases

NG-2100, NG-2200,

NG-2500, NG-4100,

 and applicable Code cases

Sec. VIII, Div. 1, Subsection

A, General Requirements;

appropriate parts of

Subsection B, Methods of

Fabrication; and Subsection

C, Classes of Materials.

Or

Sec. III, Subsection NF

 Welding and brazing

 materials

WB-2400 NG-2400

 Joint preparation WB-4200 NG-4200

 Welding WB-4400 NG-4400

 Brazing WB-4500 NG-4500

 Heat treatment WB-4600 NG-4600

 Qualification of

 procedures and 

 personnel

WB-4300 NG-4300

 Examination WB-5000 NG-5000

 Quality assurance 10 CFR Part 71, Subpart H and RG 7.10

 Fracture toughness RG 7.11 or 7.12 As justified in the SAR

1.5.2.7 Transport Index and Maximum Number of Packages

Verify that a transport index has been assigned to the packaging for the SNF contents and that a

reference is provided to the section of the SAR where a discussion of the determination of the transport

index is found.

Verify that the maximum number of SNF packages in one shipment has been assigned to the packaging

for the specified fissile contents and that a reference is provided to the section of the SAR where the

determination of the maximum number of packages is found.

1.5.3 Package Description

1.5.3.1 Packaging

Review the text description of the packaging and verify that the following information, as applicable, is

discussed. Sketches, figures, or other schematic diagrams should be used as appropriate.

• The gross weight, external dimensions, and cavity size
1-11 NUREG-1617

• Materials of construction, weights, dimensions, and fabrication methods of receptacles, neutron

absorbers or moderators, internal and external structures supporting or protecting receptacles, fuel

basket and engineered flux traps, valves, sampling ports, lifting and tie-down devices, impact limiters,

structural and mechanical means of heat dissipation, types of coolant, outer and inner protrusions,

shielding, pressure relief systems, and closures

• Identification of the containment system and boundary (see Section 4.5.1.3 for additional guidance on

containment of damaged fuel).

Examine the detailed drawings presented in the appendix. Verify that information shown on the drawings

is consistent with that discussed in the text. Drawings should be sufficiently detailed to provide a package

description that can be evaluated for compliance with 10 CFR Part 71. The packaging drawings are

incorporated by reference into the certificate of compliance and become regulatory conditions for

compliance.

Confirm that each drawing has a title block that identifies the preparing organization, drawing number,

sheet number, title, date, and signature or initials indicating approval of the drawing. Revised drawings

should identify the revision number, date, and description of the change in each revision. Proprietary

information should be clearly identified. The drawings should include:

• General arrangement of the packaging and contents, including dimensions

• Design features which affect the package evaluation

• Package markings

• Maximum allowable weight of the package

• Maximum weight of contents and secondary packaging

• Minimum weights, if appropriate.

Information on design features should include, as appropriate:

• Identification of the design feature and its components

• Materials of construction, including appropriate material specifications

• Codes, standards, or other similar specification documents for fabrication, assembly, and testing

• Dimensions with appropriate tolerances

• Operational specifications (e.g., bolt torque)

• Tamper indicating device.
NUREG-1617 1-12

Additional guidance on engineering drawings submitted in the SAR is provided in NUREG/CR-5502.
1-13 NUREG-1617

1.5.3.2 Operational Features

For complex packages, verify that all operational features and functions are discussed. A schematic

diagram should be included in the SAR showing all valves, connections, piping, openings, seals, and

containment boundaries. Detailed operational schematics should be provided and annotated in

accordance with the operations described in the Operating Procedures section of the SAR. However,

details may be referenced in the General Information section of the SAR, if provided in a later SAR

section or appendix. In this case, simplified operational schematics should be an acceptable alternative in

the General Information section of the SAR. Loading configurations for all contents should be provided

and annotated in a manner consistent with the Structural Evaluation, containment Evaluation, Thermal

Evaluation, Shielding Evaluation, Criticality Evaluation, and Operating Procedures sections of the SAR.

Confirm that a reference is provided to any other section of the SAR where evaluations of the operability

and safety of the operational features are found.

Any codes and standards proposed for regulating the operation of the package should be identified and a

reference provided to any other section of the SAR where a discussion of the proposed codes and

standards is found. Confirm that a reference is provided.

1.5.3.3 Contents

The contents should be described in the same detail as that intended for the certificate of compliance. 

Review the description of the contents and verify that, as a minimum, the following information is

presented:

• The type of SNF and maximum initial U-235 mass, its associated burnup, specific power, cooling time,

heat load, and maximum and minimum initial enrichment, including a description of non-uniform

enrichment, if applicable 

• Fuel assembly specifications, including dimensional data for the fuel rods and assembly structure

• Control assemblies or other contents (e.g., startup sources) that may be present

• Maximum quantities of radionuclides present in the SNF and the quantities estimated to be available

for immediate release within the void space of the fuel rods

• Maximum quantity of unirradiated fuel and maximum initial U-235 mass per assembly or rods and

number of assemblies or rods

• Chemical and physical form, presence of any annular pellets

• Location and configuration within the packaging

• Any material subject to chemical, galvanic, or other reaction, including the generation of combustible

gases
NUREG-1617 1-14

• Fuel densities

• Amounts of neutron absorbers or moderators in the fuel or package

• Basket or other configurations of fuel assemblies or rods

• MNOP

• Maximum weight

• Free volume of the containment vessel

• Containment fill gas

• Any unique or unusual conditions (e.g., failed fuel, non-uniform enrichment, etc.)

• For damaged fuel, the maximum quantity of damaged fuel, initial enrichment, absorption, extent of

damage, and description of the second containment system, and any other limits, as applicable are

specified (see also Section 4.5.1.3).

1.5.4 Compliance with 10 CFR Part 71

Review the summary results to determine if the packaging complies with regulations.

1.5.4.1 General Requirements of 10 CFR 71.43

Verify that a summary statement is provided indicating compliance with the general standards for all

packages and that references are provided to the sections of the SAR where discussions of compliance

with the general standards for all packages are found.

1.5.4.2 Condition of Package after Tests in 10 CFR 71.71 and 10 CFR 71.73

Verify that summary descriptions are provided for the physical condition of the package subsequent to the

tests specified in 10 CFR 71.71 (normal conditions of transport) and 10 CFR 71.73 (hypothetical accident

conditions). Verify that references are provided to all sections of the SAR where discussions of the

physical conditions of the package subsequent to testing are found.

1.5.4.3 Structural, Thermal, Containment, Shielding, Criticality

Verify that summary statements are provided attesting to the adequacy of the package design to meet the

structural, thermal, containment, shielding, and criticality requirements of 10 CFR Part 71.

1.5.4.4 Operational Procedures, Acceptance Tests and Maintenance
1-15 NUREG-1617

Verify that a summary statement is provided attesting to the adequacy of the development of the

operational procedures and acceptance tests and maintenance program to ensure compliance with the

requirements of 10 CFR Part 71.
NUREG-1617 1-16

1.5.5 Appendix

In addition to the packaging drawings discussed above, the appendix may include a list of references and

copies of any applicable references not generally available to the reviewer. The appendix may also

provide supporting details on special fabrication procedures and other appropriate supplemental

information.

1.6 EVALUATION FINDINGS

The Safety Evaluation Report (SER) does not normally include specific findings for the General

Information section of the SAR.

Before proceeding with the review of the other sections of the SAR, the reviewer should conclude, at a

minimum, that the following criteria have been demonstrated:

• The package has been described in sufficient detail to provide an adequate basis for its evaluation.

• Drawings provided contain information which provides an adequate basis for its evaluation against

10 CFR Part 71 requirements. Each drawing is identified, consistent with the text of the SAR, and

contains keys or annotation to explain and clarify information on the drawing.

• The application for package approval includes a reference to the approved quality assurance program

for the package.

• The application for package approval identifies applicable codes and standards for the package

design, fabrication, assembly, testing, maintenance, and use.

• The package meets the general requirements of 10 CFR 71.43(a) and 10 CFR 71.43(b).

• Drawings submitted with the application provide a detailed packaging description that can be

evaluated for compliance with 10 CFR Part 71 for each of the technical disciplines.

• Any restrictions on the use of the package are specified.

• Any modifications to a previously approved package do not violate the restrictions in

10 CFR 71.13(c).

1.7 REFERENCES

ANSI N14.5 Institute for Nuclear Materials Management, ANSI N14.5, “Leakage Tests

on Packages for Shipment of Radioactive Materials,” New York, NY, 1987.

ANSI N14.6 Institute for Nuclear Materials Management, ANSI N14.6, “Special Lifting

Devices for Shipping Containers Weighing 10,000 Pounds (45000 kg) or

More for Nuclear Materials,” New York, NY, 1993.
1-17 NUREG-1617

B&PV Code American Society of Mechanical Engineers, “ASME Boiler and Pressure

Vessel Code,” New York, NY, 1998.

B&PV Division 3 Code American Society of Mechanical Engineers, “ASME Boiler and Pressure

Vessel Code, Section III, Division 3, Containment Systems and Transport

Packagings For Spent Nuclear Fuel and High Level Radioactive Waste,”

New York, NY, 1998.

NUREG-0612 U.S. Nuclear Regulatory Commission, “Control of Heavy Loads at Nuclear

Power Plants,” NUREG-0612, National Technical Information Service,

Springfield, VA, July 1980.

NUREG/CR-5502 U.S. Nuclear Regulatory Commission, “Engineering Drawings for 10 CFR

Part 71 Package Approvals,” NUREG/CR-5502, U.S. Government Printing

Office, Washington, D.C., May 1998.

RG 7.9 U.S. Nuclear Regulatory Commission, Regulatory Guide 7.9, “Standard

Format and Content of 71 Applications for Approval of Packaging of type B,

Large Quantity and Fissile Radioactive Material,” U.S. Government Printing

Office, Washington, D.C., July 1979.

RG 7.10 U.S. Nuclear Regulatory Commission, Regulatory Guide 7.10, “Establishing

Quality Assurance Programs for Packaging Used in the Transport of

Radioactive Material,” U.S. Government Printing Office, Washington, D.C.,

January 1983.

RG 7.11 U.S. Nuclear Regulatory Commission, Regulatory Guide 7.11, “Fracture

Toughness Criteria of Base Material for Ferritic Steel Shipping Cask

Containment Vessels with a Maximum Wall Thickness of 4 Inches (0.1 m),”

U.S. Government Printing Office, Washington, D.C., June 1991.

RG 7.12 U.S. Nuclear Regulatory Commission, Regulatory Guide 7.12, “Fracture

Toughness Criteria of Base Material for Ferritic Steel Shipping Cask

Containment Vessels with a Wall Thickness Greater than 4 Inches (0.1 m),”

U.S. Government Printing Office, Washington, D.C., June 1991. 
NUREG-1617 1-18
2-1 NUREG-1617

2 STRUCTURAL REVIEW

2.1 REVIEW OBJECTIVE

The objective of this review is to verify that the structural performance of the package has been

adequately evaluated for the tests specified under normal conditions of transport and hypothetical accident

conditions and that the package design has adequate structural integrity to meet the requirements of

10 CFR Part 71.

2.2 AREAS OF REVIEW

The SAR should be reviewed for adequacy of the description and evaluation of the structural design. 

Areas of review include the following:

2.2.1 Description of Structural Design

2.2.1.1 Descriptive Information Including Weights and Centers of Gravity

2.2.1.2 Codes and Standards

2.2.2 Material Properties

2.2.2.1 Materials and Material Specifications

2.2.2.2 Chemical, Galvanic, or Other Reactions

2.2.2.3 Effects of Radiation on Materials

2.2.3 Lifting and Tie-down Standards for All Packages

2.2.3.1 Lifting Devices

2.2.3.2 Tie-down Devices

2.2.4 General Considerations for Structural Evaluation of Packaging

2.2.4.1 Evaluation by Analysis 

2.2.4.2 Evaluation by Test

2.2.5 Normal Conditions of Transport

2.2.5.1 Heat

2.2.5.2 Cold

2.2.5.3 Reduced External Pressure

2.2.5.4 Increased External Pressure

2.2.5.5 Vibration

2.2.5.6 Water Spray

2.2.5.7 Free Drop

2.2.5.8 Corner Drop

2.2.5.9 Compression

2.2.5.10 Penetration

2.2.6 Hypothetical Accident Conditions

2.2.6.1 Free drop
NUREG-1617 2-2

2.2.6.2 Crush

2.2.6.3 Puncture

2.2.6.4 Thermal

2.2.6.5 Immersion – Fissile Material

2.2.6.6 Immersion – All Packages

2.2.7 Special Requirement for Irradiated Nuclear Fuel Shipments

2.2.8 Internal Pressure Test

2.2.9 Appendix

2.3 REGULATORY REQUIREMENTS

Regulatory requirements of 10 CFR Part 71 applicable to the structural review are as follows:

2.3.1 Description of Structural Design

The packaging must be described in sufficient detail to provide an adequate basis for its evaluation.

[10 CFR 71.31(a)(1) and 10 CFR 71.33]

The SAR must identify established codes and standards applicable to the structural design and fabrication

of the package. [10 CFR 71.31(c)]

2.3.2 Material Properties

The package must be made of materials which assure that there will be no significant chemical, galvanic,

or other reactions among the packaging components, among package contents, or between the packaging

components and the package contents, including possible reaction resulting from inleakage of water. The

effects of radiation on the materials of construction must also be considered. [10 CFR 71.43(d)]

2.3.3 Lifting and Tie-down Standards for All Packages

The package design must meet the lifting and tie-down requirements of 10 CFR 71.45.

2.3.4 General Considerations for Structural Evaluation of Packaging

The package must be evaluated to demonstrate that it satisfies the standards specified in 10 CFR Part 71,

Subpart E, under the conditions and tests of Subpart F. [10 CFR 71.31(a)(2), 10 CFR 71.35(a), and

10 CFR 71.41(a)]

The effects on the performance of the package under the tests specified in 10 CFR 71.71 (normal

conditions of transport), in 10 CFR 71.73 (hypothetical accident conditions), and in 10 CFR 71.61 (special

requirement for irradiated nuclear fuel shipments) must be evaluated. [10 CFR 71.41(a)] by subjecting a

specimen or scale model to a specific test, or by another appropriate and acceptable method.
2-3 NUREG-1617

2.3.5 Normal Conditions of Transport

The package must be evaluated under the tests specified in 10 CFR 71.71 for normal conditions of

transport. [10 CFR 71.41(a)] The evaluation must show that under the tests, there would be no

substantial reduction in the effectiveness of the packaging. [10 CFR 71.35(a), 10 CFR 71.43(f),

10 CFR 71.51(a)(1), and 10 CFR 71.55(d)(4)]

2.3.6 Hypothetical Accident Conditions

The package must be evaluated under the tests specified in 10 CFR 71.73 for hypothetical accident

conditions. [10 CFR 71.41(a)] The evaluation must show that the packaging has adequate structural

integrity to satisfy the containment, shielding, subcriticality, and temperature requirements of 10 CFR

Part 71, Subpart E. [10 CFR 71.35(a)]

2.3.7 Special Requirement for Irradiated Nuclear Fuel Shipments

The containment vessel of a package with activity greater than 37 PBq (106 Ci) must be designed to

withstand an external pressure of 2 MPa (290 psi) for a period of not less than one hour without collapse,

buckling, or inleakge of water. [10 CFR 71.61] 

2.3.8 Internal Pressure Test

Where the maximum normal operating pressure (MNOP) will exceed 35 kPa (5 psi) gauge, the

containment design of all packages must be tested at an internal pressure at least 50 percent higher than

the MNOP to verify that the system can maintain structural integrity at that pressure. [10 CFR 71.85(b)] 

2.4 ACCEPTANCE CRITERIA

2.4.1 Description of Structural Design

In addition to the regulatory requirements identified in Section 2.3.1, the containment system should be

designed and constructed in accordance with Section III, Division 3, ASME Boiler and Pressure Vessel

(B&PV Division 3) Code. Alternate codes should be justified in the SAR.

In addition to the regulatory requirements identified in Section 2.3.1, packaging components other than the

containment vessel should be designed and constructed in accordance with the criteria identified in

Section 1.5.2.6 and Section 2.5.1.2.

In addition to the regulatory requirements identified in Section 2.3.1, load combinations for the packaging

design should be in accordance with RG 7.8.

2.4.2 Material Properties

In addition to the regulatory requirements identified in Section 2.3.2, the structural design should address

precluding brittle fracture in containments made of ferritic steels as described in RGs 7.11 and 7.12. 
NUREG-1617 2-4

Material properties should meet the material specifications applicable to the codes and standards used for

the design and fabrication of the package.

2.4.3 Lifting and Tie-Down Standards for All Packages

Section 2.3.3 discusses the acceptance criteria for lifting and tie-down standards. 

2.4.4 General Considerations for Structural Evaluation of Packaging

Section 2.3.4 discusses the structural design criteria.

2.4.5 Normal Conditions of Transport

Section 2.3.5 discusses the acceptance criteria for package testing with respect to normal transport

conditions.

2.4.6 Hypothetical Accident Conditions

Section 2.3.6 discusses the acceptance criteria for package testing with respect to hypothetical accident

conditions..

2.4.7 Special Requirement for Irradiated Nuclear Fuel Shipments

Section 2.3.7 discusses the acceptance criteria for package testing with activity greater than 37 PBq.

2.4.8 Internal Pressure Test

Section 2.3.8 discusses the acceptance criteria for internal pressure testing.

2.5 REVIEW PROCEDURES

The following procedures are generally applicable to the structural review of all spent nuclear fuel (SNF)

transportation packages.

The structural review is based in part on the descriptions and evaluations presented in the General

Information and Thermal Evaluation sections of the SAR. Similarly, results of the structural review are

considered in the review of the SAR sections on Thermal Evaluation, Containment Evaluation, Shielding

Evaluation, Criticality Evaluation, Operating Procedures, and Acceptance Tests and Maintenance

Program. Examples of SAR information flow into, within, and from the structural review are shown in

Figure 2-1.

2.5.1 Description of Structural Design

2.5.1.1 Descriptive Information Including Weights and Centers of Gravity
2-5 NUREG-1617

Review drawings and other descriptions of the structural design in the General Information and Structure

Evaluation sections of the SAR. The information should describe the function, geometry, and material of

construction of all structural components of the packaging and its lifting and tie-down devices. The

information should be sufficient for evaluating the structural performance of the packaging to meet the

regulatory requirements, which include containment, shielding, and maintaining subcriticality of the 
NUREG-1617 2-6
2-7 NUREG-1617

Figure 2-1 SAR Information Flow for Structural Review

radioactive contents under the normal conditions of transport and the hypothetical accident conditions. 

Verify that the data used in the structural evaluation are consistent with those on the drawings and

descriptions of the structural design in the SAR.

Verify that packaging drawings provided in General Information section of the SAR specify the materials

of construction, dimensions, tolerances and fabrication methods of the packaging and subassemblies,

receptacles, internal or external support structures, valves and ports, lifting devices, and tie-down devices,

and other design features relevant to the structural evaluation. Descriptive information such as the

maximum and minimum weight of the package, the maximum weight of the contents, the center of gravity

of the package, and the MNOP should be included.

2.5.1.2 Codes and Standards

The SAR should identify established codes and standards or justify the basis used for the package design

and fabrication. The codes and standard must be appropriate for the intended purpose, and must be

properly applied. The reviewer should verify that the code or standard:

• Was developed for structures of similar design and material, if not specifically for shipping packages

• Was developed for structures with similar loading conditions

• Was developed for structures which have similar consequences of failure

• Adequately addresses potential failure modes

• Adequately addresses margins of safety.

The ASME has developed a code specifically for the design and construction of the containment systems

of an SNF or high-level radioactive waste transport packaging (B&PV Division 3 Code). NRC will

accept the material, design, fabrication, welding, examination, testing, inspection, and certification of

containment systems for SNF transportation packagings in accordance with the B&PV Division 3 Code. 

If there are any deviations in any way from the B&PV Division 3 Code, the SAR should explicitly state

the applicant’s justification for the deviation, and the justification must be acceptable to NRC.

NUREG/CR-3854 identifies codes and standards which may be used for fabricating components of SNF

transportation packaging. Detailed recommendations of this report are summarized in Section 1.5.2.6,

Table 1-1.

Several RGs and NUREGs provide guidance for structural design evaluation of packages using

information from existing codes and practices: (1) RG 7.8 identifies the load combinations to be used in

package design evaluation, (2) RG 7.6 provides design stress criteria for the containment system of Type

B packages, (3) RGs 7.11 and 7.12 describe criteria for precluding brittle fracture in package containers

made of ferritic steels, (4) NUREG/CR-4554A discusses the buckling evaluation of containment vessels,

(5) NUREG/CR-6322 provides guidance for buckling analysis of SNF baskets, (6) NUREG/CR-6007
NUREG-1617 2-8

provides guidance and criteria for design analysis of closure bolts for packagings, and (7) NUREG/CR3019 presents criteria for transportation package welds.

Guidance applicable for trunnions is provided in NUREG-0612 and ANSI N14.6

2.5.2 Material Properties

2.5.2.1 Materials and Material Specifications

Review packaging materials of construction and their specifications. Material specifications and

properties should be consistent with the design code or standard selected; if no standard is available, the

SAR should provide adequately documented material properties that are important for the design and

fabrication of the packaging. A list of pertinent material properties needed to define the material for

analysis should be provided. 

Verify that the materials of structural components whose structural integrity is essential for the package

to meet regulatory requirements have sufficient fracture toughness to preclude brittle fracture under the

specified normal conditions of transport and hypothetical accident condition temperatures and loads. 

Brittle fracture must be precluded for the containment vessel under severe impact loads at the lowest

service temperature. Fracture toughness criteria for ferritic steel packaging containment vessels are

provided in RGs 7.11 and 7.12.

Verify that the material properties used are appropriate for the load condition (e.g., static or dynamic

impact loading, hot or cold temperature, wet or dry conditions, etc.). Verify that appropriate temperatures

at which allowable stress limits are defined are consistent with those temperatures expected in service

and determined in the thermal analysis. 

If the package has impact limiters, the adequacy of the method used for establishing their force-deflection

characteristics should be verified by testing. Testing of the impact limiters may be carried out statically, if

the effect of strain rates on the material crush properties is accounted for and properly included in the

force-deflection relationship for impact analysis. The force-deflection curve of the impact limiter should

be provided in the SAR for all directions evaluated for the packaging.

2.5.2.2 Prevention of Chemical, Galvanic, or Other Reactions 

Review the materials and coatings of the package to verify that they will not produce a significant

chemical or galvanic reaction among packaging components, among packaging contents, or between the

packaging components and the packaging contents. The review should also include consideration of a

possible reaction resulting from inleakage of water.

Evaluate the possible generation of hydrogen or other flammable gases; if appropriate, consider

embrittling effects of hydrogen taking into account the metallurgical state of the packaging materials.

For metallic components of the package that may come into physical contact with one another, the

possibility of eutectic reactions should be considered since such reactions can lead to melting at the
2-9 NUREG-1617

interface between the metals at a lower temperature than the melting points of the metals in contact.

Review methods used to prevent eutectic reactions.
NUREG-1617 2-10

2.5.2.3 Effects of Radiation on Materials 

Verify that any damaging effects of radiation on the packaging materials have been appropriately

considered. These effects include degradation of seals and sealing materials, and degradation of the

properties of coatings and structural materials.

2.5.3 Lifting and Tie-Down Standards for All Packages

2.5.3.1 Lifting Devices

Review the design and evaluation of those lifting devices that are a structural part of the package, their

connection with the package body, and the package body in the local area around the lifting devices. 

Verify that the design, testing, and analyses demonstrate that these devices comply with the requirements

of 10 CFR 71.45(a):

• Any lifting attachment which is a structural part of the package must be designed with a minimum

safety factor of three against yielding when used to lift the package in the intended manner

• A lifting attachment which is a structural part of the package must be designed so that its failure

under excessive load would not impair the ability of the package to meet other requirements.

The location and construction of the lifting devices should be shown on the packaging drawings. Any

other structural part of the package that could be used to lift the package must be rendered inoperable for

lifting during transport or be designed with strength equivalent to that required for lifting attachments.

2.5.3.2 Tie-Down Devices

Review the design and evaluation of the tie-down devices that are a structural part of the package, their

connection with the package body, and the package body in the local area around the tie-down devices. 

Verify that the design, testing, and analyses demonstrate that these devices comply with the requirements

of 10 CFR 71.45(b): 

• Any tie-down device which is a structural part of the package must be capable of withstanding,

without generating stress in any material of the package in excess of its yield strength, a static force

applied to the center of gravity of the package having a vertical component of 2 times the weight of

the package with its contents, a horizontal component along the direction in which the vehicle travels

of 10 times the weight of the package with its contents, and a horizontal component in the transverse

direction of 5 times the weight of the package with its contents.

• A tie-down device which is a structural part of the package must be designed so that its failure under

excessive load would not impair the ability of the package to meet other requirements.

The location and construction of the tie-down devices should be shown on the packaging drawings. Any

other structural part of the package that could be used to tie down the package must be rendered
2-11 NUREG-1617

inoperable for tying down the package during transport, or must be designed with strength equivalent to

that required for tie-down devices.

2.5.4 General Considerations for Structural Evaluation of Packaging

The SAR should demonstrate that the analyses or tests used to evaluate the package under normal

conditions of transport and hypothetical accident conditions have been adequately performed, including:

• The initial conditions (e.g., temperature, pressure, and residue heat ) used are the most limiting for test

or loading conditions of the packaging.

• The methods employed are appropriate for loading conditions considered and follow accepted

practices and precepts.

• Interpretations of evaluation results are correct.

• The drop orientations considered in the evaluation are the most damaging. Note that the most

damaging orientation for one component may not be the worst case for another component.

2.5.4.1 Evaluation by Analysis

If the structural evaluation is by analysis, the review should include the following:

• Verify that the SAR describes clearly the analysis models, methods, and results including all

assumptions and input data used. The analysis model should adequately represent the geometry,

boundary conditions, loading, material properties, and structural behavior of the packaging analyzed.

• Verify that the material model and properties are appropriate for the analyses. If the analysis is an

elastic analysis, the material should also be modeled as an elastic material. If the analysis is inelastic,

the actual material behavior or a conservative elastic-plastic material model representing the actual

material should be used. The SAR should describe how the material properties were obtained and

why the material model is appropriate for the loading conditions considered. For analysis involving

large strains, the reviewer should verify that a stress-strain curve is used.

• Verify that the applied (force and displacement) boundary conditions in the analysis model are

appropriate. For free-drop impact analyses, impact loads for package components are usually derived

from a rigid body dynamic analysis of the package and used in a quasi-static analysis of the

components. Verify that a dynamic amplification factor has been applied to the equivalent static load

to account for all vibration effects that have been ignored in the rigid-body dynamic and quasi-static

analysis. A summary of the quasi-static and dynamic analysis methods for impact analysis is provided

in NUREG/CR-3966.

• Verify that the solution method is appropriate for the evaluation. If a computer program is used, the

validity and reliability of the computer program should be verified. The SAR should describe the
NUREG-1617 2-12

solution method, the bench marking results, and the quality assurance program for maintaining and

using the computer code.

• Verify that the most critical combinations of environmental and loading conditions are evaluated. At a

minimum, the evaluation should cover all the initial and loading conditions listed in RG 7.8. In addition,

verify that all critical free-drop orientations are evaluated in the SAR, assuming that the impact can be

at any angle. In general, the drop orientations that should be evaluated consist of two groups: (1)

drops that produce the highest g-loads to be used for impact analysis of the package components and

(2) drops that attack the most vulnerable orientations and parts of the packaging (i.e., bolts, seals,

valves and ports). The first group includes drops with the package center of gravity (c.g.) located

directly above the center of the impact area. These drops are the end drops, the side drops, and the

c.g.-over-corner drops. It also includes slap-down drops where the package c.g. is not directly above

the impact area. A slap-down drop of a long package can produce a high g-load in the second impact

due to a whipping action generated by the force of the first impact. The number of drops in the

second group will depend on the vulnerable packaging components and their structural failure modes. 

Components vulnerable to impact loads should be protected from impacting directly by special design

features such as recessed construction, protective cover plate, and impact limiter. The SAR should

evaluate the consequences of all credible drops.

• Verify that the analysis results are correctly interpreted or used to demonstrate adequate margins of

safety of the structural design. The maximum stresses or strains should be compared to

corresponding design-code allowables. Verify that the response of the package to loads and load

combinations in terms of stress and strain to components and structural members is shown. The

structural stability of individual members, as applicable, should be evaluated.

2.5.4.2 Evaluation by Test

If the structural evaluation is by test, the review should include at least the following: 

• Verify that the test procedures, test equipment and the impact pad are adequate for package impact

testing. UCRL-ID-121673 provides guidelines for package drop testing including the use of reducedscale models, which are commonly used for testing SNF packages.

• Verify that the test specimen is fabricated using the same materials, methods, quality assurance, and

inspection specifications as specified in the design. Any differences should be identified and the

effects evaluated in the SAR. The specimens should include all safety components to be tested and

components that are expected to have significant effects on the test results. Substitutes for the

radioactive contents during the tests should have the same structural properties as the actual contents. 

The substitutes should have the same mass and same interaction with its surrounding packaging

component as the actual contents. The same criteria should be used for all other simulated

components to ensure that the simulated parts do not alter the test results. Verify that the scale

model test specimen is properly scaled, fabricated, and instrumented (if applicable). In general, scale

models do not provide reliable data to determine the leakage rate of the package. Verify that size

effects of the scale model test article are not significant. The SAR should provide data to show that

the size effect can be ignored, if a reduced scale model smaller than 1⁄4 scale is used.
2-13 NUREG-1617

• Verify that the selected drop orientations are sufficient for a thorough test of all critical components

of the package and the selection is supported by sound analysis or reasoning. The criteria in Section

2.5.4.1 for the selection of critical drop orientation for analysis can also be used here. The actual

drop conditions and the resulting structural response or damage should be measured and recorded

before, during, and after the tests. Verify that the methods and instruments are adequate for the

measurements and the measurements are sufficient for describing the structural response or damage. 

Both interior and exterior damage of the test specimen should be included.

• Verify that all test results are evaluated and their structural integrity implication interpreted. The test

conclusions should be valid and defensible. Unexpected or unexplainable test results indicating

possible testing problems or previously unknown specimen behavior should be discussed and

evaluated. In each test, the test measurements, damage, and observations should be consistent with

each other. Inconsistencies should be identified and their possible causes explained in the SAR. 

Unreliable results should be identified and the need for additional tests assessed. If the package is

permanently deformed or damaged, the possibility of further damage by subsequent test conditions

should be evaluated. In addition, if the final damage is severe, the margin of safety of the package

design against an unacceptable structural failure scenario such as a sudden or total collapse or rupture

should be evaluated. If the final damage indicates the possibility of an imminent unstable structural

failure, additional tests under the same test conditions should be performed to determine the

repeatability of the result. If acceptance tests are performed on the specimen after the structural

testing, the acceptance tests should be performed according to appropriate codes and standards.

2.5.5 Normal Conditions of Transport

The evaluation of the package performance under normal conditions of transport is based on the effects

of the tests specified in 10 CFR 71.71. The ambient air temperature before and after the tests must

remain near constant at that value between -29EC (-20EF) and +38EC (100EF) which are most

unfavorable for the feature under consideration. The initial internal pressure within the containment

system must be the MNOP unless a lower internal pressure consistent with the ambient temperature

assumed to precede and follow the tests is more unfavorable. Separate specimens may be used for the

free-drop test, the compression test, and the penetration test, if each specimen is subjected to the water

spray test before being subjected to any of the other tests. 

The SAR should show that the effectiveness of the package has not been reduced as a result of the

normal conditions of transport, as specified by 10 CFR 71.43(f).

2.5.5.1 Heat

Verify that the heat loading condition will not compromise the structural integrity of the package.

Review the circumferential and axial deformations and stresses (if any) that result from differential

thermal expansion. The evaluation should consider possible interferences resulting from a reduction in

gap sizes. Verify that the stresses are within the limits for normal condition loads.
NUREG-1617 2-14

The evaluations should be based on the maximum ambient temperature and the MNOP in combination

with the maximum internal heat load. For specified components of the package (e.g., elastomer seal,

neutron shield material, etc.), review the maximum temperatures, and their effect on the operation of the

package. Fatigue effects may be considered.
2-15 NUREG-1617

2.5.5.2 Cold

Verify that the evaluation for the cold condition is adequate. Confirm that the temperatures under the

cold test condition are consistent with the thermal section.

The evaluations should consider the minimum internal pressure in combination with the minimum heat

load, in combination with any residual fabrication stresses. Verify that differential thermal expansions that

could result in possible geometric interfaces have been considered.

Verify that the stresses are within the limits for normal condition loads.

2.5.5.3 Reduced External Pressure

Determine that the SAR adequately evaluates the package design for the effects of reduced external

pressure equal to 25 kPa (3.5 psi) absolute.

2.5.5.4 Increased External Pressure

Determine that the SAR adequately evaluates the package design for the effects of increased external

pressure equal to 140 kPa (20 psi) absolute. Consider this loading condition in combination with minimum

internal pressure. Consider the possibility of buckling (NUREG/CR-4554A).

2.5.5.5 Vibration

Determine that the SAR adequately evaluates the package design for the effects of vibration normally

incident to transport. The SAR should provide a determination of the acceleration due to vibration by test

or analysis. A fatigue analysis should be provided for highly stressed systems, considering the combined

stresses due to vibration, temperature, and pressure loads. If closure bolts are reused, verify that the bolt

preload is included in the fatigue evaluation (NUREG/CR-6007). Verify that a resonant vibration

condition, which can cause rapid fatigue damage, is not present in any packaging component. The effect

on package internals should be considered. References for vibration evaluation of transport packages

include NUREG/CR-0128 and NUREG/CR-2146.

2.5.5.6 Water Spray

Review the package for the effects of the water spray test that simulates exposure to rainfall of

approximately 5 cm (2 in) per hour for at least one hour. Verify that this test has no significant effects on

material properties.

2.5.5.7 Free Drop

Review the package design for the effects of the free-drop test. Review procedures for impact are

discussed in Section 2.5.4.
NUREG-1617 2-16

Review the evaluation of the closure lid bolt design, port cover plates, and other package components for

the combined effects of free-drop impact force, internal pressures, thermal stress, and all other

concurrently applied forces (e.g., seal compression force, bolt preload, etc.).

2.5.5.8 Corner Drop

This test applies only to fiberboard, wood, or fissile material rectangular packages not exceeding 50 kg

(110 lb) and fiberboard, wood, or fissile material cylindrical packages not exceeding 100 kg (220 lb). This

test is generally not applicable for SNF packages because their weight exceeds 100 kg (220 lb).

2.5.5.9 Compression

This test applies only to packages weighing up to 5000 kg (11,000 lb). This test is generally not applicable

for SNF packages because their weight exceeds 5000 kg (11,000 lb).

2.5.5.10 Penetration

Review the evaluation of the package for the penetration condition. Verify that the most vulnerable

orientations and locations of the package have been considered for this test condition.

2.5.6 Hypothetical Accident Conditions

The evaluation for hypothetical accident conditions must be based on sequential application of the tests

specified in 10 CFR 71.73, in the order indicated, to determine their cumulative effect on a package. With

respect to the initial conditions for the tests (except for the water immersion tests), the ambient air

temperature before and after the tests must remain near constant at that value between -29EC (-20EF)

and +38EC (100EF) which are most unfavorable for the feature under consideration. The initial internal

pressure within the containment system must be the MNOP unless a lower internal pressure consistent

with the ambient temperature assumed to precede and follow the tests is more unfavorable. Damage

caused by the tests is cumulative, and the evaluation of the ability of a package to withstand any one test

must consider the damage that resulted from the previous tests.

The package must have adequate structural integrity to satisfy the containment, shielding, subcriticality,

and temperature requirements of 10 CFR Part 71. Generally, inelastic deformation of the containment

closure system (e.g., bolts and flanges) is unacceptable for the containment evaluation. Deformation of

other parts of the containment vessel may be acceptable if the containment boundary is not compromised. 

Deformation of shielding components is reviewed in the shielding evaluation. Deformation of components

required for heat transfer and insulation is reviewed in terms of the thermal evaluation. Deformation of

components required for subcriticality is reviewed in the criticality evaluations.

2.5.6.1 Free Drop

Review the evaluation of the package for the free-drop test. Verify that structural integrity has been

evaluated for the drop orientation which produces the highest g-load and causes the most severe damage. 
2-17 NUREG-1617

If a feature such as a tie-down component is a structural part of the package, it must be included in the

drop-test configurations.

Review the evaluation of the closure lid bolt design, port cover plates, and other package components for

the combined effects of free-drop impact force, internal pressures, thermal stress, and all other

concurrently applied forces (e.g., seal compression, bolt preload, etc.).

Review the impact pad used for the free drop. Assure that an essentially unyielding pad, of adequate size,

has been used. For a package with lead shielding, the effects of lead slump should be evaluated for the

hypothetical accident condition free drop. The lead slump determined should be consistent with that

assumed in shielding evaluation.

2.5.6.2 Crush

This test is only specified for packages with a mass not greater than 500 kg (1100 lb), density not greater

than water, and radioactive contents greater than 1000 A2, not as special form material.

This test is generally not applicable to SNF packages.

2.5.6.3 Puncture

Review the evaluation of the package for the puncture test. Verify that the orientation and location for

which maximum damage would be expected have been considered. It should be noted that damages

resulting from the drop test must be included when evaluating the puncture test.

Generally, thin-shelled packages are susceptible to puncture damage. Verify that puncture at oblique

angles, near a support, at a valve, or at a penetration have been considered.

Although analytical methods are available for predicting punctures of plates, empirical formulas derived

from puncture test results of laminated panels are usually used for determining the package surface layer

thickness required for resisting punctures. The Nelm’s formula developed specifically for package design

provides the minimum thickness needed for preventing the puncture of the steel surface layer of a typical

steel-lead-steel laminated cask wall. NUREG/CR-4554B provides an empirical formula for puncture

evaluation based on empirical and analytical puncture studies. The formula is applicable for puncture at

an angle normal to the surface and at a location away from a stiff support under the surface. The

formula is conservative for solid packaging walls but may be nonconservative for punctures at an oblique

angle, where the delivery of the puncture energy is more concentrated than in a right angle impact. 

Fortunately, there are few oblique punctures that can involve the total impact energy of a package. In

general, oblique punctures may be critical for thin-shelled packages that require only a fraction of the total

impact energy to penetrate the packaging wall.

2.5.6.4 Thermal

Verify that the package design is evaluated for a fully engulfing fire as specified in 10 CFR 71.73(c)(4). 

Any damage resulting from the free drop or puncture conditions must be incorporated into the initial
NUREG-1617 2-18

condition of the package for the fire test. Confirm that the determination of the maximum pressure in the

package during or after the test considers the temperatures resulting from the fire and any increase in gas

inventory caused by thermal combustion or decomposition process. Verify that the maximum thermal

stresses, which can occur either during or after the fire, are evaluated and are consistent with the

Thermal Evaluation section of the SAR.
2-19 NUREG-1617

2.5.6.5 Immersion – Fissile Material 

If water inleakage has not been assumed for the criticality analysis, review the evaluation of a damaged

test specimen (i.e., after free drop, puncture, and fire) immersed under a head of water of at least 0.9 m

(3 ft) in the attitude for which maximum leakage is expected. 

2.5.6.6 Immersion – All Material 

Review the evaluation of a separate, undamaged specimen subjected to water pressure equivalent to

immersion under a head of water of at least 15 m (50 ft). For test purposes, an external pressure of

water of 150 kPa (21.7 psi) gauge is considered to meet these conditions. 

2.5.7 Special Requirement for Irradiated Nuclear Fuel Shipments

For a package of irradiated nuclear fuel with activity greater than 37 PBq (106 Ci), 10 CFR 71.61 requires

that its undamaged containment system can withstand an external water pressure of 2 MPa (290 psi) for

a period of not less than one hour without collapse, buckling, or inleakage of water. The SAR should

provide analysis or test results to show that the containment structure will not collapse or buckle within

one hour after the pressure is applied. This test applies only to the containment system. No structural

support from other packaging components should be considered unless the component is an integral part

of the containment system. The inleakage requirement has not been met if the stresses around the

closure seal region exceed the yield stress limits.

2.5.8 Internal Pressure Test

For a package with a MNOP exceeding 35 kPa (5 psi) gauge, 10 CFR 71.85(b) requires that prior to first

use the containment system be pressure tested at 150% of its MNOP. The analysis of this acceptance

test should be provided in the SAR. The analysis should show that the package containment structure

does not yield under the test pressure and the stresses are within the allowable stress limits set by the

design code. 

2.5.9 Appendix

The appendix may include a list of references, copies of any applicable references not generally available

to the reviewer, computer code descriptions, input and output files, test results, and other appropriate

supplemental information.

2.6 EVALUATION FINDINGS

The structural review should result in the following findings, as appropriate:

2.6.1 Description of Structural Design

The staff has reviewed the package structural design description and concludes that the contents of the

application meet the requirements of 10 CFR 71.31
NUREG-1617 2-20

The staff has reviewed the codes and standards used in package design and find that they are acceptable.

2.6.2 Material Properties

To the maximum credible extent, there are no significant chemical, galvanic or other reactions among the

packaging components, among package contents, or between the packaging components and the contents

in dry or wet environment conditions. The effects of radiation on materials are considered and package

containment is constructed from materials that meet the requirement of RGs 7.11 and 7.12. 

2.6.3 Lifting and Tie-down Standards for All Packages

The staff has reviewed the lifting and tie-down systems for the package and concludes that they meet

10 CFR 71.45 standards.

2.6.4 General Considerations for Structural Evaluation of Packaging

The staff has reviewed the packaging structural evaluation and concludes that the application meets the

requirements of 10 CFR 71.35.

2.6.5 Normal Conditions of Transport

The staff has reviewed the packaging structural performance under the normal conditions of transport

and concludes that there will be no substantial reduction in the effectiveness of the packaging.

2.6.6 Hypothetical Accident Conditions

The staff has reviewed the packaging structural performance under the hypothetical accident conditions

and concludes the packaging has adequate structural integrity to satisfy the subcriticality, containment,

shielding, and temperature requirements of 10 CFR Part 71.

2.6.7 Special Requirement for Irradiated Nuclear Fuel Shipments

The staff has reviewed the containment structure and concludes that it will meet the 10 CFR 71.61

requirements for irradiated nuclear fuel shipments.

2.6.8 Internal Pressure Test

The staff has reviewed the containment structure and concludes that it will meet the 10 CFR 71.85(b)

requirements for pressure test without yielding.

2.7 REFERENCES

ANSI N14.6 Institute for Nuclear Materials Management, ANSI N14.6, “Special

Lifting Devices for Shipping Containers Weighing 10,000 Pounds

(45000 kg) or More for Nuclear Materials,” New York, NY, 1993.
2-21 NUREG-1617

B&PV Division 3 Code American Society of Mechanical Engineers, “ASME Boiler and Pressure

Vessel Code, Section III, Division 3, Containment Systems and Transport

Packagings For Spent Nuclear Fuel and High Level Radioactive Waste,”

New York, NY, 1998.

NUREG-0612 U.S. Nuclear Regulatory Commission, “Control of Heavy Loads at

Nuclear Power Plants,” NUREG-0612, National Technical Information

Service, Springfield, VA, July 1980.

NUREG/CR-0128 U.S. Nuclear Regulatory Commission, “Shock and Vibration

Environments for a Large Shipping Container During Truck Transport

(part II),” NUREG/CR-0128, U.S. Government Printing Office,

Washington, D.C., August, 1978.

NUREG/CR-2146 U.S. Nuclear Regulatory Commission, “Dynamic Analysis to Establish

Normal Shock and Vibration of Radioactive Material Shipping Packages,

Volume 3: Final Summary Report,” NUREG/CR-2146, Vol. 3, U.S.

Government Printing Office, Washington, D.C., October 1983.

NUREG/CR-3854 U.S. Nuclear Regulatory Commission, “Fabrication Criteria for Shipping

Containers,” NUREG/CR-3854, Lawrence Livermore National

Laboratory, Livermore, CA, March 1985.

NUREG/CR-3966 U.S. Nuclear Regulatory Commission, “Methods for Impact Analysis of

Shipping Containers,” NUREG/CR-3966, U.S. Government Printing

Office, Washington, D.C., November 1987.

NUREG/CR-4554A U.S. Nuclear Regulatory Commission, “SCANS (Shipping Cask Analysis

System): A microcomputer Based Analysis System for Shipping Cask

Design Review, Volume 6 – Theory Manual: Buckling of Circular

Cylindrical Shells,” NUREG/CR-4554, Vol. 6, U.S. Government Printing

Office, Washington, D.C., February 1990.

NUREG/CR-4554B U.S. Nuclear Regulatory Commission, “SCANS (Shipping Cask Analysis

System): A microcomputer Based Analysis System for Shipping Cask

Design Review, Volume 7 – Theory Manual: Puncture of Shipping

Casks,” NUREG/CR-4554, Vol. 7, U.S. Government Printing Office,

Washington, D.C., February 1990.

NUREG/CR-6007 U.S. Nuclear Regulatory Commission, “Stress Analysis of Closure Bolts

for Shipping Casks,” NUREG/CR-6007, U.S. Government Printing

Office, Washington, D.C., January 1993.
NUREG-1617 2-22

NUREG/CR-6322 U.S. Nuclear Regulatory Commission, “Buckling Analysis of Spent Fuel

Basket,” NUREG/CR-6322, U.S. Government Printing Office,

Washington, D.C., May 1995.

RG 7.6 U.S. Nuclear Regulatory Commission, “Design Criteria for the Structural

Analysis of Shipping Cask Containment Vessels,” Regulatory Guide 7.6,

Rev.1, U.S. Government Printing Office, Washington, D.C., March 1978.

RG 7.8 U.S. Nuclear Regulatory Commission, “Load Combinations for the

Structural Analysis of Shipping Casks,” Regulatory Guide 7.8, U.S.

Government Printing Office, Washington, D.C., May 1977.

RG 7.11 U.S. Nuclear Regulatory Commission, “Fracture Toughness Criteria of

Base Material for Ferritic Steel Shipping Cask Containment Vessels with

a Maximum Wall Thickness of 4 Inches (0.1m),” Regulatory Guide 7.11,

U.S. Government Printing Office, Washington, D.C., June 1991.

RG 7.12 U.S. Nuclear Regulatory Commission, “Fracture Toughness Criteria of

Base Material for Ferritic Steel Shipping Cask Containment Vessels with

a Wall Thickness Greater than 4 Inches (0.1m),” Regulatory Guide 7.12,

U.S. Government Printing Office, Washington, D.C., June 1991.

UCRL-ID-121673 Mok, G.C., R.W. Carlson, S.C. Lu, and L.E. Fischer, “Guidelines for

Conducting Impact Tests on Shipping Packages for Radioactive

Material,” UCRL-ID-121673, Lawrence Livermore National Laboratory,

Livermore, CA, September 1995.
3-1 NUREG-1617

3 THERMAL REVIEW

3.1 REVIEW OBJECTIVE

The objective of this review is to verify that the thermal performance of the package has been adequately

evaluated for the tests specified under normal conditions of transport and hypothetical accident conditions

and that the package design satisfies the thermal requirements of 10 CFR Part 71.

3.2 AREAS OF REVIEW

The SAR must be reviewed for adequacy of the description and evaluation of the thermal design. Areas

of review include the following:

3.2.1 Description of the Thermal Design

3.2.1.1 Packaging Design Features

3.2.1.2 Codes and Standards

3.2.1.3 Content Heat Load Specification

3.2.1.4 Summary Tables of Temperatures

3.2.1.5 Summary Tables of Pressures in the Containment Vessel

3.2.2 Material Properties and Component Specifications

3.2.2.1 Material Thermal Properties

3.2.2.2 Technical Specifications of Components 

3.2.2.3 Thermal Design Limits of Package Materials and Components

3.2.3 General Considerations for Thermal Evaluations

3.2.3.1 Evaluation by Analyses 

3.2.3.2 Evaluation by Tests

3.2.3.3 Confirmatory Analyses

3.2.3.4 Effects of Uncertainties

3.2.4 Evaluation of Accessible Surface Temperatures

3.2.5 Evaluation under Normal Conditions of Transport

3.2.5.1 Heat and Cold

3.2.5.2 Maximum Normal Operating Pressure

3.2.5.3 Maximum Thermal Stress

3.2.6 Evaluation under Hypothetical Accident Conditions

3.2.6.1 Initial Conditions

3.2.6.2 Fire Test

3.2.6.3 Maximum Temperatures and Pressures

3.2.6.4 Maximum Thermal Stresses

3.2.7 Appendix
NUREG-1617 3-2

3.3 REGULATORY REQUIREMENTS

Regulatory requirements of 10 CFR Part 71 applicable to the thermal review are as follows:

3.3.1 Description of the Thermal Design

The packaging must be described in sufficient detail to provide an adequate basis for its evaluation.

[10 CFR 71.31(a)(1), 10 CFR 71.33(a)(5), 10 CFR 71.33(a)(6), 10 CFR 71.33(b)(1), 10 CFR 71.33(b)(3),

10 CFR 71.33(b)(5), 10 CFR 71.33(b)(7), and 10 CFR 71.33(b)(8)]

The SAR must identify established codes and standards applicable to the thermal design.

[10 CFR 71.31(c)]

The thermal design must not depend on a mechanical cooling system to meet the containment

requirements of 10 CFR 71.51(a). [10 CFR 71.51(c)]

3.3.2 Material Properties and Component Specifications

The package must be described in sufficient detail to provide an adequate basis for its evaluation.

[10 CFR 71.31(a)(1), 10 CFR 71.33(a)(5), and 10 CFR 71.33(b)(3)]

3.3.3 General Considerations for Thermal Evaluations

The package must be evaluated to demonstrate that it satisfies the thermal requirements specified in

10 CFR Part 71, Subpart E, under the conditions and tests of Subpart F. [10 CFR 71.31(a)(2),

10 CFR 71.35(a), and 10 CFR 71.41(a)]

3.3.4 Evaluation of Accessible Surface Temperatures

The package must be designed, constructed, and prepared for shipment so that the accessible surface

temperature of a package in still air at 38EC (100EF) in the shade will not exceed 85EC (185EF) in an

exclusive-use shipment. [ 10 CFR 71.43(g)] (Temperature limits for non-exclusive-use shipments are

assumed not to apply to spent nuclear fuel (SNF) packages.)

3.3.5 Thermal Evaluation under Normal Conditions of Transport

The package design must be evaluated to determine the effects of the conditions and tests under normal

conditions of transport. The ambient temperature preceding and following the tests must remain near

constant at that value between -29EC (-20EF) and +38EC (100EF) which are the most unfavorable for the

feature under consideration. The initial internal pressure within the containment system must be

considered to be the maximum normal operating pressure (MNOP), unless a lower internal pressure

consistent with the ambient temperature considered to precede and follow the tests is more unfavorable. 

The conditions and tests of 10 CFR 71.71(c)(1) and 10 CFR 71.71(c)(2) for heat and cold respectively

are the primary thermal tests for normal conditions of transport. [10 CFR 71.71]
3-3 NUREG-1617

The package must be designed, constructed, and prepared for transport so that there will be no significant

decrease in packaging effectiveness under the tests specified in 10 CFR 71.71 (normal conditions of

transport). [10 CFR 71.43(f) and 10 CFR 71.51(a)(1)]

3.3.6 Thermal Evaluation under Hypothetical Accident Conditions

The package design must be evaluated to determine the effects of the conditions and tests under a

hypothetical accident. This accident includes a sequence of incidents (impact, puncture, thermal, and

immersion) on a package (the crush test is generally not applicable to packages for SNF). Except for the

water immersion tests, the ambient temperature preceding and following the tests must remain constant at

that value between -29EC (-20EF) and +38EC (100EF) which are the most unfavorable for the feature

under consideration. The initial internal pressure within the containment system must be considered to be

the MNOP, unless a lower internal pressure consistent with the ambient temperature considered to

precede and follow the tests is more unfavorable. The 30-minute, 800EC (1475EF) fire test of

10 CFR 71.73(c)(4) on a damaged package is the primary thermal test for hypothetical accident

conditions. [10 CFR 71.73]

3.4 ACCEPTANCE CRITERIA

3.4.1 Description of the Thermal Design

The regulatory requirements in Section 3.3.1 identify the acceptance criteria.

3.4.2 Material Properties and Component Specifications

In addition to the regulatory requirements identified in Section 3.3.2, the temperatures of the materials and

components used in the package should not exceed their specified maximum allowable temperatures.

3.4.3 General Considerations for Thermal Evaluations

In addition to the regulatory requirements identified in Section 3.3.3, the models used in the thermal

evaluation must be described in sufficient detail to permit an independent review, with confirmatory

calculations, of the package thermal design.

3.4.4 Evaluation of Accessible Surface Temperature

The regulatory requirements in Section 3.3.4 identify the acceptance criteria.

3.4.5 Thermal Evaluation under Normal Conditions of Transport

The regulatory requirements in Section 3.3.5 identify the acceptance criteria.

3.4.6 Thermal Evaluation under Hypothetical Accident Conditions

The regulatory requirements in Section 3.3.6 identify the acceptance criteria.
NUREG-1617 3-4

3.5 REVIEW PROCEDURES

The following procedures are generally applicable to the thermal review of all SNF transportation

packages. Since packages for shipment of SNF are generally intended to be shipped by exclusive-use,

only exclusive-use shipments are assumed in the following SRP review procedures.

The thermal review is based in part on the descriptions and evaluations presented in the General

Information and Structural Evaluation sections of the SAR. Similarly, results of the thermal review are

considered in the review of the SAR sections on Structural Evaluation, Containment Evaluation, Shielding

Evaluation, Criticality Evaluation, Operating Procedures, and Acceptance Tests and Maintenance

Program. Examples of SAR information flow into, within, and from the thermal review are shown in

Figure 3-1.

3.5.1 Description of the Thermal Design

3.5.1.1 Packaging Design Features

Review the general description of the package presented in the General Information section of the SAR

and any additional description of the thermal design in the Thermal Evaluation section. Verify that the

package description in the General Information section of the SAR includes:

• A description of any structural and mechanical means for the transfer and dissipation of heat

• The identity and volumes of receptacles containing coolant

• The MNOP of the containment system

• The maximum amount of content decay heat

• The identity and volumes of any coolants. Verify that the thermal design does not depend on the

presence of a mechanical cooling system to ensure containment.

All text, drawings, figures, and tables describing the thermal features in the Thermal Evaluation section

should be consistent with those of the General Information section as well as those used in the applicant’s

thermal evaluation. Particular emphasis should be placed on the consistency of the component

dimensions, materials, and material properties.

3.5.1.2 Codes and Standards

Verify that the established codes and standards used in all aspects of the thermal design and evaluation of

the package, including material properties and components, are identified.

3.5.1.3 Content Heat Load Specification
3-5 NUREG-1617

Verify that the maximum decay heat of the package contents reported in the Thermal Evaluation section

of the SAR is consistent with that in the General Information section and that this heat load is

appropriately considered in all thermal evaluations.
NUREG-1617 3-6
3-7 NUREG-1617

Figure 3-1 SAR Information Flow for the Thermal Review

Review the method in which the actual heat load is determined, and ensure that it is consistent with the

SNF content specifications (e.g., burnup, enrichment, cooling time). If the heat load is based on the mass

and decay energies of the contents, verify that it has been properly determined. The computer codes

discussed in Section 5.5.2 for determination of neutron and gamma sources are often useful for

calculating content decay heat loads. These codes are especially useful for SNF that contains a large

number of radionuclide species.

3.5.1.4 Summary Tables of Temperatures

Confirm that summary tables of the temperatures of package components including, but not limited to, the

fuel/cladding, basket, impact limiters, containment vessel, seals, shielding, and neutron absorbers are

consistent with the temperatures presented in the General Information and Structural Evaluation sections

of the SAR for the normal conditions of transport and hypothetical accident conditions. Confirm that the

summary tables contain the design temperature limits for each of the components for the normal

conditions of transport and hypothetical accident conditions. For the hypothetical accident condition fire,

these summarized temperatures should additionally include the maximum temperatures after fire, the

elapsed time from the beginning of the fire to the occurrence of these maximum temperatures, and the

post-fire steady-state temperatures of each package component. Confirm that the temperatures and

design temperature limit criteria for the package components are consistent throughout the appropriate

sections of the SAR.

3.5.1.5 Summary Tables of Pressures in the Containment System

Verify that summary tables of the pressure in the containment system under the normal conditions of

transport and hypothetical accident conditions are consistent with the pressures presented in the General

Information, Structural Evaluation, Containment Evaluation, and Acceptance Tests and Maintenance

Program sections of the SAR. The design pressure limits of the package components at the temperatures

producing the pressures should be presented in the tables.

3.5.2 Material Properties and Component Specifications

3.5.2.1 Material Properties

Confirm that the thermal properties necessary to calculate thermal transport in the package as well as

from the package to the environment are presented. These properties include, but are not limited to: 

• thermal conductivity

• specific heat

• density

• thermal radiation emissivity of the package surfaces.
NUREG-1617 3-8

Verify that the thermal emissivities are appropriate for the specific package surface conditions and

radiant energy spectrums for each thermal condition being evaluated. Confirm that the type of emittances

(hemispherical vis-à-vis normal) are specified. The thermal radiation absorptivity on the external

packaging surface may be conservatively assumed to be unity to compensate for changes in the package

surface from dirt, weathering, and handling during its lifetime. Consideration of a proposed value of less

than unity in the SAR should be based on the demonstration that controls and procedures will be in place

to ensure such a value throughout the package lifetime. Periodic visual examination followed by paint

touch-up or washing may be sufficient if the absorbtivity takes adequate account of weathering. These

controls and procedures should appear in the Operating Procedures and Acceptance Tests and

Maintenance Program sections of the SAR.

Verify that, for surrounding air and any fluids present within the package, the following additional

properties are presented:

• viscosity

• Prandtl number. 

Confirm that the given fluid properties are adequate for evaluating thermal convection parameters such as

the Prandtl number (a dimensionless number defined as the ratio of the momentum diffusivity to the

thermal diffusivity) which can be determined from the other thermal properties presented.

Confirm that the thermo-mechanical properties of any packaging material that may cause temperatureinduced pressures and/or stresses within the package materials are presented. These properties include,

but are not limited to:

• coefficient of thermal expansion

• modulus of elasticity

• Poisson’s ratio.

The coefficient of thermal expansion is usually the linear coefficient for isotropic solids and the volumetric

coefficient for fluids. For an isotropic material, the linear coefficient is one-third the volumetric

coefficient. 

Ensure that the structural properties that affect thermal stresses are consistent with the values reported in

the Structural Evaluation section.

If a package material is anisotropic, confirm that the directional properties of, for example, the thermal

conductivity, modulus of elasticity, and the linear expansion coefficient are provided.

Confirm that the temperatures at which phase changes, decomposition, dehydration and combustion will

occur are presented, along with thermal and thermo-mechanical properties resulting from the change.
3-9 NUREG-1617

Confirm that the thermal properties used for the analyses of the package are appropriate for the material

specified for the package in the General Information section and are consistent with those used in the

Structural Evaluation section of the SAR. Verify that the sources of the thermal properties used in the

SAR are referenced. Authoritative sources of material properties data include, but are not limited to,

those that reference experimental measurements. In general, textbooks are an unacceptable source of

material properties data. If the applicant experimentally measures the thermal properties of the material

and components used in the package, ensure that the experiments are performed under an approved

quality assurance program. 

Confirm the appropriateness of the use of temperature-dependent thermal properties in an analysis of the

package response to thermal loads. If the material properties are not presented as a function of

temperature, verify that the value conservatively under- or over-predicts temperatures or stresses, as

appropriate, compared to the equivalent temperature-dependent property.

3.5.2.2 Technical Specifications of Components

Verify that references for the technical specifications of package components such as O-rings, pressure

relief valves, bolts, etc., are identified. Confirm that any temperature constraints on the function of the

components are identified (such as the allowable stress in a bolt). Verify that the minimum allowable

service temperature of all components is less than or equal to -40EC (-40EF) unless a minimum heat load

is specified ( see Section 3.5.5.1).

3.5.2.3 Thermal Design Limits of Package Materials and Components

Confirm that the maximum allowable temperatures for each component that could affect the containment,

shielding, and criticality functions of the package are specified.

Verify that the maximum allowable fuel/cladding temperature is justified. The justification should consider

the fuel and clad materials, irradiation conditions (e.g., the absorbed dose, neutron spectrum, and fuel

burnup), and the shipping environment including the fill gas. Other necessary considerations include the

elapsed time from removal of the SNF from the core to its placement into the transportation packaging, its

time duration in the packaging, and its post-transport disposition. Examples of temperature limits include,

but are not limited to:

• the temperature limit for metal fuel should be less than the lowest melting point eutectic of the fuel

• the temperature limit on the irradiated clad in an inert gas environment as determined by creep, creep

rupture, or diffusion controlled cavity growth (PNL-6189, UCID-21181), as appropriate. 

Verify that the temperature range of the thermal and structural properties for each package material

exceed the specified and predicted temperature limits for the material.

3.5.3 General Considerations for Thermal Evaluations
NUREG-1617 3-10

Thermal evaluations of the package can be performed by either analyses or tests, or by a combination of

both. Because of their mass and cost and the difficulty of decay-heat simulation, SNF packages are

normally evaluated by analysis. In addition, the use of analysis to evaluate the thermal performance of a

package will allow the “margin of safety” in the package design to be determined.

Review the Structural Evaluation and Thermal Evaluation sections of the SAR to determine the response

of the package to the normal conditions of transport and hypothetical accident conditions. Verify that the

corresponding models used in the thermal analyses are consistent with these effects. For example, the

package might have impact limiters or an external neutron shield that would be damaged during the

structural and thermal tests of 10 CFR 71.73.

3.5.3.1 Evaluation by Analyses

Confirm that the methods of thermal analysis are identified and sufficiently described to permit a complete

review and independent verification. The thermal analyses in the SAR can be based on simple

calculations, spreadsheet-type analyses, or detailed computer simulations. The level of detail appropriate

for each analysis, including assumptions, depends on many physical variables such as: the package

materials, SNF decay heat, geometric complexity, and package component surface conditions. Ensure

that each method of thermal analysis:

• is properly referenced or derived in the SAR as appropriate

• clearly and completely states the assumptions made in modeling heat sources and heat transfer paths

and modes

• accurately represents the physical characteristics of the package consistent with the above discussed

thermal design features (Section 3.5.1.1)

• uses appropriate thermal properties for the materials of construction (Section 3.5.2.1)

• uses appropriate expressions for conduction, convection, and thermal radiation among package

components, and from the surfaces of the package to the environment

• correctly incorporates the appropriate specified temperature and thermal boundary conditions for the

normal conditions of transport and hypothetical accident conditions.

For the 30-minute fire, the majority of the heat input to the package being tested will be radiative. For

convective heat transfer, a convective heat transfer coefficient appropriate for the conditions that would

exist if the package was exposed to the fire specified in 10 CFR 71.73, should be used. Flame velocities

in an open pool fire may be used in determining the appropriate convective heat transfer coefficient. 

Flame velocities in open pool fires are discussed by Burgess and Fry (1990), Burgess (1987), and

Schneider and Kent (1989). During the post-fire cooldown, natural convection should be assumed.

For the thermal analysis of steady-state normal conditions of transport (including the case for determining

the accessible surface temperature), confirm the presence of either the Rayleigh number (product of the
3-11 NUREG-1617

ratio of buoyancy force to the viscous force and the ratio of the momentum diffusivity to the thermal

diffusivity) for the air at the surfaces of the package, or information sufficient to calculate the Rayleigh

number. For the 30-minute fire, the appropriate radiative and convective heat transfer coefficients should

be supported by the necessary correlations. Any correlations used in the analysis should be properly

explained and justified.

Confirm that the assumptions about contact resistance at material interfaces, energy transport across gaps

or enclosures, etc. are presented and appropriate. For example, the assumption of a maximum contact

resistance between component surfaces during steady-state conditions or during a post-fire cooldown and

the assumption of no contact resistance between component surfaces during a fire will result in

maximizing the calculated component temperatures for normal conditions of transport and hypothetical

accident conditions.

Under the conditions where any of the cask component temperatures are close (within 5%) to their

limiting values during an accident or the MNOP is within 10% of its design basis pressure, or any other

special conditions, the applicant should consider, by analysis, the potential impact of the fission gas in the

canister to the cask component temperature limits and the cask internal pressurization.

In the case of computer analysis, the applicant may use standard off-the-shelf software or develop a

computer code to perform a specific analysis. Verify that the code has been benchmarked and is

maintained and operated under a quality assurance program. Verify that the code has been appropriately

used. Ensure that the SAR includes appropriate code input and output files to enable a detailed review of

the analysis.

3.5.3.2 Evaluation by Tests

For those results determined by tests, verify that a description of the test package, the test facility, and the

test procedures used for simulating either the normal conditions of transport or hypothetical accident

conditions are reported in adequate detail. Confirm that the test package was fabricated, the test facility

operated, and the test results evaluated under proper quality assurance programs.

Review the ability of both the test facilities and test procedures to meet the range of specified

temperatures: from -29EC (-20EF) to 38EC (100EF) for normal conditions of transport and both 38EC

(100EF) and 800EC (1475EF) for hypothetical accident conditions. Confirm that the facilities can simulate

the specified heat-transfer boundary conditions:

• incident heat fluxes equivalent to or exceeding the specified insolation requirements during the normal

conditions of transport or the post-fire environment for hypothetical accident conditions

• incident heat fluxes equivalent to or exceeding the specified convective and radiative heat transfer

environment, including specified emissivities, for a minimum 30-minute period representing the

hypothetical accident condition fire

• an environment that assures an adequate supply and circulation of oxygen for initiating and naturally

terminating the combustion of any burnable package component.
NUREG-1617 3-12

Confirm that the test package, with a simulated package contents and any attached test instrumentation or

hardware, adequately simulates the thermal behavior of the actual package design.

Verify that the locations of the temperature and heat flux sensing devices are shown on figures in the

SAR. Verify that the temperature sensing devices are placed on the test package: 

• on applicable components

• in such a manner that they do not unduly affect local temperatures

• in locations where maximum temperatures are expected and where other temperatures need to be

determined

• in locations that permit reasonable interpolation or extrapolation of measured temperatures for

estimating temperatures in unmonitored regions of the package.

The applicable components include, but are not limited to, the containment vessel, fuel basket, seals,

radiation shielding, criticality controls, and impact limiters. Confirm that the temperature sensing devices

are measuring the temperature of the component, not that of the component environment. 

Verify that the test time is sufficient for temperatures to reach steady-state conditions under normal

conditions of transport or their peak following cessation of the hypothetical accident condition fire. To the

extent that specified boundary conditions, the decay heat of the contents, or specified temperatures are

not achieved during a test, verify that the evaluations include appropriate corrections to the temperature

data.

Additional guidelines on reviewing thermal tests under hypothetical accident conditions are presented in

NUREG/CR-5636, SAND85-0196, Hovingh and Carlson (1994), and UCRL-ID-110445.

3.5.3.3 Confirmatory Analyses

The rigor required of the confirmatory analysis will depend on the size of the margin between the

maximum package component temperatures determined by the applicant and the maximum temperature

limit specified for a material or component or the regulatory limit determined by the type of shipment. A

conservative method of analysis of the fire portion of the hypothetical accident is to mathematically apply

an 800EC (1475EF) surface temperature for 30 minutes to the package with the appropriate initial

temperature distribution and content decay heat. This will eliminate the questions about the flame velocity

and its effect on the convection heat input into the package. The analysis will still require the appropriate

boundary conditions during cooldown to calculate the maximum component temperatures. 

3.5.3.4 Effects of Uncertainties

Verify that the thermal evaluations appropriately address the effects of uncertainties in thermal and

structural properties of materials, test conditions and diagnostics, and analytical methods, as applicable.
3-13 NUREG-1617

3.5.4 Evaluation of Accessible Surface Temperatures

Determine that the thermal model used for the calculation of the accessible surface temperature is

presented in the SAR. This model should consist of a heat balance at the surface of the package in which

the decay heat from the contents at the surface of the package is equal to the convective and radiative

heat loses to the environment at an ambient temperature of 38EC (100EF). 

If the maximum surface temperature of a package exceeds the regulatory limit, a personnel barrier can

be placed around the package. This personnel barrier becomes the accessible package surface. The

thermal impedance of the barrier should be considered when determining the package temperatures for

normal conditions of transport.

Confirm that the maximum accessible surface temperature determined by the applicant is consistent with

the General Information section of the SAR. 

When appropriate, perform an independent analysis as described in Section 3.5.3.6 to confirm the

maximum accessible surface temperature determined by the applicant.

Ensure that the maximum temperature of the accessible package surface does not exceed 85EC (185EF)

for exclusive use shipment when the package is subjected to the heat conditions of 10 CFR 71.43(g).

3.5.5 Thermal Evaluation under Normal Conditions of Transport

3.5.5.1 Heat and Cold

Confirm that the thermal evaluation demonstrates that the tests for normal conditions of transport do not

result in significant reduction in packaging effectiveness, including:

• degradation of the heat-transfer capability of the packaging (such as creation of new gaps between

components)*

• changes in material conditions or properties (e.g., expansion, contraction, gas generation, and thermal

stresses) that affect the structural performance

• changes in the packaging or contents that affect containment, shielding, or criticality such as thermal

decomposition or melting of materials

• ability of the package to withstand the tests under hypothetical accident conditions.

Verify that the SAR properly determines the maximum temperatures of the package components during

normal conditions of transport when the package is in 38EC (100EF) still air with insolation, according to

the table in 10 CFR 71.71(c)(1), and the content heat load is the maximum allowable. Temperatures of

special interest include, but are not limited to, those of the fuel/cladding, containment vessel, seals,

shielding, criticality controls, and impact limiters. Confirm that the volume-averaged temperature of gases
NUREG-1617 3-14

are determined. Verify that the results are consistent with the General Information and Structural

Evaluation sections of the SAR.

Ensure that the SAR determines the minimum temperatures of the package components during normal

conditions of transport when the package is in -40EC (-40EF) still air without insolation and the content

heat load is the minimum allowable. If the SAR does not restrict the minimum heat load, the package

should be considered at a uniform temperature of -40EC (-40EF). Verify that these temperatures are

consistent with the Structural Evaluation section of the SAR.

Confirm that the maximum and minimum temperatures do not exceed their allowable limits, as specified in

Section 3.5.2.3. 

3.5.5.2 Maximum Normal Operating Pressure (MNOP)

Confirm that the SAR determines the maximum normal operating pressure when the package has been

subjected to the heat condition for one year, as specified in 10 CFR 71.71(c)(1). Ensure that the

evaluation has considered all possible sources of gases, such as:

• gases present in the package at closure

• fill gas released from the SNF rods

• fission product gases released from the SNF

• saturated vapor from material in the containment vessel including water vapor desorbed from the

containment system components or the package contents

• helium from the a-decay of the SNF contents

• hydrogen and other gases from radiolysis or chemical reactions (e.g., sodium-water)

• hydrogen and other gases from the dehydration, combustion, or decomposition of package

components.

Guidance on release of fill gas and fission product gas for PWR and BWR fuel is provided in Table 4-1.

Verify that MNOP is consistent with the Structural Evaluation section of the SAR.

If the package has any confined volumes other than the containment vessel (e.g., coolant tanks), confirm

that their pressures are properly determined and consistent with the Structural Evaluation.

3.5.5.3 Maximum Thermal Stresses
3-15 NUREG-1617

There are two sources of thermal stresses. These stresses can be caused by either spatial temperature

gradients in constrained package components or by interference between components due to different

thermal expansions of the components.

Two cases should be investigated for interferences between package components assembled at “room”

temperature. These include:

• a steady-cold (-40EC [-40EF]) environment with maximum SNF decay heat without insolation

• a steady-hot (38EC [100EF]) environment with maximum SNF decay heat with insolation

Confirm that the dimensions of the package components, and the clearances or interferences, for the

above cases are presented in the SAR. Verify that an appropriate method for estimation of the stresses

from the interferences between components is described. Verify that the stresses from any interferences

between components are consistent with the Structural Evaluation section of the SAR.

3.5.6 Thermal Evaluation under Hypothetical Accident Conditions

Verify that the package has been evaluated to demonstrate the effects of the tests for hypothetical

accident conditions.

3.5.6.1 Initial Conditions

 

Prior to the fire test, the package must be evaluated for the effects of the drop, crush (if applicable), and

puncture tests. Ensure that the initial physical condition of the package represented in the thermal

evaluations under hypothetical accident conditions is consistent with these results from the Structural

Evaluation section of the SAR.

Verify that the SAR justifies the most unfavorable initial ambient temperature between -29EC (-20EF)

and +38EC (100EF). Unless the package is susceptible to increased structural damage at lower

temperatures, the initial ambient temperature should be 38EC (100EF). Verify that the initial steady-state

temperature distribution is consistent with the results from the thermal evaluations under normal conditions

of transport.

Confirm that the initial internal pressure of the package is the MNOP unless a lower internal pressure,

consistent with the initial ambient temperature, is more unfavorable. Similarly, confirm that the internal

heat load of the SNF contents is at its maximum allowable power unless a lower power, consistent with

the temperature and pressure, is more unfavorable.

3.5.6.2 Fire Test

Verify that the package is exposed to the 800EC (1475EF) fire environment for a minimum of 30 minutes

and that surface and fire emissivity are greater than or equal to 0.8 and 0.9, respectively. Confirm that

the flame velocities are specified and appropriate for the hydrocarbon fire and that the appropriate

correlation for convection in the fire is used as a boundary condition.
NUREG-1617 3-16

Verify that after the fire:

C No artificial cooling is applied to the package

C The package is subjected to full insolation

C The evaluation continues until the post-fire, steady-state condition is achieved

C An adequate supply of oxygen is continued throughout this period

C All combustion is allowed to proceed until it terminates naturally.

3.5.6.3 Maximum Temperatures and Pressures

Verify that the SAR appropriately evaluates the transient peak temperatures of the package components

as a function of time after the fire. The maximum temperatures in the components will occur following

cessation of the fire, with the delay time increasing with the distance inward from the package surface.

Verify also that the SAR determines the maximum temperatures of the post-fire, steady-state condition.

Confirm that the maximum temperatures do not exceed the maximum allowable temperature limits.

Verify that the evaluation of the maximum pressure in the containment vessel is based on MNOP

(Section 3.5.5.2) as it is affected by the fire-caused increases in package component temperatures. 

Confirm that possible increases in gas inventory (e.g., from fuel rod failure) have been considered in the

pressure determination.

If the package has any confined volumes other than the containment vessel (e.g.., coolant tanks), confirm

that their pressures are properly determined.

Verify that maximum temperatures and pressures are consistent with the Structural and Containment

Evaluations.

3.5.6.4 Maximum Thermal Stresses

Verify that the SAR evaluates the thermal stresses. The maximum interference between components in

a package during a hypothetical thermal accident usually occur during the post-fire cooldown. Where the

components are concentric, the tensile stresses occur in the outer component while the stresses in the

inner components are usually compressive.

Ensure that the maximum thermal stresses are consistent with those in the Structural Evaluation section of

the SAR.

3.5.7 Appendix
3-17 NUREG-1617

The appendix may include a list of references, copies of any applicable references not generally available

to the reviewer, computer code descriptions, input and output files, test facility and instrumentation

descriptions, test results, special analyses, and other appropriate supplemental information.

3.5.7.1 Justification for Assumptions or Analytical Procedures

Confirm that the applicant has stated and justified all assumptions used in the evaluation of the package. 

Review the appropriateness of and justification for the applicant’s assumptions and analytical procedures.

3.5.7.2 Computer Program Description

Confirm that the applicant describes all the computer programs used in the thermal evaluation of the

package. Verify that the space dimensionality and method of analysis (finite difference, finite element,

etc.) are identified. Verify that the range of applications and phenomena (linear, nonlinear; steady state,

transient; etc.) as well as the material properties and material models (isotropic, anisotropic, etc.) are

described. Verify that the various types of initial boundary conditions and thermal loads are described. 

Verify that solution techniques (direct or iterative for steady state; explicit, implicit, etc. for transient) are

identified. Also verify that any other capabilities (enclosure radiation with view factor calculation, thermal

stress analysis, etc.) that are applicable to the applicant’s thermal evaluation are identified and described. 

Verify that the computer programs are appropriate for the problem to which they are applied by the

applicant.

3.5.7.3 Computer Input and Output Files

Confirm that the applicant has submitted annotated input files, as applicable, for each problem (maximum

accessible surface temperature, normal conditions of transport, calculation of initial temperature

distribution for hypothetical accident, initial temperature distribution for analysis of thermal hypothetical

accident) analyzed using a computer code. Confirm that the applicant has submitted annotated output

files, as applicable, for each problem (maximum accessible surface temperature, normal conditions of

transport, calculation of initial temperature distribution for hypothetical accident conditions, and

temperature distribution histories for the thermal hypothetical accident condition during and following the

30-minute fire, until all the package component temperatures have reached their maxima).

3.5.7.4 Description of Test Facilities

Verify that the facilities used for performing thermal tests are described. The description shall include,

but is not limited to:

• the type of facility (furnace, pool fire, etc.)

• the method of heating the package (gas burners, electrical heaters, etc.).

The description of a furnace facility should include the volume and emissivity of the furnace interior as

well as the method of measuring the interior temperature. The oxygen concentration in a furnace test

should be consistent with that of a hydrocarbon-fuel fire.
NUREG-1617 3-18

For a pool fire facility, the size of the fire relative to the size of the package shall be specified. Verify that

the fire dimension conforms to the regulatory requirement that the fire thickness extend horizontally at

least one meter (but not more than three meters) beyond any external surface of the package. The

package will be positioned one meter above the surface of the fuel source. Verify that the method of

support of the package in a test facility is described and an analysis of the heat loss from the package

through the support to “ground” is presented. Review that the analysis of the heat loss from the package

through the support is appropriate.

Confirm that the sensors used to measure heat flux and temperature are identified and described. Verify

that the applicable operating ranges of the sensors are presented. Verify that the perturbation by the

sensor (due to heat losses along thermocouple leads, shadowing by heat flux measuring devices, etc.) on

the quantity to be measured (temperature, heat flux, etc.) is presented and quantified. Review that the

heat flux and temperature sensors are appropriate and that the measurements are corrected for the

perturbations by the sensors on the quantity to be measured. Verify that if calorimeters are used to

measure heat flux, the calorimeter readings are corrected to account for the difference in thermal inertia

between the calorimeter and the package. Verify that the method of correction of the calorimeter

reading is presented and review the method for appropriateness.

3.5.7.5 Test Results

Verify that test measurements including temperatures (or temperature histories) and flux (or flux 

histories) are presented. Verify that the corrected test results are presented and that appropriate methods

are used to obtain these corrections. Verify that, for the thermal portion of the hypothetical accident, the

time at which the 30-minute test starts and ends is clearly noted. Verify that the measurements (and

corrected results) are continued until steady state occurs (for tests for normal conditions of transport) or

until the maximum temperature occurs in all the package components (for test of the thermal portion of

the regulatory hypothetical accident).

Verify that photographs of the package components prior to and following the tests are presented. Verify

that photographs of regions of components with thermal damage (such as charring of the insulation,

damage to O-rings, etc.) are presented.

3.5.7.6 Applicable Supporting Documents or Specifications

Verify that the applicable sections from reference documents are included. These documents may

include the test plans used for the thermal tests, the thermal specifications of O-rings and other

components, and the documentation of the thermal properties of non-ASME approved materials used in

the package.

3.5.7.7 Special Analyses

Frequently, thermally driven special processes will occur in a package. These processes may include, but

are not limited to:

• generation of gases within the containment system
3-19 NUREG-1617

• effects of phase changes on package materials

• combustion, decomposition or dehydration of package materials.

The production of gases (e.g., hydrogen by radiolysis) or thermal decomposition of materials (e.g., a

neutron shield) may occur in the package. Phase changes of material resulting in a decrease of the

material density occurring in the containment system or in a lead shield can result in a pressure increase in

the system. The tests under hypothetical accident conditions may cause combustion, decomposition or

dehydration of components such as an impact limiter or the neutron shield material. 

Confirm that the applicant has identified all thermally driven special processes that will occur in the

package. Verify that the applicant has stated and justified all assumptions used in the quantification and

evaluation of these special processes. Review the appropriateness of and justification for the applicant’s

assumptions and analytical procedures. Verify that the results are incorporated in the appropriate

subsections of the Thermal Evaluation section.

Other supplemental calculations may be required to support evaluations presented in the Thermal

Evaluation section. Verify that all such special analyses meet the goals discussed in Section 3.5.3.1.

3.6 EVALUATION FINDINGS

The thermal review should result in the following findings, as appropriate:

3.6.1 Description of the Thermal Design

The staff has reviewed the package description and evaluation and concludes that they satisfy the thermal

requirements of 10 CFR Part 71.

3.6.2 Material Properties and Component Specifications

The staff has reviewed the material properties and component specifications used in the thermal

evaluation and concludes that they are sufficient to provide a basis for evaluation of the package against

the thermal requirements of 10 CFR Part 71.

3.6.3 General Considerations for Thermal Evaluations

The staff has reviewed the methods used in the thermal evaluation and concludes that they are described

in sufficient detail to permit an independent review, with confirmatory calculations, of the package thermal

design.

3.6.4 Evaluation of Accessible Surface Temperature

The staff has reviewed the accessible surface temperatures of the package as it will be prepared for

shipment and concludes that they satisfy 10 CFR 71.43(g) for packages transported by exclusive-use

vehicle.
NUREG-1617 3-20

3.6.5 Evaluation under Normal Conditions of Transport

The staff has reviewed the package design, construction, and preparations for shipment and concludes

that the package material and component temperatures will not extend beyond the specified allowable

limits during normal conditions of transport consistent with the tests specified in 10 CFR 71.71.

3.6.6 Evaluation under Hypothetical Accident Conditions

The staff has reviewed the package design, construction, and preparations for shipment and concludes

that the package material and component temperatures will not exceed the specified allowable short time

limits during hypothetical accident conditions consistent with the tests specified in 10 CFR 71.73.

3.7 REFERENCES

Burgess 1987 Burgess, M.H., “Heat Transfer Boundary Conditions in Pool Fires,”

Packaging and Transportation of Radioactive Material (PATRAM),

Vol. II., STI/PUB/718, IAEA, Vienna, 1987, pp. 423-431.

Burgess and Fry 1990 Burgess, M.H., and C.J. Fry, “Fire Testing for Package Approval,” Int.

J. Radioactive Materials Transport, Vol. 1, No. 1, Nuclear Technology

Publishing, Ashford, Kent, England, 1990, pp. 7-16.

60 FR 50247 U.S. Nuclear Regulatory Commission, “Compatibility With the

International Atomic Energy Agency (IAEA),” Federal Register, FR

50247, U.S. Government Printing Office, Washington, D.C., September

28, 1995.

SAND85-0196 Gregory, J.J., R. Mata, and N.R. Keltner, “Thermal Measurements in a

Series of Large Pool Fires,” SAND85-0196, TTC-0659, UC-71, Sandia

National Laboratories, Albuquerque, NM, August 1987.

Hovingh and Carlson Hovingh, J., and R.W. Carlson, “Thermal Testing Transport Packages

1994 for Radioactive Materials - Reality vs. Regulation,” ASME 1994

Pressure Vessel & Piping Conference, Minneapolis, MN, June 1994.

NUREG/CR-5636 U.S. Nuclear Regulatory Commission, “Fire and Furnace Testing of

Transportation Packages for Radioactive Materials,” NUREG/CR-5636,

U.S. Government Printing Office, Washington, D.C., January 1999.

PNL-6189 Levy, I.S., et al., “Recommended Temperature Limits for Dry Storage of

Spent Light Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas,” PNL6189, Pacific Northwest Laboratory, Richland, WA, May 1987.

Schneider and Kent Schneider, M.E., and L.A. Kent, “Measurements of Gas

Velocities and
3-21 NUREG-1617

1989 Temperatures in a Large Open Pool Fire,” Fire Technology, Vol. 25,

February 1989, pp. 51-80.

UCID-21181 Schwartz, M.W. and M.C. Witte, “Spent Fuel Cladding Integrity During

Dry Storage,” UCID-21181, Lawrence Livermore National Laboratory,

Livermore, CA, September 1987.

UCRL-ID-110445 VanSant, J.H., R.W. Carlson, L.E. Fischer, and J. Hovingh, “A Guide for

Thermal Testing Transport Packages for Radioactive Material - 

Hypothetical Accident Conditions,” UCRL-ID-110445, Lawrence

Livermore National Laboratory, Livermore, CA, February 9, 1993.
NUREG-1617 3-22
4-1 NUREG-1617

4 CONTAINMENT REVIEW

4.1 REVIEW OBJECTIVE

The objective of this review is to verify that the package design satisfies the containment requirements of

10 CFR Part 71 under normal conditions of transport and hypothetical accident conditions.

4.2 AREAS OF REVIEW

The SAR should be reviewed for adequacy of the description and evaluation of the containment design. 

Areas of review include the following:

4.2.1 Description of Containment System

4.2.1.1 Containment Boundary

4.2.1.2 Codes and Standards

4.2.1.3 Special Requirements for Damaged Spent Nuclear Fuel

4.2.2 Containment under Normal Conditions of Transport

4.2.2.1 Pressurization of Containment Vessel

4.2.2.2 Containment Criteria

4.2.2.3 Compliance with Containment Criteria

4.2.3 Containment under Hypothetical Accident Conditions

4.2.3.1 Pressurization of Containment Vessel

4.2.3.2 Containment Criteria

4.2.3.3 Compliance with Containment Criteria

4.2.4 Appendix

4.3 REGULATORY REQUIREMENTS

Regulatory requirements of 10 CFR Part 71 applicable to the containment review are as follows:

4.3.1 Description of Containment System

The packaging must be described in sufficient detail to provide an adequate basis for its evaluation.

[10 CFR 71.31(a)(1), 10 CFR 71.33(a)(4), 10 CFR 71.33(a)(5), 10 CFR 71.33(b)(1), 10 CFR 71.33(b)(3),

10 CFR 71.33(b)(5), and 10 CFR 71.33(b)(7)]

The SAR must identify established codes and standards applicable to the containment design.

[10 CFR 71.31(c)]

The package must include a containment system securely closed by a positive fastening device that

cannot be opened unintentionally or by a pressure that may arise within the package. [10 CFR 71.43(c)]
NUREG-1617 4-2

The package must be made of materials and construction that assure that there will be no significant

chemical, galvanic, or other reaction. [10 CFR 71.43(d)]

Any valve or similar device on the package must be protected against unauthorized operation and, except

for a pressure relief valve, must be provided with an enclosure to retain any leakage. [10 CFR 71.43(e)]

Spent fuel, with plutonium in excess of 0.74 TBq (20 Ci) per package, in the form of debris, particles,

loose pellets, or fragmented rods or assemblies must be packaged in a separate inner container (second

containment system) in accordance with 10 CFR 71.63(b). [10 CFR 71.63]

4.3.2 Containment under Normal Conditions of Transport 

The package must be evaluated to demonstrate that it satisfies the containment requirements of 10 CFR

Part 71, Subpart E, under the conditions and tests of Subpart F. [10 CFR 71.31(a)(2), 10 CFR 71.35(a),

and 10 CFR 71.41(a)]

A package must meet the containment requirements of 10 CFR 71.43(f) and 10 CFR 71.51(a)(1) under

the tests specified in 10 CFR 71.71 (normal conditions of transport), with no dependence on filters or a

mechanical cooling system. [10 CFR 71.51(c)]

The package may not incorporate a feature intended to allow continuous venting during transport.

[10 CFR 71.43(h)]

4.3.3 Containment under Hypothetical Accident Conditions

The package must be evaluated to demonstrate that it satisfies the containment requirements of 10 CFR

Part 71, Subpart E, under the conditions and tests of Subpart F. [10 CFR 71.31(a)(2), 10 CFR 71.35(a),

and 10 CFR 71.41(a)]

A package must meet the containment requirements of 10 CFR 71.51(a)(2) for hypothetical accident

conditions, with no dependence on filters or a mechanical cooling system. [10 CFR 71.51(c)]

4.4 ACCEPTANCE CRITERIA

4.4.1 Description of Containment System

In addition to the regulatory requirements identified in Section 4.3.1, the containment system should be

designed and constructed in accordance with Section III, Division 3, ASME Boiler and Pressure Vessel

(B&PV Division 3) Code. Alternatives codes should be justified in the SAR.

In addition to the regulatory requirements identified in Section 4.3.1, leakage from the containment system

should be determined in accordance with ANSI N14.5.

The codes, standards, and criteria for the inner containment system should generally be the same as those

of the outer containment system. Justification for differences should be presented in the SAR.
4-3 NUREG-1617

In addition to the regulatory requirements identified in Section 4.3.1, damaged fuel should be canned to

facilitate handling and to confine gross fuel particles to a known subcritical volume under normal

conditions of transport and hypothetical accident conditions.

4.4.2 Containment under Normal Conditions of Transport

In addition to the regulatory requirements identified in Section 4.3.2, combustible gases should not exceed

5% of the free gas volume in any confined region of the package while the containment vessel is sealed

and under normal transport conditions. The SAR should identify the allowable normal conditions of

transport volumetric leakage rates in accordance with ANSI N14.5.

4.4.3 Containment under Hypothetical Accident Conditions

In addition to the regulatory requirements identified in Section 4.3.3, combustible gases should not exceed

5% of the free gas volume in any confined region of the package while the containment vessel is sealed

and under hypothetical accident conditions. The SAR should identify the allowable hypothetical accident

conditions volumetric leakage rates in accordance with ANSI N14.5.

4.5 REVIEW PROCEDURES

The following procedures are generally applicable to the containment review of all SNF transportation

packages.

The containment review is based in part on the descriptions and evaluations presented in the General

Information, Structural Evaluation, and Thermal Evaluation sections of the SAR and follows the sequence

established to evaluate the packaging against applicable 10 CFR Part 71 requirements. Similarly, results

of the containment review are considered in the review of the SAR sections on Operating Procedures and

Acceptance Tests and Maintenance Program. Examples of SAR information flow into, within, and from

the containment review are shown in Figure 4-1.

4.5.1 Description of the Containment System

4.5.1.1 Containment Boundary

Review the General Information section of the SAR and any additional description of the containment

system presented in the Containment Evaluation section. All drawings, figures, and tables that describe

containment features should be consistent with the evaluation.

Verify that the SAR provides a complete description of the containment boundary, including, as

applicable, the containment vessel, welds, seals, lids, cover plates, valves, and other closure devices. The

containment boundary should be clearly depicted in a figure or sketch. Ensure that all components of the

containment boundary are shown in the drawings.
NUREG-1617 4-4

Confirm that the following information regarding components of the containment boundary is consistent

with that presented in the Structural Evaluation and Thermal Evaluation sections of the SAR:

• Materials of construction

• Welds

• Applicable codes and standards (e.g., ASME B&PV Division 3 Code specifications for the vessel)

Figure 4-1 SAR Information Flow for the Containment Review.

• Bolt torque required to maintain positive closure

• Maximum allowable temperatures of components, including seals
4-5 NUREG-1617

• Temperatures of components under normal conditions of transport and hypothetical accident

conditions.

Verify that all containment boundary penetrations and their method of closure are described in detail. 

Performance specifications for components such as valves and O-rings should be documented, and no

device may allow continuous venting. Any valve or similar device on the package must be protected

against unauthorized operation and must be provided with an enclosure to retain any leakage. Cover

plates and lids should be recessed or otherwise protected. Compliance with the permitted release limit

may not depend on filters.
NUREG-1617 4-6

Confirm that all closure devices can be leak tested. If fill, drain, or test ports utilize quick connect valves,

ensure that such valves do not preclude leakage testing of their cover-plate seals, providing such seals

form part of the containment boundary.

Determine that the seal material is compatible for its intended use and that no galvanic or chemical

reaction will occur between the seal and the packaging or its contents (NRC Bulletin 96-04). If

penetrations are closed with two seals (e.g., to enable leakage testing), verify which seal is defined as the

containment boundary. Ensure that seal grooves are appropriately sized. Verify that the temperature of

containment boundary seals will remain within their specified allowable limits under both normal conditions

of transport and hypothetical accident conditions.

Confirm that the containment system is securely closed by a positive fastening device that cannot be

opened unintentionally or by a pressure that may arise within the package.

4.5.1.2 Codes and Standards

Verify that the containment system is in full compliance with the B&PV Division 3 Code including

material, design, fabrication, examination, testing, inspection, and certification. This includes an agreement

with an Authorized Inspection Agency to provide inspection and audit services for the Design Owners,

Packaging Owners, and Class TP Certificate Holders. The SAR should justify the use of other codes, if

appropriate.

4.5.1.3 Special Requirements for Damaged Spent Nuclear Fuel

Review the condition and isotopic composition of the spent nuclear fuel proposed for the packaging. If

the contents include damaged fuel, verify that it is canned to facilitate handling and to confine gross fuel

particles to a known subcritical volume under normal conditions of transport and hypothetical accident

conditions. Ensure that appropriate material specifications and the design/fabrication criteria for the can

are justified in the SAR. These specifications and criteria should generally be the same as those for

containment or criticality support structures as discussed in Section 2 of this SRP. If a screen-type

container is used, an appropriate mesh size should be justified. Containment analysis for aluminum-based

spent fuel should be in accordance with WSRC-TR-98-00317.

Spent fuel, with more than 0.74 TBq (20 Ci) of plutonium per package, in the form of debris, particles,

loose pellets, or fragmented rods or assemblies, must be packaged in a separate inner container (second

containment system) in accordance with 10 CFR 71.63(b). Each containment system must separately

meet the requirements of 10 CFR 71.51(a)(1) under normal conditions of transport and 10 CFR

71.51(a)(2) under hypothetical accident conditions. Material specifications and design/fabrication criteria

for the inner container should be identical to those of the outer containment. In general, the inner

container should also meet all requirements of ANSI N14.5 unless otherwise justified in the SAR (e.g., for

periodic and pre-shipment leak testing). Review both containment systems as appropriate.

The determination of the fuel condition should be based, as a minimum, on review of fuel records. Fuel

which is known or suspected to be damaged should be visually inspected prior to loading. If the visual
4-7 NUREG-1617

inspection indicates no damage greater that a hairline crack or a pinhole leak, the fuel may be considered

as undamaged.

4.5.2 Containment under Normal Conditions of Transport

4.5.2.1 Pressurization of Containment Vessel

Verify that the maximum normal operating pressure is consistent with that determined in the Thermal

Evaluation section of the SAR. The pressure in the containment vessel should be based on the conditions

of the package under normal transport conditions, including temperature, release of gases and volatiles

from fuel rod cladding breaches, vaporization of contents, etc. (NRC IN 84-72).

4.5.2.2 Containment Criteria

Detailed guidance on procedures for determining the containment criteria is provided in NUREG/CR6487.

Confirm that the SNF contents are fully described, including fuel type, fuel amount, percent enrichment,

burnup, cool time, decay heat, etc. Confirm that the contents evaluated in the Containment Evaluation

section of the SAR are consistent with those presented in the General Information section of the SAR.

Verify that the SAR identifies the constituents which comprise the releasable source term, including

radioactive gases, volatiles, and powders. For SNF packages, the releasable source term is composed of

crud on the outside of the fuel rod cladding that can become aerosolized, and fuel fines, volatiles, and

gases that are released from a fuel rod in the event of a cladding breach. Although the residual

contamination on the inside surfaces of the packaging (from previous shipments) typically can be ignored

in the determination of the releasable source term, this issue should be addressed. Reasonable bounding

values for the effective surface activity density (Ci/cm2) of the crud on fuel rod clads are based on

experimental determinations. A computer code, such as ORIGEN2 (ORNL-CCC-371), is used to identify

the radionuclides present for a given percent fuel enrichment, burnup, and cool time. Using the individual

A2 values for the crud, fines, gases, and volatiles individually, the effective A2 of the releasable sourceterm mixture can be determined by using the relative release fraction for each contributor and the

methods from ANSI N14.5. The release fractions and effective specific activities for the various

releasable source term contributors for SNF with an initial enrichment of 3.2%, a burnup of 33,000

MWd/MTIHM, and a cool time of 5 years are given in Table 4-1. The release fractions presented in

Table 4-1 are considered bounding and have been developed from reasoned argument and experimental

data (NUREG/CR-6487). The SAR should justify release fractions and specific activities, as appropriate.

Based on the mass density, effective specific activity, and effective A2 of the releasable source term,

ensure that the maximum permissible release rate and the maximum permissible leakage rate are

calculated in accordance with the containment criteria specified in ANSI N14.5. Verify that the

maximum permissible leakage rate under normal transport conditions is converted into a reference air

leakage rate under standard leakage test conditions according to ANSI N14.5 and NUREG/CR-6487.
NUREG-1617 4-8

Table 4-1 Release Fractions and Specific Activities for the Contributors to the Releasable

Source Term for Packages Designed to Transport Irradiated Fuel Rods.1,2

PWR BWR

Normal

conditions of

transport

Hypothetical

accident

conditions

Normal

conditions of

transport

Hypothetical

accident

Variable conditions

Fraction of crud that spalls-off of rods, fC0.15 1.0 0.15 1.0

Crud surface activity, SC [Ci/cm2] 140 × 10-6140 × 10-61254 × 10-61254 × 10-6

Mass fraction of fuel that is released as fines

due to a cladding breach, fF

3 × 10-53 × 10-53 × 10-53 × 10-5

Specific activity of fuel rods, AR [Ci/g] 0.60 0.60 0.51 0.51

Fraction of rods that develop cladding

breaches, fB

0.03 1.0 0.03 1.0

Fraction of gases that are released due to a

cladding breach, fG

0.3 0.3 0.3 0.3

Specific activity of gases in a fuel rod, AG

[Ci/g] 

7.32 × 10-37.32 × 10-36.28 × 10-36.28 × 10-3

Specific activity of volatiles in a fuel rod, AV

[Ci/g]

0.1375 0.1375 0.1794 0.1794

Fraction of volatiles that are released due to a

cladding breach, fV

2 × 10-42 × 10-42 × 10-42 × 10-4

1. 3.2% initial enrichment, 33,000 MWd/MTIHM burnup, 5-year cooling.

2. Applicable only to undamaged fuel. Release fractions for damaged fuel should be justified in the SAR.

Verify that the following maximum permissible leakage rates are determined in accordance with

ANSI N14.5:

• Fabrication verification

• Periodic verification

• Maintenance verification

• Pre-shipment verification.
4-9 NUREG-1617

4.5.2.3 Compliance with Containment Criteria

Confirm that the SAR demonstrates that the package satisfies the containment requirements of

10 CFR 71.51(a)(1) for normal conditions of transport.

• If compliance is demonstrated by test, verify that the leakage rate of a package subjected to the tests

of 10 CFR 71.71 does not exceed the temperature- and pressure-corrected air leakage rate.

• If compliance is demonstrated by analysis, verify that the structural evaluation shows that the

containment boundary or closure bolts do not undergo any plastic deformation and that the materials

of the containment system (e.g., seals) do not exceed their maximum allowable temperature limits

when subjected to the conditions in 10 CFR 71.71.

Compliance with the leakage rates for fabrication and periodic verification is discussed in the Acceptance

Tests and Maintenance Program Review section of this SRP; compliance with the leakage rates for

assembly verification is discussed in the Operating Procedures Review section of this SRP.

4.5.3 Containment under Hypothetical Accident Conditions

The review procedures for containment under hypothetical accident conditions are analogous to those

listed in Section 4.5.2 above for normal conditions of transport. Differences relevant to hypothetical

accident conditions are noted below.

4.5.3.1 Pressurization of Containment Vessel

The pressure in the containment vessel should be based on the conditions of the package under

hypothetical accident conditions, including temperature, release of gases and volatiles from fuel rod

cladding breaches, vaporization of contents, etc. Verify that this pressure is consistent with that

determined in the Thermal Evaluation section of the SAR.

4.5.3.2 Containment Criteria

The releasable source term, maximum permissible release rate, maximum permissible leakage rate, and

conversion to the reference air leakage rate should be based on package conditions and the 10 CFR

Part 71 containment requirements under hypothetical accident conditions. Verify that the temperatures,

pressure, and physical conditions of the package (including the contents) are consistent with those

determined in the Structural Evaluation and Thermal Evaluation sections of the SAR.

Ensure that the reference air leakage rate calculated for hypothetical accident conditions is greater than

that determined in Section 4.5.2.2 for normal conditions of transport. In the rare event that this is not the

case, the containment criteria for the fabrication, periodic, and assembly verification tests should be based

on the hypothetical accident leakage, rather than normal conditions of transport.
NUREG-1617 4-10

The containment requirements of 10 CFR 71.51(a)(2) for hypothetical accident conditions shall be applied

individually for krypton-85 and the other radioactive materials. Krypton-85 shall not exceed 10 A2 in a

week. The remaining radioactive materials shall not exceed A2 in a week.
4-11 NUREG-1617

4.5.3.3 Compliance with Containment Criteria

Confirm that the SAR demonstrates that the package satisfies the containment requirements of

10 CFR 71.51(a)(2) for hypothetical accident conditions. Demonstration is similar to that discussed in

Section 4.5.2.3 above except that the package should be subjected to the tests of 10 CFR 71.73.

4.5.4 Appendix

The appendix may include a list of references, copies of any applicable references not generally available

to the reviewer, computer code descriptions, input and output files, test results, and other appropriate

supplemental information.

4.6 EVALUATION FINDINGS

The containment review should result in the following findings, as appropriate:

4.6.1 Description of Containment System

The staff has reviewed the description and evaluation of the containment system and concludes that: (1)

the SAR identifies established codes and standards for the containment system; (2) the package includes

a containment system securely closed by a positive fastening device that cannot be opened unintentionally

or by a pressure that may arise within the package; (3) the package is made of materials and construction

that assure that there will be no significant chemical, galvanic, or other reaction; (4) a package valve or

similar device, if present, is protected against unauthorized operation and, except for a pressure relief

valve, is provided with an enclosure to retain any leakage; (5) a package designed for the transport of

damaged SNF includes packaging of the damaged SNF in a separate inner container that meets the

requirements of 10 CFR 71.63(c).

4.6.2 Containment under Normal Conditions of Transport

The staff has reviewed the evaluation of the containment system under normal conditions of transport and

concludes that the package is designed, constructed, and prepared for shipment so that under the tests

specified in 10 CFR 71.71 (normal conditions of transport) the package satisfies the containment

requirements of 10 CFR 71.43(f) and 10 CFR 71.51(a)(1) for normal conditions of transport with no

dependence on filters or a mechanical cooling system.

4.6.3 Containment under Hypothetical Accident Conditions

The staff has reviewed the evaluation of the containment system under hypothetical accident conditions

and concludes that the package satisfies the containment requirements of 10 CFR 71.51(a)(2) for

hypothetical accident conditions, with no dependence on filters or a mechanical cooling system.

In summary, the staff has reviewed the Containment Evaluation section of the SAR and concludes that

the package has been described and evaluated to demonstrate that it satisfies the containment

requirements of 10 CFR Part 71, and that the package meets the containment criteria of ANSI N14.5.
NUREG-1617 4-12

4.7 REFERENCES

ANSI N14.5 Institute for Nuclear Materials Management, ANSI N14.5, “American

National Standard for Leakage Tests on Packages for Shipment of

Radioactive Materials,” New York, NY, 1997.

B&PV Division 3 Code American Society of Mechanical Engineers, “ASME Boiler and Pressure

Vessel Code, Section III, Division 3, Containment Systems and Transport

Packagings For Spent Nuclear Fuel and High Level Radioactive Waste,”

New York, NY, 1998.

NRC Bulletin 96-04 U.S. Nuclear Regulatory Commission, “Chemical, Galvanic, or Other

Reactions in Spent Fuel Storage and Transportation Casks,” OMB No.

3150-0011, Bulletin 96-04, U.S. Government Printing Office, Washington,

D.C., July 5, 1996.

NUREG/CR-6487 U.S. Nuclear Regulatory Commission, “Containment Analysis for Type

B Packages Used to Transport Various Contents,” NUREG/CR-6487,

U.S. Government Printing Office, Washington, D.C., 1996.

ORNL-CCC-371 Oak Ridge National Laboratory, “ORIGEN2.1: Isotope Generation and

Depletion Code-Matrix Exponential Method,” CCC-371, Oak Ridge, TN,

1991.

WSRC-TR-98-00317 Westinghouse Savannah River Company, “Bases for Containment

Analysis for Transportation of Aluminum-Based Spent Nuclear Fuel,”

WSRC-TR-98-00317, Westinghouse Savannah River Company, Aiken,

SC, October 1998.
5-1 NUREG-1617

5 SHIELDING REVIEW

5.1 REVIEW OBJECTIVE

The objective of this review is to verify that the package design satisfies the external radiation

requirements of 10 CFR Part 71 under normal conditions of transport and hypothetical accident

conditions.

5.2 AREAS OF REVIEW

The SAR must be reviewed for adequacy of the description and evaluation of the shielding design. Areas

of review include the following:

5.2.1 Description of the Shielding Design

5.2.1.1 Packaging Design Features

5.2.1.2 Codes and Standards

5.2.1.3 Summary Table of Maximum Radiation Levels

5.2.2 Source Specification

5.2.2.1 Gamma Source

5.2.2.2 Neutron Source

5.2.3 Model Specification

5.2.3.1 Configuration of Source and Shielding

5.2.3.2 Material Properties

5.2.4 Evaluation

5.2.4.1 Methods

5.2.4.2 Key Input and Output Data

5.2.4.3 Flux-to-Dose-Rate Conversion

5.2.4.4 Radiation Levels

5.2.5 Appendix

5.3 REGULATORY REQUIREMENTS

Regulatory requirements of 10 CFR Part 71 applicable to the shielding review are as follows:

5.3.1 Description of the Shielding Design

The packaging must be described in sufficient detail to provide an adequate basis for its evaluation. The

description of the shielding components must include dimensions, materials of construction, and materials

specifically used for neutron shielding. [10 CFR 71.31(a)(1) and 10 CFR 71.33(a)(5)]
NUREG-1617 5-2

The SAR must identify established codes and standards applicable to the shielding design.

[10 CFR 71.31(c)]
5-3 NUREG-1617

5.3.2 Source Specification

The contents must be described in sufficient detail to provide an adequate basis for their evaluation. This

description must include those radionuclides that result in the largest external radiation levels, and their

chemical and physical form. [10 CFR 71.31(a)(1), 10 CFR 71.33(b)(1), 10 CFR 71.33(b)(2), and

10 CFR 71.33(b)(3)]

5.3.3 Model Specification

The package must be described and evaluated to demonstrate that it satisfies the shielding requirements

of 10 CFR Part 71. [10 CFR 71.31(a) and 10 CFR 71.31(b)]

5.3.4 Evaluation

The package must be evaluated to demonstrate that it satisfies the shielding requirements specified in

10 CFR Part 71, Subpart E. [10 CFR 71.31(a)(2), 10 CFR 71.35(a), and 10 CFR 71.41(a)]

The package must be designed, constructed, and prepared for shipment so that the external surface

radiation levels will not significantly increase under the tests specified in 10 CFR 71.71 (normal conditions

of transport). [10 CFR 71.43(f) and 10 CFR 71.51(a)(1)]

Under the tests specified in 10 CFR 71.71 (normal conditions of transport), the external radiation levels

must satisfy the requirements of 10 CFR 71.47(b) for exclusive-use shipments. (10 CFR 71.47(a)

requirements for non-exclusive-use shipments are assumed not to apply to spent nuclear fuel (SNF)

packages.) The package and vehicle radiation limits for exclusive-use shipments are summarized in Table

5-1.

Under the tests specified in 10 CFR 71.73 (hypothetical accident conditions), the external radiation levels

at 1 m (40 in) from the package surface must not exceed 10 mSv/hr (1 rem/hr). [10 CFR 71.51(a)(2)]

5.4 ACCEPTANCE CRITERIA

5.4.1 Description of the Shielding Design

The regulatory requirements in Section 5.3.1 identify the acceptance criteria.

5.4.2 Source Specification

The regulatory requirements in Section 5.3.2 identify the acceptance criteria.

5.4.3 Model Specification

In addition to the regulatory requirements identified in Section 5.3.3, the model used in the shielding

evaluation should be described in sufficient detail to permit an independent review, with confirmatory

calculations, of the package shielding design.
NUREG-1617 5-4

Table 5-1 External Radiation Level Limits for Exclusive-Use Shipments

Package and Vehicle Radiation Level Limits (49 CFR 173.441, 10 CFR 71.47(b))

This table must not be used as a substitute for DOT or NRC regulations on transportation of radioactive

materials.

Transport Vehicle Use: Exclusive 

Transport Vehicle Type: Open (flat-bed) Open w/Enclosurea Closed

Package (or freight container) Limits: 

External Surface 2 mSv/hr

(200 mrem/hr)

10 mSv/hr

(1000 mrem/hr)

10 mSv/hr

(1000 mrem/hr)

Roadway or Railway Vehicle (or freight container) Limits:

Any point on the outer

surface

N/A N/A 2 mSv/hr

(200 mrem/hr)

Vertical planes

projected from outer

edges

2 mSv/hr

(200 mrem/hr)

2 mSv/hr

(200 mrem/hr)

N/A

Top of . . . load: 2 mSv/hr

(200 mrem/hr)

 enclosure: 2 mSv/hr

(200 mrem/hr)

 vehicle: 2 mSv/hr

(200 mrem/hr)

2 m (80 in) from. . . vertical planes:

0.1 mSv/hr

(10 mrem/hr)

vertical planes:

0.1 mSv/hr

(10 mrem/hr)

outer lateral surfaces:

0.1 mSv/hr

(10 mrem/hr)

Underside 2 mSv/hr (200 mrem/hr)

Occupied position 0.02 mSv/hr (2 mrem/hr)b

a. Securely attached (to vehicle), access-limiting enclosure; package personnel barriers are considered as

enclosures.

b. Does not apply to private carrier wearing dosimetry if under radiation protection program satisfying 10 CFR

Part 20.

5.4.4 Evaluation

In addition to the regulatory requirements identified in Section 5.3.4 and Table 5-1, shielding should not

exceed its allowable temperature limits under normal conditions of transport or hypothetical accident

conditions.
5-5 NUREG-1617

5.5 REVIEW PROCEDURES

The following procedures are generally applicable to the shielding review of SNF transportation packages. 

Since packages for shipment of SNF are generally intended to be shipped by exclusive-use, only

exclusive-use shipments are assumed in the following SRP review procedures.

The shielding review is based in part on the descriptions and evaluations presented in the General

Information, Structural Evaluation, and Thermal Evaluation sections of the SAR. Similarly, results of the

shielding review are considered in the review of the SAR sections on Operating Procedures and

Acceptance Tests and Maintenance Program. Examples of SAR information flow into, within, and from

the shielding review are shown in Figure 5-1.
NUREG-1617 5-6

Figure 5-1 SAR Information Flow for the Shielding Review.

5.5.1 Description of the Shielding Design

5.5.1.1 Packaging Design Features

Review the General Information chapter of the SAR and any additional description of the shielding design

presented in the Shielding Evaluation section. Design features important to safety include, but are not

limited to:

• Dimensions, tolerances, and densities of material for neutron and gamma shielding, including those of

structural or thermal components considered in the shielding evaluation

• Concentrations of neutron absorbers

• Structural components that maintain the contents in a fixed position within the package

• Dimensions of the conveyance that are considered in the shielding evaluation.

All information presented in the text, drawings, figures, and tables should be consistent with each other

and with that used in the shielding evaluation. Pay close attention to consistency between the drawings,

and the models and parameters used in the shielding analysis.

5.5.1.2 Codes and Standards

Verify that the established codes and standards used in the shielding design are identified. For example,

conversion of the flux to radiation levels should generally be based on ANSI 6.1.1-1977, as discussed in

Section 5.5.4.3.

5.5.1.3 Summary Table of Maximum Radiation Levels

Examine the summary table, and verify that the maximum radiation levels are within the limits of

10 CFR 71.47 and 10 CFR 71.51 for exclusive-use shipments for both normal conditions of transport and

hypothetical accident conditions (see Table 5-1). The fuel specifications (e.g., burnup, enrichment,

cooling time) at which the individual radiation levels apply should be given in the table, since the gamma

or neutron contributions could be greatest at different fuel specifications. Appendix B contains a fill-inthe-blank summary table of radiation levels for the reviewer’s use.

Examine the variation of radiation levels among the various locations for general consistency. For

example, radiation levels should decrease with increased distance from the source or greater shielding

effectiveness.

5.5.2 Source Specification
5-7 NUREG-1617

Compare the specifications for the SNF contents with those listed in the General Information section of

the SAR. The ranges of fuel type, burnup, enrichment, and cooling time should be stated. Generally, the

General Information section will specify a maximum fuel enrichment, which is important in the criticality

analysis. For shielding evaluations, however, the neutron source term increases considerably with

decreasing initial enrichment at a constant burnup. Consequently, the SAR should also specify a minimum

initial enrichment.

Generally, the applicant will use a computer code such as ORIGEN-S (NUREG/CR-0200A) [or a SAS2

sequence of SCALE], ORIGEN2 (Oak Ridge 1991), or the Department of Energy Characteristics Data

Base (TRW-CSCIID A00020002-AAX01.1) to determine the source terms. The latter two have energy

group structure limitations that are discussed below. If the applicant has chosen ORIGEN2, verify that

the cross-section library is appropriate for the fuel being considered. Many libraries are not appropriate

for burnup that exceeds 33,000 MWd/MTU.

5.5.2.1 Gamma Source

Verify that the gamma source terms are specified as a function of energy for both the SNF and activated

hardware. If the energy group structure of the source term calculation differs from that of the cross

section set of the shielding calculation, the applicant may need to regroup the photons. One method of

regrouping is to input the nuclide activities from the source term calculation to a simple decay code with a

variable group structure (e.g., GAMGEN [Gosnell 1990]). In general, only gammas from approximately

0.8 to 2.5 MeV will contribute significantly to the external radiation levels, so regrouping outside of this

range is of little consequence. Pay particular attention to whether the source terms are specified per

assembly, per total number of assemblies, or per metric ton, and ensure that the total source is correctly

used in the shielding calculation.

Determining the source terms for fuel assembly hardware is generally not as straightforward as that for

the SNF, especially if one of the ORIGEN codes is used. The activation of the hardware depends on the

impurities (e.g., 59Co) initially present and on the spatial and energy variation of the neutron flux during

burnup. The effort devoted to reviewing the calculation of the hardware source term should be

appropriate to its contribution to the radiation levels presented in the shielding analysis. Note also whether

the package is intended to transport other hardware such as control assemblies or shrouds, and ensure

that the source terms from these components are also included if applicable. Two reports that may be

helpful in reviewing the calculation of hardware activation are ORNL/TM-11018 and PNL-6906.

Depending on the packaging design, neutron interactions could result in the production of energetic

gammas near the packaging surface. If this source is not treated by the shielding analysis code, verify

that it is determined by other appropriate means.

The result of the source term calculation should be a listing of gammas per second, or MeV per second,

as a function of energy. The activity of each nuclide that contributes significantly to the source term

should be provided as supporting information.

Because the gamma radiation levels are directly proportional to the gamma source term, the review

should independently confirm the source term calculated by the applicant.
NUREG-1617 5-8

5.5.2.2 Neutron Source

Verify that the neutron source term is expressed as a function of energy. The neutron source will

generally result from both spontaneous fission of transuranics and from (a,n) reactions in the fuel. 

Depending on the methods used to calculate these source terms, the applicant might determine the energy

group structure independently. This is often accomplished by selecting the nuclide with the predominant

contribution to spontaneous fission (e.g., 244Cm) and using that spectrum for all neutrons, since the

contribution from (a,n) reactions is generally small. Assure that neutron multiplication in the fissile

material is included in the analysis. The fissile content assumed for the multiplication effect should be

justified and conservative.

The result of the source term calculation or experimental data should be a listing of neutrons per second

as a function of energy. For the spontaneous fission contribution, a listing of the significant nuclides

should also be presented.

Because the neutron radiation levels are directly proportional to the neutron source term, the review

should independently confirm the source term calculated by the applicant.

5.5.3 Model Specification

Review the Structural Evaluation and Thermal Evaluation sections of the SAR to determine the effects of

the tests for both normal conditions of transport and hypothetical accident conditions on the packaging and

its contents. For example, the package might have impact limiters or an external neutron shield that would

be damaged or destroyed during the structural and thermal tests of 10 CFR 71.73. Verify that the

corresponding models used in the shielding calculation are consistent with these effects.

5.5.3.1 Configuration of Source and Shielding

Examine the sketches or figures that indicate how the shielding is modeled. Verify that the model

dimensions and materials are consistent with those specified in the package drawings presented in the

General Information section of the SAR and the normal and accident conditions of the package. 

Dimensions should be at the conservative end of their tolerance range. Ensure that voids, streaming

paths, and irregular geometries are taken into account or otherwise modeled conservatively. Differences

between normal conditions of transport and hypothetical accident conditions should be clearly indicated.

Verify that the source term locations for both SNF and the structural support regions of the fuel

assemblies are modeled properly. Generally, at least three source regions (fuel and top/bottom assembly

hardware) are necessary. Within the SNF region, the fuel materials may generally be homogenized to

facilitate shielding calculations. In some cases, the basket material may be homogenized also. The

reviewer should watch for cases when homogenization is not appropriate, such as when it distorts the

neutron multiplication rate or when radiation streaming can occur between the basket components. 

Because of the burnup profile, a uniform source distribution is generally conservative for the top and

bottom dose points, but not for the axial center unless the source strength is appropriately adjusted. If

peaking appears to be significant, verify that it has been treated appropriately. The assembly structural
5-9 NUREG-1617

support regions (e.g., top/bottom end-pieces and plenum) should be correctly positioned relative to the

SNF. These support regions may be individually homogenized.

If transport is by exclusive use (as it typically is for SNF), dimensions of the transport vehicle should be

included, as appropriate (e.g., to determine the radiation level at 2 m from the vehicle). If the vehicle

occupants do not wear dosimetry devices under a radiation protection program in conformance with

10 CFR 20.1502, applicable vehicle dimensions will also be necessary to determine the radiation level at

normally occupied locations. (These dimensions or vehicle type, as well as positioning of the packages,

may become limiting conditions in the certificate of compliance for exclusive-use shipments.)

Verify that the dose point locations for the various calculations include all locations prescribed in

10 CFR 71.47(b) and 10 CFR 71.51(a)(2). Ensure that the dose points are chosen to identify the location

of the maximum radiation levels; the maximum might not occur at the midpoint of a package surface or

parallel plane. Radiation peaking often occurs near the edges of external neutron shields and impact

limiters. Determine that voids, streaming paths, and irregular geometries are included in the model or

otherwise treated in a conservative manner.

5.5.3.2 Material Properties

Verify that the mass densities and atom densities are provided for all materials used in the models of the

packaging, source, and conveyance (if applicable). Because most computer programs for shielding

calculations now allow input in either g/cm3 or atoms/barn-cm, the review may consider either mass or

atom densities alone to be sufficient for certain materials. Atom densities are subject to frequent error

and should be confirmed if used as input to shielding calculations. For uncommon materials, especially

foams, plastics, and other hydrocarbons, the source of the data should be referenced. Ensure that these

materials are properly controlled to achieve such densities. (Specific information on control measures

should be included in the Acceptance Tests and Maintenance Program section of the SAR.) Review

materials to assess if any shielding properties could degrade during the service life of the packaging. 

Confirm that controls are in place to ensure the long term effectiveness of the shielding, as appropriate.

Confirm that temperature-sensitive shielding materials will not be subject to temperatures at or above their

design limitations during either normal or accident conditions. Determine whether the applicant properly

examined the potential for shielding material to experience changes in material densities at temperature

extremes. (For example, elevated temperatures may reduce hydrogen content through loss of bound or

free water in hydrogenous shielding materials.)

As noted above, a common practice in shielding analyses is to homogenize the source region rather than

develop a detailed heterogeneous model of every fuel pin, pellet, or similar contents. Because an

accurate effective density of the homogenized source is important for self-shielding, a confirmatory

calculation of this density is generally warranted.

5.5.4 Evaluation

5.5.4.1 Methods
NUREG-1617 5-10

Verify that the computer program(s) used for the shielding analysis are appropriate. These codes may

use Monte Carlo transport, deterministic transport, or point-kernel techniques for problem solution. (The

latter is generally appropriate only for gammas since transportation packagings typically do not contain

sufficient hydrogenous material to apply removal cross sections for point-kernel neutron calculations.) 

Shielding codes that are typically used in SARs include, but are not limited to, TORT/DORT (ORNL6268), DANTSYS (LA-9184-M and LA-I 0049-M), MCNP (LA-12625-M), COG (Buck 1994), and

MORSE (NUREG/CR-0200B) or one of the SAS sequence codes in SCALE. For computer codes not

well established in the public domain, the SAR should describe the solution method, benchmark results,

validation procedures, and quality assurance practices.

Assess if the number of dimensions of the code is appropriate for the cask configuration. Generally, a 2-

D or 3-D calculation is necessary. One-dimensional codes provide little information about off-axis

locations and streaming paths. Even for radiation levels at the end of the package, 1-D codes require a

buckling correction that must be justified; merely using the packaging cavity diameter may underestimate

the radiation level (overestimate the radial leakage).

Verify that the cross section library used by the code is applicable for shielding calculations. Ensure that

a coupled cross section set is used and that the code has been executed in a manner that accounts for

secondary source terms, unless the evaluation has independently determined a source term for neutroninduced gamma radiation or subcritical multiplication of neutrons.

5.5.4.2 Key Input and Output Data

Verify that key input data for the shielding calculations are identified. These will depend on the type of

code (point kernel, deterministic, Monte Carlo, etc.) as well as the code itself. In addition to the source

terms, materials, and dimensions identified above, key input data can include convergence criteria, mesh

size, neutrons per generation, number of generations, etc. Verify that the information from the shielding

model is properly input into the code.

The SAR should also generally include a representative output file (or key sections of the file including

input data) for each type of calculation performed in the shielding analysis. The review should ensure that

proper convergence is achieved and that the calculated radiation levels from the output files agree with

those reported in the text.

5.5.4.3 Flux-to-Dose-Rate Conversion

The shielding analysis code will typically have the capability to perform this conversion directly with its

own data library or with one supplied by the user. Generally this conversion will use ANSI/ANS 6.1.1-

1977.

Verify the accuracy of the flux-to-dose-rate conversion factors, which should be tabulated as a function

of the energy group structure used in the shielding calculation.

5.5.4.4 Radiation Levels
5-11 NUREG-1617

Confirm that the radiation levels under both normal conditions of transport and hypothetical accident

conditions are in agreement with the summary tables in Section 5.5.1.3 above and that they satisfy the

limits in 10 CFR 71.47(b) and 10 CFR 71.51(a)(2). Verify that the analysis shows that the locations

selected are those of maximum radiation levels and include any radiation streaming paths.

For the purposes of 10 CFR 71.47(b), NRC staff considers the external surface to be that part of the

package which is shown in the drawings and has been demonstrated to remain in place under the tests in

10 CFR 71.71 (normal conditions of transport). Personnel barriers and similar devices that are attached

to the conveyance, rather than the package, can, however, qualify the vehicle as a closed vehicle

(NUREG/CR-5569A and NUREG/CR-5569B) as defined in 49 CFR 173.403.

Determine that the radiation levels appear reasonable and that their variation with location are consistent

with the geometry and shielding characteristics of the package.

Ensure that the evaluation addresses damage to the shielding under normal conditions of transport and

hypothetical accident conditions. Verify that any damage under normal conditions of transport

(10 CFR 71.71) does not result in a significant increase in external radiation levels, as required by

10 CFR 71.43(f) and 10 CFR 71.51(a)(1). Note that external neutron shielding may not be designed to

remain in place under hypothetical accident conditions.

The review may include a confirmatory analysis of the shielding calculations reported in the SAR. 

Because measurements of the actual radiation levels from packages must be performed prior to shipment

in order to show that the 10 CFR 71.47 limits are satisfied, a number of factors should be considered in

determining the level of effort of the confirmatory analysis. These factors include such items as the

expected magnitude of the radiation levels, similarity with previously reviewed packages, thoroughness of

the review of source terms and other input data, bounding assumptions in the analysis, margin from the

regulatory limits, and the contribution from difficult to measure neutrons.

At a minimum, the review should include examination of the applicant’s input to the computer program

used for the shielding analysis. Verify use of proper dimensions, material properties, and an appropriate

cross section set. In addition, independently evaluate the use of gamma and neutron source terms.

If a more detailed review is required, independently evaluate the radiation levels to ensure that the SAR

results are reasonable and conservative. As previously noted, the use of a simple code for neutron

calculations is often not appropriate. An extensive evaluation is necessary if major errors are suspected. 

To the degree possible, the use of a different shielding code with a different analytical technique and

cross section set from that of the SAR analysis will provide a more independent evaluation.

5.5.5 Appendix

The appendix may include a list of references, copies of applicable references if not generally available to

the reviewer, computer code descriptions, input and output files, test results, and other appropriate

supplemental information.

5.6 EVALUATION FINDINGS
NUREG-1617 5-12

The shielding review should result in the following findings, as appropriate:

5.6.1 Description of the Shielding Design

The staff has reviewed the package description and evaluation and concludes that they satisfy the

shielding requirements of 10 CFR Part 71.

5.6.2 Source Specification

The staff has reviewed the source specification used in the shielding evaluation and concludes that they

are sufficient to provide a basis for evaluation of the package against the 10 CFR Part 71 shielding

requirements.

5.6.3 Model Specification

The staff has reviewed the models used in the shielding evaluation and concludes that they are described

in sufficient detail to permit an independent review, with confirmatory calculations, of the package

shielding design.

5.6.4 Evaluation

The staff has reviewed the external radiation levels of the package and vehicle as it will be prepared for

shipment and concludes that they satisfy 10 CFR 71.47(b) for packages transported by exclusive-use

vehicle.

The staff has reviewed the package design, construction, and preparations for shipment and concludes

that the external radiation levels will not significantly increase during normal conditions of transport

consistent with the tests specified in 10 CFR 71.71.

The staff has reviewed the package design, construction, and preparations for shipment and concludes

that the maximum external radiation level at one meter from the external surface of the package will not

exceed 10 mSv/hr (1 rem/hr) during hypothetical accident conditions consistent with the tests specified in

10 CFR 71.73.

5.7 REFERENCES

ANSI/ANS 6.1.1-1977 American Nuclear Society, ANSI/ANS 6.1.1, “American National

Standard for Neutron and Gamma-Ray Flux to Dose Factors,” La

Grange Park, IL, 1977.

ANSI/ANS 6.1.1-1991 American Nuclear Society, ANSI/ANS 6.1.1, “American National

Standard for Neutron and Gamma-Ray Fluence to Dose Factors,” La

Grange Park, IL, 1991.
5-13 NUREG-1617

Buck 1994 Buck, R.M. et al., “COG-A Monte Carlo Neutron, Photon, Electron

Transport Code,” Lawrence Livermore National Laboratory, M-221-1,

Livermore, CA July 4, 1994.

60 FR 50247 U.S. Nuclear Regulatory Commission, “Compatibility With the

International Atomic Energy Agency (IAEA),” Federal Register, FR

50247, U.S. Government Printing Office, Washington, D.C., September

28, 1995.

Gosnell 1990 Gosnell, T.B., “Automated Calculation of Photon Source Emission From

Arbitrary Mixtures of Naturally Radioactive Heavy Nuclides,” in Editor,

Nuclear Instruments and Methods in Physics Research, A299 (1990),

Elsevier Science, Elmont, NY, pp. 682-686.

LA-I 0049-M Los Alamos National Laboratory, “User's Guide for TWODANT: A

Code Package for Two Dimensional, Diffusion Accelerated Neutral

Particle Transport,” LA-I 0049-M Rev., Los Alamos, NM, April 1992.

LA-9184-M Los Alamos National Laboratory, “Revised User's Manual for

ONEDANT: A Code for One Dimensional, Diffusion Accelerated

Neutral Particle Transport,” LA-9184-M, Rev., Los Alamos, NM,

December 1989.

LA-12625-M Los Alamos National Laboratory, “MCNP 4A, Monte Carlo N-Particle

Transport Code System,” LA-12625-M, Los Alamos, NM, December

1993.

NRC IN 80-32 U.S. Nuclear Regulatory Commission, “Clarification of Certain

Requirements for Exclusive-Use Shipments of Radioactive Materials,”

IE Information Notice 80-32, U.S. Government Printing Office,

Washington, D.C., August 29, 1980.

NUREG/CR-0200A U.S. Nuclear Regulatory Commission, “SCALE: A Modular Code

System for Performing Standardized Computer Analyses for Licensing

Evaluation,” NUREG/CR-0200, Vol. 2, Part 1, Rev. 4, U.S. Government

Printing Office, Washington, D.C., April 1995.

NUREG/CR-0200B U.S. Nuclear Regulatory Commission, “SCALE: A Modular Code

System for Performing Standardized Computer Analyses for Licensing

Evaluation,” NUREG/CR-0200, Vol. 2, Part 2, Rev. 4, U.S. Government

Printing Office, Washington, D.C., April 1995.

NUREG/CR-5569A U.S. Nuclear Regulatory Commission, “Clarification of Certain

Requirements for Exclusive-Use Shipments of Radioactive Materials,”

HPPOS-084, in Health Physics Positions Data Base, NUREG/CR-
NUREG-1617 5-14

5569, Rev. 1, U.S. Government Printing Office, Washington, D.C.,

February 1991.

NUREG/CR-5569B U.S. Nuclear Regulatory Commission, “Clarification of Certain

Requirements for Exclusive-Use Shipments,” HPPOS-085, in Health

Physics Positions Data Base, NUREG/CR-5569, Rev. 1, U.S.

Government Printing Office, Washington, D.C., February 1991.

ORNL-CCC-371 Oak Ridge National Laboratory, “ORIGEN2.1: Isotope Generation and

Depletion Code-Matrix Exponential Method,” CCC-371, Oak Ridge, TN,

1991.

ORNL-6268 Oak Ridge National Laboratory, “The TORT Three-Dimensional

Discrete Ordinates Neutron/Photon Transport Code,” ORNL-6268, Oak

Ridge, TN, November 1987.

ORNL/TM-11018 Ludwig, S.B., and Renier, J.P., “Standard- and Extended-Burnup PWR

and BWR Reactor Models for the ORIGEN2 Computer Code,”

ORNL/TM-11018, Oak Ridge National Laboratory, Oak Ridge, TN,

December 1989.

PNL-6906 Luksic, A., “Spent Fuel Assembly Hardware Characterization and

10 CFR 61 Classification for Waste Disposal,” PNL-6906, Volume 1,

Pacific Northwest Laboratory, Richland, WA, June 1989.

TRW- CSCIID TRW Environmental Safety Systems, Inc., “DOE Characteristics Data

A00020002-AAX01.0 Base, User Manual for the CDB_R,” CSCIID A00020002-AAX01.0,

Vienna, VA, November 16, 1992.
6-1 NUREG-1617

6 CRITICALITY REVIEW

6.1 REVIEW OBJECTIVE

The objective of this review is to verify that the package design satisfies the criticality safety requirements

of 10 CFR Part 71 under normal conditions of transport and hypothetical accident conditions.

6.2 AREAS OF REVIEW

The SAR should be reviewed for adequacy of the description and evaluation of the criticality design. 

Areas of review include the following:

6.2.1 Description of Criticality Design

6.2.1.1 Packaging Design Features

6.2.1.2 Codes and Standards

6.2.1.3 Summary Table of Criticality Evaluations

6.2.1.4 Transport Index

6.2.2 Spent Nuclear Fuel Contents

6.2.3 General Considerations for Criticality Evaluations

6.2.3.1 Model Configuration

6.2.3.2 Material Properties

6.2.3.3 Computer Codes and Cross Section Libraries

6.2.3.4 Demonstration of Maximum Reactivity

6.2.3.5 Confirmatory Analyses

6.2.4 Single Package Evaluation

6.2.4.1 Configuration

6.2.4.2 Results

6.2.5 Evaluation of Package Arrays under Normal Conditions of Transport

6.2.5.1 Configuration
NUREG-1617 6-2

6.2.5.2 Results

6.2.6 Evaluation of Package Arrays under Hypothetical Accident Conditions

6.2.6.1 Configuration

6.2.6.2 Results

6.2.7 Benchmark Evaluations

6.2.7.1 Experiments and Applicability

6.2.7.2 Bias Determination

6.2.8 Burnup Credit

6.2.8.1 Limits for the Licensing Basis

6.2.8.2 Code Validation

6.2.8.3 Licensing-Basis Model Assumptions

6.2.8.4 Loading Curve

6.2.8.5 Assigned Burnup Loading Value

6.2.8.6 Estimate of Additional Reactivity Margin

6.2.9 Appendix
6-3 NUREG-1617

6.3 REGULATORY REQUIREMENTS

Regulatory requirements of 10 CFR Part 71 applicable to the criticality review are as follows:

6.3.1 Description of Criticality Design

The packaging must be described in sufficient detail to provide an adequate basis for its evaluation. This

description must include types and dimensions of materials of construction and materials specifically used

as nonfissile neutron absorbers or moderators. [10 CFR 71.31(a)(1) and 10 CFR 71.33(a)(5)]

The SAR must identify established codes and standards applicable to the criticality design.

[10 CFR 71.31(c)]

The SAR must specify the allowable number of packages that may be transported in a single shipment.

[10 CFR 71.35(b)]

A fissile material package must be assigned a transport index for nuclear criticality control.

[10 CFR 71.59(b)]

6.3.2 Spent Nuclear Fuel Contents

The contents must be described in sufficient detail to provide an adequate basis for their evaluation. This

description must include the type, maximum quantity, and chemical and physical form of the spent nuclear

fuel (SNF). [10 CFR 71.31(a)(1), 10 CFR 71.33(b)(1), 10 CFR 71.33(b)(2), and 10 CFR 71.33(b)(3)]

Unknown properties of fissile material must be assumed to be those which will result in the highest

neutron multiplication. [10 CFR 71.83]

6.3.3 General Considerations for Criticality Evaluations

The package must be evaluated to demonstrate that it satisfies the criticality safety requirements of

10 CFR Part 71, Subpart E. [10 CFR 71.31(a)(2), 10 CFR 71.35(a), and 10 CFR 71.41(a)]

6.3.4 Single Package Evaluation

A single package must satisfy the specifications of 10 CFR 71.43(f), 10 CFR 71.51(a)(1), and

10 CFR 71.55(d) under normal conditions of transport. These requirements address subcriticality,

alteration of the geometric form of the contents, inleakage of water, and effectiveness of the packaging.

[10 CFR 71.35, 10 CFR 71.43(f), 10 CFR 71.51(a)(1), and 10 CFR 71.55(d)]

A single package must be designed and constructed and its contents limited so that it would be subcritical

if water were to leak into the containment system. [10 CFR 71.55(b)]

A single package must be subcritical under the tests for hypothetical accident conditions.

[10 CFR 71.55(e)]
NUREG-1617 6-4

6.3.5 Evaluation of Package Arrays under Normal Conditions of Transport

The SAR must evaluate arrays of packages under normal conditions of transport to determine the

maximum number of packages that may be transported in a single shipment. [10 CFR 71.35 and

10 CFR 71.59]

6.3.6 Evaluation of Package Arrays under Hypothetical Accident Conditions

The SAR must evaluate arrays of packages under hypothetical accident conditions to determine the

maximum number of packages that may be transported in a single shipment. [10 CFR 71.35 and

10 CFR 71.59]

6.3.7 Benchmark Evaluations

The package must be evaluated to demonstrate that it satisfies the criticality safety requirements of

10 CFR Part 71. [10 CFR 71.31(a)(2) and 10 CFR 71.35]

6.3.8 Burnup Credit

There are no regulatory requirements that are specific to burnup credit. The general criticality

requirements apply. However, based on experience, the staff has developed guidelines to facilitate the

review of burnup credit, when it is included in the analysis. Burnup credit evaluations are performed in

accordance with Sections 6.4.8.1 through 6.4.8.6.

6.4 ACCEPTANCE CRITERIA

6.4.1 Description of Criticality Design

The regulatory requirements in Section 6.3.1 identify the acceptance criteria.

6.4.2 Spent Nuclear Fuel Contents

The regulatory requirements in Section 6.3.2 identify the acceptance criteria.

6.4.3 General Considerations for Criticality Evaluations

In addition to the regulatory requirements identified in Section 6.3.3, the packaging model for the criticality

evaluation should generally consider no more than 75% of the specified minimum neutron poison

concentrations. The model for the SNF should include no burnable poisons. Methods for including fuel

burnup in the criticality calculations need to have prior approval by NRC.

The sum of the effective multiplication factor (keff), two standard deviations (95% confidence), and the

bias adjustment should not exceed 0.95 to demonstrate subcriticality by calculation. A bias that reduces

the calculated value of keff should not be applied.
6-5 NUREG-1617

6.4.4 Single Package Evaluation

In addition to the regulatory requirements identified in Section 6.3.4, the assumption of water inleakage for

the analysis pursuant to 10 CFR 71.55 (b) should consider the packaging and contents to be in their most

reactive condition, as determined by the tests in 10 CFR 71.71 and 10 CFR 71.73.

6.4.5 Evaluation of Package Arrays under Normal Conditions of Transport

The regulatory requirements in Section 6.3.5 identify the acceptance criteria.

6.4.6 Evaluation of Package Arrays under Hypothetical Accident Conditions

The regulatory requirements in Section 6.3.6 identify the acceptance criteria.

6.4.7 Benchmark Evaluations

The criticality evaluation of the package should include a comparison of the calculational methods with

applicable benchmark experiments to determine the appropriate bias and uncertainties.

6.4.8 Burnup Credit Evaluation

The staff guidance in Sections 6.4.8.1 through 6.4.8.6 identify the acceptance criteria.

6.4.8.1 Limits for the Licensing Basis

Verify that the licensing-basis analysis performed to demonstrate criticality safety limits the amount of

burnup credit to that available from actinide compositions associated with PWR irradiation of UO2 fuel to

an assembly-average burnup value of 40 GWd/MTU or less. This licensing-basis analysis should assume

an out-of-reactor cooling time of five years and should be restricted to intact assemblies that have not

used burnable absorbers. The initial enrichment of the fuel assumed for the licensing-basis analysis

should be no more than 4.0 wt% 235U unless a loading offset is applied. The loading offset is defined as

the minimum amount by which the assigned burnup loading value (see Section 6.4.8.5) must exceed the

burnup value used in the licensing safety basis analysis. The loading offset should be at least 1

GWd/MTU for every 0.1 wt% increase in initial enrichment above 4.0 wt%. In any case, the initial

enrichment shall not exceed 5.0 wt%. For example, if the applicant performs a safety analysis that

demonstrates an appropriate subcritical margin for 4.5 wt% fuel burned to the limit of 40 GWd/MTU, then

the loading curve (see Section 6.4.8.4) should be developed to ensure that the assigned burnup loading

value is at least 45 GWd/MTU (i.e., a 5 GWd/MTU loading offset resulting from the 0.5 wt% excess

enrichment over 4.0 wt%). Applicants requesting use of actinide compositions associated with fuel

assemblies, burnup values, or cooling times outside these specifications, or applicants requesting a

relaxation of the loading offset for initial enrichments between 4.0 and 5.0 wt%, should provide the
NUREG-1617 6-6

measurement data and/or justify extrapolation techniques necessary to adequately extend the isotopic

validation and quantify or bound the bias and uncertainty.

6.4.8.2. Code Validation

Ensure that the analysis methodologies used for predicting the actinide compositions and determining the

neutron multiplication factor (k-effective) are properly validated. Bias and uncertainties associated with

predicting the actinide compositions should be determined from benchmarks of applicable fuel assay

measurements. Bias and uncertainties associated with the calculation of k-effective should be derived

from benchmark experiments that represent important features of the cask design and spent fuel contents. 

The particular set of nuclides used to determine the k-effective value should be limited to that established

in the validation process. The bias and uncertainties should be applied in a way that ensures conservatism

in the licensing safety analysis. Particular consideration should be given to bias uncertainties arising from

the lack of critical experiments that are highly prototypical of spent fuel in a cask.

6.4.8.3 Licensing-Basis Model Assumptions.

Ensure that the actinide compositions used in analyzing the licensing safety basis (as described in 6.4.8.1)

are calculated using fuel design and in-reactor operating parameters selected to provide conservative

estimates of the k-effective value under cask conditions. The calculation of the k-effective value should

be performed using cask models, appropriate analysis assumptions, and code inputs that allow adequate

representation of the physics. Of particular concern should be the need to account for the axial and

horizontal variation of the burnup within a spent fuel assembly (e.g., the assumed axial burnup profiles),

the need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with

control rods fully or partly inserted, and the need for a k-effective model that accurately accounts for

local reactivity effects at the less-burned axial ends of the fuel region.

6.4.8.4 Loading Curve

Verify that the application includes one or more loading curves that plot, as a function of initial

enrichment, the assigned burnup loading value above which fuel assemblies may be loaded in the cask. 

Loading curves should be established based on a 5-year cooling time and only fuel cooled at least five

years should be loaded in a cask approved for burnup credit.

6.4.8.5 Assigned Burnup Loading Value

Verify that administrative procedures are adequately described to ensure that licensees load the cask with

fuel that is within the specifications of the approved contents. The administrative procedures should

include an assembly measurement that confirms the reactor record assembly burnup. The measurement

technique may be calibrated to the reactor records for a representative set of assemblies. For an

assembly reactor burnup record to be confirmed, the measurement should provide agreement within a 95

percent confidence interval based on the measurement uncertainty. The assembly burnup value to be
6-7 NUREG-1617

used for loading acceptance (termed the assigned burnup loading value) should be the confirmed reactor

record value as adjusted by reducing the record value by the combined uncertainties in the records and

the measurement. 

6.4.8.6 Estimate of Additional Reactivity Margin

Ensure that design-specific analyses are provided that estimate the additional reactivity margins available

from fission product and actinide nuclides not included in the licensing safety basis (as described in

Section 6.4.8.1). The analysis methods used for determining these estimated reactivity margins should be

verified using available experimental data (e.g., isotopic assay data) and computational benchmarks that

demonstrate the performance of the applicant’s methods in comparison with independent methods and

analyses. The Organization for Economic Cooperation and Development Nuclear Energy Agency’s

Working Group on Burnup Credit provides a source of computational benchmarks that may be

considered. The design-specific margins should be evaluated over the full range of initial enrichments and

burnups on the burnup credit loading curve(s). The resulting estimated margins should then be assessed

against estimates of: (a) any uncertainties not directly evaluated in the modeling or validation processes

for actinide-only burnup credit (e.g., k-effective validation uncertainties caused by a lack of critical

experiment benchmarks with either actinide compositions that match those in spent fuel or material

geometries that represent the most reactive ends of spent fuel in casks); and (b) any potential

nonconservatisms in the models for calculating the licensing-basis actinide inventories (e.g., any outlier

assemblies with higher-than-modeled reactivity caused by the use of control rod insertion during burnup).

6.5 REVIEW PROCEDURES

The following procedures are generally applicable to the criticality review of SNF transportation

packages. Since packages for shipment of SNF are generally intended to be shipped by exclusive-use,

only exclusive-use shipments are assumed in the following SRP review procedures.

The criticality review is based in part on the descriptions and evaluations presented in the General

Information, Structural Evaluation, and Thermal Evaluation sections of the SAR. Similarly, results of the

criticality review are considered in the review of the SAR sections on Operating Procedures and

Acceptance Tests and Maintenance Program. Examples of SAR information flow into, within, and from

the criticality review are shown in Figure 6-1.

6.5.1 Description of the Criticality Design

6.5.1.1 Packaging Design Features

Review the General Information section of the SAR and any additional description of the criticality design

presented in the Criticality Evaluation section. Packaging design features important for criticality include,

but are not limited to:
NUREG-1617 6-8

• Dimensions and tolerances of the containment system

• Dimensions, material composition, and tolerances of structural components (e.g., basket) that

maintain the SNF in a fixed position within the package or in a fixed position relative to neutron

absorbing material

• Dimensions, concentrations, tolerances, and location of neutron-absorbing and moderating materials,

including neutron poisons and shielding

• Dimensions and tolerances of any floodable voids, including flux traps, inside the packaging

• Dimensions and tolerances of the overall package that affect the physical separation of the SNF

contents in package arrays

• Information on control rod assemblies, shrouds, or other fuel assembly components included with

the SNF, as applicable to the criticality evaluation. All information presented in the text, drawings,

figures, and tables should be consistent with each other and with that used in the criticality

evaluation. The drawings are the authoritative source of dimensions, tolerances, and material

composition of components important to criticality safety.
6-9 NUREG-1617

Figure 6-1 SAR Information Flow for the Criticality Review.

6.5.1.2 Codes and Standards
NUREG-1617 6-10

Verify that the established codes and standards used in all aspects of the criticality design and evaluation

of the package are identified.

6.5.1.3 Summary Table of Criticality Evaluations

Review the summary table of the criticality evaluation, which should address the following cases, as

described in Sections 6.5.4 to 6.5.6 below:

• A single package, under the conditions of 10 CFR 71.55(b), (d), and (e)

• An array of 5N undamaged packages, under the conditions of 10 CFR 71.59(a)(1)

• An array of 2N damaged packages, under the conditions of 10 CFR 71.59(a)(2).

For each case, the table should at least include the maximum value of keff, the uncertainty, the bias, and the

number of packages evaluated in the arrays. The table should also show that the sum of the effective

multiplication factor (keff), two standard deviations (95% confidence), and the bias adjustment does not

exceed 0.95 for each case.

Confirm that the summary table illustrates that the package meets the above subcriticality criterion.

6.5.1.4 Transport Index

Based on the number of packages evaluated in the arrays, verify that the SAR determines the appropriate

value of N and calculates the criticality transport index correctly. Ensure that this transport index is

consistent with that reported in the General Information section of the SAR.

Confirm that the SAR identifies the maximum number of packages that can be transported in the same

exclusive-use vehicle. Ensure that this number is clearly distinguished from the value of N used in the

criticality evaluation.

6.5.2 Spent Nuclear Fuel Contents

Ensure that the specifications for the SNF used in the criticality evaluation are consistent with those in the

General Information section of the SAR. Any differences in the specifications should be clearly identified

and justified. Specifications relevant to the criticality evaluation include:

• Types of fuel assemblies or rods (e.g., BWR/PWR) and vendor/model as appropriate

• Dimensions of fuel (including any annular pellets), cladding, fuel-cladding gap, pitch, and rod length

• Number of rods per assembly and locations of guide tubes and burnable poisons (see Section

6.5.3.2)

• Materials and densities
6-11 NUREG-1617

• Active fuel length

• Enrichment (variation by rod if applicable) before irradiation

• Chemical and physical form

• Mass of initial heavy metal per assembly or rod

• Number of fuel assemblies or individual rods per package.

For BWR fuel assemblies, NRC staff does not currently allow any credit for burnup of the fissile

material or increase in actinide or fission product poisons during irradiation; therefore, the enrichment

should be that of the unirradiated fuel. However, burnup credit is allowed for PWR fuel assemblies. 

Section 6.4.8 provides guidance on applying burnup credit in the criticality safety analysis of PWR

spent fuel. If assemblies contain fuel with several enrichments, the evaluation should either assume the

maximum initial enrichment or demonstrate that another approach (e.g., average enrichment) is

bounding. Section 6.5.3.2 below discusses consideration of the depletion of burnable poisons.

Determine if the SAR includes any specifications regarding the condition of the SNF. Fuel rods that have

been removed from an assembly should be replaced with dummy rods that displace an equal amount of

water unless the criticality analyses consider the additional moderation resulting from their absence. 

(Because of the additional moderation, the contents with less fissile material might be more reactive). 

These specifications should be included as a condition of approval for the contents in the SER and

certificate of compliance.

In general, the package will be designed for numerous types of SNF. The description of the contents

should be sufficient to permit a detailed criticality evaluation of each type or to support a conclusion that

certain types are bounded by the evaluations performed. The SAR may include separate criticality

controls (e.g., number of assemblies, enrichment, transport index) for the various types of SNF evaluated. 

If the contents include damaged fuel, the maximum extent of damage should be specified and shown to be

bounded by the criticality analysis. The review procedures below should address the evaluation for each

contents as appropriate.

6.5.3 General Considerations for Criticality Evaluations

The considerations discussed below are applicable to the criticality evaluations of a single package, arrays

of packages under normal conditions of transport, and arrays under hypothetical accident conditions.

General guidance for preparing criticality evaluations of transportation packages is provided in

NUREG/CR-5661. Guidance for package arrays is provided in NUREG-1646.

6.5.3.1 Model Configuration
NUREG-1617 6-12

Examine the Structural Evaluation and Thermal Evaluation sections of the SAR to determine the effects

of the normal conditions of transport and hypothetical accident conditions on the packaging and its

contents. Verify that the models used in the criticality calculation are consistent with these effects.

Examine the sketches or figures of the model used for the criticality calculations. Verify that the

dimensions and materials are consistent with those in the drawings of the actual package. Differences

should be identified and justified. Within the specified tolerance range, dimensions should be selected to

result in the highest reactivity.

Verify that the SAR considers deviations from nominal design configurations. For example, the fuel

assemblies might not always be centered in each basket compartment, and the basket might not be

exactly centered in the package. In addition to a fully flooded package, the SAR should address

preferential flooding as appropriate. This includes flooding of the fuel-cladding gap and other regions

(e.g., flux traps) for which water density might not be uniform in a flooded package.

Determine whether the SAR includes a heterogeneous model of each fuel rod or homogenizes the entire

assembly. With current computational capability, homogenization should generally be avoided. If such

homogenization is used, the SAR must demonstrate that it is applied correctly or conservatively. As a

minimum, this demonstration should include calculation of the multiplication factor of one assembly and

several benchmark experiments (see Section 6.5.7) using both homogeneous and heterogeneous models.

6.5.3.2 Material Properties

Verify that the appropriate mass densities and atom densities are provided for all materials used in the

models of the packaging and contents. Material properties should be consistent with the condition of the

package under the tests of 10 CFR 71.71 and 10 CFR 71.73, and any differences between normal

conditions of transport and hypothetical accident conditions should be addressed. The sources of the data

on material properties should be referenced.

No more than 75% of the specified minimum neutron poison concentration of the packaging should

generally be considered in the criticality evaluation. In addition, because of differences in net reactivity

due to depletion of fissile material and burnable poisons, no credit should be taken for burnable poisons in

the fuel. Ensure that neutron absorbers and moderators (e.g., poisons and neutron shielding) are properly

controlled during fabrication to meet their specified properties. Such information should be discussed in

more detail in the Acceptance Tests and Maintenance Program section of the SAR. Additional guidance

on neutron poisons is provided in NUREG-1647.

Review materials to identify any criticality properties that could degrade during the service life of the

packaging. If appropriate, ensure that specific controls are in place to assure the effectiveness of the

packaging during its service life. Such information should also be discussed in more detail in the

Acceptance Tests and Maintenance Program or Operating Procedures sections of the SAR.

6.5.3.3 Computer Codes and Cross Section Libraries
6-13 NUREG-1617

Both Monte Carlo and deterministic computer codes may be used for criticality calculations. Monte Carlo

codes are generally better suited to analyzing three-dimensional geometry and, therefore, are more widely

used to evaluate SNF cask designs. Verify that the SAR uses an appropriate computer code for the

criticality evaluation. Commonly used codes such as SCALE/KENO (NUREG/CR-0200) and MCNP

(LA-12625-M) should be clearly referenced. KENO is a multigroup code that is part of the SCALE

sequence, while MCNP permits the use of continuous cross sections. Other codes should be described in

the SAR, and appropriate supplemental information should be provided.

Ensure that the criticality evaluations use an appropriate cross section library. If multigroup cross sections

are used, confirm that the neutron spectrum of the package has been appropriately considered for

collapsing the group structure and that the cross sections are properly processed to account for resonance

absorption and self-shielding. The use of KENO as part of the SCALE sequence will directly enable

such processing. Some cross section sets include data for fissile and fertile nuclides (based on a potential

scattering cross section, s p) that can be input by the user. If the applicant has used a stand-alone version

of KENO, ensure that potential scattering has been properly considered. Additional information

addressing cross section concerns is provided in an NRC information notice (NRC IN 91-26) and

NUREG/CR-6328.

In addition to cross section information, other key input data for the criticality calculations should be

identified. These include number of neutrons per generation, number of generations, convergence criteria,

mesh selection, etc., depending on the code used. The SAR should also include at least one

representative input file for a single package, undamaged array, and damaged array evaluation. Verify

that information regarding the model configuration, material properties, and cross sections is properly input

into the code.

At least one representative output file (or key sections) should generally also be included in the SAR.

Ensure that the calculation has properly converged and that the calculated multiplication factors from the

output files agree with those reported in the evaluation.

6.5.3.4 Demonstration of Maximum Reactivity

Verify that the SAR evaluates each type of SNF included as allowable contents or clearly demonstrates

that some types are bounded by other evaluations.

Ensure that the analysis determines the optimum combination internal moderation (within the package)

and interspersed moderation (between packages), as applicable. Confirm that preferential flooding of

different regions within the package is considered as appropriate. As noted in Section 6.5.2, the

maximum allowable amount of fissile material may not be the most reactive.

Verify that the analyses demonstrate the most reactive of the three cases listed in Section 6.5.1.3 above

(single package, array of undamaged packages, and array of damaged packages) for each of the different

types of SNF, as applicable. Assumptions and approximations should be clearly identified and justified.

Additional guidance on determining the most reactive configurations is presented in NUREG/CR-5661.
NUREG-1617 6-14

6.5.3.5 Confirmatory Analyses

The review should include a confirmatory analysis of the criticality calculations reported in the SAR. As

a minimum, the reviewer should perform an independent calculation of the most reactive case, as well as

sensitivity analyses to confirm that the most reactive case has been correctly identified. In deciding the

level of effort necessary to perform independent confirmatory calculations, the reviewer should consider

the following three factors: (1) the calculational method (computer code) used by the applicant, (2) the

degree of conservatism in the applicant’s assumptions and analyses, and (3) how large a margin exists

between the calculated result and the acceptance criterion of keff # 0.95. As with any design and review,

a small margin below the acceptance criterion and/or small degree of conservatism necessitate a more

extensive analysis. 

The reviewer should generally model the package independently and should use a different code and

cross section set from that used in the SAR. If the reported keff for the worst case is substantially lower

than the acceptance criterion of 0.95, a simple model known to produce very conservative results may be

all that is necessary for the independent calculations. A review is not expected to validate the applicant’s

calculations but should assure that the regulations and acceptance criteria are met.

When the value of keff is highly sensitive to small variations in design features, contents specifications, or

the effects of the hypothetical accident conditions, the reviewer should confirm that such variations are

appropriately considered.

6.5.4 Single Package Evaluation

6.5.4.1 Configuration

Ensure that the criticality evaluation demonstrates that a single package is subcritical under both normal

conditions of transport and hypothetical accident conditions. The evaluations should consider:

• SNF in its most reactive credible configuration consistent with the condition of package and the

chemical and physical form of the contents

• Water moderation to the most reactive credible extent, including water inleakage into the

containment system as specified in 10 CFR 71.55(b)

• Full water reflection on all sides of the package, including close reflection of the containment

system or reflection by the package materials, whichever is more reactive, as specified in

10 CFR 71.55(b)(3).

6.5.4.2 Results

Confirm that the results of the criticality calculations are consistent with the information presented in the

summary table discussed in Section 6.5.1.3.
6-15 NUREG-1617

Verify also that the package meets the additional specifications of 10 CFR 71.55(d)(2) through

10 CFR 71.55(d)(4) under normal conditions of transport. These requirements address subcriticality,

alteration of the geometric form of the contents, inleakage of water, and effectiveness of the packaging.

6.5.5 Evaluation of Package Arrays under Normal Conditions of Transport

6.5.5.1 Configuration

Ensure that the criticality evaluation demonstrates that an array of 5N packages is subcritical under

normal conditions of transport. The evaluation should consider:

• The most reactive configuration of the array, e.g., pitch, package orientation, etc., with nothing

(including moderator) between the packages

• The most reactive credible configuration of the packaging and its contents. (Because water does

not leak into a spent-fuel package under normal conditions of transport, water inleakage need

not be assumed.)

• Full water reflection on all sides of the array (unless the array is infinite).

6.5.5.2 Results

Verify that the most reactive array conditions are clearly identified and that the results of the analysis are

consistent with the information presented in the summary table discussed in Section 6.5.1.3 above.

Confirm that the appropriate N value is used in determination of the transport index. The appropriate N

should be the smaller value which assures subcriticality for 5N packages under normal conditions of

transport or 2N packages under hypothetical accident conditions, as discussed in the next section.

6.5.6 Evaluation of Package Arrays under Hypothetical Accident Conditions

6.5.6.1 Configuration

Ensure that the criticality evaluation demonstrates that an array of 2N packages is subcritical under

hypothetical accident conditions. The evaluation should consider:

• The most reactive configuration of the array, e.g., pitch, package orientation, etc.

• Optimum interspersed hydrogenous moderation (between packages)

• The most reactive credible configuration of the packaging and its contents, including inleakage of

water and internal moderation

• Full water reflection on all sides of the array (unless the array is infinite).
NUREG-1617 6-16

6.5.6.2 Results

Verify that the most reactive array conditions are clearly identified and that the results of the analysis are

consistent with the information presented in the summary table discussed in Section 6.5.1.3 above.

Confirm that the appropriate N value is used in determining the transport index. The appropriate N should

be the smaller value which assures subcriticality for 2N packages under hypothetical accident conditions

or 5N packages under normal conditions of transport, as discussed in the previous section.

6.5.7 Benchmark Evaluations

Ensure that the computer codes for criticality calculations are benchmarked against critical experiments. 

Verify that the analysis of the benchmark experiments used the same computer code, hardware, and

cross section library as those used to calculate the multiplication factor for the package evaluations. The

calculated keff of the cask should then be adjusted to include the appropriate biases and uncertainties from

the benchmark calculations.

Additional information on benchmarking criticality evaluations for SNF is provided in NUREG/CR-6361.

6.5.7.1 Experiments and Applicability

Review the general description of the benchmark experiments and confirm that they are appropriately

referenced.

The applicant should justify and the reviewer should verify that the benchmark experiments are applicable

to the actual package design. The benchmark experiments should have, to the maximum extent possible,

the same materials, neutron spectrum, and configuration as the package evaluations. Key package

parameters that should be compared with those of the benchmark experiments include type of fissile

material, enrichment, H/U ratio (dependent largely on rod pitch and diameter), poisoning, reflector

material, and configuration. Confirm that differences between the package and benchmarks are

discussed and properly considered.

In addition, the SAR should address the overall quality of the benchmark experiments and the

uncertainties in experimental data (e.g., mass, density, dimensions, etc.). Ensure that these uncertainties

are treated in a conservative manner, i.e., they result in a lower calculated multiplication factor for the

benchmark experiment.

6.5.7.2 Bias Determination

Examine the results of the calculations for the benchmark experiments and the method used to account

for biases, including the contribution from uncertainties in experimental data.

Assess that a sufficient number of appropriate benchmark experiments are analyzed and that the results

of these benchmark calculations are used to determine an appropriate bias for the package calculations. 

The applicant should check benchmark comparisons for trends in the bias with respect to parameter
6-17 NUREG-1617

variations (such as pitch-to-rod-diameter ratio, assembly separation, reflector material, neutron absorber

material, etc.). Verify that only negative biases are considered, with positive bias results (values which

decrease keff when applied) treated as zero bias.

Statistical and convergence uncertainties of both benchmark and package calculations should also be

addressed. The uncertainties should be applied to at least the 95-percent confidence level. As a general

rule, if the acceptability of the result depends on these rather small differences, reviewers should question

the overall degree of conservatism of the calculations. Considering the current availability of computer

resources, a sufficient number of neutron histories can readily be used so that the treatment of these

uncertainties should not significantly affect the results.

6.5.8 Burnup Credit

Review the burnup credit analysis to determine compliance with the staff guidance outlined in Sections

6.4.8.1 through 6.4.8.6. The guidance provides a design-specific basis for granting burnup credit, based

on actinide composition.

The staff’s guidance for burnup credit considerations are based on investigations which have been

performed, both within the United States and by other countries, in an effort to understand and document

the related phenomena. The staff will issue additional guidance, as necessary, when more information is

obtained from its research program on burnup credit and as experience is gained through future licensing

activities.

6.5.9 Appendix

The appendix may include a list of references, copies of applicable references if not generally available to

the reviewer, computer code descriptions, input and output files, test results, and any other appropriate

supplemental information.

6.6 EVALUATION FINDINGS

The criticality review should result in the following findings, as appropriate:

6.6.1 Description of Criticality Design

The staff has reviewed the description of the packaging design and concludes that it provides an adequate

basis for the criticality evaluation.

The staff has reviewed the summary information of the criticality design and concludes that it indicates

the package is in compliance with the requirements of 10 CFR Part 71.
NUREG-1617 6-18

6.6.2 Spent Nuclear Fuel Contents

The staff has reviewed the description of the SNF contents and concludes that it provides an adequate

basis for the criticality evaluation.

6.6.3 General Considerations for Criticality Evaluations

The staff has reviewed the criticality description and evaluation of the package and concludes that it

addresses the criticality safety requirements of 10 CFR Part 71.

6.6.4 Single Package Evaluation

The staff has reviewed the criticality evaluation of a single package and concludes that it is subcritical

under the most reactive credible conditions.

6.6.5 Evaluation of Package Arrays under Normal Conditions of Transport

The staff has reviewed the criticality evaluation of the most reactive array of 5N packages and concludes

that it is subcritical under normal conditions of transport.

6.6.6 Evaluation of Package Arrays under Hypothetical Accident Conditions

The staff has reviewed the criticality evaluation of the most reactive array of 2N packages and concludes

that it is subcritical under hypothetical accident conditions.

6.6.7 Benchmark Evaluations

The staff has reviewed the benchmark evaluation of the calculations and concludes that they are

sufficient to determine an appropriate bias and uncertainties for the criticality evaluation of the package.

6.6.8 Burnup Credit

The staff has reviewed the criticality evaluation for granting burnup credit and concludes that the

associated fuel loading curve is appropriate.

6.7 REFERENCES

LA-12625-M Los Alamos National Laboratory, “MCNP 4A, Monte

Carlo N-Particle Transport Code System,” LA-12625-

M, Los Alamos, NM, December 1993.
6-19 NUREG-1617

NRC IN 91-26 U.S. Nuclear Regulatory Commission, “Potential

Nonconservative Errors in the Working Format HansenRoach Cross-Section Set Provided with the KENO and

SCALE Codes,” Information Notice 91-26, U.S.

Government Printing Office, Washington, D.C., April 15,

1991.

NUREG-1646 U.S. Nuclear Regulatory Commission, “Criticality

Analysis of Transportation-Package Arrays,” NUREG1646, U.S. Government Printing Office, Washington,

D.C., January 1999.

NUREG-1647 U.S. Nuclear Regulatory Commission, “Use of Neutron

Poisons for Criticality Control in Transportation

Packages,” NUREG-1647, U.S. Government Printing

Office, Washington, D.C., January 1999.

NUREG/CR-0200 U.S. Nuclear Regulatory Commission, “SCALE: A

Modular Code System for Performing Standardized

Computer Analyses for Licensing Evaluation,”

NUREG/CR-0200, Vol. 2, Part 2, Rev. 4, U.S.

Government Printing Office, Washington, D.C., April

1995.

NUREG/CR-5661 U.S. Nuclear Regulatory Commission,

“Recommendations for Preparing the Criticality Safety

Evaluation of Transportation Packages,” NUREG/CR5661, U.S. Government Printing Office, Washington,

D.C., April 1997.

NUREG/CR-6328 U.S. Nuclear Regulatory Commission, “Adequacy of the

123-Group Cross-Section Library for Criticality Analyses

of Water-Moderated Uranium Systems,” NUREG/CR6328, U.S. Government Printing Office, Washington,

D.C., August 1995.

NUREG/CR-6361 U.S. Nuclear Regulatory Commission, “Criticality

Benchmark Guide for Light-Water-Reactor Fuel in

Transportation and Storage Packages,” NUREG/CR6361, U.S. Government Printing Office, Washington,

D.C., March 1997.
NUREG-1617 6-20
7-1 NUREG-1617

7 OPERATING PROCEDURES REVIEW

7.1 REVIEW OBJECTIVE

The objective of this review is to verify that the operating procedures comply with the requirements of

10 CFR 71 and ensure that the package will be operated in a manner consistent with the conditions

assumed in its evaluation for approval.

7.2 AREAS OF REVIEW

The SAR should be reviewed for adequacy of the operating procedures description. Areas of review

include the following:

7.2.1 Package Loading

7.2.1.1 Preparation for Loading

7.2.1.2 Loading of Contents

7.2.1.3 Preparation for Transport

7.2.2 Package Unloading

7.2.2.1 Receipt of Package from Carrier

7.2.2.2 Preparation for Unloading

7.2.2.3 Removal of Contents

7.2.3 Preparation of Empty Package for Transport

7.2.4 Other Procedures

7.2.5 Appendix

7.3 REGULATORY REQUIREMENTS

Regulatory requirements of 10 CFR Part 71 applicable to package operations and the operating

procedures review are as follows:

7.3.1 Package Loading

The SAR must identify established codes and standards applicable to the operating procedures.

[10 CFR 71.31(c)]

The SAR for a fissile material shipment must include any proposed special controls and precautions for

transport, loading, unloading, and handling and any proposed special controls in case of accident or delay.

[10 CFR 71.35(c)] 

Packages must be prepared for transport so that in still air at 38EC (100EF) and in the shade, no

accessible surface of a package would have a temperature exceeding 85EC (185EF) in an exclusive-use
NUREG-1617 7-2

shipment. [10 CFR 71.43(g)] (Temperature limits for non exclusive-use shipments are assumed not to

apply to spent nuclear fuel (SNF) packages.)

Packages which require exclusive use shipment because of external radiation levels must be controlled by

providing written instructions to the carrier. [10 CFR 71.47(b), 10 CFR 71.47(c), and 10 CFR 71.47(d)]

Before each shipment, the package must be verified to be proper for the contents to be shipped.

[10 CFR 71.87(a)]

Before each shipment, the package must be verified to be in unimpaired physical condition.

[10 CFR 71.87(b)]

Before each shipment, each closure device of the package, including any specified gasket, must be

verified to be properly installed and secured and free of defects. [10 CFR 71.87(c)]

Before each shipment, any system for containing liquid must be verified to be adequately sealed and to

have adequate space or other specified provision for expansion of the liquid. [10 CFR 71.87(d)]

Before each shipment of licensed material any pressure relief device must be verified to be operable and

properly set. [10 CFR 71.87(e)]

Before each shipment, it must be determined that the package has been loaded and closed appropriately.

[10 CFR 71.87(f)]

Before each shipment of fissile material, it must be determined that any moderator or neutron absorber, if

specified, is present and in proper condition. [10 CFR 71.87(g)]

Before each shipment, any structural part of the package that could be used to lift or to tie-down the

package during transport must be rendered inoperable for that purpose unless it satisfies the design

requirements of 10 CFR 71.45. [10 CFR 71.87(h)]

Before each shipment, the level of non-fixed (removable) radioactive contamination on the external

surfaces of each package offered for shipment must be as low as is reasonably achievable (ALARA),

and within the limits specified in DOT regulation 49 CFR 173.443. [10 CFR 71.87(i)]

External radiation levels around the package and around the vehicle, if applicable, will not exceed the

limits specified in 10 CFR 71.47 at any time during transportation. [10 CFR 71.87(j)]

Accessible package surface temperatures will not exceed the limits specified in 10 CFR 71.43(g) at any

time during transportation. [10 CFR 71.87(k)]

Before delivery of a package to a carrier for transport, the licensee must send or make available any

special instructions needed to safely open the package to the consignee for the consignee’s use in

accordance with 10 CFR 20.1906 (e). [10 CFR 71.89]
7-3 NUREG-1617

7.3.2 Package Unloading

The application for a fissile material shipment must include provisions for complying with 10 CFR 20.1906

and any proposed special controls and precautions for unloading and handling. [10 CFR 71.35(c) and

10 CFR 71.89] 

7.3.3 Preparation of Empty Package for Transport

Before each shipment, perform the necessary inspections and tests to ensure that the level of non-fixed

(removable) radioactive contamination on the external surfaces of each package offered for shipment is

ALARA, and within the limits specified in DOT regulation 49 CFR 173.443. [10 CFR 71.87(i)]

7.3.4 Other Procedures

The application for a fissile material shipment must include any proposed special controls and precautions

for transport, loading, unloading, and handling and any proposed special controls in case of accident or

delay. [10 CFR 71.35(c)] 

7.4 ACCEPTANCE CRITERIA

The operating procedures should be presented and discussed sequentially in the actual order of

performance.

7.4.1 Package Loading

In addition to the regulatory requirements identified in Section 7.3.1, leakage testing of the package should

meet the assembly verification leakage test requirements specified in ANSI N14.5.

7.4.2 Package Unloading

The regulatory requirements in Section 7.3.2 identify the acceptance criteria.

7.4.3 Preparation of Empty Package for Transport

In addition to the regulatory requirements identified in Section 7.3.3, the interior of the packaging should

be properly decontaminated and closed in accordance with the requirements of 49 CFR 173.428.

7.4.4 Other Procedures

In addition to the regulatory requirements identified in Section 7.3.4, the package should be properly

closed and delivered to the carrier in such a condition that subsequent transport will not reduce the

effectiveness of the packaging.
NUREG-1617 7-4

7.5 REVIEW PROCEDURES

The following procedures are generally applicable to the operating procedures review of all SNF

transportation packages. Since packages for shipment of SNF are generally intended to be shipped by

exclusive-use, only exclusive-use shipments are assumed in the following SRP review procedures.

The operating procedures review is based in part on the descriptions and evaluations presented in the

General Information, Structural Evaluation, Thermal Evaluation, Containment Evaluation, Shielding

Evaluation, and Criticality Evaluation sections of the SAR. Examples of SAR information flow into and

within the operating procedures review are shown in Figure 7-1.
7-5 NUREG-1617

Figure 7-1 SAR Information Flow for the Operating Procedures Review.

The operating procedures presented in the SAR should not be expected to be detailed procedures that

could be implemented without expansion. Rather, the operating procedures should be an outline that

focuses upon those steps that are important to assuring that the package is operated in a manner

consistent with its evaluation for approval. Detailed procedures not important to safety should not be

included in the SAR. Procedural steps should normally be presented in sequential order, as applicable. 

Information on both the detailed procedures and the brief procedures included with an application can be

found in NUREG/CR-4775.
NUREG-1617 7-6

7.5.1 Package Loading

7.5.1.1 Preparation for Loading

Review the procedures presented sequentially in the order of performance for loading and preparing the

package for transport. At a minimum, the procedures should ensure that:

• The package is loaded and closed in accordance with written instructions

• The contents are authorized by the certificate of compliance, including the use of a secondary

container or containment as appropriate

• The use of the package complies with the conditions of approval in the certificate of compliance,

including verification that required maintenance has been performed

• The package is in unimpaired physical condition

• Any proposed special controls and precautions for handling the package are provided.

7.5.1.2 Loading of Contents

Review the procedures presented sequentially in the order of performance for loading the package

contents for transport. At a minimum, the procedures should ensure that:

• Special handling equipment needed for loading and unloading is provided

• Any proposed special controls and precautions for loading and handling the package are provided

• Any moderator or neutron absorber, if specified, is present and in proper condition

• The package has been loaded and closed appropriately in accordance with the specified bolt torques

and bolt-tightening sequences

• Methods to drain and dry the cask are described, the effectiveness of the proposed methods is

discussed, and vacuum drying criteria are specified

• Each closure device of the package, including any specified gaskets, is properly installed and secured

and free of defects.

7.5.1.3 Preparation for Transport

Review the procedures presented sequentially in the order of performance for preparing the package for

transport. At a minimum, the procedures should ensure that:
7-7 NUREG-1617

• The level of non-fixed (removable) radioactive contamination on the external surfaces of each

package offered for shipment is ALARA, and within the limits specified in DOT regulation

49 CFR 173.443

• Radiation survey requirements of the package exterior are described to ensure that limits specified in

10 CFR 71.47 are met

• The temperature survey requirements of the package exterior ensure that limits specified in

10 CFR 71.43(g) are implemented

• Leakage testing of the package meets the assembly verification leakage test requirements specified in

ANSI N14.5

• A tamper indicating device is incorporated which, while intact, indicates that the package has not been

opened by unauthorized persons

• Any system for containing liquid is adequately sealed and has adequate space or other specified

provision for expansion of the liquid

• A check is made to ensure that any pressure relief device is operable and properly set

• Any structural part of the package that could be used to lift or to tie-down the package during

transport is rendered inoperable for that purpose unless it satisfies the design requirements of

10 CFR 71.45

• Any proposed special controls and precautions for transport, handling and any proposed special

controls in case of accident or delay are specified

• Written instructions to the carrier are provided for packages which require exclusive use shipment

because of external radiation levels [10 CFR 71.47(b), 10 CFR 71.47(c), and 10 CFR 71.47(d)]

• Before delivery of a package to a carrier for transport, the licensee has sent or made available to the

consignee any special instructions needed to safely open the package, in accordance with

10 CFR 20.1906(e).

7.5.2 Package Unloading

In general, the unloading procedures are the reverse of the loading procedures. If applicable, procedures

for special controls and precautions to ensure safe removal of fission gases, contaminated coolant, and

solid contaminants should be presented and discussed.

7.5.2.1 Receipt of Package from Carrier 
NUREG-1617 7-8

Review the procedures presented sequentially in the order of performance for receiving the package from

the carrier. At a minimum, the procedures should ensure that:

• The requirements of 10 CFR 20.1906 are met

• The package is examined for visible external damage

• Steps to define actions to be taken when the tamper indicating device is not intact, or surface

contamination or radiation survey levels are too high are provided

• A list of any special handling equipment needed for unloading and handling the package is provided

• Any proposed special controls and precautions for unloading and handling the package are provided.

7.5.2.2 Preparation for Unloading

Review the procedures presented sequentially in the order of performance for preparing the package for

unloading following receipt. At a minimum, the procedures should ensure that:

• Procedures controlling the radiation level limits on unloading operations are provided

• Procedures for the safe removal of, if any, fission gases, contaminated coolants, and solid

contaminants are provided.

7.5.2.3 Removal of Contents

Review the procedures presented sequentially in the order of performance for removing the contents

following package receipt. At a minimum, the procedures should ensure that:

• The closure is removed appropriately

• The contents are removed appropriately

• A verification is made that the contents are completely removed.

7.5.3 Preparation of Empty Package for Transport

Review the procedures presented sequentially in the order of performance for preparing an empty

package for transport. At a minimum, the procedures should ensure that:

• The packaging is empty

• Appropriate inspections and tests of the package are performed before transport, to ensure that the

requirements of 10 CFR 71.87(i) are met
7-9 NUREG-1617

• Special preparations of the packaging, to ensure that the interior of the packaging is properly

decontaminated and closed in accordance with the requirements of 49 CFR 173.428, are described.

7.5.4 Other Procedures

Other procedures, as appropriate, should be included.

7.5.5 Appendix

The appendix may include a list of references, copies of any applicable references not generally available

to the reviewer, test results, and any other appropriate supplemental information.

7.6 EVALUATION FINDINGS

The operating procedures review should conclude that the applicant’s SAR presents acceptable operating

sequences, guidance, and generic procedures for key operations and result in the following findings, as

appropriate:

7.6.1 Package Loading

The staff has reviewed the proposed special controls and precautions for transport, loading, and handling

and any proposed special controls in case of accident or delay, and concludes that they satisfy

10 CFR 71.35(c).

The staff has reviewed the description of the radiation survey requirements of the package exterior and

concludes that the limits specified in 10 CFR 71.47 will be met.

The staff has reviewed the description of the temperature survey requirements of the package exterior

and concludes that the limits specified in 10 CFR 71.43(g) will be met.

The staff has reviewed the description of the routine determinations for package use prior to transport,

and concludes that the requirements of 10 CFR 71.87 will be met.

The staff has reviewed the description of the special instructions (if applicable) needed to safely open a

package and concludes that the procedures for providing the special instruction to the consignee are in

accordance with the requirements of 10 CFR 71.89.

7.6.2 Package Unloading

The staff has reviewed the proposed special controls and precautions for unloading and handling and

concludes that they satisfy 10 CFR 71.35(c).

7.6.3 Preparation of Empty Package for Transport
NUREG-1617 7-10

The staff has reviewed the description of the routine determinations for package use prior to transport,

and concludes that the requirements of 10 CFR 71.87 will be met.

7.6.4 Other Procedures

The staff has reviewed all other applicable proposed special controls and precautions for transport,

loading, unloading, and handling and concludes that they satisfy 10 CFR 71.35(c).

7.7 REFERENCES

ANSI N14.5 Institute for Nuclear Materials Management, ANSI N14.5, “American

National Standard for Leakage Tests on Packages for Shipment of

Radioactive Materials,” New York, NY, 1987.

NUREG/CR-4775 U.S. Nuclear Regulatory Commission, “Guide for Preparing Operating

Procedures for Shipping Packages,” NUREG/CR-4775, U.S. Government

Printing Office, Washington, D.C., December 1988.
8-1 NUREG-1617

8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW

8.1 REVIEW OBJECTIVE

The objectives of this review are to verify that the acceptance tests for the packaging comply with the

requirements of 10 CFR Part 71 for the package design and that a maintenance program will ensure

acceptable packaging performance throughout its service life.

8.2 ACCEPTANCE TESTS

8.2.1 Areas of Review

The SAR should be reviewed for adequacy of the description of the acceptance tests to be performed on

the packaging. Areas of review include the following:

8.2.1.1 Visual Inspections and Measurements

8.2.1.2 Weld Inspections

8.2.1.3 Structural and Pressure Tests

8.2.1.4 Leakage Tests

8.2.1.5 Component Tests

8.2.1.6 Shielding Tests

8.2.1.7 Neutron Absorber Tests

8.2.1.8 Thermal Tests

8.2.1.9 Appendix

8.2.2 Regulatory Requirements

Regulatory requirements of 10 CFR Part 71 applicable to the acceptance tests review are as follows:

The SAR should identify established codes, standards, and specific provisions of the quality assurance

program that are applicable to the acceptance tests to be performed on the packaging. [10 CFR 71.31(c)

and 10 CFR 71.37(b)]

Before first use, the fabrication of each packaging must be verified to be in accordance with the approved

design. [10 CFR 71.85(c)]

Before first use, each packaging must be inspected for cracks, pinholes, uncontrolled voids, or other

defects that could significantly reduce its effectiveness. [10 CFR 71.85(a)]

Before first use, if the maximum normal operating pressure (MNOP) of a package exceeds 35 kPa (5

lbf/in2) gauge, the containment system of each packaging must be tested at an internal pressure at least 50

percent higher than MNOP to verify its capability to maintain structural integrity at that pressure.

[10 CFR 71.85(b)]
NUREG-1617 8-2

Before first use, if applicable, the amount and the distribution of the neutron absorbing materials or

moderators must be verified to meet the design specification. [10 CFR 71.87(g)]

Before first use, each packaging must be conspicuously and durably marked with its model number, serial

number, gross weight, and a package identification number assigned by NRC. [10 CFR 71.85(c)]

The licensee must perform any tests deemed appropriate by NRC. [10 CFR 71.93(b)]

8.2.3 Acceptance Criteria

In addition to the regulatory requirements identified in Section 8.2.2, the SAR should discuss the package

tests to be performed and the acceptance criteria to demonstrate structural, leakage, shielding, and heat

transfer performance. Fabrication, welding, and examination of components are acceptable when

performed in accordance with the recommended sections and subsections of the ASME Boiler and

Pressure Vessel (B&PV) Code given in Section 1.5.2.6, Table 1-1 and Table 1-2 of this SRP. Leakage

testing of the packaging should be accomplished in accordance with ANSI N14.5. Fabrication,

examination, and acceptance testing of lifting trunnions should be conducted in accordance with NUREG0612, ANSI N14.6, or other appropriate specification.

8.2.4 Review Procedures

The following procedures are generally applicable to the acceptance tests review of all spent nuclear fuel

(SNF) packages.

The acceptance tests review is based in part on the descriptions and evaluations presented in the General

Information, Structural Evaluation, Thermal Evaluation, Containment Evaluation, Shielding Evaluation,

Criticality Evaluation, and Operating Procedures sections of the SAR and follows the sequence

established to evaluate the packaging against applicable 10 CFR Part 71 requirements. Examples of SAR

information flow into and within the acceptance tests review are shown in Figure 8-1.

The commitments specified in the Acceptance Tests and Maintenance Program section of the SAR are

often incorporated by reference into the certificate of compliance as conditions of package approval. 

Verify that the following tests, as applicable, are performed prior to the first use of the packaging.

Information presented on each test should include, as a minimum, a description of the test, the test

procedure, and the acceptance criteria. Confirm that the established codes, standards, and specific

provisions of the quality assurance program used in all aspects of the testing of the packaging are

identified.

Additional guidance on acceptance tests is provided in NUREG/CR-3854.

8.2.4.1 Visual Inspections and Measurements
8-3 NUREG-1617

Ensure that visual inspections are performed to verify that the packaging has been fabricated and

assembled in accordance with drawings and other requirements specified in the SAR. Dimensions and

tolerances specified on the drawings should be confirmed by measurement.
NUREG-1617 8-4

Figure 8-1 SAR Information Flow for the Acceptance Tests Review.

8.2.4.2 Weld Inspections

Verify that weld inspections are performed to verify fabrication in accordance with the drawings, codes,

and standards specified in the SAR to control weld quality. Location, type, and size of the welds should

be confirmed by measurement. Other specifications for weld performance, inspection, and acceptance

should be verified as appropriate.

Additional guidance on welding criteria is provided in NUREG/CR-3019.

8.2.4.3 Structural and Pressure Tests

Verify that the structural or pressure tests are identified and described. Such tests should comply with

10 CFR 71.85(b), as well as applicable codes or standards specified in the SAR. Structural testing of

lifting trunnions should be conducted in accordance with NUREG-0612, ANSI N14.6, or other appropriate

specification.
8-5 NUREG-1617

8.2.4.4 Leakage Tests

Verify that the containment system of the packaging is subjected to the fabrication leakage tests specified

in ANSI N14.5. The acceptable leakage criterion should be consistent with that identified in the

Containment Evaluation section of the SAR.

8.2.4.5 Component Tests

Confirm that tests and acceptance criteria for other components are specified as appropriate. Such

components include valves, rupture disks, seals, etc.

8.2.4.6 Shielding Tests

Ensure that shielding tests are specified for gamma and neutron radiation, as appropriate. The tests and

acceptance criteria should be sufficient to assure no voids or streaming paths exist in the shielding.

8.2.4.7 Neutron Absorber Tests

Verify that appropriate tests are specified to verify the amount and distribution meeting the minimum

specification of neutron absorbing material described in the SAR.

8.2.4.8 Thermal Tests

Verify that appropriate tests are specified to demonstrate the heat transfer capability of the packaging. 

These tests should confirm the heat transfer properties predicted in the Thermal Evaluation section of the

SAR.

8.2.4.9 Appendix

The appendix may include a list of references, copies of any applicable references not generally available

to the reviewer, and any other appropriate supplemental information.

8.2.5 Evaluation Findings

The acceptance tests review should result in the following findings, as appropriate:

The staff has reviewed the identification of the codes, standards, and provisions of the quality assurance

program applicable to the package design and concludes that the requirements specified in

10 CFR 71.31(c) and 10 CFR 71.37 (b) will be met.

The staff has reviewed the description of the preliminary determinations for the package prior to first use

and concludes that the requirements of 10 CFR 71.85 and 10 CFR 71.87(g) will be met.
NUREG-1617 8-6

8.3 MAINTENANCE PROGRAM

8.3.1 Areas of Review

The SAR should be reviewed for adequacy of the description of the maintenance program to be

performed on the packaging. Areas of review include the following:

8.3.1.1 Structural and Pressure Tests

8.3.1.2 Leakage Tests

8.3.1.3 Component Tests

8.3.1.4 Neutron Absorber Tests

8.3.1.5 Thermal Tests

8.3.1.6 Appendix

8.3.2 Regulatory Requirements

Regulatory requirements of 10 CFR Part 71 applicable to the maintenance program review are as follows:

The SAR should identify established codes, standards, and specific provisions of the quality assurance

program that are applicable to the proper maintenance of the packaging. [10 CFR 71.31(c) and

10 CFR 71.37(b)]

The maintenance program should ensure that the packaging is maintained in unimpaired physical condition

except for superficial defects such as marks or dents. [10 CFR 71.87(b)]

The presence of a moderator or neutron absorber should be verified to be in proper condition prior to each

shipment. [10 CFR 71.87(g)]

The licensee must perform any tests deemed appropriate by NRC. [10 CFR 71.93(b)]

8.3.3 Acceptance Criteria

In addition to the regulatory requirements identified in Section 8.3.2, the maintenance program should

include periodic testing requirements, inspections, and replacement criteria and schedules for

replacements and repairs of components on an as-needed basis.

8.3.4 Review Procedures

The following procedures are generally applicable to the maintenance program review of all SNF

packages.

The maintenance program review is based in part on the descriptions and evaluations presented in the

General Information, Structural Evaluation, Thermal Evaluation, Containment Evaluation, Shielding

Evaluation, Criticality Evaluation, and Operating Procedures sections of the SAR and follows the

sequence established to evaluate the packaging against applicable 10 CFR Part 71 requirements. 
8-7 NUREG-1617

Examples of SAR information flow into and within the maintenance program review are shown in

Figure 8-2.

Figure 8-2 SAR Information Flow for the Maintenance Program Review.

The commitments specified in the Acceptance Tests and Maintenance Program section of the SAR are

often incorporated by reference into the certificate of compliance as conditions of package approval.

The maintenance program should be adequate to assure that packaging effectiveness is maintained

throughout its service life. Verify that the following maintenance tests and inspections are described with

schedules and criteria for minor refurbishment and replacement of parts, as applicable. Confirm that the

established codes, standards, and specific provisions of the quality assurance program used in all aspects

of the maintenance of the packaging are identified.

8.3.4.1 Structural and Pressure Tests
NUREG-1617 8-8

Verify that any structural or pressure tests are identified and described. Such tests would generally be

applicable to codes, standards, or other procedures specified in the SAR. Structural testing of lifting

trunnions should be conducted in accordance with NUREG-0612, ANSI N14.6, or other appropriate

specification.

8.3.4.2 Leakage Tests

Verify that the containment system of the packaging is subjected to the periodic leakage tests specified in

ANSI N14.5. The acceptable leakage criterion should be consistent with that identified in the

Containment Evaluation section of the SAR.

8.3.4.3 Component Tests

Confirm that periodic tests and replacement schedules for components are described as appropriate. 

Such components include valves, rupture disks, and seals. Elastomeric seals should be replaced at an

interval not to exceed one year. Metal seals should be replaced after each use.

8.3.4.4 Neutron Absorber Tests

Verify that the SAR identifies any process that could result in deterioration of neutron absorbing material

and that appropriate tests to ensure packaging effectiveness are specified.

8.3.4.5 Thermal Tests

Appropriate periodic tests should be performed to verify the heat transfer capability of the packaging

during its service life. Tests similar to the acceptance tests may be applicable. The typical interval for

periodic thermal tests is five years.

8.3.4.6 Appendix

The appendix may include a list of references, copies of any applicable references not generally available

to the reviewer, and any other appropriate supplemental information.

8.3.5 Evaluation Findings

The maintenance program review should result in the following findings, as appropriate:

The staff has reviewed the identification of the codes, standards, and provisions of the quality assurance

program applicable to maintenance of the packaging and concludes that the requirements specified in

10 CFR 71.31(c) and 10 CFR 71.37 (b) will be met.

The staff has reviewed the description of the routine determinations for package use prior to transport and

concludes that the requirements of 10 CFR 71.87(b) and 10 CFR 71.87(g) will be met.

8.4 REFERENCES
8-9 NUREG-1617

ANSI N14.5 Institute for Nuclear Materials Management, ANSI N14.5, “Leakage Tests

on Packages for Shipment of Radioactive Materials,” New York, NY, 1987.

ANSI N14.6 Institute for Nuclear Materials Management, ANSI N14.6, “Special Lifting

Devices for Shipping Containers Weighing 10,000 Pounds (45000 kg) or

More for Nuclear Materials,” New York, NY, 1993.

B&PV Code American Society of Mechanical Engineers, “ASME Boiler and Pressure

Vessel Code,” New York, NY, 1998.

NUREG-0612 U.S. Nuclear Regulatory Commission, “Control of Heavy Loads at Nuclear

Power Plants,” NUREG-0612, National Technical Information Service,

Springfield, VA, July 1980.

NUREG/CR-3019 U.S. Nuclear Regulatory Commission, “Recommended Welding Criteria for

Use in the Fabrication of Shipping Containers for Radioactive Materials,”

NUREG/CR-3019, U.S. Government Printing Office, Washington, D.C.,

March 1984.

NUREG/CR-3854 U.S. Nuclear Regulatory Commission, “Fabrication Criteria for Shipping

Containers,” NUREG/CR-3854, U.S. Government Printing Office,

Washington, D.C., March 1985.
A-1 NUREG-1617

APPENDIX A S STANDARD REVIEW PLAN CORRELATION WITH 10 CFR PART 71

AND REGULATORY GUIDE 7.9

The following table summarizes the correlation of the SRP review procedure sections with the appropriate

sections of 10 CFR Part 71 and RG 7.9.

Table A-1 Standard Review Plan Correlation with 10 CFR Part 71 and Regulatory Guide 7.9.

SRP Review

Procedure Section

10 CFR Part 71 Section RG 7.9 Section

1.5.1 None Introduction

1.5.2 71.13, 71.31(a)(1), 71.31(a)(2), 71.31(a)(3), 71.31(b),

71.31(c), 71.33(a)(1), 71.33(a)(3), 71.35(b), 71.37,

71.38, 71.59, 71.107(c)

1.1

1.5.3 71.31(a)(1), 71.33(a)(2), 71.33(a)(4), 71.33(a)(5),

71.33(a)(6), 71.33(b), 71.43(b)

1.2

1.5.4 71.31(a)(2), 71.35(a), 71.41(a) None

1.5.5 None 1.3

2.5.1 71.31(a)(1), 71.31(c), 71.33 2.1, 2.2

2.5.2 71.43(d) 2.3, 2.4

2.5.3 71.45 2.5

2.5.4 71.31(a)(2), 71.35(a), 71.41(a), 71.61, 71.71, 71.73 2.6, 2.7

2.5.5 71.35(a), 71.41(a), 71.43(f), 71.51(a)(1), 71.55(d)(4),

71.71

2.6

2.5.6 71.35(a), 71.41(a), 71.73 2.7

2.5.7 71.61 None

2.5.8 71.85(b) None

2.5.9 None 2.10

3.5.1 71.31(a)(1), 71.31(c), 71.33(a)(5), 71.33(a)(6),

71.33(b)(1), 71.33(b)(3), 71.33(b)(5), 71.33(b)(7),

71.33(b)(8), 71.51(c)

3.1

3.5.2 71.31(a)(1), 71.33(a)(5) 3.2, 3.3

3.5.3 71.31(a)(2), 71.35(a), 71.41(a) None
SRP Review

Procedure Section

10 CFR Part 71 Section RG 7.9 Section

NUREG-1617 A-2

3.5.4 71.43(g) None

3.5.5 71.43(f), 71.51(a)(1), 71.71 3.4

3.5.6 71.73 3.5

3.5.7 None 3.6

4.5.1 71.31(a)(1), 71.31(c), 71.33(a)(4), 71.33(a)(5),

71.33(b)(1), 71.33(b)(3), 71.33(b)(5), 71.33(b)(7),

71.43(c), 71.43(d), 71.43(e)

4.1

4.5.2 71.31(a)(2), 71.35(a), 71.41(a), 71.43(f), 71.43(h),

71.51(a)(1), 71.51(c)

4.2

4.5.3 71.31(a)(2), 71.35(a), 71.41(a), 71.51(a)(2), 71.51(c) 4.3

4.5.4 None 4.5

5.5.1 71.31(a)(1), 71.31(c), 71.33(a)(5) 5.1

5.5.2 71.31(a)(1), 71.33(b)(1), 71.33(b)(2), 71.33(b)(3) 5.2

5.5.3 71.31(a), 71.31(b) 5.3

5.5.4 71.31(a)(2), 71.35(a), 71.41(a), 71.43(f), 71.47(b),

71.51(a)(1), 71.51(a)(2)

5.4

5.5.5 None 5.5

6.5.1 71.31(a)(1), 71.31(c), 71.33(a)(5), 71.35(b), 71.59(b) 6.1

6.5.2 71.31(a)(1), 71.33(b)(1), 71.33(b)(2), 71.33(b)(3),

71.83

6.2

6.5.3 71.31(a)(2), 71.35(a), 71.41(a) 6.3

6.5.4 71.35, 71.43(f), 71.51(a)(1), 71.55(b), 71.55(d),

71.55(e)

6.4

6.5.5 71.35, 71.59 6.4

6.5.6 71.35, 71.59 6.4

6.5.7 71.31(a)(2), 71.35 6.5

6.5.8 None 6.6
SRP Review

Procedure Section

10 CFR Part 71 Section RG 7.9 Section

A-3 NUREG-1617

7.5.1 71.31(c), 71.35(c), 71.43(g), 71.47(b), 71.47(c),

71.47(d), 71.87, 71.89

7.1

7.5.2 71.35(c) 7.2

7.5.3 71.87(i) 7.3

7.5.4 71.35(c) None

7.5.5 None 7.4

8.2.4 71.31(c), 71.37(b), 71.85(a), 71.85(b), 71.85(c),

71.87(g), 71.93(b)

8.1 

8.3.4 71.31(c), 71.37(b), 71.87(b), 71.87(g), 71.93(b) 8.2
B-1 NUREG-1617

APPENDIX B S TABLE OF EXTERNAL DOSE RATES FOR EXCLUSIVE-USE

SHIPMENTS

The following table summarizes the information that should be provided by the applicant for the external

dose rates for transportation packages for spent nuclear fuel. 

Table B-1 External Dose Rates for Packages (Exclusive-Use Shipment).

Normal Conditions of Transport

Package Surfacea

Radiation Top Side Bottom

 Gammab

 Neutronc

 Total

10 CFR 71.47(b)(1) Limitd 2 (200) 2 (200) 2 (200)

Package Surfacee

Radiation Top Side Bottom

 Gammab

 Neutronc

 Total

10 CFR 71.47(b)(1)(i-iii) Limitd 10 (1000) 10 (1000) 10 (1000)

Vehicle Outer Surfacef

Radiation Top Side Bottom

 Gammab

 Neutronc

 Total

10 CFR 71.47(b)(2) Limitd 2 (200) 2 (200) 2 (200)
NUREG-1617 B-2

Table B.1 (Cont.) External Dose Rates for Packages (Exclusive-Use Shipment).

Normal Conditions of Transport

2 Meters from Vehicle Outer Surfaceg

Radiation Top Side Bottom

 Gammab

 Neutronc

 Total

10 CFR 71.47(b)(3) Limitd 0.1 (10) 0.1 (10) 0.1 (10)

Normally Occupied Positions in Vehicle h

Radiation Top Side Bottom

 Gammab

 Neutronc

 Total

10 CFR 71.47(b)(4) Limitd 0.02 (2) 0.02 (2) 0.02 (2)

Hypothetical Accident Conditions

1 Meter from Surface of Package

Radiation Top Side Bottom

 Gammab

 Neutronc

 Total

10 CFR 71.51(a)(2) Limitd 10 (1000) 10 (1000) 10 (1000)

a External surface of package.

b Gamma dose rate based on ______ MWd burnup, ___ % 235U enrichment, and ___ years cooling time.

c Neutron dose rate based on ______ MWd burnup, ___ % 235U enrichment, and ___ years cooling time.

d Dose rate in mSv/h (mrem/h).

e External surface of package provided that shipment is in a closed vehicle, package position within vehicle remains

fixed during transport, and no loading or unloading operations occur en route.

f At any point on the outer surface of the vehicle, including the upper and lower surfaces; or in a non-closed vehicle, at

any point on the vertical planes projected from the outer edges of the vehicle, on the upper surface of the load or

enclosure (if applicable), and on the lower external surface of the vehicle.

g At any point 2 meters (80 inches) from the outer lateral surface of the vehicle (excluding the top and underside of the

vehicle); or in the case of a non-closed vehicle, at any point 2 meters (6.6 feet) from the vertical planes projected by

the outer edges of the vehicle (excluding the top and underside of the vehicle).
B-3 NUREG-1617

h

In any normally occupied space, except that this provision does not apply to private carriers if exposed personnel

under their control wear radiation dosimetry devices in conformance with 10 CFR 20.1502.
NUREG-1617 B-4