ML20236S682

Standard Review Plan for Transportation Packages for Spent Nuclear Fuel.Draft Report for Comment

ML20236S682
Person / Time
Issue date:	03/31/1998
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
References
NUREG-1617, NUREG-1617-DRF-FC, NUREG-1617-DRFT, NUDOCS 9807270168
Download: ML20236S682 (123)

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l NUREG-1617 l

l Standard Review Plan for Transportation Packages for Spent Nuclear Fuel l

l l Draft Report for Comment l

1 Manuscript Completed: March 1998 l

Date Published: March 1998 1

I I

l Spert Fuel Project Omce Omce of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Wcshington, DC 20555-0001 l

5

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 nucleac 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 package approval, (2) describe the procedures by which NRC staff determines that these requirements have been satisfied, and (3) document the practices developed by the staff in previous reviews of package applications.

Comments are solicited on this draft document and will be considered and incorporated into the SRP, as appropriate. Appendix C to this NUREG contains a data form that will be used to aid NRC staffin transcribing the comments. A photocopy of the form in Appendix C or a similar form containing the same information should be used.

Comments regan"ag errors or omissions, and suggestions for improvement should be sent to the Chief, Rules Review and Directives Branch, Division of Freedom ofinformation and Publication Services, Mail Stop T-6-D-59, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001.

(

l 1

)

DRAFT iii NUREG-1617 i

l l

[

j

CONTENTS AB S TRA CT........................................................................... iii ACRONYMS AND ABBREVIATIONS.......................

.................. xii GLOS S ARY.............,........................................................... xiv INTRODUCTION..................................................................... 1 1 GENERAL INFORMATION REVIEW...................

..........................1-1 1.1 REVIEW OBJECTIVE......................................................... 1 - 1 1.2 AREA S 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

..........................................11 1.3.1 General SAR Format.............................

....................11 1.3.2 Package Design Information..........................

........... 1-2 t

1.3.3 Package Description.................................................... 1 2 1.3.4 Compliance with 10'CFR Part 71............................................ 1 2 1.4 ' ACCEPTANCE CRITERIA.........

.........................................1-2 1.4.1 General SAR Format...

.............................................1-2 1.4.2 Package Design Information............................................ 1 2 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 S AR Fonnat...................................................... 1-3 1.5.2 Package Design Information

....... 1 -4 1.5.3 Package Description........ o........................................1-6 1.5.4 Compliance with 10 CFR Part 71.............................

....... 1-11 1.5.5 Appendix...........

.. 1-12 1.6 EVALUATION FINDINGS.................................................... 1 12 1.7 REFEREN CES............................................................ 1-13 2 STRUCTURAL REVIEW...............

2-1

' 2.1 REVIEW OBJECTIVE.........................................................

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

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

2.2.2 Material Properties.................................................... 2 1 2.2.3 LiAmg and Tie-down Standards for All Packages............................... 2 1 2.2.4 General Considerations for Structural Evaluation of Packagmg.................... 21 2.2.5 Normal Conditions of Transport....................................

......21 2.2.6 Hypothetical Accident Conditions........................................... 2 1 2.2.7 Special Reqmrement for Irradiated Nuclear Fuel Shipments

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

.......22 DRAFT v

NUREG 1617 r

2.2.9 Appendix....................................................

2-2 i

2.3 REGULATORY REQUIREMENTS..............................................

2-2 l

2.3.1 Desenption of Structural Design.......................................... 2-2 2.3.2 Material Properties.............

2-2 2.3.3 Lthmg 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............................................. 2-2 2.3.6 Hypodesical Accident Conditions........................................... 2 3 2.3.7 Special Reqmrement for Irradiated Nuclear Fuel Shipmants..................... 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 LiAing and Tie-Down Standards for All Packages.............................. 2-3 2.4.4 General Considerations for Structural Evaluation of Packaging 2-3 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.................................................. 2-6 2.5.3 LiAing and Tie-Down Standards for All Packages............................ 2-7 2.5.4 General Considerations for Structural Evaluation of Packaging...................

2-8' 2.5.5 Normal Conditions of Transport

...............................2-10 2.5.6 Hypothetical Accident Conditions................................... 2-12 2.5.7 Special Requirement for Irradiated Nuclear Fuel Shipments................... 2-14 2.5.8 Internal Pressure Test.....................

.................... 2-14 2.5.9 Appendix..............................................

.... 2 14 2.6 EVALUATION FINDINGS

... 2-14 2.6.1 Description of Structur al Design................................,..... 2-15 2.6.2 Material Properties,...............

..........................215 2.6.3 LiAing and Tie-down Standards for All Packages 2 15 2.6.4 General Considerations for Structural Evaluation of Packaging

-2 15 2.6.5 Formal Conditions of Transport....................................... 2-15 2.6.6 Flypothetical Accident Conditions....................................... 2 15 2.6 7 !,pecial Requirement for Irradiated Nuclear Fuel Shipments...................... 2-15 2.uS InternalPressureTest..

...................... 2-15 2.7 RLFERENCES............................................................. 2-16 3 THERMAL REVIEW............................................................. 3-1 3.1 REVIEW OBJECTIVE........................................................

3 1 3.2 AREA S OF REVIEW......................................................... 3 1 3.2.1 Description of the Thermal Design......................................... 31 3.2.2 Material Propaties and Ca=aa=* Specifications...........................

3-1 3.2.3 Thermal Evaluation Methods............................................. 3-1 3.2.4 Evaluation of Accessible Surface Temperatures................................ 3-1 3.2.5 Evaluation under %nal Conditions of Transport...........................31 NUREG-1617 vi DRAFT

3.2.6 Evaluation under Hypahb=1 Accident Conditions.............

......... 3-1

- 3.2.7 Appendix...........................................................

3 - 1 3.3 REGULATORY REQUIREMENTS.................................

........32 3.3.1 Description of the Thennal Desip

........................................3-2 3.3.2 Material Properties and Component Sr*=%............................ 1-2 3.3.3 Thennal Evaluation Methods.....................................

......32 3.3.4 Evaluation of Accessible Surface Temperatures................................. 3-2 3.3.5 Thennal Evaluation under Normal Conditions of Transport.............

....... 3-2 3.3.6 Thennal Evaluation under Hypothetical Accident Conditions................... 3 3 3.4 ACCEPTANCE CRITERIA.................................................... 3-3 3.4.1 Description of the Thennal Desip......................................... 3 3 3.4.2 Material Properties and Ce Spe**ms..........................

33 3.4.3 Thennal Evaluation Methods.............................................. 3-3 3.4.4 Evaluation of Accessible Surface Temperature.......................

33 i

3.4.5 Thermal Evaluation under Nonnal Conditions of Transport 33 3.4.6 "Ihermal Evaluation under Hypothetical Accident Conditions

...... 3 3 3.5 REVIEW PROCEDURES................................................ 3 -3 3.5.1 Description of the Thermal Desip.....................

3-4 3.5.2 Material Properties and C=r==t Specifications...

....36 3.5.3 Thermal Evaluation Methods.....................................

3-8 3.5.4 Evaluation of Accessible Surface Temperatures.......................... 3-12 3.5.5 Thermal Evaluation under Normal Conditions of Transport......

3 12 3.5.6 Thermal Evaluation under HyM*I Accident Conditions...................

3 13 3.5. 7 Appendix......................................................

.. 3 15

- 3.6 EVALUATION FINDINGS................................................. 3 17 3.6.1 Description of the Thermal Desip....................................... 3 17 3.6.2 Material Properties and C =i==t Specifications........

........... 3 18 3.6.3 Thennal Evaluation Methods.....................................

. 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 l-4.1 REVIEW OBJECTIVE...................................................

4-1 l

4.2 AREAS OF REVIEW..................................................... 4 1 4.2.1 Description of Contamment System....................................... 41 l

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

4.2.3 Contamment under Hypothetical Accident Conditions...........................

4 1

. 4.2.4 Appendix.............................................................

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

4.3.1 Description of C~a ht System........................................ 4-1 4.3.2 Contamment muler Nonnal Conditions of Transport

...........................4-2 4.3.3 Contamment under Hypothetical Accident Conditions............................ 4-2 4.4 ACCEPTANCE CRITERIA..................................................... 4-2 4.4.1 Description of Contamment System......................................

4-2 4.4.2 Containment under Normal Conditions of Transport.........................42 4.4.3 Containment under Hypothetical Accident Conditions.................

.... 4-3 DRAFT vii NUREG 1617

4.5 REVIEW PROCEDURES................................................... 4-3 4-3 4.5.1 Description of the Contamment System....................................

3 4.5.2 Containment under Normal Conditions of Transport........

.. 4-5 j

4.5.3 Contammant under Hypothetical Accident Conditions...........................

4-7 4.5.4 Appendix..........................................................

4-8 4.6 EVALUATION FINDINGS..................................................

4-8 4-8 4.6.1 Description of Contamment Systan.........................................

4.6.2 Cantammant under Normal Conditions of Transport............................ 4 8 4.6.3 Contamment under Hypothetical Acculent Conditions............................ 4-8

4.7 REFERENCES

.......................................................... 4 -9

. 5-1 5 Spr nING REVIEW.....................................

5.1 REVIEW OBJECTIVE...............................................i.......

5-1 5.2 AREA S OF REVIEW........................................................ 5-1 5.2.1 Description of the Shielding Design....

...............................5-1 i

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

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

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..................

52 5.4.4 Evaluation.......................................................

5-3

$.5 REVIEW PROCEDURES.................................................. 5-3 5-4 5.5.1 Description of the Shielding Design 5.5.2 Source Specification...........

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

.. 5-6 5.5.4 Evaluation.......................................

.. 5-8 5.5.5 Appendix................,....................................... 5 - 10

............ 5-10 5.6 EVALUATION FINDINGC..................................

. 5-10 5.6.1 Description of the Shielding Design.

5.6.2 Source Specification................

5-10 5.6.3 Model Specification.....

5-10 5.6.4 Evaluation........

5-10

5.7 REFERENCES

5-1 1 6 CRITICALITY REVIEW..........................

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

6.2 AREA S 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 Evaluations 61 NUREG-1617 viii DRAFT 4

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.

61 6 2.8 Appendix

..................................................... 6-1 6.3 REGULATORY REQUIREMENTS..................

6-1 6.3.1 Description of Criticality Design........................

........... 6-2 6.3.2 Spent Nuclear Fuel Contents.......................................... 6 2 6.3.3 General Considerations for Evaluations.

.. 6-2 6.3.4 Single Package Evaluation.......

6-2 6.3.5 Evaluation of Package Arrays under Normal Conditions of Transport

............62 6.3.6 Evaluation of Package Arrays under Hypothetical Accident Conditions............ 6 3 6.3.7 Benchmark Evaluations..

6-3 6.4 ACCEPTANCECRITERIA 63 6.4.1 Description of Criticality Design.

........ 6-3 6.4.2 Spent Nuclear Fuel Contents 6-3 6.4.3 General Considerations for Evaluations 6-3 6.4.4 Single Package Evaluation.

63 6.4.5 Evaluation of Package Arrays under Normal Conditions of Transport........... 6-3 6.4.6 Evaluation of Package Arrays under Hypothetical Accident Conditions 6-3 6.4.7 Benenmark Evaluations.

6-4 6.5 REVIEW PROCEDURES 6-4 6.5.1 Description of the Criticality Design.......

6-4 6.5.2 Spent Nuclear Fuel Contents 6-6 6.5.3 General Considerations for Evaluations 6-7 6.5.4 Single Package Evaluation.

.. 6-9 6.5.5 Evaluation of Package Arrays under Normal Conditions of Transport.

.. 6-10 6.5.6 Evaluation of Package Arrays under Hypothetical Accident Conditions 6-10 6.5.7 Benchmark Evaluations....

6-11 6.5.8 Appendix 6-12 6.6 EVALUATION FINDINGS

.. 6-12 6.6.1 Description of Criticality Design.

. 6-12 6.6.2 Spent Nuclear Fuel Contents 6 12 6.6.3 General Considerations for Evaluations 6-12 6.6.4 Single Package Evaluation....

6-12 6.6.5 Evaluation of Package Arrays under Normal Conditions of Transport..........

6-13 6.6.6 Evaluation of Package Arrays under Hypothetical Accident Conditions 6-13 6.6.7 Benchmark Evaluations............

6-13

6.7 REFERENCES

6-13 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 71 7.2.5 Appendix

................7-1 DRAFT ix NUREG 1617

7.3 REGULATORY REQUIREMENTS 7-1 7-1 7.3.1 Package loading......

7-2 7.3.2 Package Unloading 7-3 7.3.3 Preparation of Empty Package for Transport..

7.3.4 Other Procedures 7-3 7.4 ACCEPTANCE CRITERIA.....

73

. 73 7.4.1 PackaFe Loadmg.

7-3 7.4.2 Package Unloading..

7-3 7.4.3 Preparation of Empty Package for Transport 7.4.4 Other Procedures...

7-3 7.5 REVIEW PROCEDURES............

73 7-4 7.5.1 Package leadmg.........

....76 7.5.2 PackageUnloading 7-7 7.5.3 Preparation of Empty Package for Transport 7.5.4 Other Procedures 7-7 7-7 7.5.5 Appendix 77 7.6 EVALUATION FINDINGS 7-7 7.6.1 Packageleading

. 7-8 7.6.2 Package Unloading 7.6.3 Preparation of Empty Package for Transport.

. 78 7.6.4 Other Procedures 7-8

7.7 REFERENCES

7-8 8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW 8-1 8.1 REVIEW OBJECTIVE.

.81 8.2 ACCEPTANCE TESTS 8-1 8.2.1 Areas of Review.

.81 8.2.2 Regulatory Requirements 8-1 8.2.3 Acceptance Criteria 82 8.2.4 Review Procedures.

82

. 8-4 8.2.5 Evaluation Findings.

8.3 MAINTENANCE PROGRAM.

8-4 8.3.1 Areas of Review.

8-4 8.3.2 Regulatory Requirements 8-4 8.3.3 Acceptance Criteria 8-5 8.3.4 Review Procedures..

8-5 8.3.5 Evaluation Findings..

8-7

8.4 REFERENCES

87 APPENDIX A - STANDARD REVIEW PLAN CORRELATION WITH 10 CFR PART 71 AND REGULATORY GUIDE 7.9....................

A-1 APPENDIX B - TABLE OF EXTERNAL DOSE RATES FOR EXCLUSIVE-USE SHIPMENTS... B-1 APPENDIX C - COMMENT SHEET............

C-1 NUREG-1617 x

DRAFT

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

1-4 Figure 2-1 SAR Information Flow for the Structural Review 2-5 Figure 3-1 SAR Information Flow for the Thermal Review.

3-5 l

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..

6-5 Figure 7-1 SAR Infonnation 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 Design, Fabrication, Exammation, and Ter's.g Criteria for SNF Transportation Packages based on the B&PV Code......

1-7 Table 1-2 Welding Criteria for SNF Transportation Packages 19 Table 4-1 Release Fractions and Specific Activities for the Contributors to the Releasable Source Term for Packages Designed to Transport Irradiated Fuel Rods 4-6 Table 5. l Extemal Radiation Level Limits for Exclusive-Use Shipments 53 Table A 1 Standard Review Plan Correlation with 10 CFR Part 71 and Regulatory Guide 7.9 A-1 Table B-1 Extemal Dose Rates for Packages (Exclusive-Use Shipment)...

B-1 DRAFT xi NUREG 1617

ACRONYMS AND ABBREVIATIONS ALARA' as low as is reasonably achievable (radiation exposure) i-ANS'

.American Nuclear Society

' ANSI American National Standards Institute ASME American Society of Mechanical Engmeers ASTM American Society for Testing and Materials B&PV Boiler and Pressure Vessel (ASME Code)

Bq Becquerel BWR boiling-water reactor

'C degrees Celsius Ci Curie CFR U.S. Code of Federal Regulations DOT U.S. Department of Transportation

• F degrees Fahrenheit g

gravitational unit k., -

"k" efTective LSA low specific actisity MIL -

military-MNOP maximum normal operating pressure mrem millirem mSv millisievert(1 mSv = 100 mrem)

)

NMSS NRC Office or Nuclear Material Safety and Safeguards

)

i NRC U.S. Nuclear Regulatory Commission PWR pressurized-water reactor NUREG 1617 xii DRAFT

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 DRAFT xiii NUREG-1617

GLOSSARY

' The following terms are dermed 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 defmitions from these CFR sections have not been changed in the list below, but are repeated for convenience. Standards are expressed in the laternational System of Units (SI). The U. S. standard or customary unit equivalents presented in parentheses are for infonnation only, l

)

A, the maxunum activity of special form rmlioactive material permitted in a I

Type A package.

A the maximum activity of radioactive material, other than special form, LSA 2

and SCO material, permitted in a Type A package.

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

l Benchmarking validation of the accuracy of a computer code by comparison of obtained results with those of previously determtwd values.

Bias For criticality calculations, ANSI /ANS-8.1 dermes bias as a measure of systematic differences between calculations and expenmental data and subsequently dermes uncertainty in the bias. See NUREG/CR 6361 for further discussion of bias. The detemunation of bias must adequately consider the variation in the differences between the calculations and expenmental data.

Canier 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.

1 Certificate of compliance a certificate issued by NRC which authorues the use of a specific packaging, for a specified time, and for a specified scope of activity.

Close reflection by water unmediate contact by water of sufficient thickness for maxunum reflection ofneutrons.

Closed transport vehicle a transport vehicle or conveyance equipped with a securely attached exterior enclosure that during normal transportation restricts the access of unauthorned 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 "see-through" type, and must limit access from the top, sides, and bottom 1

i 1

i l

)

i NUREG-1617 xiv DRAFT l

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.

I Confirmatory calculations calculations made by the reviewer to determme 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.

Contamment system the assembly of components of the packaging intended to retain the radioactive material during transpon.

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.

The curie is equal to 37 billion disintegrations per sccond.

Damaged spent nuclear spent nuclear fuel with defects greater than hairline cracks or pinhole fuel leaks.

Depleted uranium uranium containing less uranium 235 than the naturally occumng distnbution 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 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 DRAFT xv NUREG 1617

natural wenium and depleted wanium, and natural uranium or depleted wanium that has been irradiated in thermal reactors only, are not included in this defmition. Certam exclusions from fissile material controls are provided in 10 CFR 71.53.

Fissile matenalpackage a 6ssile material packagmg together with its fissile mataial contents.

gravitational unit. (1 g - force exerted on a mass vertically by gravity) g I=l==6* catalat==

calculations separate from the applicant's. Input data should be taken frorn 5

pnmary sources such as the package drawmss and manufactwer'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 precedmg generation.

Low specific activity radioactive material with limited specific activity that satisfies the (LSA) material descriptions and limits set forth below. Shieldmg materials swroundmg the LSA material may not be considered in deterrmmng the estimated average specific activity of the package contents. LSA material must be in one of three groups:

(1) LSA-1.

(i) Orcs 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 A 2 value is unhmited;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 '

M I

104 A/g.

(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 unifonnly distributed, and the average specific activity does not exceed 10d A/g for solids and gases, and 104 A/g foi liquids.

l (3) LSA-III. Solids (e.g., consohdated 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 contamed in a relatively insoluble material, so that, even under loss of NUREG 1617 xvi DRAFT

packaging, the loss of radioactive material per package by leaching, when placed in water for 7 days, would not exceed 0.1 A ; and (iii) The average specific activity of the solid does not exceed 2 x 10-2 A/-

28 1ow 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 ofless than 10 days.

Maximum normal operating the maximum gauge pressure that would develop in the containment pressure (MNOP) system in a period of I year under the heat condition specified in 10 CFR 71.71(c)(1), in the absence of venting, extemal 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 0.71 I weight percent uranium-235, and the remainder by weight essentially uranium-238).

Normal form s whoactive 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 prescoted for transport.

Packaging the assembly of components necessary to ensure compliance with the packaging requirements of 10 CFR Pan 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.

Radiationlevel 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 contammated liquids or gases wiein the package.

Radioactive material any material having a specific activity greater than 70 Bq per gram (0.002 microcurie per gram).

DRAFT xvii NUREG 1617

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

dose equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor (I rem = 0.0I 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 repon in the context of this SRP, the report submitted by the applicant in (SAR) compliance with 10 CFR Pan 71, Subpan D. The fundamental contents of the repon are described in 10 CFR 71.31. Guidance on format and content of the repon is provided by Regulatory Guide 7.9, " Standard Format and Content of Pan 71 Applications for Approval of Packaging for Radioactive Material." The S AR is considered to be the submitted application, along with any supplemental data and responses submitted to NRC staff to resolve questions arising during the staffs 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 repon 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 cenificate of compliance.

Special fonn 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 Pan 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 Pan 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 defmition.

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

of a radionuclides The specific activity of a material in which the radionuclides is essentially uniformly distributed is the radioactivity per unit mass of the material.

NUREG 1617 xviii DRAFT

Spent nuclear fuel fuel that has been withdrawn from a nuclear reactor following (SNF) inadiation, the constituent elements of which have not been separated by reprocessing.

l Surface contammated object a solid object that is not itself classed as radioactive material, but which (SCO) has radianctive material distributed on any ofits surfaces. SCO must be in one of two groups with surface activity not exceedmg the following limits:

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

(i) The non-fixed contammation on the accessible surface averaged 2

2 over 300 cm (or the area of the surface ifless than 300 cm ) does not 2

2 exceed 4 Bq/cm (10dmicrocurie/cm ) for beta and gamma and low toxicity alpha emitters,or 0.4 Bq/cm (10 8 microcurie /cm ) for all 2

2 other alpha emitters; (ii) The fixed contammation on the accessible surface averaged over 2

2 300 cm (or the area of the surface ifless than 300 cm ) does not exceed 4 x 10' Bq/cm (1.0 microcurie /cm ) for beta and gamma and 2

2 low toxicity alpha emitters, or 4 x 10' Bq/cm (0.1 microcurie /cm ) for 2

2 all other alpha emitters; and (iii) The non-fixed contammation plus the fixed contammation on the inaccessible surface averaged over 300 cm (or the area of the surface if 2

2 2

2 less than 300 cm ) does not exceed 4 x 10d Bq/cm (1 microcurie /cm )

for beta and gamma and low toxicity alpha emitters, or 4 x 105 Bq/cm2 2

(0.1 microcurie /cm ) for all other alpha emitters.

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

(i) The non-fixed contammation on the accessible surface averaged 2

2 over 300 cm (or the area of the surface ifless than 300 cm ) does not 2

2 exceed 400 Bq/cm (104 microcurie /cm ) for beta and gamma and low toxicity alpha emitters or 40 Bq/cm (10 8 microcurie /cm ) for allother 2

2 alpha emitters; (ii) The fixed contamination on the accessible surface averaged over 2

2 300 cm (or the area of the surface ifless than 300 cm ) does not exceed 8 x 105 Bq/cm (20 microcuries/cm ) for beta and gamma and 2

2 low toxicity alphs emitters, or 8 x 10* Bq/cm (2 microcuries/cm ) for 2

2 all other alpha emitters; and (iii) The non-fixed contammation plus the fixed contammation on the inaccessible surface averaged over 300 cm (or the area of the surface if 2

less than 300 cm ) does not exceed 8 x 105 Bq/cm 2

2 (20 microcuries/cm ) for beta and gamma and low toxicity alpha 2

emitters, or 8 x 10' Bq/cm (2 microcuries/cm ) for all other alpha 2

2 emitters.

Transportindex 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 transponation. The transpon index is detemuned as follows:

(1) For non fissile material packages, the number determmed by multiplying the maximum radiation level in millisieven (mSv) per hour at one meter (3.3 ft) from the extemal surface of the package by 100 DRAFT xix NUREG-1617

(equivalent to the maximum radiation level in adhran per hour at one meter (3.3 A));or (2) For fissile material packages, the number determmad by multiplying the maximum radiation level in raillisievert per hour at one meter (3.3 ft) froen the external swface of the package by 100 (equivalent to the =muimmu raAanan level in nulhrem per hour at one meter (3.3 A)), or, for criticality control pwpones, the number obtamed as described in 10 CFR 71.59, wlucheveris larger.

Type A quantity a quantity of radioactive mataial, the aggregate radioactivity of winch does not exceed A, for special form rasoective material, or A, for normal form 2

radioactive material, where A, and As are givenin Table A-1 of10 CFR Part 71, or may be deteramed by procedwes described in Appen&x A of 10 CFR Part 71.

Type B package a Type B packagmg together with its radioactive contcets. On approval, a Type B package design is designated by NRC as B(U) unless the package has a maximum normal operating presswe of more than 700 kPa (100 lb/in ) gauge or a pressure relief device that would allow the release of 2

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 unilateral approval ofinternational shipments; B(M) refas to the need for multilateral approval ofinternational shipments. There is no distmetion made in how packages with these designations may be used in domestic transportation.. To determme 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 ra&oactive 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"), subpans (e.g., Subpart D,

" Application for Package Appmval"), and sections (e.g.,10 CFR 71.31).

See also Title 49," Transportation."

.p NUREG 1617 xx DRAFT

INTRODUCTION The Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, referred to here as the Standard Review Plan (SRP), provides guidence 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 interpreution of NRC regulations. This SRP supplements NRC Regulatory Guide (RG) 7.9," Standard Format atd 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, 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.

I 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 ofReview 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.

l DRAFT 1

NUREG-1617 l

Subsection 3. RegulatoryRequirements 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

' cubsection includes the regulatory requirements by reference and identifies other criteria thc. :se 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. Rese 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 nis 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.

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 2

DRAFT

i 1 GENERALINFORMATIONREVIEW l.1 REVIEWOBJECTIVE 1he objective of this review is to establish (1) that the application includes an overview of relevant package infor==han includag int-lad use; and (2) a suunnary description of the packagmg, operational features, and

=va*= that provide reasonable assurance that the package can meet the regulations and operatmg objectives.

I 1.2 AREAS OF REVIEW I

.]

The SAR should be reviewed for adequacy of the package description and drawings of the packaging. Areas ofreviewinclude 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 Fabncation 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 OperationalFeaturm 1.2.3.3 Contents of Packaging 1.2.4 Compliance with 10 CFR Part 71 1.2.4.1 General Reqmrements of 10 CFR 71.43 1.2.4.2 Condition of Package aAer Tests in 10 CFR 71.71 and 10 CFR 71.73 1.2.4.3 Structural, Thermal,Contamment, Shielding, Criticality 1.2.4.4 Operational Procedures, Acceptance Tests and Maintenance 1.2.5 Appendix 1.3 REGULATORY REQUIREMENTS Regulatory reqmrements of 10 CFR Part 71 applicable to the general information review are as follows 1.3.1 GeneralSARFormat There are no speci6c regulatory regarements on the format of the SAR. SAR format provisions are give in RG 7.9, DRAFT 11 NUREG-1617 L

L __

J

1.3.2 Package Design Information The application for package approval must: (1) include the classi5caten aml model number

[10 CFR 7131(aXI),10 CFR 7133(aXI), and 10 CFR 7133(aX3)]; (2) include a quality assurance program description or a reference to a previously approved quality assurance program applicable to the l

package [10 CFR 7131(aX3) and 10 CFR 7137]; (3) identify applicable codes and usandards used in package design, fabncaten, assembly, testing, maintanaam, and use [10 CFR 7131(c)); and (4) include the transport index for nuclear criticality control. [10 CFR 7131(aX2),10 CFR 7135(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 7138]

All changes in the conditions speci6ed 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 7131(b). [10 CFR 71.107(c)]

i 1.33 Package Description The description of the packaging must include a containmen.t system, materials of construction, weights, dimensions, d~Ic of fabrication, and coolant receptacle volumes in sufficient detail to provide an adequate l

basis for their evaluation. [10 CFR 7131(a)(1),10 CFR 7133(aX2),10 CFR 7133(aX4),

l 10 CFR 7133(a)(5), and 10 CFR 7133(aX6)]

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

constituents, physical and chemical fann, neutron absorbers or moderators, extent of ref:ection, moderator-d j

fissile ratio, maximum normal operating pressure, maxunum weight, maximum decay heat, and any cooLat l

volumes. [10 CFR 7131(a)(1) and 10 CFR 7133(b)]

'l 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)]

1.3.4 Compliance with 10 CFR Part 71 The package must be evaluated to demonstrate compliance with the requirements specified in 10 CFR l

Part 71, Subpart E, under the conditions and tests of Subpart F. [10 CFR 7131(aX2),10 CFR 7135(a), and 10 CFR 71.41(a))

1.4 ACCEITANCE CRITERIA 1.4.1 GeneralSARFormat 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 13.2 identify the acceptance criteria.

NUREG-1617 12 DRAFT I

j

I A.3 Package Description in addition to the regulatory requvements identified in Section 1.3.3, a discussion of the operation of the package should be provided. [RG 7.9 (1.2.2)]

1AA Compliance with 10 CFR Part 71

- In addition to the regulatory requvements utentified in Sectmn 1.3 A, a concise statement by the applicant, that the package complies with the requuements in 10 CFR Part 71 for a Type B(U)F package, should be provided in the General Information sectmn of the SAR. This summary staeamaat 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 REVIEWPROCEDURES 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 infonuation by itself to perform a technical review of the package, the Generallnformation 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. 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.

L

. 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.

i 1.5.1 General SAR Fornist l

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

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

pnntag, binding, page numbering, separators, and number of copies).

i L

1 DRAFT l-3 NUREG 1617 i

E 1

Generallsfortsation Review Pedage Danign infunneelen Packase Description Casapliance with 10 CFR Part 71 Pwpone of applicanon Packaging

- General requirement of Quality answance program Opermoani featwes 10 CFR 71.43 uss/ general conesats t'a=*== of packaging

- Con &monof after tests in type and model 10 CFR 71.7 and 0 CFR 71.73 aumber Structural. theranal cons====at, Package caessory and shieldsg.cnticanty Operanomal procedures. =raara-*

mammen acevity Fabncanon and welding critaria test and mainannonce

- Transportindex and mari==

aumberofpackages -

1F-1P P

P aitnscoural Thennel Cantainsaaest Shielding Evaluation Evaluation Evahsation Evahsation

- Internal and external Dimensions

- Dmwnsions Dimensions structures

- Componentmatenals

- Component matenals

- Componcat maarnals Fabncahon and Decay heat Containment boundary Contents welding entena.

Heat dissipation

- Contents

- Codes and standards

- Allowable lankage rate Component snaserials

- Dimensions Weights IF l'

1F CriticaEty Operating '

Acceptance Tests and Evahsation Procedieres -

Maintenana Program

- Fiasile content

- Operationalfastures Codes and standards snessnels

- Generalrestnctions Dimennons and Dimennons and Tamper-indicaang tolerances tolerances device Component masenals

- Component masenals

- Contents Coments Neutron poison coniena Figure 1-1 SAR Infor nation Flow for the GeneralInformation Review.

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 sirc.dirs.t, 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 NUREG-1617 1-4 DRAFT

changes being requested are clearly identified. Modifications may include design changes, changes in authonzed contents, or changes in conditions of the approval. Design changes should be clearly identified and should be included in revised packaging drawmgs. 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, W-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 i

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

10 CFR Part 71. Applications for modifications may be subject to the provnions 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 Verify that the applicant has obtained NRC approval ofits 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.

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 desipated.

1.5.2.4 Package Type and Model Number Confum that the type and model number of the package are desipated. A new SNF transportation package will be designated B(U)F-85 unless it has a maximum normal operating pressure (MNOP) greater than 2

700 kPa (100 lb/in ) or a pressure relief device that would allow the release of radioactive material under the i

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 desipated for the package and that it is specified on the i

appropriate drawings.

l.5.2.5 Package Category and Maximum Activity Category l is assigned to a package whose content activity exceeds either 1.11 x 10" Bq (30,000 Ci),

3000 A, c 3000 A whichever is less. (SNF transportation packages are assumed to be Category 1 in the i

3 following SRP review procedures.) Verify that the package is desigoni Category I and that the maximum activity of the package contents is specified.

DRAFT l5 NUREG 1617

[

1.5.2.6 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 will accept full compliance with the ASME Section Ill, Division 3, Code for the containment system, including the services of an Authorized Inspection Agency.

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 design, 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 Contairunent, 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, weldmg, and examination of package cotnponents may be found.

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 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 desices, impact limiters, structural and mechanical means of heat dissipation, types of coolant, outer and inner protmsions, shielding, pressure relief systems, and closures Identification of the containment system and boundary.

NUREG-1617 1-6 DRAFT

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Table 1-2 Welding Criteria for SNF Transportation Packages.

Weld type I

E Contatemeet-related Criticality-related Other safety-related BAPV Coje section Sec. Ill, Division 3 Sec. III, Subsection NG Sec VIII,Div. I or Sec. III. Subsection NF Base materials WB-2000, WB-4100, NG-2100, NG-2200, Sec. Vill, Div.1, Subsection and applicable Code cases NG-2500,NG 4100, A GeneralRequirements; and applicable Code cases appropriate parts of Welding and brazing.

WB-2400 NG-2400 Subsection B, Methods of materials Fabrication;and Subsection Joint preparation WB-4200 NG-4200 Welding WB-4400 NG-4400 Or Brazing WB-4500 NG-4500 g3 g

Heat treatment WB-4600 NG-4600 Qualification of WB-4300 NG-4300 procedures and personnel Examination WB-5000 NG-5000 Quahty assurance 10 CFR Part 71, Subpart H and RG 7.10 Fracture toughness RG 7.11 or 7.12. as appropriate Examme 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 ofcompliance and become regulatory conditions for compliance.

Each drawing should be identified by drawing and sheet number, resision number, and page number.

Proprietary information should be clearly identified. Ensure that the following information is presented in the drawings.

General arrangement of the package Package dunensions Design and safety features of the packap Components and their materials of construction, including material specification as appropriate (e.g.,

l ASME, ASTM, MIL, DOT)

Dunensions and tolerances ef principal components Cooc, candards, or other appropriate fabrication specifications DRAFT l9 NUREG 1617 l

1 1

Weld symbols used, welding criteria, and exammation techniques and acceptance criteria Operational information (e.g., bolt torque, pressure-relief specifications)

Maximum weight ofthe package Maximum weight of contents, including'contamers, shormg, and other packaging material y

I Tamper indicating device.

i

+

i The drawings should idesitify the classification of all compeecsts important to safety. RG 7.10 provides guidance on classifying structures, systems, and components (SSCs) important to safety. Additional l

guidance and typical classifications of SSCs for SNF packagmgs are provided in NUREG/CR-6407.

l 1.5.3.2 OperationalFeatures 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, opemngs, 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 refere ;ed in the General Information section of the SAR, if provided in a later SAR section or appendix. In this case, i

simplified operational wha=='ics should be an acceptable alternative in the General Information section of the SAR. Loadmg configurations for all contents should be provided and annotated in a manner consistent l

with the Structural Evaluation, contamment 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.

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

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.

l 1.5.3.3 Contents l

\\

The contents should be described in the same detail as that intended for the certificate of compliance. Resiew the description of the contents and verify that, as a mimmum, 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 nummum initial enrichment, including a description of non-uniform l

enrichment,if applicable Fuel assembly specifications, including dunensional 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 l

l l

NUREG-1617 l-10 DRAFT i

l

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 I

gases 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 s essel Containment fill gas

+

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

For damaged fuel, the maximum quantity of damaged fuel, initial enrichment, extent of damage, and description of the second containment system, and any other limits, as applicable are specified.

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 S AR where discussions ofcompliance 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.

DRAFT l-11 NUREG 1617

1.5.4.3 Structural, Thermal, Contamment, Shielding, Criticality Verify that summary statements are provided attesting to the adequacy of the package design to meet the stmetural, thermal, containment, shielding, and criticality requirements of 10 CFR Part 71.

1.5.4.4 Operational Procedures, Acceptance Tests and Maintenance Verify that a summary statement is provided attesting to the adequacy of the development of the operational procedures and accerW.cc tests and maintenance program to ensure compliance with the reqmrements of 10 CFR Part 71.

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 resiewer. The appendix may also provide i

supporting details on special fabication procedures and other appropriate supplemental information.

j 1.6 EVALUATION F!NDINGS The Safety Evaluation Report (SER) does not normally include specific findings for the General Information section of the SAR.

i i

Before proceedmg with the review of the other sections of the SAR, the reviewer should conclude, at a mmimum, that the following criteria have been demonstrated:

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

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

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

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

I l

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

the package.

l 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).

NUREG-1617 1 12 DRAFT

t l

1.7 REFERENCES

ANSI N14.5 '

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

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

- B&PV Code American Society of Mechanical Engineers, "ASME Boiler and Pressure Vessel Code," New York, NY,'1995.

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,1997.

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.

l NUREG/CR-6407 U.S. Nuclear Regulatory Commission, " Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety," NUREG/CR-6407, U.S. Government Printing Office, j

Washington, D.C., February 1996.

i RG 7.10 U.S. Nuclear Reguistory 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 j

Containment Vessels with a Maximum Wall nickness of 4 inches (0.1 m),"

i 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 i

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

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

L l

DRAFT l-13 NUREG-1617 i

E--________.

___a

2 STRUCTURALREVIEW i

2.1 REVIEW OBJECTIVE 1

i The objective of this review is to verify that the structural perfonnance of the package has been adequately evaluated for the tests speci6ed 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

'Ihe SAR should be reviewed for adequacy of the description and evaluation of the structural design. Areas of reyww 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 Othw 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 l

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 2.2.6.2 Crush l

2.2.6.3 Puncture 2.2.6.4 Thermal 2.2.6.5 Immersion-Fissile Material DRAFT 2-1 NUREG.1617

- _ - - _ = _ _ _ _ _

i 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 regeranents of 10 CFR Part 71 appbcable to the structural revsew are as follows 2.3.1 Description of Structural Design The packagmg must be described in sufficient detail to provide an adequate basis for hs 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 fabncation of the package. [10 CFR 71.31(c)]

2.3.2 Material Properties The package must be made of materials and construction wiuch assure that there will be no significant chemical, galvanic, or other reactions among the packagmg components, among package contents, or between the packaging components and the package contents, including possible reaction resulting from inleakage of water, to the maximum credible extent. 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 reqw a 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))

By subjecting a specimen or scale model to a specific test, or by another appropriate and acceptable method, the effects on the perforrnaner of the package under the tests specified in 10 CFR 71.71 (normal conditions of transport), in 10 CFR 71.73 (byWk=1 wuiant conditions), and in 10 CFR 71.61 (special requiranent for irradiated nuclear fuel shipments) must be evaluated. [10 CFR 71.41(a))

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 ihat under the tests, there would be no substantial reduction in NUREG-1617 2-2 DRAFT

the effectiveness of the packaging. [10 CFR 71.35(a),10 CFR 71.43(f),10 C'FR 71.51(a)(1), and 10 CFR 71.55(d)(4)]

23.6 Hypothetical Accident Conditions l

The package must be evaluated under the tests specified in 10 CFR 71.73 for hypothetical accident l

conditions. [10 CFR 71.41(a)) The evaluation must show that the packaging has adequate structural integrity l

to satisfy the containment, shielding, suberiticality, and temperature reqmrements of 10 CFR Part 71, Subpart 1

E. [10 CFR 71.35(a))

23.7 Special Requirement for Irradiated Nuclear Fuel Shipments The contamment vessel of a package with activity greater :han 37 PBq (10' Ci) must be designed to withstand an external prersure of 2 MPa (290 psi) for a period of not less than one hour without collapse, buckling, or inleakge of water. [10 CFR 71.6I) 23.8 InternalPressureTest 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 structural design should comply with RGs 7.6,7.8,7.11, and 7.12 and be consistent with the package fabrication and welding criteria identified in Section 1.5.2.6.

2.4.2 Material Properties In addition to the regulatory requirements identified in Section 2.3.2, the structural design should comply with RGs 7.11 and 7.12 for precluding brittle fracture in containments made of ferritic steels. Material properties should meet the material specifications applicable to the codes and studards used for the design and fabrication of the package.

2.43 Lifting and Tie-Down Standards for All Packages Tbc regulatory requirements in Section 2.3.3 identify the acceptance criteria.

2.4.4 General Considerations for Structural Evaluation of Packaging In addition to the regulatory requirements identified in Section 2.3.4, the structural design should comply with RGs 7.6 and 7.8.

DPAFT 2-3 NUREG-1617

i

2.4.5 Normal Conditions of Transport

'Ibe regulatory requ;.4 in Section 2.3.5 idmhfy the acceptance criteria.

i

' 2.4.6 Hypothetical Accident Conditions The regulatory requiranents in Section 2.3.6 identify the acceptance criteria.

j 2.4.7 Special Requirement for Irradiated Nuclear Fuel Shipments The regulatory requuements in Section 2.3.7 identify the==*== witeria.

2.4.g laternal Pressure Test u

In addiDrm the regulatory requirements identified in Section 2.3.8, the contammmt structme should meet the 10 CFR 71.85(b) requirements for pressure test without yielding.

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 Informaten and Thermal Evaluation sections of the SAR. Similarly, results of the structural review are considered in the review of the SAR sectmas on Thermal Evaluation, Ceaht 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 ar shown in Figure 2-1.

2.5.1 Description of Structural Design 2.5.1.1 Descriptive Information including Weights and Centers of Gravity Review drawmgs and other descriptions of the structural design in the General Information and Structure Evaluation sections of the SAR. 'Ihe information should describe the function, geometry, and mataial of construction of all structural components of the packagmg and its li&g and tie <iown devices; The information should be sufficient for evaluatmg the structural performance of the packagmg to meet the regulatory reqmrements, wluch include contamment, shiciding, and maintammg subcriticality of the 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 drawags and descriptions of the structural design in the SAR.

Verify that packaging drawmgs provided in General Information section of the SAR specify the materials of construction, dunensions, tolerances and fabncation methods of the packaging and subassemblies, receptacles, internal or external support structmes, valves and ports, lihg devices, and tae down devices, and other design features relevant to the structural evaluation. Desc-iptive information such as the maxunum weight of the package, the maximum weight of the contents, the centa of gravity of the package, and the MNOP should be included.

NUREG-1617 2-4 DRAFT

General Thermal Infonanton Evahsation Insernal and external Temperatures senennes

- Pressures Fabricanon and widmg cntens Codes and standards t'a=pne=nt sessenals

- Dmwnsions Weights 1I U

j i

Structural Review l

Imading Evahsation Results

- Com saion

- Masenalproperties Bucking

- Ptyncal tesang

- Fangue

- Impact Stress analysis Strain

- Land combinations

- Differentialtherma!

- Stress Penetration expanmon

- Deformation Pressure

- Fracture Puncture

%branon ray l

i lf U

lI II Therusal Containment Shiddhig Criticality Evaluation Evaluation Evahsation Evahsation Deformanon Deformation Deformanon Deformation

- Chemical and galvanic

- Crustung' puncture Displacement of reactions

- Extnmon contents and poisons Contents condition

- Displacementof

- Flooding contents and shielding 1I U

i Operating Acaptance Tests and j

Procedures Maintenance Program Closure con 6guration Codes and standards lifhngcon6 uranon

• Structural and 3

Bolt torque Pressure test

- TMwn con 6 uration 3

- Handling resenctions i

I Figure 2-1 SAR Information Flow for the Structural Review.

DRAFT 25 NUREG 1617 l-l

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 constmetion of the contamment systems of an SNF or high-level rr.dioactive waste transport packaging.' The code is published asSection III, Division 3, ASME Boiler and Pressure Vessel Code (B&PV Division 3). 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 ' ode, the SAR must explicitly state the applicant's justification for the deviation, and thejustifica6 s :aust be acceptable to NRC.

NUREG/CR-3854 identifies codes and standards which may be used for fabricating other cunpcecnts 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 additional 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 contamment system of Type B packages,(3) RGs 7.11 and 7.12 describe criteria for precluding brittle fracture in package containers made of femtic steels, (4) NUREG/CR-4554A discusses the buckling evaluation of containment vessels, (5) j NUREG/CR-6322 provides guidance for buckling analysis of SNF baskets,(6) NUREG/CR-6007 provides guidance and criteria for design analysis of closure bolts for packagings, and (7) NUREG/CR-3019 presents criteria for transportation package welds.

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 1

i provide adequately documented material properties that are important for the design and fabrication of the packaging. A list of rertinent material properties needed to defme the material for analysis should be provided.

NUREG-1617 -

2-6 DRAFT l

(

l Verify that the materials of structural E--g==ts whose structural integrity is essential for the package to meet regulatory requirements have sufficient fracture toughness to preclude brittle fracture under the specified I

normal conditions of transport and hypothetical accident condition temperatures and loads. Brittle fracture l

must be precluded for the contet vessel under severe impact loads at the lowest service temperature.

Fracture toughness criteria for ferritic steel packagmg containment vessels are provided in RGs 7.11 and i

7.12.

Verify that the material properties used are appropriate for the load condition (e.g., static or dynanuc impact loadag, hot or cold temperature, wet or dry conditions, etc.). Verify that appropriate temperatures at which l

allowable stress limits are defined are consistent with those temperatures expected in service.

If the package has impact limiters, the adequacy of the method used for establishing their force-deflection l

charactenstacs should be verified by testing. Testing of the impact limiters may be camed out statically, if the effect of strain rates on the material crush properties is accounted for and properly included in the force-l 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 j

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 interface between the metals at a lower temperature than the melting points of the metals in contact. Review methods used to prevent eutectic reactions.

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 coatmgs 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):

DRAFT 27 NUREG-1617

I Any likg =" h-t which is a structural part of the package must be designed with a muumum safety i

factor of three agamst yielding when used to liA the package in the intended manner A lif Wriuhment which is a structural part of the package must be designed so that its failure under exet 1 would not impaa the ability of the package to meet other requirements.

The location and constructed of the libg devices should be shown on the packagmg drawings. Any other structural part of the package that could be used to lift the package must be rendered moperable for ling durmg ransport or be designed with strength equivalent to that required for likg attehmemes.

2.5.3.2' Tie-Down Devices.

Review the design and evaluaten of the tie-down devices that are a structural part of the package, their connecten 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 reqmrements of 10 CFR 71.45(b):

Any tie-down device which is a structural part of the package must be capable of withstandmg, widmut generating stress in any material of the packaee in excess ofits 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 compreent 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 prekage must be designed so that its failure under excessive load would not impair the ability of the p skage to meet other reqmrements.

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 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 loadmg conditions of the packaging.

The methods employed are appropriate for loadmg conditions considered and follow Wad practices and precepts.

1 Interpretatmos 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.

NUREG-1617 2-8 DRAFT I

. 2.5.4.1 Evalustaan by Analysis l

If the structural evaluation is by analysis, the review should include tlw followmg:

Verify that the SAR describes clearly the analysis models, mesharie, and results including all assumptions l

and input data used. The analysis model should adequately represent the wy, boundary conditions, i

s loadmg, material properties, and structural behavior of the packagmg analyzed j

Verify that the matenal model and properties are appropriate for the analyses. If the analysis is an clastic analysis, the matenal should also be modeled as an clastic material. If the analysis is inelastic, the actual matenal behavior should be uud. The SAR should describe how the material properties were obtamed and why the material model is appropriate for the loadmg conditions considered. For analysis involving large strains, the reviewer should verify that a true stress-strain curve is used.

Verify that the applied (fora and displace =aat) boundary conditions in the analysis model are appropriate. For free-drop impact analyses, impact loads for package components are usually derived from the dynamic analysis of the package and used m a quasi-static analysis of the component. Verify that a dynamic amplification factor has been applied to the equivalent static load. 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 venfied For computer codes that are not in t

the public domam and well established, the SAR should describe the solution method, the bench m arking results, and the quality assurance practices of the computer program.

Verify that the tuost critical combinations of environmental and loading conditions are evaluated. At a nummum, 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 lughest 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, arJ 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 earnart group will depend on the vulnerable packaging components and their structural failure modes in order to reduce the number of drops that must be evaluated, comporents 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 critical 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 strams should be compared to wig-r Eg 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 ofindividual members, as applicable, should be evaluated.

DRAFT 2-9 NUREG 1617 i

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 includag the use ofreduced-l scale models, wluch are mmmonly used for testing SNF packages 1

Verify that the test specimes is fabncated using the same materials, methods, quality asswance, and 6,a~daa specifications as specified in the design. Any differences should be identified and the effects evaluated in the SAR. The specunens should include all safety components to be tested and mmpaamu that are expe::ed to have significant effects on the test results. Substitutes for the radmetive contents dwing the tests should have the same structwal properties as the actual maunts. The substitutes should have the same mass and same interaction with its surroundmg packagmg 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 determme 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 % scale is used.

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 reasomng. The criteria in Secuan 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==tha4 and instruments are adequate for the measurements and the measurements are suflicient 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 specunen 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 agamst an -=+3=ble structwal failure scenario such as a sudden or total collapse or rupture should be evaluated. If the final damage indicates the possibility of an immment unstable structwal failure, additional tests under the same test conditions should be performed to deterame the repeatability of the result. If W-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 remam constant at that value between -29*C ( 20*F) and +38'C (100'F) which is most unfavorable for the feature under NUREG 1617 2-10 DRAFT

l l

considerate The initial internal pressure within the copimammt system must be the MNOP unless a lower eternal 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 sprey 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 l

Verify that the heat loadmg 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.

l Verify that the stresses are within the limits for normal condition loads.

i The evaluations should be based on the manmum ambient temperature and the MNOP in combination with i

the maximum internal heat load. For specified components of the package (e.g., elastomer seal, neutron shield material, etc.), resiew the maximum temperatures, and their effect on the operation of the package.

Fatigue effects m y be considered.

2.5.5.2 Cold l

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 muumum 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 Da J4 that the SAR adequately evaluates the package design for the effects of reduced extemal pressure equal to 25 kPa (3.5 psi) absolute.

2.5.5.4 Increased External Pressure DLde that the SAR adequately evaluates the package design for the effects ofincreased external pressure equal to 140 kPa (20 psi) absolute. Consider this loading condition in combination with muumum intemal pressure Consider the possibility of buckling (NUREG/CR-4554A).

2.5.5.5 Vibration Determme that the S AR adequately evaluates the package design for the effects of vibration normally incident to transport. The SAR should provide a determmation 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 DRAFT 2-11 NUREG-1617

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 packagmg component. The effect on packsaa 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 approxunately 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.

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, intemal pressures, thermal stress, and all other concurrently applied forces (e.g., seal compression force, bolt preload, etc.).

2.5.5.8 Comer Drop This test applies only to fiberboard, wood, or fissile material rectangular packages not exceedmg 50 kg (110 lb) and fiberboard, wood, or fissile material cylindrical packages not exceedmg 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 hypaMr=1 accident conditions must be based on sequential application of the tests specified in 10 CFR 71.73,in the order i-Ar=M to determme 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 constant at that value between -29'C ( 20'F) and +38'C (100'F) wiuch is most unfavorable for the feature under consideration. The initial internal pressure within the contamment system must be the MNOP unless a lower intemal pressure consistent with the ambient temperature assumed to procede 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.

NUREG 1617 2-12 DRAFT

The package must have adequate structwal integrity to satisfy the contamment, shielding, subcriticality, and temperatwe requirements of 10 CFR Part 71. Generally, inelastic deformation of the contamment and the closure system (e.g., bolts and flanges) is unameptable for the containment evaluaten Deformation of other parts of the contamment vessel may be acceptable if the contamment boundary is not compronused.

Deformation of stueldmg components must be reviewed in the shielding evaluation. Deformation of components requund for heat transfer must be reviewed in terms of the thermal evaluation. Deformation of components required for subcriticality must be reviewed in the criticality evaluatens 2.5.6.1 Free Drop Review the evaluauon of the package for the free-drop test. Verify that structural integrity has been evaluated for the drop orientation wiuch produces the highest g-load and causes the most severe damage If a featme such as a tie down component is a structural part of the package, it must be included in the drop-test configwauons 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 oflead slump should be evaluated for the hypothetical accident condition free drop. The lead slump determmed should be consistent with that assumed in shiciding 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 A, not as special form material.

2 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 evalcating the puncture test.

1 Generally, thin shelled packages are susceptible to puccture damage Verify that punctwe at oblique angles, l

- near a support, at a valve, or at a penetration have been considered.

l Although analytical methods are available for predictag punctwes of plates, empirical formulas derived from punctwe test results oflammated panels are usually used for determuung the package surface layer tinckness reqmrod for resisting pstuiss. The Nelm's formula dcvive specifically for package design provides the nummum tinckness needed for preventing the punctwe of the steel surface layer of a typical steel-lead-steel lammated cask wall. NUREG/CR-4554B provides an empirical formula for puncewe evaluauon based on empirical and analyucal punctwe stuches The formula is applicable for puncture at an angle normal to the i

saface and at a location away from a stiff support under the swface. The formula is conservative for solid packagmg 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.

DRAFT 2-13 NUREG-1617

2.5.6.41hermal Verify that the package design is evaluated for a fully engulfmg fire as specified in 10 CFR 71.73(c)(4). Any damage resulting from the free drop or punctwe conditions must be impcsated into the initial condition of the package for the Src test. Confirm that the determmation of the maximum pressure in the package during or aAer the test considers the temperstwes resulting from the fire and any increase in gas inventory caused by thermal combustion or decomposition process. Verify that the maximum thermal stresses, wiuch can occur ather during or aAer the fire, are evaluated and are consistent with the Thermal Evaluation section of the SAll l

4 2.5.6.5 kamarmon-Fissile Material If water inleakage has not been assumed for the criticality analysis, review the evaluation of a damaged test specunen (i.e., aAer free drop, punctwe, and fire) immersed under a head of water of at least 0.9 m (3 A) in

)

the attitude for wluch maxunum leakage is --W l

2.5.6.6 Immersion-AllMaterial Review the evaluation of a separate, undamaged specimen subjected to water pressure equivalent to unmersion under a head of water of at least 15 m (50 A). 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 ofirradiated nuclear fuel shipment with activity greater than 37 PBq (10' Ci),10 CFR 71.61 requires that its undamaged contamment system can withstand an extemal water pressure of 2 MPa (290 psi) for a penod 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 contamment stmeture will not collapse or buckle within one hour aAer the presswe is applied. This test applies only to the contamment system. No structural support from other packagmg components should be considered. The inleakage requirement has not been met if the stresses around the closwe seal regic a exceed the yield stress limits.

2.5.8 Internal Pressure Test For a package with a MNOP exceedmg 35 kPa (5 psi) gauge,10 CFR 71.85(b) requires that prior to first use the contammem system be presswe tested at 150% ofits MNOP. The analysis of this acceptance test should be provided in the SAR. The analysis should show that the package contamment structwe does not yield under the test presswe and the stresses are within the allowable stress limits set by the design code.

2.5.9. t x " s The appendix may include a list of references, copies of any applicable references not generally available to the reviewer, computer code descriptions, input and outpnt files, test results, and other appropriate supplementalinformation 2.6 EVALUATIONFINDINGS The structwal review should result in the following fmdings, as appropriate:

NUREG-1617 2 14 DRAFT

T l

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 reqmrements of 10 CFR 71.31

& staff has reviewed the codes and standards used fu 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 packagmg components, among package contents, or between the packaging waycs. cats and the contents in dry or wet environment conditions. The effects of radiation on materials are considered and package contamment is constructed from materials that meet the ryuwt of RGs 7.11 and 7.12.

I 2.6.3 Lifting and Tie-down Standards for All Packages l

The staff has reviewed the lifting and tie-down systems for the package and concludes that they meet I

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 of10 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 suberiticality, contamment, 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 contamment structure and concludes that it will meet the 10 CFR 71.85(b) requirements for pressure test without yielding.

DRAFT 2 15 NUREG.1617 1

2.7 REFERENCES

B&PV Division 3 Code Amencan Society of Mechanical F=i=s,"ASME Boiler and Pressure Vessel Code, Section Ill, Division 3, Contamment Systems and Transport Packagmgs For Spent Nuclear Fuel and High Level Radioactive Waste," New York, NY, 1997.

NUREG/CR-0128 U.S. Nuclear Ra=h~y Commission," Shock and Vibration Environments for j

a Large Shipping C--i-Durms Truck Transport (part II)," NUREG/CR-0128, U.S. Governmant Printing Office, Washington, D.C., August,1978.

NUREG/CR-2146 U.S. Nuclear Regulatory en-mienion, "Dynanuc Analysis to Establish Normal Shock and Vibration of Radioactive Matenal Shipping Packages, Volume 3:

Final Summary Report," NUKEG/CR-2146, Vol. 3, U.S. Government Piinting Office, Washington, D.C., October 1983.

NUREG/CR-3854 U.S. Nuclear Regulatory Comminion,"Fabncation Criteria for Shipping C-i-s," NUREG/CR-3854, Lawrence Livermore National Laboratory, Livermore, CA, March 1985.

NUREG/CR-3%6 U.S. Nuclear Regulatory Commission," Methods for impact Analysis of Shipping Contamers," 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 Ofnee, Washington, D.C., February 1990.

NUREG/CR-6007 U.S. Nuclear Regulatory Comminion," Stress Analysis of Closure Bolts for Shipping Casks," NUREG/CR-6007, U.S. Govemment Prmtag Office, Washmston, D.C., January 1993.

NUREG/CR-6322 U.S. Nuclear Regulatory Commission," Buckling Aaalysis 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, Wa=hiagtaa. D.C., March 1978.

NUREG-1617 2-16 DRAFT

1 i

RG 7.8 U.S. Nuclear Regulatory Cammasion,"Imad 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 l

Material for Ferritic Steel Shipping Cask C- ' =et Vessels with a j

Maximum Wall TN of 4 laches (0.im)," Regulatory Guide 7.11, U.S.

Government Pnntag Office, Washington, D.C., June 1991.

RG 7.12 U.S. Nuclear RS

-*~y Cammission," Fracture Toughness Criteria of Base A

Material for Ferritic Steel Shipping Cask Contamment Vessels with a Wall l

T1uckness Greater than 4 Inches (0.lm)," RgA-*=y Guide 7.12, U.S.

Government Prmtag Office, Washington, D.C., June 1991.

UCRL-ID-121673 Lawrence Livermore National Laboratory," Guidelines for Conducting Impact Tests on Shipping Packages for Radioactive Material,"UCRL-ID-121673, Lawrence Livermore National Laboratory, Livennore, CA, September 1995.

.l i

I DRAFT 2 17 NUREG-1617

3 THERMAL REVIEW 3.1 REVIEW OBJECTIVE The objective of this review is to verify that the thermal perfonnana of the package has been adequately evaluated for the tests specified under normal conditions of U+ t and hypotheucal accuirat conditions and that the package desip satisfies the thermal regarements of 10 CFR Part 71.

3.2 AREAS OF REVIEW The SAR must be re. viewed for adequacy of the desenption and evaluation of the thermal desip. Areas of review include the following:

3.2.1 Description of the Thermal Design l

l.

3.2.1.1 Packagmg Design Features j

l 3.2.1.2 Codes and Standards 1

3.2.1.3 Content Heat load Specification I

l.'

3.2.1.4 Summary Tables of Temperatu;es 3.2.1.5 Summary Table of Pressures in the Contamment Vessel 3.2.2 Material Properties and Component Specifications 3.2.2.1 Material Thermal Properties 3.2.2.2 Technical Specifications of Camanaants 3.2.2.3 Thermal Design Limits of Package Materials and C<- r-:- - ts 3.2.3 Thermal Evaluation Methods 3.2.3.1 Evaluation by Analyses 3.2.3.2 Evaluation by Tests 3.2.3.3 Temperatures 3.2.3.4 Pressures 3.2.3.5 Thermal Stress 3.2.3.6 Confirmatory Analyses 3.2.3.7 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 3.2.5.2 Cold 3.2.6 Evaluation under Hypothetical Accident Condstions 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 l

i j

DRAFT.

31 NUREG.1617

3.3 REGULATORY REQUIREMENTS Regulatory requir-mte of 10 CFR Part 71 applicable to the thermal review are as follows:

3.3.1 Description of the Therneal Design i

The packagmg must be described in suf5cient detail to provide an adequate basis for its evaluation.

[10 CFR 7131(aXI),10 CFR 7133(a)(5),10 CFR 7133(aX6),10 CFR 7133(b)(1),10 CFR 7133(b)(3),

j 10 CFR 7133(bX5),10 CFR 7133(bX7),10 CFR 7133(bX8)]

The SAR must identify established codes and standards applicable to the thermal design. [10 CFR 7131(c)]

- !he thermal design must not depend on a -ha+=I cooling system to meet the contamment requirements of 10 CFR 71.51(a). [10 CFR 71.51(c)]

33.2 Material Properties and Component Specifications The packaging materials and components must be described in suf5cient detail to provide an adequate basis for its evaluation. [10 CFR 71.31(a)(1) and 10 CFR 7133(aX5)]

3.33 Thermal Evaluation Methods

- 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 te sts of Subpart F. [10 CFR 7131(a)(2),10 CFR 7135(a), and 10 CFR 71.41(a))

3.3.4 Evaluation of Accessible Surface Te nperatures The package must be designed, constmeted, and prepared for shipment so that the accessible surface temperature of a package in still air at 38'C (100'F) in the shade will not exceed 85 *C (185 'F) 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 determme the effects of the conditions and tests under normal conditions of transport. The ambient temperature pr-Ang and following the tests must remam constant at that value between -29'C (-20'F) and 38'C (100'F) which is the most unfavorable for the feature under considerate 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(cXI) and 10 CFR 71.7)(c)(2) for heat and cold respectively are the primary thermal tests for normal conditions of transport. [10 CFR 71.71]

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 71.0.(a)(1)]

NUREG-1617 32 DRAFT

3.3.6 Thermal Evaluation under Hypothetical Accident Conditions The package design must be evaluated to determme the effects of the conditions and tests under a hypothetical accident. This accident includes a sequence ofincidents (impact, cmsh, puncture, thermal, and immersion) on a package. Except for the water immersion tests, the ambient temperature precedmg and following the tests must remam constant at that value between -29'C (-20'F) and 38'C (100*F) which is the most unfavor able for the feature under consideration. The initial intemal pressure within the contamment 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,800'C (1475'F) 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 Thermal Evaluation Methods 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.

3.5 REVIEW PROCEDURES The fo!!owing 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.

DRAFT 33 NUREG 1617

a The thermal review is based in part on the descriptions and evaluations presented in the General Information and Stmetural Evaluation sections of the S AR. Similarly, results of the thermal review are considered in the review of the SAR sections on Structural Evaluation, Contamment Evaluation, Shielding Evaluation, Criticality Evaluation, Operating Procedures, and Acceptance Tests and Maintenance Program. Examples of S AR 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 S AR 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 cor.::inment system The maximu.n 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, nr.d 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, includmg material properties and components, are identified.

3.5.1.3 Content Heat Load Specification 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.

Review the method in which the actual heat load is determined, and ensure that it is coraistent 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 detemuned. 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 radionuclides species.

NUREG 1617 3-4 DRAFT

l General structieral isdorenselon Evaluatlan Dunensions

- DeWon Cosoponent enssenals Decay heat

- Heatdissipesion 1f 1 I neraal Review Imading Evahistian Resides

• Decay beat

. Em.r--

- Hasttransfer

- Dogmintim g,,,,

- Phase changes

is. hon g, : 7.T

- toad combisons

- Malenalhcs Containment gas Modehng saventory

- Gas generation

- Preasure ennlysis I

lf lf lf U

Servetural Containment ShiaWing Criticanty Es ahiation Evalust6an Evaluation Evaluation Temperatures

- Temperatures

- Combusnon Combustion Preuures Pressures

- Decomposinon Decomposinon

- Gasinventory Dehydranon Dehydranon Melung Melang 1Y 1V Operathig Acceptance Tests and Procedures Maintmance Program

. Temperatures Temperatures Presures Pressures Heat transfer features Figure 3-1 SAR Information Flow for the Thermal Review.

3.5.1.4 Summary Tables of Temperatures Confirm that summary tables of the temperatures of package components including, but not limited to, the impact limiters, contamment vessel, seals, shielding, and neutron absorbers are consistent with the temperatures presented in the General Information and Stmetural 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 summarued temperatures should additionally include the elapsed time from the beguuung of the fire to the occurance of the maximum temperatures in each of the package components. Confirm that the temperatures and design temperature limit criteria for the package components are consistent throughout the appropriate sections of the SAR.

DRAFT 35 NUREG 1617

3.5.1.5 Summary Tables of Pressures in the Containment System Verify that summary tables of the pressure in the contamment system under the normal conditions of transport and hypothetical accident conditions are consistent with the pressures presented in the General Infonnation, Structural Evaluation, and Contamment Evaluation 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:

thennal conductivity specific heat a

density

+

thermal radiation emissivity of the package surfaces.

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-a'-vis normal) are specified. The thermal radiation absorptivity of the insolation incident on the package 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 ofless than unity in the S AR needs to be based on the demonstration that controls and procedures will be in place to ensure such a value throughout the package lifetime. 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.

Conftrm that the given fluid properties are adequate for evaluating thermal convection parameters such as the Prandtl number (a dimensionless number dermed as the ratio of the momentum diffusivity to the thermal diffusivity) which can be detemuned from the other thermal properties presented.

Confirm that the thermo-mechanical properties of any packaging material that may cause temperature-induced pressures and/or stresses within the package materials are presented. These properties include, but are not limited to:

coefficient of thermal expansion

+

NUREG 1617 3-6 DRAFT

modem ofclasticity Poisson's ratio.

The coefficient of thermal expansion is usually the linear coef5cient for isotropic solids and the volumetric enacient for Guids. For an isotropic matenal, the knear coef5cient is one-third the volumetric coef5cient.

Enswe that the structwal propertnes that effect thermal stresses are consinent with the values reponed in the Structural Evaluation sectaan if a package mataial is anisotropic, conGrm that the droctional propemes of, for example, the thermal conductivity, maMus ofelasticity, and the linear expansion coef5cie.it are provided.

Confirm that the t w s at which phase changes, decomposition, dehydration and combustion will s

occur are presented, along with thermal and thermo-mechamcal propertnes resulting from the change.

Confrm that the themal 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 sowces of the thermal properties used in the SAR are referenced Authoritative sources of matenal propemes data include, but are not limited to, those that reference expenmental measwements. In general, te.xtboot.s are an unacceptable source of material propemes data. If the applicant ewi w. tally measures the thermal propemes of the material and components used in the package, ensure that the e%.sts are performed under an approved quality assurance program.

If temperature Agavi-at thermal and thermo-mechanical properties are furnished by the applicant, the uncertamty (variance) in these values should be presented. Confum the appropriateness of the use of temperature dapaviaat thermal properties in an analysis of the package response to thermal loads. If the material propenies are not presented as a function of temperature, verify that the value consen'atively under-or over-predicts temperatures or stresses, as appropriate, compared to the equivalent temperatwe-dependent l

Property l

3.5.2.2 Technical Specifications of Components i

Verify that references for the technical specifications of the purchased package v-- ;---mts such as 0-rings, pressure relief valves, bolts, etc., are identified. Confum that any temperature constraints on the function of the components are identified (such as the allowable stress in a bolt). Verify that the trinimum allowable service temperature of all components is less than -40'C (-40'F).

l j

3.5.2.3 Thermal Design I imits of Package Materials and C=== =x l

Confum that the===n= allowable temperatures for each +:+_;-

that could affect the contamment, shielding, and criticality functions of the package are specified.

i Verify that the maximum allowable fuel / cladding temperature is justified. The justification should consider j

the fuel and clad materials, irradiation conditions (e.g., the absorbed dose, neutron spectrum, and fuel l

. burnup), and the shipping environment including the fill gas. Other necessr considerations include the l

elapsed time from removal of the SNF from the core to its placement into the transportation packaging, its DRAFT 37 NUREG 1617 l

l 1

time duration in the packagmg, and its post transport disposition. Examples of temperature limits include, but are notlimited to:

the temperature limit for metal fuel should be less than the lowest melting point eutoctac of the fuel l

the temperature limit on the irradiated clad in an inert gas envirnn==w as deteramed by creep or creep

=

rupture.

l l

Verify that the % 4.s range of the thennal and structural proportses for each package matenal exeerd the specified and prodzted temperature limits for the matenal 3.5J Thermal Evaluation Methods q

Thennal evaluations of the package can be performed by other analyses or tests, or by a combinaten of both.

Because of their mass and cost, and the difrzulty of decay-heat simulation, SNF packages are 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 detenmoed Review the Structural Evaluation and Thennal Evaluation sections of the SAR to decennme the response of the package to the normal conditions of transport and bypa*WI accident conditions. Verify that the wng-:=f=g models used in the thermal analyses are consistent wit these effects. For example, the package might have impact limiters or an external neutron stueld that would be damaged dunng the structural and thennal tests of 10 CFR 71.73.

3.5.3.1 Evaluation by Analyses Confum that the methods of thermal analysis are identified and suffriently described to permit a complete review and independent verification. The thermal analyses in the SAR can be based on simple calculations, sprah type analyses, or detailed @ simulations. De level of detail appropriate for each analysis, including assumptions, depends on many physical variables such as: the package materials, SNF decay heat, geometnc 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 assumpuans made in modelms heat sources and heat transfer paths and modes accurately repreeents the pbysical 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 constructed (Section 3.5.2.1)

=

uses appropriate expressions for both the convection and thermal radiation in enclosures within the package, and from the surfaces of the package to the environment correctly moorporates the appropriate specified temperature and thermal boundary conditions for the norms) conditions of transport and bypothetzal accident conditions.

NUREG 1617 3-8 DRAFT

I l

Natural (or free) convection will occw from the swface of the package to the envirn==ent for all conditions except durms the 30-nunute Gre. Forced convection based on, for example, the flame velocity that occurs l

dunng the 30-mmute Src is followed by natwal convection dwing the post-6rc cooldown. The flame velocities are discussed by Bwgess and Fry (1990), Burgess (1987), and Schneider and Kent (1989). The flow regune (lammar or turbulent) for natural convection will be detenmned by the dunensionless Rayleigh msnber (product of the ratio of buoyancy force to the viscous force and the ratio of the momentum diffusivity te the thennel diffusivity). De flow regune for forced convection will be deteramed by the dunemaionless Rt ynold's number (ratio of the inertial force to the viscous force). For steady-state normal conditions of trasport (includag the case for detenmmng the accessible swface temperstwe), con 6rm the presence of eithv the Rayleigh number for the air at the swfaces of the package or infonnation sufEcient to calculate the l

Ra3%gh number Verify that the expression for the netwal convection coef5cient for heat transfer from the surface of the package to the environment is presented and appropriate for the Reynold's number (if forced convection) or the Rayleigh number (if netwal convection).

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 surfenes dunng 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 maxmuzmg the calculated component temperatwes for normal conditions of transpost and hypothetical accident conditions.

In the can of computer analysis, the applicant may use standard off-the-shelf software or develop a computer code to perfona a specific analysis. Verify that the code has been benchmarked and is maintamed and operated under a quality assurance program. Verify that the code has been appropriately used. Ensure that

. the SAR includes code input and output files for each analyzed thennal condition that are complete enough to permit detailed review of their appropriateness and accuracy.

3.5.3.2 Evaluation by Tests For those results determmed 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 h pothetical accident conditions 3

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 -29'C ( 20'F) to 38'C (100'F) for normal conditions of transport and both 38'C (100*F) and 800*C (1475'F) for hypothetical accident conditions. Confirm that the facilities can simulate the specified heat-transfer boundary conditions:

incident beat fluxes equivalent to or exceedmg the specified insolation reqmrements during the normal conditions of transport or the post-fire environment for hypothetical accident conditions incident heat Duxes equivalent to or excendag the specified convective and radiative heat transfer environment, including specified anissivities, for a minimum 30-minute period w,r-g the hypothetical accident condition fire an environment that assures an adequate supply and circulation of oxygen for initiating and naturally a

termmatmg the combustion of any burnable package M--;-:--st.

DRAFT 39 NUREG 1617

Confm that the test package, with a simulated package contents and any attached test instrumentation or hardware, adequately simulates the thennal 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 mannar that they do not unduly aff' ct local temperatures o

=

in locations where manmum W.h are -a

• =d and where other temperatures need to be

=

deteramed in locations that permit reasonable interpolation or extrapolation of measured temperatures for esumatmg temperatures in unmonitored regions of the package.

The applicable components include, but are not limited to, the contamment vessel, fuel basket, seals, radiation shielding, criticality controls, and impact limiters. Confm 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 a number of references (Gregory et al. (1987), Hovingh and Carlson (1994), and VanSant et al. (1993)).

3.5.3.3 Temperatures Confm that the maximum temperatures of the package components for the normal conditions of transport and hypothetical accident conditions are presented. Component temperatures of special interest include, but are not limited to, those of the contamment vessel, seals, shielding, criticality controls, and impact limiters.

Confm that the volume averaged temperature of any gases in the contamment vessel is determined if the temperatures are determmed by test, confinn that the measured and the corrected maximum temperatures of the package components are detenmned. Verify that these maxunum temperatures are consistent with the General Infonnaten and Structural Evaluation sectmas of the SAR.

3.5.3.4 Pressures Confm that the SAR identifies all sources of gases released into the contamment vessel following the closure of the vessel. Such gases could include, but are not limited to:

fill gas from the SNF rods fission product gases released from the SNF NUREG-1617 3-10 DRAFT

saturated vapor from matenal in the contamment vessel including water vapor desorbed from the contamment system componets or the package contents belima from the a decay of the SNF contents hydrogen and other gases from radiolysis or chemical reactmos (e.g., sodian-water)

• hydroge and other gases from the dehydration, combusten, or decomposition of package components.

Verify that the determmation of the presswe is based on the increase in temperatwrs and the various volumes ofgases in the cont===aat vessel from the time of closwe to the mammum temperature for the normal conditions of transport and hypothetical aces conditions. Review the method for determmmg the contribution to the presswe in the containment vessel from the gases released into the vessel Verify that the total presswe in the containment vessel includes the initial fill gas plus all the gases released from 100% of the fuel rods (includmg the fuel pin fill gas plus 30% of the fission product gas associated with the fuel burnup) plus any other gases released into the vessel in the year following closure of the vessel (the MNOP).

Verify that the total pressure is determmed for the normal conditions of transport and hypothetical accident conditions and is consistent with the Structural Evaluation section of the SAR.

3.5.3.5 Thermal Stresses j

There are two sowces of thermal stresses These stresses can be caused by either spatial temperature gradients in the package components or by interference Wre. components due to different thermal expansions of the w-pe..a.ts. For the SNF heat flux and high thermal conductivities of the contamment vessel materials of construction, the spatial temperature gradients (and hence the thermal stresses) in the vessel are small.

l Confum that an appropriate method for the estunation of the temperature gradients and thermal stresses in the containment vessel is described in the SAR. Verify that an upper bound estimate of the mammum temperature gradient in the primary contamment vessel and corra=paaAng thermal stress is deteramed for the normal conditions of transport and hypothetical accident conditions and is consistent with the Structural l

Evaluation section of the SAR.

Three cases should be investigated for interferences kre. package components assembled at " room"

. temperature. These include:

a steady-cold (-40*C [-40 *F]) environment with mammen SNF decay heat without insolation I

i a a;w y-hot (38'C [100'F]) environment with man =um SNF decay heat with insolation a hypothetical acculent post-fire cooldown with mammum SNF decay best and insolation i

Confirm that the dimencions of the package components for each of the above cases are presented in the SAR. Enswe that the clearances and interferences between the components of the package for each of the above cases are provuled Verify that an appropriate method for estunation of the stresses from the t

interferences between components is described. Verify that the stresses from any interferences bEew.

j components are consistent with the Structural Evaluation section of the SAR.

DRAFT 3-11 NUREG-1617

3.5.3.6 Confirmatory Analyses Perform an ;%t analysis to confirm the temperatures deter==ad by the applicar.t. The rigor required of the mafirmatory analysis will depend on the size of the margin betweesi the maximum package component temperatures detenmoed by the applicant and the maximum te. hare limit specified for a material or component or the regulatory limit detenmned by the type of shipment. A conservative method of analysis of l

the fire portion of the by-M-1 accident is to mathematically apply an 800'C (1475'F) surface i

L temperatwe for 30 mmutes to the package with the appropriate initial : r..here distributaan and masent l

decay heat. This will ah==ane the questions about the flame velocity and itt effect on the convection heat input into the package. The analysis will still require the appropriate boundary conditions during cooldown to i

entendate the movim== component temperatures 3.5.3.7 Effects ofUncertamties Verify that the thermal evaluations appropriately address the effects of uncertamties in thermal and structural propertics of materials, test conditions and diagnostics, and analytical methods, as applicable.

3.5.4 Evaluation of Accessible Surface Temperatures Detenmne that the thennal 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 swface of the package is equal to the convective and radiative heat loses to the environment at an ambient temperstwe of 38'C (100*F).

If the maxunum surface temperature of a package exceeds the regulatory limit, a perw.ael ba Tier can be placed around the package. This personnel barrier becomes the accessible package surface. The thermal i=, i== of the barrier should be considered when determuung the package temperatwes for normal conditions of transport, but should be neglected during the hypothetical accident.

Confum that the maximum accessible surface temperature detenmned by the applicant is consistent with the General Information section of the SAR.

When appropriate, perform an independent analysis as describuct in Section 3.5.3.6 to confirm the maximum accessible surface temperature detenmned by the applicant.

Ensure that the maximum temperature of the accessible package surface is less than 85'C (185 *F) 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 Confum that the thermal evaluation demonstrates that the tests for normal conditions of transport do not result in significant reduction in packagmg effectiveness, including:

degradation of the best-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 NUREG 1617 3 12 DRAFT

changes in the packagmg that affect naatamaient, shielding, or criticality such as thermal W5ition or melting ofmaterials ability of the packagmg to withstand the tests under hypothetical accident conditions.

The peak temperstwos of the components doing normal conditions of transport will, in general, be the highest in the center of the package, decreasing to the lowest at the outer package surface. For the case where

' the surface heat flux due to the content decay heat is small relative to the insolation, the surface ^w.ure will be driven by the balance between the insolaten and the convective and radiative heat losses to the envirnamant. If the ratio of the content decay heat load to the thermal conductance from the cantant boundary j

to the package surface is small, the temperature of the package imernals is virtually the same as the surface of the M =

3.5.5.1 Heat Verify that the maximum ^w. ares of the package components are determmed during normal conditions of transport when the package is in 38'C (100'F) still air with insolation, according to the table in 10 CFR 71.7)(c)(1) and the maximum allowable content best load.

The temperatures, presmres, and thermal stresses should be reviewed as described in Sections 3.5.3.3, 3.5.3.4, and 3.5.3.5. Perform an i%t analysis as described in Section 3.5.3.6 to confirm the maximum component temperatures determmed by the applicant.

l l

Confirm that the maximum component temperatures are within the specified mammum allowable temperature i

limits of the component or material of construction as specified in Section 3.5.2.3.

3.5.5.2 Cold Ensure that the mmimum temperatures of the package components during normal conditions of transport l

when the package is in 40'C (40'F) still air without insolation, and the miniinum allowable content heat l

load is consistent with the Structural Evaluation section of the SAR. If the SAR does not restnet the muumum heat load, the package should be considered at a uniform temperature of 40*C (40*F).

)

The thermal stresses should be reviewed as described in Section 3.5.3.5.

Verify that a temperature of 40'C (-40'F) is compared to the specified mammum allowable Q. ore limit of a component or material of construction.

l 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 fee test, the peckage 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 thennal evaluations I

DRAFT 3 13 NUREG-1617

under hypothoucal accident conditions is consistent. with these results from the Structural Evaluation section of the SAR.

Verify that the SARjustifies the most unfavorable initial ambient temperature betw -29'C (-20*F) and 38'C (100 *F). Unless the package is susceptible to increased structural damage at lower temperatures, the i

initial ambient temperature should be 38'C (100'F). Verify that the initial steady-state tasaperature distribution is r=sistent with the results from the thermal evaluauons under normal conditions of ^

speit.

Confirm that the initial internal pressure of the package is the MNOP unless a lower internal pressure, consistent with the initial ambiet temperature,is more unfavorable. Similarly, confirm that the internal heat load of the SNF contents is at its mammum allowable power unless a lower power, consistent with the

{

m-ere and pressure,is more unfavorable.

Although not explicitly prescribed in 10 CFR 71.73(c)(4), insolation should be applied to the package during the post fire cooldown (60 FR 50247).

3.5.6.2 Fire Test Verify that the package surface and fn emissivity are greater than or equal to 0.8 and 0.9, respectively.

Verify that the package is exposed to the 800*C (1475'F) environment for a mmunum of 30 minutes. Verify that the flame velocities are specified and appropriate for the hydrocarbon fn. Verify that the appropriate correlation for convection in the fu is presented and used as a boundary condition.

Verify that, following the 30-minute exposure to the 800'C (1475 F) environment, the package is allowed to cool in the 38'C (100*F) still air and radiate to a 38'C (100*F) environment. Verify that the appropriate insolation is applied to the package surfaces during this post fn cooldown period. Verify that the SNF decay heat load specified in the General Information section of the SAR is included in the thermal model.

To the extent that the regulatory conditions were not achieved during the evaluation, verify the applicant has presented an acceptable treatment of the calculated or measured temperatures to correct the results to the specified regulatory conditions.

If the evaluation of the package under hypothetical accident conditions was performed by physical tests, the thermal portion of the tests should be reviewed as described in Section 3.5.3.2.

3.5.6.3 Maximum T r. ares and Pressures Confirm that the mammum temperatures of the package components during the thermal porten of the hypothetical accident conditions, as determmed above, are presented. The maximum temperatures in the components will occur following cessaten of the fire, with the delay time increasing with the distance inward frem the package surface.

The " r. ares and pressures should be reviewed as described in Sections 3.5.3.3 and 3.5.3.4.

Confirm that the mammum Q. tores in the components determmed above are compared to the specified maximum allowable temperature limits of the component or material of construction. Verify that the maximum temperature the package components are determmed to experience during the thermal portion of a NUREG-1617 3-14 DRAFT

1 regulatory hypMal accident are within the specified allowable short-term temperature limits of the components or material of construction.

I 3.5.6.4 Maximum Thermal Stresses ne maxunum interference between components in a package during a hypothetical thermal accident usually occur durmg the post fire cooldown Where the components are concentric, the tensile stresses occur in the outer mmpaaent while the stresses in the inner components are usually compressive.

De thennal stress abould be reviewed as described in Section 3.5.3.5.

3.5.7 Appendia The anM may include a list of references, copies of any applicable references not generally available to the reviewer, computer code descriptions, input and output fdes, test facility and instrumentation descriptions, test results, special analyses, and other appropriate supplemental information.

I 3.5.7.1 Justification for Assumptions or Analytical Procedures Confirm that the applicant has stated andjustified all assumptions used in the evaluation of the package.

Review the appropriateness of and justification for the applicant's assumptions and analytical procedures.

I 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 (fmite 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 materist models (isotropic, anisotropic, etc.) are described. Verify that the various types ofinitial 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 (coclosure radiation with view factor calculation, thernal stress analysis, etc.) that are applicable to the applicant's thermal evaluation are identified and described. Review that the computer I

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 (maxunum accessible surface ;

g.h.e, normal conditions of transpost, calculation ofinitial temperature distribution for hypothetical accident, initial temperature distribution for analysis of thermal hypothetical accident) analywd using a computer code. Confirm that the applicant has submitted annotated omput files, as applict.ble, for each problem (maxunum accessible surface temperature, normal conditions of transport, calculation ofinitial temperature distribution for hypothetical accident conditions, and temperature

. distr;bution histones for the thermal hypothetical accident condition during and following the 30-mmute fire, unul all the package component temperatures have reached their maw =a).

I DRAFT 3 15 NUREG 1617

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 notlimited to:

the type of facility (fwnace, pool fire, etc.)

a the method of heatmg the package (gas bwners, electrx:al heaters, etc.).

The description of a furnace facility should include the volume and emissivity of the furnace misnor as well as the method of measuring the interior ;- r.. ore. The method of measunng the oxygen macmtration and

==iataining an oxygen concentration of ~ 21 percent by volume in the fwnace atenor throughout a test shall be described.

For a pool fire facility, the size of the fue relative to the size of the package shall be specified. Verify that the fire amansion conforms to the regulatory requirement that the fire thickness extend honzontally 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. Resiew that the ansiysis of the heat loss from the package through the support is appropriate.

Confirm that the sensors used to measwe heat flux and temperatwe 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, shadowmg 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 calonmeters are used to measure heat flux, the calonmeter readmgs are corrected to account for the difference in thermal inertia between the calonmeter and the package. Verify that

)

the method of correction of the calonmeter reading is presented and review the method for appropriateness.

j 3.5.7.5 Test Results l

Verify that test measurements including temperatures (or temperatwe 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 conected results) are continued until steady state occurs (for tests for normal conditions of transport) or until the maxunum temperature occus in all the package components (for test of the thermal port on of the regulatory hypothetical accident).

! Verify that photographs of the package amycz.ir.ts prior to and following the tests are p esented. Verify that

)

photographs of regions of components with thermal damage (such as charnng of the insulation, damage to O-rings, etc.) are' presented.

i

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l I-NUREG 1617' 3-16 DRAFT

3.5.7.6 Applicable Supportag Documents or Specifications i

i Verify that the applicable sections from reference documents are included. 'Ibese h-ts 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 ofnon-ASME approved materials used in the package.

3.5.7.7 Special Analyses

)

l Frequently, thnnally driven special processes will occur in a p% These processes may include, but are notlumted to:

I generation of gases within the matainmaat system i

• effects of phase changes on package materials combustion, h- ;-wition or dehydration of package materials.

The production of gases (e g., hydrogen by radiolysis) or thermal %~=ition of materials (e.g., a neutron shic!d) may occur in the packaBe. Phase changes of material resulting in a decrease of the material density occumng 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.

Confum that the applicant has identified all thermally driven special processes that will occur in the package.

Verify that the applicant has stated andjustified all assumptions used in the quantification and evaluation of these special processes. Review the appropriateness of andjustification for the applicant's assumptions and analpical procedures. Verify that the results are incorporated in the appropriate subsections of the Thermal Evaluation section.

An example of a special process is the production of hydrogen by the radiolysis of material in the package.

The hydrogen generation rate (molecules per unit time) is the product of the "G" factor for the material (molecules of hydrogen per 100 eV ofionizing radiation) and the sum of the ionizing radiation dose rates for all types of the radiation (beta, secondary radiation, and gamma) absorbed by the material (eV per unit time).

For SNF contents with a half-life long relative to the time the material is enclosed in the containment vessel, the hydrogen produced is simply the product of the hydrogen generation rate and contamment time.

Other supplemental calculations may be required to support evaluations presented in the Thermal Evaluation j

section. Verify that all such special analyses meet the goals discussed in Section 3.5.3.1.

3.6 EVALUATIONFINDINGS The thermal review should result in the following fmdings, as appropriate:

i 3.6.1 Description of the Thermal Design 1

i The staff has reviewed the package description and evaluation and concludes that they satisfy the thermal requirements of 10 CFR Part 71.

J DRAFT 3 17 NUREG 1617 1

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

)

reqmrements of10 CFR Part 71.

3.6.3 Thermal Evaluation Methods The staff has reviewed the methods used in the thermal evaluation and concludes that they are described in sufficient detail to permit an iPt 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.

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 composest temperatures will not extend beyond the specified allowable limits during nonnal 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 ofRadioactive Material (PA TRAM), Vol. II.,

STL?UB/718, IAEA, Vienna,1987, pp. 423-431.

Burgess and Fry 1990 Burgess, M.H., and CJ. Fry," Fire Testing for Package Approval," Int. J.

Radioacrtve Materials Transport, Vol.1, No.1, Nuclear Technology Publishing, Ashford, Kent, England,1990, pp. 7-16.

60 FR50247 U.S. Nuclear Regulatory Comrmssion," Compatibility With the International Atomic Energy Agency (IAEA)," Federal Register, FR 50247, U.S.

Government Printing Office, Washington, D.C., September 28,1995.

Gregory et al.1987 Gregory, JJ., R. Mata, and N.R. Keltner," Thermal Measurements in a Series of Large Pool Fires," SAND 85-0196, TTC-0659, UC-71, Sandia Natmaal Laboratories, Albuquerque,NM, August 1987.

NUREG-1617 3-18 DRAFT

Havingh 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.

hhh and Kent Schneider, M.E., and L.A. Kent," Measurements of Gas Velocities and 1989 Tws s in a large Open Pool Fire," Fire Technology, Vol. 25, February 1989,pp.51-80.

VanSant et al.1993 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.

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DRAFT 3-19 NUREG-1617

4 CONTAINMENT REVIEW d.1 REVIEW OBJECTIVE The objective of this review is to verify that the package design satisfies the contamment requirements of 10 CFR Part 71 under normal conditions of transport and hypothetical accident conditions.

4.2 AREAS O,F REVIEW The SAR should be reviewed for adequacy of the description and evaluation of the contamment 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 Reqmrements 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 Contamment Criteria 4.2.2.3 Compliance with Contamment Criteria 4.2.3 Containment under Hypothetical Accident Conditions 4.2.3.1 Pressurization of Containment Vessel 4.2.3.2 Contamment Criteria 4.2.3.3 Compliance with Containment Cnteria 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)(S),10 CFR 71.33(b)(1),10 CFR 71.33(b)(3),

i 10 CFR 71.33(b)(S), and 10 CFR 71.33(b)(7)]

I The SAR must identify established codes and standards applicable to the contamment design.

l

[10 CFR 71.31(c)]

'Ihe 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)]

i The package must be made of materials an construction that assure that there will be no significant l

chemical, galvanic, or other reaction. [10 CFR 7L43(d)J l

DRAFT 41 NUREG.1617 l

l L

L

I 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(c))

Damaged spent nuclear fuel with plutonium in excess of 20 Ci per package must be packaged in a separate inner contamer. [10 CFR 71.63) 4.3.2 Containment under Normal Conditions of Transport The package must be evaluated to demonstrate that it satisfies the contamment reqmrements of 10 CFR Part 71, Subpart E, under the conditions and tests of Subpart F. [10 CFR 71.3 )(a)(2),10 CFR 71.35(a), and 10 CFR 71.41(a)]

)

A package must meet the contamment 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 h=='=-v on filters or a f

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 71.5 )(a)(2) for hypothetical accident conditions, with no dependence on filters or a mechanical cooling systan. [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 codes and standards applicable to the contaimnent system should include Section Ill, Division 3, ASME Boiler and Pressure Vessel (B&PV Division 3) Code for design and ccesvu: tion; and ANSI 14.5 for leak testing.

In addition to meeting the requirements for normal conditions of transport and hypothetical accidents conditions identified in Sections 4.3.2 and 4.3.3, damaged spent nuclear fuel (SNF) should be packaged in a separate inner container, which would be the package's second contamment system (double contamment).

The codes, standards, and criteria applicable to the second containment sptem should generally be the same as those applicable to the primary containment system.

4.4.2 Containment under Normal Conditions of Trnagm 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 contamment vessel is scaled and under normal transport conditions. The SAR should identify the allowable normal conditions of transport volumetric leakage rates in accordance with ANSI N14.5.

NUREG.1617 42 DRAFT

4.4.3 Contain-t under Hypothetical Accident Conditions In addition to the regulatory reqmrements 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 hyposetical accident conditions. The SAR should identify the allowable hypothetical accident conditmos volumetric leakage rates in accordance with ANSI N14.5.

4.5Property "ANSI code" (as page type) with input value "ANSI N14.5.</br></br>4.5" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. REVIEWPROCEDURES The following procedures are generally applicable to the contamment review of all SNF ti.-portation Packages The contamment 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===

established to evaluate the packaging against applicable 10 CFR Part 71 reqmrements. Similarly, results of the contamment 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 Containmen: 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 contamment boundary, including, as applicable, the containment vessel, welds, seals, lids, cover plates, valves, and other closure devices. Ensure that all components of the containment boundary are shown in the drawings.

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)

Bolt torque required to maintain positive closure j

~* - Maximum allowable temperatures of components, including seals l

Temperatwes of components under normal conditions of transport and hypothetical accident conditions.

l 1

I 1

i DRAFT 4-3 NUREG-1617 1

L____

l j

i Strucensral General hermal Evelessen laserseeen -

Evahnseen i

. Deformabon.

• Ulanenseons Tsunperature

. Chenucalandsalvanic

.e aseenals

$ Pressures reactons

. r==sh boundary

. Gasinventory

. Comissescom6tions

. Comesats

. ABomablelealase ruess I

1 V 1 V 1 Y Containment Review Dessetydenof.

Canenhisment imeder Norsnel Comenhusent asider W8 Casednmient sysens Cand6 dens af Transport Aaddent Candleens

• hboundary

, p,,,,,ng,,;,,,p

. Pressunzationof Componses acalasnment vessel conennment vessel Penetranon

• Conianmentcntens Closure

. Comai.nmentsniena

,c

.gs

. c.

, gs heds consenment enteria e namnmententeria Special requirements for demased spent nuclear fuel l I 1 I Operating Assapeance Tests and Precedures Maintenance Prograni

. Closure requirements

. Fabricanonveri6 canon Assemblymen 6 canon leslass rate.

leakage rate

. Ponodic ven6 cation leakage rete Figure 4-1 SAR Information Flow for the Containment Review.

Verify that all containment boundary penetrations and their method of closure are descJbed in detail.

Performance specifications for components such as valves and 0-rings should be documented, and no device may allow continuous venting. Any valve or similar device on the package must be protected agair.st unauthorned 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.

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 testag of their covw. plate seals, providing such seals form part of the contamment boundary.

Deterame 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 %-04). If penetrations are closed with two seals (e.g., to enable leakage testing), verify which seal is def'med as the contamment boundary. Ensure that seal grooves are appropriately sized. Verify that the temperature of contamment NUREG-1617 44 DRAFT

1 1

1 boundary seals will remam within their specified allowable limits under both normal conditions of transport

)

and hypotheocal accident conditions.

I

~ Confirm that the contamment system is securely closed by a positive fastening device that cannot be opened l

umasaneionally or by a pressure the may arise within the pday j

l 4.5.1.2 Codes and Standards Verify that the containment system is in full compliance with ASME Section III, Division 3, including material, design, fabrication, exammation, testing, inspection, and certification. This includes an agreement 1

with an Authorized Inspection Agency to provide inspection and audit services for the Design Owners, Packagag Owners, and Class TP Certifx:ste Holders.

L 4.5.1.3 Special Req

..sts for Damaged Spent Nuchar Fuel Verify that packages designed for the transport of damaged SNF include packaging of the damaged ftel in a i

separate inner contamer (second contamment system) that meets the requirements of 10 CFR 71.63/b).

Damaged fuel is typically not exempt from the double contamment requirements of 10 CFR 71.63(b) for

. plutonium shipments. The review procedures for the second containment system are generally the same three l

for the primary contamment system.

i

(

4.5.2 Containment under Normal Cenditions of Transport 4.5.2.1 Pressurization of Contaimnent Vessel l

l.

Verify that the manmum normal operating pressure is consistent with that detemuned in the Thermal Evaluation section of the SAR. The pressure in the containment vessel should be based on the conditions of-i the package under normal transport conditions, including temperature, release of gases and volatiles from fuel i

rod claddmg breaches, vaporization of contents, etc. (NRC IN 84-72).

4.5.2.2 Contamment Criteria l

Detailed guidance on procedures for determmmg the containment criteria is provided in NUREG/CR-6487.

i Confum that the SNF contents are fully described, including fuel type, fuel amount, percent enrichment, l

burnup, cool time, decay heat, etc. Confum that the contents evaluated in the Contamment Evaluation section of the SAR are consistent with those presented in the General Information section of the SAR.

l Verify that the SAR identifies the constituents which comprise the releasable source term, including L

radioactive gases, volatiles, and powders. For SNF packages, the releasable source term is e- ;-:=! of crud l

an the outside of the fuel rod cladding that can become aerosolized, and fuel fmes, 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 packagmg (frorn previous shipments) typically can be ignored in the determmation of the releasable source tenn, this issue should be addressed. Reasonable boundmg values for the effective surface activity density (Ci/cm ) of the crud on fuel rod clads are based on ew;..stal determmations. A 2

computer code, such as ORIGEN2 (ORNI CCC-371),is used to identify the radionuclides present for a

'given percent fuel enrichment, bumup, and cool time. Afler deternumng the effective A values for the crud, 2

fmes, gases, and volatiles individually, the effective A, of the releasable source term can be determmed by l

DRAFT 45 NUREG-1617 l

using the.clative release fraction for each contributor. The release fractions and effective specific activitic s for the various releasable source term contributors for SNF with an initial enrichment of 3.2%,

a burnu p of 33,000 MWU/MTIHM, and a cool time of 5 years are given in Table 4-1. The release fraction s presented in Table 4-1 are considered conservative and have been developed from reasoned argumert and experimental data (NUREC/CR-6487).

Table 4-1 C *-=.:e Fractions and Specific Activities for the Contributors to the Releasable Source Tera for Packages Designed ta Transport Irradiated Fuel Rods.

~

PWR BWR Normal Hypothetical Normal Hypothetics conditions I

conditions I accident ohranspod accident ohranspod condidons Variable conditions i

Frection of crud that spalls-off of rods, fc 0.15 1.0 0.15 1.0 i

f 2

4 4

4 Crud surface activity, Sc [Ci/cm ]

140 x 10 140 x 10' 1254 x 10 1254 x 10 l

Mass fraction of fuel that is released as fines 3 x 10 3 x 10 3 x 10 3 x 10 l

3 3

4 3

due to a cladding breach, fr Specific activity of fuel rods, Ag [Ci/g]

0.60 0.60 0.51 0.51 Fraction of rods that develop cladding 0.03 1.0 0,03 1.0 breaches, fa Fraction of gases that are released due to a 0.3 0.3 0.3 0.3 cladding breach, fo 3

3 3

3 Specific activity of gases in a fuel rod, Ao 7.32 x 10 7.32 x 10 6.28 x 10 6.28 x 10

[Ci/g]

Specific activity of volatiles in a fuel rod, Av 0.1375 0.1375 0.1794 0.1794

[Ci/g]

d d

d d

Fraction of volatiles that are released due to a 2 x 10 2 x 10 2 x 10 2 x 10 cladding breach, fy Based on the mass density, effective specific activity, and effective A of the releasable source term, 2

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 undet normal transport conditions is converted into a reference air leakage rate under standard leakage test conditions according to ANSI N14.5 and NN EG/CR-6487.

Verify that the following maximum permissible leakage rates are determined in acc; ance with ANSIN14.5:

Fabrication verification Periodic verification NUREG-1617 4-6 DRAFT

Assembly (pre-shipment) verification.

+

4.5.23 Compliance with Contamment Criteria Confmn that the SAR demonstrates that the package satisfies the contamment requirements of

.10 CFR 71'51(a)(1) for nonnal 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 contamment 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.

4.53.1 Pressurization of Containment Vessel

- The pressure in the contamment 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. (NRC IN 84-72).' Verify that this pressure is consistent with that deternuned in the Thermal Evaluation section of the SAR.

4.53.2 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 contamment requirements under hypothetical accident conditions. Verify that the temperatures, pressure, and physical conditions of the package (including *he contents) are consistent with those determmed 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 determmed in Section 4.5.2.2 for normal conditions of transport. In the rare event that this is not the case, the contamment criteria for the fabncation, penodic, and assembly verification tests should be based on the hypothetical accident leakage, rather than normal conditions of u-c+e6.

The contamment reqmrements 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 A in a week.

2 The remaining radioactive materials shall not exceed A in a week.

2 DRAFT 4-7 NUREG 1617 l

l

4.5.3.3 Compliance with Containment Criteria Confirm that the S AR 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 l

l 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 containn'ent cystem 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 requireruents 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 4-8 DRAFT

4.7 REFERENCES

- ANSIN14.5 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.

B&PV Division 3 Code American Society of Mechanical Engmeers,"ASME Boiler and Pressure Vessel Code, Secten III, Division 3, Contamment Systems and Transport Packagings For Spent Nuclear Fuel and High level Radioactive Waste," New York, NY, 1997.

NRC Bulletin %04 U.S. Nuclear Regulatory Commission," Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks," OMB No. 3150-0011, Bulletin %04, U.S. Government Printing Office, Washington, D.C., July 5, 19%.

NRC IN 84 72 U.S. Nuclear Regulatory Commission," Clarification of Conditions for Waste Shipments Subject to Hydrogen Generation," Information Notice 84-72, SSINS No. 6835, U.S. Government Printing Office, Washington, D.C., September 10, 1984.

NUREG/CR-6487 U.S. Nuclear Regulatory Commission,"Contamment Analysis for Type B Packages Used to Transport Various Contents,"NUREG/CR 6487, Lawrence Livermore National Laboratory, Livermore, CA,1996.

ORNL-CCC-37i Oak Ridge National Laboratory,"ORIGEN2.1: Isotope Generation and Depletion Code-Matrix Exponential Method," CCC-371, Oak Ridge, TN, 1991.

l DRAFT 4-9 NUREG-1617

l l

l l

5 SHIELDING REVIEW 5.1 REVIEW OBJECTIVE l

The objective of this review is to verify that the package desip satisfies the external radiation requirements of 10 CFR Part 71 under normal conditions of transport and hypothetical accident conditions.

1 I

5.2 AREAS OF REVIEW j

1he SAR must be reviewed for adequacy of the description and evaluation of the shielding desip. Areas of reywwinclude the following:

5.2.1 Description of the Shielding Design 5.2.1.1 Packaging Desip Features 1

5.2.1.2 Codes and Standards 5.2.1.3 Summary Table of Maximum Radiation levels 5.2.2 SourceSpecification 5.2.2.1 Gamma Source 5.2.2.2 Neutron Source 5.2.3 ModelSpecification 5.2.3.1 Configuration of Source and Shielding 5.2.3.2 Material Properties i

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:

l 5.3.1 Description of the Shielding Design l

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

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)]

The S AR must identify established codes and standards applicable to the shielding desip. [10 CFR 71.31(c)]

l DRAFT 51 NUREG 1617 i

k

l 5.3.2 SourceSpecification The contents must bh 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 chenucal and physical form. [10 CFR 71.31(aXI),10 CFR 71.33(bX1),10 CFR 71.33(bX2), and 10 CFR 71.33(bX3)].

5.3.3 ModelSpecification The package must be described and evaluated to demonstrate that it satisfies the shielding reqmrements 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 reqmrements 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 extemal 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 summarued in Table 5-1.

Under the tests specified in 10 CFR 71.73 (hypothetical accident conditions), the extemal radiation levels at I m (40 in) from the package surface must not exceed 10 mSv/hr () 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 ModelSpecification 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 i_%t review, with confumatory calculations, of the package shielding design.

NUREG-1617 5-2 DRAFT

l 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))

Thh table must not be used as a substitute for DOT or NRC regulations on transportation of radioactive materials.

Transport V6kk Use:

Exclustre Transport Vdkk Type:

Open (fist-bed)

Open w/ Enclosure

• Closed Package (or freight container) Ihmits:

Extemal Surface 2 mSv/hr 10 mSv/br 10 mSv/hr (200 anrem/br)

(1000 mrem /br)

(1000 meema/br)

Roadway or Railway Vehicle (or freight container) Ihmits:

l Any point on the outer N/A N/A 2 mSv/hr l

surface ofenclosure (200 mrem /br) l l

Vertical planes 2 mSv/hr 2 mSv/hr N/A projected from outer (200 mrem /br)

(200 mrem /br) edges l

Top of.

load: 2 mSv/hr enclosure: 2 mSv/hr vehicle: 2 mSv/hr (200 mresm/br)

(200 mrem /br)

(200 mrem /br) 2 m (80 in) from.

verticalplanes:

vertical planes:

outer lateral surfaces:

0.1 mSv/hr 0.1 mSv/hr 0.1 mSv/hr (10 mrem /br)

(10 mremlbr)

(10 mrem /br)

Underside 2 mSv/hr (200 mrem /br) j Occupied position 0.02 mSv/hr (2 mrem /br)'

I l

a.

Securely attached (to vehicle), access-limiting enclosure, package personnel barTiers are considered as enclosures.

b.

Does not apply to private carrier wearing dosimetry if under radiation protection program satisfying 10 CFR Pan 20 or 49 CFR 172 Subpart 1.

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 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.

DRAFT 53 NUREG 1617

The shielding review is based in part on the descriptions and evaluations presented in the General l

l Information, Structural Evaluation, and Thermal Evaluation sections of the S AR. Similarly, results of the shielding review are considered in the review of the SAR sections on Operating Procedures and Acceptance l

Tests and Maintenance Program. Examples of SAR information Dow into, within, and from the shielding l

review are shown in Figure 5-1.

I l

structurnt c wns nerwal l

Evaluatim Indonnation Evaluation Deformation

- Dunensions

- Combustion CnsMag/ puncture

- Componentmatenals

- Decomposition

- Extrumon Conients

- DehWanon

- Displacemera of McInng contents and skeldmg V

V V

Shielding Review Source Tenns Attenuation Radiation levds

- Gamrna Material properties Gamma

- Netaron

- Modehng

- Neutron

- Stucidmg analyss Gamma Neutron 1f 1P Operating Acceptana Tasts and Procedures Maintenance Program Radianon levels SLeldirs tests Streammg paths Slucidmg material Pre-dupment surveys spect6 cations Figt.rc 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, tolerr.nces, and densities of material for neutron and gamma shielding, including those of structural or thermal components considered in the shielding evaluation Concenentions of neutron absorbers NUREG 1617 5-4 DRAFT

l l

Structural components that maintain the contents in a fixed position within the package Dunensions 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 leveis apply should be given in the table, since the gamma or neutron contributions could be greatest at different fuel specifications. Appendix A contains a fill-in-the-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 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 ennchment at a constant burnup. Consequently, the SAR should also specify a muumum initial enrichment or establish other specific controls (e.g., maximum source terms) for the fuel.

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 determme 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 ShF 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 DRAFT 55 NUREG 1617 f

structure (e g., GAMGEN [Gosnell 1990J). In general, only gammas from approxunately 0.8 to 2.5 MeV will contribute significantly to the external radiation levels, so regrouping outside of this range is oflittle consequence Pay particular attention to whether the source terms are specified per assembly, per total number of assemblies, or per metnc ton, and ensure that the total source is correctly used in the shielding calculation.

C-Q the source tenns 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 unpurities (e g.,"Co) 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 imariarl 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 Ludwig and Renier (1989) and Luksic (1989).

i rhiing on the packaging dsign, neutron interactions could result in the production of energetic gammas near the packaging surface. If this source is not treated by the shielding anejysis code, verify that it is detemuned by other appropriate means l

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 supportmg information.

Because the gamma radiation levels are directly proportional to the gamma source term, the review should Letly confirm the source term calculated by the applicant.

512.2 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 transuranic and frorn (a,n) reactions in the fuel. rwaAng on the methods used to calculate these source terms, the applicant might determme the energy group structure dependently This is often accomplished by selecting the nuclide with the predommant contribution to spontaneous fission (e.g., Cm) 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.

De result of the source term calculation or expenmental data should be a listing of neutrons per second as a function of energy. For the spontaneous fission contribution, a listing of the signi6 cant nuclides should also be presented Because the neutron radiation levels are directly proportional to the neutron source term, the review should Ptly confirm the source term calculated by the applicant.

5.5.3 ModelSpecification Review the Structural Evaluation and Thensal Evaluation sections of the SAR to determme the effects of the tests for both normal conditions of transport and hypothetical accident conditions on the pe.ckaging and its contents. For example, the package might have impact limiters or an external neutron shield that would be NUREG-1617 5-6 DRAFT

i damaged or destroyed during the structural and thermal tests of 10 CFR 71.73. Verify that the correspondmg models used in the shielding calculation are consistent with these effects.

5.5.3.1 Configuration of Source and Shielding Enmma the ske*da or figwes that indicate how the shielding is modeled. Verify that the model dunensions and materials are consistent with those specifad in the package drawings presented in the General Information secuan of the SAR and the normal and accident conditions of the package. Dunensions should be at the unservative end of their tolerance range. Enswe that voids, &Q paths, and irregular geometnes are taken into account or otherwise modeled conservatively. Differences between normal conditions of transport and Hypothetical acculent conditions should be clearly indicated.

Verify that the sowce 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 homogemzed 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 radia'.on streaming can occur between the basket components.

Because of the bumup profile, a uniform source distribution is generally conservative for the top and bottom I

dose points, but not for the axial center unless the source strength is appropriately adjusted. If peakmg appears to be significant, verify that it has been treated appropriately. The assembly structural 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), dunensions of the transport vehicle should be included, as appropriate (e.g., to determme 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 determme the radiation level at i

normally occupied locations. (These dunensions or vehicle type, as well as positioning of the packages, may become limiting conditions in the certificate of compliance for exclusive-use shipments.)

l 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 maxi:num radiation levels; the maximum might not occur at the midpoint of a package surface or parallel plane. Radiation peakmg often occurs near the edges of extemal neutron shields and unpact limiters.

Deternune that voids, streammg paths, and irregular frassies are included in the model or othenvise treated in a conservative manner.

5.5.3.2 MaterialProperties 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/cm' or atoms / barn-cm, the review may consider either mess or atom L

. densities alone to be sufficient for certam 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 DRAFT 5-7 NUREG-1617

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 packagmg Confirm that specific controls are in place to ensure the long tenn effectiveness of the shielding.

Confirm that temperature-sensitive shielding materials will not be subject to 4. ares at or above their design limitations during either normal or accident conditions. Determine whether the applicant properly ev===~i the potential for shielding material to experience changes in material densities at te.wrature extremes (For example, elevated temperatures may reduce hydrogen content through loss of t.ound or free water in hydrogenous shieldmg materials.)

As noted above, a common practice in shielding analyses is to homogemze 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 homogemzed source is important for self-shielding, a confirmatory calculation of this densityis generally warranted.

1 5.5.4 Evaluation 5.5.4.1 Methods Verify that the computer program (s) used for the shielding analysis are appropriate. These codes may use 1

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 (ORNL-6268), DANTSYS (LA 9184-M and LA-10049-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 S AR should describe the solution method, benchmark results, vahdation procedures, and quality assurar.ce practices.

Assess if the number of dunensions of the code is appropriate for the cask configuration. Generally, a 2-D or 3-D calculation is necessary. One dunensional codes provide little information about off-axis locations and streanung 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 underestunate the radiation level (overestimate the radialleakage).

Verify that the cross section library used by the code is applicable for shielding calculations. Ensure that a coupled cross section se,t is used and that the code has been executed in a manner that accounts for secondary source terms, unless the evaluation has iPtly determmed a source term for neutron-induced gamma radiation or subcdtical multiplication of neutrons.

5.5.4.2 Key input and Output Data

. Verify that key input data for the shielding +d-% are identified. These will depend on the type of code l (point kernel, deterministic, Monte Carlo, etc.) as well as the code itself. In addition to the source terms, materials, and dunensions 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.

NUREG 1617 5-8 DRAFT

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 r.nd 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 slucidag analysis code will typically have the capability to perform this conversion directly with its own data library or with one supplied by the user. Generelly this conversion will use ANSI /ANS 6.1.1-1977.

Considerable confusion on these conversion factors has arisen as the result of a 1991 revision (ANSI /ANS 6.1.1-1991) to the above St-ind. This revised Standard is generally not appropriate for use in the analpis of transportation packages Title 49 CFR 173.403 indicates that radiation levels (for radioactive material shipments) are dose equivalent rates, not the effective dose equivalents as defined in ANSI /ANS 6.1.1 1991.

The conversion factors listed in 49 CFR 173.403 and those in 10 CFR 20.1004 are essentially those presented in the 1977 version. Use of the 1991 Standard will result generally in a somewhat lower estimate of the gamma radiation level compared with that from the 1977 Standard and can result in a neutron radiation level that is lower by more than a factor of 2. Finally, from a practical standpoint, the radiation levels determmed with the 1991 Standard do not correspond physically to radiation levels that are measured by dosimetry desices. Since radiation levels must be measured prior to shipment (see the Operating Procedures Review section of this SRP), radiation levels computed according to the 1991 Standard would be oflittle value for estimating the results of these measurements.

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 1.evels Confinn that the radiation levels under both normal conditions of transport and hypothetical accident i

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 l

those of maximum radiation levels and include any radiation streammg paths.

4 i

For the purposes of 10 CFR 71.47(b), NRC staficonsiders the external surface to be that part of the package l

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-i 5569A and NUREG/CR-5569B) as defined in 49 CFR 173.403.

Determme that the radiation levels appear reasonable and that their variation with location are consistent with the w;iy and shielding characteristics of the package.

s 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 t

(10 CFR 71.71) does not result in a significant increase in extemal radiation levels, as required by 10 CFR 71.43(f) ar.J 10 CFR 71.51(a)(1).

)

DRAFT 5-9 NUREG 1617

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 perfonned prior to shipment in order to show that the 10 CFR 71.47 limits are satisfied, a number of factors should be considered in determuung 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, boundmg assumptions in the analysis, margin from the regulatory limits, and the contribution from difficult to measure neutrons.

At a muumum, the review should include examination of the applicant's input to the computer program used for the shielding analysis. Verify use of proper dunensions, material properties, and an appropriate cross section set. In addition, independently evaluate the use of gamma and neutron source terms.

~ If a more detailed re iew is required, k-% =*=tly 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 l

section set from that of the S AR 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 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 rniewed 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 ModelSpecification

'The staff has rniewco the models used in the shielding evaluation and concludes that they are described in sufDcient detail to permit an i%t rniew, with confirmatory calculauons, 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.

NUREG 1617 5-10 DRAFT

The stafThas reviewed the package design, construction, and preparations fcr 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 maxunum external radiation level at one meter from the external surface of the package will not exceed 10 mSv/br (I ran/br) during hypa*W acculent conditions consistent with the tests specified in

- 10 CFR 71.73.

5 7 REFERENCES ANSUANS 6.1.1-1977 Amencan Nuclear Society, ANSUANS 6.1.1,"American National Standard for Neutron and Gamma-Ray Flux to Dose Factors,"la Grange Park, IL,1977.

ANSUANS 6.1.1-1991 American Nuclear Society, ANSUANS 6.1.1,"American National Standard for Neutron and Gamma-Ray Fluence to Dose Factors," La Grange Park, IL,1991.

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),"FederalRegister, FR 50247, U.S.

Govemment 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 Heasy Nuclides,"in Editor, Nuclearinctruments andMethods in Physics Research, A299 (l990), Elsevier Science, Ehnont, NY, pp. 682-686.

LA I 0049 M Los Alamos National Laboratory," User's Guide for TWODANT: A Code Package for Two Dunensional, Diffusion Accelerated Neutral Particle Transport,"LA-10049-M Rev.,los Alamos,NM, April 1992.

I LA 9184-M los Alamos National Laboratory," Revised User's Manual for ONEDANT: A Code for One Dunensional, DifTusion 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,14s Alamos, NM, LwM 1993.

Ludwig and Remer Ludwig, S.B., and Remer, J.P.," Standard-and Extended-Bornep PWR and 1989 BWR Reactor Models for the ORIGEN2 Computer C' E N1/TM-11018, Oak Ridge National Laboratory, Oak Ridge, TN, De her i 19.

Luksic 1989 Luksic, A.," Spent Fuel Assembly Hardware Characm.,

snd 10 CFR 61 Classification for Waste Disposal," PNL-6906, Volu.

3.cific Northwest Laboratory, Richland, WA, June 1989.

l DRAFT 5-11 NUREG 1617 l

NRC IN 80-32 U.S. Nuclear Regulatory Commission," Clarification of Certain Requirements for Exclusive-Use Shipments of Radion:tive Materials," IE Information Notice 80-32, U.S. Government Printing Office, Washington, D.C., August 29,1980.

NUREG/CR-0200A U.S. Nuclear Regulatory ramminaion," SCALE: A Modular Code System for Performmg Standardized Computer Analyses for Licensmg Evaluatm,"

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 Rag dd~y rammission," SCALE: A Modular Code System for Perfornung Standarchzed CT-Analyses for L6 6g Evaluation,"

NUREG/CR-0200, Vol. 2, Part 2, Rev. 4, U.S. Government Prmtag Office, Washington, D.C., April 1995.

NUREG/CR-5569A U.S. Nuclear Regulatory Commission," Clarification of Certam Reqw-cuts for Exclusive-Use Shipments of Radioactive Materials," HPPOS-084, in Health Physics Positions Data Base, NUREG/CR-5569, Rev.1, U.S.

Government Printing Office, Washington, D.C., February 1991.

NUREG/CR-5569B U.S. Nuclear Regulatory Commission," Clarification of Certam Requirements for Exclusive Use Shipments," HPPOS-085, in Health Physics positions Data l

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-Dunensional Discrete Ordinates Neutron / Photon Transport Code," ORNL-6268, Oak Ridge. TN, November 1987.

TRW-CSCIID TRW Environmental Safety Systems, Inc.," DOE Characteristics Data Base, A00020002 AAX01.0 User Manual for the CDB_R," CSCIID A00020002 AAX01.0, Vienna, VA, November 16,1992.

1 NUREG-1617 5-12 DRAFT

6 CRITICALITY REVIEW 6.1 REVIEW OBJECTIVE

'Ibe 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 ofreviewinclude the following:

6.2.1 Descripth,a of Criticality Design 6.2.1.1 Packaging Design Features 6.2.1.2 Codes and Standards 6.2.1.3 Suramary Table of Criticality Evaluations 6.2.1.4 TransportIndex 6.2.2 Spent Nuclear Fuel Contents 6.2.3 General Considerations for Evaluations 6.2.3.1 ModelConfiguration 6.2.3.2 MaterialProperties 6.2.3.3 Cornputer Codes and Cross Section Libraries 6.2.3.4 Demonstration of Maximtun 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 l

6.2.5.2 Results l

l 6.2.6 Evaluation of Package Arrays under Hypothetical Accident Conditions 6.2.6.1 Configuration i

6.2.6.2 Results 6.2.7 Benchmark Evaluations 6.2.7.1 Experunents and Applicability l

6.2.7.2 Bias Determmation l

6.2.8 Appendix 6.3 REGULATORY REQUIREMENTS Regulatory requirements of 10 CFR Part 71 applicable to the criticality review are as follows:

DRAFT 6-1 NUREG 1617

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 dunensions of materials of construction and materials specifically used as nanfissile neutron absorbers or moderators. [10 CFR 71.31(aXI) and 10 CFR 7133(aX5)]

The SAR must identify established codes and standards applicable to the criticality design.

[10 CFR 7131(c)]

'Ihe SAR must specify the allowable ownber of packages that may be transported in a single shipment.

[10 CFR 7135(b)]

A fissile material package must be assigned a i. spcst index for nuclear criticality control.

1

[10 CFR 71.59(b)]

l 63.2 Spent Nuclear Fuel Contents The contents must be described in sufficient detail to provde an adequate basis for their evaluation. This description must include the type, maumum quantity, and chemical and physical form of the spent nuclear fuel (SNF). [10 CFR 71.31(a)(1),10 CFR 7133(b)(1),10 CFR 7133(b)(2), and 10 CFR 7133(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]

63.3 General Considerations for 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(aX2),10 CFR 7135(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(aX1), 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 7135, 10 CFR 71.43(f),10 CFR 71.51(aXI), 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 contamment system. [10 CFR 71.55(b)]

A single package must be subcritical under the tests for hypothetical accident conditions. [10 CFR 71.55(e))

63.5 Evaluation of Package Arrays under Normal Conditions of Transport The SAR must evaluate arrays of packages under normal conditions of transport to deterame the maximum number of packages that may be transported in a single shipment. [10 CFR 7135 and 10 CFR 71.59]

l l

NUREG-1617 6-2 DRAFT 1

l 63.6 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 7135 and 10 CFR 71.59]

63.7 Benchmark Evaluations l

The package must be evaluated to demonstrate that it satisG:s the criticality safety reqmrements of 10 CFR Part 71. [10 CFR 71.3)(a)(2) and 10 CFR 7135]

l 6.4 ACCEPTANCECRITERIA l

6.4.1 Description of Criticality Design The regulatory requirements in Section 63.1 identify the acceptance criteria.

6.4.2 Spent Nuclear Fuel Contents The regulatory requirements in Section 63.2 identify the acceptance criteria.

6.4.3 General Considerations for Evaluations in addition to the regulatory requirements identified in Section 633, the packaging model for the criticality eva'uation should generally consider no more than 75% of the specified muumum 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 (k,,), two standard deviations, and the bias adjustment should not exceed 0.95 to demonstrate subcriticality by calculation. A bias that reduces the calculated value of k,,

should not be applied.

6.4.4 Single Package Evaluation In addition to the regulatory requirements identified in Section 63.4, the assumption of water inleakage for the analysis pursuant to 10 CFR 71.55 (b) should consider the packaging and contents to be io their most reactive condition, as detennined by the tests in 10 CFR 71.71 and 10 CFR 71.73.

j l

6.4.5 Evaluation of Package Arrays under Normal Conditions of Transport The regulatory reqmrements in Section 63.5 identify the acceptance critena.

6.4.6 Evaluation of Package Arrays under Hypothetical Accident Conditions The regulatory requirements in Section 63.6 identify the acceptance criteria.

DRAFT 6-3 NUREG.1617 l

l 1

6.4.7 Benchmark Evaluations The criticality evaluation of the package should include a comparison of the calculational methods with applicable h hmark experunents to determine the appropriate bias and uncertamties.

6.5 REVIEWPROCEDURES

'Ihe followmg procedures are guerally applicable to the criticality review of SNF transportation p*,,

Since packages for shipant of SNF are generally int-tad o be shipped by exclusive-use, only exclusive-t use stupmaate are assumed in the following SRP review procedures The criticality renew is based in part on the descriptions and evaluatmos presented in the General Information, Structural Evaluauon, and Thermal Evaluation sections of the SAR. Similarly, results of the criticality review are considered in the renew of the SAR sectmas on Operatmg Procedures and Acceptance Tests and Maintenance Program. Examples of SAR information flow into, within, and from the criticality

view 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. Packagmg design features unportant for criticality include, but are notlimited to:

Dunensions and tolerances of the contamment system Dunensions, 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 Dunensions, concentrations, tolerances, and location of neutron-absorbing and moderstmg materials, including neutron poisons and shielding Dunensions and tolerances of any floodable voids, including flux traps,inside the packagmg Dunensions 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 -:=g=== 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 dunensions, tolerances, and material composition of compcocets unportant to criticality safety.

NUREG-1617 6-4 DRAFT rrh

t senuenrel General h erusal Evalmenon intenmaelen Evahonen

- Deforsambon Fusile sensorial

- Coadmsson Displacessat of centents

- Decamposinon coasses and poisons h and De W Floodies mierences

- Melans minnsnals Neuaron poison osmesses 1 P 1 f F

cruncamey neview n.m Evam.n a.-i.

Fissile consent

- Mdts

- Biases

• h,%ons

- Cosnpastcodes Uncertainty derances w* s -s

is,::::;;eanalys

ft,ade.

~

n f

)

Cit ld' accident con 6nons) 1I 1 I D

b ance M

- Transportindex

- Dunensions and Neuron poisons or tolerances miodermor

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

l 6.5.1.2 Codes and Standards l

Verify that the established codes and standards used in all aspects of the criticality design and evalusten of the package are identified.

j 6.5.1.3 Summary Table of Criticality Evalustens Review the summary table of the criticality evaluatum, which should address the followmg 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 SN undamaged packages, under the conditions of 10 CFR 71.59(a)(1) l An array of 2N damaged packages, under the conditions of 10 CFR 71.59(a)(2).

DRAFT 6-5 NUREG 1617

For each case, the table should at least include the maximum value of k,,, the uncertamty, the bias, and the number of packages evaluated in the arrays. The table should also show that the sum of the effective multiplication factor (k ), two standard deviations, and the bias adjustment does not exceed 0.95 for each case.

i Confirm that the summary table illustrates that the package meets the above subcriticality critenon 6.5.1.4 TransportIndex Based on the number of packages evaluated in the arrays, verify that the SAR d%i the appropriate value of N and cale=1== 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 assemblics or rods (e.g., BWR/PWR) and vendor /model as appropriate Dunensions 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 Active fuellength Enrichment (variation by rod if applicable) before irradiation (see below)

Chenucal and physical form Mass ofinitial heavy metal per assembly or rod Number of fuel assemblies or individual rods per package.

a Because the NRC staff does not currently allow any credit for burnup of the fissile material or increase in actuude or 6ssion product poison during irradiation, the enrichment should be that of the unirradiated fuel. If assanblies contain fuel with several enrichments, the evaluation should either assume the maxunum enrichment or demonstrate that another approach (e.g., average enrichment) is boundmg. Section 6.5.3.2 below discusses consideration of the depletion of burnable poisons.

NUREG 1617 6-6 DRAFT

Deternune 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 l

I 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 pennit a detailed criticality evaluation of each type or to support a conclusion that certam l

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 MW by the criticality analysis. The review procedures below should address the evaluation for each contents as appropriate.

6.5.3 General Considerations for 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.

6.5.3.1 ModelConfiguration Exanune the Structural Evaluation and Thermal Evaluation sections of the SAR to determme 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 nommal design configurations. For example, the fuel assetablies 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 Dooded package, the SAR should address preferential flooding as appropriate. This includes floodmg of the fuel. cladding gap and other regions (e.g., flux traps) for which water density might not be uniform in a flooded package.

Deternune 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 homogeniza: ion is used, the SAR must demonstrate that it is applied correctly or conservatively. As a muumum, this demonstration should include calculation of the multiplication factor of one assembly and several benchmark expenments (see Section 6.5.7) using both homogeneous and heterogeneous models.

I 1

DRAFT 6-7 NUREG-1617

6.5.3.2 MatmalProperties Verify that the appropnate mass densities and atom densities are provided for all materials used in the models of the packagmg and n=amats 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 m., des should be referenced.

No more than 75% of the specified mmunum neutron poison raareatrauon 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 possons in the fuel. Ensure that neutron absorbers and moderators (e.g., poisons and neutron f *=g) are properly controlled during I

fabncauon to meet their speciSed propertaes. Such infonnation should be discussed in more detail in the AT-Tests and Maintenance Program secuan of the SAR.

i 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 packagmg 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 Both Monte Carlo and determmistic computer codes may bc used for criticality calculations. Monte Carlo codes are generally better suited to analyzing three-dimensional s-.wy and, therefore, are more widely used to evaluate SNF cask designs. Verify that the SAR uses an appropriate computer code for the criticality evaluation. Ct--- =!y 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 seqimn, while MCNP permits the use of continuous cross secuans 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, confum 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 q

cross section sets include data for fissile and fertile nuclides (based on a potential scattermg cross section, o,)

that can be input by the user. If the applicant has used a stand-alone version of KENO, ensure that potential scattenng has been properly considered Additional informauon addressing cross sectson concerns is provided in an NRC infonnation notice (NRC IN 91 26) and NUREG/CR-6328.

In addition to cross secuan information, other key input data for the criticality calculauons should be id = M ai These include number of neutrons per generauon, number of generauons, convergence criteria, mesh h% etc., dependag on the code used. The SAR should also include at least one representative input file for a single package, #- yi array, and damaged array evaluation. Verify that information regardag the model configurabon, matenal properues, and cross sections is property input into the code.

- At least one representative output file (or key sections) should generally also be included in the SAR. Ensure l

that the calculation has properly converged and that the calculated multiplicauon factors from the output files agree with those reported in the evaluation.

NUREG 1617 68 DRAFT 4

)

l

)

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 deterames the optimum combination internal moderation (within the package) and

%,wsed moderation (ber,~. packages), as applicable. Confirm that preferential flooding of different regmas within the package is considered as appropriate. As noted in Section 6.5.2, the maximum allowable amount of fissile matenal 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 andjustified.

Additional gnWa~ on determuung the most reactive configurations is presented in NUREG/CR 5661.

6.5.3.5 Confirmatory Analyses The review should include a confirmatory analysis of the criticality calculations reported in the SAR. As a nummum, the reviewer should perform an indaaaadant 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 k,, s 0.95. As with any design and revww, e small margin below the acceptance criterion and/or small degree of consen atism necessitate a more extensive analysis.

The reviewer should generally model the package indaaaadaatly and should use a different code and cross section r. f,m that used in the SAR. If the reported k,, for the worst case is substantially lower than the acceptance criterion of 0.95, a simple model known to produce very consen ative results may be all that is necessary for the indaaaadaat 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 k,,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 suberitical 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 I

and physical form of the contents i

l I

DRAFT 6-9 NUREG 1617 I

Water moderation to the most reactive credible extent, including water inleakage into the contamment

=

system as specified in 10 CFR 71.55(b)

Full water reflection on all sides of the package, includmg close reflection of the contamment system or reflecten by the package materials, wluchever 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.

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 suberiticality, 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 suberitical under normal conditions of transport. The evaluation should consider:

The most reactive configuration of the array, e.g., pitch, package orientation, etc., with nothmg (including moderator) between the packages The most reactive credible configuratic, & pckaging and its contents. (Because water does not leak into a spent fuel package under normaJ -

dem e oftransport, water inleakage need not be assumed.)

Full water reflection on all sides of the ara, G&s 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.

J Confirm that the appropriate N value is used in determmation of the transport index. The appropriate N should be the smaller value wluch assures suberiticality for SN 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 suberitical under hypothetical accident conditions. The evaluation should consider:

The most reactive configuration of the array, e.g., pitch, package orientation, etc.

NUREG-1617 6 10 DRAFT

l Optunum interspersed by4c as moderation (between packages)

The most reactive credible configuration of the packaginh W its contents, including inleakage of water and internal moderation Full water reDecuan on all sides of the array (unless the array is infinite).

i 6.5.6.2 Results l

l Vaify that the most reactive array conditions are clearly identifed and that the results of the analyms are consistent with the information presented in the s=nmary table discussed in Secuan 6.5.1.3 above.

I Confirm that the appropriate N value is used in determimog the ;.

g index. The appropriate N should be l

the smaller value wiuch assures subcriticality for 2N packages under hypa*W 1 accident conditions or SN packages under normal conditions of transport, as discussed in the previous section.

6.5.7 Benchmark Evaluations Ensure that the en==*r codes for criticality calculations are Wh==rked against critical experunents.

Verify that the analysis of the benchmark experunents used the same computer code, hardware, and cross section library as those used to calculate the multiplication factor for the package evaluations. The calculated l

k,,of the cask should then be adjusted to include the appropriate biases and uncertainties from the Wh==rk i

calculations.

l l

1 Additional information on benchmarkmg criticality evaluations for SNF is provided in NUREG/CR-6361.

6.5.7.1 Expenments and Applicability Review the general description of the benchmark expenments and confum that they are appropriately referenced The applicant shouldjustify and the reviewer should verify that the benchmark expenments are applicable to the actual package design. The Wh==rk expenments 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 expenments include type of fissile material, ennchment, H/U ratio (dependent largely on rod pitch and diameter), poisoning, reflector material, and configurauon 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 expenments and the uncertamties in expenmental data (e.g., mass, density, dunensions, etc.). Ensure that these uncertamties are treated in a conservative manner, i.e., they result in a lower calculated multiplication factor for the benchmark t

9.=:

6.5.7.2 Bias Determmarion Examme the results of the calculations for the benchmark expenments and the me'. hod used to account for biases, including the contribution from uncertainties in expenmental data.

DRAFT 6-11 NUREG 1617 l

Assess that a sufficient number of appropriate b L.=k experunents are analyzed and that the results of these benchmark calculations are used to determme an appropriate bias for the package calculations. The applicant should check benchmark comparisons for trends in the bias with respect to parameter 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 wiuch decrease k,, when applied) treated as zero bias.

Statistical and convergence uncertainties of both W==t and p-+=y calculations should also be addressed. 'Ihe uncertamties should be applied to at least the 95-percent confidence level. As a generW 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 uncertamties should not significantly affect the tesults.

6.5.8 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 supplementalinformation.

6.6 EVALUATION FINDINGS The criticality review should result in the following findmgs, 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.

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 evahmiion.

6.6.3 General Considerations for Evaluations The staff has reviewed the criticality description and evaluation of the package and concludes that it addresses the criticality safety requuwments of 10 CFR Part 7.1.

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.

NUREG 1617 6-12 DRAFT

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 De staff has reviewed the criticality evaluation of the most reactive array of 2N packages and -W that it is suberitical under hypothetical accident conditions.

6.6.7 Beaciunark Evaluations The staff has reviewed the umark evaluaten of the calculations and concludes that they are sufficient to i

deternune an appropriate bias and uncertamties for the criticality evaluation of the package.

6.7 REFERENCES

LA 12625-M 1.os Alamos National Laboratory,"MCNP 4A, Monte Carlo N-Particle Transport Code System," LA 12625-M, Los Alamos, NM, December 1993.

NRC IN 91-26 U.S. Nuclear Regulatsry Commission, " Potential Nonconsen ative Errors in the Working Format Hansen-Roach 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/CR-0200 U.S. Nuclear Regulatory Commission," SCALE: A Modular Code System for Performmg Standardized Computer Analyses for Licensing Evaluation,"

NUREG/CR-0200, Vol. 2, Part 2, Rev. 4, U.S. Govemment Printing Office, Washington, D.C., April 1995.

NUREG/CR-5661 U.S. Nuclear Regulatory Commission,"R*+....-mdations for Preparing the Criticality Safety Evaluation of Transportation Packages," NUREG/CR 5661 (ORN1/fM 11936), U.S. Govemment 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/CR-6328, 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/CR-6361 (ORN!/fM 13211), U.S. Government Printing Office, Washington, D.C., March 1997.

DRAFT 6-13 NUREG 1617

1 l

7 OPERATING PROCEDURES REVIEW l

7.1 REVIEWOBJECTIVE The objective of this review is to verify that the operating procedures comply with the reqmrements of 10 CFR 71 and ensure that the package will be operated in a manner consistent with the conditions assumed in its evaluauon for approval.

7.2 AREAS OF REVIEW The SAR should be reviewed for adequacy of the operating procedures description. Areas of review include the followmg 7.2.1 Package Loading 7.2.1.1 Preparation for Loadmg 7.2.1.2 loadmg 7.2.1.3 Preparation for Transport J

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 Contents Removal 7.2.3 Preparation of Empty Package for Transport 7.2.4 Other Procedures i

7.2.5 Appendix 7.3 REGULATORY REQUIREMENTS Regulatory..qmrements of 10 CFR Part 71 applicable to package operations and the operating procedures review are as follows:

7.3.1 Package Loading

'Ibe SAR must identify established codes and standards applicable to the criticality design.

[10 CFR 71.31(c)]

The SAR for a fissile material shipment must include any proposed special controls and precautions for transport, loadmg, 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 38'C (100*F) and in the shade, no accessible

' surface of a package would have a temperature exceedmg 85'C (185'F) 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.)

DRAFT 7-1 NUREG-1617 1

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.

i

[10 CFR 71.87(a))

Before each shipment, the package must be verified to be in ummpaired 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. I10 CFR 71.87(c)]

Before each shipment, any system for contammg liquid must be verified to be e-ly scaled and to have adequate space or other specified provision for expansion of the liquid. [10 CFR 71.87(d)]

Before each shipment oflicensed material any pressure relief device must be verified to be operable and properly set. [10 CFR 71.87(e)]

i Before each shipment, a verification must be made to ensure that the package has been loaded and closed appropriately. [10 CFR 71.87(f)]

Before each shipment of fissile material, a verification must be made to ensure 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 reqmrements of 10 CFR 71.45. [10 CFR 71.87(h)]

Before cach shipment, the level of non-fixed (removable) radiotetive contammation on the extemal 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 h pcitatioc. [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 carner for tr-.pcit, the hcensee must send or make available any special astructions needed to safely open the package to the consignee for the cmsignee's use in accordance with 10 CFR 20.1906 (e). [10 CFR 71.89]

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]

NUREG-1617 72 DRAFT

7J3 Preparation of Empty Package for Transport Before each shipment, perform the necessary inspections and tests to ensure that the level of non-fixed (ranovable) radioactive contammatum 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 OtherProcedures ne applicahon for a fissile material shipment must include any proposed special controls and precaubons for transport, loadas, unloadmg, and handhng and any proposed special controls in case of accident or delay.

[10 CFR 7135(c)]

7.4 ACCEPTANCECRITERIA 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 reqmrements identified in Section 7.3.1, leakage testing of the package should meet the assembly verification leakage test requirements specified in ANSI N14.5.

7.4Property "ANSI code" (as page type) with input value "ANSI N14.5.</br></br>7.4" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..2 rackage Unloading The regulatcry requirements in Section 7.3.2 identify the acceptance criteria.

7.43 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 decontammated and closed in accordance with the requirements of 49 CFR 173.428.

7.4.4 Other Procedures In addition to the regulatory reqmrements identified in Section 7.3.4, the package should be properly closed and delivered to the carner in such a condition that subsequent transport will not reduce the effectiveness of the packaging.

7.5 REVIEW PROCEDURES The followirs procedures are generally applicable to the operatmg 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 followmg SRP review procedures The operating procedures review is based in part on the descriptions and evaluations presented in the General Informauon, 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.

DRAFT 7-3 NUREG 1617

General Stuctural inferenselen Evaluetten

- Comme.

Openmoesi feeswes Liding com&gwenion~

General restrictions 8

In-lll-*

1:"cacon

- Cesssets

- Headiesreevictione 1herneal t' h ant eM=Imag Ortecauty Evaluation Eveksessa Evelmetian Evelmacion

- Temperennes Ooess requirenesis n=A=w levels

- Transportindex

- Prenewes

- Almemblyveri6cahoe

- Seensuespeabs

- Neumonpoisonsor lenkese rose

- Pre shgunset surways niederosor V

if 1F IP If 1 P Operating Procedures Iteview i

Pockese Pack realen of Empty 1

Land 6ag U

for Transport Preparamon for loading

- Receiptofpackagefrom

- Properamonfor toeding camer eensport Prepren,on for transport

- Properation for unloading

- Coseenis removal Figure 7-1 SAR Information Flew for the Operating Procedures Review.

The operating procedures presented in the SAR should not be expected to be detaded procedures that could be implemented widxnst 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.

Information on both the detailed procedures and the brief procedures included with an application can be found in NUREG/CR-4775.

73.1 PackageLoading 7.5.1.1.'.v ion for Loadmg

' Review the procedures presented sequentially in the order of performance for loadmg and preparmg the package for tr sp. At a minimum, the procedures should ensure that:

The package is proper for the cantanes to be shipped, including the second containment for damaged fuel The package is in ummpaired physical conditior.

Any proposed special controls and precautions for handhng the package are provided.

NUREG-1617 74 DRAFT

7.5.1.2 Iamdag Review the procedures presented sequentially in the order of performance for loading the package contents for transport At a mmimum, the procedures should ensure that:

Special handimg equipment needed for lon&ng and unloading is prosided Any proposed special controls and precautions for loa &ng and handimg the package are provided Any maderator 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 l

Review the procedures presented sequentially in the order of performance for preparing the package for I

transport. At a muumum, the procedures should ensure that:

1 The level of non-fixed (removable) radioactive contanunation on the external surfaces of each package offered for shipment is ALARA, and within the limits specified in DOT regulation 49 CFR 173.443 i

Radiation survey reqmrements 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 m&catmg device is irmyssted which, while intact, indicates that the package has not been Opened by unauthorized persons j'

Any system for contaming liquid is adequately sealed and has adequate space or other specified provision l

for expansion of theliquid i

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 moperable for that purpose unless it satisfies the design requirements of 10 CFR 71.45 DRAFT 7-5 NUREG 1617 1

J l

Any proposed special controls and precautions for transport, handling and any proposed special controls in case of accident or delay are specified Written instmetions to the carrier are provided for packages which require exclusive use shipment hacause 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 carriciF for 6-spect, the hcensee 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 unloadmg procedures are the reverse of the loadmg procedures if applicable, procedwes for special controls and precautions to enswe safe removal of fission gases, conemmmanad coolant, and solid ennemmmanen should be,,w;=d and discussed.

7.5.2.1 Receipt of Package from Carrier Review the procedures presented sequentially in the order of performance for receiving the package from the canier 'At a nummum, the procedures should ensure that:

The requirements of 10 CFR 20.1906 are met The package is av==i_nad for visible external damage Steps to define actions to be taken when the tamper indicating device is not intact, or surface a

contammation or radiation survey levels are too high are provided A list of any special handling equipment needed for unloading and handhng 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 ord:r of performance for preparmg the package for unloading followmg receipt. At a mmimum, the procedures should ensure that:

Procedwes controlling the radiation level limits on unloadmg operations are provided Procedures for the safe removal of, if any, fission gases, contammated coolants, and solid contammants are provided.

7.5.2.3 Contents Removal Review the procedures presented sequentially in the order of performance for removmg the contents following package receipt. At a nummum, the procedwes should ensure that:

The closure is removed appropriately NUREG 1617 7-6 DRAFT

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 mmimum, the procedures should ensure that:

The packagingis 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 Special preparations of the packaging, to ensure that the interior of tie 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 resiew should result in the following fmdings, as appropriate:

7.6.1 Packsge 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 req.rements 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 cnd concludes that the limits specified in 10 CFR 71.43(g) will be met.

The staff has resiewed the description of the routine detemunations for package use prior to transport, and concludes that the requirements of 10 CFR 71.87 will be met.

The staff has resiewed the description of the special instructions (if applicable) needed to safdy 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.

DRAFT 77 NUREG.1617

7.6.2 Package Unloading The staff has reviewed the proposed special controls and precautions for unloading and handhng and conc udes that they satisfy 10 CFR 71.35(c).

7.6.3 Preparation of Empty Package for Transport The staff has reviev od the description of the routine determmations for packsge use prior to transport, and coachides that the requirements of 10 CFR 71.87 will be met.

7.6.4 Other P nedures The staff has resiewed 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

ANSIN14.5 Institute for Nuclear Materials Management, ANSI NI4.5,"American National Standard for 1.cakage 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 (UCID-20820), U.S.

Govemment Printing OfEce, Washington, D.C., December 1988.

NUREG-1617 7-8 DRAFT

8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW 8.1 REVIEW OBJECTIVE The objectives of the 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 packagmg performance throughout its sen> ice life.

8.2 ACCEPTANCETESTS 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 VisualInspections and Measurements 8.2.1.2 WeldInspections 8.2.1.3 Structural and Pressure Tests 8.2.1.4 Leakage Tests 8.2.1.5 Component Tests 8.2.1.6 ShieldinE ests T

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. [f 0 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 operr. ting pressure (MNOP) of a package exceeds 35 kPa (5 lbf/in )

2 gauge, the containment system of each packaging must be tested at an intemal pressure at least 50 percent higher than MNOP to verify its capability to maintain structural integrity at that pressure. [10 CFR 71.85(b)]

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)]

DRAFT 81 NUREG-1617

The twensee must perform any tests damnad appropnate by NRC. [10 CFR 71.93(b)]

8.2.3 Acceptance Criteria in addition to the regulatory regerann=en ident%d in Section 8.2.2, the SAR should discuss the package tests to be perfonned and the acceptance criteria to denanetrate structural, leakage, shieldmg, and best transfer performance Fabncation, weldmg, and nammaren of components are acceptable when performed in accordance with the ra===== dad 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. Is.akage testag of the packagmg should be accomplished in accordance with ANSI N14.5. Fabncation, nammation, and acceptance testag oflifting tnanions should be canarnad in acconiance with ANSI N14.6 or other appropnate spanrwarm.

8.2.4 Review Procedures

- The followag procedures are generally applicable to the acceptance tests review of all spent nuclear fuel (SNF) packages De acceptance tests review is based in part on the descriptions and evaluations presented in the General Infonnation, Structural Evaluation, Thermal Evaluation, Contamment Evaluation, Shieldmg Evaluation, Criticality Evaluation, and Operating Procedures sections of the SAR and follows the sequence established to evaluate the packaging agamst applicable 10 CFR Part 71 reqmrements. Examples of SAR information flow into and within the acceptance tests review are shown in Figure 3 1.

The commitments specified in the Acceptance Tests and Maintenance Program section of the SAR are often ir.wporated 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 aar Information presented on each test should include, as a muumum, a description of the test, the test procedure, and the acceptance criteria. Confirm that the established codes, standards, and spreific 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 VisualInspections and Measurernents Ensure that visual inspections are performed to verify that the packaging has been fabncated and assembled in accordance with drawings and other requirements specified in the SAR. Dunensions and tolerances specified on the drawings should be confirmed by measwoment.

8.2.4.2 Weldlaspecv.ons Vwify that weld inspections are performed to vmfy fabncation in accordance with the drawings, codes, and standards specified in the SAR to control weld quality. !-% type, and size of the welds should be confinned by measurement. Other specificatmas for weld performance,i===ae*% and acceptance should be verified as appropriate.

Additional gmdance on welding criteria is provided in NUREG/CR 3019.

NUREG-1617 8-2 DRAFT

l

)

General aeructural insormeeen Evahsotion

. Codesandseendants

. Codesandseendards h and Structwal and tolermanos P' essure test

.P smaterials Comesses 1herusel rem sidsiding Cuteemacy Evahomelam Evatueehen Evahassion Evalueelen

. Temiperanse

. Febncemanvar: Seamen

. Slueidi tems

. W and veri 6annan

, ' - 'susaarial elvences

. Pressises rase

.S

. Hastseasierinsanos

- Neuseopoisoes l

leakage reas 1 F i f U

If P

U Acceptance Tests Review

. Visualinspections and

. Structwaland

. Weldinspections

. Laekase test enasweastas avesswe mas shieldias iesi

. Thennaliens

. ret tests

. Heutron absorber test Figure 81 SAR Information Flow for the Acceptance Tests Review.

8.2.4.3 Structural and Pressure Tests Verify that the structural or pressure tests are identified e4 described. Such tests should comply with

)

10 CFR 71.85(b), as well as applicable codes or standards specified in the SAR. Structural testing oflifting trunnions should be conducted in r.ceidis.ce with ANSI N14.6.

8.2Property "ANSI code" (as page type) with input value "ANSI N14.6.</br></br>8.2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..4.4 Leakage Tests Verify that the contamment system of the packaging is subjected to the fabrication leakage tests specified in ANSI N14.5. The acceptable leakage critenon should be consistent with that identified in the Contmarnent Evaluation section of the SAR.

1 8.2.4.5 Cm~=-at Tests Confinn that tests and W-criteria for other con:ponents are specified as appropriate. Such j

components include valves, rupture disks, seals, etc.

8.2.4.6 Sluekhng Tests Ensure that appropriate shielding tests are specified for gamma and neutron radiation. The tests and acceptance criteria should be sufficient to assure no voids or streanung paths exist in the shielding.

DRAFT 83 NUREG.1617 f

1 8.2.4.7 Neutron Absorber Tests Verify that appropriate tests are specified to verify the amount and distribution meetag the minimum l

specification of neutron absorbing material described in the SAR.

8.2.4.8 ThermalTests Verify that appropriate tests are specified to denanetrate the heat transfer capability of the pd=ing These tests should enafirm the heat transfer propemes predicted in the Thermal Evaluation section of the SAR.

8.2.4.9 Appendix

I The appendix may include a list of references, copies of any applicable references not generally available to the reviewer, and any other appropriate supplanental information 8.2.5 Evaluation Findings The acceptance tests review should result in the following findmgs, 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 prelimmary determmations 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.

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 Imkage Tests 8.3.1.3 C7-t Tests 8.3.1.4 Neutron Absorber Tests 8.3.1.5 Thermal Tests '

8.3.1.6 Appendix 8.3.2 Regulatory Requirements Rag"'d7 requirements of 10 CFR Part 71 applicable to the maintananer program revww are as follows The SAR should identify established codes, standards, and specific provisions of the quahty assurance program that are applicable to the proper maintenance of the packaging. [10 CFR 71.31(c) and 10 CFR 71.37(b)]

NUREG-1617 8-4 DRAFT

The maintenance program should ensure that the packaging is maintamed 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 reqmrements identified in Section 8.3.2, the maintenance program should include periodic testing reqmrements, inspections, and replacement criteria and schedules for replacements and repairs of components on an as-needed basis.

83.4 Review Procedures The following procedures are generally applicable to the maintenance program review of all SNF packages.

The maintenance program resiew 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 maintenance program resiew are shown in Figure 8 2.

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 senice life. Verify that the following maintenance tests and inspections are described with schedules and criteria for minor refurbishment and replacement ofparts, 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 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 oflifting trunnions should be conducted in accordance with ANSI N14.6.

8.3Property "ANSI code" (as page type) with input value "ANSI N14.6.</br></br>8.3" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..4.2 le.akage 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.

DRAFT 85 NUREG 1617

1..

cener :

servensral Infonnenen I' valuation

- Codesandstandards

- Codesandstandards

- Dimensions and Structwn! and tolerances prew4ure test

' ' ~ D "sassenals Contents

'Iherussi Omntainsment Sidsiding CHeicnNey Evahanelen Evakseelen Evaknauen Evahistian Temperature

- Febncabanwrdketion

- Shielding wsts Dirnensions and

- Pressures raso

- Simel&nsmasterial t>lerances

' Hem armasser tsanwes

-P veri 6caban speci6 canons

- Neutron poisons leakage rate 1V 1 V 1V 1V 1 V 1V I

Maintenance Prograan Review i

Structural and pressure tests

- taakase test Cornponent tests Neutron absor6er test Theernal tens Sisel&ng test Figure 8-2 SAR Information Flow for the Maintenance Program Review.

83.43 Component Tests Confirm that periodic tests and replacement schedules for components are described as appropriate. Such components include valves, rupture di.;ks, and seals. Elastomeric seals should be replaced at an interval not to exceed one year. Metal seals should be replaced after each use.

83.4.4 Neutron Absorber Tests Verify that the SAR identifies any process that could result in deterioratirm of neutron absorbing material and that appropriate tests to ensure packaging effectiveness are specified.

83.4.5 ThermalTests Appropriate penodic tests should be performed to verify tle heat transfer capability of the packaging during its service life. Tests similar to the acceptance tests may be applicable. The t3pical interval for periodic thermal tests is five years.

83.4.6 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.

l l

NUREG-1617 8-6 DRAFT

8.3.5 Evaluation Findings

& mamt== program review should result in the following findmgs, as appropriate:

The staff has reviewed the identification of the codes, standards, and provisions of the quality assurance program applicable to maint=== of the packagmg and concludes that the requiranents specified in 10 CFR 71.31(c) and 10 CFR 71.37 (b) will be met.

The staff has reviewed the description of the routix determinations for package use prior to h siat and concludes that the reqmrements of 10 CFR 71.87(b) and 10 CFR 71.87(g) will be met.

s.4 REFERENCES ANSIN14.5 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 Contamers Weighing 10,000 Pounds (45000 kg) or More for Nuclear Materials," New York, NY,1993.

B&PV Code American Society of Mechanical Engmeers,"ASME Boiler and Pressure Vessel Code,"New York, NY,1995.

NUREG/CR 3019 U.S. Nuclear Regulatory Commission,"Rw=W Welding Criteria for Use in the Fabrication of Shipping Contamers for Radioactive Materials,"

NUREG/CR 3019, Lawrence Livermore National laboratory, Livermore, CA, March 1984.

1 NUREG/CR-3854 U.S. Nuclear Regulatory Commission," Fabrication Criteria for Shipping 4

Containers," NUREG/CR-3854, Lawrence Livermore National Laboratory, Livermore, CA, March 1985.

l l

DRAFT 8-7 NUREG 1617

APPENDIX A-STANDARD REVIEW PLAN CORRELATION WITH 10 CFR PART 71 AND REGULATORY GUIDE 7.9 t

The following table summarues the correlation of the SRP review procedure sections with the appropriate l

acetions of 10 CFR Part 71 and RG 7.9.

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

SRP Review 10 CFR Part 71 Section RG 7.9 Section Procedure Section 1.5.1 None Introduction 1.5.2 71.13, 71.31(aX I), 71.31(a)(2), 71.31(aX3), 71.31(b),

1.1 71.31(c),71.33(a)(1),71.33(a)(3), 71.35(b),71.37, 71.38,71.59,71,107(c) 1.5.3 71.31 (a)( 1 ), 71.33 (a)(2), 71.33 (aX4), 71.33 (a)(5 ),

1.2 71.33(aX6),71.33(b),71.43(b) 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 2.4 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 (aX 1 ), 71.55(d)(4),

2.6 71.71 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 (aX I ), 71.31 (c), 71.33 (a)(5 ), 71.33 (aX6),

3.1 71.33(bX 1), 71.33(bX3), 71.33(bX5), 71.33(b)(7),

71.33(bX8),71.51(c) 3.5.2 71.31(aX1),71.33(aX5) 3.2,3.3 l

3.5.3 71.31(aX2),71.35(a),71.41(a)

None j

3.5.4 71.43(g)

None l

DRAFT A-1 NUREG-1617

l SRP Review 10 CFR Part 71 Section RG 7.3 Section Procedure Section l

1 3.5.5 71.43(0,71.51(aXI),71.71 3.4 3.5.6 71.73 3.5 3.5.7 None 3.6 l

71.31 (aX 1 ), 71.31 (c), 71.33 (aX4), 71.33 (aX5),

4.1 4.5.1 71.33(bX 1), 71.33(bX3), 7 f.33(bX5), 71.33(bX7),

71.43(c),71.43(d),71.43(e) 4.5.2 71.31 (a)(2), 71.35(a), 71.41 (a), 71.43(0, 71.43 (b),

4.2 71.51(a)(1),71.51(c) 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(0, 71.47(b),

5.4 71.5!(a)(1),71.51(a)(2) 5.5.5 None 5.5 6.5.I 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 (bX I ), 71.33(b)(2 ), 71.33(b)(3),

6.2 71.83 6.5.3 71.31(aX2),71.35(a),71.41(a) 6.3 6.5.4 71.35, 71.43(0, 71.51 (a)(1), 71.55(b), 71.55(d),

6.4 71.55(e) 6.5.5 71.35,71.59 6.4 6.5.6 71.35,71.59 6.4 6.5.7 71.31(aX2),71.35 6.5 6.5.8 None 6.6 7.5.1 71.31 (c), 71.35(c), 71.43(g), 71.47(b), 71.47(c),

7.1 71.47(d),71.87,71.89 7.5.2 71.35(c) 7.2 NUREG 1617 A2 DRAFT

SRP Review 10 CFR Part 71 Section RG 7.9 Section Procedure Section 7.5.3 71.87(i) 7.3 7.5.4 7' 35(c)

None 7.5.5 None; 7.4 8.2.4 1 1.31 (c), 71.37(b), 71.85(a), 71.85(b), 71.85(c),

8.1 7.1.87(g),71.93(b) 8.3.4 71.31(c), 71.37(b), 71.87(b), 71.87(g), 71.93(b) 8.2 l

1 l

DRAFT A3 NUREG-1617

APPENDIX B -TABLE OF EXTERNAL DOSE RATES FOR EXCLUSIVE-USE SHIPMENTS The follomng table summanzes the information that should be provided by the applicant for the external dose rates for transpatauon packages for spent nuclear fuel.

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

Normal Conditions ofTransport Package Surface

• Radiation Top Side Bottom Gamma 6 f

Neutron

• Total 10 CFR 71.47(b)(I) Limit 2 (200) 2 (200) 2 (200) d Package Surface

• 4 Radiation Top Side Bottom Gamma 6 Neutron

• Total 10 CFR 71.47(b)(1)(i-iii) Limit 10 (1000) 10 (1000) 10(1000) d Vehicle Outer Surface' Radiation Top Side Bottom Gamma 6 Neutron' Total d

10 CFR 71.47(b)(2) Limit 2 (200) 2 (200) 2 (200) l DRAFT B-1 NUREG 1617

Table B.1 (Cont.) External Dose Rates for Packages (Exclusive-Use Shipment).

Normal Conditions of Transport 2 Meters from Vehicle Outer Surface 8 Radiation Top Side Bottom 6

Gamma Neutron

• Total 10 CFR 71.47(b)(3) Limit 0.1 (10) 0.1(10) 0.1 (10) d 6

Normally Occupied Positions in Vehicle Radiation Top Side Bottom 1

Gamma 6 Neutron' Total 10 CFR 71.47(b)(4) Limit O.02 (2) 0.02 (2) 0.02 (2) d Hypothetical Accident Conditions 1 Meter from Surface of Package Radiation Top Side Bottom 6

Gamma Neutron' Total 10 CFR 71.51(a)(2) Limit 10(1000) 10 (1000) 10 (1000) d

• External surface of package.

• Gamma dose rate based on mwd burnup, _ % "'U enrichment, and _ years coolmg time.

• Neutron dose rete baand on mwd bumup, _ % "'U ennchment, and _ years coohng time.

• Dose rase in mSv/h (mrom/h).

' Extemal surface of package provided that shipment is in a closed vehicle, package postion within velucle remams fixed during transport, and no loadmg or unloading operations occur en route.

' At any point on the outer surface of the vehicle,includmg the upper and lower surfaces; or in a non< dosed volucie, at any point on the vertical planes protected from the outer edges of the vehicle, on the upper surface of the load or enclosure (if applicable), and on the lower extemal surface of the volucle At any point 2 meters (80 inches) from the outer lateral surface of the volucle (excluding the top and underade of the vehicle); or in the case of a nona:losed volucie, at any point 2 meters (6.6 feet) from the vertical planes propected by the ater edges of the vehicle (excluding the top and underside of the vehicle).

• In ey normally occupied space, except that this provision does not apply to private camers if exposed personnel under their control wear radiation dosimetry devices in conformance voth 10 CFR 20.1502.

NUREG 1617 B-2 DRAFT

APPENDIX C-COMMENT SHEET COMMENT SHEET (NUREG-1617) t l

l l

Submitted by C--t Number of i

Subject Addressed in C=-t l

l Section Paragraph Page t

l l

Type of C=-t: (Please select only one of the following categories)

I Addition Deletion Clarification / Change EditorialOnly i

l Other (specify)

Comment:

1 1

Justification.

4 l

l l

l l

i Suggested Revision:

DRAFT C-1 NUREG-1617 I

F NRC FORM 33g u.s. NUCLEAR REOuLAToRY COtmassioN

1. REPORT NuMsER p.es)

NRCM H02, (Analgned by NRC, Add Vel, Supp., Roh 32oi. am BIBLIOGRAPHIC DATA SHEET

""d^*"*"""""*"'"'4 rseemamaeone on me mwee)

2. m1E AND SuBTrTLE NUREG-1St7 Standard Review Plan for Transportation Packages for Spent Nuclear Fuel 3

DATE REPORT PusuSHED T-MONTH vtaa l

Draft Report for Comment March 1998

4. FIN OR ORANT NuMsER

5. AUTHOR (S)

s. TYPE OF REPORT Technical

7. PERIOD COVERED (ancsuerve osseet l

l 8 PF ORMlNG ORGAN ON. NAME ANo ADDRESS (NNRc, prende Duom. Once or Repm, U S NuhwRepusWwy cornnesson, andsnednp espues, toonmusw Spent FuelProject Office Office of Nuclear Material Safety and Safeguards i

U.S. Nuclear Regulatory Commission Washington, D.C.20555-0001

s. SPONSORING ORGANIZATION NAME AND ADDRESS (rNRc. Wseme es enovet acareedor, promos NRc Dumon Once orRepon, u 5 Numma andmenne sneees.)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200moros orasse)

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

{

(3) document the practices developed by the staff in previous reviews of package applications. Comments regarding errors or '

omissions, and suggestions for improvement should be sent to the Chief, Rules Review and Directives Branch, Division of Freedom of Information and Publication Services, Mail Stop T-6-D-59, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555-0001.

12. KEY WORDs/DESCRIPTORS (Last mores orphreeen met me esser resserene,e m ancoeng me sesort;

13. AVAILAssuTV STATEMENT Transportation, Packaging, SRP, Spent Nuclear Fuel, Spent Fuel, Cask

""U* *

14. SECURrTY CLASSIFICATION (The Pope) unclassified (The Renwr>

unclassified I

15. NuMsER OF PAGES l

10. PRICE NRc I oRM 336 (240)

Printed on recycled paper Federal Recycling Program

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