NUREG-1609 

 Supplement 1

 NU

Standard Review Plan for

Transportation Packages for

MOX-Radioactive Material

U.S. Nuclear Regulatory Commission

Office of Nuclear Material Safety and Safeguards

Washington, DC 20555-0001
NUREG-1609 

Supplement 1 

Standard Review Plan for 

Transportation Packages for 

MOX-Radioactive Material 

Manuscript Completed: September 2005 

Date Published: September 2005 

Prepared by 

R.S. Hafner, G.C. Mok, J. Hovingh, 

C. K. Syn, E.W. Russell, S.C. Keaton, 

 J.L. Boles, D.K. Vogt, P. Prassinos 

Lawrence Livermore National Laboratory 

7000 East Avenue 

Livermore, CA 94550-9234 

J.A. Smith, NRC Project Manager 

Spent Fuel Project Office 

Office of Nuclear Material Safety and Safeguards 

U.S. Nuclear Regulatory Commission 

Washington, D.C. 20555-0001 

NRC Job Code A0291 
 

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ABSTRACT 

The NRC contracted with LLNL to compile this supplement to NUREG-1609 to incorporate additional 

information specific to mixed uranium-plutonium oxide (MOX) radioactive material (RAM). This 

supplement provides details on package review guidance resulting from significant differences between 

low-enriched uranium (LEU) oxide radioactive material (RAM) and that from MOX RAM. While 

comparisons of MOX RAM with LEU RAM include powder, fresh-fuel rods, and fresh-fuel assemblies, 

they do not include comparisons with high-enriched uranium (HEU) or plutonium as metal, oxide, or 

nitrate in like forms that are covered by NUREG-1609. The principal purpose of this supplement is to 

ensure the quality and uniformity of staff reviews of packagings intended for transport of MOX RAM. 

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 the NRC staff review process. In particular, this supplemental 

guidance, together with NUREG-1609, assists potential applicants by indicating one or more acceptable 

means of demonstrating compliance with the applicable regulations. 
 

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CONTENTS 

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

ACRONYMS AND ABBREVIATIONS ...................................................................................................................................ix 

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

1 GENERAL INFORMATION REVIEW ..................................................................................................................................5 

1.5 Review Procedures .........................................................................................................................................................5 

1.5.1 Introduction ..........................................................................................................................................................5 

1.5.2 Package Description .............................................................................................................................................5 

2 STRUCTURAL REVIEW........................................................................................................................................................7 

2.5 Review Procedures .........................................................................................................................................................7 

2.5.1 Description of Structural Design ..........................................................................................................................7 

2.5.2 Materials...............................................................................................................................................................7 

2.5.3 Fabrication and Examination................................................................................................................................8 

2.5.4 Lifting and Tie-Down Standards for All Packages...............................................................................................8 

2.5.6 Structural Evaluation under Normal Conditions of Transport..............................................................................8 

2.5.7 Structural Evaluation under Hypothetical Accident Conditions...........................................................................8 

3 THERMAL REVIEW...............................................................................................................................................................9 

3.5 Review Procedures .........................................................................................................................................................9 

3.5.1 Description of Thermal Design ............................................................................................................................9 

3.5.3 General Considerations for Thermal Evaluations.................................................................................................9 

3.5.4 Thermal Evaluation under Normal Conditions of Transport................................................................................9 

3.5.5 Thermal Evaluation under Hypothetical Accident Conditions...........................................................................10 

3.5.6 Appendix ............................................................................................................................................................11 

4 CONTAINMENT REVIEW...................................................................................................................................................13 

4.5 Review Procedures .......................................................................................................................................................13 

4.5.1 Description of the Containment System .............................................................................................................13 

4.5.2 General Considerations.......................................................................................................................................13 

4.5.3 Containment under Normal Conditions of Transport (Type B Packages)..........................................................14 

4.5.4 Containment under Hypothetical Accident Conditions (Type B Packages).......................................................14 

5 SHIELDING REVIEW...........................................................................................................................................................17 

5.5 Review Procedures .......................................................................................................................................................17 

5.5.2 Radiation Source ................................................................................................................................................17 

5.5.4 Shielding Evaluation ..........................................................................................................................................17 

6 CRITICALITY REVIEW.......................................................................................................................................................19 

6.5 Review Procedures .......................................................................................................................................................19 

6.5.7 Benchmark Evaluations......................................................................................................................................19 

7 OPERATING PROCEDURES REVIEW ..............................................................................................................................21 

7.5 Review Procedures .......................................................................................................................................................21 

7.5.1 Package Loading ................................................................................................................................................21 

7.5.2 Package Unloading.............................................................................................................................................21 

8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW.............................................................................23 

8.5 Review Procedures .......................................................................................................................................................23 
 

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8.5.1 Acceptance Tests................................................................................................................................................23 

8.5.2 Maintenance Program.........................................................................................................................................24 

APPENDICES ...........................................................................................................................................................................27 

APPENDIX A9: MOX POWDER AND PELLET PACKAGES..........................................................................................A9-1 

A9.1 Package Type.........................................................................................................................................................A9-1 

A9.1.1 Purpose of Package......................................................................................................................................A9-1 

A9.2 Package Safety................................................................................................................................................A9-1 

APPENDIX A10: UNIRRADIATED MOX FUEL PACKAGES ......................................................................................A10-1 

A10.1 PACKAGE TYPE .............................................................................................................................................A10-1 

A10.1.1 Purpose of Package..................................................................................................................................A10-1 

A10.1.2 Description of a Typical Package ............................................................................................................A10-1 

A10.1.3 Alternative Package Design.....................................................................................................................A10-1 

A10.2 PACKAGE SAFETY ........................................................................................................................................A10-1 

A10.2.1 Safety Functions ......................................................................................................................................A10-1 

A10.2.2 Safety Features ........................................................................................................................................A10-2 

A10.2.3 Typical Areas of Review for Package Drawings.....................................................................................A10-2 

A10.2.4 Typical Areas of Safety Review ..............................................................................................................A10-3 

APPENDIX C: DIFFERENCES BETWEEN THERMAL AND RADIATION PROPERTIES OF MOX AND LEU 

RADIOACTIVE MATERIALS ..............................................................................................................................................C-1 

APPENDIX D: BENCHMARK CONSIDERATIONS FOR MOX RADIOACTIVE MATERIALS ...................................D-1 

D.1 Experimental Benchmarks.........................................................................................................................................D-1 

D.2 Summary of Bias and Uncertainty Evaluation...........................................................................................................D-1 
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LIST OF FIGURES 

Figure A9-1. MOX Powder/Pellet Package .........................................................................................................................A9-4 

Figure A10-1 MOX Fresh Fuel Package.............................................................................................................................A10-4 
 

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 LIST OF TABLES 

Table 1. Typical Isotopic Mix in Weight Percent for Various Grades of Plutonium as Specified in the 3013 

Standard................................................................................................................................................................2 

Table C-1. Typical Isotopic Mix in Weight Percent for Various Grades of Plutonium as Specified in the 3013 

Standard............................................................................................................................................................C-1 

Table C-2. Weight Percents for Heavy Metal Isotopes Chosen for Comparing MOX with LEU for Various Grades of 

Plutonium .........................................................................................................................................................C-2 

Table C-3a. Ratio of Heat Generation Rate for MOX-Fresh Fuel Composed of Various Grades of Plutonium Relative 

to LEU Fresh Fuel ............................................................................................................................................C-2 

Table C-3b. Specific Decay Parameters for MOX-Relevant Nuclides.................................................................................C-3 

Table C-4a. Gamma Emission Rates Relative to the LEU Gamma Emission Rate and Average Gamma Energies for 

Equal Weights of Some Nuclides of Uranium, Plutonium, Neptunium and Americium at 20 Years Decay 

Time..................................................................................................................................................................C-4 

Table C-4b. Ratio of Gamma Emission Rate for Gamma Energies Exceeding 100 keV for MOX-Fresh Fuel 

Composed of Various Grades of Plutonium Relative to LEU-Fresh Fuel........................................................C-5 

Table C-5a. Neutron Emission Rates Relative to the LEU Neutron Emission Rate and Average Gamma Energies for 

Equal Weights of Some Nuclides of Uranium, Plutonium and Americium for (α,n) with 17O and 18O, 

Spontaneous Fission (SF), and the Sum of All Three (Total) Neutron Emission Processes.............................C-5 

Table C-5b. Ratio of Neutron Emission Rate for MOX-Fresh Fuel Composed of Various Grades of Plutonium 

Relative to LEU-Fresh Fuel..............................................................................................................................C-6 

Table D-1. Important Characteristics of Lattice Experiments with Weight Percent of Pu/(Pu+U) to 5%..........................D-4

Table D-2. Important Characteristics of Lattice Experiments with Weight Percent of Pu/(Pu+U) from 5% to 15%.........D-5 

Table D-3. Important Characteristics of Lattice Experiments with Weight Percent of Pu/(Pu+U) Greater than 15% .......D-6 

Table D-4. Important Characteristics of Tank Experiments with Weight Percent of Pu/(Pu+U) to 31%...........................D-7 

Table D-5. Important Characteristics of Tank Experiments with Weight Percent of Pu/(Pu+U) Greater than 31% ..........D-8 
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ACRONYMS AND ABBREVIATIONS 

Code of Federal Regulations CFR 

fuel grade FG 

hypothetical accident conditions HAC 

low-enriched uranium LEU 

mixed uranium-plutonium oxide MOX 

maximum normal operating pressure MNOP 

normal conditions of transport NCT 

Nuclear Regulatory Commission NRC 

power grade PG 

radioactive materials RAM 

Regulatory Guide RG 

Safety Analysis Report SAR 

spontaneous fission SF 

weapon grade WG 

worst-case difference WCD 

1

INTRODUCTION 

The Standard Review Plan for Transportation Packages for Radioactive Material (NUREG 1609)1-1 provides 

guidance for the U.S. Nuclear Regulatory Commission (NRC) safety reviews of packages used in the transport of 

radioactive materials (RAM) under Title 10 of the U.S. Code of Federal Regulations (CFR), Part 71 

(10 CFR Part 71). It is not intended as an interpretation of NRC regulations. NUREG-1609 supplements NRC 

Regulatory Guide (RG) 7.9, “Standard Format and Content of Part 71 Applications for Approval of Packaging for 

Radioactive Material,”1-2 for review of package applications. NUREG-1609 involves guidance for reviewing 

radioactive material packagings intended for transport of radioactive materials, including low-enriched uranium, 

and low-enriched uranium oxide (LEU) in any form. 

Following long-standing, existing regulatory requirements, it should be noted that 1) the transport of large 

quantities of plutonium, in liquid form, is not authorized, and 2) that air transport is not authorized (except for 

packages that meet 10 CFR 71.88). 

This current report is not a stand-alone document but is intended as a supplement to NUREG-1609. This 

supplement to NUREG-1609 is intended to provide details on package review guidance resulting from the 

significant differences between contents of low-enriched uranium (LEU) oxide radioactive material (RAM) and 

contents of mixed uranium-plutonium oxide (MOX) RAM. While comparisons of MOX RAM with LEU RAM 

include powder, fresh-fuel rods, and fresh-fuel assemblies, they do not include comparisons with high-enriched 

uranium (HEU) or plutonium as metal, oxide, or nitrate in the forms of powder, pellets, fresh-fuel rods, or fresh 

fuel assemblies that are covered by NUREG-1609. Nothing contained in this document may be construed as 

having the force and effect of NRC regulations (except where the regulations are cited), or as indicating that 

applications supported by safety analyses and prepared in accordance with RG 7.9 will necessarily be approved, 

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

regulations. The principal purpose of this supplement to NUREG-1609 is to ensure the quality and uniformity of 

staff reviews of packagings intended for transport of MOX RAM. 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 the NRC staff review 

process. In particular, this supplemental guidance, together with NUREG-1609, assists potential applicants by 

indicating one or more acceptable means of demonstrating compliance with the applicable regulations. 

This supplement to NUREG-1609 is organized in the same manner as NUREG-1609, and has the identical 

numbering of subsections as found in that document. In addition, appendices found in this supplement are labeled 

to allow this report to be completely merged with NUREG-1609 without needing to change any labeling. For 

example, NUREG-1609 has two appendices labeled A and B with Appendix A being composed of eight parts. 

This supplement has four appendices, with two labeled A-9 and A-10, and the other two labeled C and D. 

Appendix A-9 contains information on a packaging for MOX powder and/or pellets, and Appendix A-10 contains 

information on a packaging for unirradiated MOX fuel (also referred to as MOX-fresh fuel in this document). 

Appendix C contains information on differences between thermal and radiation properties of MOX RAM and 

LEU RAM. Appendix D contains information on benchmark considerations for MOX RAM. 

The subsection numbering structure within each section in NUREG-1609 is the same. The fifth subsection is 

labeled Review Procedures, and lists different review approaches for any subsection. These different review 

approaches in each Review Procedures subsection in this supplement are consequences of significant differences 

between LEU-RAM or MOX-RAM packages that potentially affect the compliance corresponding to the section 

of the Safety Analysis Report (SAR) in question with NRC regulations. If no significant differences exist for a 

particular subsection, that particular subsection is omitted from this supplement to NUREG-1609. 

There are three generic differences that can affect each major section and potentially cause significant differences 

in review procedures between packages containing LEU RAM and MOX RAM. First, LEU RAM requires Type 

A Fissile Materials packagings, whereas MOX RAM requires Type B Radioactive Materials packagings, as 
 

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specified by 49 CFR 173.431, 10 CFR 71.0(a)(2), and 10 CFR 71.51. Second, the containment requirements for 

LEU and MOX RAM are defined in 10 CFR 71.51, 10 CFR 71.55 and 49 CFR 173.24(b)(1). These features will 

be covered in more detail in Section 4 on containment. 

Another of the potentially significant differences between LEU RAM and MOX RAM results because MOX 

RAM can have larger heat generation rates, photon emission rates, and neutron emission rates due to decay (see 

Appendix C for a discussion of these features). Another potentially significant difference between LEU RAM 

and MOX RAM results because the plutonium in MOX is a significant radiological hazard, and this can affect the 

allowable leakage requirements imposed on a package. In addition, several other differences are also noted in 

some of the sections that warrant review attention. 

The U.S. DOE Standard DOE-STD-3013-2000 (herein called the 3013 Standard)1-3 will be used to specify typical 

grades of plutonium employed to make the MOX-fresh fuel discussed in the text and in Appendix C. The actual 

plutonium compositions found in practice may not match these compositions exactly, but these grades can be 

considered typical for the purposes of this supplement to NUREG-1609. The 3013 Standard gives weight 

percents for various isotopes in various grades of plutonium. They are reproduced in the following table as 

representative values for typical grades of plutonium used to fabricate MOX-fresh fuel. 

Table 1. Typical Isotopic Mix in Weight Percent for Various Grades of 

Plutonium as Specified in the 3013 Standard 

Isotope Weapon Grade Fuel Grade Power Grade 

238Pu 0.05 0.1 1.0 

239Pu 93.50 86.1 62.0a

240Pu 6.00 12.0 22.0 

241Pu 0.40 1.6 12.0 

242Pu 0.05 0.2 3.0 

Note: 236Pu and 241Am could be present but are not included in the 3013 Standard a 63% reduced to 62% so that sum is 100% 

Comparable information can also be found on the website of the American Nuclear Society.1-4, 1-5 
3

References 

1-1. U.S. Nuclear Regulatory Commission, “Standard Review Plan for Transportation Packages for Radioactive 

Material,” NUREG-1609, U.S. Government Printing Office, Washington, D.C., 1999. 

1-2. U.S. Nuclear Regulatory Commission, “Standard Format and Content of Part 71 Applications for Approval 

of Packaging for Radioactive Material,” Regulatory Guide 7.9, Rev. 1, 1986. 

1-3. U.S. Department of Energy, “Stabilization, Packaging, and Storage of Plutonium-Bearing Materials,” U.S. 

DOE Standard, DOE-STD-3013-2000, Washington D.C., September 2000. 

1-4. American Nuclear Society, Position Statement #42, Protection and Management of Plutonium, August 

1995. 

1-5. American Nuclear Society, Position Statement #47-bi, Disposition of Surplus Weapons Plutonium Using 

Mixed Oxide Fuel, Background Report, November 2002. 
 

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1 GENERAL INFORMATION REVIEW 

1.5 Review Procedures 

The general information review of NUREG-1609 is normally applicable to the review of both MOX-RAM and 

LEU-RAM packages. In this section, no significant deviations exist in the review procedures and considerations 

for the two packages. This section considers each of the subsections of Section 1.5 (Review Procedures) of 

NUREG-1609, and highlights the special considerations or attention needed for MOX-RAM packages. In 

subsections where no significant differences were found, that particular subsection has been omitted from this 

section. 

For all packages, the general information review is based in part on the descriptions and evaluations presented in 

the Structural Evaluation, Thermal Evaluation, Containment Evaluation, Shielding Evaluation, Criticality 

Evaluation, Operating Procedures, and Acceptance Tests and Maintenance Program sections of the SAR. 

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. 

1.5.1 Introduction 

Except for the need to use Type B packages, there should be no significant differences in the general methods to 

be used for review of LEU-RAM or MOX-RAM packages. Note that essentially all packages shipping bulk 

unirradiated MOX powder and pellets will be designated as Category I Packages by RG 7.11. 

1.5.2 Package Description 

Except for the possible need to use Type B packages, there should be no significant differences in the general 

methods to be used for review of LEU-RAM or MOX-RAM packages. 
 

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2 STRUCTURAL REVIEW 

2.5 Review Procedures 

The structural review of NUREG-1609 is normally applicable to the review of both MOX-RAM and LEU-RAM 

packages. However, details of the structural review procedures and considerations for LEU-RAM and 

MOX-RAM packages can be quite different. The differences arise from the generic differences in gamma and 

neutron emission rates of the two radioactive materials (see Appendix C). The presence of plutonium makes 

MOX RAM more of a radiological hazard than is LEU. The magnitude of the difference depends on the grade of 

plutonium in the contents. (See Table 1.) These conditions lead to the requirement for a Type B Fissile Materials 

package for containment of the MOX-RAM contents, while LEU packages are usually designated a Type A 

Fissile Materials package. This section considers each of the subsections of Section 2.5 (Review Procedures) of 

NUREG-1609, and highlights the special considerations or attention needed for the MOX-RAM packages. In 

subsections where no significant differences were found, that particular subsection has been omitted from this 

section. 

For all packages, the structural review is based in part on the descriptions and evaluations presented in the 

General Information and the Thermal Evaluation sections of the SAR. Similarly, results of the structural review 

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

Evaluation, Criticality Evaluation, Operating Procedures, and Acceptance Tests and Maintenance Program. 

2.5.1 Description of Structural Design 

The MOX-RAM package usually has a containment system while almost all LEU-RAM packages do not. In 

addition, there may be more neutron absorbers for criticality control and more shielding materials in the 

MOX-RAM package than in the LEU package. The structural reviewer should identify and understand the 

precise structural design and expected structural performance of all the packaging components that serve the 

containment, shielding, and criticality control functions. 

2.5.1.2 Identification of Codes and Standards for Package Design 

There should be no significant differences in the general methods to be used for review of LEU-RAM or 

MOX-RAM packages. However, the design code sections for the LEU-RAM package may not be appropriate for 

the MOX-RAM package, and the code sections for MOX-RAM packages can be over-restrictive for LEU-RAM 

packages. 

2.5.2 Materials 

2.5.2.1 Material Properties and Specifications 

See also the additional information presented in Section X.5.2.4 of ISG-15. 

2.5.2.2 Prevention of Chemical, Galvanic, or Other Reactions 

Depending on the packaging materials, there can be significantly greater chemical reactions in a MOX-RAM 

package than in an LEU-RAM package. The structural reviewer should be alert in identifying undesirable 

conditions. Powder contents with high moisture are particularly susceptible to gas generation due to radiolysis, 

whereas LEU powder contents are not. 

See also the additional information presented in Section X.5.3.1 of ISG-15. 
 

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2.5.2.3 Effects of Radiation on Materials 

Depending on the packaging materials, there can be significantly greater radiation effects in a MOX-RAM 

package than in an LEU-RAM package. The structural reviewer should be alert in identifying undesirable 

conditions. Powder contents with high moisture are particularly susceptible to gas generation due to radiolysis, 

whereas LEU powder contents are not. 

2.5.3 Fabrication and Examination 

There should be no significant differences in the general methods to be used for review of LEU-RAM or 

MOX-RAM packages. However, the MOX-RAM and LEU-RAM packages may not need to use the same design 

code sections. The design code sections for the LEU-RAM package may not be appropriate for the MOX-RAM 

package, and the code sections for MOX-RAM packages can be over-restrictive for LEU-RAM packages. 

2.5.4 Lifting and Tie-Down Standards for All Packages 

2.5.4.1 Lifting Devices 

There should be no significant differences in the general methods to be used for review of LEU-RAM or 

MOX-RAM packages. However, if the MOX-RAM package has a containment system in addition to fuel 

cladding, the failure of the lifting devices must not impair the ability of the containment system to meet its 

containment functions during normal conditions of transport (NCT) or hypothetical accident conditions (HAC). 

2.5.4.2 Tie-Down Devices 

There should be no significant differences in the general methods to be used for review of LEU-RAM or 

MOX-RAM packages. However, if the MOX-RAM package has a containment system in addition to fuel 

cladding, the failure of the tie-down devices must not impair the ability of the containment system to meet its 

containment functions during NCT or HAC. 

2.5.6 Structural Evaluation under Normal Conditions of Transport 

The only significant difference in the structural review of the LEU RAM and MOX RAM under NCT is in the 

containment acceptance criteria. Being a Type B package, the MOX-RAM package must meet the quantitative 

containment criterion specified in 10 CFR Part 71 for NCT. On the other hand, the Type-A LEU-RAM package 

does not need to meet any quantitative containment criteria. 

2.5.7 Structural Evaluation under Hypothetical Accident Conditions 

Similar to the structural review under NCT, the only significant difference in the structural review of the LEU 

RAM and MOX RAM under the HAC is in the containment acceptance criteria. A Type-A LEU-RAM package 

does not necessarily need to meet a quantitative release limit, whereas a Type-B MOX-RAM package does need 

to meet a quantitative release limit. Thus, the structural evaluation needs to demonstrate the structural integrity of 

the containment system under HAC for the MOX-RAM package but not for the LEU-RAM package. Note that 

the packaging of a LEU-RAM package, as well as a MOX-RAM package, must be robust enough to prevent 

criticality during an HAC. 

References 

2-1. U.S. NRC, Spent Fuel Project Office, Interim Staff Guidance -15 (ISG-15), “Materials Evaluation” 
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3 THERMAL REVIEW 

3.5 Review Procedures 

The thermal review of NUREG-1609 is normally applicable to the review of both MOX-RAM and LEU-RAM 

packages. In this section, there may be some differences in emphasis in the thermal review procedures that arise 

from generic differences between LEU-RAM and MOX-RAM packagings and contents. Plutonium has a higher 

specific activity of energetic, and short-ranged decay particles (~5 MeV alphas), than does LEU RAM. This 

results in higher specific content decay heat rates in the MOX-RAM packages than in other LEU-RAM packages 

(see Appendix C). Also MOX-fresh-fuel rods and assemblies may need special attention in some of the 

subsections. This section considers each of the subsections of Section 3.5 (Review Procedures) of NUREG-1609, 

and highlights the special considerations or attention needed for MOX-RAM packages. In subsections where no 

significant differences were found, that particular subsection has been omitted from this section. 

For all packages, the thermal review is based in part on the descriptions and evaluations presented in the General 

Information and the Structural Evaluation sections of the SAR. Similarly, results of the thermal review are 

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

Evaluation, Criticality Evaluation, Operating Procedures, and Acceptance Tests and Maintenance Program. 

3.5.1 Description of Thermal Design 

3.5.1.2 Content Decay Heat 

There should be no significant differences in the general methods to be used for review of LEU-RAM or MOXRAM packages. The decay heat generation rate for MOX RAM will be larger by up to four or five orders of 

magnitude (see Appendix C), than the usually negligible decay heat generation rate for LEU RAM. 

3.5.1.3 Summary Tables of Temperatures 

For MOX-fresh-fuel rods, the summary tables of the temperatures of package components including, but not 

limited to, the fuel/cladding, basket, impact limiters, containment vessel, seals, shielding, and neutron absorbers 

shall be consistent with the temperatures presented in the General Information and Structural Evaluation sections 

of the SAR for the NCT and HAC. 

3.5.3 General Considerations for Thermal Evaluations 

3.5.3.2 Evaluation by Test 

For MOX-fresh-fuel rods and assemblies, temperature-sensing devices shall be placed on the fuel basket and fuel 

rods. 

3.5.4 Thermal Evaluation under Normal Conditions of Transport 

3.5.4.2 Maximum Normal Operating Pressure 

For MOX-fresh-fuel rods and assemblies, the thermal evaluation shall determine the maximum normal operating 

pressure (MNOP) when the package has been subjected to the heat condition for one year. The reviewer should 

ensure that the evaluation has considered all possible sources of gases such as those present in the package at 

closure, and/or fill gas released from the MOX-fresh-fuel rods. 

The evaluation of MOX powder and pellets on the MNOP should be similar to that of PuO2 powder and pellets. 

For powders, however, it should be noted that there is the possibility that hydrogen, and/or other gases, may be 

produced from the thermal- or radiation-induced decomposition of the moisture associated with impure 
 

10

plutonium-containing oxide powders. Given that the ratio of plutonium oxide powder to uranium oxide powder 

with respect to the total amount of MOX powder is expected to be small, any additional contributions from such 

gases should also expected to be small. 

By the time the MOX powders are converted to fuel pellets, the processing temperatures should have removed all 

of the impurities from the plutonium oxide. From this point on (i.e., from MOX pellets, to MOX fuel rods, to full 

fuel assemblies), the evaluations of MOX pellets and LEU pellets should be virtually identical. 

3.5.4.3 Maximum Thermal Stresses 

For MOX-fresh-fuel rods and assemblies, the thermal stresses from temperature gradients through the 

components such as the containment vessel and fuel/clad shall be determined. These thermal stresses due to the 

temperature gradients will, in general, be small because the temperature gradients through the metal are small. 

For MOX powders and fuel pellets, the processing temperatures should have removed all of the impurities from 

the plutonium oxide. From this point on (i.e., from MOX pellets, to MOX fuel rods, to full fuel assemblies), the 

evaluations of MOX pellets and LEU pellets should be virtually identical. 

3.5.5 Thermal Evaluation under Hypothetical Accident Conditions 

3.5.5.1 Initial Conditions 

For MOX-fresh-fuel rods and assemblies, the internal heat load of the MOX-fresh-fuel contents shall be at its 

maximum allowable power, unless a lower power, consistent with the temperature and pressure, is more 

unfavorable. 

For MOX powders and fuel pellets, the internal heat load of the contents shall be at its maximum allowable 

power, unless a lower power, consistent with the temperature and pressure, is more unfavorable. 

3.5.5.2 Fire Test Conditions 

For MOX-fresh-fuel rods and assemblies, the internal heat load of the MOX-fresh-fuel contents shall be at its 

maximum allowable power, unless a lower power, consistent with the temperature and pressure, is more 

unfavorable. 

For MOX powders and fuel pellets, the internal heat load of the contents shall be at its maximum allowable 

power, unless a lower power, consistent with the temperature and pressure, is more unfavorable. 

3.5.5.3 Maximum Temperatures and Pressures 

For MOX-fresh-fuel rods and assemblies, the possible increases in gas inventory (e.g., from an unlikely failure of 

a fuel rod) shall be considered in the pressure determination. 

For MOX powders and fuel pellets, the processing temperatures should have removed all of the impurities from 

the plutonium oxide. The only additional increase in pressure should be result of any helium released from the 

contents, as a result of the increased temperature. But, because any increase in temperature as a result of the 

thermal testing should be small when compared to the processing temperatures, any increase in pressure should be 

small. 
11

3.5.6 Appendix 

3.5.6.3 Applicable Supporting Documents or Specifications 

For MOX-fresh-fuel rods and assemblies, the applicable sections from reference documents shall be included. 

These documents may include the test plans used for the thermal tests, the thermal specifications of O-rings, fuel 

clad, and other components, and the documentation of the thermal properties of non-ASME-approved materials 

used in the package. 

Similar documentation should also be included for MOX powders and pellets. 
 

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4 CONTAINMENT REVIEW 

4.5 Review Procedures 

The containment review of NUREG-1609 is normally applicable to MOX-RAM and LEU-RAM packages. There 

are two major differences between the review approaches for LEU-RAM and MOX-RAM packagings: 

Unirradiated LEU packagings have traditionally come under the heading of Type A Fissile Materials packagings. 

MOX-RAM packagings, due to the intentional incorporation of the plutonium, can only be considered Type B 

Radioactive Materials packagings and requires at least one containment system, as defined in 10 CFR Part 71. 

These differences will be discussed in more detail below. This section considers each of the subsections of 

Section 4.5 (Review Procedures) of NUREG-1609, and highlights the special considerations or attention needed 

for MOX-RAM packages. In subsections where no significant differences were found, that particular subsection 

has been omitted from this section. 

For all packages, the containment review is based in part on the descriptions and evaluations presented in the 

General Information, Structural Evaluation, and Thermal Evaluation sections of the SAR. Similarly, results of the 

containment review are considered in the review of the SAR sections on Operating Procedures, and Acceptance 

Tests and Maintenance Program. 

4.5.1 Description of the Containment System 

4.5.1.1 Containment Boundary 

The containment boundary information in section 4.5.1.1 of NUREG-1609 will have to be considered for each 

containment boundary for MOX packages. (See also the additional information provided in Section X.5.2.9 

(Seals) of ISG-15.) 

4.5.1.2 Special Requirements for Plutonium 

The text of §71.63 of 10 CFR Part 71 states: 

“Shipments containing plutonium must be made with the contents in solid form, if the contents 

contain greater than 0.74 TBq (20 Ci) of plutonium.” 

4.5.2 General Considerations 

4.5.2.1 Type A Fissile Packages 

Because MOX-RAM packagings can only be considered Type B Radioactive Materials packagings, the 

containment-related information provided in Section 4.5.2.1 of NUREG-1609 can only be applied in the context 

of Type B Radioactive Materials packagings. 

4.5.2.2 Type B Packages 

Section 4.5.2.2 of NUREG-1609 specifies that Type B packagings must satisfy the quantified release rates in 

§71.51 of 10 CFR Part 71. An acceptable method for satisfying these requirements is provided in ANSI N14.5.4-1 

Additional information for determining containment criteria is provided in NUREG/CR-6487.4-2 
 

14

4.5.3 Containment under Normal Conditions of Transport (Type B Packages) 

4.5.3.1 Containment Design Criteria 

Following the methodology suggested in NUREG/CR-6487, the containment design criteria for packagings 

designed to transport MOX-RAM powders and/or MOX-RAM pellets will typically be on the order of 10-6

reference cm3

/sec for NCT. The containment design criteria will have to be applied to each containment system 

independently. A full justification for the containment design criteria will have to be provided by the applicant in 

their calculations for all containment systems. 

Although a leakage rate criterion on the order of 10-6 reference cm3

/sec may appear to be somewhat restrictive, 

leakage rates of this order of magnitude are actually relatively easy to meet in practice for packagings of this type. 

In many cases, the applicant may find it easier to default to the leaktight criterion specified in ANSI N14.5, i.e., 

≤ 1 × 10-7 reference cm3

/sec. Using this criterion also eliminates the applicant’s need to fully justify containment 

design criteria calculations. 

For packagings designed to transport MOX-fresh-fuel rods and/or MOX-fresh-fuel assemblies, which would not 

normally be considered to be dispersible radioactive material, the containment design criteria will be dependent 

on how the applicant defines the releasable source term in their calculations. Thus, although a single containment 

system will be specified for this type of packaging, the allowable leakage rate test criteria could be relatively high, 

i.e., >10-3 reference cm3

/sec. For criteria as high as these, an applicant will have to fully justify their containment 

design criteria calculations. As an alternative, the staff has determined that an NCT containment design criterion 

of ≤ 1 × 10-4 reference cm3

/sec is conservative for packagings of this type. If the applicant adopts this 

conservative criterion, the applicant is not required to provide any containment calculations to justify their 

position for NCT, beyond showing that the fuel remains largely intact such that a lower reference rate is not 

required. 

4.5.4 Containment under Hypothetical Accident Conditions (Type B Packages) 

4.5.4.1 Containment Design Criterion 

The methodology presented in NUREG/CR-6487 suggests that the containment design criteria for packagings 

designed to transport MOX-RAM powders and/or MOX-RAM pellets could easily be higher for HAC than for 

NCT. In many cases, however, the applicant may find it useful to adopt the same leaktight containment design 

criteria for HAC defined above for NCT, i.e., ≤ 1 × 10-7 reference cm3

/sec. 

If the applicant adopts this conservative criterion, the applicant is not required to provide any containment 

calculations to justify their position for HAC. For those situations where the applicant adopts a higher criteria for 

HAC, the applicant must fully justify their containment design criteria calculations. 

Because the methodology is, again, totally dependent on how the applicant defines the releasable source term in 

their calculations, the methodology presented in NUREG/CR-6487 suggests that the containment design criteria 

for packagings designed to transport MOX-fresh-fuel rods and/or MOX-fresh-fuel assemblies could easily fall 

into the 10-2 or higher reference cm3

/sec region for HAC. For these criteria, a full justification for the containment 

design criteria will have to be provided by the applicant in their calculations for HAC. As an alternative, the staff 

has determined that an HAC containment design criterion of ≤ 1 × 10-4 reference cm3

/sec is conservative for 

packagings of this type. If the applicant adopts this conservative criterion, the applicant is not required to provide 

any containment calculations to justify their position for HAC. 
15

References 

4-1. Institute for Nuclear Materials Management, “American National Standard for Radioactive Materials — 

Leakage Tests on Packages for Shipment,” ANSI N14.5-1997, New York, NY, 1998. 

4-2. U.S. Nuclear Regulatory Commission, “Containment Analysis for Type B Packages Used to Transport 

Various Contents,” NUREG/CR-6487, U.S. Government Printing Office, Washington, D.C., 1996. 

4-3. U.S. NRC, Spent Fuel Project Office, Interim Staff Guidance -15 (ISG-15), “Materials Evaluation” 
 

16

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5 SHIELDING REVIEW 

5.5 Review Procedures 

The shielding review of NUREG-1609 is normally applicable to the review of both MOX-RAM and LEU-RAM 

packages. In this section, however, a few significant deviations may exist in the review procedures and 

considerations for the two packages. This section considers each of the subsections of Section 5.5 (Review 

Procedures) of NUREG-1609 and highlights the special considerations or attention needed for MOX-RAM 

packages. In subsections where no significant differences were found, that particular subsection has been omitted 

from this section. 

For all packages, the shielding review is based in part on the descriptions and evaluations presented in the General 

Information, Structural Evaluation, and Thermal Evaluation sections of the SAR. Similarly, results of the 

shielding review are considered in the review of the SAR sections on Operating Procedures and Acceptance Tests 

and Maintenance Program. 

See also the additional information provided in Section X.5.2.6 of ISG-15. 

5.5.2 Radiation Source 

5.5.2.1 Gamma Source 

Although the decay photon emission rate for MOX RAM can be larger than the decay photon emission rate for 

LEU RAM by one or more orders of magnitude (see Appendix C), there should be no significant differences in 

the general methods to be used for review of LEU-RAM or MOX-RAM packages. Appendix C includes 

information pertinent to gamma emission rates for various isotopes of some transuranic elements. In addition, 

Appendix C includes information on gamma emission rates from MOX RAM containing different grades of 

plutonium. Note, that 236Pu, 238Pu, 241Pu and 241Am should be included in the source term when present in the 

contents. 

5.5.2.2 Neutron Source 

One potentially significant difference in the review approach between LEU RAM and MOX RAM is that the 

neutron dose rate can be much larger than the gamma dose rate. This means that particular care is necessary to 

determine the appropriate neutron source strength. The contribution from (α,n) reactions can be large relative to 

spontaneous fission for MOX RAM. Depending on the methods used to calculate these source terms, the 

applicant might still determine the energy group structure independently for spontaneous fission and (α,n) 

reactions. However, it is generally necessary to include contributions from both spontaneous fission and (α,n) 

reactions. Also neutron multiplication effects can be important sources of additional neutrons and should be 

included in the shielding analysis (most modern radiation transport codes inherently produce multiplication 

neutrons). Appendix C includes information pertinent to neutron emission rates for various isotopes of some 

transuranic elements. In addition, Appendix C includes information on neutron emission rates from MOX RAM 

containing different grades of plutonium. Note, that 238Pu, 239Pu, 240Pu, 242Pu and 241Am should be included in the 

source term when present in the contents. 

5.5.4 Shielding Evaluation 

Other than including a neutron source term in calculations along with a photon source term, there should be no 

significant differences in the general methods to be used for review of LEU-RAM or MOX-RAM packages. 
 

18

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19

6 CRITICALITY REVIEW 

6.5 Review Procedures 

The criticality review of NUREG-1609 is normally applicable to the review of both MOX-RAM and LEU-RAM 

packages. In this section, however, a few significant deviations may exist in the review procedures and 

considerations for the two packages. This section considers each of the subsections of Section 6.5 (Review 

Procedures) of NUREG-1609 and highlights the special considerations or attention needed for MOX-RAM 

packages. In subsections where no significant differences were found, that particular subsection has been omitted 

from this section. 

For all packages, the criticality review is based in part on the descriptions and evaluations presented in the 

General Information, Structural Evaluation, and Thermal Evaluation sections of the SAR. Similarly, results of the 

criticality review are considered in the review of the SAR sections on Operating Procedures and Acceptance Tests 

and Maintenance Program. 

See also the additional information provided in Section X.5.2.7 of ISG-15. 

6.5.7 Benchmark Evaluations 

6.5.7.1 Applicability of Benchmark Experiments 

There are considerably fewer criticality benchmark experiments using MOX RAM than LEU RAM. Therefore, 

differences between the package and benchmarks may be more substantial for MOX RAM than for LEU RAM, 

so it might be more difficult to properly consider them. Appendix D discusses the availability of MOX-RAM 

benchmarks and their important characteristics from a criticality perspective. Appendix D also discusses how a 

reviewer might choose a set of appropriate MOX-RAM benchmarks. 

6.5.7.2 Bias Determination 

Because of the lack of criticality benchmark experiments using MOX RAM, assigning a bias value for 

benchmarks may be more difficult than for LEU RAM. Appendix D discusses MOX-RAM benchmarks and how 

a reviewer might determine a conservative bias value from comparisons between benchmark experiments and 

criticality calculations of the multiplication coefficient for those experiments. Appendix D also discusses how a 

reviewer might determine a conservative bias value for situations when the number of MOX-RAM benchmarks is 

less than desirable. 

References 

6-1. U.S. NRC, Spent Fuel Project Office, Interim Staff Guidance -15 (ISG-15), “Materials Evaluation” 
 

20

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21

7 OPERATING PROCEDURES REVIEW 

7.5 Review Procedures 

The operating procedures review of NUREG-1609 is normally applicable to the review of both MOX-RAM and 

LEU-RAM packages. In this subsection, however, two significant deviations exist in the review procedures and 

considerations for the two packages. This section considers each of the subsections of Section 7.5 (Review 

Procedures) of NUREG-1609, and highlights the special considerations or attention needed for MOX-RAM 

packages. In subsections where no significant differences were found, that particular subsection has been omitted 

from this section. 

For all packages, the operating procedures review is based in part on the descriptions and evaluations presented in 

the General Information, Structural Evaluation, Thermal Evaluation, Containment Evaluation, Shielding 

Evaluation, Criticality Evaluation, and Acceptance Tests and Maintenance Program sections of the SAR. 

7.5.1 Package Loading 

7.5.1.2 Loading of Contents 

There are two major differences between the review approaches for LEU-RAM and MOX-RAM packagings. 

1. Unirradiated LEU packagings have traditionally come under the heading of Type A Fissile Materials 

packagings. MOX RAM packagings, due to the intentional incorporation of the plutonium, can only be 

considered Type B Radioactive Materials packagings. 

2. Because they are Type A packagings, there are no leakage test requirements that are normally associated with 

LEU-RAM packagings. MOX-RAM packagings, on the other hand, due to the intentional incorporation of 

plutonium, can only be considered Type B radioactive materials packagings. These packagings must meet the 

leakage test requirements specified in 10 CFR 71.51. (See also Section 4, above.) 

For purposes of the Operating Procedures review, the primary leakage test requirement that should be noted is the 

Preshipment Leakage Test. For MOX-RAM packagings, the Preshipment Leakage Test requirement will usually 

be 10-3 reference cm3

/sec, for all containment systems, prior to every shipment. The Preshipment Leakage Test 

rates should be determined in accordance with ANSI N14.5. 

7.5.2 Package Unloading 

7.5.2.2 Removal of Contents 

There should be no significant differences in the general methods to be used for review of LEU-RAM or 

MOX-RAM packages. 

References 

7-1. Institute for Nuclear Materials Management, “American National Standard for Radioactive Materials — 

Leakage Tests on Packages for Shipment,” ANSI N14.5-1997, New York, NY, 1998. 
 

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23

8 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW 

8.5 Review Procedures 

8.5.1 Acceptance Tests 

The acceptance tests review of NUREG-1609 is normally applicable to the review of both MOX-RAM and 

LEU-RAM packages. In this subsection, however, significant deviations exist in the review procedures and 

considerations for the two packages. This section considers each of the subsections of Section 8.5.1 (Review 

Procedures) of NUREG-1609, and highlights the special considerations or attention needed for MOX-RAM 

packages. In subsections where no significant differences were found, that particular subsection has been omitted 

from this section. 

For all packages, the acceptance tests review is based in part on the descriptions and evaluations presented in the 

General Information, Structural Evaluation, Thermal Evaluation, Containment Evaluation, Shielding Evaluation, 

Criticality Evaluation, and Operating Procedures sections of the SAR. 

The guidance in NUREG-1609 is applicable to either LEU-RAM or MOX-RAM packagings. However, 

unirradiated LEU-RAM packagings have traditionally come under the heading of Type A Fissile Materials

packagings, while MOX-RAM packagings come under the heading of Type B Radioactive Materials packagings. 

Although the information presented in Sections 8.5.1.1, 8.5.1.2, 8.5.1.3, 8.5.1.5, 8.5.1.6, and 8.5.1.7 of 

NUREG-1609 is applicable for either LEU-RAM or MOX-RAM packagings, it only applies in the context of 

Type A packaging for LEU-RAM or Type B packaging for MOX-RAM. 

8.5.1.4 Leakage Tests 

There are two major differences between the review approaches for LEU-RAM and MOX-RAM packagings. 

1. Unirradiated LEU RAM packagings have traditionally come under the heading of Type A Fissile Materials 

packagings. MOX RAM packagings, due to the intentional incorporation of the plutonium, can only be 

considered Type B Radioactive Materials packagings. 

2. Because they are Type A packagings, there are no leakage test requirements normally associated with LEURAM packagings. MOX-RAM packagings, on the other hand, must meet the leakage test requirements 

specified in 10 CFR 71.51. 

As noted in Section 4, the leakage test requirements for packages used to transport MOX-fresh-fuel rods and 

MOX-fresh-fuel assemblies may be substantially different from the requirements for packages used for the 

transport of MOX powders and pellets — they can easily be different by several orders of magnitude. For 

purposes of the Acceptance Tests review, it is important to verify that the packaging containment system(s) is 

subjected to the Fabrication Leakage Test requirements specified in ANSI N14.5. The acceptable leakage test 

criteria should be consistent with those identified in the Containment Evaluation section (i.e., Chapter 4) of the 

SAR. 
 

24

8.5.2 Maintenance Program 

The maintenance program review of NUREG-1609 is normally applicable to the review of both MOX-RAM and 

LEU-RAM packages. However, in this subsection, a few significant deviations exist in the review procedures and 

considerations for the two packages. This section considers each of the subsections of Section 8.5.2 (Review 

Procedures) of NUREG-1609 and highlights the special considerations or attention needed for MOX-RAM 

packages. In subsections where no significant differences were found, that particular subsection has been omitted 

from this section. 

For all packages, the maintenance program review is based in part on the descriptions and evaluations presented 

in the General Information, Structural Evaluation, Thermal Evaluation, Containment Evaluation, Shielding 

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

The guidance in NUREG-1609 is applicable to either LEU-RAM or MOX-RAM packagings. However, 

unirradiated LEU-RAM packagings have traditionally come under the heading of Type A Fissile Materials

packagings, while MOX-RAM packagings come under the heading of Type B Radioactive Materials packagings. 

Although the information presented in Sections 8.5.2.1, 8.5.2.3, 8.5.2.4, and 8.5.2.5 of NUREG-1609 is applicable 

for either LEU-RAM or MOX-RAM packagings, it only applies in the context of Type A packaging for LEURAM or Type B packaging for MOX-RAM. 

8.5.2.2 Leakage Tests 

There are two major differences between the review approaches for LEU-RAM and MOX-RAM packagings. 

1. Unirradiated LEU packagings have traditionally come under the heading of Type A Fissile Materials 

packagings. MOX RAM packagings, due to the intentional incorporation of the plutonium, can only be 

considered Type B Radioactive Materials packagings. 

Seal replacement procedures for Type A Fissile Materials packagings should follow typical Type A 

packaging procedures, with no additional leakage testing requirements normally specified. Seal replacement 

procedures for Type B Radioactive Materials packagings, on the other hand, should follow typical Type B 

packaging procedures, with additional leakage tests specified for all containment boundary seals. For Type B 

packagings, the requirements specified should be in agreement with the requirements specified in Section 4, 

and all leakage test requirements should be agreement with the requirements of ANSI N14.5.8-1 

2. Because they are Type A packagings, there are no leakage test requirements normally associated with LEURAM packagings. MOX-RAM packagings, on the other hand, must meet the leakage test requirements 

specified in 10 CFR 71.51. 

Seal replacement procedures for Type A Fissile Materials packagings should follow typical Type A 

packaging procedures, with no additional leakage testing requirements normally specified. Seal replacement 

procedures for Type B Radioactive Materials packagings, on the other hand, should follow typical Type B 

packaging procedures, with additional leakage tests specified for all containment boundary seals. For Type B 

packagings, the requirements specified should be in agreement with the requirements specified in Section 4, 

and all leakage test requirements should be agreement with the requirements of ANSI N14.5.8-1 

As was noted in Section 4, the leakage test requirements for packages used for the transport of MOX-fresh-fuel 

rods and MOX-fresh-fuel assemblies may be substantially different from the requirements for packages used for 

the transport of MOX powders and pellets — they can easily be different by several orders of magnitude. For 

purposes of the Maintenance Program review, it is important to verify that the packaging containment system(s) is 

subjected to the Maintenance Leakage Test and/or the Periodic Leakage Test requirements specified in 
25

ANSI N14.5. The acceptable leakage test criteria should be consistent with those identified in the Containment 

Evaluation section (i.e., Chapter 4) of the SAR. 

References 

8-1. Institute for Nuclear Materials Management, “American National Standard for Radioactive Materials — 

Leakage Tests on Packages for Shipment,” ANSI N14.5-1997, New York, NY, 1998. 
 

26

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27

APPENDICES 
 

28

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A9-1 

APPENDIX A9: MOX POWDER AND PELLET PACKAGES 

A9.1 Package Type 

A9.1.1 Purpose of Package 

The purpose of this type of package is to transport Type B quantities of MOX material (other than by air). 

A9.1.2 Description of a Typical Package 

A typical packaging consists of a containment vessels and an outer container that serves to confine the package 

internals. The outer container (confinement system) is a steel drum with a removable head and weather-tight 

gasket. The head usually is a bolted or clamped lid with a tamperproof seal. Vent holes near the top of the drum, 

which provide pressure relief from combustion gases or off-gassing from insulating materials under HAC, are 

capped or taped during NCT to prevent water inleakage. 

The inner containment vessel is a steel container, typically a stainless steel cylinder, with a maximum inner 

diameter of 0.127 m (5 in.), closed by a welded bottom cap and a welded top flange with a bolted lid. The lid, 

which is generally sealed by two O-rings, contains a leak-test port and sometimes a separate fill port for leak 

testing. 

A secondary containment vessel or product container may be used and may be designed similar to the primary 

containment vessel; and can include welded and bolted bottom cap and top flange, respectively; dual O-ring seals; 

a leak test port; and sometimes a separate fill port for leakage testing. (See, for example, Figure A9-1.) 

The contents are MOX powder or pellets. The MOX powder or pellets are generally placed in metal cans prior to 

loading into the containment vessel. Solid spacers are often used to maintain the position of the contents. 

A sketch of a typical package with an optional containment vessel is shown in Figure A9-1. 

A9.2 Package Safety 

A9.2.1 Safety Functions 

The principal functions of the package are to provide containment, shielding, and criticality control. Package 

design features that accomplish the containment and criticality functions might also provide adequate shielding to 

satisfy the requirements for nonexclusive-use shipment. Additional shielding may be required if significant 

quantities of certain isotopes, e.g., 236Pu, 238Pu, 241Pu or 241Am (from the decay of 241Pu) are present in the MOX 

material. 

A9.2.2 Safety Features 

• The steel drum and thermal insulating/impact absorbing material protect the containment vessel and contents 

under HAC and maintain a minimum spacing between packagings for criticality control. 

• The inner vessel provides containment of the radioactive material. 

• The diameter and volume of the inner containment vessel, together with limits on the fissile mass of the 

contents, ensure that a single package is subcritical, even with water inleakage. 
 

A9-2

• The containment vessel, thermal insulating/impact absorbing material, and steel drum maintain a minimum 

distance from the contents to the package surface and provide some attenuation to satisfy the shielding 

requirements. 

A9.2.3 Typical Areas of Review for Package Drawings 

• Containment vessel body 

- Materials specifications 

- Dimensions and tolerances, including maximum cavity dimensions 

- Fabrication codes or standards 

- Weld specifications, including codes or standards for nondestructive examination. 

• Containment vessel closures 

- Lid material specifications, dimensions, and tolerances 

- Bolt specifications, including number, size, material, and torque 

- Seal material specifications and size 

- Seal groove dimensions 

- Leak-test ports 

- Applicable codes and standards. 

• Spacers to position or displace fissile material 

- Material of construction 

- Dimensions and tolerances 

- Locations. 

• Thermal insulating/impact absorbing material 

- Type, and specifications 

- Dimensions and tolerances 

- Density. 

• Outer drum 

- Material specifications, including lid and closure device 

- Closure bolt specifications, including number, size, material, and torque 

- Dimensions 

- Applicable codes or standards. 

• Neutron poisons 
A9-3 

A9.2.4 Typical Areas of Safety Review 

• The structural review confirms that packaging integrity is maintained under both NCT and HAC, particularly, 

the drop, crush, and puncture tests. The review also verifies that the drum lid remains securely in place and 

the drum body and closure have no unacceptable openings that would cause the safety performance of the 

package to not meet regulatory standards, especially during the fire test. 

• The structural and thermal reviews evaluate the performance of the containment system under both NCT and 

HAC. Primary emphasis is on the structural integrity of the containment vessel and its closure, and on the 

thermal performance of the O-rings. 

• The structural and thermal reviews address the condition of the package and the minimum spacing between 

different packages under HAC. Damage to the outer drum and charring of the thermal insulating/impact 

absorbing material may result in closer spacing than that of NCT. 

• The thermal and containment reviews verify that the hydrogen concentration in any confined volume will not 

exceed 5% (by volume) during a period of one year. Shorter time periods have been approved based on 

detailed operating procedures to control and track the shipment of packages. 

• The shielding review evaluates the ability of the package to satisfy the allowed radiation levels during both 

NCT and HAC. 

• The criticality review addresses, in detail, both NCT and HAC. Key parameters for this review include the 

number of packages in the arrays, array configuration (pitch, orientation of packages, etc.), positioning of the 

containment vessel within the drum, moderation due to inleakage of water, the condition and quantity of 

spacing material, interspersed moderation between packages, preferential flooding of different regions within 

the package, packaging materials that provide moderation (e.g., plastics), and neutron poisons. 

• The review of operating procedures confirms that the containment vessel has been properly closed and its 

closure bolts are properly tightened to the specified torqued values, and that an appropriate pre-shipment leak 

test is performed. 

• The review of the acceptance tests and maintenance program verifies that appropriate fabrication and periodic 

verification leakage tests are performed. 

• The review of the acceptance tests and maintenance program also verifies that neutron poisons, if any, are 

present, and are subject to the appropriate tests to verify concentration. 
 

A9-4

Figure A9-1. MOX Powder/Pellet Package 
A10-1 

APPENDIX A10: UNIRRADIATED MOX FUEL PACKAGES 

A10.1 PACKAGE TYPE 

A10.1.1 Purpose of Package 

The purpose of this type of package is to transport unirradiated MOX-fuel assemblies and individual MOX-fuel 

rods. These packages are also referred to as “MOX fresh-fuel packages.” 

This appendix addresses those packages in which the contents are Type B quantities of fissile MOX material. The 

fissile MOX material can be in an entire assembly, or as individual fuel rods. 

A10.1.2 Description of a Typical Package 

A typical packaging consists of a metal outer shell, closed with bolts and elastomeric seals, and an impact-limiter 

system. An internal steel strongback, shock-mounted to the outer shell, supports one or more fuel assemblies, 

which are fixed in position on the strongback by clamps, separator blocks, and end support plates. Depending on 

the type of fuel, neutron poisons are sometimes used to reduce reactivity. Material surrounding the contents could 

be employed to shield against neutrons and/or gammas. If the package is used to transport individual fuel rods, a 

separate inner container is often employed. 

The contents of the package are unirradiated MOX in fuel assemblies or individual fuel rods. Because the 

majority of these packages are for commercial reactor fuel, the MOX is typically in the form of Zircaloy or 

stainless steel-clad plutonium-uranium dioxide pellets. 

A sketch of the typical package described above is shown in Figure A10-1. 

A10.1.3 Alternative Package Design 

In an alternative design for a MOX fresh-fuel package, the fuel assemblies are fixed in position by two or three 

steel channels, mounted by angle irons or a similar bracing structure to a thin-walled inner metal container. This 

inner container is in turn surrounded by a honeycomb material and enclosed in a metal outer shell. Foam 

cushioning material can be used to cushion the fuel assemblies and may be used between the inner and outer 

container. 

A10.2 PACKAGE SAFETY 

A10.2.1 Safety Functions 

The principal functions of the package are to provide containment, shielding, and criticality control. Package 

design features that accomplish the containment and criticality functions might also provide adequate shielding to 

satisfy the requirements for nonexclusive-use shipment. Additional shielding may be required if significant 

quantities of certain isotopes, e.g., 236Pu, 238Pu, 241Pu or 241Am (from the decay of 241Pu) are present in the MOX 

material. 
 

A10-2

A10.2.2 Safety Features 

• Impact limiters protect the outer shell and contents under HAC. They also provide thermal insulation for the 

O-ring seals of the outer shell. 

• A strongback with end support plates, clamps, and separators, maintains the fuel assemblies in a fixed 

position relative to each other and to any neutron poisons. 

• The metal outer shell of the packaging retains and protects the fuel assemblies, and may provide a minimum 

spacing between assemblies in an array of packages and provide some attenuation to satisfy the shielding 

requirements. 

• Neutron poisons, if present, reduce reactivity, and can provide some neutron shielding. 

• The metal outer shell also provides containment of the radioactive material. 

A10.2.3 Typical Areas of Review for Package Drawings 

• Outer shell (containment vessel body) 

- Material specifications 

- Dimensions and tolerances 

- Fabrication codes and standards 

- Weld specifications, including codes or standards for nondestructive examination. 

• Outer shell closure (containment vessel closure) 

- Lid materials, dimensions, and tolerances 

- Bolt specifications, including number, size, and torque 

- Seal material, size, and compression specifications 

- Seal groove dimensions 

- Leak-test ports 

- Applicable codes and standards. 

• Structural components (e.g., strongback, support plates, fuel clamps, and separators) that fix the position of 

fuel assemblies or relative position between fuel assemblies and poisons 

- Dimensions and material specifications 

- Methods of attachment 

- Applicable engineering codes or standards. 

• Thermal insulating/impact absorbing and/or shielding material 

- Type and specifications 

- Dimensions and tolerances 

- Density. 
A10-3 

• Neutron poisons 

- Dimensions and tolerances 

- Minimum poison content 

- Location and method of attachment 

- Material specifications 

- Applicable codes and standards. 

• Moderating materials, including plastics, wood, and foam 

- Location 

- Material properties. 

A10.2.4 Typical Areas of Safety Review 

• The general information review identifies the fuel assembly designs authorized in the package, including: 

- Number of, and arrangement of fuel assemblies 

- Number, pitch, and position of fuel rods, guide tubes, and channels 

- Overall assembly dimensions, including active fuel length 

- Authorization or restrictions on missing fuel rods or partial-length rods 

- Maximum amount of fissile material 

- Pellet dimensions and tolerances 

- Minimum cladding thickness 

- Fuel-clad gap and fill gas 

- Type, location, and concentration of burnable poisons, and other types of poisons 

- Type, location, and quantity of plastics, such as polyethylene, within the fuel assemblies. 

• The structural and thermal reviews evaluate the performance of the containment system under both NCT and 

HAC, particularly, the drop, crush (if needed), and puncture tests. Primary emphasis is on the structural 

integrity of the outer shell (containment vessel) and its closure, and on the thermal performance of the 

elastomeric seals. If the impact limiters provide thermal protection for the seals, the structural review also 

confirms the structural integrity of the impact limiters. 

• The structural review addresses possible damage to the impact limiters, outer shell, strongback, fuel assembly, 

neutron poisons (if present), clamps, separators, and end support plates to ensure that the fuel assemblies and 

neutron poisons are maintained in a fixed position relative to each other under hypothetical accident 

conditions. 

• The structural review also confirms the minimum spacing between fuel assemblies in different packages in an 

array under hypothetical accident conditions. Spacing can be affected by separation of the strongback from 

its shock mounts, failure of the shock mounts or fuel assembly clamps, and deformation of the outer shell of 

the package. 

• The thermal review evaluates the effect of the fire on outer-shell O-ring seals, neutron poisons, plastic 

sheeting, thermal insulation material (if present), or other temperature-sensitive materials under hypothetical 

accident conditions. 
 

A10-4

• The structural and thermal reviews address the condition of the package and the minimum spacing between 

different packages under HAC. Damage to the outer shell and charring of any thermal insulating/impact 

absorbing material (if present) may result in closer spacing than that of NCT. 

• The criticality review addresses both normal conditions of transport and hypothetical accident conditions. 

Key areas for this review include: 

- The number of packages in the array and the array configuration (pitch, orientation of packages, etc.). 

Because of movement of the strongback within the package and the location of poisons, the arrays might 

not be symmetrical. 

- Degree of moderation. Structural features, as well as packaging material such as plastic sheeting, are 

evaluated for the possibility of differential flooding within the package. Plastic sheeting on the fuel 

assemblies should be open at both ends to preclude differential flooding. Flooding between the fuel 

pellets and cladding is also considered. Variations in the allowable amount of lightweight packaging 

material and plastic shims inserted in the fuel assemblies can also affect criticality under normal 

conditions of transport. 

• The shielding review evaluates the ability of the package to satisfy the allowed radiation levels during both 

NCT and HAC. 

• The review of operating procedures ensures that instructions are provided so that proper clamps, separators, 

and poisons are selected for the type of fuel assemblies to be shipped and that these items are properly 

installed prior to shipment. The procedures should also address any other restrictions (e.g., limits on number 

of shims) considered in the package evaluation. The review also confirms that instructions are provided for 

the proper closure of the outer shell and for the proper completion of pre-shipment leak test. 

• The review of the acceptance tests and maintenance program verifies that the neutron poisons, if present, are 

subject to appropriate tests to verify their concentration and uniformity. The review also verifies that 

appropriate fabrication and periodic verification leakage tests of the outer shell are performed. 

Figure A10-1. MOX Fresh Fuel Package
C-1 

APPENDIX C: DIFFERENCES BETWEEN THERMAL AND RADIATION 

PROPERTIES OF MOX AND LEU RADIOACTIVE MATERIALS 

The contents considered in this SRP are unirradiated MOX RAM, in the form of powder, pellets, fresh-fuel rods, 

or fresh-reactor fuel assemblies. Unirradiated MOX RAM will also be referred to in this appendix as MOX-fresh 

fuel. This appendix summarizes the relative degree of differences between the thermal and radiation properties of 

the various MOX-RAM contents relative to similar properties for analogous LEU-RAM contents. We will use 

the 3013 StandardC-1 to specify typical grades of plutonium used to make the MOX fresh fuel discussed in this 

appendix. The actual plutonium compositions found in practice may not match these compositions exactly, but 

these grades can be considered typical for the purposes of this appendix. 

The 3013 Standard gives weight percents for various plutonium isotopes in various grades of plutonium. They 

are reproduced in the following table as representative values for typical grades of plutonium that might be used 

to fabricate MOX-fresh fuel. Pure 239Pu has been included to contrast the effect of the other plutonium isotopes. 

Note that in addition to the isotopes identified in Table C-1, plutonium will contain 236Pu and 241Am (from 241Pu 

decay). 

Table C-1. Typical Isotopic Mix in Weight Percent for Various Grades of 

Plutonium as Specified in the 3013 Standard 

Isotope Pure 

239Pu Weapon Grade Fuel 

Grade 

Power 

Grade 

238Pu 0 0.05 0.1 1.0 

239Pu 100 93.50 86.1 62.0a

240Pu 0 6.00 12.0 22.0 

241Pu 0 0.40 1.6 12.0 

242Pu 0 0.05 0.2 3.0 

a 63% reduced to 62% so that sum is 100% 

Initially, it is expected that MOX-fresh fuel will be fabricated using Weapon Grade (WG) plutonium. A more 

mature MOX-fuel program might be expected to fabricate MOX-fresh fuel from previously irradiated WG-MOX 

fuel that may have a composition similar to Fuel Grade (FG) plutonium. Fabricating MOX-fresh fuel from Power 

Grade (PG) plutonium would require a much more mature MOX-plutonium program. 

To compare MOX-fresh fuel with LEU-fresh fuel, we need to choose representative compositions for each fuel 

type. For a reference LEU-fresh fuel, we choose UO2 with 4 wt% 235U and 96 wt% 238U. For the various grades 

of plutonium in MOX-fresh fuel, we choose UO2-PuO2 having 4 wt% 239Pu with the remaining plutonium 

isotopes scaled as required by Table C-1, with the remainder being depleted uranium with 0.2 wt% 235U and 

99.8 wt% 238U. The actual composition of MOX RAM found in practice will not match these compositions, but 

they are appropriate for comparing the effects of MOX RAM using various grades of plutonium. Table C-2 lists 

the weight percents for heavy metal isotopes used in this study. 
 

C-2

Table C-2. Weight Percents for Heavy Metal Isotopes Chosen for 

Comparing MOX with LEU for Various Grades of Plutonium 

Nuclide No Plutoniuma

 Pure 239Pu Weapons 

Gradeb Fuel Gradeb

 Power Gradeb

235U 4.0000 0.1920 0.1914 0.1907 0.1871 

238U 96.0000 95.8080 95.5305 95.1653 93.3613 

238Pu 0.0000 0.0000 0.0021 0.0047 0.0645 

239Pu 0.0000 4.0000 4.0000 4.0000 4.0000 

240Pu 0.0000 0.0000 0.2567 0.5575 1.4194 

241Pu 0.0000 0.0000 0.0171 0.0743 0.7742 

242Pu 0.0000 0.0000 0.0021 0.0093 0.1935 

a No plutonium means low-enriched uranium oxide with 4 wt% 235U and 96 wt% 238U. Note that fresh 

LEU fuel will normally contain traces of 232U, 233U, 234U, and 236U from recycle and natural uranium. 

The quantities of these isotopes normally present in fresh LEU are not significant for the comparisons 

in this appendix. 

b The plutonium mixtures will also contain 236Pu and 241Am. These isotopes can have a significant effect 

on neutron and/or gamma generation rates. 

The nuclide depletion/decay code ORIGEN-ARPC-2 can be used to determine the heat generation rates for 

arbitrary compositions of plutonium with depleted uranium in MOX-fresh fuel. Table C-3a lists the ratio of heat 

generation rates for MOX-fresh fuel relative to LEU-fresh fuel using the composition weight percents for MOXfresh fuel fabricated from the various plutonium grades from Table 2 in ORIGEN-ARP. These are the values 

predicted at the initial time of MOX fuel fabrication when the composition weight percents for the various 

plutonium isotopes are as given in Table C-2. After these nuclides begin to decay, the heat generation rate 

decreases with time, so the initial heat generation rate is also the maximum rate. 

Table C-3a. Ratio of Heat Generation Rate for MOX-Fresh Fuel Composed of 

Various Grades of Plutonium Relative to LEU Fresh Fuel 

Decay 

Time No Plutonium Pure 239Pu Weapons 

Grade Fuel Grade Power 

Grade 

Initial 1 7,300 10,200 13,700 53,900 

Maximum 1 7,300 10,200 13,700 53,900 

The heat generation rate for any MOX-fresh fuel is about four orders of magnitude, or more, greater than that 

from LEU-fresh fuel. Using FG plutonium instead of WG plutonium causes the heat generation rate to increase 

by about another factor of 1.3. Using PG plutonium instead of FG plutonium causes the heat generation rate to 

increase by about another factor of 3.9. For reference, one metric ton of heavy metal of MOX fuel fabricated 

from WG plutonium will generate more than 100 watts of decay heat. 

The heat is generated predominately by alpha decay of the heavy nuclides. The average alpha energy spectrum 

for the plutonium isotopes is greater than that for the uranium isotopes by about 25%. However, the primary 

reason heat generation is greater for plutonium is that its specific activity for alpha decay is four to five orders of 

magnitude larger that that for uranium. Some specific decay parameters for MOX-relevant are shown below, in 

Table C-3b. 
C-3 

Table C-3b. Specific Decay Parameters for MOX-Relevant Nuclides 

Radionuclide Half-Life 

(Years) 

Decay Energy 

(Mev/Event) 

Decay Energy 

(Watt-yr/mole) 

Specific Heat 

Generation Rate 

(Watts/kg) 

233U 1.60E+05 4.909 15,021 5.81E-01 

235U 7.10E+08 4.681 14,333 6.00E-05 

238U 4.50E+09 4.195 12,836 8.00E-06 

238Pu 8,78E+01 5.593 17,113 5.67E+02 

239Pu 2.41E+04 5.244 16,046 1.93E+00 

240Pu 6.54E+03 5.255 16,079 7.10E+00 

241Pu 1.44E+01 0.0205 62.7 1.25E+01 

242Pu 3.76E+05 4.983 15,246 1.16E-01 

241Am 4.32E+02 5.637 17,248 1.15E+02 

The gamma emission code GAMGENC-3 can be used to determine the gamma emission rates for equal weights of 

various nuclides of uranium, plutonium, and americium. Shielding for LEU is not a significant problem as a 

function of decay time. Therefore, studying the gamma emission rate for each nuclide of interest relative to LEU 

gives a measure of how much more difficult shielding problems might be when that nuclide is present than for 

LEU. Table C-4a lists the gamma emission rates at twenty years decay time for equal weights of each nuclide, 

relative to the LEU gamma emission rate at twenty years decay time, for four energy ranges corresponding to 

different minimum gamma energies. Although gamma emission rates are not necessarily maximized at twenty 

years decay time, this decay time was chosen because it gives a better indication of the relation of the various 

nuclide emission rates relative to LEU with time. The maximum gamma energies for each nuclide are below 

3.3 MeV, and sometimes significantly below. The reason the gamma emission ratios are listed for several 

different energy ranges is to provide some indication of the energy distribution for the gammas of each nuclide as 

the minimum gamma energy increases, since shielding becomes more difficult as gamma energy increases. This 

is facilitated by listing the average gamma energy for each nuclide for each energy range in the table. The fact 

that each nuclide has a different average gamma energy for a given energy range is because each has a unique 

gamma energy spectrum. When the average energy for a nuclide is close to the minimum energy for an energy 

range, this indicates that most gammas in that range have energies near to that of the minimum energy. 

The nuclides 236Pu and 232U have very large emission ratios because of the relatively short half-lives and 

2.614 MeV gammas emitted after chain decaying to 208Tl. These gammas are extremely difficult to shield against 

and can usually be tolerated at amounts no greater than about 10-4 weight percent of heavy metal nuclides. The 

nuclides 236U, 241Am, 234U, and 237Np result from radioactive decay of 240Pu, 241Pu, 238Pu, and 241Am, respectively. 

The nuclide 233U is usually present in trace quantities. 

In Table C-4a for the minimum gamma energies corresponding to 0.041 and 0.183 MeV, most of the nuclides 

have emission ratios greater than 1.00. The nuclides 238Pu, 240Pu, 241Pu, 242Pu, 234U, 236U, and 241Am have a 

majority of gammas in the energy range between roughly 0.04 and 0.12 MeV, because their average energies for 

the first energy range are close to the minimum energy of 0.041 MeV. However, except for 235U, all nuclides 

have average energies greater than 0.28 MeV for the second energy range. Therefore, these nuclides have 

considerable gammas with energies that will require specific gamma shielding if present in sufficient quantities. 

This is reinforced by the emission ratios and average energies for the energy range with minimum gamma energy 

of 0.498 MeV, particularly for the plutonium isotopes. For the energy range with minimum gamma energy of 

1.000 MeV, only 236Pu (except for trace nuclides) has high emission rates of very high-energy gammas that may 

require substantial shielding if it is present in a significant quantity. 
 

C-4

Table C-4a. Gamma Emission Rates Relative to the LEU Gamma Emission Rate and 

Average Gamma Energies for Equal Weights of Some Nuclides of 

Uranium, Plutonium, Neptunium and Americium at 

20 Years Decay Time 

 Gamma Energies 

≥0.041 MeV

Gamma Energies 

≥0.183 MeV 

Gamma Energies 

≥0.498 MeV 

Gamma Energies 

≥1.000 MeV 

 Emission Average Emission Average Emission Average Emission Average

Select Relative to Energy Relative to Energy Relative to Energy Relative to Energy

Nuclides LEU (MeV) LEU (MeV) LEU (MeV) LEU (MeV) 

236Pu 2.73E+08 0.7929 4.46E+08 0.9927 2.90E+09 1.4300 2.26E+09 2.5212 

238Pu 5.33E+04 0.0624 1.37E+02 0.7497 1.31E+03 0.7906 7.71E+01 1.1753 

239Pu 2.89E+02 0.1173 7.08E+01 0.3883 8.39E+00 0.6968 1.06E-02 1.1750 

240Pu 9.05E+02 0.0600 1.66E+00 0.3941 5.97E+00 0.6831 4.05E-09 2.1790 

241Pu 5.71E+06 0.0544 4.74E+03 0.2805 2.64E+03 0.6771 8.49E-07 1.4577 

242Pu 1.32E+01 0.0613 3.94E-06 0.9783 3.75E-05 1.0389 3.75E-05 1.2507 

232U 2.73E+08 0.7930 4.46E+08 0.9927 2.91E+09 1.4300 2.26E+09 2.5212 

233U 2.62E+02 0.1801 1.88E+02 0.3549 8.32E+01 1.3620 1.30E+02 1.4607 

234U 7.58E+01 0.0789 2.21E-01 0.7588 1.39E+00 1.0282 1.05E+00 1.5553 

235U 1.75E+01 0.1901 2.24E+01 0.2402 7.27E-03 0.7422 1.49E-04 1.1750 

236U 4.76E-01 0.0723 1.52E-06 0.8876 1.05E-05 1.1865 5.49E-06 2.2074 

238U 3.12E-01 0.2289 1.09E-01 0.9783 1.04E+00 1.0389 1.04E+00 1.2507 

LEU 1.00E+00 0.2017 1.00E+00 0.3177 1.00E+00 1.0388 1.00E+00 1.2507 

237Np 6.06E+03 0.2096 5.94E+03 0.3374 2.63E-04 1.3575 4.09E-04 1.4597 

241Am 9.09E+06 0.0543 1.85E+03 0.3905 4.21E+03 0.6771 4.23E-06 1.4586 

ORIGEN-ARP also gives the gamma emission rates for arbitrary compositions of plutonium with depleted 

uranium in MOX-fresh fuel. Table C-4b lists the ratio of gamma emission rates for MOX-fresh fuel relative to 

LEU-fresh fuel using the composition weight percents for MOX-fresh fuel fabricated from the various plutonium 

grades from Table 2 in ORIGEN-ARP. Table C-4b lists rates for both initial time and maximum rates after some 

decay time. The decay time at maximum gamma emission rates depends on the plutonium grade in question. The 

gamma emission rates include only gammas with energies equal to or greater than 100 keV. The assumption is 

that gammas with energies less than 100 keV will be absorbed by the normal packaging materials required to 

transport MOX-fresh fuel contents, specifically the strong 59.5 keV gammas coming from any 241Am produced 

through decay of 241Pu. Note that MOX containing 236Pu at concentrations greater than about 10-4 weight percent 

of total plutonium mass or significant 241Am in-growth may have larger gamma emission rates than are shown in 

Table C-4b. 
C-5 

Table C-4b. Ratio of Gamma Emission Rate for Gamma Energies Exceeding 100 keV for MOX-Fresh 

Fuel Composed of Various Grades of Plutonium Relative to LEU-Fresh Fuel 

Decay 

Time 

No 

Plutonium 

Pure 

239Pu 

Weapon 

Grade 

Fuel 

Grade 

Power 

Grade 

Initial 1.0 6.1 6.1 6.9 15.4 

Maximum 1.0 6.1 6.1 7.2 83.5 

The gamma emission rates for MOX-fresh fuel from both WG and FG plutonium are less than an order of 

magnitude greater than those for LEU-fresh fuel. The gamma emission rates for MOX-fresh fuel from PG 

plutonium can be up to about two orders of magnitude greater than those for LEU-fresh fuel depending on the 

time since MOX-fuel fabrication. 

The neutron emission code SOURCESC-4 can be used to determine the neutron emission rates for spontaneous 

fission and alpha-induced neutrons for equal weights of various nuclides of uranium, plutonium, and americium. 

Table C-5a lists the neutron emission rates for spontaneous fission and alpha-induced neutrons from 17O and 18O, 

for equal weights of nuclides at the initial MOX fuel fabrication time relative to the LEU neutron emission rate. 

Also listed in the table is the average neutron energy for each nuclide and each neutron emission process. 

Table C-5a. Neutron Emission Rates Relative to the LEU Neutron Emission Rate 

and Average Gamma Energies for Equal Weights of Some Nuclides of Uranium, 

Plutonium and Americium for (α,n) with 17O and 18O, Spontaneous Fission (SF), 

and the Sum of All Three (Total) Neutron Emission Processes 

 17O (α,n) Average 18O (α,n) Average SF Average Total 

Select Relative to Energy Relative to Energy Relative to Energy Relative to

Nuclides LEU (MeV) LEU (MeV) LEU (MeV) LEU 

238Pu 1.03E+08 2.52 1.27E+08 2.37 1.98E+05 2.02 1.24E+06 

239Pu 2.86E+05 2.44 3.62E+05 2.25 1.67E+00 2.07 2.97E+03 

240Pu 1.06E+06 2.44 1.33E+06 2.25 7.84E+04 1.93 8.87E+04 

241Pu 9.42E+03 2.39 1.23E+04 2.19 3.77E+00 2.00 1.04E+02 

242Pu 1.49E+04 2.38 1.94E+04 2.19 1.31E+05 1.96 1.30E+05 

233U 3.50E+04 2.37 4.53E+04 2.17 6.23E-02 2.02 3.72E+02 

235U 5.96E+00 2.27 6.67E+00 2.07 2.29E-02 1.89 7.80E-02 

236U 1.87E+02 2.29 2.23E+02 2.09 4.19E-01 1.83 2.26E+00 

238U 7.93E-01 2.20 7.64E-01 1.97 1.04E+00 1.69 1.04E+00 

LEU 1.00E+00 2.22 1.00E+00 2.00 1.00E+00 1.74 1.00E+00 

241Am 2.05E+07 2.51 2.53E+07 2.36 9.03E+01 2.15 2.09E+05 

On an equal weight basis, 238Pu, 240Pu, 242Pu and 241Am are overwhelming the largest source for neutron emission 

for the nuclides listed in Table C-5a. Most nuclides listed in the table have neutron emission rates greater than 

LEU by one or more orders of magnitude. Table C-5a also shows that neutron emission from uranium isotopes is 

insignificant relative to that from plutonium isotopes on an equal weight basis. The average neutron energies 

listed in Table C-5a are between about 1.7 MeV and 2.5 MeV. This means that the spectral energy distribution 

for neutrons plays a much smaller role than does the spectral energy distribution for gammas. 
 

C-6

The neutron emission code SOURCES can also be used to determine the neutron emission rates for spontaneous 

fission and alpha-induced neutrons for arbitrary compositions of plutonium isotopes with depleted uranium in 

MOX-fresh fuel. Table C-5b lists the ratio of neutron emission rates for MOX-fresh fuel relative to LEU-fresh 

fuel using the composition weight percents for MOX-fresh fuel fabricated from the various plutonium grades 

from Table C-2 in SOURCES. Note that MOX-fresh fuel with significant 241Am in-growth (from 241Pu decay) 

can have significantly larger relative neutron emission rates as is shown in the last line of Table C-5b. 

Table C-5b. Ratio of Neutron Emission Rate for MOX-Fresh Fuel Composed of 

Various Grades of Plutonium Relative to LEU-Fresh Fuel 

Nuclide 

Composition 

No 

Plutonium 

Pure 

239Pu 

Weapons 

Grade 

Fuel 

Grade 

Power 

Grade 

Fresh Fuel 1 24 243 506 1,686 

241Pu replaced by 

241Am 1 24 250 536 1,995 

Replacing 4 wt% 235U with 4 wt% 239Pu increases the neutron emission rate by a factor of about 24. Using WG 

plutonium instead of pure 239Pu causes the neutron emission rate to increase by about another order of magnitude. 

Using FG plutonium instead of WG plutonium causes the neutron emission rate to increase by about another 

factor of two. Using PG plutonium instead of FG plutonium causes the neutron emission rate to increase by about 

another factor of three. 

Plutonium-241 decays to 241Am with a half-life of 14.35 years. Americium-241 is a stronger neutron source, so to 

get a bounding value for the expected increase in neutron emission rate, when 241Pu decays to 241Am, all 241Pu is 

replaced with 241Am, and the neutron emission rate is recalculated for each of these new artificial grades of 

plutonium.∗

 The last row of Table C-5b lists the values obtained. This approach gives an indication of what 

decay time can do to neutron emission rates. The effect on neutron emission rate of 241Pu decay to 241Am is 

expected to be rather small except for MOX-fresh fuel fabricated from PG plutonium, where it could increase by a 

factor of about 20%. 

The uncertainties in the rates of heat generation, or gamma emission, or neutron emission from analyses 

performed using radiation transport codes and cross section sets, such as those employed above, for MOX-RAM 

packages should be comparable to those performed for packages containing LEU RAM for the purposes required 

for thermal and shielding reviews. 

In summary, heat generation and neutron emission rates increase significantly when MOX RAM replaces LEU 

RAM. The alpha-energy spectrum responsible for most heat generation is somewhat different for MOX RAM 

and LEU RAM, but that is not significant in relation to the magnitude of the heat generation rates between them. 

The neutron energy spectrum from MOX RAM and LEU RAM are also somewhat different, but again this is not 

significant in relation to the magnitude of the neutron emission rates between them. The gamma emission rate 

increases between MOX RAM and LEU RAM are not as important as long as 236Pu is less than about 10-4 weight 

percent of the heavy metal present in MOX RAM. Otherwise, the strong 2.614 MeV gamma from the chain 

decay of 236Pu to 208Tl becomes an important source of external package gamma dose that is extremely hard to 

shield against. However, the gamma energy spectrum from MOX RAM and LEU RAM can be quite different 

depending on the nuclides present and this can be significant from a shielding point of view. 

 

∗

 Replacing 241Pu with 241Am is bounding for a neutron shielding evaluation, but not for a criticality evaluation. 
C-7 

References 

C-1. U.S. Department of Energy, “Stabilization, Packaging, and Storage of Plutonium-Bearing Materials,” U.S. 

DOE Standard DOE-STD-3013-2000, Washington D.C., September 2000. 

C-2. S. M. Bowman and L. C. Leal, “ORIGEN-ARP: Automatic Rapid Process for Spent Fuel Depletion, Decay, 

and Source Term Analysis,” part of SCALE: A Modular Code System for Performing Standardized 

Computer Analyses for Licensing Evaluation, Vol. I, NUREG/CR-0200, Revision 6, 

(ORNL/NUREG/CSD-2/VI/R6), Oak Ridge National Laboratory, March 2000. 

C-3. T. B. Gosnell, “Automated Calculation of Photon Source Emission From Arbitrary Mixtures of Naturally 

Radioactive Nuclides,” Nuclear Instruments and Methods in Physics Research, A299, 682–686 (1990). 

C-4. W. B. Wilson, et al., “SOURCES 4A: A Code for Calculating (α,n), Spontaneous Fission, and Delayed 

Neutron Sources and Spectra,” LA-13639-MS, Los Alamos National Laboratory, Los Alamos, NM, 

September 1999. 
 

C-8

[THIS PAGE BLANK] 
D-1

APPENDIX D: BENCHMARK CONSIDERATIONS FOR 

MOX RADIOACTIVE MATERIALS 

D.1 Experimental Benchmarks 

Substantial guidance on how to select an appropriate set of criticality benchmark experiments for LEU fissile 

systems is given in NUREG/CR-5661 and in NUREG/CR-6361.D-1, D-2 Considerably fewer benchmark 

experiments exist for MOX than for LEU, however. As a consequence, the guidance provided in 

NUREG/CR-5661 and/or in NUREG/CR-6361 cannot be applied directly to the evaluation of MOX fissile 

systems. The benchmarks needed for the criticality analyses of MOX packages are in the thermal energy range. 

This condition results because, for essentially all types of MOX, the most reactive configuration is a flooded 

containment. 

As an alternative, the International Handbook of Evaluated Criticality Safety Benchmark ExperimentsD-3

(IHECSBE) has 11 evaluated thermal-energy studies involving MOX fuel pins in various lattice experiments, and 

5 evaluated thermal-energy studies involving MOX liquids in tank experiments. These can be divided into 18 sets 

of experiments involving different fissile oxide compositions and configurations in lattices, and 13 sets of 

experiments involving different liquid fissile nitrate compositions and configurations in tanks. The total number 

of essentially different experiments is 131. Other benchmark experiments are available throughout the world, but 

are not as readily available, and the vast majority have not been rigorously evaluated in the manner of those found 

in the IHECSBE, and are consequently of limited use for benchmark criticality analyses for MOX packages. 

More evaluated MOX thermal benchmarks are expected in future editions of the IHECSBE. 

The 18 sets of experiments involving fissile oxides in lattices and 13 sets of experiments involving fissile nitrate 

liquids in tanks have been organized and shown in Tables D-1 through D-5. The various tables are separated on 

two features. The first is between lattice and tank experiments, and the second is on weight percent of plutonium 

to total plutonium plus uranium, Pu/(Pu+U). Table D-1 has lattice experiments with Pu/(Pu+U) to 5%. Table D2 has lattice experiments with Pu/(Pu+U) from 5% to 15%. Table D-3 has lattice experiments with Pu/(Pu+U) 

greater than 15%. Table D-4 has tank experiments with Pu/(Pu+U) to 31% (there are no experiments with 

Pu/(Pu+U) less than 22%). Table D-5 has tank experiments with Pu/(Pu+U) greater than 31%. Lists of 

meaningful, experimental characteristics are recorded for each set of experiments together with characteristics of 

their corresponding computational evaluations. 

Experimental plutonium benchmarks should also be taken into account as part of the initial set of benchmark 

experiments to be considered for a MOX package application. About four times as many thermal-plutoniumtank-liquid benchmarks exist in the IHECSBE as thermal-MOX-tank-liquid benchmarks. However, fewer 

thermal-plutonium-lattice benchmarks exist in the IHECSBE as thermal-MOX-lattice benchmarks. 

D.2 Summary of Bias and Uncertainty Evaluation 

There are two measures to determine the accuracy of an experiment and its associated calculation. The first 

measure is the effective bias (Eff-Bias) between calculation and benchmark experiment. The multiplication 

coefficient for a fissile system is designated as keff. Designate the calculated keff for the benchmark experiment as 

kcalc and the benchmark experimental keff as kexp. If the calculational bias, β, is defined as β = kcalc - kexp, then a 

quantity ∆k∗

 can be defined as 

 

∗

 As defined in Equation D-1, ∆k is always less than or equal to zero, and is consistent with the bias, β

_

, defined in 

Reference D-1. Typically, a calculational method is termed to have a negative bias if it under-predicts the critical 

condition. 
 

D-2

 ∆k = 

⎝

⎜

⎜

⎛

⎠

⎟

⎟ β if kcalc ≤ kexp ⎞

0 if kcalc > kexp

 . (D-1) 

For a given experimental benchmark set, ∆kmax is chosen as the largest absolute value of the ∆k given by 

Equation D-1 for all experiments in the set. The 95% confidence limit of kcalc is kcalc plus twice the calculated 

standard deviation, which is designated by 2σ. The Eff-Bias value is then given by 

 Eff-Bias = ∆kmax - 2σ. (D-2) 

Eff-Bias, as defined here, is always less than zero. If kcalc is greater than kexp for all experiments in a set, the EffBias value is just the negative of twice the calculated standard deviation. 

The second measure is the total experimental uncertainty (Exp-Uncer) that was determined by the evaluator after 

assessing all sources of uncertainty for the experiments in a set.∗

 A worst-case difference between kcalc and kexp

can be assigned as the difference of the total experimental uncertainty and the effective bias 

(Exp-Uncer – Eff-Bias) for the experimental set in question. This worst-case difference (WCD), as defined here, 

is always greater than zero. It represents the upper limit of the inherent uncertainties in the ability of the 

computer code, together with the cross-section set used, to accurately determine the keff of a critical benchmark 

experiment. Therefore, a bounding multiplication coefficient, ksafe, at the 95% confidence limit, can be chosen to 

be equal to 0.95 minus WCD, where an administrative margin of safety of 0.05 has been included.+

 

Values for the variable WCD for each experimental set vary between 0.0071 to 0.0192 (0.71% to 1.92%), 0.0043 

to 0.0328 (0.43% to 3.28%), 0.0023 to 0.0138 (0.23% to 1.38%), 0.0044 to 0.0180 (0.44% to 1.80%), and 0.0044 

to 0.0150 (0.44% to 1.50%), for the experimental sets in Tables D-1, D-2, D-3, D-4, and D-5, respectively. No 

particular correlation seems to exist between WCD and the lattice configuration or pitch. Neither does there seem 

to be a correlation with plutonium composition type (Pu type). The plutonium composition types are given in 

Table 1 in the text, and are designated as weapons grade (WG), fuel grade (FG), and power grade (PG). 

The maximum value for WCD found in the five tables is 0.0328 or 3.28% in keff. How accurately a criticality 

computer code can predict the critical value for a criticality experiment depends on the methodology employed by 

the code and the cross-section set used, together with the detail to which the experimental system is modeled in 

the computer. In addition, the basic experimental uncertainty limits the ultimate prediction accuracy possible. Of 

particular importance is the cross-section set. Values for WCD in the five tables that are significantly less than 

0.0100 are due to the fact that kcalc is greater than kexp. Therefore, the value for Eff-Bias, in that case, is just the 

negative of twice the calculated standard deviation, which is approximately 0.0020. Cross-section sets prior to 

ENDF/B-V over predict plutonium reactivity, and this represents some of the reason for the over-prediction for 

kcalc for these experiments. Values for ksafe are not expected to be much above 0.93, except when it can be 

demonstrated that the criticality code and cross-section set overestimates the reactivity of the MOX contents. 

Analyzing an acceptable number of MOX benchmarks is the preferred way to obtain a bias value for the MOX 

contents of a package. With the relatively limited number of MOX-critical experiments available for use in 

validation exercises, it is important to determine that the application of interest to the reviewer fits within the area 

of applicability for the set of critical benchmark experiments selected for validation. Guidance on how to select 

an appropriate set of benchmark experiments for LEU fissile system is given in NUREG/CR-5661 and in 

 

∗

 The evaluator included sources of experimental bias or error in each kexp. This does not represent an uncertainty and so 

is not included in the value for total experimental uncertainty. +

 If the benchmarks are applied to a package application where there is a lack of experimental data, the 0.05 administrative 

margin may not be sufficient, and the reviewer needs to be aware of this issue. In reality, the 0.05 margin should be 

sufficient, but there needs to be an assessment of the adequacy of the 0.05 to establish the basis. Guidance for deciding 

on an acceptable choice for the administrative margin is given in NUREG/CR-5561. See also NUREG/CR-6361. 
D-3

NUREG/CR-6361. An important advancement using computational methodology to select an appropriate set of 

benchmark experiments for a fissile package application is currently being developed for SCALE.D-4, D-5, D-6 

A set of sensitivity and uncertainty analysis tools are being developed for version 5 of SCALE that gives a 

measure of the similarity of the reactivity of a package application to that of an experimental benchmark. 

Sensitivity coefficients for both systems are computed and give the sensitivity of each system’s keff to the cross 

section data. These sensitivity coefficients are determined for each energy group in the cross section library 

chosen in the analysis, as well as the sum over all energy groups. Two integral parameters for the combined 

systems are produced from the sensitivity data to determine system-to-system similarities. The first parameter 

can be used as a gauge of system similarity to sensitivity-only. The second parameter can be used as a measure of 

the similarity of the systems in terms of uncertainty, not just sensitivity. The pair of integral parameter values is 

determined for every potential benchmark experiment with the package application of interest. When two 

systems produce a value of 0.8 for either integral parameter, or both, this indicates the keff response is similar 

enough that one system serves well to validate the criticality safety parameters for the other system. The 

benchmark experiments chosen for complete validation are those with high integral parameter values.D-4, D-5, D-6 

New parameters can also be constructed from the components of the integral parameters and can be used to 

explore the sensitivity of specific nuclide reactions of benchmark experiments with the package application of 

interest. For example, if low integral parameter values are found for an application with all benchmark 

experiments chosen for validation, the new parameters could serve to identify which nuclides would require 

additional experimental benchmark data for complete validation. Also, in the validation of shipping casks for 

commercial fuel, numerous benchmark experiments might serve to validate the fission reactions, and thus high 

integral parameter values would be found. However, the new parameters could be used to find benchmarks to 

ensure that any poison materials in the cask are also well validated by the benchmarks. Once these sensitivity and 

uncertainty analysis tools are released with version 5 of SCALE, the criticality safety analyst will have a powerful 

set of tools to perform detailed quantitative analyses to determine the applicability of benchmark experiments to 

help design package applications under consideration.D-4, D-5, D-6 
D-4

Table D-1. Important Characteristics of Lattice Experiments with Weight Percent of Pu/(Pu+U) to 

(from IHECSBE) 

Designation for experimentsa MCT-009 MCT-002 MCT-002 MCT-006 MCT-007 MCT-008 MCTFacility where experiments conducted Hanford Hanford Hanford Hanford Hanford Hanford ToComputer codes used in evaluationsb MCNP/KENO MCNP MCNP MCNP/KENO MCNP/KENO MCNP/KENO MCNPCross-section sets used in evaluationsc ENDF/B-V/IV ENDF/B-V ENDF/B-V ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV JENDCross-section typed cont/27grp cont cont cont/27grp cont/27grp cont/27grp cont/1Fuel compounde oxide oxide oxide oxide oxide oxide oxFuel compound form solid solid solid solid solid solid soDensity of fuelf 86.7% 86.7% 86.7% 86.7% 86.7% 86.7% 55Organization of fuelg pins pins pins pins pins pins piCladding used for fuelh Zirc-2 Zirc-2 Zirc-2 Zirc-2 Zirc-2 Zirc-2 ZirPu/(Pu+U) atom percent 1.51% 1.80% 1.80% 1.80% 2.01% 2.01% 3U-235 atom percent 0.16% 0.71% 0.71% 0.71% 0.72% 0.72% 0U-238 atom percent 99.84% 99.29% 99.29% 99.29% 99.28% 99.28% 99Pu-238 atom percent - 0.01% 0.01% 0.01% - - 0Pu-239 atom percent 91.41% 91.84% 91.84% 91.84% 81.11% 71.76% 68Pu-240 atom percent 7.83% 7.76% 7.76% 7.76% 16.54% 23.50% 22Pu-241 atom percent 0.73% 0.37% 0.37% 0.37% 2.15% 4.08% 7Pu-242 atom percent 0.03% 0.03% 0.03% 0.03% 0.20% 0.66% 2Plutonium type as given in Table 1 WG WG WG WG FG PG PShape of latticei cylinder rectangle rectangle cylinder cylinder cylinder rectaPitch of lattice triangle square square triangle triangle triangle squNumber of experiments in each set 6 3 3 6 5 6 4 Fissile moderator usedj H2O H2O B-H2O H2O H2O H2O H2Reflector used H2O H2O B-H2O H2O H2O H2O H2Maximum effective bias of experiments in set (Eff-Bias) -0.0112 -0.0052 -0.0026 -0.0089 -0.0040 -0.0068 -0.0Maximum uncertainty of experiments in set (Exp-Uncer) 0.0080 0.0059 0.0045 0.0054 0.0061 0.0065 0.0Exp-Uncer minus Eff-Bias (WCD) 0.0192 0.0111 0.0071 0.0143 0.0101 0.0133 0.0

a. Definition of acronyms is MCT = MIX-COMP-THERM. b. Codes MCNPD-7 and KENOD-8. 

c. ENDF/B-V/IV means cross-section set ENDF/B-V for MCNP and cross-section set ENDF/B-IV for KENO. J ENDL-3.2 is cross-section set for both MCNP and KENO. 

d. Cross-section type is either continuous cross sections (cont.) or group cross sections (27grp, 137grp). e. Heavy metal is as an oxide. f. MOX density given as percent of theoretical density taken as 11.00 g/cm3. 

g. Pins means organization of MOX is as pellets in fuel pins. h. Zirc-2 means zircaloy-2 cladding. i. Cylinder means shape of lattice is a cylinder. Rectangle means shape of lattice is a rectangle. j. B-H2O means borated water as moderator or reflector. 
D-5

Table D-2. Important Characteristics of Lattice Experiments with Weight Percent of Pu/(Pu+U) from 5% 

(from IHECSBE) 

Designation for experimentsa MCT-003 MCT-003 MCT-012 MCT-012 MCT-012 MCFacility where experiments conducted WREC WREC Hanford Hanford Hanford HaComputer codes used in evaluationsb MCNP MCNP MCNP/KENO MCNP/KENO MCNP/KENO MCNPCross-section sets used in evaluationsc ENDF/B-V ENDF/B-V ENDF/B-V ENDF/B-V ENDF/B-V ENDCross-section typed cont cont cont/238grp cont/238grp cont/238grp cont/Fuel compounde oxide oxide oxide-poly oxide-poly oxide-poly oxidFuel compound form solid solid solid solid solid soDensity of fuelf 94% 94% N/A N/A N/A NOrganization of fuelg pins pins cubes, slabs cubes, slabs cubes, slabs cubeCladding used for fuelh Zirc-4 Zirc-4 plastic 471 plastic 471 plastic 471 plastPu/(Pu+U) atom percent 6.63% 6.63% 7.60% 7.89% 14.62% 14.6U-235 atom percent 0.71% 0.71% 0.15% 0.15% 0.15% 0.1U-238 atom percent 99.29% 99.29% 99.85% 99.85% 99.85% 99.8Pu-238 atom percent - - 0.59% - - - Pu-239 atom percent 90.65% 90.65% 67.97% 91.25% 91.42% 91.4Pu-240 atom percent 8.55% 8.55% 22.95% 8.12% 7.97% 7.9Pu-241 atom percent 0.76% 0.76% 5.57% 0.58% 0.57% 0.5Pu-242 atom percent 0.04% 0.04% 2.92% 0.05% 0.04% 0.0Plutonium type as given in Table 1 WG-FG WG-FG PG WG WG WShape of latticei rectangle rectangle 3D cube 3D cube 3D cube 3D Pitch of lattice square square square square square sqNumber of experiments in each set 5 1 6 7 6 Fissile moderator usedj H2O B-H2O polystyrene polystyrene polystyrene polyReflector used H2O B-H2O plexiglas plexiglas plexiglas nMaximum effective bias of experiments in set (Eff-Bias) -0.0063 -0.0030 -0.0270 -0.0016 -0.0016 -0.0Maximum uncertainty of experiments in set (Exp-Uncer) 0.0071 0.0052 0.0058 0.0036 0.0027 0.0Exp-Uncer minus Eff-Bias (WCD) 0.0134 0.0082 0.0328 0.0052 0.0043 0.0

a. Definition of acronyms is MCT = MIX-COMP-THERM. b. Codes MCNPD-7 and KENOD-8. 

c. ENDF/B-V is cross-section set for MCNP and KENO. d. Cross-section type is either continuous cross sections (cont.) or group cross sections (238grp). e. Heavy metal is as an oxide. Oxide-poly means mixture of MOX particles and polystyrene pressed into cubes and slabs. f. MOX density given as percent of theoretical density taken as 11.00 g/cm3. 

g. Pins means organization of MOX is as pellets in fuel pins. Cubes, slabs means organization of MOX-polystyrene is as cubes and slabs. h. Zirc-4 means zircaloy-4 cladding. Plastic 471 means cladding is six mil plastic tape MM&M (3M) #471. i. Rectangle means shape of lattice is a rectangle. 3D cube means cubes and slabs stacked into the shape of a 3D-rectangular cube. j. B-H2O means borated water as moderator or reflector. 
D-6

Table D-3. Important Characteristics of Lattice Experiments with Weight Percent of Pu/(Pu+U) Greater t

(from IHECSBE) 

Designation for experimentsa MCT-001 MCT-011 MCT-012 MCT-012 Facility where experiments conducted Hanford Valduc Hanford Hanford Computer codes used in evaluationsb MONK MORET MCNP/KENO MCNP/KENOCross-section sets used in evaluationsc UKNDL JEF2.2 ENDF/B-V ENDF/B-V Cross-section typed cont 172gp cont/238grp cont/238grp Fuel compounde oxide oxide oxide-poly oxide-poly Fuel compound form solid solid solid solid Density of fuelf 89.4% 94.2% N/A N/A Organization of fuelg pins pins cubes, slabs cubes, slabs Cladding used for fuelh 316 SS Z3CND18.12 SS plastic 471 plastic 471 Pu/(Pu+U) atom percent 19.70% 25.80% 30.00% 30.00% U-235 atom percent 0.71% 60.15% 0.15% 0.15% U-238 atom percent 99.29% 39.85% 99.85% 99.85% Pu-238 atom percent 0.15% - - - Pu-239 atom percent 85.54% 89.00% 91.22% 91.22% Pu-240 atom percent 11.46% 9.72% 8.13% 8.13% Pu-241 atom percent 2.50% 1.21% 0.61% 0.61% Pu-242 atom percent 0.35% 0.07% 0.04% 0.04% Plutonium type as given in Table 1 FG WG-FG WG WG Shape of latticei rectangle cylinder 3D cube 3D cube Pitch of lattice square triangle square square Number of experiments in each set 4 6 8 3 Fissile moderator used H2O H2O polystyrene polystyrene Reflector used H2O H2O plexiglas none Maximum effective bias of experiments in set (Eff-Bias) -0.0103 -0.0006 -0.0018 -0.0086 Maximum uncertainty of experiments in set (Exp-Uncer) 0.0025 0.0017 0.0049 0.0052 Exp-Uncer minus Eff-Bias (WCD) 0.0128 0.0023 0.0067 0.0138 

a. Definition of acronyms is MCT = MIX-COMP-THERM. b. Codes MCNPD-7 and KENOD-8 MONK,D-9 and MORET.D-10

c. ENDF/B-V is cross-section set for MCNP and KENO. UKNDL is cross-section set for MONK. JEF2.2 is cross-section set for M

d. Cross-section type is either continuous cross sections (cont.) or group cross sections (172grp, 238grp) e. Heavy metal is as an oxide. Oxide-poly means mixture of MOX particles and polystyrene pressed into cubes and slabs. 

f. MOX density given as percent of theoretical density taken as 11.00 g/cm3. 

g. Pins means organization of MOX is as pellets in fuel pins. Cubes, slabs means organization of MOX-polystyrene is as cubes and 

h. SS means stainless steel cladding. Plastic 471 means cladding is six mil plastic tape MM&M (3M) #471. 

i. Cylinder means shape of lattice is a cylinder. Rectangle means shape of lattice is a rectangle cube. 3D cube means cubes and slab

rectangular cube. 
D-7

Table D-4. Important Characteristics of Tank Experiments with Weight Percent of Pu/(Pu+U) to 31

(from IHECSBE) 

Designation for experimentsa MST-001 MST-001 MST-001 MST-001 MST-002 Facility where experiments conducted Hanford Hanford Hanford Hanford Hanford Computer codes used in evaluationsb MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO Cross-section sets used in evaluationsc ENDF/B-V /IV ENDF/B-V /IV ENDF/B-V /IV ENDF/B-V /IV ENDF/B-V /IV Cross-section typed cont/27grp cont/27grp cont/27grp cont/27grp cont/27grp cFuel compounde nitrate nitrate nitrate nitrate nitrate Fuel compound form liquid liquid liquid liquid liquid Density of fuelf 1.31-1.68 1.31-1.68 1.31-1.48 1.70 1.09 Pu/(Pu+U) atom percent 22% 22% 22% 22% 23% U-235 atom percent 0.70% 0.70% 0.70% 0.70% 0.70% U-238 atom percent 99.30% 99.30% 99.30% 99.30% 99.30% Pu-238 atom percent 0.03% 0.03% 0.03% 0.03% 0.03% Pu-239 atom percent 91.12% 91.12% 91.12% 91.12% 91.12% Pu-240 atom percent 8.34% 8.34% 8.34% 8.34% 8.31% Pu-241 atom percent 0.42% 0.42% 0.42% 0.42% 0.45% Pu-242 atom percent 0.09% 0.09% 0.09% 0.09% 0.09% Plutonium type as given in Table 1 WG WG WG WG WG Tank fissile liquid is ing N/A cylinder cylinder cylinder cylinder Auxiliary tank additional fissile liquid is inh annular annular annular N/A N/A Number of experiments in each set 2 5 2 1 1 Fissile moderator usedi soln H2O soln H2O soln H2O soln H2O soln H2O Reflector usedj B4C-concrete B4C-concrete poly-Cd cover none H2O HMaximum effective bias of experiments in set (Eff-Bias) -0.0101 -0.0164 -0.0028 -0.0068 -0.0020 Maximum uncertainty of experiments in set (Exp-Uncer) 0.0016 0.0016 0.0016 0.0016 0.0024 Exp-Uncer minus Eff-Bias (WCD) 0.0117 0.0180 0.0044 0.0084 0.0044 

a. Definition of acronyms is MST = MIX-SOL-THERM. b. Codes MCNP,D-7 KENO,D-8 and MONK.D-9

c. ENDF/B-V/IV means cross-section set ENDF/B-V for MCNP and ENDF/B-IV for KENO. UKNDL is cross-section set for MONK. d. Cross-section type is either continuous cross sections (cont.) or group cross sections (27grp). e. Heavy metal is as a nitrate dissolved in dilute nitric acid solution. 

f. Solution density is in g/ml. g. Containers for fissile solution are cylinders or slabs. h. Annular tank surrounding central cylindrical tank or just an annular tank. i. Soln H2O means the moderator is the fissile nitrate solution. 

j B4C-concrete means borated concrete. Poly-Cd cover means polyethylene reflector coated with Cd. 
D-8

Table D-5. Important Characteristics of Tank Experiments with Weight Percent of Pu/(Pu+U) Greater th

(from IHECSBE) 

Designation for experimentsa MST-004 MST-004 MST-004 MST-005 MST-005 MST-0Facility where experiments conducted Hanford Hanford Hanford Hanford Hanford HanfoComputer codes used in evaluationsb MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KCross-section sets used in evaluationsc ENDF/B-V /IV ENDF/B-V /IV ENDF/B-V /IV ENDF/B-V /IV ENDF/B-V /IV ENDF/B-Cross-section typed cont/27grp cont/27grp cont/27grp cont/27grp cont/27grp cont/27Fuel compounde nitrate nitrate nitrate nitrate nitrate nitratFuel compound form liquid liquid liquid liquid liquid liquid Density of fuelf 1.17-1.67 1.17-1.67 1.17-1.67 1.17-1.67 1.17-1.67 1.05 Pu/(Pu+U) atom percent 40% 40% 40% 40% 40% 52% U-235 atom percent 0.56% 0.56% 0.56% 0.56% 0.56% 0.70% U-238 atom percent 99.44% 99.44% 99.44% 99.44% 99.44% 99.30Pu-238 atom percent 0.03% 0.03% 0.03% 0.03% 0.03% 0.03% Pu-239 atom percent 91.12% 91.12% 91.12% 91.12% 91.12% 91.12Pu-240 atom percent 8.34% 8.34% 8.34% 8.34% 8.34% 8.34% Pu-241 atom percent 0.42% 0.42% 0.42% 0.42% 0.42% 0.42% Pu-242 atom percent 0.09% 0.09% 0.09% 0.09% 0.09% 0.09% Plutonium type as given in Table 1 WG WG WG WG WG WG Tank fissile liquid is ing cylinder cylinder cylinder slab slab cylindAuxiliary tank additional fissile liquid is inh N/A N/A N/A N/A N/A N/A Number of experiments in each set 3 3 3 3 4 2 Fissile moderator usedi soln H2O soln H2O soln H2O soln H2O soln H2O soln HReflector usedj none H2O concrete none H2O H2O Maximum effective bias of experiments in set (Eff-Bias) -0.0060 -0.0048 -0.0024 -0.0114 -0.0026 -0.002Maximum uncertainty of experiments in set (Exp-Uncer) 0.0033 0.0033 0.0078 0.0036 0.0037 0.002Exp-Uncer minus Eff-Bias (WCD) 0.0093 0.0081 0.0102 0.0150 0.0063 0.004

a. Definition of acronyms is MST = MIX-SOL-THERM. b. Codes MCNPD-7 and KENO.D-8

c. ENDF/B-V/IV means ENDF/B-V for MCNP and ENDF/B-IV for KENO. d. Cross-section type is either continuous cross sections (cont.) or group cross sections (27grp). e. Heavy metal is as a nitrate dissolved in dilute nitric acid solution. 

f. Solution density is in g/ml. g. Containers for fissile solution are cylinders or slabs. h. Annular tank surrounding central cylindrical tank. i. Soln H2O means the moderator is the fissile nitrate solution. 

j. B4C-concrete means borated concrete. 
D-9

References 

D-1. H. R. Dyer and C. V. Parks, “Recommendations for Preparing the Criticality Safety Evaluation of 

Transportation Packages,” NUREG/CR-5661, April 1997. 

D-2 J. J. Lichtenwalter, S. M. Bowman, M. D. DeHart, and C. M. Hopper, “Criticality Benchmark Guide for 

Light-Water-Reactor Fuel in Transportation and Storage Packages,” NUREG/CR-6361, March 1997. 

D-3. Nuclear Energy Agency, “International Handbook of Evaluated Criticality Safety Benchmark 

Experiments,” Organization for Economic Co-operation and Development, NEA/NSC/DOC(95)03, 

September 2001 Edition. 

D-4. B. L. Broadhead, C. M. Hopper, R. L. Childs, and C. V. Parks, “Sensitivity and Uncertainty Analyses 

Applied to Criticality Safety Validation,” NUREG/CR-6655, Vols. 1 and 2 (ORNL/TM-13692/V1 and V2), 

U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, November 1999. 

D-5. B. T. Rearden and R. L. Childs, “Prototypical Sensitivity and Uncertainty Analysis Codes for Criticality 

Safety with the SCALE Code System,” Trans. Am. Nucl. Soc., Washington, D.C., November 2000. 

D-6. M. E. Dunn and B. T. Rearden, “Application of Sensitivity and Uncertainty Analysis Methods to a 

Validation Study for Weapons-Grade Mixed-Oxide Fuel,” 2001 ANS Embedded Topical Meeting on 

Practical Implementation of Nuclear Criticality Safety, Reno, NV, November 11–15, 2001. 

D-7. MCNP is a three-dimensional, Monte Carlo radiation transport code that uses point-wise cross sections, 

developed by LANL. MCNP – A General Monte Carlo N-Particle Transport Code, Version 4B, Judith F. 

Briesmeister, Editor, Los Alamos Report, LA-12625-M, March 1997. Various versions of MCNP are 

available. This reference is for version 4B. 

D-8. KENO is a three-dimensional Monte Carlo criticality module in the SCALE system that uses multigroup 

cross sections, developed by ORNL. SCALE: A Modular Code System for Performing Standardized 

Analyses for Licensing Evaluations, NUREG/CR-0200, Rev. 4 (ORNL/NUREG/CSD-2/R4), Vols. I, II, III, 

October 1995. Various versions of SCALE are available. This reference is for version 4.3. 

D-9. MONK is a three-dimensional Monte Carlo radiation transport code that uses point-wise cross sections, 

developed by A.E.A Technology of the United Kingdom. 

D-10. MORET is a three-dimensional Monte Carlo criticality code that uses multigroup cross sections, developed 

by C.E.A. of France. 
 

D-10

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