ML21193A320

Nd Chapter 33.1-10-13.1

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CHAPTER 33.1-10-13.1 PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Section 33.1-10-13.1-01 Adoption by Reference of Several Sections in 10 Code of Federal Regulations Part 71 33.1-10-13.1-01. Adoption by reference of several sections in 10 Code of Federal Regulations part 71.

10 Code of Federal Regulations 71.0, 71.3, 71.4, 71.5, 71.7, 71.8, 71.9, 71.10, 71.12, 71.13, 71.14, 71.15, 71.17, 71.21, 71.22, 71.23, 71.47, 71.81, 71.83, 71.85, 71.87, 71.88, 71.89, 71.91, 71.93, 71.95, 71.97, 71.101, 71.103, 71.105, 71.106, 71.127, 71.129, 71.131, 71.133, 71.135, and 71.137 and appendix A to part 71 are adopted by reference as they exist on December 30, 2019, with the following exceptions:

1. Not adopted by reference are 10 Code of Federal Regulations 71.0(d), 71.14(b),

71.85(a)-(c), 71.91(b), 71.101(c)(2), (d), and (e).

2. Requirements in 10 Code of Federal Regulations part 71 that apply to "licensed material" or "byproduct material" also apply to naturally occurring or accelerator-produced radioactive material.

3. Where the words "NRC", "commission", "nuclear regulatory commission", "United States nuclear regulatory commission", or "administrator of the appropriate regional office" appear in 10 Code of Federal Regulations part 71, substitute the words "department of environmental quality" except when used in 10 Code of Federal Regulations 71.5(b),

71.10, 71.17(c)(3) and (e), 71.85(c), 71.88(a)(4), 71.93(c), 71.95, 71.97(c), and (c)(3)(iii),

and (f).

4. Where the words ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards appear in 10 Code of Federal Regulations 71.101(c)(1), substitute the words department of environmental quality.

5. The terms certificate of compliance, compliance holder or applicant used in 10 Code of Federal Regulations 71.91(c) and (d), 71.101(a)-(c), 71.103(a), and 71.135 apply only to the U.S. Nuclear Regulatory Commission (NRC) as the NRC is the sole authority for issuing a packages Certificate of Compliance.

6. 10 Code of Federal Regulations 71.9 employee protection also applies to violations of North Dakota Century Code chapters 23.1-02 and 23.1-03.

7. State form number 8414, "notice to employees", must be posted instead of United States nuclear regulatory commission form 3 that is specified in 10 Code of Federal Regulations part 71.

History: Effective January 1, 2019; amended effective July 1, 2021.

General Authority: NDCC 28-32-02; S.L. 2017, ch. 199, § 1 Law Implemented: NDCC 28-32-02 1

PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Subpart A--General Provisions 71.0 Purpose and scope.

71.1 Communications and records.

71.2 Interpretations.

71.3 Requirement for license.

71.4 Definitions.

71.5 Transportation of licensed material.

71.6 Information collection requirements: OMB approval.

71.7 Completeness and accuracy of information.

71.8 Deliberate misconduct.

71.9 Employee Protection.

71.10 Public Inspection of application.

71.11 Protection of Safeguards Information Subpart B--Exemptions 71.12 Specific exemptions.

71.13 Exemption of physicians.

71.14 Exemption for low-level materials.

71.15 Exemption from classicification as fissile material.

71.16 [Reserved]

Subpart C--General Licenses 2

71.17 General license: NRC-approved package.

71.18 [Reserved]

71.19 Previously approved package.

71.21 General license: Use of foreign approved package.

71.22 General license: Fissile material.

71.23 General license: Plutonium-beryllium special form material.

71.24 [Reserved]

71.25 [Reserved]

Subpart D--Application for Package Approval 71.31 Contents of application.

71.33 Package description.

71.35 Package evaluation.

71.37 Quality assurance.

71.38 Renewal of a certificate of compliance or quality assurance program approval.

71.39 Requirement for additional information.

Subpart E--Package Approval Standards 71.41 Demonstration of compliance.

71.43 General standards for all packages.

71.45 Lifting and tie-down standards for all packages.

71.47 External radiation standards for all packages.

71.51 Additional requirements for Type B packages.

71.53 [Reserved]

71.55 General requirements for fissile material packages.

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71.57 [Reserved]

71.59 Standards for arrays of fissile material packages.

71.61 Special requirements for Type B packages containing more than 105A2.

71.63 Special requirements for plutonium shipments.

71.64 Special requirements for plutonium air shipments.

71.65 Additional requirements.

Subpart F--Package, Special Form, and LSA-III Tests 71.70 Incorporations by reference.

71.71 Normal conditions of transport.

71.73 Hypothetical accident conditions.

71.74 Accident conditions for air transport of plutonium.

71.75 Qualification of special form radioactive material.

71.77 Qualification of LSA-III Material.

Subpart G--Operating Controls and Procedures 71.81 Applicability of operating controls and procedures.

71.83 Assumptions as to unknown properties.

71.85 Preliminary determinations.

71.87 Routine determinations.

71.88 Air transport of plutonium.

71.89 Opening instructions.

71.91 Records.

71.93 Inspection and tests.

71.95 Reports.

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71.97 Advance notification of shipment of irradiated reactor fuel and nuclear waste.

71.99 Violations.

71.100 Criminal penalties.

Subpart H--Quality Assurance 71.101 Quality assurance requirements.

71.103 Quality assurance organization.

71.105 Quality assurance program.

71.106 Changes to quality assurance program.

71.107 Package design control.

71.109 Procurement document control.

71.111 Instructions, procedures, and drawings.

71.113 Document control.

71.115 Control of purchased material, equipment, and services.

71.117 Identification and control of materials, parts, and components.

71.119 Control of special processes.

71.121 Internal inspection.

71.123 Test control.

71.125 Control of measuring and test equipment.

71.127 Handling, storage, and shipping control.

71.129 Inspection, test, and operating status.

71.131 Nonconforming materials, parts, or components.

71.133 Corrective action.

71.135 Quality assurance records.

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71.137 Audits.

Appendix A to Part 71--Determination of A1 and A2 Authority: Secs. 53, 57, 62, 63, 81, 161, 182, 183, 68 Stat. 930, 932, 933, 935, 948, 953, 954, as amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 2201, 2232, 2233, 2297f); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); Energy Policy Act of 2005, Pub. L. No. 109-58, 119 Stat. 594 (2005). Section 71.97 also issued under sec. 301, Pub.

L. 96- 295, 94 Stat. 789-790.

Source: 60 FR 50264, Sept. 28, 1995, unless otherwise noted.

[72 FR 63974, Nov. 14, 2007; 73 FR 63572, Oct. 24, 2008]

Subpart A--General Provisions Source: 69 FR 3786, Jan. 26, 2004, unless otherwise noted.

§ 71.0 Purpose and scope.

(a) This part establishes--

(1) Requirements for packaging, preparation for shipment, and transportation of licensed material; and (2) Procedures and standards for NRC approval of packaging and shipping procedures for fissile material and for a quantity of other licensed material in excess of a Type A quantity.

(b) The packaging and transport of licensed material are also subject to other parts of this chapter (e.g., 10 CFR parts 20, 21, 30, 40, 70, and 73) and to the regulations of other agencies (e.g., the U.S. Department of Transportation (DOT) and the U.S. Postal Service)1 having jurisdiction over means of transport. The requirements of this part are in addition to, and not in substitution for, other requirements.

(c) The regulations in this part apply to any licensee authorized by specific or general license issued by the Commission to receive, possess, use, or transfer licensed material, if the licensee delivers that material to a carrier for transport, transports the material outside the site of usage as specified in the NRC license, or transports that material on public highways. No provision of this part authorizes possession of licensed material.

(d)(1) Exemptions from the requirement for license in § 71.3 are specified in § 71.14. General licenses for which no NRC package approval is required are issued in §§ 71.21 through 71.23.

The general license in § 71.17 requires that an NRC certificate of compliance or other package approval be issued for the package to be used under this general license.

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(2) Application for package approval must be completed in accordance with subpart D of this part, demonstrating that the design of the package to be used satisfies the package approval standards contained in subpart E of this part, as related to the tests of subpart F of this part.

(3) A licensee transporting licensed material, or delivering licensed material to a carrier for transport, shall comply with the operating control requirements of subpart G of this part; the quality assurance requirements of subpart H of this part; and the general provisions of subpart A of this part, including DOT regulations referenced in § 71.5.

(e) The regulations of this part apply to any person holding, or applying for, a certificate of compliance, issued pursuant to this part, for a package intended for the transportation of radioactive material, outside the confines of a licensee's facility or authorized place of use.

(f) The regulations in this part apply to any person required to obtain a certificate of compliance, or an approved compliance plan, pursuant to part 76 of this chapter, if the person delivers radioactive material to a common or contract carrier for transport or transports the material outside the confines of the person's plant or other authorized place of use.

(g) This part also gives notice to all persons who knowingly provide to any licensee, certificate holder, quality assurance program approval holder, applicant for a license, certificate, or quality assurance program approval, or to a contractor, or subcontractor of any of them, components, equipment, materials, or other goods or services, that relate to a licensee's, certificate holder's, quality assurance program approval holder's, or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 71.8.

1 Postal Service Manual (Domestic Mail Manual), section 124, which is incorporated by reference at 39 CFR 111.1.

[80 FR 34011, Jun. 12, 2015]

§ 71.1 Communications and records.

(a) Except where otherwise specified, all communications and reports concerning the regulations in this part and applications filed under them should be sent by mail addressed: ATTN:

Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the submission date falls on a Saturday, Sunday, or a Federal 7

holiday, the next Federal working day becomes the official due date.

(b) Each record required by this part must be legible throughout the retention period specified by each Commission regulation. The record may be the original or a reproduced copy or a microform provided that the copy or microform is authenticated by authorized personnel and that the microform is capable of producing a clear copy throughout the required retention period. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.

[69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004; 70 FR 69421, Nov. 16, 2005; 72 FR 33386, Jun. 18, 2007; 74 FR 62683, Dec. 1, 2009; 75 FR 73945, Nov. 30, 2010; 79 FR 75741, Dec. 19, 2014; 80 FR 74981, Dec. 1, 2015; 84 FR 65639, Nov. 29, 2109; 84 FR 66561, Dec. 5, 2019]

§ 71.2 Interpretations.

Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission, other than a written interpretation by the General Counsel, will be recognized to be binding upon the Commission.

§ 71.3 Requirement for license.

Except as authorized in a general license or a specific license issued by the Commission, or as exempted in this part, no licensee may--

(a) Deliver licensed material to a carrier for transport; or (b) Transport licensed material.

§ 71.4 Definitions.

The following terms are as defined here for the purpose of this part. To ensure compatibility with international transportation standards, all limits in this part are given in terms of dual units: The International System of Units (SI) followed or preceded by U.S. standard or customary units.

The U.S. customary units are not exact equivalents but are rounded to a convenient value, providing a functionally equivalent unit. For the purpose of this part, either unit may be used.

A1 means the maximum activity of special form radioactive material permitted in a Type A package. This value is either listed in Appendix A, Table A-1, of this part, or may be derived in accordance with the procedures prescribed in Appendix A of this part.

A2 means the maximum activity of radioactive material, other than special form material, LSA, and SCO material, permitted in a Type A package. This value is either listed in Appendix A, Table A-1, of this part, or may be derived in accordance with the procedures prescribed in 8

Appendix A of this part.

Carrier means a person engaged in the transportation of passengers or property by land or water as a common, contract, or private carrier, or by civil aircraft.

Certificate holder means a person who has been issued a certificate of compliance or other package approval by the Commission.

Certificate of Compliance (CoC) means the certificate issued by the Commission under subpart D of this part which approves the design of a package for the transportation of radioactive material.

Close reflection by water means immediate contact by water of sufficient thickness for maximum reflection of neutrons.

Consignment means each shipment of a package or groups of packages or load of radioactive material offered by a shipper for transport.

Containment system means the assembly of components of the packaging intended to retain the radioactive material during transport.

Contamination means the presence of a radioactive substance on a surface in quantities in excess of 0.4 Bq/cm2 (1x10-5 µCi/cm2) for beta and gamma emitters and low toxicity alpha emitters, or 0.04 Bq/cm2 (1x10-6 µCi/cm2) for all other alpha emitters.

(1) Fixed contamination means contamination that cannot be removed from a surface during normal conditions of transport.

(2) Non-fixed contamination means contamination that can be removed from a surface during normal conditions of transport.

Conveyance means:

(1) For transport by public highway or rail any transport vehicle or large freight container; (2) For transport by water any vessel, or any hold, compartment, or defined deck area of a vessel including any transport vehicle on board the vessel; and (3) For transport by any aircraft.

Criticality Safety Index (CSI) means the dimensionless number (rounded up to the next tenth) assigned to and placed on the label of a fissile material package, to designate the degree of control of accumulation of packages, overpacks or freight containers containing fissile material during transportation. Determination of the criticality safety index is described in §§ 71.22, 71.23, and 71.59. The criticality safety index for an overpack, freight container, consignment or conveyance containing fissile material packages is the arithmetic sum of the criticality safety 9

indices of all the fissile material packages contained within the overpack, freight container, consignment or conveyance.

Deuterium means, for the purposes of §§ 71.15 and 71.22, deuterium and any deuterium compounds, including heavy water, in which the ratio of deuterium atoms to hydrogen atoms exceeds 1:5000.

DOT means the U.S. Department of Transportation.

Exclusive use means the sole use by a single consignor of a conveyance for which all initial, intermediate, and final loading and unloading are carried out in accordance with the direction of the consignor or consignee. The consignor and the carrier must ensure that any loading or unloading is performed by personnel having radiological training and resources appropriate for safe handling of the consignment. The consignor must issue specific instructions, in writing, for maintenance of exclusive use shipment controls, and include them with the shipping paper information provided to the carrier by the consignor.

Fissile material means the radionuclides uranium-233, uranium-235, plutonium-239, and plutonium-241, or any combination of these radionuclides. Fissile material means the fissile nuclides themselves, not material containing fissile nuclides. Unirradiated natural uranium and depleted uranium and natural uranium or depleted uranium, that has been irradiated in thermal reactors only, are not included in this definition. Certain exclusions from fissile material controls are provided in §71.15.

Graphite means, for the purposes of §§ 71.15 and 71.22, graphite with a boron equivalent content less than 5 parts per million and density greater than 1.5 grams per cubic centimeter.

Indian Tribe means and Indian of Alaska native Tribe, band, nation, pueblo, village, or community that the Secretary of the Interior acknowledges to exist as an Indian Tribe pursuant to the Federally Recognized Indian Tribe List Act of 1994, 25 U.S.C. 479a.

Licensed material means byproduct, source, or special nuclear material received, possessed, used, or transferred under a general or specific license issued by the Commission pursuant to the regulations in this chapter.

Low Specific Activity (LSA) material means radioactive material with limited specific activity which is nonfissile or is excepted under § 71.15, and which satisfies the descriptions and limits set forth in the following section. Shielding materials surrounding the LSA material may not be considered in determining the estimated average specific activity of the package contents. The LSA material must be in one of three groups:

(1) LSA-I.

(i) Uranium and thorium ores, concentrates of uranium and thorium ores, and other ores containing naturally occurring radionuclides that are intended to be processed for the use of these radionuclides; 10

(ii) Natural uranium, depleted uranium, natural thorium or their compounds or mixtures, provided they are unirradiated and in solid or liquid form; (iii) Radioactive material other than fissile material, for which the A2 value is unlimited; or (iv) Other radioactive material in which the activity is distributed throughout and the estimated average specific activity does not exceed 30 times the value for exempt material activity concentration determined in accordance with appendix A.

(2) LSA-II.

(i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/liter); or (ii) Other radioactive material in which the activity is distributed throughout and the estimated average specific activity does not exceed 104 A2/g for solids and gases, and 105 A2/g for liquids.

(3) LSA-III. Solids (e.g., consolidated wastes, activated materials), excluding powders, that satisfy the requirements of § 71.77, in which:

(i) The radioactive material is distributed throughout a solid or a collection of solid objects, or is essentially uniformly distributed in a solid compact binding agent (such as concrete, bitumen, ceramic, etc.);

(ii) The radioactive material is relatively insoluble, or it is intrinsically contained in a relatively insoluble material, so that even under loss of packaging, the loss of radioactive material per package by leaching when placed in water for 7 days will not exceed 0.1 A2; and (iii) The estimated average specific activity of the solid, excluding any shielding material, does not exceed 2 x 103 A2/g.

Low toxicity alpha emitters means natural uranium, depleted uranium, natural thorium; uranium-235, uranium-238, thorium-232, thorium-228 or thorium-230 when contained in ores or physical or chemical concentrates or tailings; or alpha emitters with a half-life of less than 10 days.

Maximum normal operating pressure means the maximum gauge pressure that would develop in the containment system in a period of 1 year under the heat condition specified in §71.71(c)(1),

in the absence of venting, external cooling by an ancillary system, or operational controls during transport.

Natural thorium means thorium with the naturally occurring distribution of thorium isotopes (essentially 100 weight percent thorium-232).

Normal form radioactive material means radioactive material that has not been demonstrated to qualify as "special form radioactive material."

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Optimum interspersed hydrogenous moderation means the presence of hydrogenous material between packages to such an extent that the maximum nuclear reactivity results.

Package means the packaging together with its radioactive contents as presented for transport.

(1) Fissile material package or Type AF package, Type BF package, Type B(U)F package, or Type B(M)F package means a fissile material packaging together with its fissile material contents.

(2) Type A package means a Type A packaging together with its radioactive contents. A Type A package is defined and must comply with the DOT regulations in 49 CFR part 173.

(3) Type B package means a Type B packaging together with its radioactive contents. On approval, a Type B package design is designated by NRC as B(U) unless the package has a maximum normal operating pressure of more than 700 kPa (100 lbs/in2) gauge or a pressure relief device that would allow the release of radioactive material to the environment under the tests specified in §71.73 (hypothetical accident conditions), in which case it will receive a designation B(M). B(U) refers to the need for unilateral approval of international shipments; B(M) refers to the need for multilateral approval of international shipments. There is no distinction made in how packages with these designations may be used in domestic transportation. To determine their distinction for international transportation, see DOT regulations in 49 CFR Part 173. A Type B package approved before September 6, 1983, was designated only as Type B. Limitations on its use are specified in §71.19.

Packaging means the assembly of components necessary to ensure compliance with the packaging requirements of this part. It may consist of one or more receptacles, absorbent materials, spacing structures, thermal insulation, radiation shielding, and devices for cooling or absorbing mechanical shocks. The vehicle, tie-down system, and auxiliary equipment may be designated as part of the packaging.

Special form radioactive material means radioactive material that satisfies the following conditions:

(1) It is either a single solid piece or is contained in a sealed capsule that can be opened only by destroying the capsule; (2) The piece or capsule has at least one dimension not less than 5 mm (0.2 in); and (3) It satisfies the requirements of §71.75. A special form encapsulation designed in accordance with the requirements of § 71.4 in effect on June 30, 1983 (see 10 CFR part 71, revised as of January 1, 1983), and constructed before July 1, 1985; a special form encapsulation designed in accordance with the requirements of § 71.4 in effect on March 31, 1996 (see 10 CFR part 71, revised as of January 1, 1996), and constructed before April 1, 1998; and special form material that was successfully tested before September 10, 2015 in accordance with the requirements of

§ 71.75(d) of this section in effect before September 10, 2015 may continue to be used. Any other special form encapsulation must meet the specifications of this definition.

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Specific activity of a radionuclide means the radioactivity of the radionuclide per unit mass of that nuclide. The specific activity of a material in which the radionuclide is essentially uniformly distributed is the radioactivity per unit mass of the material.

Spent nuclear fuel or Spent fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, has undergone at least 1 year's decay since being used as a source of energy in a power reactor, and has not been chemically separated into its constituent elements by reprocessing. Spent fuel includes the special nuclear material, byproduct material, source material, and other radioactive materials associated with fuel assemblies.

State means a State of the United States, the District of Columbia, the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American Samoa, and the Commonwealth of the Northern Mariana Islands.

Surface Contaminated Object (SCO) means a solid object that is not itself classed as radioactive material, but which has radioactive material distributed on any of its surfaces. SCO must be in one of two groups with surface activity not exceeding the following limits:

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

(i) The nonfixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 Bq/cm2 (104 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 0.4 Bq/cm2 (10-5 microcurie/cm2) for all other alpha emitters; (ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 x 104 Bq/cm2 (1.0 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 4 x 103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha emitters; and (iii) The nonfixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 x 104 Bq/cm2 (1 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 4 x 103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha emitters.

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

(i) The nonfixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 400 Bq/cm2 (102 microcurie/cm2) for beta and gamma and low toxicity alpha emitters or 40 Bq/cm2 (103 microcurie/cm2) for all other alpha emitters; (ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8 x 105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8 x 104 Bq/cm2 (2 microcuries/cm2) for all other alpha 13

emitters; and (iii) The nonfixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8 x 105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8 x 104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters.

Transport index (TI) means the dimensionless number (rounded up to the next tenth) placed on the label of a package, to designate the degree of control to be exercised by the carrier during transportation. The transport index is the number determined by multiplying the maximum radiation level in millisievert (mSv) per hour at 1 meter (3.3 ft) from the external surface of the package by 100 (equivalent to the maximum radiation level in millirem per hour at 1 meter (3.3 ft)).

Tribal official means the highest ranking individual that represents Tribal leadership, such as the Chief, President, or Tribal Council leadership.

Type A quantity means a quantity of radioactive material, the aggregate radioactivity of which does not exceed A1 for special form radioactive material, or A2, for normal form radioactive material, where A1 and A2 are given in Table A-1 of this part, or may be determined by procedures described in Appendix A of this part.

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

Unirradiated uranium means uranium containing not more than 2 x 103 Bq of plutonium per gram of uranium-235, not more than 9 x 106 Bq of fission products per gram of uranium-235, and not more than 5 x 10-3 g of uranium-236 per gram of uranium-235.

Uranium - natural, depleted, enriched.

(1) Natural uranium means uranium (which may be chemically separated) with the naturally occurring distribution of uranium isotopes (approximately 0.711 weight percent uranium-235 and the remainder by weight essentially uranium-238).

(2) Depleted uranium means uranium containing less uranium-235 than the naturally occurring distribution of uranium isotopes.

(3) Enriched uranium means uranium containing more uranium-235 than the naturally occurring distribution of uranium isotopes.

[69 FR 3787, Jan. 26, 2004; 69 FR 58038, Sep. 29, 2004; 77 FR 34204, Jun. 11, 2012; 80 FR 34011, Jun. 12, 2015; 80 FR 48684, Aug. 14, 2015; 80 FR 74981, Dec. 1, 2015; 82 FR 52825, Nov. 15, 2017]

§ 71.5 Transportation of licensed material.

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(a) Each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the DOT regulations in 49 CFR parts 107, 171 through 180, and 390 through 397, appropriate to the mode of transport.

(1) The licensee shall particularly note DOT regulations in the following areas:

(i) Packaging--49 CFR part 173: subparts A, B, and I.

(ii) Marking and labeling--49 CFR part 172: subpart D; and §§ 172.400 through 172.407 and

§§ 172.436 through 172.441 of subpart E.

(iii) Placarding--49 CFR part 172: subpart F, especially §§ 172.500 through 172.519 and 172.556; and appendices B and C.

(iv) Accident reporting--49 CFR part 171: §§ 171.15 and 171.16.

(v) Shipping papers and emergency information--49 CFR part 172: subparts C and G.

(vi) Hazardous material employee training--49 CFR part 172: subpart H.

(vii) Security plans--49 CFR part 172: subpart I.

(viii) Hazardous material shipper/carrier registration--49 CFR part 107: subpart G.

(2) The licensee shall also note DOT regulations pertaining to the following modes of transportation:

(i) Rail--49 CFR part 174: subparts A through D and K.

(ii) Air--49 CFR part 175.

(iii) Vessel--49 CFR part 176: subparts A through F and M.

(iv) Public Highway--49 CFR part 177 and parts 390 through 397.

(b) If DOT regulations are not applicable to a shipment of licensed material, the licensee shall conform to the standards and requirements of the DOT specified in paragraph (a) of this section to the same extent as if the shipment or transportation were subject to DOT regulations. A request for modification, waiver, or exemption from those requirements, and any notification referred to in those requirements, must be filed with, or made to, the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

§ 71.6 Information collection requirements: OMB approval.

15

(a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0008.

(b) The approved information collection requirements contained in this part appear in §§ 71.5, 71.7, 71.9, 71.12, 71.17, 71.19, 71.22, 71.23, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.41, 71.47, 71.85, 71.87, 71.89, 71.91, 71.93, 71.95, 71.97, 71.101, 71.103, 71.105, 71.106, 71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, 71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137, and appendix A, paragraph II.

[75 FR 73945, Nov. 30, 2010; 80 FR 34012, Jun. 12, 2015]

§ 71.7 Completeness and accuracy of information.

(a) Information provided to the Commission by a licensee, certificate holder, or an applicant for a license or CoC; or information required by statute or by the Commission's regulations, orders, license or CoC conditions, to be maintained by the licensee or certificate holder, must be complete and accurate in all material respects.

(b) Each licensee, certificate holder, or applicant for a license or CoC must notify the Commission of information identified by the licensee, certificate holder, or applicant for a license or CoC as having, for the regulated activity, a significant implication for public health and safety or common defense and security. A licensee, certificate holder, or an applicant for a license or CoC violates this paragraph only if the licensee, certificate holder, or applicant for a license or CoC fails to notify the Commission of information that the licensee, certificate holder, or applicant for a license or CoC has identified as having a significant implication for public health and safety or common defense and security. Notification must be provided to the Administrator of the appropriate Regional Office within 2 working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.

§ 71.8 Deliberate misconduct.

(a) This section applies to any--

(1) Licensee; (2) Certificate holder; (3) Quality assurance program approval holder; (4) Applicant for a license, certificate, or quality assurance program approval; 16

(5) Contractor (including a supplier or consultant) or subcontractor, to any person identified in paragraph (a)(4) of this section; or (6) Employees of any person identified in paragraphs (a)(1) through (a)(5) of this section.

(b) A person identified in paragraph (a) of this section who knowingly provides to any entity, listed in paragraphs (a)(1) through (a)(5) of this section, any components, materials, or other goods or services that relate to a licensee's, certificate holder's, quality assurance program approval holder's, or applicant's activities subject to this part may not:

(1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee, certificate holder, quality assurance program approval holder, or any applicant to be in violation of any rule, regulation, or order; or any term, condition or limitation of any license, certificate, or approval issued by the Commission; or (2) Deliberately submit to the NRC, a licensee, a certificate holder, quality assurance program approval holder, an applicant for a license, certificate or quality assurance program approval, or a licensee's, applicant's, certificate holder's, or quality assurance program approval holder's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.

(c) A person who violates paragraph (b)(1) or (b)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.

(d) For the purposes of paragraph (b)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:

(1) Would cause a licensee, certificate holder, quality assurance program approval holder, or applicant for a license, certificate, or quality assurance program approval to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license or certificate issued by the Commission; or (2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, certificate holder, quality assurance program approval holder, applicant, or the contractor or subcontractor of any of them.

§ 71.9 Employee protection.

(a) Discrimination by a Commission licensee, certificate holder, an applicant for a Commission license or a CoC, or a contractor or subcontractor of any of these, against an employee for engaging in certain protected activities, is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act of 1954, as amended, or the Energy Reorganization Act of 1974, as amended.

17

(1) The protected activities include, but are not limited to:

(i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) of this section or possible violations of requirements imposed under either of those statutes; (ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) of this section or under these requirements if the employee has identified the alleged illegality to the employer; (iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements; (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) of this section; and (v) Assisting or participating in, or is about to assist or participate in, these activities.

(2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee's assistance or participation.

(3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent),

deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.

(b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.

(c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, certificate holder, applicant for a Commission license or a CoC, or a contractor or subcontractor of any of these may be grounds for:

(1) Denial, revocation, or suspension of the license or the CoC; (2) Imposition of a civil penalty on the licensee, applicant, or a contractor or subcontractor of the licensee or applicant; or (3) Other enforcement action.

18

(d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.

(e)(1) Each licensee, certificate holder, and applicant for a license or CoC must prominently post the current revision of NRC Form 3, "Notice to Employees," referenced in §19.11(c) of this chapter. This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. The premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license or CoC, and for 30 days following license or CoC termination.

(2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in Appendix D to part 20 of this chapter or by calling the NRC Publishing Services Branch at 301-415-5877.

(f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in a protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.

[72 FR 63975, Nov. 14, 2007; 79 FR 66605, Nov. 10, 2014]

§ 71.10 Public inspection of application.

Applications for approval of a package design under this part, which are submitted to the Commission, may be made available for public inspection, in accordance with provisions of parts 2 and 9 of this chapter. This includes an application to amend or revise an existing package design, any associated documents and drawings submitted with the application, and any responses to NRC requests for additional information.

§ 71.11 Protection of Safeguards Information Each licensee, certificate holder, or applicant for a Certificate of Compliance for a transportation package for transport of irradiated reactor fuel, strategic special nuclear material, a critical mass of special nuclear material, or byproduct material in quantities determined by the Commission through order or regulation to be significant to the public health and safety or the common defense and security, shall protect Safeguards Information against unauthorized disclosure in accordance with the requirements in § 73.21 and the requirements of § 73.22 or § 73.23 of this chapter, as applicable.

19

[73 FR 63572, Oct. 24, 2008]

Subpart B--Exemptions Source: 69 FR 3786, Jan. 26, 2004, unless otherwise noted.

§ 71.12 Specific exemptions.

On application of any interested person or on its own initiative, the Commission may grant any exemption from the requirements of the regulations in this part that it determines is authorized by law and will not endanger life or property nor the common defense and security.

§ 71.13 Exemption of physicians.

Any physician licensed by a State to dispense drugs in the practice of medicine is exempt from § 71.5 with respect to transport by the physician of licensed material for use in the practice of medicine. However, any physician operating under this exemption must be licensed under 10 CFR part 35 or the equivalent Agreement State regulations.

§ 71.14 Exemption for low-level materials.

(a) A licensee is exempt from all the requirements of this part with respect to shipment or carriage of the following low-level materials:

(1) Natural material and ores containing naturally occurring radionuclides that are either in their natural state, or have only been processed for purposes other than for the extraction of the radionuclides, and which are not intended to be processed for the use of these radionuclides, provided the activity concentration of the material does not exceed 10 times the applicable radionuclide activity concentration values specified in appendix A, Table A-2, or Table A-3 of this part.

(2) Materials for which the activity concentration is not greater than the activity concentration values specified in appendix A, Table A-2, or Table A-3 of this part, or for which the consignment activity is not greater than the limit for an exempt consignment found in appendix A, Table A-2, or Table A-3 of this part.

(3) Non-radioactive solid objects with radioactive substances present on any surfaces in quantities not in excess of the levels cited in the definition of contamination in § 71.4.

(b) A licensee is exempt from all the requirements of this part, other than §§ 71.5 and 71.88, with respect to shipment or carriage of the following packages, provided the packages do not contain any fissile material, or the material is exempt from classification as fissile material under

§ 71.15:

(1) A package that contains no more than a Type A quantity of radioactive material; 20

(2) A package transported within the United States that contains no more than 0.74 TBq (20 Ci) of special form plutonium-244; or (3) The package contains only LSA or SCO radioactive material, provided--

(i) That the LSA or SCO material has an external radiation dose of less than or equal to 10 mSv/h (1 rem/h), at a distance of 3 m from the unshielded material; or (ii) That the package contains only LSA-I or SCO-I material.

[80 FR 34012, Jun. 12, 2015]

§ 71.15 Exemption from classification as fissile material.

Fissile material meeting the requirements of at least one of the paragraphs (a) through (f) of this section are exempt from classification as fissile material and from the fissile material package standards of §§ 71.55 and 71.59, but are subject to all other requirements of this part, except as noted.

(a) Individual package containing 2 grams or less fissile material.

(b) Individual or bulk packaging containing 15 grams or less of fissile material provided the package has at least 200 grams of solid nonfissile material for every gram of fissile material.

Lead, beryllium, graphite, and hydrogenous material enriched in deuterium may be present in the package but must not be included in determining the required mass for solid nonfissile material.

(c)(1) Low concentrations of solid fissile material commingled with solid nonfissile material, provided that:

(i) There is at least 2000 grams of solid nonfissile material for every gram of fissile material, and (ii) There is no more than 180 grams of fissile material distributed within 360 kg of contiguous nonfissile material.

(2) Lead, beryllium, graphite, and hydrogenous material enriched in deuterium may be present in the package but must not be included in determining the required mass of solid nonfissile material.

(d) Uranium enriched in uranium-235 to a maximum of 1 percent by weight, and with total plutonium and uranium-233 content of up to 1 percent of the mass of uranium-235, provided that the mass of any beryllium, graphite, and hydrogenous material enriched in deuterium constitutes less than 5 percent of the uranium mass, and that the fissile material is distributed homogeneously and does not form a lattice arrangement within the package.

(e) Liquid solutions of uranyl nitrate enriched in uranium-235 to a maximum of 2 percent by mass, with a total plutonium and uranium-233 content not exceeding 0.002 percent of the mass 21

of uranium, and with a minimum nitrogen to uranium atomic ratio (N/U) of 2. The material must be contained in at least a DOT Type A package.

(f) Packages containing, individually, a total plutonium mass of not more than 1000 grams, of which not more than 20 percent by mass may consist of plutonium-239, plutonium-241, or any combination of these radionuclides.

[80 FR 34012, Jun. 12, 2015]

§ 71.16 [Reserved]

Subpart C--General Licenses Source: 69 FR 3792, Jan. 26, 2004, unless otherwise noted.

§ 71.17 General license: NRC-approved package.

(a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package for which a license, certificate of compliance (CoC), or other approval has been issued by the NRC.

(b) This general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.

(c) Each licensee issued a general license under paragraph (a) of this section shall (1) Maintain a copy of the Certificate of Compliance, or other approval of the package, and the drawings and other documents referenced in the approval relating to the use and maintenance of the packaging and to the actions to be taken before shipment; (2) Comply with the terms and conditions of the license, certificate, or other approval, as applicable, and the applicable requirements of subparts A, G, and H of this part; and (3) Submit in writing before the first use of the package to: ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards, using an appropriate method listed in § 71.1(a), the licensee's name and license number and the package identification number specified in the package approval.

(d) This general license applies only when the package approval authorizes use of the package under this general license.

(e) For a Type B or fissile material package, the design of which was approved by NRC before April 1, 1996, the general license is subject to the additional restrictions of § 71.19.

[75 FR 73945, Nov. 30, 2010; 79 FR 75741, Dec. 19, 2014; 80 FR 34012, Jun. 12, 2015; 84 FR 65639, Nov. 29, 2019; 84 FR 66561, Dec. 5, 2019]

22

§ 71.18 [Reserved]

§ 71.19 Previously approved package.

(a) [Reserved]

(b) A Type B(U) package, a Type B(M) package, or a fissile material package, previously approved by the NRC but without the designation "- 85" in the identification number of the NRC CoC, may be used under the general license of § 71.17 with the following additional conditions:

(1) Fabrication of the package is satisfactorily completed by April 1, 1999, as demonstrated by application of its model number in accordance with § 71.85(c);

(2) A package used for a shipment to a location outside the United States is subject to multilateral approval as defined in DOT regulations at 49 CFR 173.403; and (3) A serial number which uniquely identifies each packaging which conforms to the approved design is assigned to and legibly and durably marked on the outside of each packaging.

(c) A Type B(U) package, a Type B(M) package, or a fissile material package previously approved by the NRC with the designation "-85" in the identification number of the NRC CoC, may be used under the general license of § 71.17 with the following additional conditions:

(1) Fabrication of the package must be satisfactorily completed by December 31, 2006, as demonstrated by application of its model number in accordance with § 71.85(c); and (2) After December 31, 2003, a package used for a shipment to a location outside the United States is subject to multilateral approval as defined in DOT regulations at 49 CFR 173.403.

(d) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided--

(1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified in §§ 71.71 and 71.73; (2) The modifications of a fissile material package are not significant, with respect to the prevention of criticality, when the package is subjected to the tests specified in §§ 71.71 and 71.73; and (3) The modifications to the package satisfy the requirements of this part.

(e) NRC will revise the package identification number to designate previously approved package designs as B, BF, AF, B(U), B(M), B(U)F, B(M)F, B(U)-85, B(U)F-85, B(M)-85, B(M)F-85, or AF-85 as appropriate, and with the identification number suffix "-96" after receipt of an application demonstrating that the design meets the requirements of this part.

23

[80 FR 34012, Jun. 12, 2015]

§ 71.21 General license: Use of foreign approved package.

(a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package, the design of which has been approved in a foreign national competent authority certificate, that has been revalidated by the DOT as meeting the applicable requirements of 49 CFR 171.23.

(b) Except as otherwise provided in this section, the general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the applicable provisions of subpart H of this part.

(c) This general license applies only to shipments made to or from locations outside the United States.

(d) Each licensee issued a general license under paragraph (a) of this section shall (1) Maintain a copy of the applicable certificate, the revalidation, and the drawings and other documents referenced in the certificate, relating to the use and maintenance of the packaging and to the actions to be taken before shipment; and (2) Comply with the terms and conditions of the certificate and revalidation, and with the applicable requirements of subparts A, G, and H of this part.

[80 FR 34012, Jun. 12, 2015]

§ 71.22 General license: Fissile material.

(a) A general license is issued to any licensee of the Commission to transport fissile material, or to deliver fissile material to a carrier for transport, if the material is shipped in accordance with this section. The fissile material need not be contained in a package which meets the standards of subparts E and F of this part; however, the material must be contained in a Type A package. The Type A package must also meet the DOT requirements of 49 CFR 173.417(a).

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.

(c) The general license applies only when a package's contents:

(1) Contain no more than a Type A quantity of radioactive material; and (2) Contain less than 500 total grams of beryllium, graphite, or hydrogenous material enriched in deuterium.

(d) The general license applies only to packages containing fissile material that are labeled with a 24

CSI which:

(1) Has been determined in accordance with paragraph (e) of this section; (2) Has a value less than or equal to 10; and (3) For a shipment of multiple packages containing fissile material, the sum of the CSIs must be less than or equal to 50 (for shipment on a nonexclusive use conveyance) and less than or equal to 100 (for shipment on an exclusive use conveyance).

(e)(1) The value for the CSI must be greater than or equal to the number calculated by the following equation:

(2) The calculated CSI must be rounded up to the first decimal place; (3) The values of X, Y, and Z used in the CSI equation must be taken from Tables 71-1 or 71-2, as appropriate; (4) If Table 71-2 is used to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium must be assumed to be zero; and (5) Table 71-1 values for X, Y, and Z must be used to determine the CSI if:

(i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight percent enrichment; or (iv) Substances having a moderating effectiveness (i.e., an average hydrogen density greater than H2O) (e.g., certain hydrocarbon oils or plastics) are present in any form, except as polyethylene used for packing or wrapping.

Table 71-1. Mass Limits for General License Packages Containing Mixed Quantities of Fissile Material or Uranium-235 of Unknown Enrichment per § 71.22(e)

Fissile material mass Fissile material mass mixed with moderating mixed with moderating Fissile material substances having an substances having an average hydrogen density average hydrogen density less than or equal to H2O greater than H2Oa 25

(grams) (grams) 235 U (X) 60 38 233 U (Y) 43 27 239 Pu or 241Pu (Z) 37 24 a

When mixtures of moderating substances are present, the lower mass limits shall be used if more than 15 percent of the moderating substance has an average hydrogen density greater than H2O.

Table 71-2. Mass Limits for General License Packages Containing Uranium-235 of Known Enrichment per § 71.22(e)

Uranium enrichment in weight percent of 235U not Fissile material mass of 235U (X) exceeding (grams) 24 60 20 63 15 67 11 72 10 76 9.5 78 9 81 8.5 82 8 85 7.5 88 7 90 6.5 93 6 97 5.5 102 5 108 4.5 114 4 120 3.5 132 3 150 2.5 180 2 246 26

1.5 408 1.35 480 1 1,020 0.92 1,800

[69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004]

§ 71.23 General license: Plutonium-beryllium special form material.

(a) A general license is issued to any licensee of the Commission to transport fissile material in the form of plutonium-beryllium (Pu-Be) special form sealed sources, or to deliver Pu-Be sealed sources to a carrier for transport, if the material is shipped in accordance with this section. This material need not be contained in a package which meets the standards of subparts E and F of this part; however, the material must be contained in a Type A package. The Type A package must also meet the DOT requirements of 49 CFR 173.417(a).

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.

(c) The general license applies only when a package's contents:

(1) Contain no more than a Type A quantity of radioactive material; and (2) Contain less than 1000 g of plutonium, provided that: plutonium-239, plutonium-241, or any combination of these radionuclides, constitutes less than 240 g of the total quantity of plutonium in the package.

(d) The general license applies only to packages labeled with a CSI which:

(1) Has been determined in accordance with paragraph (e) of this section; (2) Has a value less than or equal to 100; and (3) For a shipment of multiple packages containing Pu-Be sealed sources, the sum of the CSIs must be less than or equal to 50 (for shipment on a nonexclusive use conveyance) and less than or equal to 100 (for shipment on an exclusive use conveyance).

(e)(1) The value for the CSI must be greater than or equal to the number calculated by the following equation:

27

(2) The calculated CSI must be rounded up to the first decimal place.

§ 71.24 [Reserved]

§ 71.25 [Reserved]

Subpart D--Application for Package Approval

§ 71.31 Contents of application.

(a) An application for an approval under this part must include, for each proposed packaging design, the following information:

(1) A package description as required by § 71.33; (2) A package evaluation as required by § 71.35; and (3) A quality assurance program description, as required by § 71.37, or a reference to a previously approved quality assurance program.

(b) Except as provided in § 71.13, an application for modification of a package design, whether for modification of the packaging or authorized contents, must include sufficient information to demonstrate that the proposed design satisfies the package standards in effect at the time the application is filed.

(c) The applicant shall identify any established codes and standards proposed for use in package design, fabrication, assembly, testing, maintenance, and use. In the absence of any codes and standards, the applicant shall describe and justify the basis and rationale used to formulate the package quality assurance program.

[80 FR 34012, Jun. 12, 2015]

§ 71.33 Package description.

The application must include a description of the proposed package in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation of the package. The description must include --

(a) With respect to the packaging --

(1) Classification as Type B(U), Type B(M), or fissile material packaging; (2) Gross weight; (3) Model number; 28

(4) Identification of the containment system; (5) Specific materials of construction, weights, dimensions, and fabrication methods of --

(i) Receptacles; (ii) Materials specifically used as nonfissile neutron absorbers or moderators; (iii) Internal and external structures supporting or protecting receptacles; (iv) Valves, sampling ports, lifting devices, and tie-down devices; and (v) Structural and mechanical means for the transfer and dissipation of heat; and (6) Identification and volumes of any receptacles containing coolant.

(b) With respect to the contents of the package --

(1) Identification and maximum radioactivity of radioactive constituents; (2) Identification and maximum quantities of fissile constituents; (3) Chemical and physical form; (4) Extent of reflection, the amount and identity of nonfissile materials used as neutron absorbers or moderators, and the atomic ratio of moderator to fissile constituents; (5) Maximum normal operating pressure; (6) Maximum weight; (7) Maximum amount of decay heat; and (8) Identification and volumes of any coolants.

§ 71.35 Package evaluation.

The application must include the following:

(a) A demonstration that the package satisfies the standards specified in subparts E and F of this part; (b) For a fissile material package, the allowable number of packages that may be transported in the same vehicle in accordance with § 71.59; and (c) For a fissile material shipment, any proposed special controls and precautions for transport, 29

loading, unloading, and handling and any proposed special controls in case of an accident or delay.

§ 71.37 Quality assurance.

(a) The applicant shall describe the quality assurance program (see Subpart H of this part) for the design, fabrication, assembly, testing, maintenance, repair, modification, and use of the proposed package.

(b) The applicant shall identify any specific provisions of the quality assurance program that are applicable to the particular package design under consideration, including a description of the leak testing procedures.

§ 71.38 Renewal of a certificate of compliance or quality assurance program approval.

(a) Except as provided in paragraph (b) of this section, each Certificate of Compliance or Quality Assurance Program Approval expires at the end of the day, in the month and year stated in the approval.

(b) In any case in which a person, not less than 30 days before the expiration of an existing Certificate of Compliance or Quality Assurance Program Approval issued pursuant to the part, has filed an application in proper form for renewal of either of those approvals, the existing Certificate of Compliance or Quality Assurance Program Approval for which the renewal application was filed shall not be deemed to have expired until final action on the application for renewal has been taken by the Commission.

(c) In applying for renewal of an existing Certificate of Compliance or Quality Assurance Program Approval, an applicant may be required to submit a consolidated application that incorporates all changes to its program that, are incorporated by reference in the existing approval or certificate, into as few referenceable documents as reasonably achievable.

[80 FR 34012, Jun. 12, 2015]

§ 71.39 Requirement for additional information.

The Commission may at any time require additional information in order to enable it to determine whether a license, certificate of compliance, or other approval should be granted, renewed, denied, modified, suspended, or revoked.

Subpart E--Package Approval Standards

§ 71.41 Demonstration of compliance.

(a) The effects on a package of the tests specified in § 71.71 ("Normal conditions of transport"),

and the tests specified in § 71.73 ("Hypothetical accident conditions"), and § 71.61 ("Special requirements for Type B packages containing more than 105 A2"), must be evaluated by 30

subjecting a specimen or scale model to a specific test, or by another method of demonstration acceptable to the Commission, as appropriate for the particular feature being considered.

(b) Taking into account the type of vehicle, the method of securing or attaching the package, and the controls to be exercised by the shipper, the Commission may permit the shipment to be evaluated together with the transporting vehicle.

(c) Environmental and test conditions different from those specified in §§ 71.71 and 71.73 may be approved by the Commission if the controls proposed to be exercised by the shipper are demonstrated to be adequate to provide equivalent safety of the shipment.

(d) Packages for which compliance with the other provisions of these regulations is impracticable shall not be transported except under special package authorization. Provided the applicant demonstrates that compliance with the other provisions of the regulations is impracticable and that the requisite standards of safety established by these regulations have been demonstrated through means alternative to the other provisions, a special package authorization may be approved for one-time shipments. The applicant shall demonstrate that the overall level of safety in transport for these shipments is at least equivalent to that which would be provided if all the applicable requirements had been met.

[60 FR 50264, Sept. 28, 1995 as amended at 69 FR 3794, Jan. 26, 2004]

§ 71.43 General standards for all packages.

(a) The smallest overall dimension of a package may not be less than 10 cm (4 in).

(b) The outside of a package must incorporate a feature, such as a seal, that is not readily breakable and that, while intact, would be evidence that the package has not been opened by unauthorized persons.

(c) Each package must include a containment system securely closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package.

(d) A package must be made of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the package contents, including possible reaction resulting from inleakage of water, to the maximum credible extent. Account must be taken of the behavior of materials under irradiation.

(e) A package valve or other device, the failure of which would allow radioactive contents to escape, must be protected against unauthorized operation and, except for a pressure relief device, must be provided with an enclosure to retain any leakage.

(f) A package must be designed, constructed, and prepared for shipment so that under the tests specified in § 71.71 ("Normal conditions of transport") there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no 31

substantial reduction in the effectiveness of the packaging.

(g) A package must be designed, constructed, and prepared for transport so that in still air at 38°C (100°F) and in the shade, no accessible surface of a package would have a temperature exceeding 50°C (122°F) in a nonexclusive use shipment, or 85°C (185°F) in an exclusive use shipment.

(h) A package may not incorporate a feature intended to allow continuous venting during transport.

§ 71.45 Lifting and tie-down standards for all packages.

(a) Any lifting attachment that is a structural part of a package must be designed with a minimum safety factor of three against yielding when used to lift the package in the intended manner, and it must be designed so that failure of any lifting device under excessive load would not impair the ability of the package to meet other requirements of this subpart. Any other structural part of the package that could be used to lift the package must be capable of being rendered inoperable for lifting the package during transport, or must be designed with strength equivalent to that required for lifting attachments.

(b) Tie-down devices:

(1) If there is a system of tie-down devices that is a structural part of the package, the system must be capable of withstanding, without generating stress in any material of the package in excess of its yield strength, a static force applied to the center of gravity of the package having a vertical component of 2 times the weight of the package with its contents, a horizontal component along the direction in which the vehicle travels of 10 times the weight of the package with its contents, and a horizontal component in the transverse direction of 5 times the weight of the package with its contents.

(2) Any other structural part of the package that could be used to tie down the package must be capable of being rendered inoperable for tying down the package during transport, or must be designed with strength equivalent to that required for tie-down devices.

(3) Each tie-down device that is a structural part of a package must be designed so that failure of the device under excessive load would not impair the ability of the package to meet other requirements of this part.

§ 71.47 External radiation standards for all packages.

(a) Except as provided in paragraph (b) of this section, each package of radioactive materials offered for transportation must be designed and prepared for shipment so that under conditions normally incident to transportation the radiation level does not exceed 2 mSv/h (200 mrem/h) at any point on the external surface of the package, and the transport index does not exceed 10.

(b) A package that exceeds the radiation level limits specified in paragraph (a) of this section 32

must be transported by exclusive use shipment only, and the radiation levels for such shipment must not exceed the following during transportation:

(1) 2 mSv/h (200 mrem/h) on the external surface of the package, unless the following conditions are met, in which case the limit is 10 mSv/h (1000 mrem/h):

(i) The shipment is made in a closed transport vehicle; (ii) The package is secured within the vehicle so that its position remains fixed during transportation; and (iii) There are no loading or unloading operations between the beginning and end of the transportation; (2) 2 mSv/h (200 mrem/h) at any point on the outer surface of the vehicle, including the top and underside of the vehicle; or in the case of a flat-bed style vehicle, at any point on the vertical planes projected from the outer edges of the vehicle, on the upper surface of the load or enclosure, if used, and on the lower external surface of the vehicle; and (3) 0.1 mSv/h (10 mrem/h) at any point 2 meters (80 in) from the outer lateral surfaces of the vehicle (excluding the top and underside of the vehicle); or in the case of a flat-bed style vehicle, at any point 2 meters (6.6 feet) from the vertical planes projected by the outer edges of the vehicle (excluding the top and underside of the vehicle); and (4) 0.02 mSv/h (2 mrem/h) in any normally occupied space, except that this provision does not apply to private carriers, if exposed personnel under their control wear radiation dosimetry devices in conformance with 10 CFR 20.1502.

(c) For shipments made under the provisions of paragraph (b) of this section, the shipper shall provide specific written instructions to the carrier for maintenance of the exclusive use shipment controls. The instructions must be included with the shipping paper information.

(d) The written instructions required for exclusive use shipments must be sufficient so that, when followed, they will cause the carrier to avoid actions that will unnecessarily delay delivery or unnecessarily result in increased radiation levels or radiation exposures to transport workers or members of the general public.

§ 71.51 Additional requirements for Type B packages.

(a) A Type B package, in addition to satisfying the requirements of §§ 71.41 through 71.47, must be designed, constructed, and prepared for shipment so that under the tests specified in:

(1) Section 71.71 ("Normal conditions of transport"), there would be no loss or dispersal of radioactive contents--as demonstrated to a sensitivity of 10-6 A2 per hour, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging; and 33

(2) Section 71.73 ("Hypothetical accident conditions"), there would be no escape of krypton-85 exceeding 10 A2 in 1 week, no escape of other radioactive material exceeding a total amount A2 in 1 week, and no external radiation dose rate exceeding 10 mSv/h (1 rem/h) at 1 m (40 in) from the external surface of the package.

(b) Where mixtures of different radionuclides are present, the provisions of appendix A, paragraph IV of this part shall apply, except that for Krypton-85, an effective A2 value equal to 10 A2 may be used.

(c) Compliance with the permitted activity release limits of paragraph (a) of this section may not depend on filters or on a mechanical cooling system.

(d) For packages which contain radioactive contents with activity greater than 105 A2, the requirements of § 71.61 must be met.

[60 FR 50264, Sept. 28, 1995 as amended at 69 FR 3794, Jan. 26, 2004]

§ 71.53 [Reserved]

[62 FR 5913, Feb. 10, 1997; 69 FR 3794, January 26, 2004]

§ 71.55 General requirements for fissile material packages.

(a) A package used for the shipment of fissile material must be designed and constructed in accordance with §§ 71.41 through 71.47. When required by the total amount of radioactive material, a package used for the shipment of fissile material must also be designed and constructed in accordance with § 71.51.

(b) Except as provided in paragraph (c) or (g) of this section, a package used for the shipment of fissile material must be so designed and constructed and its contents so limited that it would be subcritical if water were to leak into the containment system, or liquid contents were to leak out of the containment system so that, under the following conditions, maximum reactivity of the fissile material would be attained:

(1) The most reactive credible configuration consistent with the chemical and physical form of the material; (2) Moderation by water to the most reactive credible extent; and (3) Close full reflection of the containment system by water on all sides, or such greater reflection of the containment system as may additionally be provided by the surrounding material of the packaging.

(c) The Commission may approve exceptions to the requirements of paragraph (b) of this section if the package incorporates special design features that ensure that no single packaging error would permit leakage, and if appropriate measures are taken before each shipment to ensure that 34

the containment system does not leak.

(d) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in § 71.71 ("Normal conditions of transport")

(1) The contents would be subcritical; (2) The geometric form of the package contents would not be substantially altered; (3) There would be no leakage of water into the containment system unless, in the evaluation of undamaged packages under § 71.59(a)(1), it has been assumed that moderation is present to such an extent as to cause maximum reactivity consistent with the chemical and physical form of the material; and (4) There will be no substantial reduction in the effectiveness of the packaging, including:

(i) No more than 5 percent reduction in the total effective volume of the packaging on which nuclear safety is assessed; (ii) No more than 5 percent reduction in the effective spacing between the fissile contents and the outer surface of the packaging; and (iii) No occurrence of an aperture in the outer surface of the packaging large enough to permit the entry of a 10 cm (4 in) cube.

(e) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in § 71.73 ("Hypothetical accident conditions"), the package would be subcritical. For this determination, it must be assumed that:

(1) The fissile material is in the most reactive credible configuration consistent with the damaged condition of the package and the chemical and physical form of the contents; (2) Water moderation occurs to the most reactive credible extent consistent with the damaged condition of the package and the chemical and physical form of the contents; and (3) There is full reflection by water on all sides, as close as is consistent with the damaged condition of the package.

(f) For fissile material package designs to be transported by air:

(1) The package must be designed and constructed, and its contents limited so that it would be subcritical, assuming reflection by 20 cm (7.9 in) of water but no water inleakage, when subjected to sequential application of:

(i) The free drop test in § 71.73(c)(1);

35

(ii) The crush test in § 71.73(c)(2);

(iii) A puncture test, for packages of 250 kg or more, consisting of a free drop of the specimen through a distance of 3 m (120 in) in a position for which maximum damage is expected at the conclusion of the test sequence, onto the upper end of a solid, vertical, cylindrical, mild steel probe mounted on an essentially unyielding, horizontal surface. The probe must be 20 cm (7.9 in) in diameter, with the striking end forming the frustum of a right circular cone with the dimensions of 30 cm height, 2.5 cm top diameter, and a top edge rounded to a radius of not more than 6 mm (0.25 in). For packages less than 250 kg, the puncture test must be the same, except that a 250 kg probe must be dropped onto the specimen which must be placed on the surface; and (iv) The thermal test in § 71.73(c)(4), except that the duration of the test must be 60 minutes.

(2) The package must be designed and constructed, and its contents limited, so that it would be subcritical, assuming reflection by 20 cm (7.9 in) of water but no water inleakage, when subjected to an impact on an unyielding surface at a velocity of 90 m/s normal to the surface, at such orientation so as to result in maximum damage. A separate, undamaged specimen can be used for this evaluation.

(3) Allowance may not be made for the special design features in paragraph (c) of this section, unless water leakage into or out of void spaces is prevented following application of the tests in paragraphs (f)(1) and (f)(2) of this section, and subsequent application of the immersion test in § 71.73(c)(5).

(g) Packages containing uranium hexafluoride only are excepted from the requirements of paragraph (b) of this section provided that:

(1) Following the tests specified in § 71.73 ("Hypothetical accident conditions"), there is no physical contact between the valve body and any other component of the packaging, other than at its original point of attachment, and the valve remains leak tight; (2) There is an adequate quality control in the manufacture, maintenance, and repair of packagings; (3) Each package is tested to demonstrate closure before each shipment; and (4) The uranium is enriched to not more than 5 weight percent uranium-235.

[60 FR 50264, Sept. 28, 1995; 61 FR 28724, June 6, 1996; 69 FR 3794, Jan. 26, 2004]

§ 71.57 [Reserved]

§ 71.59 Standards for arrays of fissile material packages.

(a) A fissile material package must be controlled by either the shipper or the carrier during transport to assure that an array of such packages remains subcritical. To enable this control, the 36

designer of a fissile material package shall derive a number "N" based on all the following conditions being satisfied, assuming packages are stacked together in any arrangement and with close full reflection on all sides of the stack by water:

(1) Five times "N" undamaged packages with nothing between the packages would be subcritical; (2) Two times "N" damaged packages, if each package were subjected to the tests specified in § 71.73 ("Hypothetical accident conditions") would be subcritical with optimum interspersed hydrogenous moderation; and (3) The value of "N" cannot be less than 0.5.

(b) The CSI must be determined by dividing the number 50 by the value of "N" derived using the procedures specified in paragraph (a) of this section. The value of the CSI may be zero provided that an unlimited number of packages are subcritical, such that the value of "N" is effectively equal to infinity under the procedures specified in paragraph (a) of this section. Any CSI greater than zero must be rounded up to the first decimal place.

(c) For a fissile material package which is assigned a CSI value--

(1) Less than or equal to 50, that package may be shipped by a carrier in a nonexclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 50.

(2) Less than or equal to 50, that package may be shipped by a carrier in an exclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 100.

(3) Greater than 50, that package must be shipped by a carrier in an exclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 100.

[69 FR 3795, Jan. 26, 2004]

§ 71.61 Special requirements for Type B packages containing more than 105A2.

A Type B package containing more than 105A2 must be designed so that its undamaged containment system can withstand an external water pressure of 2 MPa (290 psi) for a period of not less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without collapse, buckling, or inleakage of water.

[69 FR 3795, Jan. 26, 2004]

§ 71.63 Special requirement for plutonium shipments.

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.

[69 FR 3795, Jan. 26, 2004]

37

§ 71.64 Special requirements for plutonium air shipments.

(a) A package for the shipment of plutonium by air subject to § 71.88(a)(4), in addition to satisfying the requirements of §§ 71.41 through 71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in --

(1) Section 71.74 ("Accident conditions for air transport of plutonium") --

(i) The containment vessel would not be ruptured in its post-tested condition, and the package must provide a sufficient degree of containment to restrict accumulated loss of plutonium contents to not more than an A2 quantity in a period of 1 week; (ii) The external radiation level would not exceed 10 mSv/h (1 rem/h) at a distance of 1 m (40 in) from the surface of the package in its post-tested condition in air; and (iii) A single package and an array of packages are demonstrated to be subcritical in accordance with this part, except that the damaged condition of the package must be considered to be that which results from the plutonium accident tests in § 71.74, rather than the hypothetical accident tests in § 71.73; and (2) Section 71.74(c), there would be no detectable leakage of water into the containment vessel of the package.

(b) With respect to the package requirements of paragraph (a), there must be a demonstration or analytical assessment showing that --

(1) The results of the physical testing for package qualification would not be adversely affected to a significant extent by --

(i) The presence, during the tests, of the actual contents that will be transported in the package; and (ii) Ambient water temperatures ranging from 0.6°C (+33°F) to 38°C (+100°F) for those qualification tests involving water, and ambient atmospheric temperatures ranging from -40°C (-

40°F) to +54°C (+130°F) for the other qualification tests.

(2) The ability of the package to meet the acceptance standards prescribed for the accident condition sequential tests would not be adversely affected if one or more tests in the sequence were deleted.

§ 71.65 Additional requirements.

The Commission may, by rule, regulation, or order, impose requirements on any licensee, in addition to those established in this part, as it deems necessary or appropriate to protect public health or to minimize danger to life or property.

38

Subpart F--Package, Special Form, and LSA-III Tests2

§ 71.70 Incorporations by reference.

(a) The materials listed in this section are incorporated by reference in the corresponding sections noted and made a part of the regulations in part 71. These incorporations by reference were approved by the Director of the Federal Register under 5 U.S.C. 552(a) and 1 CFR part 51.

These materials are incorporated as they exist on the date of the approval. A notice of any changes made to the material incorporated by reference will be published in the Federal Register, and the material must be available to the public. The materials can be examined, by appointment, at the NRC's Technical Library, which is located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-7000; email: Library.Resource@nrc.gov.

The materials are also available from the sources listed below. All approved material is available for inspection at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 1-202-741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.

(b) International Organization for Standardization, ISO Central Secretariat, Chemin de Blandonnet 8 CP 401, 1214 Vernier, Geneva, Switzerland; email: central@iso.org; phone: +41 22 749 01 11; Web site: http://www.iso.org.

(1) ISO 9978:1992(E), "Radiation protectionSealed radioactive sourcesLeakage test methods," First Edition (February 15, 1992), incorporation by reference approved for § 71.75(a),

is available for purchase from the American National Standards Institute, 25 West 43rd Street, 4th Floor, New York, NY 10036, 212-642-4900, http://www.ansi.org, or info@ansi.org.

(2) ISO 2919:1999(E), "Radiation protectionSealed radioactive sourcesGeneral requirements and classification," Second Edition (February 15, 1999), incorporation by reference approved for § 71.75(d), is available on http://www.amazon.com.

[80 FR 34013, Jun. 12, 2015; 80 FR 48684, Aug. 14, 2015]

§ 71.71 Normal conditions of transport.

(a) Evaluation. Evaluation of each package design under normal conditions of transport must include a determination of the effect on that design of the conditions and tests specified in this section. Separate specimens may be used for the free drop test, the compression test, and the penetration test, if each specimen is subjected to the water spray test before being subjected to any of the other tests.

(b) Initial conditions. With respect to the initial conditions for the tests in this section, the demonstration of compliance with the requirements of this part must be based on the ambient temperature preceding and following the tests remaining constant at that value between -29°C (-

20°F) and +38°C (+100°F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be considered to be the maximum normal operating pressure, unless a lower internal pressure consistent with the ambient 39

temperature considered to precede and follow the tests is more unfavorable.

(c) Conditions and tests.

(1) Heat. An ambient temperature of 38°C (100°F) in still air, and insolation according to the following table:

INSOLATION DATA Form and location of surface Total insolation for a 12-hour period (g cal/cm2)

Flat surfaces transported horizontally; Base None Other surfaces 800 Flat surfaces not transported horizontally 200 Curved surfaces 400 (2) Cold. An ambient temperature of -40°C (-40°F) in still air and shade.

(3) Reduced external pressure. An external pressure of 25 kPa (3.5 lbf/in2) absolute.

(4) Increased external pressure. An external pressure of 140 kPa (20 lbf/in2) absolute.

(5) Vibration. Vibration normally incident to transport.

(6) Water spray. A water spray that simulates exposure to rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(7) Free drop. Between 1.5 and 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the conclusion of the water spray test, a free drop through the distance specified below onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.

Criteria for Free Drop Test (Weight/Distance)

Package weight Free drop distance Kilograms (Pounds) Meters (Feet)

Less than 5,000 (Less than 11,000) 1.2 (4) 5,000 to 10,000 (11,000 to 22,000) 0.9 (3) 10,000 to 15,000 (22,000 to 33,100) 0.6 (2)

More than 15,000 (More than 33,100) 0.3 (1)

(8) Corner drop. A free drop onto each corner of the package in succession, or in the case of a 40

cylindrical package onto each quarter of each rim, from a height of 0.3 m (1 ft) onto a flat, essentially unyielding, horizontal surface. This test applies only to fiberboard, wood, or fissile material rectangular packages not exceeding 50 kg (110 lbs) and fiberboard, wood, or fissile material cylindrical packages not exceeding 100 kg (220 lbs).

(9) Compression. For packages weighing up to 5000 kg (11,000 lbs), the package must be subjected, for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to a compressive load applied uniformly to the top and bottom of the package in the position in which the package would normally be transported. The compressive load must be the greater of the following:

(i) The equivalent of 5 times the weight of the package; or (ii) The equivalent of 13 kPa (2 lbf/in2) multiplied by the vertically projected area of the package.

(10) Penetration. Impact of the hemispherical end of a vertical steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13 lbs) mass, dropped from a height of 1 m (40 in) onto the exposed surface of the package that is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface.

2 The package standards related to the tests in this subpart are contained in subpart E of this part.

[81 FR 86910, Dec. 2, 2016]

§ 71.73 Hypothetical accident conditions.

(a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the tests specified in this section, in the order indicated, to determine their cumulative effect on a package or array of packages. An undamaged specimen may be used for the water immersion tests specified in paragraph (c)(6) of this section.

(b) Test conditions. With respect to the initial conditions for the tests, except for the water immersion tests, to demonstrate compliance with the requirements of this part during testing, the ambient air temperature before and after the tests must remain constant at that value between -

29°C (-20°F) and +38°C (+100°F) which is most unfavorable for the feature under consideration.

The initial internal pressure within the containment system must be the maximum normal operating pressure, unless a lower internal pressure, consistent with the ambient temperature assumed to precede and follow the tests, is more unfavorable.

(c) Tests. Tests for hypothetical accident conditions must be conducted as follows:

(1) Free Drop. A free drop of the specimen through a distance of 9 m (30 ft) onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.

(2) Crush. Subjection of the specimen to a dynamic crush test by positioning the specimen on a 41

flat, essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 500-kg (1100-lb) mass from 9 m (30 ft) onto the specimen. The mass must consist of a solid mild steel plate 1 m (40 in) by 1 m (40 in) and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 500 kg (1100 lb), an overall density not greater than 1000 kg/m3 (62.4 lb/ft3) based on external dimension, and radioactive contents greater than 1000 A2 not as special form radioactive material. For packages containing fissile material, the radioactive contents greater than 1000 A2 criterion does not apply.

(3) Puncture. A free drop of the specimen through a distance of 1 m (40 in) in a position for which maximum damage is expected, onto the upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizontal surface. The bar must be 15 cm (6 in) in diameter, with the top horizontal and its edge rounded to a radius of not more than 6 mm (0.25 in), and of a length as to cause maximum damage to the package, but not less than 20 cm (8 in) long. The long axis of the bar must be vertical.

(4) Thermal. Exposure of the specimen fully engulfed, except for a simple support system, in a hydrocarbon fuel/air fire of sufficient extent, and in sufficiently quiescent ambient conditions, to provide an average emissivity coefficient of at least 0.9, with an average flame temperature of at least 800°C (1475°F) for a period of 30 minutes, or any other thermal test that provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 800°C. The fuel source must extend horizontally at least 1 m (40 in), but may not extend more than 3 m (10 ft), beyond any external surface of the specimen, and the specimen must be positioned 1 m (40 in) above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the convective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified. Artificial cooling may not be applied after cessation of external heat input, and any combustion of materials of construction, must be allowed to proceed until it terminates naturally.

(5) Immersion--fissile material. For fissile material subject to § 71.55, in those cases where water inleakage has not been assumed for criticality analysis, immersion under a head of water of at least 0.9 m (3 ft) in the attitude for which maximum leakage is expected.

(6) Immersion--all packages. A separate, undamaged specimen must be subjected to water pressure equivalent to immersion under a head of water of at least 15 m (50 ft). For test purposes, an external pressure of water of 150 kPa (21.7 lbf/in2) gauge is considered to meet these conditions.

[69 FR 3795, Jan. 26, 2004]

§ 71.74 Accident conditions for air transport of plutonium.

(a) Test conditions--Sequence of tests. A package must be physically tested to the following conditions in the order indicated to determine their cumulative effect.

42

(1) Impact at a velocity of not less than 129 m/sec (422 ft/sec) at a right angle onto a flat, essentially unyielding, horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage at the conclusion of the test sequence.

(2) A static compressive load of 31,800 kg (70,000 lbs) applied in the orientation expected to result in maximum damage at the conclusion of the test sequence. The force on the package must be developed between a flat steel surface and a 5 cm (2 in) wide, straight, solid, steel bar. The length of the bar must be at least as long as the diameter of the package, and the longitudinal axis of the bar must be parallel to the plane of the flat surface. The load must be applied to the bar in a manner that prevents any members or devices used to support the bar from contacting the package.

(3) Packages weighing less than 227 kg (500 lbs) must be placed on a flat, essentially unyielding, horizontal surface, and subjected to a weight of 227 kg (500 lbs) falling from a height of 3 m (10 ft) and striking in the position expected to result in maximum damage at the conclusion of the test sequence. The end of the weight contacting the package must be a solid probe made of mild steel. The probe must be the shape of the frustum of a right circular cone, 30 cm (12 in) long, 20 cm (8 in) in diameter at the base, and 2.5 cm (1 in) in diameter at the end. The longitudinal axis of the probe must be perpendicular to the horizontal surface. For packages weighing 227 kg (500 lbs) or more, the base of the probe must be placed on a flat, essentially unyielding horizontal surface, and the package dropped from a height of 3 m (10 ft) onto the probe, striking in the position expected to result in maximum damage at the conclusion of the test sequence.

(4) The package must be firmly restrained and supported such that its longitudinal axis is inclined approximately 45° to the horizontal. The area of the package that made first contact with the impact surface in paragraph (a)(1) of this section must be in the lowermost position. The package must be struck at approximately the center of its vertical projection by the end of a structural steel angle section falling from a height of at least 46 m (150 ft). The angle section must be at least 1.8 m (6 ft) in length with equal legs at least 13 cm (5 in) long and 1.3 cm (0.5 in) thick. The angle section must be guided in such a way as to fall end-on, without tumbling.

The package must be rotated approximately 90° about its longitudinal axis and struck by the steel angle section falling as before.

(5) The package must be exposed to luminous flames from a pool fire of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes. The luminous flames must extend an average of at least 0.9 m (3 ft) and no more than 3 m (10 ft) beyond the package in all horizontal directions. The position and orientation of the package in relation to the fuel must be that which is expected to result in maximum damage at the conclusion of the test sequence. An alternate method of thermal testing may be substituted for this fire test, provided that the alternate test is not of shorter duration and would not result in a lower heating rate to the package. At the conclusion of the thermal test, the package must be allowed to cool naturally or must be cooled by water sprinkling, whichever is expected to result in maximum damage at the conclusion of the test sequence.

(6) Immersion under at least 0.9 m (3 ft) of water.

43

(b) Individual free-fall impact test.

(1) An undamaged package must be physically subjected to an impact at a velocity not less than the calculated terminal free-fall velocity, at mean sea level, at a right angle onto a flat, essentially unyielding, horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage.

(2) This test is not required if the calculated terminal free-fall velocity of the package is less than 129 m/sec (422 ft/sec), or if a velocity not less than either 129 m/sec (422 ft/sec) or the calculated terminal free-fall velocity of the package is used in the sequential test of paragraph (a)(1) of this section.

(c) Individual deep submersion test. An undamaged package must be physically submerged and physically subjected to an external water pressure of at least 4 MPa (600 lbs/in2).

§ 71.75 Qualification of special form radioactive material.

(a) Special form radioactive materials must meet the test requirements of paragraph (b) of this section. Each solid radioactive material or capsule specimen to be tested must be manufactured or fabricated so that it is representative of the actual solid material or capsule that will be transported, with the proposed radioactive content duplicated as closely as practicable. Any differences between the material to be transported and the test material, such as the use of non-radioactive contents, must be taken into account in determining whether the test requirements have been met. In addition:

(1) A different specimen may be used for each of the tests; (2) The specimen may not break or shatter when subjected to the impact, percussion, or bending tests; (3) The specimen may not melt or disperse when subjected to the heat test; (4) After each test, leaktightness or indispersibility of the specimen must be determined by a method no less sensitive than the leaching assessment procedure prescribed in paragraph (c) of this section. For a capsule resistant to corrosion by water, and which has an internal void volume greater than 0.1 milliliter, an alternative to the leaching assessment is a demonstration of leaktightness of x10-4 torr-liter/s (1.3xx10-4 atm-cm3/s) based on air at 25°C (77°F) and one atmosphere differential pressure for solid radioactive content, or x10-6 torr-liter/s (1.30xx10-6 atm-cm3/s) for liquid or gaseous radioactive content; and (5) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to the leaktightness procedure specified in this section, provided it is alternatively subjected to any of the tests prescribed in ISO/TR4826-1979(E), "Sealed radioactive sources leak test methods" which is available from the American National Standards Institute, 1430 Broadway, New York, N.Y. 10018.

44

(b) Test methods. (1) Impact Test. The specimen must fall onto the target from a height of 9 m (30 ft) or greater in the orientation expected to result in maximum damage. The target must be a flat, horizontal surface of such mass and rigidity that any increase in its resistance to displacement or deformation, on impact by the specimen, would not significantly increase the damage to the specimen.

(2) Percussion Test. (i) The specimen must be placed on a sheet of lead that is supported by a smooth solid surface, and struck by the flat face of a steel billet so as to produce an impact equivalent to that resulting from a free drop of 1.4 kg (3 lbs) through 1 m (40 in);

(ii) The flat face of the billet must be 25 millimeters (mm) (1 inch) in diameter with the edges rounded off to a radius of 3 mm+/-0.3 mm(.12 in+/-0.012 in);

(iii) The lead must be hardness number 3.5 to 4.5 on the Vickers scale and thickness 25 mm (1 in) or greater, and must cover an area greater than that covered by the specimen; (iv) A fresh surface of lead must be used for each impact; and (v) The billet must strike the specimen so as to cause maximum damage.

(3) Bending test. (i) This test applies only to long, slender sources with a length of 10 cm (4 inches) or greater and a length to width ratio of 10 or greater; (ii) The specimen must be rigidly clamped in a horizontal position so that one half of its length protrudes from the face of the clamp; (iii) The orientation of the specimen must be such that the specimen will suffer maximum damage when its free end is struck by the flat face of a steel billet; (iv) The billet must strike the specimen so as to produce an impact equivalent to that resulting from a free vertical drop of 1.4 kg (3 lbs) through 1 m (40 in); and (v) The flat face of the billet must be 25 mm (1 inch) in diameter with the edges rounded off to a radius of 3 mm+/-0.3 mm (.12 in+/-0.012 in).

(4) Heat test. The specimen must be heated in air to a temperature of not less than 800°C (1475°F), held at that temperature for a period of 10 minutes, and then allowed to cool.

(c) Leaching assessment methods. (1) For indispersible solid material --

(i) The specimen must be immersed for 7 days in water at ambient temperature. The water must have a pH of 6-8 and a maximum conductivity of 10 micromho per centimeter at 20° (68°F);

(ii) The water with specimen must then be heated to a temperature of 50°C+/-5°C (122°F+/-9°F) and maintained at this temperature for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

45

(iii) The activity of the water must then be determined; (iv) The specimen must then be stored for at least 7 days in still air of relative humidity not less than 90 percent at 30°C (86°F);

(v) The specimen must then be immersed in water under the same conditions as in paragraph (c)(1)(i) of this section, and the water with specimen must be heated to 50°C+/-5°C (122°F+/-9°F) and maintained at that temperature for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; (vi) The activity of the water must then be determined. The sum of the activities determined here and in paragraph (c)(1)(iii) of this section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie (µCi)).

(2) For encapsulated material --

(i) The specimen must be immersed in water at ambient temperature. The water must have a pH of 6-8 and a maximum conductivity of 10 micromho per centimeter; (ii) The water and specimen must be heated to a temperature of 50°C+/-5°C (122°F+/-9°F) and maintained at this temperature for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; (iii) The activity of the water must then be determined; (iv) The specimen must then be stored for at least 7 days in still air at a temperature of 30°C (86°F) or greater; (v) The process in paragraph (c)(2)(i), (ii), and (iii) of this section must be repeated; and (vi) The activity of the water must then be determined. The sum of the activities determined here and in paragraph (c)(2)(iii) of this section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie (Ci)).

(d) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to --

(1) The impact test and the percussion test of this section, provided that the specimen is alternatively subjected to the Class 4 impact test prescribed in ISO 2919-1980(e), "Sealed Radioactive Sources Classification" (see § 71.75(a)(5) for statement of availability); and (2) The heat test of this section, provided the specimen is alternatively subjected to the Class 6 temperature test specified in the International Organization for Standardization document ISO 2919-1980(e), "Sealed Radioactive Sources Classification."

[80 FR 34013, Jun. 12, 2015]

§ 71.77 Qualification of LSA-III Material 46

(a) LSA-III material must meet the test requirements of paragraph (b) of this section. Any differences between the specimen to be tested and the material to be transported must be taken into account in determining whether the test requirements have been met.

(b) Leaching Test. (1) The specimen, representing no less than the entire contents of the package, must be immersed for 7 days in water at ambient temperature; (2) The volume of water to be used in the test must be sufficient to ensure that at the end of the test period the free volume of the unabsorbed and unreacted water remaining will be at least 10%

of the volume of the specimen itself; (3) The water must have an initial pH of 6-8 and a maximum conductivity 10 micromho/cm at 20°C (68°F); and (4) The total activity of the free volume of water must be measured following the 7 day immersion test and must not exceed 0.1 A2.

Subpart G--Operating Controls and Procedures

§ 71.81 Applicability of operating controls and procedures.

A licensee subject to this part, who, under a general or specific license, transports licensed material or delivers licensed material to a carrier for transport, shall comply with the requirements of this subpart G, with the quality assurance requirements of subpart H of this part, and with the general provisions of subpart A of this part.

§ 71.83 Assumptions as to unknown properties.

When the isotopic abundance, mass, concentration, degree of irradiation, degree of moderation, or other pertinent property of fissile material in any package is not known, the licensee shall package the fissile material as if the unknown properties have credible values that will cause the maximum neutron multiplication.

§ 71.85 Preliminary determinations.

Before the first use of any packaging for the shipment of licensed material --

(a) The licensee shall ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could significantly reduce the effectiveness of the packaging; (b) Where the maximum normal operating pressure will exceed 35 kPa (5 lbf/in2) gauge, the licensee shall test the containment system at an internal pressure at least 50 percent higher than the maximum normal operating pressure, to verify the capability of that system to maintain its structural integrity at that pressure; and (c) The licensee shall conspicuously and durably mark the packaging with its model number, 47

serial number, gross weight, and a package identification number assigned by NRC. Before applying the model number, the licensee shall determine that the packaging has been fabricated in accordance with the design approved by the Commission.

(d) The licensee shall ascertain that the determinations in paragraphs (a) through (c) of this section have been made.

[80 FR 34013, Jun. 12, 2015]

§ 71.87 Routine determinations.

Before each shipment of licensed material, the licensee shall ensure that the package with its contents satisfies the applicable requirements of this part and of the license. The licensee shall determine that --

(a) The package is proper for the contents to be shipped; (b) The package is in unimpaired physical condition except for superficial defects such as marks or dents; (c) Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects; (d) Any system for containing liquid is adequately sealed and has adequate space or other specified provision for expansion of the liquid; (e) Any pressure relief device is operable and set in accordance with written procedures; (f) The package has been loaded and closed in accordance with written procedures; (g) For fissile material, any moderator or neutron absorber, if required, is present and in proper condition; (h) Any structural part of the package that could be used to lift or tie down the package during transport is rendered inoperable for that purpose, unless it satisfies the design requirements of § 71.45; (i) The level of non-fixed (removable) radioactive contamination on the external surfaces of each package offered for shipment is as low as reasonably achievable, and within the limits specified in DOT regulations in 49 CFR 173.443; (j) External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in § 71.47 at any time during transportation; and (k) Accessible package surface temperatures will not exceed the limits specified in § 71.43(g) at any time during transportation.

48

§ 71.88 Air transport of plutonium.

(a) Notwithstanding the provisions of any general licenses and notwithstanding any exemptions stated directly in this part or included indirectly by citation of 49 CFR chapter I, as may be applicable, the licensee shall assure that plutonium in any form, whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for air transport unless:

(1) The plutonium is contained in a medical device designed for individual human application; or (2) The plutonium is contained in a material in which the specific activity is less than or equal to the activity concentration values for plutonium specified in Appendix A, Table A-2, of this part, and in which the radioactivity is essentially uniformly distributed; or (3) The plutonium is shipped in a single package containing no more than an A2 quantity of plutonium in any isotope or form, and is shipped in accordance with § 71.5; or (4) The plutonium is shipped in a package specifically authorized for the shipment of plutonium by air in the Certificate of Compliance for that package issued by the Commission.

(b) Nothing in paragraph (a) of this section is to be interpreted as removing or diminishing the requirements of § 73.24 of this chapter.

(c) For a shipment of plutonium by air which is subject to paragraph (a)(4) of this section, the licensee shall, through special arrangement with the carrier, require compliance with 49 CFR 175.704, U.S. Department of Transportation regulations applicable to the air transport of plutonium.

[69 FR 3795, Jan. 26, 2004]

§ 71.89 Opening instructions.

Before delivery of a package to a carrier for transport, the licensee shall ensure that any special instructions needed to safely open the package have been sent to, or otherwise made available to, the consignee for the consignee's use in accordance with 10 CFR 20.1906(e).

§ 71.91 Records.

(a) Each licensee shall maintain, for a period of 3 years after shipment, a record of each shipment of licensed material not exempt under § 71.14, showing where applicable --

(1) Identification of the packaging by model number and serial number; (2) Verification that there are no significant defects in the packaging, as shipped; (3) Volume and identification of coolant; 49

(4) Type and quantity of licensed material in each package, and the total quantity of each shipment; (5) For each item of irradiated fissile material --

(i) Identification by model number and serial number; (ii) Irradiation and decay history to the extent appropriate to demonstrate that its nuclear and thermal characteristics comply with license conditions; and (iii) Any abnormal or unusual condition relevant to radiation safety; (6) Date of the shipment; (7) For fissile packages and for Type B packages, any special controls exercised; (8) Name and address of the transferee; (9) Address to which the shipment was made; and (10) Results of the determinations required by § 71.87 and by the conditions of the package approval.

(b) Each certificate holder shall maintain, for a period of 3 years after the life of the packaging to which they apply, records identifying the packaging by model number, serial number, and date of manufacture.

(c) The licensee, certificate holder, and an applicant for a CoC, shall make available to the Commission for inspection, upon reasonable notice, all records required by this part. Records are only valid if stamped, initialed, or signed and dated by authorized personnel, or otherwise authenticated.

(d) The licensee, certificate holder, and an applicant for a CoC shall maintain sufficient written records to furnish evidence of the quality of packaging. The records to be maintained include results of the determinations required by § 71.85; design, fabrication, and assembly records; results of reviews, inspections, tests, and audits; results of monitoring work performance and materials analyses; and results of maintenance, modification, and repair activities. Inspection, test, and audit records must identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted. These records must be retained for 3 years after the life of the packaging to which they apply.

[69 FR 3796, Jan. 26, 2004; 80 FR 34013, Jun. 12, 2015]

§ 71.93 Inspection and tests.

(a) The licensee, certificate holder, and applicant for a CoC shall permit the Commission, at all 50

reasonable times, to inspect the licensed material, packaging, premises, and facilities in which the licensed material or packaging is used, provided, constructed, fabricated, tested, stored, or shipped.

(b) The licensee, certificate holder, and applicant for a CoC shall perform, and permit the Commission to perform, any tests the Commission deems necessary or appropriate for the administration of the regulations in this chapter.

(c) The certificate holder and applicant for a CoC shall notify the NRC, in accordance with § 71.1, 45 days in advance of starting fabrication of the first packaging under a CoC. This paragraph applies to any packaging used for the shipment of licensed material which has either--

(1) A decay heat load in excess of 5 kW; or (2) A maximum normal operating pressure in excess of 103 kPa (15 lbf/in2) gauge.

[69 FR 3796, Jan. 26, 2004]

§ 71.95 Reports.

(a) The licensee, after requesting the certificate holder's input, shall submit a written report to the Commission of--

(1) Instances in which there is a significant reduction in the effectiveness of any NRC-approved Type B or Type AF packaging during use; or (2) Details of any defects with safety significance in any NRC-approved Type B or fissile material packaging, after first use.

(3) Instances in which the conditions of approval in the Certificate of Compliance were not observed in making a shipment.

(b) The licensee shall submit a written report to the Commission of instances in which the conditions in the certificate of compliance were not followed during a shipment.

(c) Each licensee shall submit, in accordance with § 71.1, a written report required by paragraph (a) or (b) of this section within 60 days of the event or discovery of the event. The licensee shall also provide a copy of each report submitted to the NRC to the applicable certificate holder.

Written reports prepared under other regulations may be submitted to fulfill this requirement if the reports contain all the necessary information, and the appropriate distribution is made. Using an appropriate method listed in § 71.1(a), the licensee shall report to: ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards. These written reports must include the following:

(1) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed to the event and significant corrective action taken or planned 51

to prevent recurrence.

(2) A clear, specific, narrative description of the event that occurred so that knowledgeable readers conversant with the requirements of part 71, but not familiar with the design of the packaging, can understand the complete event. The narrative description must include the following specific information as appropriate for the particular event.

(i) Status of components or systems that were inoperable at the start of the event and that contributed to the event; (ii) Dates and approximate times of occurrences; (iii) The cause of each component or system failure or personnel error, if known; (iv) The failure mode, mechanism, and effect of each failed component, if known; (v) A list of systems or secondary functions that were also affected for failures of components with multiple functions; (vi) The method of discovery of each component or system failure or procedural error; (vii) For each human performance-related root cause, a discussion of the cause(s) and circumstances; (viii) The manufacturer and model number (or other identification) of each component that failed during the event; and (ix) For events occurring during use of a packaging, the quantities and chemical and physical form(s) of the package contents.

(3) An assessment of the safety consequences and implications of the event. This assessment must include the availability of other systems or components that could have performed the same function as the components and systems that failed during the event.

(4) A description of any corrective actions planned as a result of the event, including the means employed to repair any defects, and actions taken to reduce the probability of similar events occurring in the future.

(5) Reference to any previous similar events involving the same packaging that are known to the licensee or certificate holder.

(6) The name and telephone number of a person within the licensee's organization who is knowledgeable about the event and can provide additional information.

(7) The extent of exposure of individuals to radiation or to radioactive materials without identification of individuals by name.

52

(d) Report legibility. The reports submitted by licensees and/or certificate holders under this section must be of sufficient quality to permit reproduction and micrographic processing.

[60 FR 50264, Sept. 28, 1995, as amended at 67 FR 3585, Jan. 25, 2002; 68 FR 58818, Oct. 10, 2003; 69 FR 3796, Jan. 26, 2004; 75 FR 73945, Nov. 30, 2010; 79 FR 75741, Dec. 19, 2014; 84 FR 65639, Nov. 29, 2019; 84 FR 66561, Dec. 5, 2019]

§ 71.97 Advance notification of shipment of irradiated reactor fuel and nuclear waste.

(a)(1) As specified in paragraphs (b), (c) and (d) of this section, each licensee shall provide advance notification to the governor of a State, or the governor's designee, of the shipment of licensed material, within or across the boundary of the State, before the transport, or delivery to a carrier, for transport, of licensed material outside the confines of the licensee's plant or other place of use or storage.

(2) As specified in paragraphs (b), (c), and (d) of this section, after June 11, 2013, each licensee shall provide advance notification to the Tribal official of participating Tribes referenced in paragraph (c)(3)(iii) of this section, or the officials designee, of the shipment of licensed material, within or across the boundary of the Tribes reservation, before the transport, of delivery to a carrier, for transport, of licensed material outside the confines of the licensees plant or other place of use or storage.

(b) Advance notification is also required under this section for the shipment of licensed material, other than irradiated fuel, meeting the following three conditions:

(1) The licensed material is required by this part to be in Type B packaging for transportation; (2) The licensed material is being transported to or across a State boundary en route to a disposal facility or to a collection point for transport to a disposal facility; and (3) The quantity of licensed material in a single package exceeds the least of the following:

(i) 3000 times the A1 value of the radionuclides as specified in appendix A, Table A-1 for special form radioactive material; (ii) 3000 times the A2 value of the radionuclides as specified in appendix A, Table A-1 for normal form radioactive material; or (iii) 1000 TBq (27,000 Ci).

(c) Procedures for submitting advance notification.

(1) The notification must be made in writing to:

(i) The office of each appropriate governor or governor's designee; 53

(ii) The office of each appropriate Tribal official of Tribal officials designee; and (iii) The Director, Office of Nuclear Security and Incident Response.

(2) A notification delivered by mail must be postmarked at least 7 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.

(3) A notification delivered by any other means than mail must reach the office of the governor or of the governor's designee or the Tribal official or Tribal officials designee at least 4 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.

(i) A list of the names and mailing addresses of the governors' designees receiving advance notification of transportation of nuclear waste was published in the Federal Register on June 30, 1995 (60 FR 34306).

(ii) Contact information for each State, including telephone and mailing addresses of governors and governors designees, and participating Tribes, including telephone and mailing addresses of Tribal officials and Tribal officials designees, is available on the NRC Web site at:

https://scp.nrc.gov/special/designee.pdf.

(iii) A list of the names and mailing addresses of the governors' designees and Tribal officials designees of participating Tribes is available on request from the Director, Division of Materials Safety, Security, State, and Tribal Programs, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

(4) The licensee shall retain a copy of the notification as a record for 3 years.

(d) Information to be furnished in advance notification of shipment. Each advance notification of shipment of irradiated reactor fuel or nuclear waste must contain the following information:

(1) The name, address, and telephone number of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the regulations of DOT in 49 CFR 172.202 and 172.203(d);

(3) The point of origin of the shipment and the 7-day period during which departure of the shipment is estimated to occur; (4) The 7-day period during which arrival of the shipment at State boundaries or Tribal reservation boundaries is estimated to occur; (5) The destination of the shipment, and the 7-day period during which arrival of the shipment is estimated to occur; and 54

(6) A point of contact, with a telephone number, for current shipment information.

(e) Revision notice. A licensee who finds that schedule information previously furnished to a governor or governor's designee or a Tribal official of Tribal officials designee, in accordance with this section, will not be met, shall telephone a responsible individual in the office of the governor of the State or of the governor's designee or the Tribal official or the Tribal officials designee and inform that individual of the extent of the delay beyond the schedule originally reported. The licensee shall maintain a record of the name of the individual contacted for 3 years.

(f) Cancellation notice. (1) Each licensee who cancels an irradiated reactor fuel or nuclear waste shipment for which advance notification has been sent shall send a cancellation notice to the governor of each State or to the governor's designee previously notified, each Tribal official or to the Tribal officials designee previously notified, and to the Director, Office of Nuclear Security and Incident Response.

(2) The licensee shall state in the notice that it is a cancellation and identify the advance notification that is being canceled. The licensee shall retain a copy of the notice as a record for 3 years.

[60 FR 50264, Sept. 28, 1995, as amended at 67 FR 3586, Jan. 25, 2002; 68 FR 58818, Oct. 10, 2003; 74 FR 62683, Dec. 1, 2009; 75 FR 73945, Nov. 30, 2010; 77 FR 34204, Jun. 11, 2012; 78 FR 17021, Mar. 19, 2013; 79 FR 75741, Dec. 19, 2014; 80 FR 74981, Dec. 1, 2015; 83 FR 30285, Jun. 28, 2018; 83 FR 57231, Nov. 21, 2018]

§ 71.99 Violations.

(a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of --

(1) The Atomic Energy Act of 1954, as amended; (2) Title II of the Energy Reorganization Act of 1974, as amended; or (3) A regulation or order issued pursuant to those Acts.

(b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:

(1) For violations of --

(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended; (ii) Section 206 of the Energy Reorganization Act; (iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section; or 55

(iv) Any term , condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.

(2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.

§ 71.100 Criminal penalties.

(a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 71 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.

(b) The regulations in part 71 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 71.0, 71.2, 71.4, 71.6, 71.7, 71.10, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.40, 71.41, 71.43, 71.45, 71.47, 71.51, 71.55, 71.59, 71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, and 71.100.

[69 FR 3796, Jan. 26, 2004]

Subpart H--Quality Assurance Source: 69 FR 3797, Jan. 26, 2004, unless otherwise noted.

§ 71.101 Quality assurance requirements.

(a) Purpose. This subpart describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging that are important to safety. As used in this subpart, quality assurance comprises all those planned and systematic actions necessary to provide adequate confidence that a system or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to control of the physical characteristics and quality of the material or component to predetermined requirements. Each certificate holder and applicant for a package approval is responsible for satisfying the quality assurance requirements that apply to design, fabrication, testing, and modification of packaging subject to this subpart. Each licensee is responsible for satisfying the quality assurance requirements that apply to its use of a packaging for the shipment of licensed material subject to this subpart.

(b) Establishment of program. Each licensee, certificate holder, and applicant for a CoC shall establish, maintain, and execute a quality assurance program satisfying each of the applicable criteria of §§ 71.101 through 71.137 and satisfying any specific provisions that are applicable to the licensee's activities including procurement of packaging. The licensee, certificate holder, and applicant for a CoC shall execute the applicable criteria in a graded approach to an extent that is commensurate with the quality assurance requirement's importance to safety.

56

(c) Approval of program. (1) Before the use of any package for the shipment of licensed material subject to this subpart, each licensee shall obtain Commission approval of its quality assurance program. Using an appropriate method listed in § 71.1(a), each licensee shall file a description of its quality assurance program, including a discussion of which requirements of this subpart are applicable and how they will be satisfied, by submitting the description to: ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards.

(2) Before the fabrication, testing, or modification of any package for the shipment of licensed material subject to this subpart, each licensee, certificate holder, or applicant for a CoC shall obtain Commission approval of its quality assurance program. Each certificate holder or applicant for a CoC shall, in accordance with § 71.1, file a description of its quality assurance program, including a discussion of which requirements of this subpart are applicable and how they will be satisfied.

(d) Existing package designs. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979, and which have been designed in accordance with the provisions of this part in effect at the time of application for package approval. Those packages will be accepted as having been designed in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section.

(e) Existing packages. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979, have been at least partially fabricated before that date, and for which the fabrication is in accordance with the provisions of this part in effect at the time of application for approval of package design. These packages will be accepted as having been fabricated and assembled in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section.

(f) Previously approved programs. A Commission-approved quality assurance program that satisfies the applicable criteria of subpart H of this part, Appendix B of part 50 of this chapter, or subpart G of part 72 of this chapter, and that is established, maintained, and executed regarding transport packages, will be accepted as satisfying the requirements of paragraph (b) of this section. Before first use, the licensee, certificate holder, and applicant for a CoC shall notify the NRC, in accordance with § 71.1, of its intent to apply its previously approved subpart H, Appendix B, or subpart G quality assurance program to transportation activities. The licensee, certificate holder, and applicant for a CoC shall identify the program by date of submittal to the Commission, Docket Number, and date of Commission approval.

(g) Radiography containers. A program for transport container inspection and maintenance limited to radiographic exposure devices, source changers, or packages transporting these devices and meeting the requirements of § 34.31(b) of this chapter or equivalent Agreement State requirement, is deemed to satisfy the requirements of §§ 71.17(b) and 71.101(b).

[75 FR 73945, Nov. 30, 2010; 79 FR 75741, Dec. 19, 2014; 80 FR 34013, Jun. 12, 2015; 84 FR 65639, Nov. 29, 2019; 84 FR 66561, Dec. 5, 2019]

57

§ 71.103 Quality assurance organization.

(a) The licensee,2 certificate holder, and applicant for a Certificate of Compliance shall be responsible for the establishment and execution of the quality assurance program. The licensee, certificate holder, and applicant for a Certificate of Compliance may delegate to others, such as contractors, agents, or consultants, the work of establishing and executing the quality assurance program, or any part of the quality assurance program, but shall retain responsibility for the program. These activities include performing the functions associated with attaining quality objectives and the quality assurance functions.

(b) The quality assurance functions are--

(1) Assuring that an appropriate quality assurance program is established and effectively executed; and (2) Verifying, by procedures such as checking, auditing, and inspection, that activities affecting the functions that are important to safety have been correctly performed.

(c) The persons and organizations performing quality assurance functions must have sufficient authority and organizational freedom to--

(1) Identify quality problems; (2) Initiate, recommend, or provide solutions; and (3) Verify implementation of solutions.

(d) The persons and organizations performing quality assurance functions shall report to a management level that assures that the required authority and organizational freedom, including sufficient independence from cost and schedule, when opposed to safety considerations, are provided.

(e) Because of the many variables involved, such as the number of personnel, the type of activity being performed, and the location or locations where activities are performed, the organizational structure for executing the quality assurance program may take various forms, provided that the persons and organizations assigned the quality assurance functions have the required authority and organizational freedom.

(f) Irrespective of the organizational structure, the individual(s) assigned the responsibility for assuring effective execution of any portion of the quality assurance program, at any location where activities subject to this section are being performed, must have direct access to the levels of management necessary to perform this function.

2 While the term "licensee" is used in these criteria, the requirements are applicable to whatever design, fabrication, assembly, and testing of the package is accomplished with respect to a package before the time a package approval is issued.

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[80 FR 34014, Jun. 12, 2015]

§ 71.105 Quality assurance program.

(a) The licensee, certificate holder, and applicant for a CoC shall establish, at the earliest practicable time consistent with the schedule for accomplishing the activities, a quality assurance program that complies with the requirements of §§ 71.101 through 71.137. The licensee, certificate holder, and applicant for a CoC shall document the quality assurance program by written procedures or instructions and shall carry out the program in accordance with those procedures throughout the period during which the packaging is used. The licensee, certificate holder, and applicant for a CoC shall identify the material and components to be covered by the quality assurance program, the major organizations participating in the program, and the designated functions of these organizations.

(b) The licensee, certificate holder, and applicant for a CoC, through its quality assurance program, shall provide control over activities affecting the quality of the identified materials and components to an extent consistent with their importance to safety, and as necessary to assure conformance to the approved design of each individual package used for the shipment of radioactive material. The licensee, certificate holder, and applicant for a CoC shall assure that activities affecting quality are accomplished under suitably controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanliness; and assurance that all prerequisites for the given activity have been satisfied. The licensee, certificate holder, and applicant for a CoC shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test.

(c) The licensee, certificate holder, and applicant for a CoC shall base the requirements and procedures of its quality assurance program on the following considerations concerning the complexity and proposed use of the package and its components:

(1) The impact of malfunction or failure of the item to safety; (2) The design and fabrication complexity or uniqueness of the item; (3) The need for special controls and surveillance over processes and equipment; (4) The degree to which functional compliance can be demonstrated by inspection or test; and (5) The quality history and degree of standardization of the item.

(d) The licensee, certificate holder, and applicant for a CoC shall provide for indoctrination and training of personnel performing activities affecting quality, as necessary to assure that suitable proficiency is achieved and maintained. The licensee, certificate holder, and applicant for a CoC shall review the status and adequacy of the quality assurance program at established intervals.

Management of other organizations participating in the quality assurance program shall review regularly the status and adequacy of that part of the quality assurance program they are 59

executing.

§ 71.106 Changes to quality assurance program.

(a) Each quality assurance program approval holder shall submit, in accordance with § 71.1(a), a description of a proposed change to its NRC-approved quality assurance program that will reduce commitments in the program description as approved by the NRC. The quality assurance program approval holder shall not implement the change before receiving NRC approval.

(1) The description of a proposed change to the NRC-approved quality assurance program must identify the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the applicable requirements of subpart H of this part.

(2) [Reserved]

(b) Each quality assurance program approval holder may change a previously approved quality assurance program without prior NRC approval, if the change does not reduce the commitments in the quality assurance program previously approved by the NRC. Changes to the quality assurance program that do not reduce the commitments shall be submitted to the NRC every 24 months, in accordance with § 71.1(a). In addition to quality assurance program changes involving administrative improvements and clarifications, spelling corrections, and non-substantive changes to punctuation or editorial items, the following changes are not considered reductions in commitment:

(1) The use of a quality assurance standard approved by the NRC that is more recent than the quality assurance standard in the certificate holders or applicants current quality assurance program at the time of the change; (2) The use of generic organizational position titles that clearly denote the position function, supplemented as necessary by descriptive text, rather than specific titles, provided that there is no substantive change to either the functions of the position or reporting responsibilities; (3) The use of generic organizational charts to indicate functional relationships, authorities, and responsibilities, or alternatively, the use of descriptive text, provided that there is no substantive change to the functional relationships, authorities, or responsibilities; (4) The elimination of quality assurance program information that duplicates language in quality assurance regulatory guides and quality assurance standards to which the quality assurance program approval holder has committed to on record; and (5) Organizational revisions that ensure that persons and organizations performing quality assurance functions continue to have the requisite authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations.

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(c) Each quality assurance program approval holder shall maintain records of quality assurance program changes.

[80 FR 34014, Jun. 12, 2015]

§ 71.107 Package design control.

(a) The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that applicable regulatory requirements and the package design, as specified in the license or CoC for those materials and components to which this section applies, are correctly translated into specifications, drawings, procedures, and instructions. These measures must include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from standards are controlled. Measures must be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the functions of the materials, parts, and components of the packaging that are important to safety.

(b) The licensee, certificate holder, and applicant for a CoC shall establish measures for the identification and control of design interfaces and for coordination among participating design organizations. These measures must include the establishment of written procedures, among participating design organizations, for the review, approval, release, distribution, and revision of documents involving design interfaces. The design control measures must provide for verifying or checking the adequacy of design, by methods such as design reviews, alternate or simplified calculational methods, or by a suitable testing program. For the verifying or checking process, the licensee shall designate individuals or groups other than those who were responsible for the original design, but who may be from the same organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, the licensee, certificate holder, and applicant for a CoC shall include suitable qualification testing of a prototype or sample unit under the most adverse design conditions. The licensee, certificate holder, and applicant for a CoC shall apply design control measures to the following:

(1) Criticality physics, radiation shielding, stress, thermal, hydraulic, and accident analyses; (2) Compatibility of materials; (3) Accessibility for inservice inspection, maintenance, and repair; (4) Features to facilitate decontamination; and (5) Delineation of acceptance criteria for inspections and tests.

(c) The licensee, certificate holder, and applicant for a CoC shall subject design changes, including field changes, to design control measures commensurate with those applied to the original design. Changes in the conditions specified in the CoC require prior NRC approval.

§ 71.109 Procurement document control.

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The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that adequate quality is required in the documents for procurement of material, equipment, and services, whether purchased by the licensee, certificate holder, and applicant for a CoC or by its contractors or subcontractors. To the extent necessary, the licensee, certificate holder, and applicant for a CoC shall require contractors or subcontractors to provide a quality assurance program consistent with the applicable provisions of this part.

§ 71.111 Instructions, procedures, and drawings.

The licensee, certificate holder, and applicant for a CoC shall prescribe activities affecting quality by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall require that these instructions, procedures, and drawings be followed.

The instructions, procedures, and drawings must include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

§ 71.113 Document control.

The licensee, certificate holder, and applicant for a CoC shall establish measures to control the issuance of documents such as instructions, procedures, and drawings, including changes, that prescribe all activities affecting quality. These measures must assure that documents, including changes, are reviewed for adequacy, approved for release by authorized personnel, and distributed and used at the location where the prescribed activity is performed.

§ 71.115 Control of purchased material, equipment, and services.

(a) The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents. These measures must include provisions, as appropriate, for source evaluation and selection, objective evidence of quality furnished by the contractor or subcontractor, inspection at the contractor or subcontractor source, and examination of products on delivery.

(b) The licensee, certificate holder, and applicant for a CoC shall have available documentary evidence that material and equipment conform to the procurement specifications before installation or use of the material and equipment. The licensee, certificate holder, and applicant for a CoC shall retain, or have available, this documentary evidence for the life of the package to which it applies. The licensee, certificate holder, and applicant for a CoC shall assure that the evidence is sufficient to identify the specific requirements met by the purchased material and equipment.

(c) The licensee, certificate holder, and applicant for a CoC shall assess the effectiveness of the control of quality by contractors and subcontractors at intervals consistent with the importance, complexity, and quantity of the product or services.

§ 71.117 Identification and control of materials, parts, and components.

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The licensee, certificate holder, and applicant for a CoC shall establish measures for the identification and control of materials, parts, and components. These measures must assure that identification of the item is maintained by heat number, part number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, installation, and use of the item. These identification and control measures must be designed to prevent the use of incorrect or defective materials, parts, and components.

§ 71.119 Control of special processes.

The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that special processes, including welding, heat treating, and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

§ 71.121 Internal inspection.

The licensee, certificate holder, and applicant for a CoC shall establish and execute a program for inspection of activities affecting quality by or for the organization performing the activity, to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. The inspection must be performed by individuals other than those who performed the activity being inspected. Examination, measurements, or tests of material or products processed must be performed for each work operation where necessary to assure quality. If direct inspection of processed material or products is not carried out, indirect control by monitoring processing methods, equipment, and personnel must be provided. Both inspection and process monitoring must be provided when quality control is inadequate without both. If mandatory inspection hold points, which require witnessing or inspecting by the licensee's designated representative and beyond which work should not proceed without the consent of its designated representative, are required, the specific hold points must be indicated in appropriate documents.

§ 71.123 Test control.

The licensee, certificate holder, and applicant for a CoC shall establish a test program to assure that all testing required to demonstrate that the packaging components will perform satisfactorily in service is identified and performed in accordance with written test procedures that incorporate the requirements of this part and the requirements and acceptance limits contained in the package approval. The test procedures must include provisions for assuring that all prerequisites for the given test are met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions. The licensee, certificate holder, and applicant for a CoC shall document and evaluate the test results to assure that test requirements have been satisfied.

§ 71.125 Control of measuring and test equipment.

The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that tools, gauges, instruments, and other measuring and testing devices used in activities affecting 63

quality are properly controlled, calibrated, and adjusted at specified times to maintain accuracy within necessary limits.

§ 71.127 Handling, storage, and shipping control.

The licensee, certificate holder, and applicant for a CoC shall establish measures to control, in accordance with instructions, the handling, storage, shipping, cleaning, and preservation of materials and equipment to be used in packaging to prevent damage or deterioration. When necessary for particular products, special protective environments, such as inert gas atmosphere, and specific moisture content and temperature levels must be specified and provided.

§ 71.129 Inspection, test, and operating status.

(a) The licensee, certificate holder, and applicant for a CoC shall establish measures to indicate, by the use of markings such as stamps, tags, labels, routing cards, or other suitable means, the status of inspections and tests performed upon individual items of the packaging. These measures must provide for the identification of items that have satisfactorily passed required inspections and tests, where necessary to preclude inadvertent bypassing of the inspections and tests.

(b) The licensee shall establish measures to identify the operating status of components of the packaging, such as tagging valves and switches, to prevent inadvertent operation.

§ 71.131 Nonconforming materials, parts, or components.

The licensee, certificate holder, and applicant for a CoC shall establish measures to control materials, parts, or components that do not conform to the licensee's requirements to prevent their inadvertent use or installation. These measures must include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations.

Nonconforming items must be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures.

§ 71.133 Corrective action.

The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that conditions adverse to quality, such as deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. In the case of a significant condition adverse to quality, the measures must assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken must be documented and reported to appropriate levels of management.

§ 71.135 Quality assurance records.

The licensee, certificate holder, and applicant for a Certificate of Compliance shall maintain sufficient written records to describe the activities affecting quality. These records must include changes to the quality assurance program as required by § 71.106, the instructions, procedures, 64

and drawings required by § 71.111 to prescribe quality assurance activities, and closely related specifications such as required qualifications of personnel, procedures, and equipment. The records must include the instructions or procedures that establish a records retention program that is consistent with applicable regulations and designates factors such as duration, location, and assigned responsibility. The licensee, certificate holder, and applicant for a Certificate of Compliance shall retain these records for 3 years beyond the date when the licensee, certificate holder, and applicant for a Certificate of Compliance last engage in the activity for which the quality assurance program was developed. If any portion of the quality assurance program, written procedures or instructions is superseded, the licensee, certificate holder, and applicant for a Certificate of Compliance shall retain the superseded material for 3 years after it is superseded.

[80 FR 34014, Jun. 12, 2015]

§ 71.137 Audits.

The licensee, certificate holder, and applicant for a CoC shall carry out a comprehensive system of planned and periodic audits to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program. The audits must be performed in accordance with written procedures or checklists by appropriately trained personnel not having direct responsibilities in the areas being audited. Audited results must be documented and reviewed by management having responsibility in the area audited. Followup action, including reaudit of deficient areas, must be taken where indicated.

Appendix A to Part 71--Determination of A1 and A2 I. Values of A1 and A2 for individual radionuclides, which are the bases for many activity limits elsewhere in these regulations, are given in Table A-1. The curie (Ci) values specified are obtained by converting from the Terabecquerel (TBq) value. The Terabecquerel values are the regulatory standard. The curie values are for information only and are not intended to be the regulatory standard. Where values of A1 and A2 are unlimited, it is for radiation control purposes only. For nuclear criticality safety, some materials are subject to controls placed on fissile material.

II. a. For individual radionuclides whose identities are known, but which are not listed in Table A-1, the A1 and A2 values contained in Table A-3 may be used. Otherwise, the licensee shall obtain prior Commission approval of the A1 and A2 values for radionuclides not listed in Table A-1, before shipping the material.

b. For individual radionuclides whose identities are known, but which are not listed in Table A-2, the exempt material activity concentration and exempt consignment activity values contained in Table A-3 may be used. Otherwise, the licensee shall obtain prior Commission approval of the exempt material activity concentration and exempt consignment activity values for radionuclides not listed in Table A-2, before shipping the material.

c. The licensee shall submit requests for prior approval, described under paragraphs II(a) and II(b) of this Appendix, to the Commission, in accordance with § 71.1 of this part.

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III. In the calculations of A1 and A2 for a radionuclide not in Table A-1, a single radioactive decay chain, in which radionuclides are present in their naturally occurring proportions, and in which no daughter radionuclide has a half-life either longer than 10 days, or longer than that of the parent radionuclide, shall be considered as a single radionuclide, and the activity to be taken into account, and the A1 or A2 value to be applied, shall be those corresponding to the parent radionuclide of that chain. In the case of radioactive decay chains in which any daughter radionuclide has a half-life either longer than 10 days, or greater than that of the parent radionuclide, the parent and those daughter radionuclides shall be considered as mixtures of different radionuclides.

IV. For mixtures of radionuclides whose identities and respective activities are known, the following conditions apply:

a. For special form radioactive material, the maximum quantity transported in a Type A package is as follows:

B(i)

A (i) 1 i 1 where B(i) is the activity of radionuclide i in special form, and A1(i) is the A1 value for radionuclide i.

b. For normal form radioactive material, the maximum quantity transported in a Type A package is as follows:

B(i)

A (i) 1 i 2 where B(i) is the activity of radionuclide i in normal form, and A2(i) is the A2 value for radionuclide i.

c. If the package contains both special and normal form radioactive material, the activity that may be transported in a Type A package is as follows:

B(i ) C( j)

A (i) + A 1 i 1 j 2 (j) where B(i) is the activity of radionuclide i as special form radioactive material, A1(i) is the A1 value for radionuclide i, C(j) is the activity of radionuclide j as normal form radioactive material, and A2(j) is the A2 value for radionuclide j.

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d. Alternatively, the A1 value for mixtures of special form material may be determined as follows:

A1 for mixture = 1 f (i )

A (i) i 1 where f(i) is the fraction of activity for radionuclide i in the mixture and A1(i) is the appropriate A1 value for radionuclide i.

e. Alternatively, the A2 value for mixtures of normal form material may be determined as follows:

1 A2 for mixture =

f(i) i A2(i) where f(i) is the fraction of activity for radionuclide i in the mixture and A2(i) is the appropriate A2 value for radionuclide i.

f. The exempt activity concentration for mixtures of nuclides may be determined as follows:

1 Exempt activity concentration for mixture =

f(i) i [A](i) where f(i) is the fraction of activity concentration of radionuclide i in the mixture and [A](i) is the activity concentration for exempt material containing radionuclide i.

g. The activity limit for an exempt consignment for mixtures of radionuclides may be determined as follows:

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1 Exempt consignment activity limit for mixture =

f(i) i A(i) where f(i) is the fraction of activity of radionuclide i in the mixture and A(i) is the activity limit for exempt consignments for radionuclide i.

V. a. When the identity of each radionuclide is known, but the individual activities of some of the radionuclides are not known, the radionuclides may be grouped, and the lowest A1 or A2 value, as appropriate, for the radionuclides in each group may be used in applying the formulas in paragraph IV. Groups may be based on the total alpha activity and the total beta/gamma activity when these are known, using the lowest A1 or A2 values for the alpha emitters and beta/gamma emitters.

b. When the identity of each radionuclide is known but the individual activities of some of the radionuclides are not known, the radionuclides may be grouped and the lowest [A] (activity concentration for exempt material) or A (activity limit for exempt consignment) value, as appropriate, for the radionuclides in each group may be used in applying the formulas in paragraph IV of this appendix. Groups may be based on the total alpha activity and the total beta/gamma activity when these are known, using the lowest [A] or A values for the alpha emitters and beta/gamma emitters, respectively.

Table A-1A1 and A2 VALUES FOR RADIONUCLIDES Symbol of Element and atomic Specific activity A1 (TBq) A1(Ci)b A2 (TBq) A2(Ci)b radionuclide number (TBq/g) (Ci/g)

Ac-225 (a) Actinium (89) 8.0X10-1 2.2X101 6.0X10-3 1.6X10-1 2.1X103 5.8X104 Ac-227 (a) 9.0X10-1 2.4X101 9.0X10-5 2.4X10-3 2.7 7.2X101 Ac-228 6.0X10-1 1.6X101 5.0X10-1 1.4X101 8.4X104 2.2X106 Ag-105 Silver (47) 2.0 5.4X101 2.0 5.4X101 1.1X103 3.0X104 Ag-108m (a) 7.0X10-1 1.9X101 7.0X10-1 1.9X101 9.7X10-1 2.6X101 Ag-110m (a) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.8X102 4.7X103 Ag-111 2.0 5.4X101 6.0X10-1 1.6X101 5.8X103 1.6X105 Al-26 Aluminum (13) 1.0X10-1 2.7 1.0X10-1 2.7 7.0X10-4 1.9X10-2 Am-241 Americium (95) 1.0X101 2.7X102 1.0X10-3 2.7X10-2 1.3X10-1 3.4 Am-242m (a) 1.0X101 2.7X102 1.0X10-3 2.7X10-2 3.6X10-1 1.0X101 Am-243 (a) 5.0 1.4X102 1.0X10-3 2.7X10-2 7.4X10-3 2.0X10-1 Ar-37 Argon (18) 4.0X101 1.1X103 4.0X101 1.1X103 3.7X103 9.9X104 Ar-39 4.0X101 1.1X103 2.0X101 5.4X102 1.3 3.4X101 Ar-41 3.0X10-1 8.1 3.0X10-1 8.1 1.5X106 4.2X107 As-72 Arsenic (33) 3.0X10-1 8.1 3.0X10-1 8.1 6.2X104 1.7X106 As-73 4.0X101 1.1X103 4.0X101 1.1X103 8.2X102 2.2X104 68

As-74 1.0 2.7X101 9.0X10-1 2.4X101 3.7X103 9.9X104 As-76 3.0X10-1 8.1 3.0X10-1 8.1 5.8X104 1.6X106 As-77 2.0X101 5.4X102 7.0X10-1 1.9X101 3.9X104 1.0X106 At-211 (a) Astatine (85) 2.0X101 5.4X102 5.0X10-1 1.4X101 7.6X104 2.1X106 Au-193 Gold (79) 7.0 1.9X102 2.0 5.4X101 3.4X104 9.2X105 Au-194 1.0 2.7X101 1.0 2.7X101 1.5X104 4.1X105 Au-195 1.0X101 2.7X102 6.0 1.6X102 1.4X102 3.7X103 Au-198 1.0 2.7X101 6.0X10-1 1.6X101 9.0X103 2.4X105 Au-199 1.0X101 2.7X102 6.0X10-1 1.6X101 7.7X103 2.1X105 Ba-131 (a) Barium (56) 2.0 5.4X101 2.0 5.4X101 3.1X103 8.4X104 Ba-133 3.0 8.1X101 3.0 8.1X101 9.4 2.6X102 Ba-133m 2.0X101 5.4X102 6.0X10-1 1.6X101 2.2X104 6.1X105 Ba-140 (a) 5.0X10-1 1.4X101 3.0X10-1 8.1 2.7X103 7.3X104 Be-7 Beryllium (4) 2.0X101 5.4X102 2.0X101 5.4X102 1.3X104 3.5X105 Be-10 4.0X101 1.1X103 6.0X10-1 1.6X101 8.3X10-4 2.2X10-2 Bi-205 Bismuth (83) 7.0X10-1 1.9X101 7.0X10-1 1.9X101 1.5X103 4.2X104 Bi-206 3.0X10-1 8.1 3.0X10-1 8.1 3.8X103 1.0X105 Bi-207 7.0X10-1 1.9X101 7.0X10-1 1.9X101 1.9 5.2X101 Bi-210 1.0 2.7X101 6.0X10-1 1.6X101 4.6X103 1.2X105 Bi-210m (a) 6.0X10-1 1.6X101 2.0X10-2 5.4X10-1 2.1X10-5 5.7X10-4 Bi-212 (a) 7.0X10-1 1.9X101 6.0X10-1 1.6X101 5.4X105 1.5X107 Bk-247 Berkelium (97) 8.0 2.2X102 8.0X10-4 2.2X10-2 3.8X10-2 1.0 Bk-249 (a) 4.0X101 1.1X103 3.0X10-1 8.1 6.1X101 1.6X103 Br-76 Bromine (35) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 9.4X104 2.5X106 Br-77 3.0 8.1X101 3.0 8.1X101 2.6X104 7.1X105 Br-82 4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.0X104 1.1X106 C-11 Carbon (6) 1.0 2.7X101 6.0X10-1 1.6X101 3.1X107 8.4X108 C-14 4.0X101 1.1X103 3.0 8.1X101 1.6X10-1 4.5 Ca-41 Calcium (20) Unlimited Unlimited Unlimited Unlimited 3.1X10-3 8.5X10-2 Ca-45 4.0X101 1.1X103 1.0 2.7X101 6.6X102 1.8X104 Ca-47 (a) 3.0 8.1X101 3.0X10-1 8.1 2.3X104 6.1X105 Cd-109 Cadmium (48) 3.0X101 8.1X102 2.0 5.4X101 9.6X101 2.6X103 Cd-113m 4.0X101 1.1X103 5.0X10-1 1.4X101 8.3 2.2X102 Cd-115 (a) 3.0 8.1X101 4.0X10-1 1.1X101 1.9X104 5.1X105 Cd-115m 5.0X10-1 1.4X101 5.0X10-1 1.4X101 9.4X102 2.5X104 Ce-139 Cerium (58) 7.0 1.9X102 2.0 5.4X101 2.5X102 6.8X103 Ce-141 2.0X101 5.4X102 6.0X10-1 1.6X101 1.1X103 2.8X104 Ce-143 9.0X10-1 2.4X101 6.0X10-1 1.6X101 2.5X104 6.6X105 Ce-144 (a) 2.0X10-1 5.4 2.0X10-1 5.4 1.2X102 3.2X103 69

Cf-248 Californium (98) 4.0X101 1.1X103 6.0X10-3 1.6X10-1 5.8X101 1.6X103 Cf-249 3.0 8.1X101 8.0X10-4 2.2X10-2 1.5X10-1 4.1 Cf-250 2.0X101 5.4X102 2.0X10-3 5.4X10-2 4.0 1.1X102 Cf-251 7.0 1.9X102 7.0X10-4 1.9X10-2 5.9X10-2 1.6 Cf-252 1.0x10-1 2.7 3.0x10-3 8.1x10-2 2.0x101 5.4x102 Cf-253 (a) 4.0X101 1.1X103 4.0X10-2 1.1 1.1X103 2.9X104 Cf-254 1.0X10-3 2.7X10-2 1.0X10-3 2.7X10-2 3.1X102 8.5X103 Cl-36 Chlorine (17) 1.0X101 2.7X102 6.0X10-1 1.6X101 1.2X10-3 3.3X10-2 Cl-38 2.0X10-1 5.4 2.0X10-1 5.4 4.9X106 1.3X108 Cm-240 Curium (96) 4.0X101 1.1X103 2.0X10-2 5.4X10-1 7.5X102 2.0X104 Cm-241 2.0 5.4X101 1.0 2.7X101 6.1X102 1.7X104 Cm-242 4.0X101 1.1X103 1.0X10-2 2.7X10-1 1.2X102 3.3X103 Cm-243 9.0 2.4X102 1.0X10-3 2.7X10-2 1.9X10-3 5.2X101 Cm-244 2.0X101 5.4X102 2.0X10-3 5.4X10-2 3.0 8.1X101 Cm-245 9.0 2.4X102 9.0X10-4 2.4X10-2 6.4X10-3 1.7X10-1 Cm-246 9.0 2.4X102 9.0X10-4 2.4X10-2 1.1X10-2 3.1X10-1 Cm-247 (a) 3.0 8.1X101 1.0X10-3 2.7X10-2 3.4X10-6 9.3X10-5 Cm-248 2.0X10-2 5.4X10-1 3.0X10-4 8.1X10-3 1.6X10-4 4.2X10-3 Co-55 Cobalt (27) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.1X105 3.1X106 Co-56 3.0X10-1 8.1 3.0X10-1 8.1 1.1X103 3.0X104 Co-57 1.0X101 2.7X102 1.0X101 2.7X102 3.1X102 8.4X103 Co-58 1.0 2.7X101 1.0 2.7X101 1.2X103 3.2X104 Co-58m 4.0X101 1.1X103 4.0X101 1.1X103 2.2X105 5.9X106 Co-60 4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.2X101 1.1X103 Cr-51 Chromium (24) 3.0X101 8.1X102 3.0X101 8.1X102 3.4X103 9.2X104 Cs-129 Cesium (55) 4.0 1.1X102 4.0 1.1X102 2.8X104 7.6X105 Cs-131 3.0X101 8.1X102 3.0X101 8.1X102 3.8X103 1.0X105 Cs-132 1.0 2.7X101 1.0 2.7X101 5.7X103 1.5X105 Cs-134 7.0X10-1 1.9X101 7.0X10-1 1.9X101 4.8X101 1.3X103 Cs-134m 4.0X101 1.1X103 6.0X10-1 1.6X101 3.0X105 8.0X106 Cs-135 4.0X101 1.1X103 1.0 2.7X101 4.3X10-5 1.2X10-3 Cs-136 5.0X10-1 1.4X101 5.0X10-1 1.4X101 2.7X103 7.3X104 Cs-137 (a) 2.0 5.4X101 6.0X10-1 1.6X101 3.2 8.7X101 Cu-64 Copper (29) 6.0 1.6X102 1.0 2.7X101 1.4X105 3.9X106 Cu-67 1.0X101 2.7X102 7.0X10-1 1.9X101 2.8X104 7.6X105 Dy-159 Dysprosium (66) 2.0X101 5.4X102 2.0X101 5.4X102 2.1X102 5.7X103 Dy-165 9.0X10-1 2.4X101 6.0X10-1 1.6X101 3.0X105 8.2X106 Dy-166 (a) 9.0X10-1 2.4X101 3.0X10-1 8.1 8.6X103 2.3X105 Er-169 Erbium (68) 4.0X101 1.1X103 1.0 2.7X101 3.1X103 8.3X104 70

Er-171 8.0X10-1 2.2X101 5.0X10-1 1.4X101 9.0X104 2.4X106 Eu-147 Europium (63) 2.0 5.4X101 2.0 5.4X101 1.4X103 3.7X104 Eu-148 5.0X10-1 1.4X101 5.0X10-1 1.4X101 6.0X102 1.6X104 Eu-149 2.0X101 5.4X102 2.0X101 5.4X102 3.5X102 9.4X103 Eu-150 (short 2.0 5.4X101 7.0X10-1 1.9X101 6.1X104 1.6X106 lived)

Eu-150 (long 7.0X10-1 1.9X101 7.0X10-1 1.9X101 6.1X104 1.6X106 lived)

Eu-152 1.0 2.7X101 1.0 2.7X101 6.5 1.8X102 Eu-152m 8.0X10-1 2.2X101 8.0X10-1 2.2X101 8.2X104 2.2X106 Eu-154 9.0X10-1 2.4X101 6.0X10-1 1.6X101 9.8 2.6X102 Eu-155 2.0X101 5.4X102 3.0 8.1X101 1.8X101 4.9X102 Eu-156 7.0X10-1 1.9X101 7.0X10-1 1.9X101 2.0X103 5.5X104 F-18 Fluorine (9) 1.0 2.7X101 6.0X10-1 1.6X101 3.5X106 9.5X107 Fe-52 (a) Iron (26) 3.0X10-1 8.1 3.0X10-1 8.1 2.7X105 7.3X106 Fe-55 4.0X101 1.1X103 4.0X101 1.1X103 8.8X101 2.4X103 Fe-59 9.0X10-1 2.4X101 9.0X10-1 2.4X101 1.8X103 5.0X104 Fe-60 (a) 4.0X101 1.1X103 2.0X10-1 5.4 7.4X10-4 2.0X10-2 Ga-67 Gallium (31) 7.0 1.9X102 3.0 8.1X101 2.2X104 6.0X105 Ga-68 5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.5X106 4.1X107 Ga-72 4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.1X105 3.1X106 Gd-146 (a) Gadolinium (64) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 6.9X102 1.9X104 Gd-148 2.0X101 5.4X102 2.0X10-3 5.4X10-2 1.2 3.2X101 Gd-153 1.0X101 2.7X102 9.0 2.4X102 1.3X102 3.5X103 Gd-159 3.0 8.1X101 6.0X10-1 1.6X101 3.9X104 1.1X106 Ge-68 (a) Germanium (32) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 2.6X102 7.1X103 Ge-71 4.0X101 1.1X103 4.0X101 1.1X103 5.8X103 1.6X105 Ge-77 3.0X10-1 8.1 3.0X10-1 8.1 1.3X105 3.6X106 Hf-172 (a) Hafnium (72) 6.0X10-1 1.6X101 6.0X10-1 1.6X101 4.1X101 1.1X103 Hf-175 3.0 8.1X101 3.0 8.1X101 3.9X102 1.1X104 Hf-181 2.0 5.4X101 5.0X10-1 1.4X101 6.3X102 1.7X104 Hf-182 Unlimited Unlimited Unlimited Unlimited 8.1X10-6 2.2X10-4 Hg-194 (a) Mercury (80) 1.0 2.7X101 1.0 2.7X101 1.3X10-1 3.5 Hg-195m (a) 3.0 8.1X101 7.0X10-1 1.9X101 1.5X104 4.0X105 Hg-197 2.0X101 5.4X102 1.0X101 2.7X102 9.2X103 2.5X105 Hg-197m 1.0X101 2.7X102 4.0X10-1 1.1X101 2.5X104 6.7X105 Hg-203 5.0 1.4X102 1.0 2.7X101 5.1X102 1.4X104 Ho-166 Holmium (67) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 2.6X104 7.0X105 Ho-166m 6.0X10-1 1.6X101 5.0X10-1 1.4X101 6.6X10-2 1.8 I-123 Iodine (53) 6.0 1.6X102 3.0 8.1X101 7.1X104 1.9X106 71

I-124 1.0 2.7X101 1.0 2.7X101 9.3X103 2.5X105 I-125 2.0X101 5.4X102 3.0 8.1X101 6.4X102 1.7X104 I-126 2.0 5.4X101 1.0 2.7X101 2.9X103 8.0X104 I-129 Unlimited Unlimited Unlimited Unlimited 6.5X10-6 1.8X10-4 I-131 3.0 8.1X101 7.0X10-1 1.9X101 4.6X103 1.2X105 I-132 4.0X10-1 1.1X101 4.0X10-1 1.1X101 3.8X105 1.0X107 I-133 7.0X10-1 1.9X101 6.0X10-1 1.6X101 4.2X104 1.1X106 I-134 3.0X10-1 8.1 3.0X10-1 8.1 9.9X105 2.7X107 I-135 (a) 6.0X10-1 1.6X101 6.0X10-1 1.6X101 1.3X105 3.5X106 In-111 Indium (49) 3.0 8.1X101 3.0 8.1X101 1.5X104 4.2X105 In-113m 4.0 1.1X102 2.0 5.4X101 6.2X105 1.7X107 In-114m (a) 1.0X101 2.7X102 5.0X10-1 1.4X101 8.6X102 2.3X104 In-115m 7.0 1.9X102 1.0 2.7X101 2.2X105 6.1X106 Ir-189 (a) Iridium (77) 1.0X101 2.7X102 1.0X101 2.7X102 1.9X103 5.2X104 Ir-190 7.0X10-1 1.9X101 7.0X10-1 1.9X101 2.3X103 6.2X104 Ir-192 c1.0 c2.7x101 6.0x10-1 1.6x101 3.4x102 9.2x103 Ir-194 3.0X10-1 8.1 3.0X10-1 8.1 3.1X104 8.4X105 K-40 Potassium (19) 9.0X10-1 2.4X101 9.0X10-1 2.4X101 2.4X10-7 6.4X10-6 K-42 2.0X10-1 5.4 2.0X10-1 5.4 2.2X105 6.0X106 K-43 7.0X10-1 1.9X101 6.0X10-1 1.6X101 1.2X105 3.3X106 Kr-79 Krypton (36) 4.0 1.1x102 2.0 5.4x101 4.2x104 1.1x106 Kr-81 4.0x101 1.1x103 4.0x101 1.1x103 7.8x10-4 2.1x10-2 Kr-85 1.0X101 2.7X102 1.0X101 2.7X102 1.5X101 3.9X102 Kr-85m 8.0 2.2X102 3.0 8.1X101 3.0X105 8.2X106 Kr-87 2.0X10-1 5.4 2.0X10-1 5.4 1.0X106 2.8X107 La-137 Lanthanum (57) 3.0X101 8.1X102 6.0 1.6X102 1.6X10-3 4.4X10-2 La-140 4.0X10-1 1.1X101 4.0X10-1 1.1X101 2.1X104 5.6X105 Lu-172 Lutetium (71) 6.0X10-1 1.6X101 6.0X10-1 1.6X101 4.2X103 1.1X105 Lu-173 8.0 2.2X102 8.0 2.2X102 5.6X101 1.5X103 Lu-174 9.0 2.4X102 9.0 2.4X102 2.3X101 6.2X102 Lu-174m 2.0X101 5.4X102 1.0X101 2.7X102 2.0X102 5.3X103 Lu-177 3.0X101 8.1X102 7.0X10-1 1.9X101 4.1X103 1.1X105 Mg-28 (a) Magnesium (12) 3.0X10-1 8.1 3.0X10-1 8.1 2.0X105 5.4X106 Mn-52 Manganese (25) 3.0X10-1 8.1 3.0X10-1 8.1 1.6X104 4.4X105 Mn-53 Unlimited Unlimited Unlimited Unlimited 6.8X10-5 1.8X10-3 Mn-54 1.0 2.7X101 1.0 2.7X101 2.9X102 7.7X103 Mn-56 3.0X10-1 8.1 3.0X10-1 8.1 8.0X105 2.2X107 Mo-93 Molybdenum (42) 4.0X101 1.1X103 2.0X101 5.4X102 4.1X10-2 1.1 Mo-99 ah 1.0 2.7x101 6.0x10-1 1.6x101 1.8x104 4.8x105 72

N-13 Nitrogen (7) 9.0X10-1 2.4X101 6.0X10-1 1.6X101 5.4X107 1.5X109 Na-22 Sodium (11) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 2.3X102 6.3X103 Na-24 2.0X10-1 5.4 2.0X10-1 5.4 3.2X105 8.7X106 Nb-93m Niobium (41) 4.0X101 1.1X103 3.0X101 8.1X102 8.8 2.4X102 Nb-94 7.0X10-1 1.9X101 7.0X10-1 1.9X101 6.9X10-3 1.9X10-1 Nb-95 1.0 2.7X101 1.0 2.7X101 1.5X103 3.9X104 Nb-97 9.0X10-1 2.4X101 6.0X10-1 1.6X101 9.9X105 2.7X107 Nd-147 Neodymium (60) 6.0 1.6X102 6.0X10-1 1.6X101 3.0X103 8.1X104 Nd-149 6.0X10-1 1.6X101 5.0X10-1 1.4X101 4.5X105 1.2X107 Ni-59 Nickel (28) Unlimited Unlimited Unlimited Unlimited 3.0X10-3 8.0X10-2 Ni-63 4.0X101 1.1X103 3.0X101 8.1X102 2.1 5.7X101 Ni-65 4.0X10-1 1.1X101 4.0X10-1 1.1X101 7.1X105 1.9X107 Np-235 Neptunium (93) 4.0X101 1.1X103 4.0X101 1.1X103 5.2X101 1.4X103 Np-236 2.0X101 5.4X102 2.0 5.4X101 4.7X10-4 1.3X10-2 (short-lived)

Np-236 (long-9.0X100 2.4X102 2.0X10-2 5.4X10-1 4.7X10-4 1.3X10-2 lived)

Np-237 2.0X101 5.4X102 2.0X10-3 5.4X10-2 2.6X10-5 7.1X10-4 Np-239 7.0 1.9X102 4.0X10-1 1.1X101 8.6X103 2.3X105 Os-185 Osmium (76) 1.0 2.7X101 1.0 2.7X101 2.8X102 7.5X103 Os-191 1.0X101 2.7X102 2.0 5.4X101 1.6X103 4.4X104 Os-191m 4.0X101 1.1X103 3.0X101 8.1X102 4.6X104 1.3X106 Os-193 2.0 5.4X101 6.0X10-1 1.6X101 2.0X104 5.3X105 Os-194 (a) 3.0X10-1 8.1 3.0X10-1 8.1 1.1X101 3.1X102 P-32 Phosphorus (15) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.1X104 2.9X105 P-33 4.0X101 1.1X103 1.0 2.7X101 5.8X103 1.6X105 Pa-230 (a) Protactinium (91) 2.0 5.4X101 7.0X10-2 1.9 1.2X103 3.3X104 Pa-231 4.0 1.1X102 4.0X10-4 1.1X10-2 1.7X10-3 4.7X10-2 Pa-233 5.0 1.4X102 7.0X10-1 1.9X101 7.7X102 2.1X104 Pb-201 Lead (82) 1.0 2.7X101 1.0 2.7X101 6.2X104 1.7X106 Pb-202 4.0X101 1.1X103 2.0X101 5.4X102 1.2X10-4 3.4X10-3 Pb-203 4.0 1.1X102 3.0 8.1X101 1.1X104 3.0X105 Pb-205 Unlimited Unlimited Unlimited Unlimited 4.5X10-6 1.2X10-4 Pb-210 (a) 1.0 2.7X101 5.0X10-2 1.4 2.8 7.6X101 Pb-212 (a) 7.0X10-1 1.9X101 2.0X10-1 5.4 5.1X104 1.4X106 Pd-103 (a) Palladium (46) 4.0X101 1.1X103 4.0X101 1.1X103 2.8X103 7.5X104 Pd-107 Unlimited Unlimited Unlimited Unlimited 1.9X10-5 5.1X10-4 Pd-109 2.0 5.4X101 5.0X10-1 1.4X101 7.9X104 2.1X106 Pm-143 Promethium (61) 3.0 8.1X101 3.0 8.1X101 1.3X102 3.4X103 Pm-144 7.0X10-1 1.9X101 7.0X10-1 1.9X101 9.2X101 2.5X103 73

Pm-145 3.0X101 8.1X102 1.0X101 2.7X102 5.2 1.4X102 Pm-147 4.0X101 1.1X103 2.0 5.4X101 3.4X101 9.3X102 Pm-148m (a) 8.0X10-1 2.2X101 7.0X10-1 1.9X101 7.9X102 2.1X104 Pm-149 2.0 5.4X101 6.0X10-1 1.6X101 1.5X104 4.0X105 Pm-151 2.0 5.4X101 6.0X10-1 1.6X101 2.7X104 7.3X105 Po-210 Polonium (84) 4.0X101 1.1X103 2.0X10-2 5.4X10-1 1.7X102 4.5X103 Pr-142 Praseodymium (59) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.3X104 1.2X106 Pr-143 3.0 8.1X101 6.0X10-1 1.6X101 2.5X103 6.7X104 Pt-188 (a) Platinum (78) 1.0 2.7X101 8.0X10-1 2.2X101 2.5X103 6.8X104 Pt-191 4.0 1.1X102 3.0 8.1X101 8.7X103 2.4X105 Pt-193 4.0X101 1.1X103 4.0X101 1.1X103 1.4 3.7X101 Pt-193m 4.0X101 1.1X103 5.0X10-1 1.4X101 5.8X103 1.6X105 Pt-195m 1.0X101 2.7X102 5.0X10-1 1.4X101 6.2X103 1.7X105 Pt-197 2.0X101 5.4X102 6.0X10-1 1.6X101 3.2X104 8.7X105 Pt-197m 1.0X101 2.7X102 6.0X10-1 1.6X101 3.7X105 1.0X107 Pu-236 Plutonium (94) 3.0X101 8.1X102 3.0X10-3 8.1X10-2 2.0X101 5.3X102 Pu-237 2.0X101 5.4X102 2.0X101 5.4X102 4.5X102 1.2X104 Pu-238 1.0X101 2.7X102 1.0X10-3 2.7X10-2 6.3X10-1 1.7X101 Pu-239 1.0X101 2.7X102 1.0X10-3 2.7X10-2 2.3X10-3 6.2X10-2 Pu-240 1.0X101 2.7X102 1.0X10-3 2.7X10-2 8.4X10-3 2.3X10-1 Pu-241 (a) 4.0X101 1.1X103 6.0X10-2 1.6 3.8 1.0X102 Pu-242 1.0X101 2.7X102 1.0X10-3 2.7X10-2 1.5X10-4 3.9X10-3 Pu-244 (a) 4.0X10-1 1.1X101 1.0X10-3 2.7X10-2 6.7X10-7 1.8X10-5 Ra-223 (a) Radium (88) 4.0X10-1 1.1X101 7.0X10-3 1.9X10-1 1.9X103 5.1X104 Ra-224 (a) 4.0X10-1 1.1X101 2.0X10-2 5.4X10-1 5.9X103 1.6X105 Ra-225 (a) 2.0X10-1 5.4 4.0X10-3 1.1X10-1 1.5X103 3.9X104 Ra-226 (a) 2.0X10-1 5.4 3.0X10-3 8.1X10-2 3.7X10-2 1.0 Ra-228 (a) 6.0X10-1 1.6X101 2.0X10-2 5.4X10-1 1.0X101 2.7X102 Rb-81 Rubidium (37) 2.0 5.4X101 8.0X10-1 2.2X101 3.1X105 8.4X106 Rb-83 (a) 2.0 5.4X101 2.0 5.4X101 6.8X102 1.8X104 Rb-84 1.0 2.7X101 1.0 2.7X101 1.8X103 4.7X104 Rb-86 5.0X10-1 1.4X101 5.0X10-1 1.4X101 3.0X103 8.1X104 Rb-87 Unlimited Unlimited Unlimited Unlimited 3.2X10-9 8.6X10-8 Rb(nat) Unlimited Unlimited Unlimited Unlimited 6.7X106 1.8X108 Re-184 Rhenium (75) 1.0 2.7X101 1.0 2.7X101 6.9X102 1.9X104 Re-184m 3.0 8.1X101 1.0 2.7X101 1.6X102 4.3X103 Re-186 2.0 5.4X101 6.0X10-1 1.6X101 6.9X103 1.9X105 Re-187 Unlimited Unlimited Unlimited Unlimited 1.4X10-9 3.8X10-8 Re-188 4.0X10-1 1.1X101 4.0X10-1 1.1X101 3.6X104 9.8X105 74

Re-189 (a) 3.0 8.1X101 6.0X10-1 1.6X101 2.5X104 6.8X105 Re(nat) Unlimited Unlimited Unlimited Unlimited 0.0 2.4X10-8 Rh-99 Rhodium (45) 2.0 5.4X101 2.0 5.4X101 3.0X103 8.2X104 Rh-101 4.0 1.1X102 3.0 8.1X101 4.1X101 1.1X103 Rh-102 5.0X10-1 1.4X101 5.0X10-1 1.4X101 4.5X101 1.2X103 Rh-102m 2.0 5.4X101 2.0 5.4X101 2.3X102 6.2X103 Rh-103m 4.0X101 1.1X103 4.0X101 1.1X103 1.2X106 3.3X107 Rh-105 1.0X101 2.7X102 8.0X10-1 2.2X101 3.1X104 8.4X105 Rn-222 (a) Radon (86) 3.0X10-1 8.1 4.0X10-3 1.1X10-1 5.7X103 1.5X105 Ru-97 Ruthenium (44) 5.0 1.4X102 5.0 1.4X102 1.7X104 4.6X105 Ru-103 (a) 2.0 5.4X101 2.0 5.4X101 1.2X103 3.2X104 Ru-105 1.0 2.7X101 6.0X10-1 1.6X101 2.5X105 6.7X106 Ru-106 (a) 2.0X10-1 5.4 2.0X10-1 5.4 1.2X102 3.3X103 S-35 Sulphur (16) 4.0X101 1.1X103 3.0 8.1X101 1.6X103 4.3X104 Sb-122 Antimony (51) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.5X104 4.0X105 Sb-124 6.0X10-1 1.6X101 6.0X10-1 1.6X101 6.5X102 1.7X104 Sb-125 2.0 5.4X101 1.0 2.7X101 3.9X101 1.0X103 Sb-126 4.0X10-1 1.1X101 4.0X10-1 1.1X101 3.1X103 8.4X104 Sc-44 Scandium (21) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 6.7X105 1.8X107 Sc-46 5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.3X103 3.4X104 Sc-47 1.0X101 2.7X102 7.0X10-1 1.9X101 3.1X104 8.3X105 Sc-48 3.0X10-1 8.1 3.0X10-1 8.1 5.5X104 1.5X106 Se-75 Selenium (34) 3.0 8.1X101 3.0 8.1X101 5.4X102 1.5X104 Se-79 4.0X101 1.1X103 2.0 5.4X101 2.6X10-3 7.0X10-2 Si-31 Silicon (14) 6.0X10-1 1.6X101 6.0X10-1 1.6X101 1.4X106 3.9X107 Si-32 4.0X101 1.1X103 5.0X10-1 1.4X101 3.9 1.1X102 Sm-145 Samarium (62) 1.0X101 2.7X102 1.0X101 2.7X102 9.8X101 2.6X103 Sm-147 Unlimited Unlimited Unlimited Unlimited 8.5X10-1 2.3X10-8 Sm-151 4.0X101 1.1X103 1.0X101 2.7X102 9.7X10-1 2.6X101 Sm-153 9.0 2.4X102 6.0X10-1 1.6X101 1.6X104 4.4X105 Sn-113 (a) Tin (50) 4.0 1.1X102 2.0 5.4X101 3.7X102 1.0X104 Sn-117m 7.0 1.9X102 4.0X10-1 1.1X101 3.0X103 8.2X104 Sn-119m 4.0X101 1.1X103 3.0X101 8.1X102 1.4X102 3.7X103 Sn-121m (a) 4.0X101 1.1X103 9.0X10-1 2.4X101 2.0 5.4X101 Sn-123 8.0X10-1 2.2X101 6.0X10-1 1.6X101 3.0X102 8.2X103 Sn-125 4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.0X103 1.1X105 Sn-126 (a) 6.0X10-1 1.6X101 4.0X10-1 1.1X101 1.0X10-3 2.8X10-2 Sr-82 (a) Strontium (38) 2.0X10-1 5.4 2.0X10-1 5.4 2.3X103 6.2X104 Sr-85 2.0 5.4X101 2.0 5.4X101 8.8X102 2.4X104 75

Sr-85m 5.0 1.4X102 5.0 1.4X102 1.2X106 3.3X107 Sr-87m 3.0 8.1X101 3.0 8.1X101 4.8X105 1.3X107 Sr-89 6.0X10-1 1.6X101 6.0X10-1 1.6X101 1.1X103 2.9X104 Sr-90 (a) 3.0X10-1 8.1 3.0X10-1 8.1 5.1 1.4X102 Sr-91 (a) 3.0X10-1 8.1 3.0X10-1 8.1 1.3X105 3.6X106 Sr-92 (a) 1.0 2.7X101 3.0X10-1 8.1 4.7X105 1.3X107 T(H-3) Tritium (1) 4.0X101 1.1X103 4.0X101 1.1X103 3.6X102 9.7X103 Ta-178 (long-Tantalum (73) 1.0 2.7X101 8.0X10-1 2.2X101 4.2X106 1.1X108 lived)

Ta-179 3.0X101 8.1X102 3.0X101 8.1X102 4.1X101 1.1X103 Ta-182 9.0X10-1 2.4X101 5.0X10-1 1.4X101 2.3X102 6.2X103 Tb-157 Terbium (65) 4.0X101 1.1X103 4.0X101 1.1X103 5.6X10-1 1.5X101 Tb-158 1.0 2.7X101 1.0 2.7X101 5.6X10-1 1.5X101 Tb-160 1.0 2.7X101 6.0X10-1 1.6X101 4.2X102 1.1X104 Tc-95m (a) Technetium (43) 2.0 5.4X101 2.0 5.4X101 8.3X102 2.2X104 Tc-96 4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.2X104 3.2X105 Tc-96m (a) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.4X106 3.8X107 Tc-97 Unlimited Unlimited Unlimited Unlimited 5.2X10-5 1.4X10-3 Tc-97m 4.0X101 1.1X103 1.0 2.7X101 5.6X102 1.5X104 Tc-98 8.0X10-1 2.2X101 7.0X10-1 1.9X101 3.2X10-5 8.7X10-4 Tc-99 4.0X101 1.1X103 9.0X10-1 2.4X101 6.3X10-4 1.7X10-2 Tc-99m 1.0X101 2.7X102 4.0 1.1X102 1.9X105 5.3X106 Te-121 Tellurium (52) 2.0 5.4X101 2.0 5.4X101 2.4X103 6.4X104 Te-121m 5.0 1.4X102 3.0 8.1X101 2.6X102 7.0X103 Te-123m 8.0 2.2X102 1.0 2.7X101 3.3X102 8.9X103 Te-125m 2.0X101 5.4X102 9.0X10-1 2.4X101 6.7X102 1.8X104 Te-127 2.0X101 5.4X102 7.0X10-1 1.9X101 9.8X104 2.6X106 Te-127m (a) 2.0X101 5.4X102 5.0X10-1 1.4X101 3.5X102 9.4X103 Te-129 7.0X10-1 1.9X101 6.0X10-1 1.6X101 7.7X105 2.1X107 Te-129m (a) 8.0X10-1 2.2X101 4.0X10-1 1.1X101 1.1X103 3.0X104 Te-131m (a) 7.0X10-1 1.9X101 5.0X10-1 1.4X101 3.0X104 8.0X105 Te-132 (a) 5.0X10-1 1.4X101 4.0X10-1 1.1X101 1.1X104 3.0X105 Th-227 Thorium (90) 1.0X101 2.7X102 5.0X10-3 1.4X10-1 1.1X103 3.1X104 Th-228 (a) 5.0X10-1 1.4X101 1.0X10-3 2.7X10-2 3.0X101 8.2X102 Th-229 5.0 1.4X102 5.0X10-4 1.4X10-2 7.9X10-3 2.1X10-1 Th-230 1.0X101 2.7X102 1.0X10-3 2.7X10-2 7.6X10-4 2.1X10-2 Th-231 4.0X101 1.1X103 2.0X10-2 5.4X10-1 2.0X104 5.3X105 Th-232 Unlimited Unlimited Unlimited Unlimited 4.0X10-9 1.1X10-7 Th-234 (a) 3.0X10-1 8.1 3.0X10-1 8.1 8.6X102 2.3X104 76

Th(nat) Unlimited Unlimited Unlimited Unlimited 8.1X10-9 2.2X10-7 Ti-44 (a) Titanium (22) 5.0X10-1 1.4X101 4.0X10-1 1.1X101 6.4 1.7X102 Tl-200 Thallium (81) 9.0X10-1 2.4X101 9.0X10-1 2.4X101 2.2X104 6.0X105 Tl-201 1.0X101 2.7X102 4.0 1.1X102 7.9X103 2.1X105 Tl-202 2.0 5.4X101 2.0 5.4X101 2.0X103 5.3X104 Tl-204 1.0X101 2.7X102 7.0X10-1 1.9X101 1.7X101 4.6X102 Tm-167 Thulium (69) 7.0 1.9X102 8.0X10-1 2.2X101 3.1X103 8.5X104 Tm-170 3.0 8.1X101 6.0X10-1 1.6X101 2.2X102 6.0X103 Tm-171 4.0X101 1.1X103 4.0X101 1.1X103 4.0X101 1.1X103 U-230 (fast Uranium (92) 4.0X101 1.1X103 1.0X10-1 2.7 1.0X103 2.7X104 lung absorption)

(a)(d)

U-230 4.0X101 1.1X103 4.0X10-3 1.1X10-1 1.0X103 2.7X104 (medium lung absorption)

(a)(e)

U-230 (slow 3.0X101 8.1X102 3.0X10-3 8.1X10-2 1.0X103 2.7X104 lung absorption)

(a)(f)

U-232 (fast 4.0X101 1.1X103 1.0X10-2 2.7X10-1 8.3X10-1 2.2X101 lung absorption)

(d)

U-232 4.0X101 1.1X103 7.0X10-3 1.9X10-1 8.3X10-1 2.2X101 (medium lung absorption)

(e)

U-232 (slow 1.0X101 2.7X102 1.0X10-3 2.7X10-2 8.3X10-1 2.2X101 lung absorption) (f)

U-233 (fast 4.0X101 1.1X103 9.0X10-2 2.4 3.6X10-4 9.7X10-3 lung absorption)

(d)

U-233 4.0X101 1.1X103 2.0X10-2 5.4X10-1 3.6X10-4 9.7X10-3 (medium lung absorption)

(e)

U-233 (slow 4.0X101 1.1X103 6.0X10-3 1.6X10-1 3.6X10-4 9.7X10-3 lung absorption) (f)

U-234 (fast 4.0X101 1.1X103 9.0X10-2 2.4 2.3X10-4 6.2X10-3 lung absorption)

(d)

U-234 4.0X101 1.1X103 2.0X10-2 5.4X10-1 2.3X10-4 6.2X10-3 (medium lung absorption) 77

(e)

U-234 (slow 4.0X101 1.1X103 6.0X10-3 1.6X10-1 2.3X10-4 6.2X10-3 lung absorption) (f)

U-235 (all Unlimited Unlimited Unlimited Unlimited 8.0X10-8 2.2X10-6 lung absorption types)

(a),(d),(e),(f)

U-236 (fast Unlimited Unlimited Unlimited Unlimited 2.4X10-6 6.5X10-5 lung absorption)

(d)

U-236 4.0X101 1.1X103 2.0X10-2 5.4X10-1 2.4X10-6 6.5X10-5 (medium lung absorption)

(e)

U-236 (slow lung 4.0X101 1.1X103 6.0X10-3 1.6X10-1 2.4X10-6 6.5X10-5 absorption) (f)

U-238 (all Unlimited Unlimited Unlimited Unlimited 1.2X10-8 3.4X10-7 lung absorption types)

(d),(e),(f)

U (nat) Unlimited Unlimited Unlimited Unlimited 2.6X10-8 7.1X10-7 U (enriched to Unlimited Unlimited Unlimited Unlimited See Table A-4 See Table A-4 20% or less)

(g)

U (dep) Unlimited Unlimited Unlimited Unlimited See Table A-4 (See Table A-3)

V-48 Vanadium (23) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 6.3X103 1.7X105 V-49 4.0X101 1.1X103 4.0X101 1.1X103 3.0X102 8.1X103 W-178 (a) Tungsten (74) 9.0 2.4X102 5.0 1.4X102 1.3X103 3.4X104 W-181 3.0X101 8.1X102 3.0X101 8.1X102 2.2X102 6.0X103 W-185 4.0X101 1.1X103 8.0X10-1 2.2X101 3.5X102 9.4X103 W-187 2.0 5.4X101 6.0X10-1 1.6X101 2.6X104 7.0X105 W-188 (a) 4.0X10-1 1.1X101 3.0X10-1 8.1 3.7X102 1.0X104 Xe-122 (a) Xenon (54) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.8X104 1.3X106 Xe-123 2.0 5.4X101 7.0X10-1 1.9X101 4.4X105 1.2X107 Xe-127 4.0 1.1X102 2.0 5.4X101 1.0X103 2.8X104 Xe-131m 4.0X101 1.1X103 4.0X101 1.1X103 3.1X103 8.4X104 Xe-133 2.0X101 5.4X102 1.0X101 2.7X102 6.9X103 1.9X105 Xe-135 3.0 8.1X101 2.0 5.4X101 9.5X104 2.6X106 Y-87 (a) Yttrium (39) 1.0 2.7X101 1.0 2.7X101 1.7X104 4.5X105 Y-88 4.0X10-1 1.1X101 4.0X10-1 1.1X101 5.2X102 1.4X104 Y-90 3.0X10-1 8.1 3.0X10-1 8.1 2.0X104 5.4X105 78

Y-91 6.0X10-1 1.6X101 6.0X10-1 1.6X101 9.1X102 2.5X104 Y-91m 2.0 5.4X101 2.0 5.4X101 1.5X106 4.2X107 Y-92 2.0X10-1 5.4 2.0X10-1 5.4 3.6X105 9.6X106 Y-93 3.0X10-1 8.1 3.0X10-1 8.1 1.2X105 3.3X106 Yb-169 Ytterbium (70) 4.0 1.1X102 1.0 2.7X101 8.9X102 2.4X104 Yb-175 3.0X101 8.1X102 9.0X10-1 2.4X101 6.6X103 1.8X105 Zn-65 Zinc (30) 2.0 5.4X101 2.0 5.4X101 3.0X102 8.2X103 Zn-69 3.0 8.1X101 6.0X10-1 1.6X101 1.8X106 4.9X107 Zn-69m (a) 3.0 8.1X101 6.0X10-1 1.6X101 1.2X105 3.3X106 Zr-88 Zirconium (40) 3.0 8.1X101 3.0 8.1X101 6.6X102 1.8X104 Zr-93 Unlimited Unlimited Unlimited Unlimited 9.3X10-5 2.5X10-3 Zr-95 (a) 2.0 5.4X101 8.0X10-1 2.2X101 7.9X102 2.1X104 Zr-97 (a) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 7.1X104 1.9X106 a

A1 and/or A2 values include contributions from daughter nuclides with half-lives less than 10 days, as listed in the following:

Mg-28 Al-28 Ca-47 Sc-47 Ti-44 Sc-44 Fe-52 Mn-52m Fe-60 Co-60m Zn-69m Zn-69 Ge-68 Ga-68 Rb-83 Kr-83m Sr-82 Rb-82 Sr-90 Y-90 Sr-91 Y-91m Sr-92 Y-92 Y-87 Sr-87m Zr-95 Nb-95m Zr-97 Nb-97m, Nb-97 Mo-99 Tc-99m Tc-95m Tc-95 Tc-96m Tc-96 Ru-103 Rh-103m Ru-106 Rh-106 Pd-103 Rh-103m Ag-108m Ag-108 Ag-110m Ag-110 Cd-115 In-115m In-114m In-114 Sn-113 In-113m Sn-121m Sn-121 79

Sn-126 Sb-126m Te-127m Te-127 Te-129m Te-129 Te-131m Te-131 Te-132 I-132 I-135 Xe-135m Xe-122 I-122 Cs-137 Ba-137m Ba-131 Cs-131 Ba-140 La-140 Ce-144 Pr-144m, Pr-144 Pm-148m Pm-148 Gd-146 Eu-146 Dy-166 Ho-166 Hf-172 Lu-172 W-178 Ta-178 W-188 Re-188 Re-189 Os-189m Os-194 Ir-194 Ir-189 Os-189m Pt-188 Ir-188 Hg-194 Au-194 Hg-195m Hg-195 Pb-210 Bi-210 Pb-212 Bi-212, Tl-208, Po-212 Bi-210m Tl-206 Bi-212 Tl-208, Po-212 At-211 Po-211 Rn-222 Po-218, Pb-214, At-218, Bi-214, Po-214 Ra-223 Rn-219, Po-215, Pb-211, Bi-211, Po-211, Tl-207 Ra-224 Rn-220, Po-216, Pb-212, Bi-212, Tl-208, Po-212 Ra-225 Ac-225, Fr-221, At-217, Bi-213, Tl-209, Po-213, Pb-209 Ra-226 Rn-222, Po-218, Pb-214, At-218, Bi-214, Po-214 Ra-228 Ac-228 Ac-225 Fr-221, At-217, Bi-213, Tl-209, Po-213, Pb-209 Ac-227 Fr-223 Th-228 Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208, Po-212 Th-234 Pa-234m, Pa-234 Pa-230 Ac-226, Th-226, Fr-222, Ra-222, Rn-218, Po-214 U-230 Th-226, Ra-222, Rn-218, Po-214 U-235 Th-231 Pu-241 U-237 Pu-244 U-240, Np-240m Am-242m Am-242, Np-238 Am-243 Np-239 Cm-247 Pu-243 80

Bk-249 Am-245 Cf-253 Cm-249 b

The values of A1 and A2 in Curies (Ci) are approximate and for information only; the regulatory standard units are Terabecquerels (TBq) (see Appendix A to Part 71 - Determination of A1 and A2,Section I).

c The activity of Ir-192 in special form may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the source.

d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.

e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.

f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.

g These values apply to unirradiated uranium only.

h A2 = 0.74 TBq (20 Ci) for Mo-99 for domestic use.

Table A-2EXEMPT MATERIAL ACTIVITY CONCENTRATIONS AND EXEMPT CONSIGNMENT ACTIVITY LIMITS FOR RADIONUCLIDES Activity Activity Activity Activity limit concentration concentration limit for Symbol of for exempt Element and atomic number for exempt for exempt exempt radionuclide consignment material material consignment (Ci)

(Bq/g) (Ci/g) (Bq)

Ac-225 Actinium (89) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Ac-227 1.0X10-1 2.7X10-12 1.0X103 2.7X10-8 Ac-228 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Ag-105 Silver (47) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ag-108m (b) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Ag-110m 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Ag-111 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Al-26 Aluminum (13) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Am-241 Americium (95) 1.0 2.7X10-11 1.0X104 2.7X10-7 Am-242m (b) 1.0 2.7X10-11 1.0X104 2.7X10-7 Am-243 (b) 1.0 2.7X10-11 1.0X103 2.7X10-8 Ar-37 Argon (18) 1.0X106 2.7X10-5 1.0X108 2.7X10-3 Ar-39 1.0X107 2.7X10-4 1.0X104 2.7X10-7 Ar-41 1.0X102 2.7X10-9 1.0X109 2.7X10-2 As-72 Arsenic (33) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 As-73 1.0X103 2.7X10-8 1.0X107 2.7X10-4 As-74 1.0X101 2.7X10-10 1.0X106 2.7X10-5 81

As-76 1.0X102 2.7X10-9 1.0X105 2.7X10-6 As-77 1.0X103 2.7X10-8 1.0X106 2.7X10-5 At-211 Astatine (85) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Au-193 Gold (79) 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Au-194 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Au-195 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Au-198 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Au-199 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ba-131 Barium (56) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ba-133 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ba-133m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ba-140 (b) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Be-7 Beryllium (4) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Be-10 1.0X104 2.7X10-7 1.0X106 2.7X10-5 Bi-205 Bismuth (83) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Bi-206 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Bi-207 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Bi-210 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Bi-210m 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Bi-212 (b) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Bk-247 Berkelium (97) 1.0 2.7X10-11 1.0X104 2.7X10-7 Bk-249 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Br-76 Bromine (35) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Br-77 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Br-82 1.0X101 2.7X10-10 1.0X106 2.7X10-5 C-11 Carbon (6) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 C-14 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Ca-41 Calcium (20) 1.0X105 2.7X10-6 1.0X107 2.7X10-4 Ca-45 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Ca-47 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Cd-109 Cadmium (48) 1.0X104 2.7X10-7 1.0X106 2.7X10-5 Cd-113m 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Cd-115 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Cd-115m 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Ce-139 Cerium (58) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ce-141 1.0X102 2.7X10-9 1.0X107 2.7X10-4 82

Ce-143 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ce-144 (b) 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Cf-248 Californium (98) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Cf-249 1.0 2.7X10-11 1.0X103 2.7X10-8 Cf-250 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Cf-251 1.0 2.7X10-11 1.0X103 2.7X10-8 Cf-252 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Cf-253 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Cf-254 1.0 2.7X10-11 1.0X103 2.7X10-8 Cl-36 Chlorine (17) 1.0X104 2.7X10-7 1.0X106 2.7X10-5 Cl-38 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Cm-240 Curium (96) 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Cm-241 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Cm-242 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Cm-243 1.0 2.7X10-11 1.0X104 2.7X10-7 Cm-244 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Cm-245 1.0 2.7X10-11 1.0X103 2.7X10-8 Cm-246 1.0 2.7X10-11 1.0X103 2.7X10-8 Cm-247 1.0 2.7X10-11 1.0X104 2.7X10-7 Cm-248 1.0 2.7X10-11 1.0X103 2.7X10-8 Co-55 Cobalt (27) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Co-56 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Co-57 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Co-58 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Co-58m 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Co-60 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Cr-51 Chromium (24) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Cs-129 Cesium (55) 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Cs-131 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Cs-132 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Cs-134 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Cs-134m 1.0X103 2.7X10-8 1.0X105 2.7X10-6 Cs-135 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Cs-136 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Cs-137 (b) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Cu-64 Copper (29) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 83

Cu-67 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Dy-159 Dysprosium (66) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Dy-165 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Dy-166 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Er-169 Erbium (68) 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Er-171 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Eu-147 Europium (63) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Eu-148 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Eu-149 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Eu-150 (short lived) 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Eu-150 (long lived) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Eu-152 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Eu-152m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Eu-154 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Eu-155 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Eu-156 1.0X101 2.7X10-10 1.0X106 2.7X10-5 F-18 Fluorine (9) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Fe-52 Iron (26) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Fe-55 1.0X104 2.7X10-7 1.0X106 2.7X10-5 Fe-59 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Fe-60 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Ga-67 Gallium (31) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ga-68 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Ga-72 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Gd-146 Gadolinium (64) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Gd-148 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Gd-153 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Gd-159 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Ge-68 Germanium (32) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Ge-71 1.0X104 2.7X10-7 1.0X108 2.7X10-3 Ge-77 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Hf-172 Hafnium (72) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Hf-175 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Hf-181 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Hf-182 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Hg-194 Mercury (80) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 84

Hg-195m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Hg-197 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Hg-197m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Hg-203 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Ho-166 Holmium (67) 1.0X103 2.7X10-8 1.0X105 2.7X10-6 Ho-166m 1.0X101 2.7X10-10 1.0X106 2.7X10-5 I-123 Iodine (53) 1.0X102 2.7X10-9 1.0X107 2.7X10-4 I-124 1.0X101 2.7X10-10 1.0X106 2.7X10-5 I-125 1.0X103 2.7X10-8 1.0X106 2.7X10-5 I-126 1.0X102 2.7X10-9 1.0X106 2.7X10-5 I-129 1.0X102 2.7X10-9 1.0X105 2.7X10-6 I-131 1.0X102 2.7X10-9 1.0X106 2.7X10-5 I-132 1.0X101 2.7X10-10 1.0X105 2.7X10-6 I-133 1.0X101 2.7X10-10 1.0X106 2.7X10-5 I-134 1.0X101 2.7X10-10 1.0X105 2.7X10-6 I-135 1.0X101 2.7X10-10 1.0X106 2.7X10-5 In-111 Indium (49) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 In-113m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 In-114m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 In-115m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ir-189 Iridium (77) 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Ir-190 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Ir-192 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Ir-194 1.0X102 2.7X10-9 1.0X105 2.7X10-6 K-40 Potassium (19) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 K-42 1.0X102 2.7X10-9 1.0X106 2.7X10-5 K-43 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Kr-79 Krypton (36) 1.0x103 2.7x10-8 1.0x105 2.7x10-6 Kr-81 1.0x104 2.7x10-7 1.0x107 2.7x10-4 Kr-85 1.0X105 2.7X10-6 1.0X104 2.7X10-7 Kr-85m 1.0X103 2.7X10-8 1.0X1010 2.7X10-1 Kr-87 1.0X102 2.7X10-9 1.0X109 2.7X10-2 La-137 Lanthanum (57) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 La-140 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Lu-172 Lutetium (71) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Lu-173 1.0X102 2.7X10-9 1.0X107 2.7X10-4 85

Lu-174 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Lu-174m 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Lu-177 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Mg-28 Magnesium (12) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Mn-52 Manganese (25) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Mn-53 1.0X104 2.7X10-7 1.0X109 2.7X10-2 Mn-54 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Mn-56 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Mo-93 Molybdenum (42) 1.0X103 2.7X10-8 1.0X108 2.7X10-3 Mo-99 1.0X102 2.7X10-9 1.0X106 2.7X10-5 N-13 Nitrogen (7) 1.0X102 2.7X10-9 1.0X109 2.7X10-2 Na-22 Sodium (11) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Na-24 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Nb-93m Niobium (41) 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Nb-94 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Nb-95 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Nb-97 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Nd-147 Neodymium (60) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Nd-149 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ni-59 Nickel (28) 1.0X104 2.7X10-7 1.0X108 2.7X10-3 Ni-63 1.0X105 2.7X10-6 1.0X108 2.7X10-3 Ni-65 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Np-235 Neptunium (93) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Np-236 (short-1.0X103 2.7X10-8 1.0X107 2.7X10-4 lived)

Np-236 (long-lived) 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Np-237 (b) 1.0 2.7X10-11 1.0X103 2.7X10-8 Np-239 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Os-185 Osmium (76) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Os-191 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Os-191m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Os-193 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Os-194 1.0X102 2.7X10-9 1.0X105 2.7X10-6 P-32 Phosphorus (15) 1.0X103 2.7X10-8 1.0X105 2.7X10-6 P-33 1.0X105 2.7X10-6 1.0X108 2.7X10-3 Pa-230 Protactinium (91) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Pa-231 1.0 2.7X10-11 1.0X103 2.7X10-8 86

Pa-233 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Pb-201 Lead (82) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Pb-202 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Pb-203 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Pb-205 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Pb-210 (b) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Pb-212 (b) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Pd-103 Palladium (46) 1.0X103 2.7X10-8 1.0X108 2.7X10-3 Pd-107 1.0X105 2.7X10-6 1.0X108 2.7X10-3 Pd-109 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Pm-143 Promethium (61) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Pm-144 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Pm-145 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Pm-147 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Pm-148m 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Pm-149 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Pm-151 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Po-210 Polonium (84) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Pr-142 Praseodymium (59) 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Pr-143 1.0X104 2.7X10-7 1.0X106 2.7X10-5 Pt-188 Platinum (78) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Pt-191 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Pt-193 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Pt-193m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Pt-195m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Pt-197 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Pt-197m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Pu-236 Plutonium (94) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Pu-237 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Pu-238 1.0 2.7X10-11 1.0X104 2.7X10-7 Pu-239 1.0 2.7X10-11 1.0X104 2.7X10-7 Pu-240 1.0 2.7X10-11 1.0X103 2.7X10-8 Pu-241 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Pu-242 1.0 2.7X10-11 1.0X104 2.7X10-7 Pu-244 1.0 2.7X10-11 1.0X104 2.7X10-7 Ra-223 (b) Radium (88) 1.0X102 2.7X10-9 1.0X105 2.7X10-6 87

Ra-224 (b) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Ra-225 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Ra-226 (b) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Ra-228 (b) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Rb-81 Rubidium (37) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Rb-83 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Rb-84 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Rb-86 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Rb-87 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Rb(nat) 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Re-184 Rhenium (75) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Re-184m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Re-186 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Re-187 1.0X106 2.7X10-5 1.0X109 2.7X10-2 Re-188 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Re-189 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Re(nat) 1.0X106 2.7X10-5 1.0X109 2.7X10-2 Rh-99 Rhodium (45) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Rh-101 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Rh-102 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Rh-102m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Rh-103m 1.0X104 2.7X10-7 1.0X108 2.7X10-3 Rh-105 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Rn-222 (b) Radon (86) 1.0X101 2.7X10-10 1.0X108 2.7X10-3 Ru-97 Ruthenium (44) 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Ru-103 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Ru-105 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Ru-106 (b) 1.0X102 2.7X10-9 1.0X105 2.7X10-6 S-35 Sulphur (16) 1.0X105 2.7X10-6 1.0X108 2.7X10-3 Sb-122 Antimony (51) 1.0X102 2.7X10-9 1.0X104 2.7X10-7 Sb-124 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Sb-125 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Sb-126 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Sc-44 Scandium (21) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Sc-46 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Sc-47 1.0X102 2.7X10-9 1.0X106 2.7X10-5 88

Sc-48 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Se-75 Selenium (34) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Se-79 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Si-31 Silicon (14) 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Si-32 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Sm-145 Samarium (62) 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Sm-147 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Sm-151 1.0X104 2.7X10-7 1.0X108 2.7X10-3 Sm-153 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Sn-113 Tin (50) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Sn-117m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Sn-119m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Sn-121m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Sn-123 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Sn-125 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Sn-126 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Sr-82 Strontium (38) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Sr-85 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Sr-85m 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Sr-87m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Sr-89 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Sr-90 (b) 1.0X102 2.7X10-9 1.0X104 2.7X10-7 Sr-91 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Sr-92 1.0X101 2.7X10-10 1.0X106 2.7X10-5 T(H-3) Tritium (1) 1.0X106 2.7X10-5 1.0X109 2.7X10-2 Ta-178 (long-lived) Tantalum (73) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Ta-179 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Ta-182 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Tb-157 Terbium (65) 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Tb-158 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Tb-160 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Tc-95m Technetium (43) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Tc-96 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Tc-96m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Tc-97 1.0X103 2.7X10-8 1.0X108 2.7X10-3 Tc-97m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 89

Tc-98 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Tc-99 1.0X104 2.7X10-7 1.0X107 2.7X10-4 Tc-99m 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Te-121 Tellurium (52) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Te-121m 1.0x102 2.7x10-9 1.0x106 2.7x10-5 Te-123m 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Te-125m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Te-127 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Te-127m 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Te-129 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Te-129m 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Te-131m 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Te-132 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Th-227 Thorium (90) 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Th-228 (b) 1.0 2.7X10-11 1.0X104 2.7X10-7 Th-229 (b) 1.0 2.7X10-11 1.0X103 2.7X10-8 Th-230 1.0 2.7X10-11 1.0X104 2.7X10-7 Th-231 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Th-232 1.0X101 2.7X10-10 1.0X104 2.7X10-7 Th-234 (b) 1.0X103 2.7X10-8 1.0X105 2.7X10-6 Th (nat) (b) 1.0 2.7X10-11 1.0X103 2.7X10-8 Ti-44 Titanium (22) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 Tl-200 Thallium (81) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Tl-201 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Tl-202 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Tl-204 1.0X104 2.7X10-7 1.0X104 2.7X10-7 Tm-167 Thulium (69) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Tm-170 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Tm-171 1.0X104 2.7X10-7 1.0X108 2.7X10-3 U-230 (fast lung Uranium (92) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 absorption) (b),(d)

U-230 (medium 1.0X101 2.7X10-10 1.0X104 2.7X10-7 lung absorption) (e)

U-230 (slow lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption) (f)

U-232 (fast lung 1.0 2.7X10-11 1.0X103 2.7X10-8 absorption) (b),(d)

U-232 (medium 1.0X101 2.7X10-10 1.0X104 2.7X10-7 90

lung absorption) (e)

U-232 (slow lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption) (f)

U-233 (fast lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption) (d)

U-233 (medium 1.0X102 2.7X10-9 1.0X105 2.7X10-6 lung absorption) (e)

U-233 (slow lung 1.0X101 2.7X10-10 1.0X105 2.7X10-6 absorption) (f)

U-234 (fast lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption) (d)

U-234 (medium 1.0X102 2.7X10-9 1.0X105 2.7X10-6 lung absorption) (e)

U-234 (slow lung 1.0X101 2.7X10-10 1.0X105 2.7X10-6 absorption) (f)

U-235 (all lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption types)

(b),(d),(e),(f)

U-236 (fast lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption) (d)

U-236 (medium 1.0X102 2.7X10-9 1.0X105 2.7X10-6 lung absorption) (e)

U-236 (slow lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption) (f)

U-238 (all lung 1.0X101 2.7X10-10 1.0X104 2.7X10-7 absorption types)

(b),(d),(e),(f)

U (nat) (b) 1.0 2.7X10-11 1.0X103 2.7X10-8 U (enriched to 20% 1.0 2.7X10-11 1.0X103 2.7X10-8 or less) (g)

U (dep) 1.0 2.7X10-11 1.0X103 2.7X10-8 V-48 Vanadium (23) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 V-49 1.0X104 2.7X10-7 1.0X107 2.7X10-4 W-178 Tungsten (74) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 W-181 1.0X103 2.7X10-8 1.0X107 2.7X10-4 W-185 1.0X104 2.7X10-7 1.0X107 2.7X10-4 W-187 1.0X102 2.7X10-9 1.0X106 2.7X10-5 W-188 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Xe-122 Xenon (54) 1.0X102 2.7X10-9 1.0X109 2.7X10-2 Xe-123 1.0X102 2.7X10-9 1.0X109 2.7X10-2 Xe-127 1.0X103 2.7X10-8 1.0X105 2.7X10-6 Xe-131m 1.0X104 2.7X10-7 1.0X104 2.7X10-7 91

Xe-133 1.0X103 2.7X10-8 1.0X104 2.7X10-7 Xe-135 1.0X103 2.7X10-8 1.0X1010 2.7X10-1 Y-87 Yttrium (39) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Y-88 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Y-90 1.0X103 2.7X10-8 1.0X105 2.7X10-6 Y-91 1.0X103 2.7X10-8 1.0X106 2.7X10-5 Y-91m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Y-92 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Y-93 1.0X102 2.7X10-9 1.0X105 2.7X10-6 Yb-169 Ytterbium (70) 1.0X102 2.7X10-9 1.0X107 2.7X10-4 Yb-175 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Zn-65 Zinc (30) 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Zn-69 1.0X104 2.7X10-7 1.0X106 2.7X10-5 Zn-69m 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Zr-88 Zirconium (40) 1.0X102 2.7X10-9 1.0X106 2.7X10-5 Zr-93 (b) 1.0X103 2.7X10-8 1.0X107 2.7X10-4 Zr-95 1.0X101 2.7X10-10 1.0X106 2.7X10-5 Zr-97 (b) 1.0X101 2.7X10-10 1.0X105 2.7X10-6 a

[Reserved]

b Parent nuclides and their progeny included in secular equilibrium are listed as follows:

Sr-90 Y-90 Zr-93 Nb-93m Zr-97 Nb-97 Ru-106 Rh-106 Ag-108m Ag-108 Cs-137 Ba-137m Ce-144 Pr-144 Ba-140 La-140 Bi-212 Tl-208 (0.36), Po-212 (0.64)

Pb-210 Bi-210, Po-210 Pb-212 Bi-212, Tl-208 (0.36), Po-212 (0.64)

Rn-222 Po-218, Pb-214, Bi-214, Po-214 Ra-223 Rn-219, Po-215, Pb-211, Bi-211, Tl-207 Ra-224 Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)

Ra-226 Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210 Ra-228 Ac-228 Th-228 Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212(0.64)

Th-229 Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209 Th-nat Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, 92

Tl-208 (0.36), Po-212 (0.64)

Th-234 Pa-234m U-230 Th-226, Ra-222, Rn-218, Po-214 U-232 Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)

U-235 Th-231 U-238 Th-234, Pa-234m U-nat Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210 Np-237 Pa-233 Am-242m Am-242 Am-243 Np-239 c

[Reserved]

d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.

e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.

f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.

g These values apply to unirradiated uranium only.

TABLE A-3GENERAL VALUES FOR A1 AND A2 A1 A2 Activity Activity Activity Activity concentration concentration limits for limits for Contents for exempt for exempt exempt exempt (TBq) (Ci) (TBq) (Ci) material material consignments consignments (Bq/g) (Ci/g) (Bq) (Ci)

Only beta or 1 x 10-1 2.7 x 100 2 x 10-2 5.4 x 10- 1 x 101 2.7 x 10-10 1 x 104 2.7 x 10-7 gamma emitting 1 radionuclides are known to be present Alpha emitting 2 x 10-1 5.4 x 100 9 x 10-5 2.4 x 10- 1 x 10-1 2.7 x 10-12 1 x 103 2.7 x 10-8 nuclides, but no 3 neutron emitters, are known to be present a Neutron 1 x 10-3 2.7 x 10- 9 x 10-5 2.4 x 10- 1 x 10-1 2.7 x 10-12 1 x 103 2.7 x 10-8 emitting 2 3 nuclides are known to be present or no 93

relevant data are available a

If beta or gamma emitting nuclides are known to be present, the A1 value of 0.1 TBq (2.7 Ci) should be used.

TABLE A-4ACTIVITY-MASS RELATIONSHIPS FOR URANIUM Uranium Enrichment1 Specific Activity wt % U-235 present TBq/g Ci/g 0.45 1.8 x 10-8 5.0 x 10-7 0.72 2.6 x 10-8 7.1 x 10-7 1 2.8 x 10-8 7.6 x 10-7 1.5 3.7 x 10-8 1.0 x 10-6 5 1.0 x 10-7 2.7 x 10-6 10 1.8 x 10-7 4.8 x 10-6 20 3.7 x 10-7 1.0 x 10-5 35 7.4 x 10-7 2.0 x 10-5 50 9.3 x 10-7 2.5 x 10-5 90 2.2 x 10-6 5.8 x 10-5 93 2.6 x 10-6 7.0 x 10-5 95 3.4 x 10-6 9.1 x 10-5 1

The figures for uranium include representative values for the activity of the uranium-234 that is concentrated during the enrichment process.

[60 FR 50264, Sept. 28, 1995 as amended at 61 FR 28724, June 6, 1996; 69 FR 3800, Jan. 26, 2004; 77 FR 39908, Jul. 6, 2012; 80 FR 34014, Jun. 12, 2015]

94