[Federal Register Volume 60, Number 188 (Thursday, September 28, 1995)]
[Rules and Regulations]
[Pages 50248-50289]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-23538]
[[Page 50247]]
_______________________________________________________________________
Part II
Nuclear Regulatory Commission
_______________________________________________________________________
10 CFR Part 71
Compatibility With the International Atomic Energy Agency (IAEA); Final
Rule
Federal Register / Vol. 60, No. 188 / Thursday, September 28, 1995 /
Rules and Regulations
[[Page 50248]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 71
RIN 3150-AC41
Compatibility With the International Atomic Energy Agency (IAEA)
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is revising the
regulations governing the transportation of radioactive material. The
final rule conforms NRC regulations with those of the International
Atomic Energy Agency, and codifies criteria for packages used to
transport plutonium by air. This action is necessary to ensure that NRC
regulations reflect accepted international standards and comply with
current legislative requirements.
EFFECTIVE DATE: April 1, 1996. Section 71.52 expires April 1, 1999.
ADDRESSES: Single copies of the regulatory analysis for this rule may
be obtained on request from the contact. Copies of the regulatory
analysis may be examined and copied, for a fee, in the Commission's
Public Document Room, at 2120 L Street (Lower Level), NW., Washington,
DC.
FOR FURTHER INFORMATION CONTACT: John R. Cook, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, telephone: (301) 415-8521.
SUPPLEMENTARY INFORMATION:
Background
The U.S. Nuclear Regulatory Commission is revising its regulations,
for the safe transportation of radioactive material to make them
compatible with those of the International Atomic Energy Agency (IAEA)
and to incorporate new criteria for packages used to transport
plutonium by air. The revised rule, in combination with a corresponding
amendment of Title 49, Code of Federal Regulations, by the U.S.
Department of Transportation (DOT), would bring U.S. regulations into
general accord with IAEA regulations (Regulations for the Safe
Transport of Radioactive Material, 1985 Edition, Safety Series No. 6).
The final rule also adopts approval criteria for packages used to
transport plutonium by air. These criteria were developed in response
to Public Law 94. Except for these revisions, NRC's basic standards for
packaging and transportation remain essentially unchanged. These
regulations apply to all NRC licensees who transport, or offer for
transport, byproduct, source, or special nuclear material, and will
help ensure the continued safe transportation of radioactive materials
in domestic and international commerce.
In addition, three Petitions for Rulemaking, concerning the
transportation of Low Specific Activity (LSA) radioactive material, are
denied in this action.
In 1969, the IAEA, recognizing that its international transport
regulations should be revised from time to time on the basis of
scientific and technical advances, as well as accumulated experience,
invited member states to submit comments and suggested changes to the
regulations. As a result of this initiative, the IAEA issued revised
regulations in 1973 (Regulations for the Safe Transport of Radioactive
Material, 1973 Edition, Safety Series No. 6). The IAEA also decided to
periodically review its transportation regulations, at intervals of
about 10 years, to ensure that the regulations are kept current. As a
result, a review of IAEA regulations was initiated, in 1979, that
resulted in the publication of revised regulations in 1985 (Regulations
for the Safe Transport of Radioactive Material, 1985 Edition, Safety
Series No. 6).
On August 5, 1983 (48 FR 35600) NRC published, in the Federal
Register a final revision to 10 CFR Part 71, ``Packaging and
Transportation of Radioactive Material.'' That revision, in combination
with a parallel revision of the hazardous materials transportation
regulations of DOT, brought U.S. domestic transport regulations at the
Federal level into general accord with the 1973 edition of IAEA
transport regulations. Some of the revisions that were eventually
included in the 1985 IAEA regulations were anticipated by NRC and DOT
when they were finalizing their transportation regulations in 1983.
These changes were incorporated in Titles 10 and 49 of the Code of
Federal Regulations at that time.
On June 8, 1988 (53 FR 21550) NRC published a proposed revision to
its regulations in 10 CFR Part 71 in the Federal Register for the
purpose of making U.S. transportation regulations compatible with the
1985 edition of the IAEA regulations. In a parallel rulemaking, DOT
published a proposed revision to its radioactive material
transportation regulations on November 14, 1989 (54 FR 47454). Several
corrections to the NRC proposed rule were published in the Federal
Register on June 22, 1988 (53 FR 23484). Interested persons were
invited to submit written comments and suggestions on the NRC proposal
and/or the supporting regulatory analysis by October 6, 1988. The
public comment period was subsequently extended to February 9, 1990. On
December 8, 1994, the NRC staff provided a briefing on the proposed LSA
requirements and the other revisions at the 416th meeting of the
Advisory Committee on Reactor Safeguards (ACRS). This meeting also
provided industry and the public another opportunity to present their
views on the revisions. Based on the public comments, consultations
with DOT, and other considerations, the Commission is adopting the
proposed rule, with some modifications.
Discussion of Major Changes From Current Requirements
Most of the revisions presented in the proposed rule are being
adopted in the final rule. These include additional hypothetical
accident test criteria for certain types of packages, an increase in
the number of radionuclides with listed A1 and A2 values,
changes in the currently listed A1 and A2 values for some
radionuclides, simplification of fissile material transport classes,
revised requirements for shipment of LSA materials, and inclusion of
criteria for packages used to transport plutonium by air. These changes
are discussed in more detail in the following paragraphs.
Additional Accident Test Requirements
IAEA deep-water immersion and dynamic crush tests are adopted in
the final rule. The 200 meter (656 ft) deep-water immersion test has
been added to the requirements for Type B packages (casks) authorized
for irradiated fuel content in excess of 37 PBq (10\6\ Ci)(Sec. 71.61
Special requirement for irradiated nuclear fuel shipments). The purpose
of the deep immersion test, which can be satisfied through engineering
evaluation or actual physical test (Sec. 71.41), is to ensure that the
cask containment system does not collapse, buckle, nor allow inleakage
of water, if submerged at 200 m (656 ft).
A dynamic crush test (Sec. 71.73(c)(2) Crush) has also been added
to Type B package requirements, for certain lightweight packages that
are minimally vulnerable to damage in the 9 m (30 ft) drop test, but
which have a high potential for radiation hazard, if package failure
occurs. IAEA regulations require the crush test in place of the 9 m (30
ft) drop test, for these packages. NRC is requiring both the crush test
and drop test, for lightweight packages, to ensure that package
response to both crush and drop forces is within applicable limits.
These requirements only apply to package designs certified after
this final
[[Page 50249]]
rule becomes effective. Further, this rule does not apply to packages
fabricated under previous versions of Part 71; however, previously
fabricated packages are subject to multilateral approval, when used for
international transport (Sec. 71.13(b)).
Expansion of Radionuclide List and Changes in Radionuclide Limits
Table A-1, in 10 CFR Part 71, Appendix A, lists the Type A package
quantity limits (A1 and A2 values) for many radionuclides.
The final rule increases the number of radionuclides listed, from 284
to 378. The final rule also adopts the revised A1 and A2
values contained in the 1985 edition of the IAEA regulations. As a
result, 144 A1 values previously listed in Table A-1 are being
increased, and 73 are being decreased, while 129 A2 values are
being increased, and 95 decreased. In addition, the final rule modifies
the method used to determine A1 and A2 values for unlisted
radionuclides.
Simplification of Fissile Material Classes
The final rule revises the criteria for shipment of fissile
material. Specifically, the rule eliminates the three fissile class
designations currently used establishes a single set of criteria for
all packages of fissile material, uses the transport index as the
primary control for the number of fissile packages that may be
transported together, and requires special arrangements for fissile
packages that do not meet the established criteria.
Inclusion of Criteria for Air Shipment of Plutonium
The final rule amends Part 71 to include approval criteria for
packages used to transport plutonium by air (Secs. 71.64, 71.74, and
71.88). These criteria were developed as a result of Pub. L. 94-79,
which prohibited NRC from licensing the air shipment of plutonium, in
any form, until NRC certified to the Congress that a safe container had
been developed. The NRC subsequently developed and certified package
criteria to Congress and published the criteria in NUREG-0360,
Qualification Criteria to Certify a Package for Air Transport of
Plutonium, dated January 1978. This final rule incorporates these
criteria. There are no corresponding criteria in IAEA regulations.
Modifications From Proposed Rule
The final rule differs from the proposed rule in several
significant respects and are described as follows:
1. Package limit for Shipment of LSA and Surface-Containment-Object
(SCO) Material. In its 1985 regulations, the IAEA added a limit of 10
mSv/hour (1 rem/hour) at 3 meters for the radiation level from the
unshielded contents of LSA and SCO (Surface Contaminated Object)
packages not designed to withstand accidents. This radiation level
limit controls the external radiation exposures to individuals if an
LSA package is severely damaged in a transportation accident.
The IAEA limit considers the loss of package shielding during an
accident but it does not consider the possibility that a package's
contents might be released and redistributed, causing a reduction in
self-shielding of the contents. The reduction in self-shielding could
result in potential accident radiation levels that significantly exceed
IAEA's 10 mSv/hour (1 rem /hour) at 3 meters limit.
The IAEA dose rate limit provides a significant added degree of
protection over the 1973 IAEA regulations (which specify no quantity
limit for LSA packages). NRC and DOT did not believe, however that the
IAEA limit provided the same level of safety for all types of LSA
material, particularly for relatively large quantities of radioactive
materials contained in dispersible LSA materials (e.g., resins and
other media used in liquid radioactive waste treatment).
In lieu of the radiation level limit, DOT and NRC proposed a
2A1 quantity limit for all LSA packages. Although this proposal
addressed the accident concern by directly limiting package quantity,
it was not compatible with the IAEA provisions. Both agencies received
many comments from industry on the proposed 2A1 quantity limit
that objected to the impacts on occupational dose and shipping costs.
Further, after a briefing on the draft final rule on December 8, 1994,
the Advisory Committee on Reactor Safeguards (ACRS) issued a letter
report, dated December 19, 1994, recommending, inter alia, that the
requirements again be reevaluated with the objective of making them
equivalent to the IAEA regulations.
After consideration of comments from ACRS and industry, DOT and NRC
have agreed to adopt the IAEA LSA provisions. Accordingly, the final
rule imposes a limit on the external radiation level at 3 meters from
the unshielded contents of LSA-II, LSA-III, or SCO-II packages of 10
mSv/hour (1 rem/hour) (Sec. 71.10(b)).
2. The final rule delays imposing the LSA package external
radiation level limit for 3 years. The effect of imposing the LSA
package limit is to reduce the quantity of LSA materials that can be
transported in non-Type B, LSA packages. The final rule may increase
demand for Type B packages, and there are very few currently available.
NRC had proposed a 1 year delay in implementing the new LSA rules.
Industry comments expressed the view that 1 year is not an adequate
period of time to design a package, have it approved by NRC, and
manufacture a reasonable number of Type B waste packages. NRC agrees,
and has included a delay of 3 years from the effective date of this
rule for implementation of this provision of the final rule
(Sec. 71.52).
3. The proposed rule would have adopted 2A1 as the threshold
below which licensees are exempt from NRC requirements for packages
containing LSA material (except for Secs. 71.5, 71.88 and 71.53).
Because NRC and DOT are adopting the IAEA LSA package limit, the final
rule changes the exemption threshold to 1 rem/h at 3 m
(Sec. 71.10(b)(2)). Thus, designs for packages used to ship LSA or SCO
in quantities where the external dose rate exceeds 1 rem/h at 3 m from
the unshielded material will be subject to NRC Type B package
regulations. Package designs for lesser quantities of LSA or SCO will
be self-certified, by package designers, as meeting applicable DOT IP-
1, IP-2, IP-3, Type A, or strong tight, package regulations. [Licensees
should note that DOT has prescribed, in its final rule, the use of IAEA
Industrial Packages (IP-1, IP-2, and IP-3) for LSA and SCO material.
For domestic transportation only, DOT also provides for the use of Type
A, and strong tight, containers.]
4. For compatibility with IAEA and DOT requirements, a new,
``Sec. 71.77 Qualification of LSA-III Material,'' has been added to
Subpart F. This section prescribes assessment of LSA-III material
leaching. (In the proposed rule, Sec. 71.77 contained ``Tests for
special form radioactive material.'' Those requirements have been moved
to Sec. 71.75 ``Qualification of special form material,'' in the final
rule.)
Other Administrative Actions
The final rule corrects numerical errors in Secs. 71.20(b)(3) and
71.24(b)(4) of the current rule (Secs. (71.20(c)(3) and 71.24(c)(4),
respectively, of the proposed rule). These errors, which were not
identified at the time the proposed rule was published, resulted when
the limit for graphite was expressed as an atomic ratio, instead of a
mass ratio. The errors were inadvertently adopted, in Part 71, during a
rulemaking in 1983, to make
[[Page 50250]]
NRC regulations compatible with 1973 IAEA transportation regulations.
IAEA has subsequently corrected these errors in the 1985 edition of its
transportation regulations.
Section 71.20(b)(3), as currently written, limits the mass of
graphite to ``* * * 150 times the total mass of uranium-235 plus
plutonium.'' Section 71.20(c)(3), in the final rule, would be amended
to read as follows: ``The total mass of graphite present does not
exceed 7.7 times the total mass of uranium-235 plus plutonium.''
Section 71.24(c)(4) would be similarly revised to change the limits on
graphite from 150 to 7.7 times the total mass of uranium-235 plus
plutonium.
NRC is correcting these errors in this final rule. The affected
sections may bear on the criticality safety of fissile materials in
transport. In addition, these corrections are expected to have minimal
impact because there are no shipping casks currently being used that
were designed using the erroneous provisions.
Summary and Resolution of Public Comments
There were 171 letters of comment received on the proposed rule
from industry, State, and local governments; environmental
organizations; medical facilities; and members of the public. A
discussion of general comments is presented below, followed by
responses to comments on specific sections of the proposed rule.
One of the most frequent comments noted differences among NRC, DOT,
and IAEA definitions and requirements where there were no reasons for
the differences. Many of the differences between NRC and DOT
requirements resulted from the long period of time between publication
of the NRC proposed rule (June 8, 1988) and publication of the DOT
proposed rule (November 14, 1989; 54 FR 47454). The two proposed rules
were intended to be published on or about the same date but
circumstances did not permit concurrent publication. Between
publication of the NRC and DOT rules, IAEA published a complete set of
minor changes and changes of detail to its regulations. These changes
were not contained in the NRC proposed rule, but were introduced in the
DOT proposed rule. In addition, a large number of printing errors
appeared in the text of the NRC proposed rule. Only the most
significant errors were rectified in a correction notice published June
22, 1988 (53 FR 23484). The remaining inconsistencies have been
corrected in the final rule.
Another frequently raised comment was in response to NRC's
inclusion of new criteria for the air transportation of plutonium. Out
of 171 total letters of comment on the proposed rule, 119 of those
letters were concerned with the single issue of air transportation of
plutonium. In general, these letters requested that NRC codify the
NUREG-0360 criteria for the safe air transportation of plutonium,
notwithstanding urging by the U.S. Department of Energy (DOE) that NRC
withhold codification until it could consider rules being developed by
IAEA for the safe air transportation of plutonium. Many of these
letters, primarily from residents of Alaska, attributed development of
the NUREG-0360 1 criteria to U.S. Senator Frank Murkowski.
However, the criteria in NUREG-0360 were developed by the NRC in
response to Public Law 94-79, enacted in 1975. (Senator Murkowski
sponsored much more recent legislation on transportation of plutonium
by air, identified as Section 5062 of Public Law 100-203, for which
regulatory criteria have not been developed.) NRC has relied on the
NUREG-0360 criteria for plutonium transportation by air since the
criteria were published in 1978. DOE's request that NRC withhold the
codification of the NUREG-0360 criteria while NRC considers the IAEA
alternative cannot be accommodated because there is no existing IAEA
alternative to consider and none is expected for several years.
Although the IAEA development process has begun, the process is long
and multifaceted. Predictions as to final content of an IAEA
alternative cannot be made at this time. It also should be noted that,
under Public Law 94-79, the proposed criteria would apply to any U.S.
import, export, or domestic plutonium air transport regardless of IAEA
regulations. Accordingly, the plutonium air transport criteria are
incorporated in the final rule.
\1\ Copies of NUREG-0360 may be purchased from the
Superintendent of Documents, U.S. Government Printing Office, P.O.
Box 37082, Washington, DC 20013-7082. Copies are also available from
the National Technical Information Service, 5285 Port Royal Road,
Springfield, VA 22161. A copy is also available for inspection and
copying for a fee in the NRC Public Document Room, 2120 L Street,
NW. (Lower Level), Washington, DC.
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Section 71.0 Purpose and Scope
One comment suggested that Sec. 71.0 (a) could be clarified by
referring to the need for a Type B package rather than to licensed
material in excess of a Type A quantity. Section 71.0 (a)(2) would then
read ``Procedures and standards for NRC approval of packaging and
shipping procedures for fissile material and for other licensed
material required by this Part to be transported in a Type B
packaging.''
Although the suggested wording may be a good description of Part
71, Fissile Type A packages are still subject to NRC approval.
Therefore a scope based on quantity of radioactive material is better
than a scope based on a single type of package.
Section 71.4 Definitions
One comment noted that the term ``licensed material'' is used in
Part 71, in several locations, but is not defined in Part 71. In
response to this comment, NRC has added the definition of ``licensed
material,'' as codified in 10 CFR Part 39, to the definitions in Part
71. The term ``licensed material'' only includes radioactive material
licensed by the NRC. One comment noted that in defining the term
``exclusive use,'' the parenthetical note ``* * * also referred to in
other regulations as `sole use' or `full load' '' is no longer
necessary. Those other terms have been almost completely phased out,
and IAEA has eliminated the clarifying note. NRC agrees and also has
eliminated the clarifying note.
One comment noted that the definition of ``exclusive use'' requires
that loading and unloading be performed by personnel having
radiological training and resources appropriate for safe handling of
the consignment. However, the definition provides no criteria to
indicate what that training should be. NRC believes this is an area
where the regulation includes a sufficient level of detail to define
the intent of the provision. NRC further notes that DOT has established
requirements for hazardous material employee training (see 49 CFR Part
172, Subpart H, Secs. 172.700-172.704, effective July 2, 1992).
One comment suggested that the term ``transport index'' specify
that the number be rounded up ``to the next tenth'' rather than ``to
the first decimal place.'' NRC believes that either terminology is
adequately clear, and is retaining the original wording for uniformity.
This wording has been used satisfactorily over a number of years.
One comment suggested that the ``Natural uranium'' definition
should be clarified to indicate that the phrase ``the remainder being
uranium-238'' refers strictly to a weight basis, not to a radioactivity
basis. NRC has made the clarification.
One comment raised the question whether ``licensee'' and ``licensee
of the Commission'' are synonymous, and whether the terms include
``persons
[[Page 50251]]
licensed by an Agreement State,'' so that the general licenses of
Secs. 71.12-71.24 could apply. NRC asserts that the terms ``licensee''
and ``licensee of the Commission'' are synonymous. For uniformity, the
NRC has eliminated the longer of the two terms in the final rule.
Neither term includes Agreement State licensees. However, Agreement
State licensees engaging in activities in non-Agreement States, or in
offshore waters, under the reciprocity provisions of 10 CFR Part 150,
``Exemptions and Continued Regulatory Authority in Agreement States and
in Offshore Waters under Section 274,'' are subject to the requirements
of 10 CFR Part 71. In such instances, the NRC general licenses
mentioned above apply to Agreement State licensees.
One comment noted that the term ``specific activity'' should only
be used when describing the radioactivity of a radionuclide per unit
mass of the element. When describing the radioactivity per unit mass of
a material in general, the comment suggested the use of the words
``concentration of radioactivity.'' NRC has been unable to confirm any
preferred limited use of the term ``specific activity,'' and, in view
of the years of successful international use of the term in its broader
sense, plans to continue that broader use.
One comment noted that the NRC and DOT definitions of ``exclusive
use'' are not identical, and that the DOT definition appears
preferable. In the final rules promulgated by NRC and DOT, the
definitions of ``exclusive use'' are identical.
One comment noted a difference in quantities, for DOT's proposed
rule ``highway route controlled quantities,'' in 49 CFR 173.403, and
for NRC's ``advanced notification of shipment of nuclear waste''
requirements in 10 CFR 71.97. The limits were intended to be the same.
As the comment suggested, the error (by NRC) was caused by the rounding
of the International System (of units) (SI) and customary units and has
been corrected in this final rule.
Section 71.4 Definitions (Dual Unit System--The International System
of Units Followed or Preceded by U.S. Standard or Customary Units).
Ten comments suggested both support for the dual unit system used
in both NRC and DOT proposed regulations and potential problems that
might result from a dual unit system. Several other comments suggested
that NRC and DOT be consistent in the use of units. NRC and DOT intend
to use dual units in specifying the regulatory requirements. The
introductory language to Sec. 71.4 states that the different units are
functionally equivalent and can be used interchangeably for purposes of
this part. There are no paperwork requirements in Part 71 (e.g.,
records, reports) where the mandatory use of units is specified. DOT
regulations also specify regulatory requirements in terms of dual
units. In 49 CFR 171.10, DOT specifies that the SI units are intended
to serve as the standard, but that the customary units (rounded) are
included to provide a functionally equivalent limit. The dual unit
approaches used by NRC and DOT are compatible.
In addition, DOT specifies, in 49 CFR Part 172, the units that must
be used to satisfy the communication standards for shipping papers and
package labels. Sections 172.203(d)(4)and 172.403(g)(2) require that
shipping papers and package labels be completed either in SI units
alone or in SI units and customary units. These requirements also
permit, for a period of one year after the effective date of the final
rule, the use of customary units on shipping papers and package labels
for domestic shipments only.
One comment noted that the double conversion from customary units
to SI units, and back to customary units produces specifications that
are out of line with standard material sizes. For example, a test with
what was a standard 6-inch-diameter mild steel bar, with an edge radius
of \1/4\ inch, was proposed as a test with a 5.91-inch diameter mild
steel bar, with an edge radius of 0.236 inch. The converted customary
units of length and weight have been returned to their original values
in the final rule.
One comment suggested greater consistency of units between the NRC
and DOT transportation regulations and the Commission's ``Standards for
Protection against Radiation'' in 10 CFR Part 20. Since the NRC and DOT
transportation rules were proposed, NRC has revised 10 CFR 20.1004,
``Units of Radiation Dose,'' and 10 CFR 20.1005, ``Units of
Radioactivity,'' to permit the use of either customary or SI units,
These revisions achieve greater consistency of units among
transportation and radiation protection regulations.
One comment noted that differences between IAEA and Part 71 A
values (expressed in conventional units) may cause problems in
international transport. The curie values in Safety Series #6, Table I
are approximate, rounded down from the TBq values after conversion to
Ci, whereas the curie values in Table A-1 Part 71 are converted from
the TBq values to three significant figures without rounding down. The
Part 71 method was used because it yields values that more closely
approximate previous Table A-1 values. As noted earlier in this
preamble, DOT regulations will require the use of the SI units in
shipping papers and labels for international shipments (although
conventional units may be used in addition to the SI units). The use of
SI units should retain consistency with the IAEA regulations.
One comment suggested that the term ``transport index'' be defined
using both customary and SI units, as IAEA has done. The proposed
definition was expressed only in customary units. NRC agrees with this
suggestion and has adopted the DOT definition of ``transport index''
which includes both customary and SI units.
Section 71.4 Definitions (LSA and SCO in Particular)
Several comments related to clarification of LSA definitions.
Two comments noted the typographical error in the proposed rule in
which the ``water with tritium'' concentrations for LSA-II were printed
as 27.0 Ci/ (1 TBq/), rather than as 27.0 Ci/l (1
TBq/l). Two other comments noted that the numerical values differed
from those in the DOT proposed rule (20 Ci/l and 0.8 TBq/l,
respectively). One comment stated a preference for the 27.0 Ci/l limit.
NRC values in the proposed rule were derived from the IAEA and DOT
values by rounding up the terabequerel limit and then converting to
curies. For consistency, NRC has adopted the IAEA and DOT values in the
final rule.
Three comments were concerned with the definition of LSA-I. The
first comment noted that material generated from the extraction of
uranium or thorium was not classified into any LSA category. The
comment recommended an LSA-I classification for this material. Another
comment recommended that the term ``contaminated earth'' in LSA-I be
expanded to include ``soil, earth, concrete rubble, and other bulk
debris.'' A third comment expressed concern that mill tailings
exceeding 10-6 A2/g could not be shipped in bulk under the
proposed rule. The comment recommended that either mill tailings be
specifically included in the definition of LSA-I without an activity or
concentration limit, or the specific activity limit for LSA-I be
increased to 4x10-6 A2/g.
NRC agrees that ore-like materials (materials with highly uniform
distribution of small quantities of radionuclides) should be
transported as LSA-I material. Accordingly, the definition of LSA-I has
been changed from ``contaminated earth * * * `` to
[[Page 50252]]
``contaminated earth, mill tailings, concrete rubble and other bulk
debris * * *'' Further, NRC believes that mill tailings will meet the
proposed 10-6 A2/g specific activity limit, and therefore has
not increased the limit.
Two comments suggested that NRC include a definition of the term
``closed transport vehicle'' used in the definition of LSA-I. This term
has been removed from the definition of LSA-I because NRC and DOT
concluded the use of a vehicle-based term in the definition of a
material was inappropriate. ``Closed transport vehicle'' is defined in
DOT's rule (49 CFR 173.403(c)).
One comment suggested that LSA-II material definition be expanded
to include activated materials, consolidated wastes, and materials
intrinsically contained in a relatively insoluble matrix. LSA-II is
expected to include primarily unsolidified material in which the
radioactive material may or may not be uniformly distributed, including
lesser activity resins and filter sludges, other similar materials from
reactor operations, similar materials from other fuel cycle operations,
scintillation vials, and hospital, biological, and decommissioning
wastes. There is, however, no prohibition against activated materials,
consolidated wastes, and materials intrinsically contained in a
relatively insoluble matrix in group LSA-II, provided the specific
activity limit is met. The IAEA established the LSA-III group
principally for irradiated reactor parts and other activated, or
activated and contaminated, equipment that exceed the limits for the
other LSA groups. NRC does not believe it is necessary to expand the
LSA-II group definition to include these materials. The NRC believes
that to do so might cause confusion with the LSA-III definition.
One comment stated that dewatered material should be defined as a
solid for LSA-II. NRC agrees that dewatered resins should be subject to
the specific activity for solids under LSA-II and notes that there is
no prohibition against dewatered resins in LSA-II.
One comment asked whether the specific activity limits for LSA-II
and LSA-III materials were pre- or post-solidification. The specific
activity limits apply to materials as prepared for shipment, i.e.,
post-solidification. However, licensees should note that packaging or
shielding material may not be considered in determining either the
specific activity or the radiation level at 3 m.
One comment recommended that NRC remove the criterion for leaching
that is applicable to LSA-III solids. The criterion limits the loss of
radioactive material per package, when the package is placed in water
for 7 days, to 0.1 A2. Another comment stated that the criterion
for leaching in the definition of LSA-III needed to be compatible with
the leachability index requirements for solidified waste in 10 CFR
Parts 60 and 61.
A control on the potential intake of these LSA-III materials is
necessary because the radioactivity is not entirely insoluble. Because
non-Type A packaging might be used in transporting these materials, a
release of 10-2A in an accident is assumed, with a possible
bystander uptake of 10-3 A2, under the standard model for
determining A2 values. Because the total body uptake must be
limited to 10-6 A2, the package's dispersible radioactive
contents (i.e., the leachate liquid), must not exceed 0.1 A2. For
purposes of compatibility with IAEA and DOT requirements, a new
Sec. 71.77, ``Qualification of LSA-III Material,'' has been added to
Subpart F. This section prescribes testing requirements for assessment
of LSA-III material leaching. The hazard from the transportation of
these materials is different from that posed by their disposal;
therefore, no attempt has been made to achieve compatibility between
transportation and disposal leachability limits.
One comment found the proposed rule unclear on the need for three
LSA categories and how to classify materials under the criteria,
including compacted dry active waste. IAEA developed the three LSA
groups to differentiate controls based on the activity, distribution,
and form of LSA material. The LSA-I group accommodates very uniformly
distributed materials, such as ores. LSA-III accommodates large
activated parts or solidified materials. LSA-II accommodates less
uniformly distributed materials, such as compacted dry active waste.
One comment described radioactive atoms in activated products as
inherently non-dispersible and relatively non-leachable. The comment
recommended that activated materials be authorized for shipment as LSA-
I, provided other transportation requirements are met. Although
activated materials do not pose a dispersibility hazard, these
materials are subject to localized concentrations of non-uniformly
distributed material. Consequently activated materials are included in
groups LSA-III and LSA-II.
One comment suggested changing the definition of SCO from ``* * *
not itself radioactive * * *'' to ``* * * not classed as radioactive
material under these rules * * *,'' since nothing is free of
radioactive material. NRC and DOT have adopted this comment.
Several comments identified a typographical error in the limit for
non-fixed contamination from beta and gamma emitters on the accessible
surface of SCO-I objects. That value has been changed from 1.08 x
10-5 Ci/cm \2\ to 10-4 microcurie/cm \2\. These comments also
noted inconsistencies in the NRC and DOT contamination limits e.g.,
(1.08 x 10-4 Ci/cm \2\ and 10-4 microcurie/cm \2\,
respectively). NRC has adopted the DOT convention for these limits in
the final rule.
One comment inquired as to whether it was consistent for NRC not to
exempt SCO-I from transportation requirements when facilities with
similar contamination levels may be released for unrestricted use
according to NRC Regulatory Guide 1.86. Under the final rule, SCO-I
group materials are exempt from NRC regulations, except for one
Sec. 71.5 requirement that licensees comply with DOT requirements.
Further, the SCO-I non-fixed surface contamination limits are greater
than, not similar to, the corresponding acceptable surface
contamination levels in Table 1 of NRC Regulatory Guide 1.86.
Several comments noted that the term ``inaccessible surface'' used
in the SCO-I definition is not defined and that it was not clear how to
comply with a limit for surfaces that were inaccessible. This provision
provides for the disposal of materials that have contaminated surfaces
that are not readily accessible. Examples of inaccessible surfaces
include: inner surfaces of pipes, inner surfaces of maintenance
equipment for nuclear facilities, and inner surfaces of glove boxes.
Compliance can be achieved by sampling a small area of the surface that
may be accessible or by a documented estimate of the inaccessible
surface contamination.
One comment stated a belief that the implementation of SCO groups
would: (a) Further complicate the preparation and shipment process,
without an increase in the safety and quality of waste shipments; (b)
result in a significant increase in personnel exposure costs, and
delays for preparation and disposal of radioactive waste; (c) require
substantial initial personnel training; and (d) require extensive
revisions of existing procedures and waste shipping computer programs.
NRC acknowledges that the introduction of multiple LSA and SCO groups
complicates the transportation of LSA materials. The IAEA consensus was
that it was appropriate to regulate SCO separately from LSA materials.
The purpose of
[[Page 50253]]
these groups is to recognize the lesser hazard of LSA and SCO relative
to other radioactive materials, and to provide relief from shipment
requirements that would otherwise apply to these materials, while still
assuring safety.
With regard to exposure, it is true that the LSA groups will
require some increased material treatment or handling. However, this
handling is necessary to eliminate the current practice in which there
is no quantity limit on LSA packages. This situation poses a risk to
the public during transport. Costs will increase, but not by an amount
considered significant for the industry. Training with regard to the
LSA groups, or any new provision, will be required. Periodic training
of hazardous material employees regarding the safe transportation of
hazardous materials is required by DOT regulations (49 CFR Part 172
Subpart H); instruction with regard to the LSA and SCO groups may be
included at that time.
Implementing the LSA groups will require revision of procedures and
computer codes. These costs are judged to be acceptable in order to
achieve compatibility with the IAEA regulations for the safe transport
of radioactive materials.
A comment noted that the SCO classification ``appears to be well-
meaning,'' but that the proposed criteria (presumably the proposed
2A1 limit) ``detract from its potential benefit and utility,'' and
that it would be easier and less expensive for both producers and
consumers of electricity to enjoy the benefits of new transportation
systems without the related restrictions. As stated previously, NRC has
adopted the IAEA 10 mSv/h (1 rem/h) at 3 m limit for LSA packages, and
believes that a limit is needed to protect the public from the
potential for excessive external radiation exposure in the case of a
severe transportation accident.
One comment suggested that the rule make clear that not every SCO
needs to be surveyed and that a random representative survey is
adequate. There is no requirement that each SCO in a package be
surveyed. The shipper must be able to demonstrate, however, that the
package contents comply with applicable SCO definitions.
One comment objected to the upper limit for removable surface
contamination for SCO-II (10-2 Ci/cm \2\ for beta and
gamma emitters) because this limit is a factor of 90 less than current
LSA limits, and would require extensive decontamination of reactor
outage equipment at each site. The comment stated such decontamination
is not warranted because it violates the as low as reasonably
achievable (ALARA) principle, and is not justified based on shipping
experience. The comment suggested that an SCO-III group be defined for
materials exceeding SCO-II, and that Type A packaging be required for
such materials.
Apparently, this comment is comparing the SCO-II limit for
removable (non-fixed) surface contamination with the current LSA limit
that applies to nonradioactive material objects that are externally
contaminated with radioactive material that is not readily dispersible.
The SCO-II limit for fixed surface contamination is a more appropriate
comparison with the current limit for not readily dispersible
contamination. The SCO-II fixed contamination limit is 20 times greater
than the current LSA limit for not readily dispersible contamination.
Section 71.5 Transportation of Licensed Material
Two comments asked for clarification of the specification ``* *
*outside of the confines of its plant or other place of use,'' when
describing transportation made subject to DOT regulations. One of those
comments suggested that the provision be reworded as ``* * *outside the
site of usage, as specified in the NRC license, or where transport is
on public highways.'' This wording clarifies the provision and has been
included in the final rule. Similar wording has been substituted in
Sec. 71.0(c).
A comment asked whether Sec. 71.5(b) means ``that an approval must
be obtained when the shipment is covered by local State regulations and
those regulations will be followed.'' The purpose of Sec. 71.5(b) is to
impose, by NRC authority, pertinent DOT requirements on shipments, by
NRC licensees, that are not normally subject to DOT requirements. There
is no exemption from the requirement of Sec. 71.5(b) regarding
compliance with State or local regulations.
Section 71.10 Exemption for Low Level Materials
A comment noted that the SI unit specification of 74 kBq/kg
(0.002Ci/g) for exempted low-level radioactive material in
Sec. 71.10(a) is not consistent with the 70 Bq value specified in the
DOT proposed rule. The specification in Sec. 71.10(a) has been changed
to 70 Bq/g, the value in the DOT's final rule. This exemption is
applicable only with respect to transportation, and is not generally
applicable to other Commission-regulated activities.
A comment noted that it would be useful to have an exemption for
small quantities of radioactive material in Sec. 71.10(a) as well as
the exemption for LSA material. The safety rationale developed by IAEA
2 for LSA material does not extend to other radioactive materials.
IAEA has been informed that a small quantity exemption may be a useful
concept. However, this exemption has not been developed yet.
\2\ International Atomic Energy Agency Safety Series #7--
``Explanatory Material for the IAEA Regulations for the Safe
Transport of Radioactive Material'' (1985 Edition). Available from
Bernam-Unipub, 4611-F Assembly Drive, Lanham, MD 20706-4391. Tel.
(301) 459-7666.
---------------------------------------------------------------------------
One comment asked that NRC clarify the use of a reference to
Sec. 71.53 in the ``Exemption for low-level materials'' provision of
Sec. 71.10(b), a provision that pertains to Type A and LSA packages. In
addition to control over excessive radiation, the Commission's
responsibility with respect to fissile material is to provide
reasonable controls to avoid the occurrence of accidental criticality.
The regulatory standards for this are found in Secs. 71.55 and 71.59.
There are some relatively common types of fissile material packages for
which there is no credible risk of criticality in transport, even in
the absence of controls. These packages are described in Sec. 71.53,
and are exempted from the criticality controls of Secs. 71.55 and
71.59, because the controls are unnecessary.
The provisions of Sec. 71.10, ``Exemption for low-level
materials,'' provide broad exemptions from 10 CFR Part 71 rules that
relinquish to DOT the control of types of shipments that are of low
risk both from radiation and criticality standpoints. To ensure that
only low criticality risk shipments are included in Sec. 71.10(b), NRC
restricts the exemption to Type A and LSA packages that either contain
no fissile material or satisfy the fissile material exemptions in
Sec. 71.53. It should be noted that the exemption does not relieve
licensees from DOT transportation requirements by reason of NRC
authority, nor does the exemption relieve licensees from the
restrictions on air transportation of plutonium imposed by Congress.
The proposed rule introduced a 2A1 quantity limit, for LSA
packages not designed to withstand accidents (non-Type B packages), to
control potential external radiation exposures. Thirty comments were
received requesting that the limit be changed in the final rule. Two
comments supported no limit; nine supported the IAEA dose limit of 10
mSv/h (1 rem/h)r at a distance of 3 meters for an unshielded package; 4
supported higher multiples of A1; and 15 supported the optional
use of either the IAEA limit or a higher multiple of A1. As
described previously in this
[[Page 50254]]
preamble, NRC and DOT have decided that the best overall response on
the LSA issue and these comments is to drop the proposed 2A1
quantity limit, and to adopt the IAEA radiation level limit of 10 mSv/h
(1 rem/h) at 3 m from the unshielded contents.
One comment suggested that the need for labels on LSA packages
should be reconsidered. Package labelling falls under DOT jurisdiction.
In its final rule, DOT has retained the exception from package marking
and labeling requirements for domestic LSA shipments consigned as
exclusive use (see 49 CFR 173.427).
One comment expressed concern over the transition of control of
packages for shipping Type B quantities of LSA radioactive material
from NRC to DOT. NRC has a centralized package design approval
authority, whereas DOT authority allows a shipper to determine
acceptable package designs (i.e., self-certify package designs). The
comment expressed apprehension about permitting each shipper to review
package and shipping restrictions against DOT regulations, a situation
that could result in some confusion and different interpretations of
the regulations.
In the final rule, the IAEA limit of 1 rem/h at 3 m from the
unshielded material contents has been established as the threshold for
NRC regulation of LSA or SCO package designs. NRC will review and
approve, if adequate, designs for packages that contain quantities of
LSA or SCO material that exceed that limit. The review by regulatory
authority of package designs for quantities that exceed the IAEA limit
is consistent with the approach used by other IAEA member states.
Section 71.13 Previously Approved Package
One comment proposed that the date specified in Sec. 71.13(b)(2) be
December 31, 1990, instead of December 31, 1992, to be consistent with
IAEA transportation regulations. The original 1985 IAEA transport
regulations specified December 31, 1990, as the cutoff date for the
routine use of packages manufactured under the 1973 edition of the
regulations. That date was subsequently extended for 2 years by one of
the periodic updates of IAEA regulations and was properly used in the
proposed rule. However, since the proposed date of December 31, 1992,
has passed, the final rule has been revised (by eliminating reference
to any particular date) to make this provision effective on the date
that the final rule becomes effective.
Two comments noted that the preamble to the proposed Part 71
indicated that Type B and fissile packages fabricated before a certain
date and not used internationally could continue to be used
domestically until the end of their useful lives. The licensee would
not need to demonstrate that the packages satisfy the new crush test or
deep-immersion test. The comments would take that provision one step
further and require the crush and deep-immersion tests only for
international use packages.
NRC believes that the international package standards should be
used by the United States for both domestic and international
shipments, to the extent practicable. However, based on a history of
safe use under earlier safety standards, and the absence of unfavorable
operational data, NRC will allow the continued use of existing packages
in domestic transport until the end of their useful lives. NRC will not
allow, however, the continued fabrication of packages to the old
designs. This action permits use of existing packages. It does not
perpetuate package designs that can be discarded or upgraded to satisfy
the new standards.
Another comment suggested grandfathering the existing Type A casks
now approved for transporting Type B quantities of LSA radioactive
material, until the Type B waste casks required to satisfy the new
standards become available. NRC has adopted the suggestion, extending
the proposed provisions in Sec. 71.52, ``Exemption for low-specific-
activity (LSA) packages,'' to a 3 year period, to give the industry
time to design, receive approval, and fabricate new Type B waste
packages.
Section 71.22 General license: Fissile Material, Limited Quantity,
Controlled Shipment
One comment requested clarification as to whether the Type A limit
imposed in Sec. 71.22(c) also applies to Sec. 71.22(d).
The requirements of Secs. 71.22(a) through 71.22(e) are cumulative,
each imposing additional requirements on the use of the general
license. The radioactivity limit and mass limits of Sec. 71.22(c) apply
to packages, whereas the mass and mass ratio limits of Sec. 71.22(d)
apply to shipments.
A comment noted an error, in Sec. 71.22(d)(3), which changed the
intent of the section. The commenter suggests that the phrase ``exceeds
unity'' at the end of Sec. 71.22(d)(3) be replaced by the phrase ``does
not exceed unity.'' NRC agrees and has made that change.
Section 71.24 General License: Fissile Material, Limited Moderator,
Controlled Shipment
One commenter asked if the statement in Sec. 71.24(b), ``* * * a
quality assurance program approved by the Commission as satisfying the
provisions of Subpart H of this part,'' is any different from ``* * * a
quality assurance program approved by the Commission.'' The two
statements are different in that the first is more specific and
provides more detail. There are several different quality assurance
programs, in different licensing areas, approved by the Commission.
Specifying that the program must satisfy Subpart H makes it clear as to
the type of quality assurance program is required.
One commenter recommended inserting ``by weight'' after ``1
percent'' in Sec. 71.24(c)(6). NRC agrees and has made this change in
Sec. 71.24(c)(7), as well.
With respect to a general license for a package containing fissile
contents, one commenter requested clarification of what is meant by
``no uranium-233'' in Sec. 71.24(c)(6). For a general license under
Sec. 71.24(c)(6), a package containing fissile contents must have no
detectable U-233. The method for making this determination can be
decided by the licensee. For example, the licensee can make this
determination by performing an assay or by knowing the history of the
material.
Subpart D--Application for Package Approval
One comment suggested changing the title of Subpart D to
``Application for Type B Package Approval'' for clarity. Because NRC
also approves Type A packages for fissile material, the title of
Subpart D continues to refer to ``Package Approval.''
Section 71.38 Renewal
One comment suggested that NRC provide some administrative
acknowledgment when a timely application for renewal of a certificate
of compliance has been received to provide proof that timely renewal is
in effect. The Commission does not believe that proof of timely renewal
is particularly important and that providing an acknowledgment to each
registered user of a package would be too burdensome for the benefit
gained.
Section 71.43 General Standards for All Packages
Four comments suggested the addition of IAEA regulations relating
to packaging of liquids and gases to Part 71, including those
pertaining to the special free drop and penetration tests
[[Page 50255]]
for liquids and gases. The NRC approves only Type B and fissile
material packages. The NRC also notes that fissile material packages
must be evaluated for hypothetical accident conditions more severe than
the tests for liquids. Furthermore, there are currently no NRC-licensed
packages designed for gaseous fissile materials and NRC does not
anticipate any future applications for such packages. These additional
provisions would complicate regulations that are presently adequate.
IAEA standards on absorbent material and double containment have been
selectively included in DOT regulations.
Eight comments disagreed with the NRC view that Sec. 71.43(f)
should continue to restrict to ``no significant increase'' any change
in external surface radiation levels, as a result of subjecting a
package to the defined normal conditions of transport. The comments
argued that the 20 percent increase specified in IAEA regulations is a
safe, reasonable, and practical number that could not reasonably be
lower, and that specifying a value in the rule provides the package
design engineer and the NRC review engineer a measurable goal that is
consistent both with IAEA and with engineering practice.
Type B and fissile material packages can be readily designed so
that normal conditions of transport result in no significant increase
in dose rates, and that a twenty percent increase in dose rates because
of normal handling is excessive. In addition, if a package were
designed so that the external dose rate could increase 20 percent
during normal handling, the package could exceed the dose rate limits
in Sec. 71.47 during transport, and would be an item of non-compliance.
NRC and DOT have therefore decided to not adopt the IAEA ``20 percent
increase'' provision, and to retain the current ``no significant
increase'' provision.
Four comments suggest the addition of the special provisions of
IAEA regulations pertaining to the transportation of radioactive
material by the air mode. NRC has determined that special requirements
for transport of packages by air should be excluded from Part 71
because these provisions are properly incorporated in the carrier
restrictions imposed by the Department of Transportation.
Two comments suggested that the phrase ``Account must be taken of
the behavior of materials under irradiation'' be clarified and
quantified, perhaps in a regulatory guide, or deleted from Part 71.
Although there is no regulatory guidance now available relating this
requirement to transportation packages, it is clear that any effects of
irradiation on materials used in the package must be taken into
account. These effects could be the accelerated aging or embrittlement
of elastomers or elastics and may result in requiring a frequent change
of gaskets, for example.
One comment suggested the performance requirement of Sec. 71.43(f)
be changed to include a numerical sensitivity for the requirement that
there be ``no loss or dispersal of radioactive contents'' as a result
of subjecting a package to the specified normal conditions of
transport. The equivalent paragraph in the IAEA regulations for Type A
packages is paragraph 537, and does not contain a numerical
sensitivity. Paragraph 548, of IAEA Safety Series #6, is the equivalent
of 10 CFR 71.51, for Type B package leaktight sensitivity. Both those
provisions require Type B packages to be leaktight to a sensitivity of
10-6 A2/h.
Three comments noted that IAEA no longer prohibits continuous
venting of packages in its 1985 edition and urged the NRC to allow the
practice domestically for Type B packages. The commenters argued that
although NRC took a strong position, in the preamble to the proposed
rule, that continuous package venting is ``poor engineering practice,''
NRC did not explain why. The commenters noted that DOT regulations do
not prohibit continuous venting for Type A packages, leaving the
acceptability of continuous venting to be decided by performance
requirements. The commenters stated that in some cases it would make
good sense to allow continuous venting to provide pressure equalization
and discharge of organically generated hydrogen gas.
NRC is continuing its ban on continuous venting of Type B packages
for the following reasons:
1. Venting of a package containment system during normal conditions
of transport defeats the purpose of the containment system;
2. It is practical to design packages that do not rely on venting,
to relieve pressure under normal conditions of transport;
3. The use of a vent does not necessarily prevent the generation of
potentially flammable or explosive gas mixtures; and
4. The reliability of filters under temperature extremes, varied
operating conditions, and sustained service has not been established.
Two comments stated that Mo-99/Tc-99m radiopharmaceutical
generators are open to the atmosphere to allow changes in ambient
pressure and that the generators do not vent radioactive material. The
comments recommended that the prohibition against venting be limited to
venting radioactive material only and that NRC continue current
practices.
NRC believes these comments arise from concern over the reduction
in the A2 quantity for Mo-99 from 20 curies to 13.5 curies in the
proposed rule. NRC recognizes that the shipment of Mo-99/Tc-99m
generators is a special case, and is retaining the 20 curie A2
value for Mo-99, to permit the continuation of current practices.
Section 71.47 External Radiation Standards for All Packages
NRC used the term ``accessible external surface'' in its proposed
rule for determining radiation levels on package surfaces, whereas DOT
used the term ``external surface'' in its proposed rule. Four comments
argued that the NRC and DOT regulations for radiation level limits on
package surfaces should be identical. Most believed that a limit on
accessible surfaces was the more reasonable standard.
DOT has indicated that it is considering a petition for rulemaking
to add the word ``accessible'' to its radiation level regulations and
will consider that complex issue in a separate action. Pending
completion of the DOT separate action, NRC has deleted the word
``accessible'' from this section of the final rule but does not intend
to alter its practices regarding this provision.
One comment stated that this paragraph tends to be confusing in
that it establishes a limit of 2 mSv/h (200 mrem/h) for package surface
radiation levels, yet Sec. 71.47(b)(2) seems to state that packages
transported on a flatbed trailer can exceed 2 mSv/h (200 mrem/h),
provided the radiation level at the planar edges of the trailer is less
than or equal to 2 mSv/h (200 mrem/h).
Section 71.47 establishes a generally applicable 2 mSv/h (200 mrem/
h) Package surface radiation-level limit. The section further
establishes that, if a package is shipped as exclusive use, the
radiation level may exceed 2 mSv/h (200 mrem/h), provided the
applicable provisions of paragraphs (a) (with repect to Transport
Index) through (d) are met. Paragraph (b)(2) restricts the radiation
level at any point on the vertical planes projected by the outer edges
of a flat-bed style vehicle to 2 mSv/h (200 mrem/h) (the same limit
imposed in paragraph (a) for the outer surfaces of closed transport
vehicles). Thus, provided packages are shipped as exclusive use,
external radiation levels may exceed 2 mSv/h (200 mrem/h) at the
surface of packages on flatbed trailers, but not at the outer-edge
planes of the vehicle.
[[Page 50256]]
Section 71.51 Additional Requirements for Type B Packages
One comment suggested that the clarifying provision following
paragraphs 548(a) and (b) of IAEA regulations be added to Part 71 for
consistency. The clarifying provision pertains to allowable releases of
radioactive material from a package containing a mixture of
radionuclides. This is the case, for example, with spent nuclear fuel
casks. That clarifying provision has been added.
Section 71.52 Exemption for LSA Packages
Twelve comments expressed concern that the proposed Part 71 affords
only a 1-year delay in applying the new LSA rules. NRC established the
1-year delay to give the industry an opportunity to design and build
the Type B waste casks that would be required under the new rules. The
comments uniformly argued that 1 year was not a sufficient period of
time to design a waste cask, to have it reviewed and approved by NRC,
and to fabricate an adequate number of casks, to approved designs, that
satisfy the needs of the new LSA rule. The commenters differed in how
long they thought that process would take, varying over 2, 3, and 5
year periods. NRC agrees with the thrust of this comment and has
established the exemption period at 3 years. Thus existing packagings
may be used for 3 years and new packagings may be fabricated from
existing designs for 3 years.
A consequence of establishing the IAEA LSA/SCO package limit as the
delineator between NRC and DOT regulation of LSA and SCO packaging [see
Sec. 71.10(b)(2)] is that, after the 3 year exemption period, LSA will
be shipped either in DOT authorized packagings, or in NRC certified
Type B packagings. Accordingly, NRC is discontinuing the practice of
certifying Type A LSA packages. NRC has therefore not adopted a
proposed exemption (Sec. 71.52(a)) that only would have applied to NRC
certification of new Type A LSA package designs.
One comment stated that the demand for waste casks would rise until
1993 and then fall again because few of the low-level radioactive waste
disposal site compacts will permit disposal access. Vendors will
hesitate to invest in casks that will not be used after 1993 and waste
will need to be stored onsite.
NRC is unwilling to accept this proposition and believes that as
long as NRC specifies the requirements for transportation of waste,
given adequate time, industry will continue to develop disposal
options.
One comment argues that the specific reference to Sec. 71.43(f)
should be deleted because it is included in the broader reference to
Secs. 71.41-71.47.
Section 71.52 exempts exclusive use LSA and SCO packages from the
additional requirements for Type B packages for a period of 3 years
from the effective date of the final rule. These LSA packages are still
subject to other requirements that apply to all packages. The referral
to these other package requirements includes Secs. 71.41-71.47, plus a
specific reference to. An argument could also be made for deleting the
entire reference because those requirements apply regardless of the
reference in this section. However, NRC chose to include the reference
in Sec. 71.52 as a reminder that the exemption is only from Sec. 71.51,
not from all packaging requirements. NRC believes the reference to
Sec. 71.43(f) (normal conditions of transport tests) is important and
has decided that it will be retained.
One comment suggested that SCO be included within the scope of
Sec. 71.52, and that the 2A1 limit be included in the section for
clarity. NRC agrees with the comment and has made the clarifications,
substituting the IAEA LSA limit for 2A1.
Section 71.53 Fissile Material Exemptions
One comment suggested spelling out the word ``liter'' instead of
using ``l'' as the abbreviation. Considering the typing errors caused
by the use of that abbreviation, the final rule spells out the word
``liter'' wherever it appears.
Section 71.55 General Requirements for Fissile Material Packages
One comment suggested that by adding the word ``full'' to the water
reflection criterion of Sec. 71.55(b)(3), the NRC has added more cost
with no apparent benefit ``* * * since transport limits already take
this consideration into account.'' The latter part of this comment
probably refers to the ``transport index'' controls that limit the
number of packages which can be transported and stored together, but do
not consider the safety of an individual package in isolation. Addition
of the word ``full'' in Sec. 71.55(b)(3) is a matter of clarification.
NRC has always required ``full'' reflection wherever reflection is
required. IAEA regulations required ``full'' reflection in the 1973
edition, and go a step further in the 1985 edition, to define ``full''
as ``water 20-cm thick (or its equivalent).'' NRC has retained the word
``full,'' in Sec. 71.55(b)(3), and has added the word ``full,'' in
Sec. 71.55(e)(3), for consistency.
A commenter agrees that the proposed Part 71 begins to simplify the
system of shipping fissile material but that most of the difficulties
still exist. The commenter advocates development of ``a system of
performance-oriented packaging,'' to reduce the current complexity of
the ``design-oriented package choices.'' NRC agrees that there are a
number of radiation control design requirements that apply to the
fissile material packages as well as to packages of other radioactive
material. However, NRC views the criticality control provisions as
performance-oriented rather than design-oriented. NRC must specify the
conditions against which the package must be designed. Without the
environmental tests and package objectives, there would be no level of
protection against which to design packages.
Section 71.61 Special Requirement for Irradiated Nuclear Fuel
Shipments
One comment recommended that the rule clarify that the deep
immersion test is to be applied to an otherwise undamaged package. This
important detail is implied, but not specifically stated. The
Commission agrees and has made that clarification.
In the final rule, this section has been modified to require that
the external pressure test be applied directly to the containment
system of a package. NRC does not believe the external structure should
play a part in helping the containment system of a package withstand an
external pressure test and has chosen to ignore its existence in
specifying the requirement.
A comment recommended that the word ``rupture,'' as used in this
requirement, be defined as a gross structural collapse and not just an
inleakage of water. Although the word ``rupture'' in the proposed rule
did mean gross structural collapse, NRC has since decided that the term
``rupture'' cannot be determined by engineering analysis. NRC has
decided to change the acceptance criteria for the deep immersion test
from ``rupture'' to ``collapse, buckling, or inleakage of water.''
A comment stated that this requirement should include the 1-hour
time specification included in the IAEA requirement to avoid later
misinterpretation of the test. The NRC agrees that adding the 1-hour
test specification would help prevent confusion between IAEA and
domestic regulations, and has included the time specification.
[[Page 50257]]
A comment noted that the term ``at least'' is used two times in the
proposed requirement, thereby creating an opportunity for
misinterpretation. Although the term is used in the IAEA text, the NRC
agrees with the commenter that it serves no useful purpose and has
deleted the term.
A comment stated that the deep-water immersion test should be
clarified to ensure that an engineering evaluation is an acceptable
alternative to a physical test because an actual 200-m test would be
costly and difficult. NRC believes it is clear that an engineering
evaluation is acceptable because the equivalent external gauge pressure
is specified in the text of the requirement. The provisions of
Sec. 71.41(a) are intended to allow the use of engineering evaluations
when they are reasonably applied.
The remaining three comments relating to this section all deal with
transition periods and special provisions for casks for which there
will be no further fabrication and that are not used internationally.
The earlier portion of this preamble dealing with the provisions of
Sec. 71.13 presents the NRC view on these matters.
Section 71.63 Special Requirements for Plutonium Shipments
Four comments argued that the extension of this provision to
radionuclides other than plutonium is unjustified and that the
provision, even without the extension to other radionuclides, differs
from IAEA rules and is inconsistent with the principles of IAEA rules.
Two of the commenters argued further that the existing provisions, if
examined in the light of current regulatory analyses, probably could
not be justified.
NRC recognizes that some requirements have been added to the
regulations over the years strictly on the basis of prudent judgment.
Because the basis for current rules is not a part of this rulemaking
action, NRC will simply refrain from extending the present rule to
other radionuclides.
One commenter argued that the rule should be rewritten using
multiples of the A2 values, not only to define radionuclides
subject to the rule, but also to define the level of activity at which
the extra requirements come into effect. Because the extension to other
radionuclides is being withdrawn, the inclusion of A values does not
appear to improve the requirement.
Section 71.71 Normal Conditions of Transport
Three comments noted that the provision of IAEA's paragraph 528
requiring consideration of a temperature range from -40 deg.C to +70
deg.C for the components of the packaging is not reflected in Part 71.
NRC omitted this provision because NRC does not want to limit the high
end temperature consideration to 70 deg.C because that would imply
that +70 deg.C is the highest temperature that has to be considered
for package design. This does not take into account the considerably
higher temperatures resulting from decay heat in certain Type B
packages.
Three comments noted that 10 CFR 71.71(c)(4) prescribes an
increased external pressure specification of 140 kPa absolute but IAEA
regulations do not have that exact requirement. NRC believes there is a
need for an external pressure test for normal conditions to ensure that
a package filled at low pressure or high altitude will withstand an
external pressure increase. The additional pressure test has been
retained.
Three comments observed that Sec. 71.71(c)(7) states that the free
drop test be conducted between 1.5 and 2.5 hours after the conclusion
of the water spray test but the same requirement is not included in the
IAEA regulations. The IAEA rules, however, do include restrictions, in
paragraph 620, on the timing of the mechanical tests after the water
spray test. NRC has retained the water spray test as is and believes
the NRC test meets the intent of the IAEA test.
One comment noted that with the deletion of the fissile classes,
the corner drop test, which was required only for Fissile Class II
packages, is proposed to be applied to all fissile packages. The
commenter argued that for a large and heavy package, such as a spent
fuel shipping cask, ``it is considered highly implausible for a package
to undergo a one-foot corner drop as a normal condition of transport.
Only a free drop with the package in its normal orientation should be
specified as a normal condition of transport for large and heavy
packages, therefore saving valuable analysis effort and time.''
NRC agrees with the comment and has deleted the corner drop test
for fiberboard, wood, or fissile material rectangular packages weighing
more than 50 kg (110 lb), and for fissile material cylindrical packages
weighing more than 100 kg (220 lb). For these packages, NRC does not
believe that the corner drop tests are significant in developing a safe
fissile material package.
Section 71.73 Hypothetical Accident Conditions
One comment stated that reversing the order of the two immersion
tests in Secs. 71.73 (c)(5) and (c)(6) would restore the order of the
tests, which must be run consecutively, and would therefore clarify the
text. NRC agrees and has made the change.
One comment recommended that the temperature extremes specified for
the initial test conditions in Sec. 71.73(b) be given a reasonable
tolerance because ambient air temperatures cannot be controlled. NRC
agrees that temperatures, as with other required parameters of the test
conditions, cannot be accurately controlled. NRC's position, however,
is not to establish tolerances, but to require that the effects of test
conditions different from those specified be analyzed as part of the
overall evaluation. Every analysis would then be normalized to the same
set of specifications.
One comment recommended that the word ``single,'' in the second
line of the thermal test in Sec. 71.73(c)(4), should be ``simple''. NRC
agrees and has made that change.
Two comments asked that NRC include some information as to how the
effects of solar radiation should be treated. One comment stated, ``The
solar insolation can be a significant factor and should be consistently
evaluated.'' Others have argued that the effects of solar insolation
are insignificant compared with the thermal effects of the fire test
and should be ignored.
NRC adopts the view of the thermal experts who participated in
developing the IAEA regulations. Those experts thought the effects of
solar radiation may be neglected before and during the thermal test but
that such effects should be considered in the subsequent evaluation of
the package response.
One comment recommended the development of guidance on how
designers should interpret the revised thermal test requirement.
Although there is guidance provided in the IAEA's companion documents
to its transportation regulations (IAEA Safety Series No. 7,
``Explanatory Material for the IAEA Regulations for the Safe Transport
of Radioactive Material--1985 Edition,'' and IAEA Safety Series No. 37,
``Advisory Material for the IAEA Regulations for the Safe Transport of
Radioactive Material--1985 Edition''), further guidance may be
necessary. If so, it is the industry that can best propose guidance,
based on its capabilities. If coordinated under the auspices of the
American National Standards Institute (ANSI), Committee N-14, with NRC
representation, there is a good chance that a consensus standard could
be developed that could be endorsed by NRC as a satisfactory means to
satisfy regulatory requirements.
[[Page 50258]]
One comment stated that packages that are subjected to the crush
test should not also be subjected to the 30-foot free drop test, as
required in the proposed rule. Instead, consistent with IAEA, the crush
test should be in lieu of the 30-foot free drop test.
NRC believes that the crush test and the free drop test impart
different types of loadings onto the package. Having sufficient crush
resistance for the crush test does not ensure the adequacy of the
package under the inertial loadings that occur during the 30-foot drop
tests. NRC believes that it is important for packages to have
resistance to impact and that the crush test should not be a substitute
for the impact test.
One comment stated that a crush scenario is not likely during
``dedicated'' shipments because heavy loads are not placed above the
shipment at any time during transport. The comment questioned the
applicability of the test for dedicated shipments, and requested that
at least an engineering evaluation be allowed as an alternative to a
physical test. NRC has made it clear (see Sec. 71.41) that appropriate
analyses may be used to demonstrate the ability of a package to meet
crush test conditions.
Section 71.75 Qualifications of Special Form Radioactive Material
One comment indicates that changes in Sec. 71.75(a) from the
current rule have changed the concept of special form from being a
provision for special properties of the radioactive material contents
of the package to being a provision for special properties of the
package--a change from qualifying a ``special form source'' to
qualifying a ``special form package.''
NRC regrets the confusion, but intended no substantive change to
the concept of special form. Special form criteria in this final rule
have been brought closer to those of DOT, but still without any basic
changes.
One comment noted that the reference in Sec. 71.75(e)
[Sec. 71.75(d), in the final rule], to a standard of the International
Standard Organization (ISO) is vague and should be made more specific.
Although the ISO standard could be written in all its detail in
Part 71, rather than simply referenced there, most comments over the
years have encouraged NRC to have less repetition and more simple
references to other requirements.
Section 71.83 Assumptions as to Unknown Properties
One comment pointed out an error in line 7 of Sec. 71.83, where the
proposed rule referred to ``known properties'', where it should have
referred to ``unknown properties.'' That error has been corrected.
Section 71.85 Preliminary Determinations
One comment recommended that the term ``durable'' in the context of
``durably mark the packaging,'' as in Sec. 71.85, be defined in terms
of the conditions that the markings on the packaging must be able to
withstand. When developing its regulations, NRC must decide at what
level of detail they are to be written. Sometimes that level of detail
is changed as a result of experience if a widespread misuse of a
standard becomes known because of a lack of detail. NRC is not aware of
any problem with the term ``durably,'' even though it has been used
since 1968 in the preliminary determinations section. In the absence of
a significant problem, NRC prefers to leave the term as is.
Section 71.87 Routine Determinations
One comment recommended that NRC's Table V ``Removable External
Radioactive Contamination Wipe Limits,'' be used by DOT in place of its
Table 11. NRC notes that the only significant difference between the
two tables is that the term ``low toxicity alpha emitters'' is replaced
by its definition in the NRC table. The NRC final rule simply refers to
the DOT requirement (49 CFR 173.443) for maximum permissible
contamination limits.
Section 71.88 Air Transport of Plutonium
One comment recommended that the forward tie-down specification of
9 g detailed in Sec. 71.88(c)(2) be reduced to 1.5 g for plutonium
packages transported on a Boeing 747 aircraft. The reason for this
recommendation has to do with the 14 CFR 25.561 regulatory requirement
of the Federal Aviation Administration (FAA), that the supporting
structure of an airplane must be designed to restrain, up to specified
inertial forces, including 9-g in the forward direction, ``* * * each
item of mass that could injure an occupant if it came loose in a minor
crash landing.'' NRC, in prescribing tie-down requirements for
plutonium packages in aircraft, took note of the supporting structure
requirements of the FAA and required a 9-g tie-down system for the
package on the main deck of the aircraft. The Boeing 747 cargo
aircraft, however, with no passengers and the cockpit located above the
main deck, is not subject to the requirements of 14 CFR 25.561 because
there are no occupants to injure if ``* * * the package came loose in a
minor crash landing.'' Thus, the Boeing 747 ``Weight and Balance
Manual,'' DG-13700, shows a load factor of 1.5 g in the forward
direction.
The purpose of the NRC tie-down requirement was not to protect
occupants of the aircraft from cargo that has come loose in a minor
crash landing. Therefore, the comparison with the FAA supporting
structure requirement is not germane. The purpose of the NRC
requirement was to protect the plutonium package from the uncontrolled
potential for damage inherent in having the package unrestrained in a
crash landing.
Paragraph (c) of Sec. 71.88 proposed a requirement that the
licensee make special arrangements with the carrier on where to place
the plutonium cargo in the aircraft, how to tie it down, and what
restrictions are to be placed on other cargo. Recognizing that these
restrictions would be more appropriately placed directly on the carrier
rather than through the shipper, the DOT has placed these restrictions
in its air carrier regulations (Sec. 175.704 of 49 CFR Part 175,
``Carriage By Aircraft.'') These regulations are now referenced in
Sec. 71.88.
Section 71.95 Reports
All three public comments on this section were directed at the
newly proposed provisions of paragraph (c), which require a 30-day
report of ``* * * instances in which the conditions of approval in the
certificate of compliance were not observed in making a shipment.''
One comment requested clarification whether Sec. 71.95(c) applies
to shippers or receivers.
The scope of Part 71 (Sec. 71.0(c)) makes the regulation applicable
only to shippers of radioactive material. Therefore, Sec. 71.95(c)
applies only to shippers of radioactive material. However, shipment
deficiency may be detected by the receiver of the shipment. If the
receiver reports that deficiency to the shipper, the shipper is
obligated to report it to NRC. Further, note that 10 CFR Part 21,
``Reporting of Defects and Noncompliance'', is applicable to receiving
facilities.
The other two comments dealt with the substance of the event that
would prompt the report. One suggested the regulation be more specific
on conditions that would require a report. The second comment suggested
that the report include the consequences of the deficient shipment such
as radioactive contamination, a loosened sealing cap, etc.
Although both of these suggestions have merit, neither has been
[[Page 50259]]
incorporated in the final rule. The purpose of the requirement is to
provide feedback to NRC on quality assurance program effectiveness by
an indication of the number and type of packaging and other mistakes
and on the safety significance of those mistakes by an indication of
the mistake consequences. NRC believes the reporting requirement should
retain its broad scope. A large number of reports is not expected. NRC
also believes that individual follow-up is the only reasonable way to
uncover any procedural deficiency that might cause mistakes.
One comment questioned whether this type of report is important
enough to be required within 30 days. NRC judges that the timing is
about right, and expects the staff's review of submitted reports to be
completed within a similar time frame.
Section 71.97 Advance Notification of Shipment of Irradiated Reactor
Fuel and Nuclear Waste
Of the five comments submitted on this notification requirement,
two suggested changing the value for the number of curies in
Sec. 71.97(b)(3)(iii), so it corresponds to the same limit in the
regulations of DOT and IAEA. That change has been made.
The other three comments stated that this requirement was not
clearly expressed. The requirement has been reorganized in the final
rule, and consists of the following parts:
1. Paragraph (a) provides a broad general requirement that
licensees pre-notify governors of States of any shipments of
radioactive material going to, through, or across the boundary of the
State;
2. Paragraph (b) limits the prenotification requirement to certain
types of shipments. All the conditions of paragraph (b) must be
satisfied for the prenotification requirement to apply. The licensed
material must be required to be in a Type B package, limiting the
requirement to shipments of relatively high potential hazard. The
shipment must be destined to a disposal site or to a collection point
for transport to a disposal site, further limiting the requirement to
waste material. The quantity of radioactive waste in a single package
must exceed the limits specified in the DOT regulations for highway-
route controlled quantities. Lastly, for irradiated fuel, the quantity
contained in a single package must be less than that subject to the
similar advance notification requirement of 10 CFR 73.37(f).
3. Paragraphs (c), (d), (e) and (f) contain the details for timing,
information in the notification, revisions, and cancellation.
One comment noted that from the wording in Sec. 71.97(a), a reader
would expect to find exceptions in Sec. 71.97(b). The comment notes
that the provision does not contain exceptions. NRC agrees with this
comment and has revised Sec. 71.97(a) for clarity.
One comment questioned the value of proposed Sec. 71.97(b)(4)
[Sec. 71.97(b) in the final rule] which required that ``* * * the
quantity of irradiated fuel is less than that subject to advance
notification requirements of Sec. 73.37(f) of this chapter.'' Paragraph
73.37(f) refers to a separate part of the Commission's regulations, 10
CFR Part 73, ``Physical Protection of Plants and Materials,'' and
imposes an advance notification requirement for irradiated fuel
shipments similar to the one under discussion. The scope of Part 73
(see Sec. 73.1(b)(5)) limits its applicability regarding shipments of
irradiated reactor fuel to ``* * * quantities that in a single shipment
both exceed 100 grams in net weight of irradiated fuel, exclusive of
cladding or other structural or packaging material, and have a total
radiation dose rate in excess of 100 rems per hour at a distance of 3
feet from any accessible surface without intervening shielding.'' If
the quantity of irradiated fuel in a shipment exceeded the quantity
specified in Sec. 73.1(b)(5), the notification would be made under
Sec. 73.37(f). If not, the notification would be made under Sec. 71.97.
The proposed provision in Sec. 71.97(b)(4) was intended to prevent
duplicate notifications for some shipments.
The final comment on Sec. 71.97 included a clear rewrite of
Sec. 71.97(b) that has been used in its entirety in the final rule.
Comments on Appendix A
Five comments supported the inclusion of new radionuclides in Table
A-1 of Appendix A as useful and justified. Five other comments pointed
out errors and inconsistencies between NRC and DOT for the A1/
A2 values in Table A-1. These inconsistencies have been corrected
in the NRC and DOT final rules.
Three comments recommended a grandfathering provision for the
continued authority to transport molybdenum (Mo) 99/technetium (Tc) 99m
generators, in Type A packages, with radioactivity between the current
A2 value of 20 Ci and the new A2 value of 13.5 Ci for Mo-99.
The lower A2 value is the result of a new dosimetric model, for
beta-emitting radionuclides, to address skin contamination. In the
preamble to the NRC proposed rule, the NRC noted, with respect to the
changes in the A1 and A2 values:
Based on our most current knowledge of radioactive material
shipments in the United States, the economic impacts of these
changes are not likely to be large. However, any situations where a
potential exists for significant economic impacts as a result of
changes in the A1 or A2 values should be brought to the
NRC's attention in public comments.
NRC agrees that this is a situation where health care in the United
States could be significantly impacted as a result of forcing the
larger quantity Mo-99/Tc-99m generators now transported in Type A
packages into Type B packages. In view of the favorable experience over
the years with these generators, NRC and DOT will allow the continued
domestic transportation of generators that contain up to 20 Ci of
radioactive material in Type A packages.
Two similar proposals to grandfather the transportation of carbon-
14, phosphorus-32, sulfur-35, and iodine-125 at existing levels were
not as persuasive and have not been adopted. The decrease in A1
and A2 values would apparently force many shipments out of the
``limited quantity'' category, where they are excepted from
specification packaging, shipping papers and certification, and marking
and labeling requirements, and into the ``Type A'' category.
Although there are clearly more packaging and communication
requirements associated with the ``Type A'' category than with the
``limited quantity'' category, NRC does not view that change as
creating the same economic impact as a change from the ``Type A'' to
the ``Type B'' category.
One comment suggested that the radionuclides einsteinium-253 and
einsteinium-254 be added to Table A-1 because shipment of those
transuranics are increasing in number and the default values are not
expected to be adequate. NRC has added those radionuclides and will
also propose them for addition to the IAEA regulations. Until they are
included in IAEA Safety Series No. 6, however, multilateral approval is
required for international shipments. This limitation is identified by
footnote in Table A-1.
One comment objected to having to obtain NRC approval of A1/
A2 values that are not in Table A-1. In addition to NRC approval,
international shipments require multilateral approval of A values that
are not included in the IAEA regulations by each country through or
into which the consignment is to be transported. The development of A
values may not be a simple matter, requiring consideration of daughter
[[Page 50260]]
radionuclides and differing radioactive emissions. Although a competent
health physicist or nuclear engineer should not have too much
difficulty determining an A value, NRC must assure that a system exists
to protect against faulty determinations. Use of the conservative A
values from Table A-2 does not require regulatory approval.
One commenter questioned the unlimited values, for A1 and
A2 in Table A-1, for uranium-235 enriched less than 5 percent. The
comment argued that U-235 is a fissile material and the unlimited
values may not be appropriate. The A1/A2 values are for
radiological, not fissile, considerations. The A1/A2 values
set the maximum quantity of radioactive material that can be shipped in
a Type A package (except for LSA); other package characteristics, such
as heat generation, weight, criticality, external radiation, etc., can
further limit the quantity of radioactive material in that Type A
package. Limitations with respect to fissile characteristics, for
example, are addressed in Secs. 71.53, 71.55, and 71.59. NRC has
decided to add a clarifying note, currently in the IAEA regulations, to
the A1/A2 Table in Appendix A of Part 71. The Appendix A note
reads ``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.''
Finally, one comment suggested that we eliminate the specific
activity column from Table A-1. The comment argues that ``Specific
activity information is not required or explained in the regulations,
and it is difficult to keep the information accurate.''
Although the NRC is in basic agreement with the comment and would
have no problem in eliminating the specific activity data from Part 71
if there were a good source of comparable data available for the times
it is needed to implement the transportation regulations. NRC is not
familiar with any good substitute source. Though IAEA Safety Series No.
37, ``Advisory Material for the IAEA Regulations for the Safe Transport
of Radioactive Material (1985 Edition),'' third edition, published in
June 1987, includes a table of half-lives and specific-activities,
there is no indication yet of a system of periodic reviews that would
keep that information up to date.
Comments on Draft Regulatory Analysis
Ten persons commented on the impacts associated with the proposed
changes to limit the content of LSA/SCO packages to 2A1. The main
thrust of these comments is that the impacts are much greater than
presented. In part in response to these comments, NRC has adopted in
the final rule the IAEA LSA/SCO package limit of 10 mSv/h (1 rem/h) at
3 m, in lieu of the proposed 2A1 limit.
Because the NRC data base for determining the additional shipments
expected to be caused by the proposed rule dated back to 1980, and
because a clear preference was developing in the public comments for
the IAEA radiation level limit rather than the 2A1 limit, NRC
repeated its analysis using more recent data. An NRC contractor
gathered 1989 data from the 3 shallow land burial facilities for all
waste shipments of resins, evaporator bottoms, and filter media. The
contractor analyzed the characteristics of those 4600 Type A cask
shipments and found that approximately 150 of those shipments would
have exceeded the IAEA limit. NRC assumes that each shipment exceeding
the limit is split into 2 shipments due to the smaller capacity of Type
B packaging. Thus 150 additional shipments are caused by the LSA limit.
The impacts of preparing additional packages of LSA waste for
shipment and receiving those additional shipments at the burial ground
were absent from the draft regulatory analysis. One comment advised the
NRC of the results of an exposure study which concluded that the extent
of the collective exposure for preparation and receipt of waste casks
was approximately 0.5 person-rem per shipment. The NRC noted that half
of the 0.5 person-rem per shipment factor multiplied by the 4600 waste
cask shipments per year from the new data base corresponds fairly well
to a large portion of the 1726 person-rem collective exposure reported
for all light water reactors for 1986 under the category ``waste
processing'' by Barbara G. Brooks, NRC, and D. Hagemeyer, SAIC in
NUREG-0713, Vol. 8, dated August 1989 (this version was current at the
time the contractor prepared the regulatory analysis). On the basis of
this data, NRC has accepted the 0.5 man-rem per shipment number as a
reasonable estimate. Multiplying that 0.5 man-rem per shipment
conversion factor by the 150 additional shipments which the limit of 1
rem per hour at 3 meters would cause, the effect of the limit would be
75 person-rem per year.
Because the IAEA LSA provisions permit a greater quantity of LSA/
SCO material to be shipped in a package, fewer packages and shipments
are needed to transport a given quantity of material. The estimated
burden on industry from the final rule is therefore less than that for
the proposed rule. The NRC draft regulatory analysis dated November,
1987 developed industry costs resulting from a 2A1 limit on LSA
shipments of $1.7 million per year. These costs consist of package
costs and shipment costs resulting from an estimated 311 additional
cask shipments per year. Through the same simple modeling used in the
older analysis, the new NRC regulatory analysis shows increased dollar
costs associated with the 150 additional LSA/SCO shipments of $1.0
million per year. These estimates include differential package costs
and differential shipping and handling costs, annualizing and summing
each component. These estimates do not include cost components
recognized but not quantified in the public comments as training,
procedure revisions, computer program changes and upgrades, insurance
premiums, and disposal costs.
There were no significant comments related to the projected number
of non-radiological deaths and injuries associated with the increased
shipments caused by the new standards.
Agreement State Compatibility
Section 274d.(2) of the Atomic Energy Act of 1954, as amended,
requires that before entering into an agreement with any State, the
Commission shall make a determination that the State's program is
compatible with the Commission's program. Section 274g authorizes and
directs the Commission to cooperate with the States in the formulation
of standards to assure that State and Commission programs will be
coordinated and compatible. The basic objective of NRC's State
Agreements Program has been to achieve uniformity among the various
programs to the maximum extent practicable recognizing that the States
must be allowed some flexibility to accommodate local conditions. Under
this Program, procedures have established criteria for better defining
compatibility, and for determining the degree to which States
regulations must show uniformity with Commission regulations. In
practice, the Commission's regulations are categorized as Division 1-4
Rules according to the degree of State regulation uniformity required,
as summarized in the following table:
------------------------------------------------------------------------
Division Agreement State regulation uniformity
------------------------------------------------------------------------
1............ Agreement States are expected to adopt, essentially
verbatim, the regulation to provide consistency between
Federal and State requirements.
[[Page 50261]]
2............ Agreement States have the flexibility to adopt similar or
more stringent requirements based on their radiation
protection experience, professional judgements, and
community values.
3............ Agreement States should adopt the requirement, but there
is no degree of uniformity between NRC and Agreement
States required.
4............ Agreement States should not adopt the requirement since
these are regulatory functions reserved to NRC.
------------------------------------------------------------------------
The final rule does not affect the current compatibility
categorization of Part 71 regulations. The following table lists the
Part 71 Sections and corresponding rule categorization (Division 1-4):
----------------------------------------------------------------------------------------------------------------
Division Section Title
----------------------------------------------------------------------------------------------------------------
1......................... 71.4...................... Definitions.
1......................... 71.5...................... Transportation of Licensed Material.
1......................... 71.10..................... Exemption for Low-Level Materials.
1......................... Appendix A................ Determination of A1 and A2.
2......................... 71.12..................... General License: NRC-Approved Package.
2......................... 71.13..................... Previously Approved Package.
2......................... 71.14..................... General License: DOT Specification Container.
2......................... 71.16..................... General License: Use of Foreign Approved Package.
2......................... 71.81..................... Applicability of Operating Controls and Procedures.
2......................... 71.85..................... Preliminary Determinations.
2......................... 71.87..................... Routine Determinations.
2......................... 71.88..................... Air Transport of Plutonium.
2......................... 71.89..................... Opening Instructions.
2......................... 71.97..................... Advance Notification of Shipment of Irradiated Reactor
Fuel and Nuclear Waste.
3......................... 71.0...................... Purpose and Scope.
3......................... 71.1...................... Communications.
3......................... 71.2...................... Interpretations.
3......................... 71.3...................... Requirement for License.
3......................... 71.7...................... Completeness and Accuracy of Information.
3......................... 71.8...................... Specific Exemptions.
3......................... 71.9...................... Exemption of Physicians.
3......................... 71.91..................... Records.
3......................... 71.93..................... Inspections and Tests.
3......................... 71.95..................... Reports.
3......................... 71.99..................... Violations.
3......................... 71.101.................... Quality Assurance Requirements.
3......................... 71.103.................... Quality Assurance Organization.
3......................... 71.105.................... Quality Assurance Program.
3......................... 71.107.................... Package Design Control.
3......................... 71.109.................... Procurement Document Control.
3......................... 71.111.................... Instructions, Procedures, and Drawings.
3......................... 71.113.................... Document Control.
3......................... 71.115.................... Control of Purchased Material, Equipment, and Services.
3......................... 71.117.................... Identification and Control of Materials, Parts, and
Components.
3......................... 71.119.................... Control of Special Process.
3......................... 71.121.................... Internal Inspection.
3......................... 71.123.................... Test Control.
3......................... 71.125.................... Control of Measuring and Test Equipment.
3......................... 71.127.................... Handling, Storage, and Shipping Control.
3......................... 71.129.................... Inspection, Test and Operating Status.
3......................... 71.131.................... Nonconforming Materials, Parts, or Components.
3......................... 71.133.................... Corrective Action.
3......................... 71.135.................... Quality Assurance Records.
3......................... 71.137.................... Audits.
4......................... 71.6...................... Information Collection Requirements: OMB Approval.
4......................... 71.18..................... General License: Fissile Material, Limited Quantity per
Package.
4......................... 71.20..................... General license: Fissile Material, Limited Moderator per
Package.
4......................... 71.22..................... General License: Fissile Material, Limited Quantity,
Controlled Shipment.
4......................... 71.24..................... General License: Fissile Material, Limited Moderator,
Controlled Shipment.
4......................... 71.31..................... Contents of Application.
4......................... 71.33..................... Package Description.
4......................... 71.35..................... Package Evaluation.
4......................... 71.37..................... Quality Assurance.
4......................... 71.38..................... Renewal of a Certificate of Compliance or Quality
Assurance Program Approval.
4......................... 71.39..................... Requirement for Additional Information.
4......................... 71.41..................... Demonstration of Compliance.
4......................... 71.43..................... General Standards for all Packages.
4......................... 71.45..................... Lifting and Tie-down Standards for all Packages.
4......................... 71.47..................... External Radiation Standards for all Packages.
4......................... 71.51..................... Additional Requirements for Type B Packages.
[[Page 50262]]
4......................... 71.52..................... Exemption for Low-Specific-Activity (LSA) Packages.
4......................... 71.53..................... Fissile Material Exemptions.
4......................... 71.55..................... General Requirements for Fissile Material Packages.
4......................... 71.59..................... Standards for Arrays of fissile Material Packages.
4......................... 71.61..................... Special Requirement for Irradiated Nuclear Fuel
Shipments.
4......................... 71.63..................... Special Requirements for Plutonium Shipments.
4......................... 71.64..................... Special Requirements for Plutonium Air Shipments.
4......................... 71.65..................... Additional Requirements.
4......................... 71.71..................... Normal Conditions of Transport.
4......................... 71.73..................... Hypothetical Accident Conditions.
4......................... 71.74..................... Accident Conditions for Air Transport of Plutonium.
4......................... 71.75..................... Qualification of Special Form Radioactive Material.
4......................... 71.77..................... Qualification of LSA-III Material.
4......................... 71.83..................... Assumptions as to Unknown Properties.
4......................... 71.100.................... Criminal Penalties.
----------------------------------------------------------------------------------------------------------------
Petitions for Rulemaking
Three petitions for rulemaking were filed with the NRC in
connection with the rules for transporting LSA radioactive material.
The substance of each of the three petitions was essentially the same,
to request that NRC exempt LSA materials from its requirements in Part
71.
The petitioners were the Energy Research and Development
Administration (now the U.S. Department of Energy) in its letter dated
July 23, 1975 (PRM-71-1); ANSI Committee N14, in its letter dated March
10, 1976 (PRM-71-2); and Chem-Nuclear Systems, Inc., in its letter
dated November 22, 1976 (PRM-71-4). At the time these petitions were
filed, DOT regulated carriers and shippers of small quantities of all
radioactive materials (including LSA materials) through provisions in
its regulations in 49 CFR Parts 170-189, whereas NRC regulated shippers
of fissile material and of larger quantities of other radioactive
materials (including LSA materials) through its regulations in Part 71
and its licensing program. All three petitioners argued that the
control NRC was exerting over transportation of LSA materials created
an inconsistency between NRC regulations and those of the IAEA and
should be discontinued. A proposed rule that would have provided the
exemption for LSA materials requested in the petitions was published by
NRC for public comment on August 17, 1979 (44 FR 48234). Before
finalization of that rule, however, a deficiency in the new LSA
requirements, as proposed, was recognized so that the entire LSA
proposal, including the exemption, was withdrawn. In the interim, the
corresponding deficiency in the LSA requirements in the IAEA
regulations was recognized and corrected. That correction is discussed
under the ``major modifications from proposed rule'' section of this
preamble. This correction is implemented in both DOT regulations and
NRC regulations.
The exemption requested in the three petitions has been superseded
by the changes in LSA requirements. The LSA requirements imposed in NRC
regulations are an integral part of the NRC/DOT regulatory scheme for
LSA materials. This scheme is based on IAEA regulations. There is an
exemption provided for LSA materials in Sec. 71.10 that clearly defines
the level where NRC regulations impose additional packaging
requirements. For the above reasons, NRC has denied the petitions.
Administrative Correction
At about the same time the Notice of Proposed Rulemaking regarding
compatibility with IAEA transportation regulations was published for
public comment on June 8, 1988 (53 FR 21550), a separate notice of
final rulemaking was issued, by NRC, affecting the retention periods
for records (53 FR 19240, May 27, 1988). Included in that separate
notice were changes to the transportation regulations in Part 71,
specifically to Secs. 71.105, ``Quality assurance program,'' and
71.135, ``Quality assurance records.'' Because the two rules were being
processed at the same time by different organizations, NRC's internal
controls failed to recognize that the new quality assurance provisions
needed to be incorporated in the June 8, 1988, notice of proposed
rulemaking. No written comments were filed with respect to the quality
assurance sections proposed, although two phone calls were received
advising NRC of its error. The quality assurance changes that were made
effective by the final rule, published on May 27, 1988, are included in
this final rule.
Finding of No Significant Environmental Impact: Availability
The Commission has determined, under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
Subpart A of 10 CFR Part 51, that this rule is not a major Federal
action significantly affecting the quality of the human environment,
and therefore an environmental impact statement (EIS) is not required.
The Commission's ``Final Environmental Statement on the
Transportation of Radioactive Material by Air and Other Modes,'' NUREG-
0170,3 dated December 1977, is NRC's generic EIS, covering all
types of radioactive material transportation by all modes (road, rail,
air, and water). From the Commission's latest survey of radioactive
material shipments and their characteristics, ``Transport of
Radioactive Material in the United States,'' SAND 84-7174, April 1985,
it can be concluded that current radioactive material shipments are not
so different from those evaluated in NUREG-0170 as to invalidate the
results or conclusions of that EIS. Environmental impacts associated
with this rulemaking are evaluated in ``Regulatory Analysis of Changes
to 10 CFR Part 71--NRC Regulations on Packaging and Transportation of
Radioactive Material,'' dated April 1995.
\3\ Copies of NUREG-0170 may be purchased from the
Superintendent of Documents, U.S. Government Printing Office, P.O.
Box 37082, Washington, DC 20013-7082. Copies are also available from
the National Technical Information Service, 5285 Port Royal Road,
Springfield, VA 22161. A copy is also available for inspection and
copying for a fee in the NRC Public Document Room, 2120 L Street,
NW. (Lower Level), Washington, DC.
---------------------------------------------------------------------------
NUREG-0170 established the non-accident related radiation exposures
associated with transportation of radioactive material in the United
States as 98 person-Sv (9800 person-rem) which, based on the
conservative linear
[[Page 50263]]
radiation dose hypothesis, resulted in a maximum of 1.7 genetic effects
and 1.2 latent cancer effects per year. More than half this impact
resulted from shipment of medical-use radioactive materials. Accident
related impacts were established at a maximum of one genetic effect and
one latent cancer fatality for 200 years of transporting radioactive
materials. The principal nonradiological impacts were found to be two
injuries per year, and less than one accidental death per 4 years. In
contrast, non-accident related radiation exposures associated with this
rulemaking would be increased by 0.75 person-Sv/y (75.0 person-rem/y),
whereas accident related impacts would be decreased by approximately
0.006 person-Sv/y (0.6 person-rem/y). Nonradiological traffic injuries
would be increased by 0.06 per year and nonradiological traffic deaths
by 0.003 per year (less than 1 accidental death per 330 years). These
impacts are judged to be insignificant compared with the baseline
impacts established in NUREG-0170.
The environmental assessment and finding of no significant impact
on which this determination is based are available, for inspection, at
the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC. Single copies of the environmental assessment and
finding of no significant impact are also available from the contact
listed under the Addresses heading.
Paperwork Reduction Act Statement
This final rule amends information collection requirements that are
subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et
seq.). These requirements were approved by the Office of Management and
Budget, Approval Number 3150-0008.
The public reporting burden for this collection of information is
estimated to average 7 hours per response, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
collection of information. Send comments regarding this burden estimate
or any other aspect of this collection of information, including
suggestions for reducing this burden, to the Information and Records
Management Branch (T-6F33), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-10202, (3150-0008), Office of
Management and Budget, Washington, D.C. 20503.
Regulatory Analysis
The NRC has prepared a regulatory analysis on this final
regulation. The analysis examines the costs and benefits of the
alternatives considered by NRC. Interested persons may examine a copy
of the regulatory analysis at the NRC Public Document Room at 2120 L
Street NW. (Lower Level), Washington, DC. Single copies of the analysis
may be obtained from the contact listed under the Addresses heading.
Regulatory Flexibility Act Certification
In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C.
605(b)), the Commission certifies that this rule does not have a
significant economic impact on a substantial number of small entities.
This final rule affects NRC licensees, including operators of nuclear
power plants, who transport or deliver to a carrier, for transport,
relatively large quantities of radioactive material, in a single
package. These companies do not generally fall within the scope of the
definition of ``small entities'' set forth in the Regulatory
Flexibility Act or the size standards adopted by the NRC (10 CFR
2.810).
Backfit Analysis
The Commission has determined that the backfit rule does not apply
to the Part 71 final rule because the final rule is not a backfit under
10 CFR Part 50.109. However, NRC analyzed the accident-resistant
packaging requirement for the specified LSA shipments and found that
there is an increase in overall protection to be derived from the
requirement and that direct and indirect costs of implementation are
justified in view of this increased protection.
The factors normally considered in a backfit analysis are evaluated
in the ``Regulatory Analysis of Changes to 10 CFR Part 71--NRC
Regulations on Packaging and Transportation of Radioactive Material,''
dated April 1995. That evaluation shows very small changes in accident
risks as a result of the adoption of the revision, but some reduction
in maximum consequences given an accident. The evaluation shows broad
improvement in NRC regulatory consistency with IAEA, at an initial cost
of $1.375 million to industry, and continual annual costs to industry
of $1.0 million (See Table S.1 of Regulatory Analysis). NRC costs are
estimated at $0.463 million.
The continuing costs are associated with the addition of new limits
on the quantity of LSA radioactive material allowed in a single
transportation package. Internationally, a new limit is considered to
be a necessary safety requirement to limit the consequences of a severe
transportation accident involving LSA material.
The one-time costs are chiefly associated with industry upgrading
of its package safety analyses to include the proposed new accident
crush and immersion tests and with NRC review of those new analyses.
The estimated costs are overstated because of the assumption that all
licensees using packages approved under earlier regulatory standards
would take immediate steps to upgrade the package analyses so the
package approvals would reflect approval, under the latest revised
standards. Although that is a prudent assumption, absent any reasonable
basis for predicting actual licensee reaction, there is little reason
licensees would take any immediate action to upgrade their package
approvals. Both domestic and international regulations are based on the
responsible agency's confidence that packages built to a design
approved under earlier standards are adequately safe for continued use,
although new package construction to that design would be limited, and
international use requires approval by all countries through which the
package is to be transported. In actual practice, some package
approvals would never be upgraded. Those that would be upgraded would
be done over a period of several years as guidance and experience in
upgrading become available.
Although the regulatory analysis shows a small reduction in
accident risks from the amendments to this rule and some reduction in
maximum consequences given an accident, the primary benefit of this
rulemaking is to achieve consistency in radioactive material
transportation regulations between the United States and the rest of
the world. This consistency would not only facilitate the free movement
of radioactive materials between countries for medical, research,
industrial, and nuclear fuel cycle purposes, but it would also
contribute to safety by concentrating the efforts of the world's
experts on a single set of safety standards and guidance (those of the
IAEA) from which individual countries could develop their domestic
regulations. In addition, the accident experience of every country that
bases its domestic regulations on those of the IAEA could be applied to
every other country with consistent regulations to improve its safety
program.
In summary, the effort to make U.S. regulations compatible with
those of the IAEA provides major benefits including
[[Page 50264]]
a substantial increase in the overall protection of the public health
and safety, and it is associated with short-term and relatively minor
costs that are justified in view of this increased protection. This
effort is associated with ongoing costs, but the new limit is
considered to be a justified safety requirement, to limit the
consequences of a severe transportation accident involving LSA
material.
List of Subjects in 10 CFR Part 71
Criminal penalties, Hazardous materials transportation, Nuclear
materials, Packaging and containers, Reporting and recordkeeping
requirements.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended, the Energy Reorganization
Act of 1974, as amended, and 5 U.S.C. 552 and 553, 10 CFR part 71 is
revised to read as follows:
PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL
Subpart A--General Provisions
Sec.
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.
Subpart B--Exemptions
71.6 Information collection requirements: OMB approval.
71.7 Completeness and accuracy of information.
71.8 Specific exemptions.
71.9 Exemption of physicians.
71.10 Exemption for low-level materials.
71.11 [Reserved]
Subpart C--General Licenses
71.12 General license: NRC-approved package.
71.13 Previously approved package.
71.14 General license: DOT specification container.
71.16 General license: Use of foreign approved package.
71.18 General license: Fissile material, limited quantity per
package.
71.20 General license: Fissile material, limited moderator per
package.
71.22 General license: Fissile material, limited quantity,
controlled shipment.
71.24 General license: Fissile material, limited moderator,
controlled shipment.
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.
71.52 Exemption for low-specific-activity (LSA) packages.
71.53 Fissile material exemptions.
71.55 General requirements for fissile material packages.
71.57 [Reserved]
71.59 Standards for arrays of fissile material packages.
71.61 Special requirement for irradiated nuclear fuel shipments.
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
71.71 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.
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.
71.107 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.
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).
Section 71.97 also issued under sec. 301, Pub. L. 96-295, 94
Stat. 789-790.
Subpart A--General Provisions
Sec. 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.
\1\ Postal Service Manual (Domestic Mail Manual), section 124.3,
which is incorporated by reference at 39 CFR 111.1.
---------------------------------------------------------------------------
(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) Exemptions from the requirement for license in Sec. 71.3 are
specified in Sec. 71.10. General licenses for which no NRC package
approval is required are issued in Secs. 71.14 through 71.24. The
general license in Sec. 71.12 requires that an NRC certificate of
compliance or other package approval be issued for the package to be
used under the general license. Application for package
[[Page 50265]]
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. The transport of
licensed material or delivery of licensed material to a carrier for
transport is subject to the operating controls and procedures
requirements of subpart G of this part, to the quality assurance
requirements of subpart H of this part, and to the general provisions
of subpart A of this part, including DOT regulations referenced in
Sec. 71.5.
(e) 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.
Sec. 71.1 Communications and records.
(a) All communications concerning the regulations in this part
should be addressed to the Director, Office of Nuclear Material Safety
and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, or may be delivered in person, at the Commission offices,
at 11545 Rockville Pike, Rockville, Maryland.
(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, 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.
Sec. 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.
Sec. 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.
Sec. 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. A2 means the maximum
activity of radioactive material, other than special form, LSA and SCO
material, permitted in a Type A package. These values are either listed
in Appendix A of this part, Table A-1, or may be derived in accordance
with the procedure prescribed in 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.
Close reflection by water means immediate contact by water of
sufficient thickness for maximum reflection of neutrons.
Containment system means the assembly of components of the
packaging intended to retain the radioactive material during 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 aircraft any aircraft.
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 plutonium-238, plutonium-239, plutonium-241,
uranium-233, uranium-235, or any combination of these radionuclides.
Unirradiated natural uranium and depleted uranium, and natural uranium
or depleted uranium that has been irradiated in thermal reactors only
are not included in this definition. Certain exclusions from fissile
material controls are provided in Sec. 71.53.
Licensed material means by-product, 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 that satisfies the descriptions and
limits set forth below. Shielding materials surrounding the LSA
material may not be considered in determining the estimated average
specific activity of the package contents. LSA material must be in one
of three groups:
(1) LSA-I.
(i) Ores containing only naturally occurring radionuclides (e.g.,
uranium, thorium) and uranium or thorium concentrates of such ores; or
(ii) Solid unirradiated natural uranium or depleted uranium or
natural thorium or their solid or liquid compounds or mixtures; or
(iii) Radioactive material, other than fissile material, for which
the A2 value is unlimited; or
(iv) Mill tailings, contaminated earth, concrete, rubble, other
debris, and activated material in which the radioactive material is
essentially uniformly distributed, and the average specific activity
does not exceed 10-6 A2/g.
(2) LSA-II.
(i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/
liter); or
(ii) Material in which the radioactive material is essentially
uniformly distributed, and the average specific activity does not
exceed 10-4 A2/g for solids and gases, and 10-5 A2/
g for liquids.
(3) LSA-III. Solids (e.g., consolidated wastes, activated
materials) in which:
(i) The radioactive material is essentially uniformly distributed
[[Page 50266]]
throughout a solid or a collection of solid objects, or is essentially
uniformly distributed in a solid compact binding agent (such as
concrete, bitumen, ceramic, etc.);
(ii) The radioactive material is relatively insoluble, or it is
intrinsically contained in a relatively insoluble material, so that,
even under loss of packaging, the loss of radioactive material per
package by leaching, when placed in water for 7 days, would not exceed
0.1 A2; and
(iii) The average specific activity of the solid does not exceed 2
x 10-3
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 Sec. 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.''
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 means a fissile material packaging
together with its fissile material contents.
(2) 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 lb/in2) gauge or a
pressure relief device that would allow the release of radioactive
material to the environment under the tests specified in Sec. 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 Sec. 71.13.
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 Sec. 71.75. A special form
encapsulation designed in accordance with the requirements of Sec. 71.4
in effect on June 30, 1983, (see 10 CFR part 71, revised as of January
1, 1983), and constructed before July 1, 1985, and a special form
encapsulation designed in accordance with the requirements of Sec. 71.4
in effect on March 31, 1996, (see 10 CFR part 71, revised as of January
1, 1983), and constructed before April 1, 1998, may continue to be
used. Any other special form encapsulation must meet the specifications
of this definition.
Specific activity 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.
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 non-fixed contamination on the accessible surface averaged
over 300 cm2 (or the area of the surface if less than 300
cm2) does not exceed 4 Bq/cm2 (10-4 microcurie/cm2)
for beta and gamma and low toxicity alpha emitters, or 0.4 Bq/cm2
(10-5 microcurie/cm2) for all other alpha emitters;
(ii) The fixed contamination on the accessible surface averaged
over 300 cm2 (or the area of the surface if less than 300
cm2) does not exceed 4x104 Bq/cm2 (1.0 microcurie/
cm2) for beta and gamma and low toxicity alpha emitters, or
4x103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha
emitters; and
(iii) The non-fixed contamination plus the fixed contamination on
the inaccessible surface averaged over 300 cm2 (or the area of the
surface if less than 300 cm2) does not exceed 4x104 Bq/
cm2 (1 microcurie/cm2) for beta and gamma and low toxicity
alpha emitters, or 4x103 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 non-fixed contamination on the accessible surface averaged
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does
not exceed 400 Bq/cm\2\ (10-2 microcurie/cm\2\) for beta and gamma
and low toxicity alpha emitters or 40 Bq/cm\2\ (10-3 microcurie/
cm\2\) for all other alpha emitters;
(ii) The fixed contamination on the accessible surface averaged
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does
not exceed 8 x 10\5\ Bq/cm\2\ (20 microcuries/cm\2\) for beta and gamma
and low toxicity alpha emitters, or 8 x 10 \4\ Bq/cm\2\ (2 microcuries/
cm\2\) for all other alpha emitters; and
(iii) The non-fixed contamination plus the fixed contamination on
the inaccessible surface averaged over 300 cm\2\ (or the area of the
surface if less than 300 cm\2\) does not exceed 8 x 10\5\ Bq/cm\2\ (20
microcuries/cm\2\) for beta and gamma and low toxicity alpha emitters,
or 8 x 10\4\ Bq/cm\2\ (2 microcuries/cm\2\) for all other alpha
emitters.
Transport index 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 determined as follows:
(1) For non-fissile material packages, the number determined by
multiplying the maximum radiation level in millisievert (mSv) per hour
at one meter (3.3 ft) from the external surface of the package by 100
(equivalent to the
[[Page 50267]]
maximum radiation level in millirem per hour at one meter (3.3 ft)); or
(2) For fissile material packages, the number determined by
multiplying the maximum radiation level in millisievert per hour at one
meter (3.3 ft) from the external surface of the package by 100
(equivalent to the maximum radiation level in millirem per hour at one
meter (3.3 ft)), or, for criticality control purposes, the number
obtained as described in Sec. 71.59, whichever is larger.
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.
Uranium--natural, depleted, enriched
(1) Natural uranium means uranium 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.
Sec. 71.5 Transportation of licensed material.
(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 170 through 189 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 and B and I.
(ii) Marking and labeling--49 CFR part 172: Subpart D,
Secs. 172.400 through 172.407, Secs. 172.436 through 172.440, and
subpart E.
(iii) Placarding--49 CFR part 172: Subpart F, especially
Secs. 172.500 through 172.519, 172.556, and appendices B and C.
(iv) Accident reporting--49 CFR part 171: Secs. 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) 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.
Subpart B--Exemptions
Sec. 71.6 Information collection requirements: OMB approval.
(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 of 1980 (44 U.S.C. 3501 et seq.). 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 Secs. 71.5, 71.6a, 71.7, 71.12, 71.13, 71.31,
71.33, 71.35, 71.37, 71.38, 71.39, 71.47, 71.85, 71.87, 71.89, 71.91,
71.93, 71.95, 71.97, 71.101, 71.103, 71.105, 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, and 71.137.
Sec. 71.7 Completeness and accuracy of information.
(a) Information provided to the Commission by an applicant for a
license, or by a licensee, or information required by statute or by the
Commission's regulations, orders, or license conditions to be
maintained by the applicant or the licensee must be complete and
accurate in all material respects.
(b) Each applicant or licensee shall notify the Commission of
information identified by the applicant or licensee as having, for the
regulated activity, a significant implication for public health and
safety or common defense and security. An applicant or licensee
violates this requirement only if the applicant or licensee fails to
notify the Commission of information that the applicant or licensee 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 two working
days of identifying the information. This requirement is not applicable
to information that is already required to be provided to the
Commission by other reporting or updating requirements.
Sec. 71.8 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.
Sec. 71.9 Exemption of physicians.
Any physician licensed by a State to dispense drugs in the practice
of medicine is exempt from Sec. 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.
Sec. 71.10 Exemption for low-level materials.
(a) A licensee is exempt from all requirements of this part with
respect to shipment or carriage of a package containing radioactive
material having a specific activity not greater than 70 Bq/g (0.002
Ci/g).
(b) A licensee is exempt from all requirements of this part, other
than Sec. 71.5 and Sec. 71.88, with respect to shipment or carriage of
the following packages, provided the packages contain no fissile
material, or the fissile material exemption standards of Sec. 71.53 are
satisfied:
(1) A package containing no more than a Type A quantity of
radioactive material;
(2) A package in which the only radioactive material is low
specific activity (LSA) material or surface contaminated objects (SCO),
provided the external radiation level at 3 m from the unshielded
material or objects does not exceed 10 mSv/h (1 rem/h); or
[[Page 50268]]
(3) A package transported within locations within the United States
which contains only americium or plutonium in special form with an
aggregate radioactivity not to exceed 20 curies.
(c) A licensee is exempt from all requirements of this part, other
than Secs. 71.5 and 71.88, with respect to shipment or carriage of low-
specific-activity (LSA) material in group LSA-I, or surface
contaminated objects (SCOs) in group SCO-I.
Sec. 71.11 [Reserved]
Subpart C--General Licenses
Sec. 71.12 General license: NRC-approved package.
(a) A general license is hereby 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, 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) This general license applies only to a licensee who--
(1) Has a copy of the certificate of compliance, or other approval
of the package, and has 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) Complies 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) Submits in writing to the Director, Office of Nuclear Material
Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001, before the licensee's first use of the package, 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 Sec. 71.13.
Sec. 71.13 Previously approved package.
(a) A Type B package previously approved by NRC but not designated
as B(U) or B(M) in the identification number of the NRC Certificate of
Compliance, may be used under the general license of Sec. 71.12 with
the following additional conditions:
(1) Fabrication of the packaging was satisfactorily completed by
August 31, 1986, as demonstrated by application of its model number in
accordance with Sec. 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 that uniquely identifies each packaging which
conforms to the approved design is assigned to, and legibly and durably
marked on, the outside of each packaging.
(b) A Type B(U) package, a Type B(M) package, a low specific
activity (LSA) material package or a fissile material package,
previously approved by the NRC but without the designation ``-85'' in
the identification number of the NRC Certificate of Compliance, may be
used under the general license of Sec. 71.12 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 Sec. 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) 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 Secs. 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 Secs. 71.71 and 71.73;
and
(3) The modifications to the package satisfy the requirements of
this part.
(d) NRC will revise the package identification number to designate
previously approved package designs as B(U), B(M), AF, BF, or A as
appropriate, and with the identification number suffix ``-85'' after
receipt of an application demonstrating that the design meets the
requirements of this part.
Sec. 71.14 General license: DOT specification container.
(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 specification container for fissile material or for a
Type B quantity of radioactive material as specified in DOT regulations
at 49 CFR parts 173 and 178.
(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) This general license applies only to a licensee who--
(1) Has a copy of the specification; and
(2) Complies with the terms and conditions of the specification and
the applicable requirements of subparts A, G, and H of this part.
(d) This general license is subject to the limitation that the
specification container may not be used for a shipment to a location
outside the United States, except by multilateral approval, as defined
in DOT regulations at 49 CFR 173.403.
Sec. 71.16 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 DOT as meeting the applicable requirements of 49 CFR
171.12.
(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) This general license applies only to a licensee who--
(1) Has 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) Complies with the terms and conditions of the certificate and
revalidation, and with the applicable requirements of subparts A, G,
and H of
[[Page 50269]]
this part. With respect to the quality assurance provisions of subpart
H of this part, the licensee is exempt from design, construction, and
fabrication considerations.
Sec. 71.18 General license: Fissile material, limited quantity per
package.
(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, without complying with the package standards of
subparts E and F of this part, if the material is shipped in accordance
with this section.
(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) This general license applies only when a package contains no
more than a Type A quantity of radioactive material, including only one
of the following:
(1) Up to 40 g of uranium-235;
(2) Up to 30 g of uranium-233;
(3) Up to 25 g of the fissile radionuclides of plutonium, except
that for encapsulated plutonium-beryllium neutron sources in special
form, an A1 quantity of plutonium may be present; or
(4) A combination of fissile radionuclides in which the sum of the
ratios of the amount of each radionuclide to the corresponding maximum
amounts in paragraphs (c)(1), (2), and (3) of this section does not
exceed unity.
(d) (1) This general license applies only when, except as specified
below for encapsulated plutonium-beryllium sources, a package
containing more than 15 g of fissile radionuclides is labeled with a
transport index not less than the number given by the following
equation, where the package contains x grams of uranium-235, y grams of
uranium-233, and z grams of the fissile radionuclides of plutonium:
Minimum Transport Index = (0.40x+0.67y+z) (1-15 ).x+y+z
(2) For a package in which the only fissile material is in the form
of encapsulated plutonium-beryllium neutron sources in special form,
the transport index based on criticality considerations may be taken as
0.026 times the number of grams of the fissile radionuclides of
plutonium in excess of 15 g. In all cases, the transport index must be
rounded up to one decimal place and may not exceed 10.0.
Sec. 71.20 General license: Fissile material, limited moderator per
package.
(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, without complying with the package standards of
subparts E and F of this part if the material is shipped in accordance
with this section.
(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) This general license applies only when--
(1) The package contains no more than a Type A quantity of
radioactive material;
(2) Neither beryllium nor hydrogenous material enriched in
deuterium is present;
(3) The total mass of graphite present does not exceed 7.7 times
the total mass of uranium-235 plus plutonium;
(4) Substances having a higher hydrogen density than water (e.g.,
certain hydrocarbon oils), are not present, except that polyethylene
may be used for packing or wrapping;
(5) Uranium-233 is not present, and the amount of plutonium does
not exceed 1 percent of the amount of uranium-235;
(6) The amount of uranium-235 is limited as follows:
(i) If the fissile radionuclides are not uniformly distributed, the
maximum amount of uranium-235 per package may not exceed the value
given in Table I of this part; or
(ii) If the fissile radionuclides are distributed uniformly (i.e.,
cannot form a lattice arrangement within the packaging), the maximum
amount of uranium-235 per package may not exceed the value given in
Table II of this part; and
(7) The transport index of each package, based on criticality
considerations, is taken as 10 times the number of grams of uranium-235
in the package divided by the maximum allowable number of grams per
package in accordance with Table I or Table II of this part, as
applicable.
Table I.--Permissible Mass of Uranium-235 per Fissile Material Package,
Applicable to Sec. 71.20(c)(6)(i)
[Nonuniform distribution]
------------------------------------------------------------------------
Permissible
maximum
Uranium enrichment in weight percent of uranium-235 not grams of
exceeding uranium-235
per package
------------------------------------------------------------------------
24......................................................... 40
20......................................................... 42
15......................................................... 45
11......................................................... 48
10......................................................... 51
9.5........................................................ 52
9.......................................................... 54
8.5........................................................ 55
8.......................................................... 57
7.5........................................................ 59
7.......................................................... 60
6.5........................................................ 62
6.......................................................... 65
5.5........................................................ 68
5.......................................................... 72
4.5........................................................ 76
4.......................................................... 80
3.5........................................................ 88
3.......................................................... 100
2.5........................................................ 120
2.......................................................... 164
1.5........................................................ 272
1.35....................................................... 320
1.......................................................... 680
0.92....................................................... 1,200
------------------------------------------------------------------------
Table II.--Permissible Mass of Uranium-235 per Fissile Material Package,
Applicable to Sec. 71.20(c)(6)(ii)
[Uniform Distribution]
------------------------------------------------------------------------
Permissible
maximum
Uranium enrichment in weight percent of uranium-235 not grams of
exceeding uranium-235
per package
------------------------------------------------------------------------
4.......................................................... 84
3.5........................................................ 92
3.......................................................... 112
2.5........................................................ 148
2.......................................................... 240
1.5........................................................ 560
1.35....................................................... 800
------------------------------------------------------------------------
Sec. 71.22 General license: Fissile material, limited quantity,
controlled shipment.
(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, without complying with the package standards of
Subparts E and F of this part, if limited material is shipped in
accordance with this section.
(b) The general license applies only to a licensee who has a
quality assurance
[[Page 50270]]
program approved by the Commission as satisfying the provisions of
Subpart H of this part.
(c) This general license applies only when a package contains no
more than a Type A quantity of radioactive material and no more than
400 g total of the fissile radionuclides of plutonium encapsulated as
plutonium-beryllium neutron sources in special form.
(d) This general license applies only when the fissile
radionuclides in the shipment exceed none of the following:
(1) 500 g of uranium-235;
(2) 300 g total of uranium-233, and the fissile radionuclides of
plutonium;
(3) A total quantity of uranium-233, uranium-235, and the fissile
radionuclides of plutonium so that the sum of the ratios of the
quantity of each radionuclide to the quantity specified in paragraphs
(d)(1) and (d)(2) of this section does not exceed unity; or
(4) 2500 g total of the fissile radionuclides of plutonium
encapsulated as plutonium-beryllium neutron sources in special form.
(e) This general license applies only when shipment of these
packages is made under procedures specifically authorized by DOT, in
accordance with 49 CFR part 173 of its regulations, to prevent loading,
transport, or storage of these packages with other fissile material
shipments.
Sec. 71.24 General license: Fissile material, limited moderator,
controlled shipment.
(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, without complying with the package standards of
subparts E and F of this part, if limited material is shipped in
accordance with this section.
(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) This general license applies only when--
(1) No package contains more than a Type A quantity of radioactive
material;
(2) The packaging does not incorporate lead shielding exceeding 5
cm in thickness, tungsten shielding, or uranium shielding;
(3) Neither beryllium nor hydrogenous material enriched in
deuterium is present;
(4) The total mass of graphite present does not exceed 7.7 times
the total mass of uranium-235 and plutonium;
(5) Substances having a higher hydrogen density than water (e.g.,
certain hydrocarbon oils), are not present, except that polyethylene
may be used for packing or wrapping;
(6) For fissile contents containing no uranium-233 and less than 1
percent by weight total plutonium, if the fissile radionuclides are--
(i) Not uniformly distributed, the maximum amount of uranium-235
per consignment does not exceed the value given in Table III of this
part; or
(ii) Distributed uniformly and cannot form a lattice arrangement
within the packaging, the maximum amount of uranium-235 per shipment
does not exceed the value given in Table IV of this part;
(7) For fissile contents containing uranium-233 or more than 1
percent by weight plutonium, the total mass of fissile material per
shipment is limited so that the sum of the number of grams of uranium-
235 divided by 400, the number of grams of plutonium divided by 225,
and the number of grams of uranium-233 divided by 250, does not exceed
unity, as expressed in the formula:
[GRAPHIC][TIFF OMITTED]TR28SE95.000
(8) The transport must be direct to the consignee without any
intermediate transit storage; and
(9) Shipment of these packages is made under procedures
specifically authorized by DOT in accordance with 49 CFR part 173 of
its regulations to prevent loading, transport, or storage of these
packages with other fissile material shipments.
Table III.--Permissible Mass of Uranium-235 per Fissile Material
Shipment Applicable to Sec. 71.24(c)(6)(i)
[Nonuniform distribution]
------------------------------------------------------------------------
Permissible
Uranium enrichment in weight percent of uranium-235 maximum grams
notexceeding of uranium-235
perconsignment
------------------------------------------------------------------------
20...................................................... 520
15...................................................... 560
11...................................................... 600
10...................................................... 640
9.5..................................................... 655
9....................................................... 675
8.5..................................................... 690
8....................................................... 710
7.5..................................................... 730
7....................................................... 750
6.5..................................................... 780
6....................................................... 810
5.5..................................................... 850
5....................................................... 900
4.5..................................................... 950
4....................................................... 1,000
3.5..................................................... 1,100
3....................................................... 1,250
2.5..................................................... 1,500
2....................................................... 2,050
1.5..................................................... 3,400
1.35.................................................... 4,000
1....................................................... 8,500
0.92.................................................... 15,000
------------------------------------------------------------------------
Table IV.--Permissible Mass of Uranium-235 per Fissile Material Shipment
Applicable to Sec. 71.24(c)(6)(ii)
[Uniform distribution]
------------------------------------------------------------------------
Permissible
maximum
Uranium enrichment in weight percent of uranium-235 not grams of
exceeding uranium-235
per
consignment
------------------------------------------------------------------------
4......................................................... 1,050
3.5....................................................... 1,150
3......................................................... 1,400
2.5....................................................... 1,800
2......................................................... 3,000
1.5....................................................... 7,000
1.35...................................................... 10,000
------------------------------------------------------------------------
Subpart D--Application for Package Approval
Sec. 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 Sec. 71.33;
[[Page 50271]]
(2) A package evaluation as required by Sec. 71.35; and
(3) A quality assurance program description, as required by
Sec. 71.37, or a reference to a previously approved quality assurance
program.
(b) Except as provided in Sec. 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.
Sec. 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;
(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.
Sec. 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
Sec. 71.59; and
(c) For a fissile material shipment, any proposed special controls
and precautions for transport, loading, unloading, and handling and any
proposed special controls in case of an accident or delay.
Sec. 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.
Sec. 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.
Sec. 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
Sec. 71.41 Demonstration of compliance.
(a) The effects on a package of the tests specified in Sec. 71.71
(``Normal conditions of transport''), and the tests specified in
Sec. 71.73 (``Hypothetical accident conditions''), and Sec. 71.61
(Special requirement for irradiated nuclear fuel shipments''), must be
evaluated by 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 Secs. 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.
Sec. 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
[[Page 50272]]
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 Sec. 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 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 deg.C (100 deg.F) and in the
shade, no accessible surface of a package would have a temperature
exceeding 50 deg.C (122 deg.F) in a nonexclusive use shipment, or
85 deg.C (185 deg.F) in an exclusive use shipment.
(h) A package may not incorporate a feature intended to allow
continuous venting during transport.
Sec. 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.
Sec. 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 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.
Sec. 71.51 Additional requirements for Type B packages.
(a) Except as provided in Sec. 71.52, a Type B package, in addition
to satisfying the requirements of Secs. 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
(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.
Sec. 71.52 Exemption for low-specific-activity (LSA) packages.
A package need not satisfy the requirements of Sec. 71.51 if it
contains only LSA or SCO material, and is transported as exclusive use,
but is subject to Secs. 71.41 through 71.47, including Sec. 71.43(f).
This section expires April 1, 1999.
Sec. 71.53 Fissile material exemptions.
The following packages are exempt from fissile material
classification and
[[Page 50273]]
from the fissile material standards of Sec. 71.55 and Sec. 71.59, but
are subject to all other requirements of this part:
(a) A package containing not more than 15 g of fissile material. If
material is transported in bulk, the quantity limitation applies to the
conveyance;
(b) A package containing homogeneous hydrogenous solutions or
mixtures where:
(1) The minimum ratio of the number of hydrogen atoms to the number
of atoms of fissile radionuclides (H/X) is 5200;
(2) The maximum concentration of fissile radionuclides is 5 g/
liter; and
(3) The maximum mass of fissile radionuclides in the package is 800
g, with an exception for a mixture where the total mass of plutonium
and uranium-233 exceeds 1 percent of the mass of uranium-235, the limit
is 500 g. If the material is transported in bulk, other than by
aircraft, the quantity limitations apply to the conveyance;
(c) A package containing uranium enriched in uranium-235 to a
maximum of 1 percent by weight, and with a total plutonium and uranium-
233 content of up to 1 percent of the mass of uranium-235, if the
fissile radionuclides are distributed homogeneously throughout the
package contents and do not form a lattice arrangement within the
package;
(d) A package containing any fissile material if it does not
contain more than 5 g of fissile radionuclides in any 10 liter volume,
and if the material is packaged so as to maintain this limit of fissile
radionuclide concentration during normal transport;
(e) A package containing not more than 1 kg of plutonium of which
not more than 20 percent by mass may consist of plutonium-239,
plutonium-241, or any combination of those radionuclides; or
(f) A package containing liquid solutions of uranyl nitrate
enriched in uranium-235 to a maximum of 2 percent by weight, with total
plutonium and uranium-233 not more than 0.1 percent of the mass of
uranium-235 and with a minimum nitrogen-to-uranium atomic ratio (N/U)
of 2.
Sec. 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 Secs. 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 Sec. 71.51.
(b) Except as provided in paragraph (c) 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 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 Sec. 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 Sec. 71.59(b)(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 Sec. 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.
Sec. 71.57 [Reserved]
Sec. 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 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 Sec. 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 transport index based on nuclear criticality control must
be obtained 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 transport index for nuclear criticality control may be
zero provided that an unlimited number of packages is subcritical such
that the value of ``N'' is effectively equal to infinity under the
procedures specified in paragraph (a) of this section. Any transport
index greater than zero must be rounded up to the first decimal place.
(c) Where a fissile material package is assigned a nuclear
criticality control transport index--
(1) Not in excess of 10, that package may be shipped by any
carrier, and that carrier provides adequate criticality control by
limiting the sum of the transport indexes to 50 in a non-exclusive use
vehicle, and to 100 in an exclusive use vehicle.
(2) In excess of 10, that package may only be shipped by exclusive
use
[[Page 50274]]
vehicle or other shipper controlled system specified by DOT for fissile
material packages. The shipper provides adequate criticality control by
limiting the sum of the transport indexes to 100 in an exclusive use
vehicle.
Sec. 71.61 Special requirement for irradiated nuclear fuel shipments.
A package for irradiated nuclear fuel with activity greater than 37
PBq (106 Ci) must be so designed that its undamaged containment
system can withstand an external water pressure of 2 MPa (290 psi) for
a period of not less than one hour without collapse, buckling, or
inleakage of water.
Sec. 71.63 Special requirements for plutonium shipments.
(a) Plutonium in excess of 20 Ci (0.74 TBq) per package must be
shipped as a solid.
(b) Plutonium in excess of 20 Ci (0.74 TBq) per package must be
packaged in a separate inner container placed within outer packaging
that meets the requirements of subparts E and F of this part for
packaging of material in normal form. If the entire package is
subjected to the tests specified in Sec. 71.71 (``Normal conditions of
transport''), the separate inner container must not release plutonium
as demonstrated to a sensitivity of 10-6 A2/h. If the entire
package is subjected to the tests specified in Sec. 71.73
(``Hypothetical accident conditions''), the separate inner container
must restrict the loss of plutonium to not more than A2 in 1 week.
Solid plutonium in the following forms is exempt from the requirements
of this paragraph:
(1) Reactor fuel elements;
(2) Metal or metal alloy; and
(3) Other plutonium bearing solids that the Commission determines
should be exempt from the requirements of this section.
Sec. 71.64 Special requirements for plutonium air shipments.
(a) A package for the shipment of plutonium by air subject to
Sec. 71.88(a)(4), in addition to satisfying the requirements of
Secs. 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 Sec. 71.74, rather than the
hypothetical accident tests in Sec. 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 deg.C (+33 deg.F)
to 38 deg.C (+100 deg.F) for those qualification tests involving water,
and ambient atmospheric temperatures ranging from -40 deg.C (-40 deg.F)
to +54 deg.C (+130 deg.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.
Sec. 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.
Subpart F--Package, Special Form, and LSA-III Tests \2\
\2\ The package standards related to the tests in this subpart
are contained in subpart E of this part.
---------------------------------------------------------------------------
Sec. 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 deg.C (-20 deg.F) and +38 deg.C (+100 deg.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 temperature considered to precede and
follow the tests is more unfavorable.
(c) Conditions and tests.
(1) Heat. An ambient temperature of 38 deg.C (100 deg.F) in still
air, and insolation according to the following table:
Insolation Data
------------------------------------------------------------------------
Total
insolation for
Form and location of surface 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 deg.C (-40 deg.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 hour.
(7) Free drop. Between 1.5 and 2.5 hours 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)
[[Page 50275]]
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 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 hours, 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.
Sec. 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 deg.C (-20 deg.F) and +38 deg.C (+100 deg.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 flat, essentially unyielding, horizontal
surface so as to suffer maximum damage by the drop of a 500 kg (1100
pound) 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 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 lbs), an overall density not
greater than 1000 kg/m3 (62.4 lbs/ft3) based on external
dimensions, and radioactive contents greater than 1000 A2 not as
special form radioactive material.
(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 deg.C (1475 deg.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 deg.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
Sec. 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.
Sec. 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.
(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.
[[Page 50276]]
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 deg. 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 deg. 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.
(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/in \2\).
Sec. 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
x 10-4 torr-liter/s (1.3 x x 10-4 atm-cm\3\/s) based on air
at 25 deg.C (77 deg.F) and one atmosphere differential pressure for
solid radioactive content, or x 10-6 torr-liter/s
(1.3 x x 10-6 atm-cm\3\/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.
(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 deg.C (1475 deg.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
[[Page 50277]]
temperature. The water must have a pH of 6-8 and a maximum conductivity
of 10 micromho per centimeter at 20 deg. (68 deg.F);
(ii) The water with specimen must then be heated to a temperature
of 50 deg.C 5 deg.C (122 deg.F 9 deg.F) and
maintained at this temperature for 4 hours.
(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 deg.C
(86 deg.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 deg.C 5 deg.C
(122 deg.F 9 deg.F) and maintained at that temperature for
4 hours;
(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
(Ci)).
(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 deg.C 5 deg.C (122 deg.F 9 deg.F) and
maintained at this temperature for 4 hours;
(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 deg.C (86 deg.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 Sec. 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.''
Sec. 71.77 Qualification of LSA-III Material
(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 deg.C (68 deg.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
Sec. 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.
Sec. 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.
Sec. 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/in\2\) 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, 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.
Sec. 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
Sec. 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
Sec. 71.47 at any time during transportation; and
(k) Accessible package surface temperatures will not exceed the
limits specified in Sec. 71.43(g) at any time during transportation.
Sec. 71.88 Air transport of plutonium.
(a) Notwithstanding the provisions of any general licenses and
[[Page 50278]]
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 not greater than 0.002 Ci/g (70 Bq/g) of material
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 Sec. 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 Sec. 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.
Sec. 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).
Sec. 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 Sec. 71.10, 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;
(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 Sec. 71.87 and by
the conditions of the package approval.
(b) The licensee 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.
(c) The licensee 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
Sec. 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. The records must be retained for three years
after the life of the packaging to which they apply.
Sec. 71.93 Inspection and tests.
(a) The licensee or certificate holder shall permit the Commission,
at all 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 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 licensee shall notify the Administrator of the appropriate
NRC Regional Office listed in appendix A of part 73 of this chapter, at
least 45 days before fabrication of a package to be used for the
shipment of licensed material having a decay heat load in excess of 5
kW or with a maximum normal operating pressure in excess of 103 kPa (15
lbf/in2) gauge.
Sec. 71.95 Reports.
The licensee shall report to the Director, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, within 30 days--
(a) Any instance in which there is significant reduction in the
effectiveness of any approved Type B, or fissile, packaging during use;
(b) Details of any defects with safety significance in Type B, or
fissile, packaging after first use, with the means employed to repair
the defects and prevent their recurrence; or
(c) Instances in which the conditions of approval in the
certificate of compliance were not observed in making a shipment.
Sec. 71.97 Advance notification of shipment of irradiated reactor fuel
and nuclear waste.
(a) 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, through, 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.
(b) Advance notification is required under this section for
shipments of irradiated reactor fuel in quantities less than that
subject to advance notification requirements of Sec. 73.37(f) of this
chapter. Advance notification is also required under this section for
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 the office of each
appropriate governor or governor's designee and to the Administrator of
the appropriate
[[Page 50279]]
NRC Regional Office listed in appendix A to part 73 of this chapter.
(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 messenger must reach the office of
the governor or of the governor's 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) The list will be published annually in the Federal Register on
or about June 30 to reflect any changes in information.
(iii) A list of the names and mailing addresses of the governors'
designees is available on request from the Director, Office of State
Programs, 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 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
(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, 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 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, and to the Administrator of
the appropriate NRC Regional Office listed in appendix A of part 73 of
this chapter.
(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.
Sec. 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
(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.
Sec. 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:
Secs. 71.0, 71.2, 71.4, 71.6, 71.7, 71.9, 71.10, 71.31, 71.33, 71.35,
71.37, 71.38, 71.39, 71.41, 71.43, 71.45, 71.47, 71.51, 71.52, 71.53,
71.55, 71.59, 71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, and
71.100.
Subpart H--Quality Assurance
Sec. 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.
(b) Establishment of program. Each licensee shall establish,
maintain, and execute a quality assurance program satisfying each of
the applicable criteria of Secs. 71.101 through 71.137 and satisfying
any specific provisions that are applicable to the licensee's
activities including procurement of packaging. The licensee shall apply
each of the applicable criteria in a graded approach, i.e., to an
extent that is consistent with its importance to safety.
(c) Approval of program. 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. 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, with the Director, Office of
Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001.
(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.
[[Page 50280]]
(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 prior
to 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 Appendix B
of Part 50 of this chapter, and that is established, maintained, and
executed with regard to transport packages, will be accepted as
satisfying the requirements of paragraph (b) of this section. Before
first use, the licensee shall notify the Director, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, of its intent to apply its previously
approved Appendix B program to transportation activities. The licensee
shall identify the program by date of submittal to the Commission,
Docket Number, and date of Commission approval.
Sec. 71.103 Quality assurance organization.
(a) The licensee 3 shall be responsible for the establishment
and execution of the quality assurance program. The licensee 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. The licensee shall clearly establish
and delineate, in writing, the authority and duties of persons and
organizations performing activities affecting the safety-related
functions of structures, systems, and components. These activities
include performing the functions associated with attaining quality
objectives and the quality assurance functions.
\3\ 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 prior to the time a package approval is issued.
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(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 safety-related functions have
been performed correctly.
(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.
Sec. 71.105 Quality assurance program.
(a) The licensee 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
Secs. 71.101 through 71.137. The licensee 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 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, 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 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 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 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 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 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
which they are executing.
Sec. 71.107 Package design control.
(a) The licensee shall establish measures to assure that applicable
regulatory requirements and the package design, as specified in the
license 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 safety-related functions of the
materials, parts, and components of the packaging.
[[Page 50281]]
(b) The licensee 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 shall include
suitable qualification testing of a prototype or sample unit under the
most adverse design conditions. The licensee shall apply design control
measures to items such as 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 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 package
approval require NRC approval.
Sec. 71.109 Procurement document control.
The licensee 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 or by its
contractors or subcontractors. To the extent necessary, the licensee
shall require contractors or subcontractors to provide a quality
assurance program consistent with the applicable provisions of this
part.
Sec. 71.111 Instructions, procedures, and drawings.
The licensee 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.
Sec. 71.113 Document control.
The licensee shall establish measures to control the issuance of
documents such as instructions, procedures, and drawings, including
changes, which 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. These measures must assure that changes to documents are
reviewed and approved.
Sec. 71.115 Control of purchased material, equipment, and services.
(a) The licensee 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 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 shall
retain, or have available, this documentary evidence for the life of
the package to which it applies. The licensee shall assure that the
evidence is sufficient to identify the specific requirements met by the
purchased material and equipment.
(c) The licensee 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.
Sec. 71.117 Identification and control of materials, parts, and
components.
The licensee 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.
Sec. 71.119 Control of special processes.
The licensee 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.
Sec. 71.121 Internal inspection.
The licensee 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.
Sec. 71.123 Test control.
The licensee 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 shall document and evaluate the test results to assure that
test requirements have been satisfied.
[[Page 50282]]
Sec. 71.125 Control of measuring and test equipment.
The licensee shall establish measures to assure that tools, gauges,
instruments, and other measuring and testing devices used in activities
affecting quality are properly controlled, calibrated, and adjusted at
specified times to maintain accuracy within necessary limits.
Sec. 71.127 Handling, storage, and shipping control.
The licensee 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.
Sec. 71.129 Inspection, test, and operating status.
(a) The licensee 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.
Sec. 71.131 Nonconforming materials, parts, or components.
The licensee 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.
Sec. 71.133 Corrective action.
The licensee 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.
Sec. 71.135 Quality assurance records.
The licensee shall maintain sufficient written records to describe
the activities affecting quality. The records must include the
instructions, procedures, and drawings required by Sec. 71.111 to
prescribe quality assurance activities and must include closely related
specifications such as required qualifications of personnel,
procedures, and equipment. The records must include the instructions or
procedures which establish a records retention program that is
consistent with applicable regulations and designates factors such as
duration, location, and assigned responsibility. The licensee shall
retain these records for 3 years beyond the date when the licensee last
engages in the activity for which the quality assurance program was
developed. If any portion of the written procedures or instructions is
superseded, the licensee shall retain the superseded material for 3
years after it is superseded.
Sec. 71.137 Audits.
The licensee 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. Follow-up 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) figure. The
curie values are expressed to three significant figures to assure
that the difference in the TBq and Ci quantities is one tenth of one
percent or less. Where values of A1 or 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. For individual radionuclides whose identities are known, but
which are not listed in Table A-1, the determination of the values
of A1 and A2 requires Commission approval, except that the
values of A1 and A2 in Table A-2 may be used without
obtaining Commission approval.
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 nuclide has a half-life either
longer than 10 days, or longer than that of the parent nuclide,
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 nuclide of that chain. In
the case of radioactive decay chains in which any daughter nuclide
has a half-life either longer than 10 days, or greater than that of
the parent nuclide, the parent and those daughter nuclides shall be
considered as mixtures of different nuclides.
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:
[GRAPHIC][TIFF OMITTED]TR28SE95.001
(b) For normal form radioactive material, the maximum quantity
transported in a Type A package:
[[Page 50283]]
[GRAPHIC][TIFF OMITTED]TR28SE95.002
Where B(i) is the activity of radionuclide I and A1(i) and
A2(i) are the A1 and A2 values for radionuclide I,
respectively.
Alternatively, an A1 value for mixtures of special form
material may be determined as follows:
[GRAPHIC][TIFF OMITTED]TR28SE95.003
Where f(i) is the fraction of activity of nuclide I in the mixture
and A1(i) is the appropriate A1 value for nuclide I.
An A2 value for mixtures of normal form material may be
determined as follows:
[GRAPHIC][TIFF OMITTED]TR28SE95.004
Where f(i) is the fraction of activity of nuclide I in the mixture
and A2(i) is the appropriate A2 value for nuclide I.
V. 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.
Table A-1.--A1 and A2 Values for Radionuclides
--------------------------------------------------------------------------------------------------------------------------------------------------------
Specific activity
Symbol of Element and A1 (TBq) A1 (Ci) A2 (TBq) A2 (Ci) -------------------------------------
radionuclide atomic number (TBq/g) (Ci/g)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225........... Actinium(89)..... 0.6 16.2 1 x 10-2 0.270 2.1 x 103 5.8 x 104
Ac-227........... 40 1080 2 x 10-5 5.41 x 10-4 2.7 7.2 x 101
Ac-228........... 0.6 16.2 0.4 10.8 8.4 x 104 2.2 x 106
Ag-105........... Silver(47)....... 2 54.1 2 54.1 1.1 x 103 3.0 x 104
Ag-108m.......... 0.6 16.2 0.6 16.2 9.7 x 10-1 2.6 x 101
Ag-110m.......... 0.4 10.8 0.4 10.8 1.8 x 103 4.7 x 103
Ag-111........... 0.6 16.2 0.5 13.5 5.8 x 103 1.6 x 105
Al-26............ Aluminum(13)..... 0.4 10.8 0.4 10.8 7.0 x 10-4 1.9 x 10-2
Am-241........... Americium(95).... 2 54.1 2 x 10-4 5.41 x 10-3 1.3 x 10-1 3.4
Am-242m.......... 2 54.1 2 x 10-4 5.41 x 10-3 3.6 x 10-1 9.7 x 105
Am-243........... 2 54.1 2 x 10-4 5.41 x 10-3 7.4 x 10-3 2.0 x 10-1
Ar-37............ Argon(18)........ 40 1080 40 1080 3.7 x 103 9.9 x 104
Ar-39............ 20 541 20 541 1.3 x 103 3.4 x 101
Ar-41............ 0.6 16.2 0.6 16.2 1.5 x 106 4.2 x 107
Ar-42............ 0.2 5.41 0.2 5.41 9.6 2.6 x 102
As-72............ Arsenic(33)...... 0.2 5.41 0.2 5.41 6.2 x 104 1.7 x 106
As-73............ 40 1080 40 1080 8.2 x 102 2.2 x 104
As-74............ 1 27.0 0.5 13.5 3.7 x 103 9.9 x 104
As-76............ 0.2 5.41 0.2 5.41 5.8 x 104 1.6 x 106
As-77............ 20 541 0.5 13.5 3.9 x 104 1.0 x 106
At-211........... Astatine(85)..... 30 811 2 54.1 7.6 x 104 2.1 x 10