# Compatibility With the International Atomic Energy Agency (IAEA)

> Briefs, arguments, decisions, and more.

URL: https://www.frixlaw.com/law-library/documents/fr%3A95-23538

## Record

- **Collection:** Federal Register
- **Document type:** Rule
- **Published:** September 28, 1995
- **Citation:** 60 FR 50248

## Text

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.
---------------------------------------------------------------------------

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 Inte

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A95-23538. Public record. Not legal advice.