[Federal Register Volume 61, Number 14 (Monday, January 22, 1996)]
[Notices]
[Pages 1608-1625]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-703]
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NUCLEAR REGULATORY COMMISSION
Disposition of Cesium-137 Contaminated Emission Control Dust and
Other Incident-Related Material; Proposed Staff Technical Position
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice: Proposed Staff Technical Position.
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[[Page 1609]]
SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing guidance,
in the form of a Technical Position, that may be used in case-by-case
requests by appropriate licensees to dispose of a specific mixed waste.
Mixed waste is a waste that is not only radioactive, but also
classified as hazardous under the Resource Conservation and Recovery
Act (RCRA). The specific mixed waste is emission control dust from
electric arc furnaces and foundries that has been contaminated with
cesium-137 (Cs-137). The contamination results from the inadvertent
melting of a Cs-137 source, that: (1) has been improperly disposed of
by an NRC or Agreement State licensee; (2) has been commingled with the
steel scrap supply; (3) has not been detected as it progresses to the
steel producing process; and (4) is volatilized in production process
and thereby can and has contaminated large volumes of emission control
dust and the emission control systems at steel producing facilities.
The proposed position, which has been coordinated with the U.S.
Environmental Protection Agency (EPA), provides the possibility of a
public health-protective, environmentally sound, and cost-effective
alternative for the disposal of much of this mixed waste that contains
Cs-137, in concentrations similar to values that frequently occur in
the environment. The position provides the bases that, with the
approval of appropriate regulatory authorities (e.g., State-permitting
agencies) and others (e.g., disposal site operators), and with public
input, could be used to allow disposal of treated (stabilized) waste at
Subtitle C, RCRA-permitted, hazardous waste disposal facilities. NRC
believes that disposal, under the provisions of the position or other
acceptable alternatives, is preferable to allowing this mixed waste to
remain indefinitely at steel company sites.
The proposed position has been developed through a very ``open''
process in which working draft documents have been routinely shared
with EPA, and also placed in NRC's Public Document Room (Subject File:
204.1.23) to allow interested party access. In keeping with this
process, NRC, rather than noticing the availability of the proposed
position, is publishing the entire position for public comment.
DATES: Submit comments by March 22, 1996. Comments received after this
date will be considered if it is practical to do so, but the Commission
is able to assure consideration only for comments received on or before
this date.
ADDRESSES: Send comments to Chief, Rules Review and Directives Branch,
U.S. Nuclear Regulatory Commission, Washington, DC 20555. A final
position will be issued following NRC staff review of the comments
received.
FOR FURTHER INFORMATION CONTACT:
W.R. Lahs, Division of Waste Management, Office of Nuclear Material
Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington,
DC 20555, Telephone (301) 415-6756.
SUPPLEMENTARY INFORMATION:
Disposition of Cesium-137 Contaminated Emission Control Dust and Other
Incident-Related Materials; Proposed Branch Technical Position
A. Introduction
Emission control (baghouse) dust and other incident-related
materials (e.g., cleanup materials or recycle process streams)
contaminated with cesium-137 (Cs-137) 1 are currently being stored
as mixed radioactive and hazardous waste at several steel company sites
across the country. At any single site, this material typically
contains a total Cs-137 quantity ranging downward from a little more
than one curie (37 gigabecquerels (GBq)) of activity, distributed
within several hundred to a few thousand tons of iron/zinc-rich dust,
as well as within much smaller quantities of cleanup or dust-recycle,
process stream materials.2
\1\ The byproduct material Cs-137 does not include the Cs-137,
from global fallout, that exists in the environment from the testing
of nuclear explosive devices (See Footnote 3).
\2\ The term, ``incident-related material,'' is frequently used
in this position to refer to the total spectrum of Cs-137-
contaminated materials resulting from an inadvertent melting event.
Because of its widespread use in radioactive devices and its
volatility when subjected to steel melting temperatures, the
position is directed solely at incident-related materials involving
this nuclide.
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The radioactivity is not evenly distributed among these materials.
Typically, a small fraction (e.g., one-tenth) of the material contains
most (e.g., 95 percent) of the radioactivity. Most of the material
contains a small quantity of radioactivity at low concentrations and
makes up most of the mixed-waste volume. This material is generally
classified as hazardous waste under RCRA because it contains lead,
cadmium, and chromium that are common to the recycle metal supply. The
Cs-137 contamination of this hazardous waste, on the other hand,
results from a series of three principal events: (1) the loss of
control of a radioactive source by an NRC or Agreement State licensee;
(2) the inclusion of the source within the recycle metal scrap supply
used by the steel producers; and (3) the inability to screen out the
radioactive source as it progresses along the typical scrap collection-
to-melt pathway (e.g., including radiation detectors used at most
furnaces and foundries). Consequently, irrespective of the quantity or
concentration of the radioactivity, all the material is subject to
joint regulation as mixed waste under RCRA and the Atomic Energy Act of
1954, as amended, or the equivalent law of an Agreement State.
The disposal options for these materials, specifically the large
volumes of material with the lower concentrations of Cs-137, have been
limited because of their ``mixed-waste'' classification and the costs
associated with the disposition of large volumes of mixed or
radioactive waste. Long-term solutions addressing the control and
accountability of licensed radioactive sources are being considered by
NRC and its Agreement States. Solutions addressing the disposition of
mixed wastes are being considered by various Federal and State
regulatory authorities and the U.S. Department of Energy. Nevertheless,
the Commission believes that, pending decisions on improved licensee
accountability and the ultimate disposition of mixed waste, appropriate
disposal of the existing incident-related, mixed-waste material is
preferable to indefinite onsite storage.
As a result, this technical position defines the bases that the NRC
staff would generally find acceptable for: (1) authorizing a licensee,
possessing Cs-137 contaminated emission control dust and other
incident-related materials (e.g., the steel company or its service
contractor), to transfer Cs-137 contaminated material, below levels
specified in this position, to a Subtitle C, RCRA-permitted hazardous
waste disposal facility; and (2) exempting the possession and disposal
of these incident-related materials (e.g., by the RCRA-permitted
disposal facility) from NRC or Agreement State licensing requirements.
Because of its radioactivity (i.e., Cs-137 concentration levels), some
of the incident-related material may not be suitable for disposal at a
Subtitle C, RCRA-permitted disposal facility. This material may be
disposed of either: (1) at a licensed low-level radioactive waste
disposal facility following ``delisting'' (e.g., after appropriate
treatment of its hazardous constituents) or (2) at a mixed waste
disposal facility, if applicable acceptance criteria are met.
The regulatory basis for the first action is found at 10 CFR
20.2001(a)(1). This paragraph authorizes a licensee to
[[Page 1610]]
dispose of licensed material as provided in the regulations in 10 CFR
Parts 30, 40, 60, 61, 70, or 72. Paragraph 30.41(b) states the
conditions under which licensees are allowed to transfer byproduct
material. Paragraph 30.41(b)(7) of Part 30 specifically provides that
licensees may transfer byproduct material if authorized, by the
Commission, in writing.
The regulatory basis for the second action is found at Sec. 30.11
(``Specific exemptions''), which states that the Commission may, on its
own initiative, grant exemptions (from the requirements of the
regulations in 10 CFR Parts 30 through 36, and 39) as it determines are
authorized by law and will not endanger life or property and are
otherwise in the public interest. It should be noted that additional
acceptance requirements, beyond those covered in this NRC position for
disposal of Cs-137-contaminated hazardous waste at a Subtitle C RCRA-
permitted disposal facility, may be established by: (1) an Agreement
State; (2) the permit conditions or policies of the RCRA-permitted
disposal facility; (3) the regulatory requirements of the RCRA disposal
facility's permitting agency; or (4) other authorized parties,
including State and local governments. These requirements may be more
stringent than those covered in the guidance described in this
technical position. The licensed entity transferring the Cs-137-
contaminated incident-related materials should consult with these
parties, and obtain all necessary approvals, before making the
transfers defined in this technical position. Nothing in this position
shall be or is intended to be construed as a waiver of any RCRA permit
condition or term, of any State or local statute or regulation, or of
any Federal RCRA regulation.
B. Discussion
Over the past decade, there has been an increasing number of
instances in which radioactive material has been inadvertently
commingled with scrap metal that subsequently has entered the steel-
recycle production process. If this radioactive material is not removed
before the melting process, it could contaminate the finished metal
product, associated dust-recycle process streams, equipment
(principally air effluent treatment systems), and the dust generated
during the process. Some of the contaminant radioactivity is a result
of naturally occurring radionuclides that deposit in oil and gas
transmission piping. Other radioactivity may be associated with
radioactive sources that are contained in industrial or medical
devices. In this latter case, the commingling of the radioactive source
with metal destined for recycling can occur if the regulatorily
required accountability of these sources fails and a radioactive source
is included within the metal scrap supply used by the steel producers.
In cases where the radionuclide is naturally occurring, or is already
present in the environment as a result of global fallout, the
inadvertent melting of a radioactive source could increase the
contaminant concentration above that caused by these background
environmental levels.3
\3\ In a letter to William Guerry, Jr. from NRC's Executive
Director for Operations, James M. Taylor, dated May 25, 1993, NRC
made a preliminary determination that Cs-137 levels in baghouse dust
can reasonably be attributed to fallout from past nuclear weapons
testing, if concentrations are less than about 2 pCi/g (0.074 Bq/g).
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Although many of the steel producers have installed equipment to
detect incoming radioactivity, this equipment cannot provide absolute
protection because of the shielding of radioactive emissions that may
be provided by uncontaminated scrap metal or the shielded ``pig'' that
contains the radioactive source. Of special concern, because of the
nature and magnitude of the involved radioactivity, are NRC- or
Agreement State-licensed sources containing Cs-137.
When Cs-137 sources are inadvertently melted with a load of scrap
metal, a significant amount of the Cs-137 activity contaminates the
metal-rich dust that is collected in the highly efficient emission
control systems that steel mills have installed to comply with air
pollution regulations. Because of toxic constituents--specifically
lead, cadmium, and chromium--electric arc furnace (EAF) and foundry
emission control dust are subject to regulation under RCRA. If this
dust becomes contaminated with Cs-137, the resulting material would be
classified as a mixed waste. Emission control dust, generated
immediately after the melting of a Cs-137 source with the scrap metal,
can contain cesium concentrations in the range of hundreds or thousands
of picocuries per gram (pCi/g) or a few to a few tens of becquerels
(Bq) per gram of dust, above typical levels in dust caused by Cs-137 in
the environment (e.g., 2 pCi/g or 0.074 Bq/g). Several thousand cubic
feet (several tens of cubic meters) of dust could be contaminated at
these levels. Dust generated days or weeks after a melt of a source
(containing hundreds of millicuries or a few curies of Cs-137) will
contain reduced concentrations, typically less than 100 pCi/g (3.7 Bq/
g).
Even after extensive decontamination and remediation activities,
newly generated dust may still contain concentrations greater than 2
pCi/g (0.074 Bq/g) background levels, but generally less than 10 pCi/g
(0.37 Bq/g). When the melting of a source is not immediately detected,
materials related to downstream processes have also been contaminated
with relatively low concentrations of Cs-137 (e.g., 10 pCi/g (0.37 Bq/
g)). In addition, materials used during decontamination may also be
contaminated with dust containing Cs-137 concentrations at similar
levels above background.
As the result of past inadvertent meltings of Cs-137 sources, a
number of steel producers possess a total of over 10,000 tons of
incident-related materials, most of which contains Cs-137
concentrations of less than 100 pCi/g (3.7 Bq/g). This material is
typically being stored onsite because of the lack of disposal options
that are considered cost effective by the steel companies.4 It is
the disposition of material at these concentration levels that is the
subject of this technical position.
\4\ In April 1995, Envirocare of Utah, Inc., an operator of a
mixed-waste disposal site, received authorization from the State of
Utah and initiated operations to treat and dispose of Cs-137-
contaminated incident-related (mixed waste) materials at
concentrations not exceeding 560 pCi/g (20.7 Bq/g).
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C. Regulatory Position
General
Because of the ``incident-related'' origin of the Cs-137
contaminated materials, the Commission has approved a course of action
that includes: (1) exploration of approaches to improve licensee
control and accountability to reduce the likelihood of sealed sources
entering the scrap metal supply; (2) cooperation with the steel
manufacturers and other appropriate organizations to identify the
magnitude and character of the problem (with particular emphasis on
improving the capability to detect sealed sources before their
inadvertent melting); and (3) development of interim guidelines for the
disposal of Cs-137 contaminated dust and other incident-related
materials (the subject of this technical position).
Specific
Bases for Allowing Transfer and Possession of Cs-137 Contaminated
Incident-Related Material. The bases for allowing transfer and
possession of Cs-137 contaminated emission control dust and other
incident-related materials, under the provisions of existing
regulations, are as follows: (1) Any
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person at a Subtitle C, RCRA-permitted disposal facility involved with
the receipt, movement, storage, or disposal of contaminated materials
should not receive an exposure greater than 1 millirem (mrem) or 10
micro-sievert (Sv) per year (i.e., one-hundredth of the dose
limit for individual members of the public as defined at 10 CFR
20.1301(a)(1)), above natural background levels; 5 (2) members of
the general public in the vicinity of storage or disposal facilities
should not receive exposures and no individual member of the public
should be likely to receive a dose greater than 1 mrem (10 Sv)
per year above background as a result of any and all transfers and
disposals of contaminated materials; (3) handling or processing of the
contaminated materials, undertaken as a result of its radioactivity,
should not compromise the effectiveness of permitted hazardous waste
disposal operations; (4) treatment of contaminated materials should be
accomplished by persons operating under a licensee's radiation
protection program; and (5) transportation of contaminated materials
should be performed by hazardous material employees, as defined in U.S.
Department of Transportation (DOT) regulations (49 CFR Part 172,
Subpart H).
\5\ The use of 1 mrem (10 Sv) has no significance or
precedential value as a health and safety goal. It was selected only
for the purpose of analysis of the levels at which the referenced
materials could be partitioned to allow the bulk of the material to
be transferred to unlicensed persons. It does not represent an NRC
position on the generic acceptability of dose levels. Such levels
are established only by rule.
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Definition of Contaminated Materials and Initial Incident Response.
A melting event generally necessitates extensive decontamination and
remediation operations at the EAF or foundry (e.g., replacing
refractory bricks and duct work). Subsequent operations include the
proper interim handling and management (e.g., accumulation and
containment) of emission control dust and other incident-related
contaminated materials. Based on a review of several recent incidents,
the dust may contain Cs-137 concentrations up to hundreds or thousands
of pCi/g (a few to a few tens of Bq/g), whereas the other generally
limited-volume, incident-related materials typically contain lower
concentrations. As a result, the initial cleanup and collection/
treatment/ packaging of the contaminated emission control dust and
other materials at the EAF or foundry should be performed by an NRC or
Agreement State licensee operating under an approved radiation
protection program. The licensee would also be responsible for
compliance with other non-radiological regulatory requirements (e.g.,
those of the Occupational Safety and Health Administration and RCRA
Treatment Permitting requirements).
Provisions for Disposal at a Subtitle C, RCRA-Permitted, Disposal
Facility. Once the decontamination/remediation and collection/
treatment/packaging activities have been completed, one of two paths
may be followed for the disposal of the incident-related materials,
dependent on Cs-137 concentration levels and whether the final land
disposal operation involves the burial of packaged or unpackaged
materials.
1. Packaged Disposal of Treated Waste. On this disposal path,
contaminated materials would be treated through stabilization to comply
with all EPA and/or State waste treatment requirements for land
disposal of regulated hazardous waste. The treatment operations would
be undertaken by either (i) The owner/operator of the EAF or foundry
(licensed by NRC or appropriate Agreement State to possess, treat, and
transfer Cs-137 contaminated incident-related materials); or (ii) an
NRC-or Agreement State-licensed service contractor. Based on the
radiological impact assessment provided in the appendix, the licensee
could be authorized to transfer the treated incident-related materials
to a Subtitle C, RCRA-permitted, disposal facility, provided that all
the following conditions are met:
(a) The Cs-137-contaminated emission control dust and other
incident-related materials are the result of an inadvertent melting of
a sealed source or device;
(b) The emission control dust and other incident-related materials
have been treated (stabilized) to meet requirements for land disposal
of RCRA-regulated waste, and have been stored (if applicable) and
transferred in compliance with a radiation protection program as
specified at 10 CFR 20.1101;
(c) The total Cs-137 activity, contained in emission control dust
and other incident-related materials to be transferred to a Subtitle C,
RCRA-permitted, disposal facility, has been specifically approved by
NRC or the appropriate Agreement State(s) and does not exceed the total
activity associated with the inadvertent melting incident. Moreover,
NRC or the appropriate Agreement State should maintain a public record
of the total incident-related Cs-137 activity, received by the facility
over its operating life, to ensure that this total-disposed Cs-137
activity does not exceed 1 curie (37 GBq); 6
\6\ The 1 curie (37 GBq) value represents a reasonable bounding
activity, associated with several incidents, that could be
transferred to an RCRA-permitted facility under the provisions of
this position. It also represents a quantity that would be less than
the activity disposed of over the operating life of the RCRA-
permitted facility, if the facility routinely disposed of non-
incident-related emission control dust containing background
concentrations of Cs-137.
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(d) The RCRA disposal facility operator has been notified in
writing of the impending transfer of the incident-related materials and
has agreed in writing to receive and dispose of the packaged materials;
(e) The licensee providing the radiation protection program
required in paragraph (b), notifies, in writing, the Commission or
Agreement State(s) in which the transferor and transferee are located,
of the impending transfer, at least 30 days before the transfer;
(f) The treated (stabilized) material has been packaged for
transportation and disposal in non-bulk steel packagings as defined in
DOT regulations at 49 CFR 173.213. (Note that this is a condition
established under this technical position and is not a DOT requirement.
Under DOT regulations, material with concentrations of less than 2
thousand picocuries per gram (74 Bq/g) is not considered radioactive);
(g) In any package, the emission control dust and other incident-
related materials, that have been treated (stabilized) and packaged as
defined in (b) and (f) above, contain pretreatment average
concentrations of Cs-137 that did not exceed 130 pCi/g (4.8 Bq/g) of
material; 7 and
\7\ The 130 pCi/g (4.8 Bq/g) value is the concentration, based
on the analysis in the appendix and including a regulatory margin of
1.5, that would result in a calculated potential exposure less than
1 mrem (10 Sv). The disposal of incident-related materials
in packaged form allows compliance with this position to be
demonstrated through measurement of Cs-137 concentrations, as well
as direct radiation levels external to the package. Notwithstanding
the redundant approaches to ensure compliance with the exposure
criterion, the regulatory margin of 1.5 has been included in
determining the acceptable measurables defined in the position.
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(h) The dose rate at 3.28 feet (1 meter) from the surface of any
package containing treated (stabilized) waste does not exceed 20
rem per hour or 0.20 Sv per hour, above
background.8
\8\ At this exposure rate, for the exposure period as defined in
the appendix, total exposure would not exceed 1 mrem (10
Sv) with a regulatory margin of 1.5.
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Note that, in defining the pretreatment Cs-137 concentration value
stated in paragraph (1)(g), a factor of 1.5 has been included as a
regulatory margin. This factor adds further
[[Page 1612]]
assurance to the certainty in protection provided by the licensee's (1)
Sampling of Cs-137 concentrations in contaminated materials, (2)
measurements of dose rate external to the disposal (and transportation)
packagings, and (3) other assumptions included in the radiological
impacts assessment.
2. Disposal of Unpackaged (i.e., Bulk) Treated Waste. On this
disposal path, contaminated materials would also be treated through
stabilization to comply with all EPA and State waste treatment
requirements for land disposal of RCRA-regulated hazardous waste. The
treatment operations would be undertaken by either (i) The owner/
operator of the EAF or foundry (licensed to possess, treat, and
transfer Cs-137-contaminated incident-related materials), or (ii) a
licensed service contractor. Based on the radiological impact
assessment provided in the appendix, the licensee could be authorized
to transfer the treated (stabilized) incident-related materials to a
Subtitle C, RCRA-permitted, disposal facility, provided that all the
following conditions are met. (Note that conditions (a) through (e) are
identical to those applicable to packaged disposal of treated waste):
(a) The Cs-137 contaminated emission control dust and other
incident-related materials are the result of an inadvertent melting of
a sealed source or device;
(b) The emission control dust and other incident-related materials
have been treated (stabilized) to meet requirements for land disposal
of RCRA-regulated waste, and have been stored (if applicable), and
transferred in compliance with a radiation protection program as
specified at 10 CFR 20.1101;
(c) The total Cs-137 activity, contained in emission control dust
and other incident-related materials to be transferred to a Subtitle C,
RCRA-permitted, disposal facility, has been specifically approved by
NRC or the appropriate Agreement State(s) and does not exceed the total
activity associated with the inadvertent melting incident. Moreover,
NRC or the appropriate Agreement State should maintain a public record
of the total incident-related Cs-137 activity, received by the facility
over its operating life, to ensure that this total disposed Cs-137
activity does not exceed 1 curie (37 GBq); 9
\9\ See footnote 6.
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(d) The RCRA disposal facility operator has been notified in
writing of the impending transfer of the incident-related materials and
has agreed in writing to receive and dispose of these materials;
(e) The licensee providing the radiation protection program
required in paragraph (b) notifies, in writing, the Commission or
Agreement State(s) in which the transferor and transferee are located,
of the impending transfer, at least 30 days before the transfer; and
(f) The emission control dust and other incident-related materials,
that have been treated (stabilized) as defined in (b) above, contain
pretreatment average concentrations of Cs-137 that did not exceed 100
pCi/g (3.7 Bq/g) of material.10
\10\ The 100 pCi/g (3.7 Bq/g) value is the concentration, based
on the analysis in the appendix and including a regulatory margin of
2, that would result in a calculated potential exposure of less than
1 mrem (10 Sv). The disposal of incident-related material
in unpackaged (bulk) form dictates that compliance with this
position would be demonstrated through measurement of Cs-137
concentrations. Without the redundant approach to ensure compliance
with the exposure criterion inherent with the packaged-disposal
approach (see Footnote 7), the regulatory margin, included in
determining the acceptable measurables defined in the position, has
been increased to 2.0.
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Note that, in defining the pretreatment Cs-137 concentration value
in paragraph (2)(f), a factor of 2 has been included as a regulatory
margin. The factor adds further assurance to the certainty of
protection provided by the licensee's (1) sampling of Cs-137
concentrations in contaminated materials; and (2) other assumptions
included in the radiological impacts assessment.
Treatment, Storage, and Transfer of Emission Control Dust or Other
Incident-Related Materials with Cs-137 Concentrations Indistinguishable
from Background Levels (i.e., 2 pCi/g (0.074 Bq/g) or Less). The EAF or
foundry licensed to possess and transfer Cs-137 contaminated emission
control dust or a licensed service contractor is authorized to transfer
emission control dust and other incident-related materials as if they
were not radioactive, provided that the Cs-137 concentration within the
emission control dust and other incident-related materials is 2 pCi/g
(0.074 Bq/g) of material or less.
Aggregation of Cs-137 Contaminated Emission Control Dust and Other
Incident-Related Materials. Aggregation of Cs-137 contaminated emission
control dust and other incident-related material, before stabilization
treatment, is acceptable if performed in compliance with a radiation
protection program, as described at 10 CFR 20.1101, and provided that:
(1) Aggregation involves the same characteristic or listed
hazardous waste and the wastes must be amenable to and undergo the same
appropriate treatment for land-disposal restricted waste;
(2) Aggregation does not increase the overall total volume nor the
radioactivity of the incident-related mixed waste; and
(3) Materials, when aggregated, are subjected to a sampling
protocol that demonstrates compliance with Cs-137 concentration
criteria on a package-average 11 basis.
\11\ The term package, as used here, refers to packages used by
the licensee to transfer the material to the disposal facility,
irrespective of whether this package is also the disposal container.
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Determination of Cs-137 Concentrations and Radiation Measurements.
Cs-137 concentrations may be determined by the licensee by direct or
indirect (e.g., external radiation) measurements, through an NRC- or
Agreement State-approved sampling program. The program should be
sufficient to ensure that Cs-137 contamination in stabilized treated
emission control dust and in other incident-related materials, on a
package-average basis, is consistent with the concentration criteria in
this technical position. The sampling program should provide assurance
that the quantity of Cs-137 in any package (see footnote 11) does not
exceed the product of the applicable concentration criterion times the
net weight of contaminated material in a package.
Appendix--Assessment of Radiological Impact of Disposal of Cs-137
Contaminated Emission Control Dust and Other Incident-related Materials
at a Subtitle C RCRA-Permitted Disposal Facility
Background
In the normal process of producing recycled steel, scrap steel is
subjected to a melting process. In this process, most impurities in the
scrap steel are removed and generally contained within process-
generated slag or off-gas. Typically, the off-gas carries dust,
containing iron and zinc, together with certain heavy metals, through
an emission control system to a ``baghouse,'' where the dust is
captured in ``bag-type'' filters. Hazardous constituents within the
dust, principally lead, cadmium, and chromium, cause the dust to be
designated by EPA as a hazardous waste, under RCRA, often as the listed
waste K061.
Typically, when the scrap consists largely of junk automobiles, the
dust contains a high percentage (greater than 20 percent) of zinc,
which can be a valuable recovery product. Moreover, the zinc recovery
process produces slag and other byproducts that have recycle potential.
If economic (e.g., low zinc content) or process considerations
[[Page 1613]]
preclude these recycle options, the dust may be treated and disposed of
in a hazardous waste disposal facility. Treatment standards for the
various hazardous constituents of the dust have been specified by EPA
in 40 CFR 268.40. Solidification is the treatment process typically
used to meet these standards.
Because the recycling of steel involves the addition of natural
materials (primarily lime and ferromanganese), very low levels of
radioactivity, ubiquitous in the environment, are involved in the
production process. One of these radionuclides is Cs-137 which now
occurs in the environment as a result of global fallout from past
weapons-testing programs.
Cs-137 has a 30-year half-life (i.e., a quantity of this
radionuclide and its associated radioactivity will decrease by half
every 30 years). The decay of Cs-137 and its very short-lived daughter
produces emissions of beta particles and gamma rays.
The principal hazard from the beta particles can only be realized
when it enters the human body. The principal hazard from the gamma rays
is as an external source of penetrating radiation similar to the type
of exposure received from an X-ray. Because of its volatility in the
very high-temperature (typically 3000 degrees fahrenheit) steel-making
process, Cs-137 is volatilized and transported in the furnace off-gas
and, as it condenses, becomes a constituent of the emission control
(baghouse) dust. Normal background Cs-137 concentrations in dust have
been measured at picocurie per gram levels (0.024 to 1.23 pCi/g)
12 or thousandths of a becquerel per gram (Bq/g). This
concentration is consistent with the general range of background levels
measured in soils within the United States whereas concentrations of 10
pCi/g (0.37 Bq/g) are relatively common in drainage areas.13 As a
result of this information, NRC has determined that Cs-137
concentrations in emission control dust below 2 pCi/g (0.074 Bq/g) can
be attributed to fallout from past weapons testing.14
\12\ A picocurie is one-trillionth of a curie and represents a
decay rate of one disintegration every 27 seconds or 1/27 of a
becquerel.
\13\ Letter to William Lahs, Nuclear Regulatory Commission,
from Andrew Wallo III, Department of Energy, dated May 20, 1993.
\14\ Letter from James M. Taylor, Nuclear Regulatory
Commission, to William Guerry, Jr., Collier, Shannon, Rill, and
Scott, dated May 25, 1993.
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Statement of Problem
The inadvertent melting of a licensed Cs-137 sealed source with
scrap steel at an EAF or foundry typically results in the contamination
of the steel producer's emission control system and the generation of
potentially large quantities (e.g., of the order of 1000 tons) of Cs-
137 contaminated emission control dust. Facility cleanup operations
will produce an additional quantity of contaminated material and,
depending on the effectiveness of cleanup operations, further
generation of contaminated dust or cleanup-related materials can occur.
Furthermore, if the occurrence of the melting event is not immediately
detected, contamination can unknowingly be carried forward with the
dust into zinc-recovery process streams. In one case, for example, this
has led to Cs-137 contamination of the zinc-rich, splash condenser
dross residue, referred to as SCDR material. In the incidents to date,
total quantities of these contaminated materials have not exceeded 2000
tons per event. The Cs-137 concentration in all these materials can
vary, but in typical past events, much of the material is contaminated
at levels ranging from 2 pCi/g (0.074 Bq/g) to a few hundred pCi/g
(most below approximately 100 pCi/g or 3.7 Bq/g). Smaller volumes
(typically less than 5 percent of the total volume) have included
concentrations at nanocurie/gram levels (thousands of pCi/g or a few
tens of Bq/g).
The intent of this analysis is to characterize the potential
radiological impacts associated with the alternative options for
disposal of Cs-137 contaminated emission control dust and other
incident-related materials at a Subtitle C, RCRA-permitted facility.
Because these RCRA hazardous wastes must be treated to comply with the
requirements for land disposal of restricted waste, the potential
radiological impacts associated with treatment processes required
consideration. To protect against these radiological impacts, the
position includes the provision that treatment of Cs-137 contaminated
emission control dust and other incident-related materials be performed
by an NRC or Agreement State licensee. The licensee would operate under
an approved radiation protection program, as well as any required RCRA
treatment permit. Such controls are necessary because of the wide range
of contaminated materials and their physical forms, together with the
variability in EPA-approved treatment processes. Under this decision,
the Subtitle C, RCRA-permitted disposal facility would be receiving the
emission control dust and other incident-related materials after their
treatment to stabilize the RCRA-hazardous constituents (specifically,
lead, cadmium, and chromium) in a non-dispersible,15 solid (e.g.,
cement-type) form. As a result, the potential radiological hazard from
the ``treated'' material during disposal operations is associated with
its characteristic as an external source of radiation.
\15\ In the context used, the term ``non-dispersible'' means
that any radiological impacts from resuspended material are
inconsequential in comparison to the impacts from direct external
exposures resulting from the emission of gamma radiation in the Cs-
137 decay process.
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After disposal, Cs-137 could only become a hazard through water
pathways if a sufficient quantity and concentration of Cs-137 were to:
(1) become available, (2) be leached from its solid form, (3) be
released from the disposal facility, and (4) enter a drinking water
supply. No significant radiological hazard would be expected to result
from inadvertent intrusion into the disposed waste after facility
closure. Notwithstanding the hazard to the intruder from the hazardous
waste constituents, constraints placed on the total Cs-137 activity and
concentration, and the waste form, can ensure that radiological
exposures would not exceed those that would be received from residing
over commonly-measured background Cs-137 concentrations in the United
States (see discussion under ``Intruder Considerations'').
The following analyses will therefore be directed at an evaluation
of the potential direct, water pathway, and intruder hazards and will
provide a perspective on their significance.
Direct Exposure
After the inadvertent melting of a Cs-137 sealed source at an EAF
or foundry, the relatively volatile Cs-137 will leave the furnace as an
offgas and be commingled with the normal emission control dust. As a
result, concentrations of Cs-137 contained in this dust (and other
materials associated with furnace cleanup operations or subsequent dust
recycle process streams) will increase. Thus, the rate of radiological
exposure from this material will be similar in type, but different in
magnitude, than that received from the typical background levels of Cs-
137. Any change in magnitude of the exposures to workers at the
disposal facility from this contaminated material when compared to the
exposure received from typical emission control dust would depend on:
(1) differences in Cs-137 concentrations; (2) variations in the
physical/chemical properties of the materials disposed of; and (3)
changes in worker time-integrated interactions with contaminated
materials.
[[Page 1614]]
The three key variables above are particularly important in the
development of this technical position. Of significance to all three
variables, the approach defined in the position calls for treatment
(stabilization) of incident-related materials (to comply with
requirements for land disposal of restricted waste) to take place
``under license,'' at the location where the material was generated, or
at the site of a service contractor permitted for stabilization
treatment of the material. Complying with the ``Treatment Standards for
Hazardous Wastes,'' defined at 40 CFR 268.40, will result in a solid
waste form from which exposure rates will be smaller than those
originating from the hazardous waste form (e.g., dust) before
treatment. More importantly, treatment of the contaminated materials,
under license, will obviate the need to specifically address potential
radiological exposures at unlicensed, RCRA-permitted, treatment
facilities. Thus, under the approach of this technical position, any
minimal exposure to workers who have not been trained in radiation
safety would be limited to disposal operations.
Furthermore, because the origin of the Cs-137 contaminated
materials is the result of a melting incident, upper bound values can
be established for the volume, weight, radioactive material
concentration, and total activity of the contaminated material, on an
incident basis. The base case analysis in this appendix presumes that
the contaminated material involves a volume of 40,000 cubic feet (1132
cubic meters), a weight of 2000 tons, and a total activity content of
less than a 1 curie (37 gigabecquerels (GBq)) of Cs-137. These values
are generally consistent with the particulars from the incidents that
have occurred to date.
Within these constraints, the starting point in the direct exposure
calculation is to estimate the radiation dose rate at a distance of
3.28 feet (1 meter) from the surface of a semi-infinite volume (i.e.,
infinite in areal extent and depth from the point of exposure) of
solidified contaminated material.16 The calculations assume that
the initial Cs-137 contamination in all untreated dust is 100 pCi/g
(3.7 Bq/g). Direct exposure results scale linearly for other
concentration levels, if the waste configuration is unchanged.
\16\ This assessment is generally consistent with the approach
employed in ``Risk Assessment of Options for Disposition of EAF Dust
Following a Meltdown Incident of a Radioactive Cesium Source in
Scrap Steel,'' SELA-9301, Stanley E. Logan, April 1993.
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Stabilization treatment,17 conducted under a licensed
radiation protection program, is achieved by mixing moist dust with
additives (e.g., liquid reagent to adjust oxidation potential and
portland cement/fly ash).18 These additives (typically presumed to
add 30 parts by weight to 100 parts of dust or contaminated material)
would result in a solidified product that would contain Cs-137
concentrations at about 77 percent of initial concentrations (e.g., 77
pCi/g (2.84 Bq/g)). Because of allowable variations in the
solidification processes (e.g., from the production of granularized
aggregate to solidified monoliths), the bulk density of the solidified
material can range from about 1.4 to 2.5 g/cm3. A representative
dose conversion factor 19 under these conditions (calculated at a
density of 1.5 g/cm3) would typically be less than 49 microrem/
hour (rem/hr) or 0.49 microsieverts/hour (Sv/hr), at
a distance of 3.28 feet (1 meter) from the surface of a hypothetical
semi-infinite volume of the solidified material.20
\17\ In the context of this position, stabilized treatment does
not include either onsite or offsite high-temperature metals
recycling processes.
\18\ This treatment may include the addition of special
stabilization reagents, such as clays, or involve other RCRA-
approved stabilization technologies, that reduce the leachability of
Cs-137, although the radiological impacts analysis indicates that
such processes are not necessary to protect public health and
safety, and the environment.
\19\ A dose conversion factor represents a value that allows a
radionuclide contamination level to be converted to an estimated
exposure rate.
\20\ The dose rates in this appendix have been calculated
through use of the Microshield computer program, Grove Engineering,
Inc., version 4.2, 1995. The value of 49 rem/hour
represents 0.77 of the 62.9 value shown on Figure 1.
---------------------------------------------------------------------------
Because the quantities of treated dust and other incident-related
materials are not semi-infinite in volume, the actual dose rate/
distance relationships from finite volumes of contaminated materials
will be less. The reduction can be calculated for various volumetric
sources through the use of shape factors. Shape factors have been
calculated for several configurations that are likely to occur during
operations from the time the contaminated treated material is received
at the RCRA-permitted disposal facility through its disposal. The shape
factors can be determined from Figures 1 through 6 for various
distances between a specific source configuration and an exposed
individual. Typically, at a distance of 3.28 feet (1 meter), these
factors range from about 0.03 to 0.5 (Figures 1 through 5), and have
been calculated without accounting for the limited shielding provided
by any packaging. As the distance from the contaminated materials
increases to 9.84 feet (3 meters), the shape factors for these similar
geometries become smaller, ranging from about 0.004 to 0.2. The largest
likely dose rate potentially experienced by an individual involved in
the disposal process, measured at 3.28 feet (1 meter), would be from
the sides of large containers or shipments of contaminated materials,
and would be expected to range from about 10 to less than 14
rem/hour (0.14 Sv/hr) above background (typically 8
to 12 rem/hr (0.08 to 0.12 Sv/hr).21 From an
open trench (Figure 4), filled with contaminated materials, the
calculated dose rate would also be somewhat less than 13 rem/
hr (0.13 Sv/hr) measured directly over the trench at a 3.28
feet (1 meter) distance. Again, these values represent 0.77 of the
respective values indicated on the figures because of solidification
additives. Figures 6 and 7, respectively, show the variation in dose
rate with the width of the trench and depth of the waste. Figure 8 is
provided to show the change in dose rate versus the distance offset
from the side of the trailer-type container considered in Figure 3.
\21\ The two-thirds loading of the 30-cubic yard box is related
to the typical maximum payload weight that can be transported by
truck without an overweight permit. If the boxes referred to in
Figures 1 and 2 were full, the dose rate would increase by less than
a factor of 1.5. Similarly, if the assumed additive weight percent
(i.e., 30 percent) is varied over a reasonable range from 20 to 40
percent, the resulting dose rate would change in an inversely
proportional manner.
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A typical disposal rate at a trench within an RCRA-permitted
facility would typically exceed 500 tons per shift.22 Assuming
this disposal rate of 500 tons per shift applies to the disposal of
treated, Cs-137-contaminated, incident-related material (approximately
20 to 25 truckloads in 8 hours), it would require approximately 4 times
this period of time to dispose of 2000 tons. (Note that the rate of
arriving material would likely be dictated by transportation
arrangements, so that the 32 hours required to dispose of the
contaminated material could be spread over several days or weeks.)
Facility workers, therefore, would, on average, only be exposed to
finite volumes of contaminated material for a maximum period of 32
worker-hours. Applying the highest likely dose rate (approximately 13
rem/hr (0.13 Sv/hr) from the side of a trailer
containing the contaminated materials), and presuming exposure at a
3.28-ft (1-meter) distance for the entire 32-hour period, a worker
would receive
[[Page 1615]]
a dose of less than 0.5 mrem (5 Sv) above background.
\22\ Note that if treatment at an RCRA-permitted facility were
required, the limiting operational handling rate for the treated
materials may be limited to 100 to 200 tons per shift.
---------------------------------------------------------------------------
Qualitatively descriptive time and motion data gathered from three
RCRA-permitted disposal facilities indicate that the above-calculated
dose is conservative for two principal reasons: (1) the workers having
the most significant exposure to materials, from receipt to disposal,
are effectively at greater distances than 3.28 feet (1 meter); and (2)
their exposure is over time periods significantly less than the assumed
receipt through disposal time period of 32 hours. As a result, actual
exposures are expected to be significantly less than 0.5 mrem (5
Sv).
This conservative estimate of potential exposure is based on the
aforementioned time-distance assumptions and is expected to bound
reasonable interactions of disposal facility workers with the treated
(stabilized) incident-related materials. For example, incident-related
material could be stored at the disposal site or samples of the treated
material could be subjected to sampling activities. In the first case,
if a 90-day storage period is presumed, the average exposure distance
over the entire period needed to ensure a dose less than the position's
exposure criteria would be on the order of 10 to 20 meters (see Figures
1 through 3 which illustrate the decrease in dose rate as a function of
distance from the source). In the second case, the typical activity in
a 100 gram sample would be no greater than about 10-2 Ci
(370 Bq). The dose rate from such a sample would be less than 0.1
rem/hr (0.001 Sv/hr) at a distance of 1 foot (0.3
meters).
To place the significance of this calculation into perspective, an
estimate can be made of worker exposure from the presumed handling,
treatment, and disposal of normal emission control dust (i.e., dust
that has not been contaminated with Cs-137 from a melted source). This
dust would contain background levels of Cs-137 (approximately 1 pCi/g
(0.037 Bq/g)). Therefore, a worker interacting with this material at an
effective distance of 3.28 feet (1 meter) over about 300 8-hour shifts
(a little more than a working year) would receive a total maximum
exposure about 0.5 mrem (5 Sv). The magnitude of this exposure
is in the same range as the exposure calculated for the disposal of the
contaminated materials from a single melting event. Moreover, the
potential exposure from the ``melting event'' was estimated under the
extremely conservative assumption that all materials were contaminated
at levels of 100 pCi/g (3.7 Bq/g).
The imposition of a 1-curie (37 GBq) criterion on the total
incident-related activity that could be disposed of at any one Subtitle
C, RCRA facility (see following discussion on water-pathway
considerations) should further ensure that worker exposures from Cs-137
contaminated emission control dust and other incident-related materials
will not exceed 1 mrem/year (10Sv/year) integrated over the
lifetime of the facility.
Water-Pathway Considerations
The proposed approach to manage Cs-137 contaminated emission
control dust and other incident-related materials presumes licensee
treatment of these materials to comply with requirements for land
disposal of restricted waste. Thus, the hazardous radiological and
chemical constituents of these materials will be incorporated into a
stable, solid (e.g., cement-type) form, similar to that required for
routine RCRA-permitted disposal of emission control dust. As a result,
the possibility of Cs-137 presenting a hazard through a water pathway
requires consideration of: (1) the quantity of Cs-137 available; (2)
the degree to which the Cs-137 could be leached from its waste matrix;
and (3) the extent that any leached Cs-137 could migrate into a water
supply.
The disposal of Cs-137 in treated emission control dust and other
incident-related materials would be constrained by this policy to a
total activity of 1 curie (37 GBq). In the previous reference-basis
analysis, an effective concentration, in the treated waste, of 77 pCi/g
(2.84 Bq/g) was evaluated--the originally assumed contaminated material
concentration reduced by 30 percent as a result of the added mass
associated with treatment. Both the quantity and position-defined
concentration values place bounds on any potential water pathway
hazard. In the actual wastes that are subject to potential disposal
under the provisions of this position, the concentration of Cs-137
averaged over all the treated waste would typically be significantly
less than the defined concentration criteria.
Furthermore, because the Cs-137 is contained in a solid matrix and
buried within a facility in which the amount of water infiltration is
minimized, any Cs-137 removal from its final disposal location would be
limited while these conditions remain in effect. The chemistry of any
water interacting with the solidified, Cs-137-contaminated waste would
also be expected to limit the leaching process (e.g., avoidance of
acidic environments), because of the controlled nature of the Subtitle
C, RCRA-permitted disposal site and the types and nature (e.g., no
liquids) of the wastes accepted for disposal. Any water that leached
Cs-137 from the waste would normally be collected in a leachate
collection system at volumetric concentrations expected to be far less
than that existing in the treated waste. The chemistry of the fill
materials used at the disposal site could also provide a sorbing medium
if any Cs-137 leached from the solidified waste. Finally, the location
of Subtitle C, RCRA-permitted disposal sites is such that the source of
any water supply would typically be some distance from the disposal
site.
These chemistry and distance factors are also likely to be major
factors in delaying the arrival of Cs-137 at a receptor well because of
retardation effects. This retardation, in terms of its effect on the
time required, under a worst-case scenario, for the Cs-137 to reach a
water supply, is such that significant radioactive decay of the Cs-137
inventory is likely (the radioactive half-life of Cs-137 is 30 years)
before this pathway could potentially pose a hazard.
Although qualitative in nature, and based on considerations that
can vary among Subtitle C, RCRA-permitted disposal sites, the
discussion has focused on the factors that are likely to prevent any
significant water-pathway hazard. The following, more quantitative
assessment, is provided to conservatively bound any water-pathway
hazard that could potentially occur under extremely unlikely
conditions, and provides the technical basis for NRC's position.
The leachability of Cs-137 from any solid waste form that allows
compliance with the land disposal restrictions for the waste's non-
radiological hazardous constituents is likely to be extremely limited
after initial waste placement. After the end of operations and a post-
closure care period of 30 years, a worst-case scenario presumes that
processes take place to degrade the site so that infiltrating water
from the surface passes unimpeded through the contaminated waste. In
predicting the dissolution of Cs-137 under these conditions, a critical
process is the partitioning of the Cs-137 that takes place between the
waste, soil, and infiltrating water. Conservatively assuming that the
partitioning from the solid waste form is similar to that from the
interstitial backfill soil to water, an estimate can be made of the
amount of Cs-137 that can leach into the infiltrating water.
The most important parameter in estimating this transfer, as well
as the subsequent movement of the Cs-137 in groundwater, is the
distribution
[[Page 1616]]
coefficient, Kd. This parameter expresses the ratio at equilibrium
of Cs-137 sorbed onto a given weight of soil particles to the amount
remaining in a given volume of water. The higher the value of the
distribution coefficient, the greater the concentration of Cs-137
remaining in the soil. The Kd value can be affected by factors
such as soil texture, pH, competing cation effects, soil porewater
concentration, and soil organic matter content.23 For the non-
acidic, sand/clay/soil environments presumed to represent the RCRA-
permitted disposal facilities, a Kd value of 270 milliliter (ml)/g
was selected from the Footnote 23 reference as being appropriate for
the subsequent bounding, conservative analysis.
\23\ ``Default Soil Solid/Liquid Partition Coefficients,
Kds, for Four Major Soil Types: A Compendium,'' M. Sheppard and
D. Thibault, Health Physics, Vol. 59, No. 4, October, 1990, pp. 471-
482.
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To model the potential groundwater impacts, the RESRAD 24 code
was used. For the representative case, the bounding 40,000 cubic feet
(ft\3\) or 1132 cubic meters (m\3\) of treated material were presumed
to be disposed of in a volume measuring 100-ft (30.4-m) length x 20-ft
(6.09-m) width x 20-ft (6.09-m) depth. All this material was assumed
to contain a Cs-137 concentration of 77 pCi/g (2.84 Bq/g).
Notwithstanding the actual layouts of Subtitle C, RCRA-permitted
facilities, a well was presumed to be located and centered at the
downgradient edge of this specific volume of waste. To maximize the
hazard as calculated by the RESRAD model, the hydraulic gradient was
considered to be parallel to the length of the disposed volume.
Infiltration representative of a humid site was presumed and a minimal
unsaturated zone thickness of 3.28 ft (1 m) was assumed to separate the
contaminated zone from the saturated zone. The value assigned to
Kd in the unsaturated zone was 270 ml/g. Assessments beyond this
representative case evaluation are subsequently discussed.
\24\ RESRAD, Version 5.0, Argonne National Laboratory,
September 1993.
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The results from this bounding analysis indicate that drinking
water dose rate would be insignificant (e.g., far less than a microrem
(10-2 Sv) per year). This result is not surprising
because the retardation provided, even in the 3.28-ft (1-m) deep
unsaturated zone and the saturated zone, are sufficient to preclude
drinking water doses for almost 700 years. During this period, the
activity of Cs-137 would decay (i.e., be reduced by radioactive decay)
by a factor of about 10 million.
Note that, although it is considered an unrealistic scenario, the
drinking of the leachate directly from the disposal trench after a
period of 30 years would only result in a calculated exposure of about
7 mrem/year (70 Sv/year).25
\25\ This dose estimate is based on comparing leachate
concentrations with the water effluent concentration in 10 CFR Part
20, Appendix B.
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To consider the effects of a range of parameters, including other
Kd values, on the results of this bounding analysis, the following
analyses are presented. Based on the typical existing volumes and Cs-
137 concentrations of incident-related materials, the imposition of a
constraint on Cs-137 concentration effectively bounds the total
activity that could be disposed of at a Subtitle C, RCRA-permitted
facility from a single steel company site to a few tens of
millicuries.26 Material at higher concentrations would require
disposal at either a mixed-waste disposal facility or a licensed low-
level radioactive waste disposal site. Thus, for the potential
disposals at the Subtitle C, RCRA-permitted site to approach the 1
curie (37 GBq) incident-related material constraint in this position,
disposals of materials from several incidents would have to occur. The
total volume of material, in this case, would still represent only a
small fraction of a RCRA-permitted facility's disposal capacity.
Repeating the RESRAD analysis discussed above under these assumptions,
but respectively considering lower Kd values in the contaminated,
unsaturated, and saturated zones, would still result in drinking water
doses of less than 1 mrem (10 Sv) per year unless the Kd
values in all zones approach single digit values. Even in these cases
(e.g., Kd equal to 2.7), separation of the hypothesized well
location from the disposed material by about 100 meters (328 ft) would
reduce dose rates below 1 mrem (10 Sv) per year because of the
decay of Cs-137 brought about by the increased retardation times.
\26\ For example, the total activity contained in 2000 tons of
material, contaminated at a level of 77 pCi/g, would be about 0.14
curies (5.2 GBq). It would be unlikely that all the material from a
particular incident would be at the maximum concentration defined in
the technical position.
---------------------------------------------------------------------------
The concentration constraints in this position, coupled with the
limited number of inadvertent melting situations to which this position
could be applicable, and the case-by-case NRC or Agreement State
approval of the proposed material transfers are believed to provide a
sufficient basis to ensure protection of public health and safety, and
the environment from water-pathway considerations. Nevertheless, to
provide further protection, should a single Subtitle C, RCRA-permitted
disposal facility accept incident-related material from more than one
incident, the position includes a total Cs-137 incident-related
activity constraint of 1 curie (37 GBq). The magnitude of this
constraint is based on the typical bounding activity associated with an
inadvertent melting of Cs-137 sources that have occurred to date at
EAFs or foundries. In large measure, it has been included to provide
assurance that the position is only directed at the ultimate
disposition of radioactive material that exists in the environment as a
result of specific inadvertent melting incidents. However, it also
provides a constraint on the extent of volumetric contamination as a
function of concentration. The practical effect, as previously alluded
to, is to limit the disposal volumes of incident-related contaminated
materials to a small fraction of total disposal site capacity for
hazardous waste. As a result of this volumetric limit, the constraint
would further ensure that any exposures occurring offsite over the
operating life of the Subtitle C, RCRA-permitted facility would be
equal to or less than 1 mrem/year (10 Sv/year), if integrated
over the facility's operating life.
Again, the activity constraint and the water pathway considerations
can be placed in perspective by evaluating the potential normal
disposal of EAF emission control dust at a Subtitle C, RCRA-permitted
facility. If this dust includes a background Cs-137 concentration of 1
pCi/g (0.037 Bq/g), and the facility can treat 200 tons of dust per
day, the total quantity of Cs-137 disposed of annually would be about
50 mCi (1.85 GBq). Thus, over a facility operating period of about 20
years, the total quantity of Cs-137 disposed of could equal the 1-curie
(37 GBq) incident-related material activity constraint.
Intruder Considerations
In the development of its licensing requirements for land disposal
of radioactive waste in 10 CFR Part 61, NRC considered protection for
individuals who might inadvertently intrude into the disposal site,
occupy the site, and contact the waste. In the context of this
position, this possibility has been considered although the greater
risk to the intruder would likely result from the non-radiological
hazardous constituents at the site.
In the intruder scenarios applied in the development of NRC's low-
level
[[Page 1617]]
waste standards,27 an inadvertent intruder was assumed to dig a 3-
meter (9.9 ft) deep foundation hole for construction of a house. The
top 2 meters (6.6 ft) of the foundation were assumed to be trench cover
material and the bottom 1 meter (3.28 ft) was assumed to be waste.
Based on the details of the scenarios, which included these and other
considerations, the intruder interacted with material whose
concentration had been reduced from the waste concentration by a factor
of 10. Presuming similar scenarios and assuming intrusion occurs
immediately after a post-closure care period of 30 years, the intruder
would be exposed to a Cs-137 concentration of about 4 pCi/g (0.15 Bq/
g); that is, 77 pCi/g (2.84 Bq/g) reduced by the factor of 10 and an
additional factor of 2 to account for radioactive decay). Even for this
worst-case situation in which all the incident-related waste was
presumed to have initial Cs-137 concentrations of 77 pCi/g (2.84 Bq/g),
the projected intruder exposure would range from 0.8 to 3.8 mrem (8 to
38 Sv/year).28 As noted above, the average concentrations
over large volumes of incident-related material would be expected to be
far less than 77 pCi/g (2.84 Bq/g).
\27\ See NUREG-0782, vol. 4, Draft Environmental Impact
Statement on 10 CFR Part 61, ``Licensing Requirements for Land
Disposal of Radioactive Waste,'' September 1981.
\28\ These estimates are based on the concentration to dose
conversion values in NUREG-1500, ``Working Draft Regulatory Guide on
Release Criteria for Decommissioning: NRC Staff's Draft for
Comment,'' August 1994. Appropriate adjustments of the tabulated
information were made to reflect the occupancy and shielding
assumptions made in NUREG-0782 (see Footnote 24).
---------------------------------------------------------------------------
Conclusions
These bounding analyses indicate that some significant volume of
Cs-137-contaminated emission control dust and other incident-related
materials from an inadvertent melting of a sealed source can be
disposed of at a Subtitle C, RCRA-permitted facility with negligible
impacts to public and worker health and safety and the environment.
This method for disposal, if implemented according to the limitations
stipulated in this position, is very unlikely to cause worst-case
exposures that exceed 1 mrem (10 Sv) to any worker at the
disposal facility or to any member of the public in the vicinity of the
facility. The design, operations, and post-closure activities that take
place at Subtitle C, RCRA-permitted facilities will ensure that
radiological impacts from Cs-137 will also be negligible in future
timeframes. Proper disposal of these materials would protect public
health and safety, and the environment to a greater degree than the
alternative of indefinitely storing these materials at a steel company
facility. The calculated public health and safety and environmental
impacts of disposition of specified incident-related materials at a
Subtitle C, RCRA-permitted facility can also be used to determine an
optimum course for disposal, if disposition alternatives exist.
BILLING CODE 7590-01-P
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BILLING CODE 7590-01-P
Dated at Rockville, Maryland, this 11th day of January, 1996.
For the Nuclear Regulatory Commission.
Michael F. Weber,
Chief, Low-Level Waste and Decommissioning Projects Branch, Division of
Waste Management, Office of Nuclear Material Safety and Safeguards.
[FR Doc. 96-703 Filed 1-19-96; 8:45 am]
BILLING CODE 7590-01-O
