[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 

[[Page 1611]]
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.
---------------------------------------------------------------------------

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

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

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

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

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

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

    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.

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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