[Federal Register Volume 62, Number 53 (Wednesday, March 19, 1997)]
[Notices]
[Pages 13176-13198]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-6884]


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NUCLEAR REGULATORY COMMISSION

Disposition of Cesium-137 Contaminated Emission Control Dust and 
Other Incident-Related Material; Final Staff Technical Position

AGENCY: U.S. Nuclear Regulatory Commission.

ACTION: Notice: final staff technical position.

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SUMMARY: The U.S. Nuclear Regulatory Commission is issuing 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 incident-
related mixed waste. Mixed waste is a waste that not only is 
radioactive, but also is classified as hazardous under the Resource 
Conservation and Recovery Act (RCRA). The specific mixed waste 
addressed in this position is emission control dust from electric arc 
furnaces (EAFs) or foundries that has been contaminated with cesium-137 
(\137\Cs). The contamination results from the inadvertent melting of a 
\137\Cs 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 the production process and 
thereby can and has contaminated large volumes of emission control dust 
and the emission control systems at steel-producing facilities.
    The 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 a large part of this mixed waste, much 
of which contains \137\Cs 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 possible public input, could be used to allow disposal of 
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 position has been developed through an open public process in 
which working draft documents have been routinely shared with EPA, and 
also placed in NRC's Public Document Room to allow interested party 
access. NRC published the proposed position in the Federal Register for 
comment (61 FR 1608, dated January 22, 1996). NRC is now publishing the 
entire final position, together with its responses to the comments 
received.

FOR FURTHER INFORMATION CONTACT: Dominick A. Orlando, Division of Waste 
Management, Office of Nuclear Material Safety and Safeguards, Mail Stop 
TWFN 8F-37, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 
Telephone (301) 415-6749.

SUPPLEMENTARY INFORMATION:

Disposition of Cesium-137 Contaminated Emission Control Dust and Other 
Incident-Related Materials; Branch Technical Position

A. Introduction

    Emission control (baghouse) dust and other incident-related 
materials (e.g., clean-up materials or recycle process streams) 
contaminated with \137\Cs <SUP>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 \137\Cs quantity ranging downward from a little more than 1 curie 
(Ci) or 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 clean-up or dust-recycle, process-
stream materials. In current situations, most, but not all, of this 
material would be classified as mixed waste and this technical position 
is intended as a potential disposition alternative for this incident-
related material.<SUP>2
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    \1\ The byproduct material \137\Cs does not include the \137\ 
Cs, 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 \137\Cs-
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 radioisotope.
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    Typically, the radioactivity is not evenly distributed among the 
incident-related materials. Rather, 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, incident-related 
material volume. This material is classified as hazardous waste under 
RCRA because it contains lead, cadmium, and chromium which are common 
to the recycle metal supply. The \137\Cs contamination of this 
hazardous waste results from a series of three principal events: (1) 
The loss of control of a radioactive source by an NRC or an 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 (including radiation detectors used at 
most furnaces, foundries and many ferrous metal recycling facilities). 
Consequently, irrespective of the quantity or concentration of the 
radioactivity, most of the current 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.

[[Page 13177]]

    The disposal options for these materials, specifically the large 
volumes of material with the lower concentrations of \137\Cs, 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 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 find acceptable for: (1) Authorizing a licensee, possessing 
\137\Cs-contaminated emission control dust and other incident-related 
materials (e.g., the steel company or its service contractor), to 
transfer treated \137\Cs-contaminated material, below levels specified 
in this position, to a Subtitle C, RCRA-permitted hazardous waste 
disposal facility; and (2) not licensing the possession and disposal of 
these incident-related materials by the RCRA-permitted disposal 
facility. The position does not address disposal at a Subtitle D 
facility. Because of its radioactivity (i.e., \137\Cs 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: (a) at a licensed low-level 
radioactive waste disposal facility after appropriate treatment of its 
hazardous constituents; or (b) at a mixed-waste disposal facility, if 
applicable acceptance criteria are met.
    The regulatory basis for the action is found at 10 CFR 
20.2001(a)(1) and 20.2002. The first paragraph authorizes a licensee to 
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. In the case of the \137\Cs-contaminated 
material, the licensing action under 10 CFR 20.2002 would constitute 
the written authorization required by paragraph 30.41(b)(7).
    It should be noted that additional acceptance requirements, beyond 
those covered in this NRC position for disposal of \137\Cs-contaminated 
incident-related 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 \137\Cs-contaminated incident-related 
materials must consult with these parties, and obtain all necessary 
approvals, in addition to those of NRC and/or appropriate Agreement 
States, for 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. The position applies to 
both hazardous and non-hazardous incident-related waste as specifically 
defined. In addition, the conditions established in this position 
pertain to NRC staff and licensee actions. Therefore, in those 
instances where an Agreement State is the sole regulatory authority for 
the radioactive material, the Agreement State has the option of using 
this guidance in reviewing requests for the disposal of the material.

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 are deposited 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.<SUP>3
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    \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 \137\Cs 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 \137\Cs.
    When \137\Cs sources are inadvertently melted with a load of scrap 
metal, a significant amount of the \137\Cs 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 hazardous constituents--specifically 
lead, cadmium, and chromium--EAF emission control dust is a listed 
waste, KO61, which is subject to regulation under RCRA. If this dust 
becomes contaminated with \137\Cs, the resulting material is classified 
as a mixed waste. Emission control dust, generated immediately after 
the melting of a \137\Cs 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 \137\Cs 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 \137\Cs 
(<difference>37 GBq)) 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

[[Page 13178]]

to downstream processes have also been contaminated with relatively low 
concentrations of \137\Cs (e.g., 10 pCi/g (0.37 Bq/g)). In addition, 
materials used during decontamination may also be contaminated with 
dust containing \137\Cs concentrations at similar levels above 
background.
    As the result of past inadvertent meltings of \137\Cs sources, a 
number of steel producers possess a total of about 10,000 tons (9000 
metric tons) of incident-related materials, most of which contains 
\137\Cs 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.<SUP>4 It is the disposition of material at these 
concentration levels that is the subject of this technical position.
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    \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 \137\Cs-
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 \137\Cs-
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 \137\Cs contaminated dust and other incident-related 
materials (the subject of this technical position).
Specific
Bases for Allowing Transfer and Possession of <SUP>137Cs-Contaminated, 
Incident-Related Material
    The bases for allowing transfer and possession of \137\Cs-
contaminated emission control dust and other incident-related 
materials, under the provisions of existing regulations, are as 
follows: (1) Any 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 microsievert (<greek-m>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; 
<SUP>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 <greek-m>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 must be accomplished by persons operating under a licensee's 
radiation protection program (note that the licensee can be the steel 
facility or the entity that treats the incident-related material, 
either on- or offsite); and (5) transportation of contaminated 
materials will be subject to U.S. Department of Transportation (DOT) 
regulations and, as applicable, transportation of contaminated, 
hazardous materials must be performed by hazardous material employees, 
as defined in DOT regulations (49 CFR Part 172, Subpart H).
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    \5\ The use of 1 mrem/yr (10 <greek-m>Sv/yr) 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 \137\Cs 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 clean-up and collection/
treatment/packaging of the contaminated emission control dust and other 
materials at the EAF or foundry must be performed by an NRC or 
Agreement State licensee operating under an approved radiation 
protection program. The licensee is also responsible for compliance 
with other 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 \137\Cs-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 must be treated 
through stabilization to comply with all EPA and/or State waste 
treatment requirements for land disposal of regulated hazardous 
waste.<SUP>6 The treatment operations must 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 \137\Cs-
contaminated, incident-related materials); or (ii) an NRC- or Agreement 
State-licensed service contractor (operating either on- or offsite). 
Based on the radiological impact assessment provided in Appendix A, the 
licensee could be authorized by NRC or an Agreement State to transfer 
the treated incident-related materials to a Subtitle C, RCRA-permitted, 
disposal facility, provided that all the following conditions are met:
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    \6\ For non-hazardous material covered by this position, 
stabilization equivalent to that provided for hazardous waste would 
be necessary.
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    (a) The \137\Cs-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 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 \137\Cs 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.

[[Page 13179]]

Moreover, NRC or the appropriate Agreement State will maintain a public 
record of the total incident-related \137\Cs activity, received by the 
facility over its operating life, to ensure that the total disposed of 
\137\Cs activity does not exceed 1 curie (37 GBq); <SUP>7
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    \7\ The 1-curie (37-GBq) value represents a reasonable maximum 
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 \137\Cs.
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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; 
<SUP>8
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    \8\ The NRC staff believes the contract between the licensed 
facility and the RCRA facility operator is an appropriate vehicle 
for complying with this provision, provided that the contract 
specifies the volume of waste, the radionuclide and its average 
concentration in the waste in picocuries per gram or becquerels per 
gram, the total aggregated amount of radioactive material in the 
shipment, the hazardous waste code of the waste, and the EPA 
identification number of the RCRA disposal facility receiving the 
waste. The NRC staff will evaluate requests for license amendments 
to transfer incident-related material based upon the licensee 
demonstrating that the RCRA disposal facility operator has agreed to 
the transfer and has made provisions to retain the information about 
the radioactive material in the waste, along with the information 
that is required to be retained by the RCRA facility operator under 
40 CFR 263.22.
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    (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 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 2000 pCi/g (74 
Bq/g) is not considered radioactive);
    (g) In any package, the emission control dust and other incident-
related materials, that have been stabilized and packaged as defined in 
(b) and (f) above, contain pretreatment average concentrations of 
\137\Cs that did not exceed 130 pCi/g (4.8 Bq/g) of material; <SUP>9 
and
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    \9\ 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 of less 
than 1 mrem (10 <greek-m>Sv). The disposal of incident-related 
materials in packaged form allows compliance with this position to 
be demonstrated through measurement of \137\Cs 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 stabilized waste does not exceed 20 <greek-m>rem per 
hour or 0.20 <greek-m>Sv per hour, above background.<SUP>10

    \10\ At this exposure rate, for the exposure period as defined 
in the appendix, total exposure would not exceed 1 mrem (10 
<greek-m>Sv) with a regulatory margin of 1.5.
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Note that, in defining the pretreatment \137\Cs-concentration value 
stated in paragraph (1)(g), a factor of 1.5 has been included as a 
regulatory margin. This factor adds further assurance to the certainty 
in protection provided by the licensee's: (1) Sampling of \137\Cs 
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 must also be treated 
through stabilization to comply with all EPA and State waste treatment 
requirements for land disposal of RCRA-regulated hazardous 
waste.<SUP>11 The treatment operations must be undertaken by either (i) 
the owner/operator of the EAF or foundry (licensed to possess, treat, 
and transfer <SUP>137Cs-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 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):
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    \11\ See footnote 6.
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    (a) The <SUP>137Cs-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 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 <SUP>137Cs 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 will 
maintain a public record of the total incident-related <SUP>137Cs 
activity, received by the facility over its operating life, to ensure 
that the total disposed of <SUP>137Cs activity does not exceed 1 curie 
(37 GBq); <SUP>12
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    \12\ See footnote 7.
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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; 
<SUP>13
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    \13\ The NRC staff believes the contract between the licensed 
facility and the RCRA facility operator is an appropriate vehicle 
for complying with this provision, provided that the contract 
specifies the volume of waste, the radionuclide and its average 
concentration in the waste in picocuries per gram or becquerels per 
gram, the total aggregated amount of radioactive material in the 
shipment, the hazardous waste code of the waste and the EPA 
identification number of the RCRA disposal facility receiving the 
waste. The NRC staff will evaluate requests for license amendments 
to transfer incident-related material based upon the licensee 
demonstrating that the RCRA disposal facility operator has agreed to 
the transfer and has made provisions to retain the information about 
the radioactive material in the waste along with the information 
that is required to be retained by the RCRA facility operator under 
40 CFR 263.22.
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    (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 stabilized as defined in (b) above, contain pretreatment 
average concentrations of <SUP>137Cs that did not exceed 100 pCi/g (3.7 
Bq/g) of material.<SUP>14
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    \14\ 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 <greek-m>Sv). The disposal of incident-related material 
in unpackaged (bulk) form dictates that compliance with this 
position would be demonstrated through measurement of <SUP>137Cs 
concentrations. Without the redundant approach to ensure compliance 
with the exposure criterion inherent with the packaged-disposal 
approach (see footnote 8), 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 <SUP>137Cs-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 <SUP>137Cs 
concentrations in

[[Page 13180]]

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 <SUP>137Cs 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 <SUP>137Cs-
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 
<SUP>137Cs concentration within the emission control dust and other 
incident-related materials is 2 pCi/g (0.074 Bq/g) of material or less. 
The foundry or licensed service contractor must determine the 
<SUP>137Cs concentration using the sampling program discussed below.

Aggregation of <SUP>137Cs-Contaminated Emission Control Dust and Other 
Incident-Related Materials

    If applicable, aggregation of <SUP>137Cs-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 waste; and
    (3) Materials, when aggregated, are subjected to a sampling 
protocol that demonstrates compliance with <SUP>137Cs-concentration 
criteria on a package-average <SUP>15 basis.
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    \15\ 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 <SUP>137Cs Concentrations and Radiation Measurements

    <SUP>137Cs 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 sampling program 
must be sufficient to ensure that <SUP>137Cs contamination in the 
stabilized 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 
must provide assurance that the quantity of <SUP>137Cs in any package 
(see footnote 15) does not exceed the product of the applicable 
concentration criterion times the net weight of contaminated material 
in a package.

Appendix A--Assessment of Radiological Impact of Disposal of 
<SUP>137Cs-Contaminated Emission Control Dust and Other Incident-
Related Materials at a Subtitle C RCRA-Permitted Disposal Facility

1. 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, that can contain 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, can cause 
the U.S. Environmental Protection Agency (EPA) to designate the dust 
as a hazardous waste, under the Resource Conservation and Recovery 
Act (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 preclude these recycle options, the dust may be 
treated and disposed of in a hazardous waste disposal facility. EPA 
has specified treatment standards for the various hazardous 
constituents of the dust in 40 CFR 268.40. Solidification is the 
treatment process typically used to meet these standards. On the 
other hand, dust from steel production at basic oxygen furnaces and 
open hearth furnaces is excluded from regulation as hazardous waste 
(40 CFR 261.4(b)(7)(xvii)).
    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 cesium-137 
(\137\Cs) which now occurs in the environment as a result of global 
fallout from past weapons-testing programs. \137\Cs 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 
\137\Cs 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 or <difference>1650 degrees celsius) steel-making 
process, \137\Cs is volatilized and transported in the furnace off-
gas and, as it condenses, becomes a constituent of the emission 
control (baghouse) dust. Normal background \137\Cs concentrations in 
dust have been measured at picocurie per gram levels (0.024 to 1.23 
pCi/g) <SUP>1 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.<SUP>2 As a result of this information, the U.S. 
Nuclear Regulatory Commission has determined that \137\Cs 
concentrations in emission control dust below 2 pCi/g (0.074 Bq/g) 
can be attributed to fallout from past weapons testing.<SUP>3
---------------------------------------------------------------------------

    \1\ 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.
    \2\ Letter to William Lahs, U.S. Nuclear Regulatory Commission, 
from Andrew Wallo III, U.S. Department of Energy, dated May 20, 
1993.
    \3\ Letter from James M. Taylor, NRC, to William Guerry, Jr., 
Collier, Shannon, Rill, and Scott, dated May 25, 1993.
---------------------------------------------------------------------------

2. Statement of Problem

    The inadvertent melting of a licensed \137\Cs sealed source with 
scrap steel at an electric arc furnace (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 or 900 metric tons) of \137\Cs-
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 \137\Cs 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 (1800 metric 
tons) per event. The \137\Cs 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 \137\Cs-contaminated emission control dust and other 
incident-related materials at a Subtitle C, RCRA-permitted facility. 
Because 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 \137\Cs-
contaminated emission control dust and other incident-related

[[Page 13181]]

materials be performed by an NRC or Agreement State licensee. The 
licensee would operate, either on- or offsite, 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 incident-related 
material. This stabilized material would be, or would be equivalent 
to, the form necessary to stabilize the RCRA-hazardous constituents 
(specifically, lead, cadmium, and chromium); that is, a non-
dispersible,<SUP>4 solid (e.g., cement-type) form. As a result, the 
potential radiological hazard from the ``treated'' (stabilized) 
material during disposal operations is associated with its 
characteristic as an external source of radiation.
---------------------------------------------------------------------------

    \4\ 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 
\137\Cs decay process.
---------------------------------------------------------------------------

    After disposal, \137\Cs could only become a hazard through water 
pathways if a sufficient quantity and concentration of \137\Cs 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 of waste after 
facility closure. Notwithstanding the hazard to the intruder from 
the hazardous waste constituents, or other hazardous wastes, 
constraints placed on the total \137\Cs 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 \137\Cs 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.

3. Direct Exposure

    After the inadvertent melting of a \137\Cs sealed source at an 
EAF or foundry, the relatively volatile \137\Cs will leave the 
furnace as an offgas and be commingled with the normal emission 
control dust. As a result, concentrations of \137\Cs contained in 
this dust (and other materials associated with furnace clean-up 
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 \137\Cs. 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 \137\Cs concentrations; (2) variations in the 
physical/chemical properties of the materials disposed of; and (3) 
changes in worker time-integrated interactions with contaminated 
materials.
    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 who has been permitted for 
stabilization treatment of the material either on or off the steel 
company site. 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 treatment-related 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 \137\Cs-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 (1800 metric tons), 
and a total activity content of less than a 1 curie (Ci) or 37 GBq 
of \137\Cs. 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.<SUP>5 The 
calculations assume that the initial \137\Cs 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.
---------------------------------------------------------------------------

    \5\ 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,<SUP>6 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).<SUP>7 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 
\137\Cs 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/cm<SUP>3. A 
representative dose [rate] conversion factor <SUP>8 under these 
conditions (calculated at a density of 1.5 g/cm<SUP>3) would 
typically be less than 49 microem/hour (<greek-m>rem/hr) or 0.49 
microsieverts/hour (<greek-m>Sv/hr), at a distance of 3.28 feet (1 
meter) from the surface of a hypothetical semi-infinite volume of 
the solidified material.\9\
---------------------------------------------------------------------------

    \6\ In the context of this position, stabilized treatment does 
not include either onsite or offsite high-temperature metals 
recycling processes.
    \7\ 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 
\137\Cs, although the radiological impacts analysis indicates that 
such processes are not necessary to protect public health and 
safety, and the environment.
    \8\ A dose conversion factor represents a value that allows a 
radionuclide contamination level to be converted to an estimated 
exposure rate.
    \9\ 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 <greek-m>rem (0.49<greek-m>Sv)/
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 <greek-m>rem/hour 
(0.14 <greek-m>Sv/hr) above background (typically 8 to 
12<greek-m>rem/hr (0.08 to 0.12 <greek-m>Sv/hr).<SUP>10 From an open 
trench (Figure 4), filled with

[[Page 13182]]

contaminated materials, the calculated dose rate would also be 
somewhat less than 13 <greek-m>rem/hr (0.13 <greek-m>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.
---------------------------------------------------------------------------

    \10\ 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 generally exceed 500 tons (450 metric tons) per 
shift.<SUP>11 Assuming this disposal rate of 500 tons (450 metric 
tons) per shift applies to the disposal of treated,<SUP>137Cs-
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 (1800 metric 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 <greek-m>rem/hr (0.13 <greek-m>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 a dose of less than 0.5 mrem 
(5 <greek-m>Sv) above background.
---------------------------------------------------------------------------

    \11\ 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 (90 to 180 metric 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, at this distance, 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 <greek-m>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 
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-g sample would be no 
greater than about 10<SUP>-2 <greek-m>Ci (370 Bq). The dose rate 
from such a sample would be less than 0.1 <greek-m>rem/hr (0.001 
<greek-m>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 <SUP>137Cs from a 
melted source). This dust would contain background levels of 
<SUP>137Cs (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 of about 0.5 mrem (5 
<greek-m>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-Ci (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 
<SUP>137Cs-contaminated emission control dust and other incident-
related materials will not exceed 1 mrem/year (10 <greek-m>Sv/year) 
integrated over the lifetime of the facility.
    4. Water-Pathway Considerations
    The proposed approach to manage <SUP>137Cs-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 radiological, and 
potentially hazardous 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 <SUP>137Cs 
presenting a hazard through a water pathway requires consideration 
of: (1) the quantity of <SUP>137Cs available; (2) the degree to 
which the <SUP>137Cs could be leached from its waste matrix; and (3) 
the extent that any leached <SUP>137Cs could migrate into a water 
supply.
    The disposal of <SUP>137Cs in treated emission control dust and 
other incident-related materials would be constrained by this policy 
to a total activity of 1 Ci (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 <SUP>137 Cs averaged over all the treated waste 
would typically be significantly less than the defined concentration 
criteria.
    Furthermore, because the <SUP>137Cs is contained in a solid 
matrix and buried within a facility in which the amount of water 
infiltration is minimized, any <SUP>137Cs removal from its final 
disposal location would be limited while these conditions remain in 
effect. The chemistry of any water interacting with the solidified, 
<SUP>137Cs-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 <SUP>137Cs from the 
waste would normally be collected in a leachate collection system at 
volumetric concentrations expected to be far less than those 
existing in the treated waste. The chemistry of the fill materials 
used at the disposal site could also provide a sorbing medium if any 
<SUP>137 Cs 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 <SUP>137Cs 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 
<SUP>137Cs to reach a water supply, is such that significant 
radioactive decay of the <SUP>137Cs inventory is likely (the 
radioactive half-life of <SUP>137Cs is 30 years) before the 
<SUP>137Cs could potentially reach the water supply.
    Although qualitative in nature, and based on considerations that 
can vary among Subtitle C, RCRA-permitted disposal sites, the 
previous 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.
    The leachability of <SUP>137Cs from any solid waste form that 
complies 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 <SUP>137Cs 
under these conditions, a critical process is the partitioning of 
the <SUP>137Cs 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 
<SUP>137Cs that can leach into the infiltrating water.
    The most important parameter in estimating this transfer, as 
well as the subsequent movement of the <SUP>137Cs in groundwater, is 
the distribution coefficient, ``K<INF>d.'' This parameter expresses 
the ratio at equilibrium of <SUP>137Cs sorbed onto a given weight of 
soil particles to the amount

[[Page 13183]]

remaining in a given volume of water. The higher the value of the 
distribution coefficient, the greater the concentration of 
<SUP>137Cs remaining in the soil. The K<INF>d value can be affected 
by factors such as soil texture, pH, competing cation effects, soil 
porewater concentration, and soil organic matter content.<SUP>12 For 
the non-acidic, sand/clay/soil environments presumed to represent 
the RCRA-permitted disposal facilities, a K<INF>d value of 270 
milliliter (ml)/g was selected from the Footnote 12 reference as 
being appropriate for the subsequent bounding, conservative 
analysis.
---------------------------------------------------------------------------

    \12\ ``Default Soil Solid/Liquid Partition Coefficients, 
K<INF>ds, 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 <SUP>13 
code was used. For the representative case, the bounding 40,000 
cubic feet (ft \3\) or 1132 cubic meters (m<SUP>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 <SUP>137Cs 
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 of material. 
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 K<INF>d in the unsaturated zone was 270 ml/g. 
Assessments beyond this representative case evaluation are 
subsequently discussed.
---------------------------------------------------------------------------

    \13\ 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<SUP>-2 <greek-m>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, is sufficient to 
preclude drinking water doses for almost 700 years. During this 
period, the activity of <SUP>137Cs 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 <greek-m>Sv/year).<SUP>14
---------------------------------------------------------------------------

    \14\  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 K<INF>d values, on the results of this bounding analysis, the 
following analyses are presented. Based on the typical existing 
volumes and <SUP>137Cs concentrations of incident-related materials, 
the imposition of a constraint on <SUP>137Cs 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 (a few GBq).<SUP>15 Material at 
higher concentrations would require disposal at either a mixed-waste 
disposal facility or a licensed low-level waste disposal site. Thus, 
for the potential disposals at the Subtitle C, RCRA-permitted site 
to approach the 1-Ci (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 an RCRA-permitted 
facility's disposal capacity. Repeating the RESRAD analysis 
discussed above under these assumptions, but respectively 
considering lower K<INF>d values in the contaminated, unsaturated, 
and saturated zones, would still result in drinking water doses of 
less than 1 mrem (10 <greek-m>Sv) per year unless the K<INF>d values 
in all zones approach single-digit values. Even in these cases 
(e.g., K<INF>d equal to 2.7), separation of the hypothesized well 
location from the disposed material by about 328 feet (100 meters) 
would reduce dose rates below 1 mrem (10 <greek-m>Sv) per year 
because of the decay of <SUP>137 Cs brought about by the increased 
retardation times.
---------------------------------------------------------------------------

    \15\ For example, the total activity contained in 2000 tons 
(1800 metric tons) of material, contaminated at a level of 77 pCi/g 
(2.84 Bq/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 
incident-related <SUP>137 Cs activity constraint of 1 Ci (37 GBq). 
The magnitude of this constraint is based on the typical bounding 
activity associated with an inadvertent melting of <SUP>137 Cs 
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 <greek-m>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 
<SUP>137Cs concentration of 1 pCi/g (0.037 Bq/g), and the facility 
can treat 200 tons (180 metric tons) of dust per day, the total 
quantity of <SUP>137Cs 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 <SUP>137Cs disposed of could equal the 
1-Ci (37 GBq) incident-related material activity constraint.

5. 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 
LLW standards, <SUP>16 an inadvertent intruder was assumed to dig a 
3-m (9.9-ft) deep foundation hole for construction of a house. The 
top 2 m (6.6 ft) of the foundation were assumed to be trench cover 
material and the bottom 1 m (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 137Cs 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 137Cs 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 <greek-m>Sv/year).<SUP>17 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).
---------------------------------------------------------------------------

    \16\ See NUREG-0782, Vol. 4, Draft Environmental Impact 
Statement on 10 CFR Part 61, ``Licensing Requirements for Land 
Disposal of Radioactive Waste,'' September 1981.
    \17\ 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 16).
---------------------------------------------------------------------------

6. Conclusions

    These bounding analyses indicate that some significant volume of 
<SUP>137Cs-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

[[Page 13184]]

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 
<greek-m>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 <SUP>137Cs 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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Analysis of and Response to Comments on Staff Technical Position 
``Disposition of Cesium-137 Contaminated Emission Control Dust and 
Other Incident-Related Material''

    On January 22, 1996, the Nuclear Regulatory Commission published 
a proposed technical position on the disposition of Cesium-137 
(<SUP>137Cs) contaminated emission control dust and other incident-
related materials in the Federal Register (61 FR 1608). Comments 
were solicited and, in response, 22 comment letters were received. 
These comment letters included: six from State nuclear safety, human 
resources, environmental conservation, and health offices; five from 
the steel industry (three from industry associations); four from 
hazardous waste disposal facility operators; one from a mixed-waste 
disposal facility operator; two from other industry associations; 
two from environmental groups; one from a member of the public; and 
one from a member of Congress. These letters raised a number of 
issues ranging from policy and legal concerns to specific comments 
on the conservatisms in the position's supporting radiological 
assessments. The responses to these submitted comments have been 
grouped into the following categories: (1) Position justification, 
precedence, and relationship to ``below regulatory concern'' (BRC) 
policy considerations and constraints; (2) regulatory approach 
(i.e., rulemaking versus technical position) and the implementation 
process; (3) legal considerations; (4) related health, safety, and 
environmental concerns; (5) technical considerations; (6) other 
issues; and (7) clarifications.

1. Justification, Precedence, and Relationship to BRC

    a. Comment: The comments on the justification issue were 
intertwined with several other issues. Basically, however, the 
comments from the steel and other industries (associations), three 
States, the Subtitle C, hazardous waste facility operators, and the 
Congressman supported the disposal concept, proposed in the 
position, as a necessary adjunct to a regulatory program that should 
improve licensee control of the devices, whose inadvertent meltings 
have caused the problem. They believe that the position is safe, 
environmentally sound, and cost-effective, and a reasonable 
alternative to disposal at the existing mixed-waste disposal 
facility (available since April 1995 for disposal of the subject 
waste). These commenters, plus those from the disposal facility 
operators, expressed the view that there was a serious lack of 
competition in the business of disposing of mixed waste, resulting 
in unacceptably high disposal costs. Several of these commenters, 
including one State, suggested that the disposal costs could affect 
the financial viability of certain facilities.
    Comments from two of the States, the environmental groups, the 
member of the public, and the mixed-waste disposal facility operator 
stated that the justification (or combined justification and 
regulatory approach) for the position was weak, with most 
emphasizing the conflict with the current policy on, and approach 
to, mixed-waste disposal. One State and the mixed-waste disposal 
facility operator noted that the time, effort, and resources 
expended to effect mixed-waste disposal at the existing facility 
will be undermined by the precedent being established in the 
position. The disposal facility operator also believed that the 
position relied on unsubstantiated economic assumptions and 
assertions, which NRC had accepted at face value without any 
independent investigation, and that the justification had changed 
with position development from a public-health to a cost-effective 
rationale. This commenter also stated the belief that the position 
was not in the public's interest because of potential exposures to 
transportation workers, members of the public, and Subtitle C 
facility workers. One of the environmental group commenters stated 
that saving a few cents per ton of steel may not be in the public's 
best interests.
    a. Response: Because the subject disposition option was being 
proposed in the form of a ``permissive'' technical position (i.e., 
an option, likely requiring multiple approvals, that would be 
initiated by a request from an NRC or Agreement State licensee on a 
case-by-case basis) and not a rule, the broad policy justification, 
stated in the Federal Register notice (61 FR 1609, column 1) was as 
follows: ``NRC believes that disposal, under the provisions of the 
position or other acceptable alternatives (emphasis added), is 
preferable to allowing this mixed waste to remain indefinitely at 
steel company sites.'' Another acceptable alternative referred to in 
the Federal Register notice (61 FR 1610, column 3) is the mixed 
waste disposal facility operated by Envirocare of Utah. As the 
footnote indicated, this facility received authorization to accept 
the subject waste at concentrations not exceeding 560 picocuries 
(pCi)/gram (g) (20.7 becquerel (Bq)/g) during April 1995, after the 
work on the technical position was initiated. With the availability 
of this facility, the NRC staff believes its aforementioned public 
health justification became enmeshed with cost considerations, as 
described below.
    Notwithstanding the availability of the mixed-waste disposal 
facility option, which has been used by several steel facilities 
with <SUP>137Cs-contaminated, incident-related material, other steel 
companies did not consider this option cost-effective. Contrary to 
the commenter's statement, NRC staff did contact the industry, as 
well as the mixed-waste and Subtitle C disposal facility operators, 
to comprehend the possible cost differentials of the disposal 
alternatives. Although contractual privacy and market considerations 
prevented exact determination, NRC staff concluded that the 
differential costs between the mixed-waste and Subtitle C disposal 
options could be significant. In fact, the comment letters from the 
mixed-waste disposal facility operator and the Subtitle C facility 
operators appear to confirm this assessment. But whatever the actual 
cost differences may be for specific situations, the process, as 
envisioned in the potential use of the position, would identify the 
cost differentials, if any, in the environmental assessment that 
would support any decision to implement the position's disposition 
alternative (One State commenter stated that it should be made clear 
that an environmental assessment, under the National Environmental 
Policy Act, would be required for each disposal). The significance 
of any cost differential could be judged by appropriate regulatory 
authorities, in their selection of the most reasonable and proper 
disposal alternative (e.g., whether saving a few cents per ton of 
steel is in the public interest).
    In its decision to pursue what, the Commission believes, is a 
health-protective and environmentally sound disposition alternative, 
the Commission also considered the origin of the radioactive source 
melting problem (a problem being addressed under a separate NRC 
program) and the significant efforts of the steel industry to detect 
incoming radioactive material. In the Commission's view, these 
factors provided further justification for its ongoing actions. 
Thus, although the Commission is aware of the substantial efforts, 
time, and resources expended by all parties involved in the 
licensing of the mixed-waste disposal facility, the Commission's 
primary focus is to achieve ultimate disposition of the incident-
related material. The Commission believes that the real or imagined 
cost differentials, from lack of competition or other causes, may be 
resolved through issuance of the position, and lead to a resolution 
of the disposal problem. The Commission has coordinated its actions 
with the U.S. Environmental Protection Agency (EPA) and believes it 
has the support of EPA in the position, at this proposal stage.
    The potential exposures to workers and members of the public are 
addressed in the responses to comments 3.a. and 4.f.
    b. Comment: Two State commenters, the mixed-waste disposal 
facility operator, the environmental group commenters, and the 
member of the public raised concerns about the precedent-setting 
nature of the technical position. By establishing a ``default'' 
value for <SUP>137Cs in incident-related material, it was questioned 
why a similar argument could not be made for <SUP>137Cs in soil, or 
some other radionuclide in another medium. It was further pointed 
out that the cumulative effect of similar actions would need to be 
addressed. An environmental group commenter opposed the creation of 
exemptions that could be used by others, specifically the Department 
of Energy, as applicable or relevant and appropriate requirements, 
in settings for which the drafters of the [position] did not intend 
or anticipate. The member of the public claimed a possible 
relationship to issues involving disposals from the U.S. Enrichment 
Corporation.
    Commenters from the steel and other industry associations 
supported the position as a proper precedent, and suggested that 
other circumstances could justify similar actions. It was pointed 
out, for example, that although Basic Oxygen Furnaces (BOFs) also 
process scrap and are subject to the same kinds of incidents as 
Electric Arc Furnaces, BOF dust may be neither a listed hazardous 
waste nor a characteristic hazardous waste. It was suggested that 
the position be clarified regarding its applicability to the 
potential

[[Page 13194]]

disposal of <SUP>137Cs-contaminated BOF material and incident-
related material that may not be classified as mixed waste. Another 
industry commenter questioned whether the position would apply to a 
steel producer who is not an NRC or Agreement State licensee. This 
same commenter questioned what approach would be used for meltings 
involving other radionuclides, and whether the position could not be 
broadened to other industries that have large volumes of mixed 
waste. In a broader sense, a few of these commenters applauded NRC 
and EPA efforts to minimize dual regulation of mixed waste.
    b. Response: The Commission strongly believes that broad-based 
exemptions, or creation of specific positions outside of established 
policies, should be implemented through the rulemaking process. In 
fact, efforts to provide the technical analyses to support a broad 
recycle rulemaking, that would include consideration of incident-
related material, are underway. However, under its specific 
regulations, cited in the technical position, the Commission can, 
and has, in case-by-case determinations, approved actions that it 
believes are in the best interests of public health and safety and 
protection of the environment. In the case of this ``permissive'' 
technical position, NRC is putting forward a disposition option, 
whose implementation and approval can be considered by applicable 
regulatory authorities and others. The advantages and disadvantages 
of alternatives would be addressed in appropriate environmental 
assessments that would accompany license amendment requests, and the 
choice would require acceptance by various regulatory authorities 
and others, and would be contingent on State laws and permit 
conditions.
    The Commission believes the precedent being suggested, in this 
case, is reasonable and proper, based on the circumstances and the 
justification, as described in 1.a. above. The disposition option, 
however, applies only to disposals at Subtitle C facilities; only to 
treated (stabilized) <SUP>137Cs-contaminated, incident-related 
material (inclusive of material that may not be classified as mixed 
waste) that constitutes the greatest part of the problem; and only 
to companies, or their service contractors, that will treat the 
incident-related material, under NRC or Agreement State license, to 
meet the land disposal requirements that would apply if the material 
contained hazardous constituents. This last provision was considered 
necessary to avoid the difficult task of generically defining bounds 
on the potential radiological exposures that could occur during 
treatment or disposal, and could involve consideration of inhalation 
and ingestion, as well as direct exposure pathways.
    With regard to other possible situations for which this position 
may be considered a precedent, such as the disposal of BOF material, 
the Commission believes these situations should be judged on their 
own merit. Any interactions among combined actions would require 
consideration, as one of the State commenters pointed out. The staff 
is also aware of EPA interests, identified in its proposed Hazardous 
Waste Identification Rule, and has encouraged EPA efforts to 
identify mixed wastes that may be regulated as low-level radioactive 
waste (LLRW), outside of Resource Conservation and Recovery Act 
(RCRA) regulations. The response to comment 2 is also pertinent to 
the question of developing a broader technical position. The staff 
has had no interactions with U.S. Enrichment Corporation issues that 
affected development of this position.
    c. Comment: Several commenters either requested clarification on 
the relationship of this position to BRC policy or stated their 
belief that the position contravenes public law.
    c. Response: In 1992, in response to the Commission's 
publication of a BRC policy statement in 1990, Congress, in Public 
Law 102-486, Energy Policy Act of 1992, stated that the [BRC] policy 
shall have no effect. The BRC policy basically stated the bases that 
the Commission would apply to determine if broad practices should be 
considered for exemption from regulatory control. The NRC staff does 
not believe that the subject technical position is a BRC policy for 
the following reasons: (1) The technical position is a 
``permissive'' guidance statement, basically stating Commission 
views on safe implementation of existing regulations on licensed 
material disposal in 10 CFR Part 20; (2) the position specifically 
``directs'' the disposal to a regulated disposal entity, and 
includes approval, notification, and total activity provisions that, 
the staff believes, are inconsistent with the concept of BRC; (3) 
the position is narrow in scope (i.e., directed at specific 
material, caused by specific circumstances); and (4) if implemented, 
the actions under the position are consistent with other case-by-
case determinations made by the Commission.

2. Regulatory Approach and the Implementation Process

    Comment: Although related to the BRC issue discussed above, a 
State commenter questioned why, if the proposal is sound in 
protecting public health and safety, the regulatory approach is a 
technical position, as opposed to a rulemaking--the latter providing 
a broader review process. Another State commenter believed NRC 
should define the ``life expectancy'' of the guidelines in the 
technical position. Industry comments generally supported the 
technical position as the approach needed to address a real problem 
in a timely manner, as opposed to a rulemaking that would be very 
time-consuming. They believe the steel companies should not be put 
in the middle of a political tug-of-war over appropriate 
administrative procedures to follow, given that the position has 
been made available for public scrutiny in a manner similar to a 
proposed rule.
    Response: As referred to in 1.b. above, NRC staff intends to re-
address the Subtitle C-disposal option, proposed in the technical 
position, in conjunction with a broad recycle rulemaking. At that 
time the need to broadly address disposal options will be revisited. 
However, because this rulemaking is in an early development stage, 
with finalization unlikely in the next couple of years, and because 
of the desirability of properly disposing of specific incident-
related material in a timely manner, the Commission directed that 
the staff should work with EPA to develop interim guidelines and 
associated technical bases. This is the process that has been 
followed to date. The guidelines proposed in the technical position 
would be in effect until this rulemaking is finalized.
    To address the concern of the State commenter regarding a 
broader review process, the staff has not only worked with EPA, but 
has made early versions of the position available directly to a 
number of affected parties and States. The Commission's intentions 
were openly discussed and the early versions of the position, 
together with early exchanges of views, were placed in NRC's public 
document room. The staff published the proposed technical position 
in its entirety in the Federal Register to obtain the broader review 
that the commenter suggests. Furthermore, contrary to interpretation 
of one commenter, NRC is not asserting the adoption of the technical 
position as a matter of Agreement State compatabilty. In fact, 
recognizing the likely involvement of many parties, if the 
position's alternative is implemented, the staff's intent was that 
this wide review and approval could be helpful in gaining general 
understanding and acceptance of the merits of the proposed 
alternative. Case-by-case reviews and approvals of individual 
applications to use the position's disposal approach will still be 
necessary even with the final technical position in place.

3. Legal Considerations

    a. Comment: In several comment letters from the States and a 
Subtitle C disposal facility operator, and in staff discussions with 
other Subtitle C disposal facility representatives, it was pointed 
out that the legal applicability of the technical position's 
disposal alternative, in specific States, could be determined by how 
the incident-related material is defined. If the waste were defined 
as LLRW, requiring disposal as specified in the Low-Level Waste 
Policy Amendments Act of 1985, the disposal alternative described in 
the position could be precluded, absent an appropriate change to 
State law or regulations, or permit conditions. One State commenter 
stated that, if the treated incident-related material is considered 
contaminated ash, it would be subject to permit and manifesting 
requirements.
    Another State commenter pointed out that State LLRW regulations 
require demonstration that design, operation, and closure of any 
class of LLRW facility ensure protection against inadvertent 
intrusion and provide for an institutional control period. There was 
concern about States being open to lawsuits if the incident-related 
material were considered LLRW and if the aforementioned provisions 
were not addressed. The mixed-waste disposal facility operator 
pointed out that Subtitle C facilities are not required to have 
radiation training programs. Another State commenter questioned the 
differences that would exist between the Subtitle C and mixed waste 
disposal facility requirements and their rationale, in the context 
of the position.
    a. Response: In the ``Regulatory Position'' text in Section C, 
the waste that could be

[[Page 13195]]

transferred to the Subtitle C disposal facility was described as 
incident-related material, and was not referred to as low-level 
radioactive waste. In developing the proposed position, this was not 
a decision based on legal considerations, but the terminology 
selected to best characterize the waste, in a technical position 
whose principal purpose was to demonstrate, through a conservative 
assessment, the minimal radiological significance of the proposed 
disposal option. It was recognized, however, that State laws and 
permit conditions would need to be satisfied, and that numerous 
approvals may be required, including those of appropriate State 
regulatory bodies and the disposal facility operator.
    Among other provisions, implementation of the disposal option 
proposed in the position: (1) Involves a licensee's request and 
regulatory approvals on a case-by-case basis pursuant to 10 CFR 
20.2002; (2) includes notification and disposal-site operator-
approval provisions; and (3) includes accounting of the single and 
total incident-related material received at a Subtitle C disposal 
site. As a result, the position does not allow a licensee to dispose 
of the incident-related material as if it were not radioactive, a 
concept that applies only to disposal of certain wastes defined in 
NRC regulations at 10 CFR 20.2005(a). Instead, if the provisions of 
the position are followed, including the specific provision for 
disposal at a Subtitle C facility, the position provides a basis for 
disposing of incident-related material at a site other than one 
specifically licensed for disposal of low-level radioactive waste.
    Furthermore, although not taking a position on what LLW disposal 
requirements could be reasonably applied to the disposal of this 
incident-related material at a Subtitle C hazardous waste disposal 
facility, the staff did specifically address groundwater and 
intruder considerations. Groundwater and intruder assessments were 
provided to allow others to judge the significance of these 
scenarios and the need for additional regulatory provisions 
(including radiation protection training). NRC staff has concluded 
that, with the constraints provided in the position, specific 
regulatory actions (e.g., groundwater monitoring for \137\Cs, 
intruder barriers, institutional controls beyond those applicable to 
Subtitle C disposal facilities) directed at these scenarios are not 
necessary.
    The NRC staff has also concluded that the position's dose 
criterion, and the conservative assessment of allowable \137\Cs 
concentrations, obviates the need for radiation protection training 
for the Subtitle C facility workers. In this regard, the staff would 
point out that the material defined by the technical position would 
not be considered radioactive, for transportation purposes, under 
the U.S. Department of Transportation's (DOT's) regulations. In 
fact, the concentration criteria in the position are a factor of 
about 20 less than the value used by DOT to define radioactive 
material.
    b. Comment: Two State commenters pointed out that the position 
does not address specific permitting provisions pertaining to dust 
treatment to meet land disposal requirements for the dust's 
hazardous constituents. One of these commenters stated that NRC 
cannot assume sole jurisdiction for [hazardous] waste treatment, if 
such treatment were conducted at the steel company sites.
    b. Response: The commenter is correct. The position calls for 
compliance with RCRA land disposal requirements. In the situations 
being addressed, the NRC staff believes appropriate RCRA authorities 
may approve various options for carrying out the treatment of the 
incident-related material. Therefore, only a general statement of 
compliance was included in the technical position. The staff 
acknowledges and agrees with the comment regarding NRC's 
jurisdiction over hazardous waste treatment, no matter where 
conducted. This would be an issue for the State-permitting agencies 
or EPA to decide. In essence, the presumption in the position is 
that the Subtitle C disposal facility would be disposing of waste 
that had been treated under applicable RCRA requirements.

4. Related Safety, Health, and Environmental Concerns

    a. Comment: An environmental group, a State commenter, and the 
member of the public suggested that the best approach to solve the 
problem is a better accounting of the sources causing the incidents, 
and more rigorous regulation appears warranted. The suggestion was 
made that worker exposure at the foundries should be a principal NRC 
concern. As indicated in the discussion in comment 1.a., the steel 
industry commenters also strongly requested NRC action to improve 
accountability.
    a. Response: The Commission, in its directions to the staff on 
October 18, 1994, approved several concurrent courses of action. One 
of these has led to the development of the proposed position, while 
another has led to an Agreement State-NRC Working Group that is 
developing recommendations to address the accountability issue. The 
Working Group has held several meetings and a workshop, and 
recommendations were sent to the Commission in late 1996. The NRC 
staff is in the process of evaluating the NRC/Agreement States 
Working Group's recommendations for increased control over, and 
accountability for, devices containing radioactive material. Once 
the NRC staff completes its evaluation, it will submit an action 
plan to the Commission outlining measures to improve control over, 
and accountability for, devices. Thus, the Commission agrees with 
the commenter's worker safety and ``front-end'' concerns but, 
recognizing that incident-related material currently exists, and 
future incidents may not be prevented with 100 percent confidence, 
believes the ``back end'' of the problem also requires Commission 
action.
    b. Comment: A State and an environmental group commenter viewed 
the policy, in its granting of a ``regulatory exemption'' for the 
incident-related waste, as counterproductive to the desire to 
improve detection capabilities at the steel facilities. Three 
industry commenters, one who responded directly to the State view, 
pointed out that the steel company facilities have installed state-
of-the-art radiation-detection capabilities at considerable expense, 
not to meet any regulatory requirement, but to reduce the likelihood 
of experiencing the consequences of inadvertent-melting events that 
result in significant shutdown, cleanup, and disposal costs, as well 
as the possibility of incident-related exposures to plant personnel. 
Furthermore, these detection systems have been coupled with 
comprehensive scrap inspection programs.
    b. Response: Although the policy provisions may require NRC or 
appropriate Agreement States to not require licensing of the 
Subtitle C facility for the radioactive material, the main feature 
of the policy is the NRC determination that the incident-related 
material can be transferred, under existing regulations (10 CFR 
20.2001 and 20.2002), from a licensed to an unlicensed entity. The 
position not only provides a conservative NRC assessment of the 
radiological impacts of the disposal alternative, but also evaluates 
certain hypothetical situations to provide a frame of reference for 
the calculated impacts. Contrary to the connotation, ``regulatory 
exemption,'' used by the commenters, NRC staff does not consider the 
proposed position to be an exemption action, but an assessment that 
could allow case-by-case decisions on incident-related material 
disposals under current regulations (also see response 1.c.).
    The staff also believes that this policy has no impact on a 
steel company's selection of ``source'' detection capabilities. The 
costs associated with shutdown (downtime) and cleanup alone can 
exceed millions of dollars, far in excess of the costs of effective 
detection systems and programs.
    c. Comment: The environmental group commenters and the mixed-
waste disposal facility operator suggested that the position could 
lead the steel companies to continue operations after a melting for 
the purpose of generating additional contaminated dust in sufficient 
volume to meet the position's concentration criteria. A State 
commenter stated that this issue should be addressed. The mixed-
waste disposal facility operator postulated other abuses (e.g., the 
mixing of other regulated waste with [incident-related] material) 
and asked whether prevention measures were being proposed.
    c. Response: The staff believes that the cost disincentives 
alone are sufficient to consider the former suggestion unreasonable. 
For example, the dilution necessary at one of the facilities with 
this material, such that all the contaminated material would comply 
with the position's criteria, would be about a factor of 5. The 
costs of disposing of this increased volume at a Subtitle C 
facility, even with an optimistic estimate of disposal costs, could 
reach millions of dollars. The staff would note that its development 
of this position has been enmeshed with cost-effective 
considerations because of the real or imagined excessive 
differential costs of the disposal alternatives. Furthermore, based 
on the operation of the steel facilities' emission control systems, 
with their dust-collection systems, the staff can not conceive of a 
scenario that would allow real time comprehension of the extent of 
the contamination or total quantity of \137\Cs involved in an 
incident.

[[Page 13196]]

    With regard to the question of protective measures, the staff 
believes the NRC, Agreement State(s), permitting agencies, or the 
Subtitle C disposal facility operator could, if warranted, require 
or strongly recommend testing requirements to address any concerns 
on disposal of unauthorized radioactive material. The NRC staff 
believes that a licensee's measurement and sampling program, as 
approved by NRC or the Agreement State, will be sufficient to 
preclude unauthorized radioactive material disposals.
    d. Comment: An environmental group commenter stated that the 
concentration criteria in the position appear to be inconsistent and 
less strict than criteria imposed by EPA on mill tailings at 40 CFR 
Part 192. The mixed-waste disposal facility commenter questioned the 
position's comparisons with environmental <SUP>137Cs concentrations. 
The member of the public claimed the proposal would exempt 10 times 
the amount of material that would have been exempted under the BRC 
policy.
    d. Response: The staff presumes that reference is being made to 
the 5 and 15
pCi/g or 135 and 405 Bq/g remedial action criterion for radium=226 
(\226\ Ra) in soil. These are criteria that would apply to soil that 
could be released for unrestricted use. The concentrations in the 
position are those for material that would be disposed of at a 
hazardous waste disposal facility. Because radium is about 2.5 times 
more hazardous from a direct exposure standpoint than <SUP>137Cs, 
the position's bounding <SUP>137Cs values for Subtitle C facility 
disposal are only about 3 to 4 times a value that would be found 
acceptable for unrestricted release. In fact, the typical incident-
related material at under 20 pCi/g (540 Bq/g) would be within the 
criteria range cited and applicable to unrestricted release 
situations. Note also that the position contains a total-quantity 
criterion which is not a part of the 40 CFR Part 192 regulations.
    The comparison referred to by the mixed-waste disposal facility 
commenter was between ``much of the mixed waste'' that contains 
concentrations below 20 pCi/g (540 Bq/g). This concentration was 
being compared with actual environmentally measured concentrations 
of 11 and
12 pCi/g (6300 Bq/g) and statistically-predicted concentrations (95 
percent value of distribution) up to 19 pCi/g (513 Bq/g). The 
reference in footnote 13 of the final technical position is the 
source of these values.
    The staff was not certain about the intended context of the 
comment from the member of the public, but has presumed it is 
related to other issues addressed in the response to this comment, 
comment 1.a., 4.e., or 4.f..
    e. Comment: The mixed-waste disposal facility operator, among 
others, suggested that the position, if adopted, may have adverse 
health, safety, and environmental consequences. One issue involved 
the disposition of higher-activity material that would not be 
covered by the position's criteria. The commenter cites an example 
where the <SUP>137Cs concentration, if averaged over all the 
incident-related material, could be 551 pCi/g (14,900 Bq/g)--[below 
the acceptance criteria at the mixed waste facility]. If the 
material with concentrations below the position's values is disposed 
of under the position's provisions, the commenter asks what would be 
the disposition of the higher concentration material and, if it 
remains onsite, would this violate NRC's intent in promulgating the 
position.
    In a somewhat related comment, a State questioned whether 
material delisted from hazardous material regulations, and meeting 
the concentration values in the position, could be disposed of at a 
Subtitle D facility.
    e. Response: For incident-related material remaining after 
``position-allowed-'' and economically feasible blending of 
contaminated material, the staff is aware of only one disposition 
option at this time (see 61 FR 1616, column 2). That option would 
involve treatment and delisting of the material under hazardous 
material regulations, and disposal of the material as LLW. In two 
situations where incident-related material existed or currently 
exists at steel facilities, about 90 percent of the activity was 
contained in a few percent of the material volume. Given that, in 
many cases, it may not be feasible to blend the <SUP>137Cs in this 
small volume to concentrations acceptable at either the mixed waste 
or the Subtitle C facility (under the provisions of the position), 
treatment and delisting of this small volume may not be onerous. In 
any event, the staff does not believe the uncertainty or current 
feasibility of addressing a small percentage of the problem affects 
the merits of the position, especially as it relates to the mixed-
waste or Subtitle C disposal alternatives.
    In response to the State query, the position does not justify 
disposal at a Subtitle D sanitary waste landfill because the 
radiological assessment was based on a Subtitle C facility disposal. 
Any such disposal, if justified, would have to address the 
differences, if any, between facilities and their operations.
    f. Comment: A series of comments from the mixed-waste disposal 
facility operator questioned NRC's appreciation of the potential 
effects of exposure to low levels of radiation. On the other hand, 
most other commenters either considered the regulatory basis for the 
position of 1 mrem (10 microsievert (<greek-m>Sv)) per year (yr) to 
be reasonable or very conservative. Among several comments, one 
commenter suggested a modest increase in the position's dose basis 
from 1 mrem (10 <greek-m>Sv)/yr to 4 mrem
(40 <greek-m>Sv)/yr, corresponding to the value in EPA drinking 
water standards.
    f. Response: For a number of years, the Commission has used the 
linear no-threshold hypothesis as providing a reasonable and prudent 
basis to assess the radiological risk associated with its actions. 
In essence, this hypothesis involves an extrapolation of the 
statistically significant health effects that can be attributed to 
high-level, short-duration exposures (e.g., the Japanese atomic bomb 
survivors) to levels of exposure at or below what the earth's 
population receives from background sources (e.g., cosmic radiation 
and exposure to radiation emanating from naturally occurring 
materials). Notwithstanding the scientific controversy regarding the 
reality of these hypothetical risks, the Commission's radiation 
protection standards are consistent with standards, recommended by 
international and national advisory bodies, that reflect this 
hypothesis.
    In the case of the technical position, a dose rate of 1 mrem/yr 
(one-hundredth of the public dose limit and about one three-
hundredth of the average exposure rate received year in and year out 
by the population of the United States) was chosen as the regulatory 
basis, because, in the staff's view, it was suitably conservative 
and, from a practical standpoint, provided a disposition solution 
for most of the incident-related material currently existing at 
steel company sites. Footnote 5 of the final technical position 
reflected this view.
    With respect to the mixed waste disposal facility operator's 
comments on the NRC staff's appreciation of the effects of low-level 
radiation and the 1 mrem/yr (10 <greek-m>Sv/yr) regulatory basis, 
the staff believes that the conservatisms in its selection of a dose 
criterion, with appropriate regulatory margins, can be appropriate, 
if the resulting position can lead to resolution of an outstanding 
incident-related waste disposition problem. Although selection of 4 
mrem (40 <greek-m>Sv)/yr could be justified, staff's view is that 
selecting a drinking water standard for this position, which staff 
believes does not present a drinking water issue, would create more 
concern and confusion than the value selected, and its associated 
basis.

5. Technical Considerations

    a. Comment: A State commenter suggested that the position should 
specify acceptable methods for averaging the waste within a 
container.
    a. Response: The staff recognizes that the incident-related 
material in a particular container may not be homogeneous in terms 
of <SUP>137Cs concentration. However, because the principal 
radiological hazard being addressed is related to direct exposure, 
complying with the concentration values, as determined on a 
container average basis, is acceptable. The specifics of the 
characterization program directed at defining treated-material 
(<SUP>137Cs concentrations) would be defined when approving the 
licensee's request for transfer of the incident-related material. 
The characteristics of the treated material, the decision to pursue 
packaged or unpackaged disposal, the statistical confidence desired, 
the regulatory margins provided in the position, and the views of 
the approving parties would need to be considered. The response to 
comment 7.c. could also be applicable in determining a 
characterization program.
    b. Comment: The mixed waste facility operator noted that if one 
considered exposure to a plane source of 60 <greek-m>rem (0.6 
<greek-m>Sv) per hour for 8 hours per day for over 4 weeks, the 
result would be a total exposure exceeding EPA's maximum allowable 
dose. An industry association commenter noted that the dose rate 
limit applied to shipments of radioactive material is a factor 500 
times higher than the value applied in the position to packaged 
disposal.

[[Page 13197]]

    b. Response: The staff does not believe this calculation is 
pertinent. Although the staff is not certain what maximum allowable 
dose is being referred to, the critical point in the calculation is 
that it presumes continuous exposure at 1 meter (<difference>3 feet) 
to a plane of material that is all at the maximum concentration 
criterion. As a point of reference, exposure to ``normal'' dust 
could be calculated to cause an exposure that would be a factor of 
65 or lower, or presuming the possibility of greater exposure 
periods associated with the greater volumes of material, equivalent 
exposure would be reached over a period of about 5 years. The need 
to consider the applicable exposure scenario on which a regulatory 
position is based is brought out by the industry association 
commenter. To make this point, the staff would note that under 
similar assumptions, DOT's allowable exposure rate of 10 mrem (0.1 
mSv) per hour at 1 meter (<difference>3 feet) could be translated 
into a dose estimate of 1.6 rem (16 mSv).
    c. Comment: An industrial association commenter suggested that 
the 1-curie (Ci) or 3.7 x 10 \4\ MBecquerel (MBq) total activity 
limit be modified to a per disposal cell basis (i.e., if the cell 
were larger than 100,000 cubic meters (3.5 x 10 \6\ ft \3\)), on the 
grounds that the proposed constraint may be too limiting if one 
facility would accept the incident-related material from more than a 
single event.
    c. Response: Although this change could be justified, it has not 
been accepted for the following reasons: (1) The procedural 
difficulties for the NRC or Agreement State to require a particular 
disposal constraint at an unlicensed facility, and (2) the belief 
that individual incident disposals under the position's provisions 
are, in most cases, unlikely to approach the quantity constraint 
(one-tenth is expected to be more typical).
    d. Comment: An industrial association commenter suggested that 
the area/shape factors used by NRC were overly restrictive by a 
factor of 2. d.
    d. Response: NRC became aware of area/shape factor differences 
between different codes. Staff has checked its calculations and does 
not believe its estimates are in error.
    e. Comment: A State commenter questioned whether a discrepancy 
existed in NRC's source term assumption, in that dividing a 1-Ci 
(3.7 x 10 \4\ MBq) source over 2000 tons (1814 metric tons) of 
contaminated material would result in an average concentration of 
551 pCi/g (1.49 x 10 \4\ Bq/g).
    e. Response: The commenter's calculation is correct. However, in 
actual events, a significant fraction of the activity is generally 
contained in a small volume of incident-related material at high 
concentrations. As discussed in the response to comment 4.e., the 
disposition of this material will likely require treatment of its 
hazardous properties, so that the material can be delisted and 
disposed of at a licensed LLRW disposal facility. Although the 
position's provisions do allow blending of contaminated material, 
NRC staff recognized that providing the required reduction in 
average concentration to meet the position's concentration criteria 
would likely not be practical in all cases. Staff believed this was 
reflected in the proposed position (e.g., see ``Introduction'' (61 
FR 1609, column 3) and ``Discussion'' (61 FR 1610, columns 2 and 
3)). This reality is why the activity that could be disposed of at 
the Subtitle C facility, for the specific events that have taken 
place to date, is unlikely to exceed 100 mCi (3.7 x 10 \3\ MBq).
    f. Comment: A State commenter raised several questions about the 
groundwater modeling and the input parameters.
    f. Response: The commenter noted that these comments applied to 
an earlier version of the position; however, a few still have 
relevance to the proposed version. In the context of this position, 
the staff was faced with the task of bounding a specific potential 
radiological impact, that staff believed was relegated to a status 
of insignificance by the position's defined concentration and 
quantity criteria. Nevertheless, the approach taken in the position 
was to perform simple bounding analyses and comparisons, so as to 
provide a perspective on the specific hazard. For example, in 
staff's view, a very conservative dose estimate was provided under 
the hypothesis that an individual could and would drink trench 
leachate. Contrary to the commenter's apparent view, staff considers 
the very conservatively calculated 7-mrem (70-<greek-m>Sv) dose from 
directly drinking trench leachate, with a bounding concentration of 
radioactive material, to be a prima facie rationale for claiming 
that EPA's drinking water standards would be met with significant 
margin, not only at the ``tap,'' but at any point in the 
groundwater.

6. Other Issues

    Comment: A State commenter suggested that the position should 
state whether NRC [would] allow import or export of incident-related 
material for disposal.
    Response: The position did not address the import/export issue. 
To the extent that the position's assumptions remain valid, the 
technical basis could be applied to export. However, any imports or 
exports could involve decisions by responsible parties, beyond NRC, 
including non-U.S. regulatory authorities. To the extent that 
appropriate U.S. regulatory authorities agree, and determine that 
they can legally support NRC's views that the treated incident-
related material is not LLRW, the material could be considered for 
disposal under the provisions of the position, giving consideration 
to its hazardous properties, if applicable. The staff does not 
believe this issue needs to be addressed in the context of the 
position itself.

7. Clarifications

    a. Comment: A State commenter stated that the licensee 
transferring the treated incident-related material should notify the 
Agreement State Program or, in the case where an Agreement State 
Program does not exist, the appropriate solid or hazardous waste 
regulatory authority.
    a. Response: The position's provisions are intended to ensure 
such notifications. In the case of Agreement States, their approval 
of the transfer is called for in the position's provisions, as is 
written notification from the licensee at least 30 days before any 
actual transfer. The position also calls for disposal facility 
operator notification and acceptance, in writing. Thus, there are 
two avenues through which the solid or hazardous waste regulatory 
authorities would likely be apprised of actions to implement the 
position. In non-Agreement States, NRC would be the initial, but 
possibly not the only, radiological approving authority. In these 
cases, State-permitting authorities may seek the advice and approval 
of their respective State radiological or public health 
organizations. NRC would work with these authorities and others to 
determine if implementing the position's disposition alternative is 
reasonable and prudent, and legally acceptable.
    b. Comment: In the comments from one State, there appeared to be 
some confusion on what entity would track the total quantity 
constraint (i.e., 1 Ci or 3.7 x 10 \4\ MBq).
    b. Response: Under the position's provisions, the total quantity 
constraint would be tracked by NRC or the appropriate Agreement 
State, although others could also track this inventory value.
    c. Comment: A State commenter queried who would confirm that the 
position's concentration constraints were being met. An 
environmental group commenter suggested that accurate 
characterization presents a considerable challenge.
    c. Response: In the staff's view, NRC or the appropriate 
Agreement State would have a significant incentive to provide some 
independent verification of the concentration criteria. However, the 
specifics of this verification would be addressed when approving the 
licensee's request to make the transfer of incident-related material 
under the provisions of the position. Other parties, including the 
Subtitle C facility operator and the permitting agency, whose 
approvals are required, could also dictate a specific confirmation 
process. On this point, the staff would note the inclusion of 
regulatory margins in the position that, staff believes, should be 
considered in developing a reasonable confirmation program.
    d. Comment: An industry association commenter requested 
clarification regarding the shipment of pretreated incident-related 
material to offsite licensed treatment facilities.
    d. Response: Under the provisions of the position, NRC would 
have no objection to incident-related material being transferred 
offsite for permitted treatment by an NRC or Agreement State 
licensed entity. The position only addresses the transfer of 
incident-related material that has been properly treated, under a 
Commission or Agreement State license, to a Subtitle C disposal 
facility.
    e. Comment: An industry group commenter suggested that the 
position should provide allowance for licensed service contractors 
to be brought in to supervise implementation operations. It was 
further suggested that treatment should not be a prior condition to 
transport.
    e. Response: The position, and NRC regulations, allow the 
possibility of service contractors operating under the contracting 
entity's license. Treatment is only required before transport to an 
unlicensed Subtitle C disposal facility. See the response to comment 
7.d. above.
    f. Comment: An industrial association commenter questioned the 
accuracy of the

[[Page 13198]]

dose rates associated with the 55-gallon drum.
    f. Response: The publication of the figures in the Federal 
Register caused some blurring that has caused the commenter to 
misread the indicated dose rate. Comparisons with the scale on the 
ordinate indicate that the commenter's figure is high by a factor of 
10.

    Dated at Rockville, Maryland, this 13th day of March, 1997.

    For the U.S. Nuclear Regulatory Commission
John W. N. Hickey,
Chief, Low-Level Waste and Decommissioning Projects Branch, Division of 
Waste Management, Office of Nuclear Material Safety and Safeguards.
[FR Doc. 97-6884 Filed 3-18-97; 8:45 am]
BILLING CODE 7590-01-P