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

NUCLEAR REGULATORY COMMISSION

Disposition of Cesium-137 Contaminated Emission Control Dust and

Other Incident-Related Material; Proposed Staff Technical Position

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice: Proposed Staff Technical Position.

-----------------------------------------------------------------------

[[Page 1609]]

SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing guidance,

in the form of a Technical Position, that may be used in case-by-case

requests by appropriate licensees to dispose of a specific mixed waste.

Mixed waste is a waste that is not only radioactive, but also

classified as hazardous under the Resource Conservation and Recovery

Act (RCRA). The specific mixed waste is emission control dust from

electric arc furnaces and foundries that has been contaminated with

cesium-137 (Cs-137). The contamination results from the inadvertent

melting of a Cs-137 source, that: (1) has been improperly disposed of

by an NRC or Agreement State licensee; (2) has been commingled with the

steel scrap supply; (3) has not been detected as it progresses to the

steel producing process; and (4) is volatilized in production process

and thereby can and has contaminated large volumes of emission control

dust and the emission control systems at steel producing facilities.

The proposed position, which has been coordinated with the U.S.

Environmental Protection Agency (EPA), provides the possibility of a

public health-protective, environmentally sound, and cost-effective

alternative for the disposal of much of this mixed waste that contains

Cs-137, in concentrations similar to values that frequently occur in

the environment. The position provides the bases that, with the

approval of appropriate regulatory authorities (e.g., State-permitting

agencies) and others (e.g., disposal site operators), and with public

input, could be used to allow disposal of treated (stabilized) waste at

Subtitle C, RCRA-permitted, hazardous waste disposal facilities. NRC

believes that disposal, under the provisions of the position or other

acceptable alternatives, is preferable to allowing this mixed waste to

remain indefinitely at steel company sites.

The proposed position has been developed through a very ``open''

process in which working draft documents have been routinely shared

with EPA, and also placed in NRC's Public Document Room (Subject File:

204.1.23) to allow interested party access. In keeping with this

process, NRC, rather than noticing the availability of the proposed

position, is publishing the entire position for public comment.

DATES: Submit comments by March 22, 1996. Comments received after this

date will be considered if it is practical to do so, but the Commission

is able to assure consideration only for comments received on or before

this date.

ADDRESSES: Send comments to Chief, Rules Review and Directives Branch,

U.S. Nuclear Regulatory Commission, Washington, DC 20555. A final

position will be issued following NRC staff review of the comments

received.

FOR FURTHER INFORMATION CONTACT:

W.R. Lahs, Division of Waste Management, Office of Nuclear Material

Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington,

DC 20555, Telephone (301) 415-6756.

SUPPLEMENTARY INFORMATION:

Disposition of Cesium-137 Contaminated Emission Control Dust and Other

Incident-Related Materials; Proposed Branch Technical Position

A. Introduction

Emission control (baghouse) dust and other incident-related

materials (e.g., cleanup materials or recycle process streams)

contaminated with cesium-137 (Cs-137) 1 are currently being stored

as mixed radioactive and hazardous waste at several steel company sites

across the country. At any single site, this material typically

contains a total Cs-137 quantity ranging downward from a little more

than one curie (37 gigabecquerels (GBq)) of activity, distributed

within several hundred to a few thousand tons of iron/zinc-rich dust,

as well as within much smaller quantities of cleanup or dust-recycle,

process stream materials.2

\1\ The byproduct material Cs-137 does not include the Cs-137,

from global fallout, that exists in the environment from the testing

of nuclear explosive devices (See Footnote 3).

\2\ The term, ``incident-related material,'' is frequently used

in this position to refer to the total spectrum of Cs-137-

contaminated materials resulting from an inadvertent melting event.

Because of its widespread use in radioactive devices and its

volatility when subjected to steel melting temperatures, the

position is directed solely at incident-related materials involving

this nuclide.

---------------------------------------------------------------------------

The radioactivity is not evenly distributed among these materials.

Typically, a small fraction (e.g., one-tenth) of the material contains

most (e.g., 95 percent) of the radioactivity. Most of the material

contains a small quantity of radioactivity at low concentrations and

makes up most of the mixed-waste volume. This material is generally

classified as hazardous waste under RCRA because it contains lead,

cadmium, and chromium that are common to the recycle metal supply. The

Cs-137 contamination of this hazardous waste, on the other hand,

results from a series of three principal events: (1) the loss of

control of a radioactive source by an NRC or Agreement State licensee;

(2) the inclusion of the source within the recycle metal scrap supply

used by the steel producers; and (3) the inability to screen out the

radioactive source as it progresses along the typical scrap collection-

to-melt pathway (e.g., including radiation detectors used at most

furnaces and foundries). Consequently, irrespective of the quantity or

concentration of the radioactivity, all the material is subject to

joint regulation as mixed waste under RCRA and the Atomic Energy Act of

1954, as amended, or the equivalent law of an Agreement State.

The disposal options for these materials, specifically the large

volumes of material with the lower concentrations of Cs-137, have been

limited because of their ``mixed-waste'' classification and the costs

associated with the disposition of large volumes of mixed or

radioactive waste. Long-term solutions addressing the control and

accountability of licensed radioactive sources are being considered by

NRC and its Agreement States. Solutions addressing the disposition of

mixed wastes are being considered by various Federal and State

regulatory authorities and the U.S. Department of Energy. Nevertheless,

the Commission believes that, pending decisions on improved licensee

accountability and the ultimate disposition of mixed waste, appropriate

disposal of the existing incident-related, mixed-waste material is

preferable to indefinite onsite storage.

As a result, this technical position defines the bases that the NRC

staff would generally find acceptable for: (1) authorizing a licensee,

possessing Cs-137 contaminated emission control dust and other

incident-related materials (e.g., the steel company or its service

contractor), to transfer Cs-137 contaminated material, below levels

specified in this position, to a Subtitle C, RCRA-permitted hazardous

waste disposal facility; and (2) exempting the possession and disposal

of these incident-related materials (e.g., by the RCRA-permitted

disposal facility) from NRC or Agreement State licensing requirements.

Because of its radioactivity (i.e., Cs-137 concentration levels), some

of the incident-related material may not be suitable for disposal at a

Subtitle C, RCRA-permitted disposal facility. This material may be

disposed of either: (1) at a licensed low-level radioactive waste

disposal facility following ``delisting'' (e.g., after appropriate

treatment of its hazardous constituents) or (2) at a mixed waste

disposal facility, if applicable acceptance criteria are met.

The regulatory basis for the first action is found at 10 CFR

20.2001(a)(1). This paragraph authorizes a licensee to

[[Page 1610]]

dispose of licensed material as provided in the regulations in 10 CFR

Parts 30, 40, 60, 61, 70, or 72. Paragraph 30.41(b) states the

conditions under which licensees are allowed to transfer byproduct

material. Paragraph 30.41(b)(7) of Part 30 specifically provides that

licensees may transfer byproduct material if authorized, by the

Commission, in writing.

The regulatory basis for the second action is found at Sec. 30.11

(``Specific exemptions''), which states that the Commission may, on its

own initiative, grant exemptions (from the requirements of the

regulations in 10 CFR Parts 30 through 36, and 39) as it determines are

authorized by law and will not endanger life or property and are

otherwise in the public interest. It should be noted that additional

acceptance requirements, beyond those covered in this NRC position for

disposal of Cs-137-contaminated hazardous waste at a Subtitle C RCRA-

permitted disposal facility, may be established by: (1) an Agreement

State; (2) the permit conditions or policies of the RCRA-permitted

disposal facility; (3) the regulatory requirements of the RCRA disposal

facility's permitting agency; or (4) other authorized parties,

including State and local governments. These requirements may be more

stringent than those covered in the guidance described in this

technical position. The licensed entity transferring the Cs-137-

contaminated incident-related materials should consult with these

parties, and obtain all necessary approvals, before making the

transfers defined in this technical position. Nothing in this position

shall be or is intended to be construed as a waiver of any RCRA permit

condition or term, of any State or local statute or regulation, or of

any Federal RCRA regulation.

B. Discussion

Over the past decade, there has been an increasing number of

instances in which radioactive material has been inadvertently

commingled with scrap metal that subsequently has entered the steel-

recycle production process. If this radioactive material is not removed

before the melting process, it could contaminate the finished metal

product, associated dust-recycle process streams, equipment

(principally air effluent treatment systems), and the dust generated

during the process. Some of the contaminant radioactivity is a result

of naturally occurring radionuclides that deposit in oil and gas

transmission piping. Other radioactivity may be associated with

radioactive sources that are contained in industrial or medical

devices. In this latter case, the commingling of the radioactive source

with metal destined for recycling can occur if the regulatorily

required accountability of these sources fails and a radioactive source

is included within the metal scrap supply used by the steel producers.

In cases where the radionuclide is naturally occurring, or is already

present in the environment as a result of global fallout, the

inadvertent melting of a radioactive source could increase the

contaminant concentration above that caused by these background

environmental levels.3

\3\ In a letter to William Guerry, Jr. from NRC's Executive

Director for Operations, James M. Taylor, dated May 25, 1993, NRC

made a preliminary determination that Cs-137 levels in baghouse dust

can reasonably be attributed to fallout from past nuclear weapons

testing, if concentrations are less than about 2 pCi/g (0.074 Bq/g).

---------------------------------------------------------------------------

Although many of the steel producers have installed equipment to

detect incoming radioactivity, this equipment cannot provide absolute

protection because of the shielding of radioactive emissions that may

be provided by uncontaminated scrap metal or the shielded ``pig'' that

contains the radioactive source. Of special concern, because of the

nature and magnitude of the involved radioactivity, are NRC- or

Agreement State-licensed sources containing Cs-137.

When Cs-137 sources are inadvertently melted with a load of scrap

metal, a significant amount of the Cs-137 activity contaminates the

metal-rich dust that is collected in the highly efficient emission

control systems that steel mills have installed to comply with air

pollution regulations. Because of toxic constituents--specifically

lead, cadmium, and chromium--electric arc furnace (EAF) and foundry

emission control dust are subject to regulation under RCRA. If this

dust becomes contaminated with Cs-137, the resulting material would be

classified as a mixed waste. Emission control dust, generated

immediately after the melting of a Cs-137 source with the scrap metal,

can contain cesium concentrations in the range of hundreds or thousands

of picocuries per gram (pCi/g) or a few to a few tens of becquerels

(Bq) per gram of dust, above typical levels in dust caused by Cs-137 in

the environment (e.g., 2 pCi/g or 0.074 Bq/g). Several thousand cubic

feet (several tens of cubic meters) of dust could be contaminated at

these levels. Dust generated days or weeks after a melt of a source

(containing hundreds of millicuries or a few curies of Cs-137) will

contain reduced concentrations, typically less than 100 pCi/g (3.7 Bq/

g).

Even after extensive decontamination and remediation activities,

newly generated dust may still contain concentrations greater than 2

pCi/g (0.074 Bq/g) background levels, but generally less than 10 pCi/g

(0.37 Bq/g). When the melting of a source is not immediately detected,

materials related to downstream processes have also been contaminated

with relatively low concentrations of Cs-137 (e.g., 10 pCi/g (0.37 Bq/

g)). In addition, materials used during decontamination may also be

contaminated with dust containing Cs-137 concentrations at similar

levels above background.

As the result of past inadvertent meltings of Cs-137 sources, a

number of steel producers possess a total of over 10,000 tons of

incident-related materials, most of which contains Cs-137

concentrations of less than 100 pCi/g (3.7 Bq/g). This material is

typically being stored onsite because of the lack of disposal options

that are considered cost effective by the steel companies.4 It is

the disposition of material at these concentration levels that is the

subject of this technical position.

\4\ In April 1995, Envirocare of Utah, Inc., an operator of a

mixed-waste disposal site, received authorization from the State of

Utah and initiated operations to treat and dispose of Cs-137-

contaminated incident-related (mixed waste) materials at

concentrations not exceeding 560 pCi/g (20.7 Bq/g).

---------------------------------------------------------------------------

C. Regulatory Position

General

Because of the ``incident-related'' origin of the Cs-137

contaminated materials, the Commission has approved a course of action

that includes: (1) exploration of approaches to improve licensee

control and accountability to reduce the likelihood of sealed sources

entering the scrap metal supply; (2) cooperation with the steel

manufacturers and other appropriate organizations to identify the

magnitude and character of the problem (with particular emphasis on

improving the capability to detect sealed sources before their

inadvertent melting); and (3) development of interim guidelines for the

disposal of Cs-137 contaminated dust and other incident-related

materials (the subject of this technical position).

Specific

Bases for Allowing Transfer and Possession of Cs-137 Contaminated

Incident-Related Material. The bases for allowing transfer and

possession of Cs-137 contaminated emission control dust and other

incident-related materials, under the provisions of existing

regulations, are as follows: (1) Any

[[Page 1611]]

person at a Subtitle C, RCRA-permitted disposal facility involved with

the receipt, movement, storage, or disposal of contaminated materials

should not receive an exposure greater than 1 millirem (mrem) or 10

micro-sievert (Sv) per year (i.e., one-hundredth of the dose

limit for individual members of the public as defined at 10 CFR

20.1301(a)(1)), above natural background levels; 5 (2) members of

the general public in the vicinity of storage or disposal facilities

should not receive exposures and no individual member of the public

should be likely to receive a dose greater than 1 mrem (10 Sv)

per year above background as a result of any and all transfers and

disposals of contaminated materials; (3) handling or processing of the

contaminated materials, undertaken as a result of its radioactivity,

should not compromise the effectiveness of permitted hazardous waste

disposal operations; (4) treatment of contaminated materials should be

accomplished by persons operating under a licensee's radiation

protection program; and (5) transportation of contaminated materials

should be performed by hazardous material employees, as defined in U.S.

Department of Transportation (DOT) regulations (49 CFR Part 172,

Subpart H).

\5\ The use of 1 mrem (10 Sv) has no significance or

precedential value as a health and safety goal. It was selected only

for the purpose of analysis of the levels at which the referenced

materials could be partitioned to allow the bulk of the material to

be transferred to unlicensed persons. It does not represent an NRC

position on the generic acceptability of dose levels. Such levels

are established only by rule.

---------------------------------------------------------------------------

Definition of Contaminated Materials and Initial Incident Response.

A melting event generally necessitates extensive decontamination and

remediation operations at the EAF or foundry (e.g., replacing

refractory bricks and duct work). Subsequent operations include the

proper interim handling and management (e.g., accumulation and

containment) of emission control dust and other incident-related

contaminated materials. Based on a review of several recent incidents,

the dust may contain Cs-137 concentrations up to hundreds or thousands

of pCi/g (a few to a few tens of Bq/g), whereas the other generally

limited-volume, incident-related materials typically contain lower

concentrations. As a result, the initial cleanup and collection/

treatment/ packaging of the contaminated emission control dust and

other materials at the EAF or foundry should be performed by an NRC or

Agreement State licensee operating under an approved radiation

protection program. The licensee would also be responsible for

compliance with other non-radiological regulatory requirements (e.g.,

those of the Occupational Safety and Health Administration and RCRA

Treatment Permitting requirements).

Provisions for Disposal at a Subtitle C, RCRA-Permitted, Disposal

Facility. Once the decontamination/remediation and collection/

treatment/packaging activities have been completed, one of two paths

may be followed for the disposal of the incident-related materials,

dependent on Cs-137 concentration levels and whether the final land

disposal operation involves the burial of packaged or unpackaged

materials.

1. Packaged Disposal of Treated Waste. On this disposal path,

contaminated materials would be treated through stabilization to comply

with all EPA and/or State waste treatment requirements for land

disposal of regulated hazardous waste. The treatment operations would

be undertaken by either (i) The owner/operator of the EAF or foundry

(licensed by NRC or appropriate Agreement State to possess, treat, and

transfer Cs-137 contaminated incident-related materials); or (ii) an

NRC-or Agreement State-licensed service contractor. Based on the

radiological impact assessment provided in the appendix, the licensee

could be authorized to transfer the treated incident-related materials

to a Subtitle C, RCRA-permitted, disposal facility, provided that all

the following conditions are met:

(a) The Cs-137-contaminated emission control dust and other

incident-related materials are the result of an inadvertent melting of

a sealed source or device;

(b) The emission control dust and other incident-related materials

have been treated (stabilized) to meet requirements for land disposal

of RCRA-regulated waste, and have been stored (if applicable) and

transferred in compliance with a radiation protection program as

specified at 10 CFR 20.1101;

(c) The total Cs-137 activity, contained in emission control dust

and other incident-related materials to be transferred to a Subtitle C,

RCRA-permitted, disposal facility, has been specifically approved by

NRC or the appropriate Agreement State(s) and does not exceed the total

activity associated with the inadvertent melting incident. Moreover,

NRC or the appropriate Agreement State should maintain a public record

of the total incident-related Cs-137 activity, received by the facility

over its operating life, to ensure that this total-disposed Cs-137

activity does not exceed 1 curie (37 GBq); 6

\6\ The 1 curie (37 GBq) value represents a reasonable bounding

activity, associated with several incidents, that could be

transferred to an RCRA-permitted facility under the provisions of

this position. It also represents a quantity that would be less than

the activity disposed of over the operating life of the RCRA-

permitted facility, if the facility routinely disposed of non-

incident-related emission control dust containing background

concentrations of Cs-137.

---------------------------------------------------------------------------

(d) The RCRA disposal facility operator has been notified in

writing of the impending transfer of the incident-related materials and

has agreed in writing to receive and dispose of the packaged materials;

(e) The licensee providing the radiation protection program

required in paragraph (b), notifies, in writing, the Commission or

Agreement State(s) in which the transferor and transferee are located,

of the impending transfer, at least 30 days before the transfer;

(f) The treated (stabilized) material has been packaged for

transportation and disposal in non-bulk steel packagings as defined in

DOT regulations at 49 CFR 173.213. (Note that this is a condition

established under this technical position and is not a DOT requirement.

Under DOT regulations, material with concentrations of less than 2

thousand picocuries per gram (74 Bq/g) is not considered radioactive);

(g) In any package, the emission control dust and other incident-

related materials, that have been treated (stabilized) and packaged as

defined in (b) and (f) above, contain pretreatment average

concentrations of Cs-137 that did not exceed 130 pCi/g (4.8 Bq/g) of

material; 7 and

\7\ The 130 pCi/g (4.8 Bq/g) value is the concentration, based

on the analysis in the appendix and including a regulatory margin of

1.5, that would result in a calculated potential exposure less than

1 mrem (10 Sv). The disposal of incident-related materials

in packaged form allows compliance with this position to be

demonstrated through measurement of Cs-137 concentrations, as well

as direct radiation levels external to the package. Notwithstanding

the redundant approaches to ensure compliance with the exposure

criterion, the regulatory margin of 1.5 has been included in

determining the acceptable measurables defined in the position.

---------------------------------------------------------------------------

(h) The dose rate at 3.28 feet (1 meter) from the surface of any

package containing treated (stabilized) waste does not exceed 20

rem per hour or 0.20 Sv per hour, above

background.8

\8\ At this exposure rate, for the exposure period as defined in

the appendix, total exposure would not exceed 1 mrem (10

Sv) with a regulatory margin of 1.5.

---------------------------------------------------------------------------

Note that, in defining the pretreatment Cs-137 concentration value

stated in paragraph (1)(g), a factor of 1.5 has been included as a

regulatory margin. This factor adds further

[[Page 1612]]

assurance to the certainty in protection provided by the licensee's (1)

Sampling of Cs-137 concentrations in contaminated materials, (2)

measurements of dose rate external to the disposal (and transportation)

packagings, and (3) other assumptions included in the radiological

impacts assessment.

2. Disposal of Unpackaged (i.e., Bulk) Treated Waste. On this

disposal path, contaminated materials would also be treated through

stabilization to comply with all EPA and State waste treatment

requirements for land disposal of RCRA-regulated hazardous waste. The

treatment operations would be undertaken by either (i) The owner/

operator of the EAF or foundry (licensed to possess, treat, and

transfer Cs-137-contaminated incident-related materials), or (ii) a

licensed service contractor. Based on the radiological impact

assessment provided in the appendix, the licensee could be authorized

to transfer the treated (stabilized) incident-related materials to a

Subtitle C, RCRA-permitted, disposal facility, provided that all the

following conditions are met. (Note that conditions (a) through (e) are

identical to those applicable to packaged disposal of treated waste):

(a) The Cs-137 contaminated emission control dust and other

incident-related materials are the result of an inadvertent melting of

a sealed source or device;

(b) The emission control dust and other incident-related materials

have been treated (stabilized) to meet requirements for land disposal

of RCRA-regulated waste, and have been stored (if applicable), and

transferred in compliance with a radiation protection program as

specified at 10 CFR 20.1101;

(c) The total Cs-137 activity, contained in emission control dust

and other incident-related materials to be transferred to a Subtitle C,

RCRA-permitted, disposal facility, has been specifically approved by

NRC or the appropriate Agreement State(s) and does not exceed the total

activity associated with the inadvertent melting incident. Moreover,

NRC or the appropriate Agreement State should maintain a public record

of the total incident-related Cs-137 activity, received by the facility

over its operating life, to ensure that this total disposed Cs-137

activity does not exceed 1 curie (37 GBq); 9

\9\ See footnote 6.

---------------------------------------------------------------------------

(d) The RCRA disposal facility operator has been notified in

writing of the impending transfer of the incident-related materials and

has agreed in writing to receive and dispose of these materials;

(e) The licensee providing the radiation protection program

required in paragraph (b) notifies, in writing, the Commission or

Agreement State(s) in which the transferor and transferee are located,

of the impending transfer, at least 30 days before the transfer; and

(f) The emission control dust and other incident-related materials,

that have been treated (stabilized) as defined in (b) above, contain

pretreatment average concentrations of Cs-137 that did not exceed 100

pCi/g (3.7 Bq/g) of material.10

\10\ The 100 pCi/g (3.7 Bq/g) value is the concentration, based

on the analysis in the appendix and including a regulatory margin of

2, that would result in a calculated potential exposure of less than

1 mrem (10 Sv). The disposal of incident-related material

in unpackaged (bulk) form dictates that compliance with this

position would be demonstrated through measurement of Cs-137

concentrations. Without the redundant approach to ensure compliance

with the exposure criterion inherent with the packaged-disposal

approach (see Footnote 7), the regulatory margin, included in

determining the acceptable measurables defined in the position, has

been increased to 2.0.

---------------------------------------------------------------------------

Note that, in defining the pretreatment Cs-137 concentration value

in paragraph (2)(f), a factor of 2 has been included as a regulatory

margin. The factor adds further assurance to the certainty of

protection provided by the licensee's (1) sampling of Cs-137

concentrations in contaminated materials; and (2) other assumptions

included in the radiological impacts assessment.

Treatment, Storage, and Transfer of Emission Control Dust or Other

Incident-Related Materials with Cs-137 Concentrations Indistinguishable

from Background Levels (i.e., 2 pCi/g (0.074 Bq/g) or Less). The EAF or

foundry licensed to possess and transfer Cs-137 contaminated emission

control dust or a licensed service contractor is authorized to transfer

emission control dust and other incident-related materials as if they

were not radioactive, provided that the Cs-137 concentration within the

emission control dust and other incident-related materials is 2 pCi/g

(0.074 Bq/g) of material or less.

Aggregation of Cs-137 Contaminated Emission Control Dust and Other

Incident-Related Materials. Aggregation of Cs-137 contaminated emission

control dust and other incident-related material, before stabilization

treatment, is acceptable if performed in compliance with a radiation

protection program, as described at 10 CFR 20.1101, and provided that:

(1) Aggregation involves the same characteristic or listed

hazardous waste and the wastes must be amenable to and undergo the same

appropriate treatment for land-disposal restricted waste;

(2) Aggregation does not increase the overall total volume nor the

radioactivity of the incident-related mixed waste; and

(3) Materials, when aggregated, are subjected to a sampling

protocol that demonstrates compliance with Cs-137 concentration

criteria on a package-average 11 basis.

\11\ The term package, as used here, refers to packages used by

the licensee to transfer the material to the disposal facility,

irrespective of whether this package is also the disposal container.

---------------------------------------------------------------------------

Determination of Cs-137 Concentrations and Radiation Measurements.

Cs-137 concentrations may be determined by the licensee by direct or

indirect (e.g., external radiation) measurements, through an NRC- or

Agreement State-approved sampling program. The program should be

sufficient to ensure that Cs-137 contamination in stabilized treated

emission control dust and in other incident-related materials, on a

package-average basis, is consistent with the concentration criteria in

this technical position. The sampling program should provide assurance

that the quantity of Cs-137 in any package (see footnote 11) does not

exceed the product of the applicable concentration criterion times the

net weight of contaminated material in a package.

Appendix--Assessment of Radiological Impact of Disposal of Cs-137

Contaminated Emission Control Dust and Other Incident-related Materials

at a Subtitle C RCRA-Permitted Disposal Facility

Background

In the normal process of producing recycled steel, scrap steel is

subjected to a melting process. In this process, most impurities in the

scrap steel are removed and generally contained within process-

generated slag or off-gas. Typically, the off-gas carries dust,

containing iron and zinc, together with certain heavy metals, through

an emission control system to a ``baghouse,'' where the dust is

captured in ``bag-type'' filters. Hazardous constituents within the

dust, principally lead, cadmium, and chromium, cause the dust to be

designated by EPA as a hazardous waste, under RCRA, often as the listed

waste K061.

Typically, when the scrap consists largely of junk automobiles, the

dust contains a high percentage (greater than 20 percent) of zinc,

which can be a valuable recovery product. Moreover, the zinc recovery

process produces slag and other byproducts that have recycle potential.

If economic (e.g., low zinc content) or process considerations

[[Page 1613]]

preclude these recycle options, the dust may be treated and disposed of

in a hazardous waste disposal facility. Treatment standards for the

various hazardous constituents of the dust have been specified by EPA

in 40 CFR 268.40. Solidification is the treatment process typically

used to meet these standards.

Because the recycling of steel involves the addition of natural

materials (primarily lime and ferromanganese), very low levels of

radioactivity, ubiquitous in the environment, are involved in the

production process. One of these radionuclides is Cs-137 which now

occurs in the environment as a result of global fallout from past

weapons-testing programs.

Cs-137 has a 30-year half-life (i.e., a quantity of this

radionuclide and its associated radioactivity will decrease by half

every 30 years). The decay of Cs-137 and its very short-lived daughter

produces emissions of beta particles and gamma rays.

The principal hazard from the beta particles can only be realized

when it enters the human body. The principal hazard from the gamma rays

is as an external source of penetrating radiation similar to the type

of exposure received from an X-ray. Because of its volatility in the

very high-temperature (typically 3000 degrees fahrenheit) steel-making

process, Cs-137 is volatilized and transported in the furnace off-gas

and, as it condenses, becomes a constituent of the emission control

(baghouse) dust. Normal background Cs-137 concentrations in dust have

been measured at picocurie per gram levels (0.024 to 1.23 pCi/g)

12 or thousandths of a becquerel per gram (Bq/g). This

concentration is consistent with the general range of background levels

measured in soils within the United States whereas concentrations of 10

pCi/g (0.37 Bq/g) are relatively common in drainage areas.13 As a

result of this information, NRC has determined that Cs-137

concentrations in emission control dust below 2 pCi/g (0.074 Bq/g) can

be attributed to fallout from past weapons testing.14

\12\ A picocurie is one-trillionth of a curie and represents a

decay rate of one disintegration every 27 seconds or 1/27 of a

becquerel.

\13\ Letter to William Lahs, Nuclear Regulatory Commission,

from Andrew Wallo III, Department of Energy, dated May 20, 1993.

\14\ Letter from James M. Taylor, Nuclear Regulatory

Commission, to William Guerry, Jr., Collier, Shannon, Rill, and

Scott, dated May 25, 1993.

---------------------------------------------------------------------------

Statement of Problem

The inadvertent melting of a licensed Cs-137 sealed source with

scrap steel at an EAF or foundry typically results in the contamination

of the steel producer's emission control system and the generation of

potentially large quantities (e.g., of the order of 1000 tons) of Cs-

137 contaminated emission control dust. Facility cleanup operations

will produce an additional quantity of contaminated material and,

depending on the effectiveness of cleanup operations, further

generation of contaminated dust or cleanup-related materials can occur.

Furthermore, if the occurrence of the melting event is not immediately

detected, contamination can unknowingly be carried forward with the

dust into zinc-recovery process streams. In one case, for example, this

has led to Cs-137 contamination of the zinc-rich, splash condenser

dross residue, referred to as SCDR material. In the incidents to date,

total quantities of these contaminated materials have not exceeded 2000

tons per event. The Cs-137 concentration in all these materials can

vary, but in typical past events, much of the material is contaminated

at levels ranging from 2 pCi/g (0.074 Bq/g) to a few hundred pCi/g

(most below approximately 100 pCi/g or 3.7 Bq/g). Smaller volumes

(typically less than 5 percent of the total volume) have included

concentrations at nanocurie/gram levels (thousands of pCi/g or a few

tens of Bq/g).

The intent of this analysis is to characterize the potential

radiological impacts associated with the alternative options for

disposal of Cs-137 contaminated emission control dust and other

incident-related materials at a Subtitle C, RCRA-permitted facility.

Because these RCRA hazardous wastes must be treated to comply with the

requirements for land disposal of restricted waste, the potential

radiological impacts associated with treatment processes required

consideration. To protect against these radiological impacts, the

position includes the provision that treatment of Cs-137 contaminated

emission control dust and other incident-related materials be performed

by an NRC or Agreement State licensee. The licensee would operate under

an approved radiation protection program, as well as any required RCRA

treatment permit. Such controls are necessary because of the wide range

of contaminated materials and their physical forms, together with the

variability in EPA-approved treatment processes. Under this decision,

the Subtitle C, RCRA-permitted disposal facility would be receiving the

emission control dust and other incident-related materials after their

treatment to stabilize the RCRA-hazardous constituents (specifically,

lead, cadmium, and chromium) in a non-dispersible,15 solid (e.g.,

cement-type) form. As a result, the potential radiological hazard from

the ``treated'' material during disposal operations is associated with

its characteristic as an external source of radiation.

\15\ In the context used, the term ``non-dispersible'' means

that any radiological impacts from resuspended material are

inconsequential in comparison to the impacts from direct external

exposures resulting from the emission of gamma radiation in the Cs-

137 decay process.

---------------------------------------------------------------------------

After disposal, Cs-137 could only become a hazard through water

pathways if a sufficient quantity and concentration of Cs-137 were to:

(1) become available, (2) be leached from its solid form, (3) be

released from the disposal facility, and (4) enter a drinking water

supply. No significant radiological hazard would be expected to result

from inadvertent intrusion into the disposed waste after facility

closure. Notwithstanding the hazard to the intruder from the hazardous

waste constituents, constraints placed on the total Cs-137 activity and

concentration, and the waste form, can ensure that radiological

exposures would not exceed those that would be received from residing

over commonly-measured background Cs-137 concentrations in the United

States (see discussion under ``Intruder Considerations'').

The following analyses will therefore be directed at an evaluation

of the potential direct, water pathway, and intruder hazards and will

provide a perspective on their significance.

Direct Exposure

After the inadvertent melting of a Cs-137 sealed source at an EAF

or foundry, the relatively volatile Cs-137 will leave the furnace as an

offgas and be commingled with the normal emission control dust. As a

result, concentrations of Cs-137 contained in this dust (and other

materials associated with furnace cleanup operations or subsequent dust

recycle process streams) will increase. Thus, the rate of radiological

exposure from this material will be similar in type, but different in

magnitude, than that received from the typical background levels of Cs-

137. Any change in magnitude of the exposures to workers at the

disposal facility from this contaminated material when compared to the

exposure received from typical emission control dust would depend on:

(1) differences in Cs-137 concentrations; (2) variations in the

physical/chemical properties of the materials disposed of; and (3)

changes in worker time-integrated interactions with contaminated

materials.

[[Page 1614]]

The three key variables above are particularly important in the

development of this technical position. Of significance to all three

variables, the approach defined in the position calls for treatment

(stabilization) of incident-related materials (to comply with

requirements for land disposal of restricted waste) to take place

``under license,'' at the location where the material was generated, or

at the site of a service contractor permitted for stabilization

treatment of the material. Complying with the ``Treatment Standards for

Hazardous Wastes,'' defined at 40 CFR 268.40, will result in a solid

waste form from which exposure rates will be smaller than those

originating from the hazardous waste form (e.g., dust) before

treatment. More importantly, treatment of the contaminated materials,

under license, will obviate the need to specifically address potential

radiological exposures at unlicensed, RCRA-permitted, treatment

facilities. Thus, under the approach of this technical position, any

minimal exposure to workers who have not been trained in radiation

safety would be limited to disposal operations.

Furthermore, because the origin of the Cs-137 contaminated

materials is the result of a melting incident, upper bound values can

be established for the volume, weight, radioactive material

concentration, and total activity of the contaminated material, on an

incident basis. The base case analysis in this appendix presumes that

the contaminated material involves a volume of 40,000 cubic feet (1132

cubic meters), a weight of 2000 tons, and a total activity content of

less than a 1 curie (37 gigabecquerels (GBq)) of Cs-137. These values

are generally consistent with the particulars from the incidents that

have occurred to date.

Within these constraints, the starting point in the direct exposure

calculation is to estimate the radiation dose rate at a distance of

3.28 feet (1 meter) from the surface of a semi-infinite volume (i.e.,

infinite in areal extent and depth from the point of exposure) of

solidified contaminated material.16 The calculations assume that

the initial Cs-137 contamination in all untreated dust is 100 pCi/g

(3.7 Bq/g). Direct exposure results scale linearly for other

concentration levels, if the waste configuration is unchanged.

\16\ This assessment is generally consistent with the approach

employed in ``Risk Assessment of Options for Disposition of EAF Dust

Following a Meltdown Incident of a Radioactive Cesium Source in

Scrap Steel,'' SELA-9301, Stanley E. Logan, April 1993.

---------------------------------------------------------------------------

Stabilization treatment,17 conducted under a licensed

radiation protection program, is achieved by mixing moist dust with

additives (e.g., liquid reagent to adjust oxidation potential and

portland cement/fly ash).18 These additives (typically presumed to

add 30 parts by weight to 100 parts of dust or contaminated material)

would result in a solidified product that would contain Cs-137

concentrations at about 77 percent of initial concentrations (e.g., 77

pCi/g (2.84 Bq/g)). Because of allowable variations in the

solidification processes (e.g., from the production of granularized

aggregate to solidified monoliths), the bulk density of the solidified

material can range from about 1.4 to 2.5 g/cm3. A representative

dose conversion factor 19 under these conditions (calculated at a

density of 1.5 g/cm3) would typically be less than 49 microrem/

hour (rem/hr) or 0.49 microsieverts/hour (Sv/hr), at

a distance of 3.28 feet (1 meter) from the surface of a hypothetical

semi-infinite volume of the solidified material.20

\17\ In the context of this position, stabilized treatment does

not include either onsite or offsite high-temperature metals

recycling processes.

\18\ This treatment may include the addition of special

stabilization reagents, such as clays, or involve other RCRA-

approved stabilization technologies, that reduce the leachability of

Cs-137, although the radiological impacts analysis indicates that

such processes are not necessary to protect public health and

safety, and the environment.

\19\ A dose conversion factor represents a value that allows a

radionuclide contamination level to be converted to an estimated

exposure rate.

\20\ The dose rates in this appendix have been calculated

through use of the Microshield computer program, Grove Engineering,

Inc., version 4.2, 1995. The value of 49 rem/hour

represents 0.77 of the 62.9 value shown on Figure 1.

---------------------------------------------------------------------------

Because the quantities of treated dust and other incident-related

materials are not semi-infinite in volume, the actual dose rate/

distance relationships from finite volumes of contaminated materials

will be less. The reduction can be calculated for various volumetric

sources through the use of shape factors. Shape factors have been

calculated for several configurations that are likely to occur during

operations from the time the contaminated treated material is received

at the RCRA-permitted disposal facility through its disposal. The shape

factors can be determined from Figures 1 through 6 for various

distances between a specific source configuration and an exposed

individual. Typically, at a distance of 3.28 feet (1 meter), these

factors range from about 0.03 to 0.5 (Figures 1 through 5), and have

been calculated without accounting for the limited shielding provided

by any packaging. As the distance from the contaminated materials

increases to 9.84 feet (3 meters), the shape factors for these similar

geometries become smaller, ranging from about 0.004 to 0.2. The largest

likely dose rate potentially experienced by an individual involved in

the disposal process, measured at 3.28 feet (1 meter), would be from

the sides of large containers or shipments of contaminated materials,

and would be expected to range from about 10 to less than 14

rem/hour (0.14 Sv/hr) above background (typically 8

to 12 rem/hr (0.08 to 0.12 Sv/hr).21 From an

open trench (Figure 4), filled with contaminated materials, the

calculated dose rate would also be somewhat less than 13 rem/

hr (0.13 Sv/hr) measured directly over the trench at a 3.28

feet (1 meter) distance. Again, these values represent 0.77 of the

respective values indicated on the figures because of solidification

additives. Figures 6 and 7, respectively, show the variation in dose

rate with the width of the trench and depth of the waste. Figure 8 is

provided to show the change in dose rate versus the distance offset

from the side of the trailer-type container considered in Figure 3.

\21\ The two-thirds loading of the 30-cubic yard box is related

to the typical maximum payload weight that can be transported by

truck without an overweight permit. If the boxes referred to in

Figures 1 and 2 were full, the dose rate would increase by less than

a factor of 1.5. Similarly, if the assumed additive weight percent

(i.e., 30 percent) is varied over a reasonable range from 20 to 40

percent, the resulting dose rate would change in an inversely

proportional manner.

---------------------------------------------------------------------------

A typical disposal rate at a trench within an RCRA-permitted

facility would typically exceed 500 tons per shift.22 Assuming

this disposal rate of 500 tons per shift applies to the disposal of

treated, Cs-137-contaminated, incident-related material (approximately

20 to 25 truckloads in 8 hours), it would require approximately 4 times

this period of time to dispose of 2000 tons. (Note that the rate of

arriving material would likely be dictated by transportation

arrangements, so that the 32 hours required to dispose of the

contaminated material could be spread over several days or weeks.)

Facility workers, therefore, would, on average, only be exposed to

finite volumes of contaminated material for a maximum period of 32

worker-hours. Applying the highest likely dose rate (approximately 13

rem/hr (0.13 Sv/hr) from the side of a trailer

containing the contaminated materials), and presuming exposure at a

3.28-ft (1-meter) distance for the entire 32-hour period, a worker

would receive

[[Page 1615]]

a dose of less than 0.5 mrem (5 Sv) above background.

\22\ Note that if treatment at an RCRA-permitted facility were

required, the limiting operational handling rate for the treated

materials may be limited to 100 to 200 tons per shift.

---------------------------------------------------------------------------

Qualitatively descriptive time and motion data gathered from three

RCRA-permitted disposal facilities indicate that the above-calculated

dose is conservative for two principal reasons: (1) the workers having

the most significant exposure to materials, from receipt to disposal,

are effectively at greater distances than 3.28 feet (1 meter); and (2)

their exposure is over time periods significantly less than the assumed

receipt through disposal time period of 32 hours. As a result, actual

exposures are expected to be significantly less than 0.5 mrem (5

Sv).

This conservative estimate of potential exposure is based on the

aforementioned time-distance assumptions and is expected to bound

reasonable interactions of disposal facility workers with the treated

(stabilized) incident-related materials. For example, incident-related

material could be stored at the disposal site or samples of the treated

material could be subjected to sampling activities. In the first case,

if a 90-day storage period is presumed, the average exposure distance

over the entire period needed to ensure a dose less than the position's

exposure criteria would be on the order of 10 to 20 meters (see Figures

1 through 3 which illustrate the decrease in dose rate as a function of

distance from the source). In the second case, the typical activity in

a 100 gram sample would be no greater than about 10-2 Ci

(370 Bq). The dose rate from such a sample would be less than 0.1

rem/hr (0.001 Sv/hr) at a distance of 1 foot (0.3

meters).

To place the significance of this calculation into perspective, an

estimate can be made of worker exposure from the presumed handling,

treatment, and disposal of normal emission control dust (i.e., dust

that has not been contaminated with Cs-137 from a melted source). This

dust would contain background levels of Cs-137 (approximately 1 pCi/g

(0.037 Bq/g)). Therefore, a worker interacting with this material at an

effective distance of 3.28 feet (1 meter) over about 300 8-hour shifts

(a little more than a working year) would receive a total maximum

exposure about 0.5 mrem (5 Sv). The magnitude of this exposure

is in the same range as the exposure calculated for the disposal of the

contaminated materials from a single melting event. Moreover, the

potential exposure from the ``melting event'' was estimated under the

extremely conservative assumption that all materials were contaminated

at levels of 100 pCi/g (3.7 Bq/g).

The imposition of a 1-curie (37 GBq) criterion on the total

incident-related activity that could be disposed of at any one Subtitle

C, RCRA facility (see following discussion on water-pathway

considerations) should further ensure that worker exposures from Cs-137

contaminated emission control dust and other incident-related materials

will not exceed 1 mrem/year (10Sv/year) integrated over the

lifetime of the facility.

Water-Pathway Considerations

The proposed approach to manage Cs-137 contaminated emission

control dust and other incident-related materials presumes licensee

treatment of these materials to comply with requirements for land

disposal of restricted waste. Thus, the hazardous radiological and

chemical constituents of these materials will be incorporated into a

stable, solid (e.g., cement-type) form, similar to that required for

routine RCRA-permitted disposal of emission control dust. As a result,

the possibility of Cs-137 presenting a hazard through a water pathway

requires consideration of: (1) the quantity of Cs-137 available; (2)

the degree to which the Cs-137 could be leached from its waste matrix;

and (3) the extent that any leached Cs-137 could migrate into a water

supply.

The disposal of Cs-137 in treated emission control dust and other

incident-related materials would be constrained by this policy to a

total activity of 1 curie (37 GBq). In the previous reference-basis

analysis, an effective concentration, in the treated waste, of 77 pCi/g

(2.84 Bq/g) was evaluated--the originally assumed contaminated material

concentration reduced by 30 percent as a result of the added mass

associated with treatment. Both the quantity and position-defined

concentration values place bounds on any potential water pathway

hazard. In the actual wastes that are subject to potential disposal

under the provisions of this position, the concentration of Cs-137

averaged over all the treated waste would typically be significantly

less than the defined concentration criteria.

Furthermore, because the Cs-137 is contained in a solid matrix and

buried within a facility in which the amount of water infiltration is

minimized, any Cs-137 removal from its final disposal location would be

limited while these conditions remain in effect. The chemistry of any

water interacting with the solidified, Cs-137-contaminated waste would

also be expected to limit the leaching process (e.g., avoidance of

acidic environments), because of the controlled nature of the Subtitle

C, RCRA-permitted disposal site and the types and nature (e.g., no

liquids) of the wastes accepted for disposal. Any water that leached

Cs-137 from the waste would normally be collected in a leachate

collection system at volumetric concentrations expected to be far less

than that existing in the treated waste. The chemistry of the fill

materials used at the disposal site could also provide a sorbing medium

if any Cs-137 leached from the solidified waste. Finally, the location

of Subtitle C, RCRA-permitted disposal sites is such that the source of

any water supply would typically be some distance from the disposal

site.

These chemistry and distance factors are also likely to be major

factors in delaying the arrival of Cs-137 at a receptor well because of

retardation effects. This retardation, in terms of its effect on the

time required, under a worst-case scenario, for the Cs-137 to reach a

water supply, is such that significant radioactive decay of the Cs-137

inventory is likely (the radioactive half-life of Cs-137 is 30 years)

before this pathway could potentially pose a hazard.

Although qualitative in nature, and based on considerations that

can vary among Subtitle C, RCRA-permitted disposal sites, the

discussion has focused on the factors that are likely to prevent any

significant water-pathway hazard. The following, more quantitative

assessment, is provided to conservatively bound any water-pathway

hazard that could potentially occur under extremely unlikely

conditions, and provides the technical basis for NRC's position.

The leachability of Cs-137 from any solid waste form that allows

compliance with the land disposal restrictions for the waste's non-

radiological hazardous constituents is likely to be extremely limited

after initial waste placement. After the end of operations and a post-

closure care period of 30 years, a worst-case scenario presumes that

processes take place to degrade the site so that infiltrating water

from the surface passes unimpeded through the contaminated waste. In

predicting the dissolution of Cs-137 under these conditions, a critical

process is the partitioning of the Cs-137 that takes place between the

waste, soil, and infiltrating water. Conservatively assuming that the

partitioning from the solid waste form is similar to that from the

interstitial backfill soil to water, an estimate can be made of the

amount of Cs-137 that can leach into the infiltrating water.

The most important parameter in estimating this transfer, as well

as the subsequent movement of the Cs-137 in groundwater, is the

distribution

[[Page 1616]]

coefficient, Kd. This parameter expresses the ratio at equilibrium

of Cs-137 sorbed onto a given weight of soil particles to the amount

remaining in a given volume of water. The higher the value of the

distribution coefficient, the greater the concentration of Cs-137

remaining in the soil. The Kd value can be affected by factors

such as soil texture, pH, competing cation effects, soil porewater

concentration, and soil organic matter content.23 For the non-

acidic, sand/clay/soil environments presumed to represent the RCRA-

permitted disposal facilities, a Kd value of 270 milliliter (ml)/g

was selected from the Footnote 23 reference as being appropriate for

the subsequent bounding, conservative analysis.

\23\ ``Default Soil Solid/Liquid Partition Coefficients,

Kds, for Four Major Soil Types: A Compendium,'' M. Sheppard and

D. Thibault, Health Physics, Vol. 59, No. 4, October, 1990, pp. 471-

482.

---------------------------------------------------------------------------

To model the potential groundwater impacts, the RESRAD 24 code

was used. For the representative case, the bounding 40,000 cubic feet

(ft\3\) or 1132 cubic meters (m\3\) of treated material were presumed

to be disposed of in a volume measuring 100-ft (30.4-m) length x 20-ft

(6.09-m) width x 20-ft (6.09-m) depth. All this material was assumed

to contain a Cs-137 concentration of 77 pCi/g (2.84 Bq/g).

Notwithstanding the actual layouts of Subtitle C, RCRA-permitted

facilities, a well was presumed to be located and centered at the

downgradient edge of this specific volume of waste. To maximize the

hazard as calculated by the RESRAD model, the hydraulic gradient was

considered to be parallel to the length of the disposed volume.

Infiltration representative of a humid site was presumed and a minimal

unsaturated zone thickness of 3.28 ft (1 m) was assumed to separate the

contaminated zone from the saturated zone. The value assigned to

Kd in the unsaturated zone was 270 ml/g. Assessments beyond this

representative case evaluation are subsequently discussed.

\24\ RESRAD, Version 5.0, Argonne National Laboratory,

September 1993.

---------------------------------------------------------------------------

The results from this bounding analysis indicate that drinking

water dose rate would be insignificant (e.g., far less than a microrem

(10-2 Sv) per year). This result is not surprising

because the retardation provided, even in the 3.28-ft (1-m) deep

unsaturated zone and the saturated zone, are sufficient to preclude

drinking water doses for almost 700 years. During this period, the

activity of Cs-137 would decay (i.e., be reduced by radioactive decay)

by a factor of about 10 million.

Note that, although it is considered an unrealistic scenario, the

drinking of the leachate directly from the disposal trench after a

period of 30 years would only result in a calculated exposure of about

7 mrem/year (70 Sv/year).25

\25\ This dose estimate is based on comparing leachate

concentrations with the water effluent concentration in 10 CFR Part

20, Appendix B.

---------------------------------------------------------------------------

To consider the effects of a range of parameters, including other

Kd values, on the results of this bounding analysis, the following

analyses are presented. Based on the typical existing volumes and Cs-

137 concentrations of incident-related materials, the imposition of a

constraint on Cs-137 concentration effectively bounds the total

activity that could be disposed of at a Subtitle C, RCRA-permitted

facility from a single steel company site to a few tens of

millicuries.26 Material at higher concentrations would require

disposal at either a mixed-waste disposal facility or a licensed low-

level radioactive waste disposal site. Thus, for the potential

disposals at the Subtitle C, RCRA-permitted site to approach the 1

curie (37 GBq) incident-related material constraint in this position,

disposals of materials from several incidents would have to occur. The

total volume of material, in this case, would still represent only a

small fraction of a RCRA-permitted facility's disposal capacity.

Repeating the RESRAD analysis discussed above under these assumptions,

but respectively considering lower Kd values in the contaminated,

unsaturated, and saturated zones, would still result in drinking water

doses of less than 1 mrem (10 Sv) per year unless the Kd

values in all zones approach single digit values. Even in these cases

(e.g., Kd equal to 2.7), separation of the hypothesized well

location from the disposed material by about 100 meters (328 ft) would

reduce dose rates below 1 mrem (10 Sv) per year because of the

decay of Cs-137 brought about by the increased retardation times.

\26\ For example, the total activity contained in 2000 tons of

material, contaminated at a level of 77 pCi/g, would be about 0.14

curies (5.2 GBq). It would be unlikely that all the material from a

particular incident would be at the maximum concentration defined in

the technical position.

---------------------------------------------------------------------------

The concentration constraints in this position, coupled with the

limited number of inadvertent melting situations to which this position

could be applicable, and the case-by-case NRC or Agreement State

approval of the proposed material transfers are believed to provide a

sufficient basis to ensure protection of public health and safety, and

the environment from water-pathway considerations. Nevertheless, to

provide further protection, should a single Subtitle C, RCRA-permitted

disposal facility accept incident-related material from more than one

incident, the position includes a total Cs-137 incident-related

activity constraint of 1 curie (37 GBq). The magnitude of this

constraint is based on the typical bounding activity associated with an

inadvertent melting of Cs-137 sources that have occurred to date at

EAFs or foundries. In large measure, it has been included to provide

assurance that the position is only directed at the ultimate

disposition of radioactive material that exists in the environment as a

result of specific inadvertent melting incidents. However, it also

provides a constraint on the extent of volumetric contamination as a

function of concentration. The practical effect, as previously alluded

to, is to limit the disposal volumes of incident-related contaminated

materials to a small fraction of total disposal site capacity for

hazardous waste. As a result of this volumetric limit, the constraint

would further ensure that any exposures occurring offsite over the

operating life of the Subtitle C, RCRA-permitted facility would be

equal to or less than 1 mrem/year (10 Sv/year), if integrated

over the facility's operating life.

Again, the activity constraint and the water pathway considerations

can be placed in perspective by evaluating the potential normal

disposal of EAF emission control dust at a Subtitle C, RCRA-permitted

facility. If this dust includes a background Cs-137 concentration of 1

pCi/g (0.037 Bq/g), and the facility can treat 200 tons of dust per

day, the total quantity of Cs-137 disposed of annually would be about

50 mCi (1.85 GBq). Thus, over a facility operating period of about 20

years, the total quantity of Cs-137 disposed of could equal the 1-curie

(37 GBq) incident-related material activity constraint.

Intruder Considerations

In the development of its licensing requirements for land disposal

of radioactive waste in 10 CFR Part 61, NRC considered protection for

individuals who might inadvertently intrude into the disposal site,

occupy the site, and contact the waste. In the context of this

position, this possibility has been considered although the greater

risk to the intruder would likely result from the non-radiological

hazardous constituents at the site.

In the intruder scenarios applied in the development of NRC's low-

level

[[Page 1617]]

waste standards,27 an inadvertent intruder was assumed to dig a 3-

meter (9.9 ft) deep foundation hole for construction of a house. The

top 2 meters (6.6 ft) of the foundation were assumed to be trench cover

material and the bottom 1 meter (3.28 ft) was assumed to be waste.

Based on the details of the scenarios, which included these and other

considerations, the intruder interacted with material whose

concentration had been reduced from the waste concentration by a factor

of 10. Presuming similar scenarios and assuming intrusion occurs

immediately after a post-closure care period of 30 years, the intruder

would be exposed to a Cs-137 concentration of about 4 pCi/g (0.15 Bq/

g); that is, 77 pCi/g (2.84 Bq/g) reduced by the factor of 10 and an

additional factor of 2 to account for radioactive decay). Even for this

worst-case situation in which all the incident-related waste was

presumed to have initial Cs-137 concentrations of 77 pCi/g (2.84 Bq/g),

the projected intruder exposure would range from 0.8 to 3.8 mrem (8 to

38 Sv/year).28 As noted above, the average concentrations

over large volumes of incident-related material would be expected to be

far less than 77 pCi/g (2.84 Bq/g).

\27\ See NUREG-0782, vol. 4, Draft Environmental Impact

Statement on 10 CFR Part 61, ``Licensing Requirements for Land

Disposal of Radioactive Waste,'' September 1981.

\28\ These estimates are based on the concentration to dose

conversion values in NUREG-1500, ``Working Draft Regulatory Guide on

Release Criteria for Decommissioning: NRC Staff's Draft for

Comment,'' August 1994. Appropriate adjustments of the tabulated

information were made to reflect the occupancy and shielding

assumptions made in NUREG-0782 (see Footnote 24).

---------------------------------------------------------------------------

Conclusions

These bounding analyses indicate that some significant volume of

Cs-137-contaminated emission control dust and other incident-related

materials from an inadvertent melting of a sealed source can be

disposed of at a Subtitle C, RCRA-permitted facility with negligible

impacts to public and worker health and safety and the environment.

This method for disposal, if implemented according to the limitations

stipulated in this position, is very unlikely to cause worst-case

exposures that exceed 1 mrem (10 Sv) to any worker at the

disposal facility or to any member of the public in the vicinity of the

facility. The design, operations, and post-closure activities that take

place at Subtitle C, RCRA-permitted facilities will ensure that

radiological impacts from Cs-137 will also be negligible in future

timeframes. Proper disposal of these materials would protect public

health and safety, and the environment to a greater degree than the

alternative of indefinitely storing these materials at a steel company

facility. The calculated public health and safety and environmental

impacts of disposition of specified incident-related materials at a

Subtitle C, RCRA-permitted facility can also be used to determine an

optimum course for disposal, if disposition alternatives exist.

BILLING CODE 7590-01-P

[[Page 1618]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.001

[[Page 1619]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.002

[[Page 1620]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.003

[[Page 1621]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.004

[[Page 1622]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.005

[[Page 1623]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.006

[[Page 1624]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.007

[[Page 1625]]

[GRAPHIC] [TIFF OMMITTED] TN22JA96.008

BILLING CODE 7590-01-P

Dated at Rockville, Maryland, this 11th day of January, 1996.

For the Nuclear Regulatory Commission.

Michael F. Weber,

Chief, Low-Level Waste and Decommissioning Projects Branch, Division of

Waste Management, Office of Nuclear Material Safety and Safeguards.

[FR Doc. 96-703 Filed 1-19-96; 8:45 am]

BILLING CODE 7590-01-O