ML20205H952

Onsite Disposal of Radioactive Waste.Methodology for the Radiological Assessment of Disposal by Subsurface Burial

ML20205H952
Person / Time
Issue date:	02/28/1987
From:	William Kennedy, Neuder S NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
References
NUREG-1101, NUREG-1101-V02, NUREG-1101-V2, NUDOCS 8704010173
Download: ML20205H952 (50)

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NUREG-1101 Vol. 2 Onsite Disposal of Radioactive Waste Msthodology for the Radiological Assessment of Disposal by Subsurface Burial U.S. Nuclear Regulatory Commission office of Nuclear Material San and Safeguards i

s. M. Neuder and W. E. Kennedy, Jr.

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' Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

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The NRC Public Document Room,-1717 H Street, N.W.'

' Washington, DC 20656 l

2. The Superintendent of Documents, U.S. Government Printing' Office, Post Office Box' 37082, Washington,'DC 20013-7082

3. : The National Technical Information Service, Springfield, VA' 22161 h

' Although the listing that follows represents the majority of documents cited in NRC publications, L

it is not intended to be exhaustive.-

[

~ Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection -

and Enforcement. bulletins, circulars, information notices,' inspection and. investigation, notices;.

Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the GPO Sales -

l Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and NucAser Regulatory Commission Issuances.

- Documents available ftom the - National Technical-Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

-j Documents available from public and special technical libraries include all open literature items, such as books, joumal and periodical articles, and transactions. Feoieral Register notices, federal and state legislation,'and congressional reports can usually be obtained from these libraries.

' Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Division of Technical information and Document Control, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555.

I-Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if thsy are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.

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NUREG-1101 Vol. 2 Onsite Disposal of Radioactive Waste Msthodology for the Radiological Assessment of Disposal by Subsurface Burial Manuscript Completed: January 1987 Dits Published: February 1987

s. M. Neuder, U.s. Nuclear Regulatory Commission W. E. Kennedy, Jr., Pacific Northwest Laboratory Division of Waste Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Wcshington, D.C. 20566 s *=%,

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ABSTRACT Volume 1 of this NUREG provides guidance for academic, medical, and industrial licensees seeking authorization to dispose of small quantities of ra!!oactive material by onsite subsurface disposal.

Licensee requests for such authoriza-tions are made pursuant to Section 20.302 of 10 CFR Part 20 " Standards for Pro-tection Against Radiation." Volume 1 supplements Section 20.302 to assure that appropriate information is provided by the licerssbe and describes disposal methods and techniques acceptable to the NRC s % ff in its evaluation of the application.

In addition, Volume 1 identifies categories of radionuclides defined for subsur-face disposal.

Limiting conditions are described for each category of radionu-clides with respect to total radioactivity, waste packaging, burial frequency and other conditions likely to be acceptable for subsurface disposal.

The categories of radionuclides and associated disposal conditions and criteria are the primary data by which the NRC staff will make an initial evaluation of in-formation provided in the application.

Applications for proposed disposal activities that do not fit any of the disposal categories may need to be eval-uated against more detailed guidance described in this volume (Volume 2).

This volume describes the criteria and technical methodology used by NRC staff to evaluate requests by licensees for approval of onsite disposal by burial in soil.

The technical methodology includes the 0NSITE/MAXIl code for calculating radiological exposure from various pathways, the M0CMOD84 code, and analytical methods for calculating contaminant transport and concentration of radionuclides in flowing groundwater.

Radiological exposure analyses include the following pathways:

(1) exposure to direct gamma from any surface contamination or buried waste, (2) drinking water from a well contaminated by migration of radionuclides, (3) ingesting agricultural products derived from radionuclide-contaminated soil, and (4) inhaling radionuclides resuspended at the burial site.

Licensee proposed disposal activities are evaluated in terms of radiological impact on public health and safety and the environment.

The estimated committed effective dose equivalent resulting from the technical evaluation will usually be the determining factor in the authorization of the proposed disposal.

NUREG-1101, Vol 2 iii

TABLE OF CONTENTS PaSe ABSTRACT.............................................................

iii ACKNOWLEDGEMENTS.....................................................

vii 1

INTRODUCTION....................................................

1-1 2

DISCUSSION.....................................................

2-1 2.1 Disposal Pursuant to 10 CFR 20.302.........................

2-1 2.2 Description of Onsite Disposals............................

2-1 2.3 Generators and Types of Waste..............................

2-2 2.4 NRC Review and Assessment..................................

2-4 2.5 Pathways and Scenarios.....................................

2-5 2.6 Radiological Dose Assessment and Recommendations...........

2-7 3

THEORETICAL CONSIDERATIONS AND COMPUTATIONAL METHODS............

3-1 3.1 The ONSITE/MAXII Code......................................

3-1 3.2 Data Base for the ONSITE/MAXII Computer Program......

3-3 3.3 External Exposure Pathway - The 0NSITE/MAXII...............

3-5 3.4 Ingestion Pathway - The ONSITE/MAXII.......................

3-7 3.5 Inhalation Dose-Conversion Factors - The Task Group Lung Model......................................................

3-9 3.6 Comoarison of ICRP 2 and ICRP 26/30 Dose-Conversion Factors.

3-10 3.7 Groundwater Pathway Modeling...............................

3-11 3.8 Groundwater Pathway - Numerical Model......................

3-12 3.9 Groundwater Pathway - Analytical Mode1......................

3-14 3.10 Onsite Well................................................

3-17 4

DOSE CALCULATIONS AND ADDITIONAL COMPUTATIONAL CONSIDERATIONS...

4-1 4.1 Usage Parameters...........................................

4-1 4.2 Dose From External Radiation...............................

4-3 4.3 Area Correction Factor for External Exposure...............

4-3 4.4 Dose From Ingestion of Food Products.......................

4-3 4.5 Dose From Inhalation.......................................

4-6 4.6 Dose From Drinking Water...................................

4-7 4.7 Dose From an Onsite Well...................................

4-7 4.8 Source Term Considerations.................................

4-7 4.9 Comparison With 10 CFR Part 61 Methods and Results.........

4-10 5

REFERENCES.....................................................

5-1 NUREG-1101, Vol 2 v

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TABLE OF CONTENTS (Continued)

LIST OF FIGURES P,ag 3.1 Conceptual Groundwater Model...................................

3-15 4.1 Area Correction Factors........................................

4-4 LIST OF TABLES 2.1 Principal Radionuclides in Institutional Wastes.................

2-3 2.2 Types of Waste Proposed for Disposal by NRC Licensees Pursuant to 10 CFR 20.302.......................................

2-3 3.1 Computational Methods and Associated Pathways...................

3-1 3.2 Radionuclides Available for ONSITE/MAXII Dose Assessments.......

3-4 3.3 Sensitivity Study Results for the External Exposure 3 for Selected Waste Form Pathway - mrem /hr per Ci/m Densities and Slab Thicknesses..................................

3-7 4.1 Parameters Used for Calculation of Radiation Dose Factors From Consumption of Foods.......................................

4-2 4.2 Comparison of Calculated and 10 CFR Part 61 Low-Level Waste Disposal Concentrations...................................

4-11 i

NUREG-1101, Vol 2 vi

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ACKNOWLEDGEMENT The authors wish to acknowledge the valuable support of Mr. Robert E. Browning, Director, Division of Waste Management, Mr. Richard E. Cunningham, Director, Division of Fuel Cycle and Material Safety, and Mr. Vandy L. Miller, Chief, Material Licensing Branch, Division of Fuel Cycle and Material Safety, during the development and implementation of 10 CFR 30.302 disposal assessment methodology.

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NUREG-1101, Vol 2 vii

1 INTRODUCTION Pursuant to Section 20.302 of 10 CFR Part 20 [ Code of Federal Regulations, Title 10, Part 20], any licensee or applicant for a license may apply to the U.S. Nuclear Regulatory Commission (NRC) for approval of proposed procedures to dispose of licensed or any other radioactive material in a manner not otherwise authorized in the regulations. Section 20.302 of the regulations provides the licensee with an alternative to the more conventional means of disposal of their i

radioactive waste.

For example, the conventional disposal of a large volume of soil contaminated with trace quantities of radioactivity may require that several hundred drums be shipped to a commercial waste disposal site.

In this situation, the licensee may wish to seek NRC authorization under 620.302 to bury the slightly contaminated radioactive soil in a trench located on the licensee's own property. Licensee proposals pursuant to $20.302 are reviewed by the NRC on a case-by-case basis and are assessed primarily in terms of potential radio-logical impacts on public health and safety. These impact assessments will strongly depend on the site-specific and other information provided by the licensee when submitting a request for disposal authorization.

The specific information required from the licensee and a description of disposal methods and techniques likely to be acceptable to the NRC staff in its evaluation of the application is described in Volume 1 of this NUREG [Neuder, 1986).

In addition, Volume 1 identifies three categories of radionuclides defined for sub-surface disposal.

Limiting conditions are described for each category of radio-nuclides with respect to total radioactivity, waste packaging, burial frequency, and other conditions acceptable for subsurface disposal.

The application process is simplified if the requested disposal activity falls within a given disposal category and the associated criteria, as described in Volume 1, can be met by the applicant.

For proposed disposal activities which do not fall within any of the three categories, as for example unpackaged dis-posals or disposal quantities which exceed those specified, a more detailed application may be necessary and technical assessments of the types described in this document should be performed by the licensee.

Radiological impact assessments of licensee-proposed disposal activities include consideration of site-specific exposure pathways and exposure scenarios.

In general, the critical exposure pathways include (1) external radiation exposure directly from contaminated surface soil and from buried waste, (2) internal exposure from resuspension and inhalation of airborne radionuclides, (3) inter-nal exposure from ingestion of food crops and animal products derived from the site, (4) internal exposure from drinking contaminated water from a nearby well, and (5) internal exposure from drinking water from a well constructed within the waste disposal area itself. Dose rates to an individual located on or near the site are calculated from the quantity and concentration of radioactivity in the soil, water, and air medium near the disposal site and the individual's rate of exposure to the radiation from appropriate pathways.

NUREG-1101, Vol 2 1-1

In general, the committed effective dose equivalent will be the primary factor in determining whether the licensee's proposed disposal activities will or will not be authorized. When authorized, burials by the licensee under prescribed conditions and restrictions on surrounding land use will provide reasonable assurance that the potential dose to an individual on or near the site will be within acceptable limits.

Section 20.302 of the regulationt is applicable to all NRC licensees including industrial, medical, academic crd reactor licensees.

For example, a nuclear power reactor licensee may request aui.horization for onsite burial of equipment or materials which has become slightly contaminated during routine reactor opera-tions. Although differences exist between power reactor sites and sites of other types of facilities, radiological impact assessments of reactor waste disposals pursuant to S20,302 proceed along the same lines as those described in this document.

Disposal of radioactive waste by burial on the licensee's property or by other means, pursuant to $20.302, and the methodology for the radiological impact assessments of these types of disposal activities are the subjects of this volume (Volume 2 of this NUREG series).

Volume 3 of this NUREG series [Goode, Neuder, Pennifill, and Ginn, 1986]

describes several conceptual models and solution techniques for estimating potential groundwater contamination due to migration of radionuclides from a burial site.

Volume 4 of this NUREG series [Brenneman and Neuder, in preparation] will pro-vide guidance to reactor licensees for the disposal of very low level radio-active waste pursuant to S20.302.

Volume 4 is in preparation and is not available at this time.

NUREG-1101, Vol 2 1-2

2 DISCUSSION 2.1 Disposal Pursuant to 10 CFR 20.302 As described earlier, licensees may apply for authorization to dispose of small quantities of radioactive materials in a manner not otherwise authorized in the regulations. The licensee may, for example, seek approval for burials of tritium-contaminated, laboratory-type trash on a quarterly basis or may seek approval for a one-time burial of soil slightly contaminated with carbon-14 from radio-labled plant-uptake studies.

In most cases, proposed disposals pursuant to S20.302 are for burial of low-activity waste in soil.

Incineration of radioactive waste at the licensee's facility has been authorized under S20.302 and S20.305 in a few instances.

Guidelines for disposal by incineration are established and will therefore not be considered in this document.

Several reactor as well as non-reactor licen-sees have requested authorization for disposal of contaminated waste by means other than incineration or onsite burials [Brenneman and Neuder, in preparation].

In all cases, however, an NRC determination that the radiological risk to public health and safety is negligible does not relieve the licensee of the responsi-bility for satisfying all other applicable federal, state, and local disposal requirements [Neuder, 1986; Brenneman and Neuder, in preparation].

This document addresses radioactive waste disposal by burial in soil.

The pathways, exposure scenarios and assessment methodology described in the next chapter are therefore applicable to near-surface waste burial considerations for any licensee.

2.2 Description of Onsite Disposals Licensees seeking authorization for onsite waste disposal generally propose to The bury radioactive waste within the first 20 feet beneath the soil surface.

proposed burial pit varies in size, shape, and depth depending on the types of These burial pits waste, waste volume, and the licensee's method of excavation.

are usually filled with waste to within 4 to 6 feet of the original soil surface and the remaining space above the waste is generally backfilled with the excava-ted soil and mounded above the original surface level.

In most instances, the licensee will propose to use the burial site for infre-quent disposals, with consecutive burials usually made in adjacent, noncontig-Frequency of burials varies from licensee to licensee, ranging uous areas.

anywhere from once each month to once a year.

Occasionally, a licensee will request authorization for a one-time release or burial of contaminated material.

Long-term isolation is normally not a consideration in onsite disposal authori-zations.

In general, an authorization for onsite burial would not be issued if it required controls beyond the time the licensee can reasonably be expected to occupy the site [Neuder, 1986].

NUREG-1101, Vol 2 2-1

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Except for proposed disposals at reactor sites, radioactive wastes are usually packaged in disposal drums prior to burial on site.

Surface slumping, depres-sions, or cave-ins (collectively referred to as subsidence) will frequently occur at onsite disposal areas.

Causes of subsidence include infiltration of water into the ground, erosion from surface runoff, void spaces inside and between waste packages, and waste material which degrades in time.

Leaching of radioactive constituents from buried waste will ultimately occur in varying degrees depending on the particular isotopes present in the waste, the amount of water percolatir.g into the disposal unit, the physical condition of the waste package, the physical and chemical forms of the waste, and the con-tact time between the water and the waste itself.

2.3 Generators and Types of Waste Institutional waste refers to radioactive waste generated at universities, col-leges, medical schools, hospitals, testing, and research laboratories.

Several institutional waste streams which have relatively high volume but low levels of radioactivity include absorbed liquid scintillation fluids, solidified or absorbed organic and aqueous liquids, biological waste, and trash.

1 Scintillation fluids are classified as hazardous wastes and would therefore not be acceptable for onsite burials in accordance with Environmental Protection Agency (EPA) regulations [ Code of Federal Regulations, Title 40, Part 261].

Biological wastes include animal carcasses (for the most part), plant and animal tissue, animal bedding, excreta, and culture media.

Carcass disposals generate large volumes since they are usually packed in absorbent materials, lime, and in double containers.

Typical institutional trash consists of paper, plastic, and glass.

The largest volume contributors to institutional wastes are institutions con-ducting biological research with radioactive materials.

Typical volumes vary f rom 50 f t3 to 5000 ft3 The most common radionuclides associated with bio-research wastes are H-3, C-14, P-32, S-35, Ca-45, Cr-51, and I-125 of which H-3 is dominant [ Beck et al., 1977].

A more complete list of radioruclides asso-ciated with institutional wastes, their half-lives and principal radioemissions is given in Table 2.1.

Note that most of these radionuclides have half-lives less than 90 days.

Radionuclide concentrations in institutional type waste generally do not exceed 0.2 pCi/cm3 (approximately 6 mci /ft3) in liquid scintil-lation media and in contaminated trash [ Wild et al., 1981].

The predominant radionuclides found in very low level reactor wastes proposed for $20.302 disposals are isotopes of cesium, cobalt, and manganese.

Typical volumes vary from 1000 ft3 to 100,000 ft3 In almost all cases, concentrations are in the range 0.1 to 100 pCi/gm.

Generators of industrial waste include producers and distributors of radioactive isotopes, manufacturers of materials and devices containing radioactive isotopes, and users of materials, instruments, and devices containing these isotopes.

Certain industrial waste streams generate high volume but relatively low levels of radioactivity, with radionuclide concentration levels generally below a few nanocuries per cubic centimeter [ Wild et al., 1981].

Such wastes are generated by industrial laboratories and radiopharmaceutical companies and have character-istics similar to institutional wastes.

NUREG-1101, Vol 2 2-2

i Table 2.1 Principal radionuclides in institutional wastes Radionuclide Half-life Emission H-3 12.3y p

C-14 5730y p

P-32 14.3d p

S-35 88d p

Ca-45 165d p

Cr-51 27.8d p,y Fe-59 45d p,y Co-60 5.26y p,y Ga-67 78.7h p,y Se-75 120d p,y Rb-86 18.7d p,y Sr-90 28.ly S

Mo-99 66.7h p,y Tc-99" 6.05h p,y In-111 2.82d p,y I-125 60.0d p,y I-131 8.05d p,y Xe-133 5.27d p,y Cs-137 30.0y p,y Yb-169 32d p,y T1-201 73h p,y Table 2.2 lists the types of waste proposed for disposal by NRC licensees.

Concentrations proposed for disposal in these wastes are typically of the order of.01 pCi/cm3 for non-reactor licensees and 10 pCi/cm3 or less for reactor licensees.

Volume of waste, in cubic feet, ranges over four orders of magnitude.

Table 2.2 Types of waste proposed for disposal by NRC licensees pursuant to 10 CFR 20.302 i

Reactor Licensees Institutional / Industrial Licensees l

Soil Paper, plastic, g! ass i

Sludge Animal bedding, carcass l

Sand Sand, soil, rock l

Wood Liquid scintillation media Waste oil Plant matter 7

Ash residue l

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2.4 NRC Review and Assessment In accordance with the regulation, licensees must provide information which includes a description of the contaminated material, the disposal conditions, the physical parameters of the site, and the nature of the environment.

Volume 1 of this NUREG [Neuder, 1986] provides a comprehensive list of information needed from the licensee when requesting authorization for onsite disposal by burial in soil. The licensee's submittal should include sufficient information to enable the NRC to assess the potential hazard to public health and safety and to determine whether the proposed action will have a significant effect on the human environment.

The submittal is initially reviewed for regulatory conformance, i.e., the request should be appropriate under $20.302.

For example, requesting a generic deregu-lation of a specific waste contaminated with certain small concentrations of short-lived radioisotopes would be inappropriate under S20.302 since generic deregulation requires regulatory rulemaking procedures.

An example of an appro-priate request under S20.302 would be seeking authorization to dispose of radioactive incinerator ash by burial on the licensee's own property.

The second step in the review of the licensee's submittal is the determination of completeness of information.

For disposals by burial in soil, the licensee is expected to provide the information listed in Volume 1 of this NUREG.

If the information in the submittal is determined to be complete, the technical assessment based on this information may then be performed.

Volume 1 of this NUREG also provides general guidance to the licensee regarding disposal conditions to be met in order that the proposed disposal be acceptable.

For example, the waste proposed for burial should not contain hazardous chemical constituents as defined in the Environmental Protection Agency (EPA) regulation j

40 CFR Part 261.

Hazardous waste not exempted by statute or rule must be managed

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in compliance with EPA regulations or the requirements of an approved state regulatory program.

Still other conditions to be followed by the licensee require that the location of the disposal site and its contents should be such as to preclude potential-offsite migration or ready access to critical pathways.

For example, the waste should be buried well above the water table.

Locating the site adjacent to a neighbor's property line or next to animal grazing areas would ordinarily not be acceptable.

The NRC reviews the submittal to ensure that these and other conditions described in Volume 1 are met by the licensee.

The next step in the review is to determine whether the proposed quantity of radioactivity fits any of the categories of radionuclides defined in Volume 1

[Neuder, 1986].

Authorization would likely be granted if all requirements and associated disposal conditions described in Volume 1 are met.

Following the preliminary screening of the contents of the submittal, the licensee's radiological impact assessment is reviewed.

The assessment generally consists of modeling the radionuclide release to the environment via site-specific critical exposure pathways and the projection of potential radiologi-cal dose to an onsite individual and to an offsite member of the public.

NUREG-1101, Vol 2 2-4

2.5 Pathways and Scenarios' For burial of radioactive waste in soil, the critical exposure pathways usually include one or_more of the following:

External exposure to direct radiation from the buried waste.

Internal exposure from ingestion of agriculture products grown in radioactive soil.

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Internal exposure from inhalation of resuspended radionuclides.

Internal exposure'due to drinking water from a downgradient well con-taminated by migrating radionuclides.

Internal exposure due to drinking water from an onsite well after cessation of disposal activities.

i Site-specific parameter values necessary to permit modeling the critical path-ways and radiation exposure scenarios are provided by the licensee in accordance with guidance provided in Volume 1.

For example, the depth of burial, the spa-tial array of the waste pits, the cover thickness, and the overall size of the burial site will influence the dose to an inadvertent intruder from direct exter-nal radiation.

The size of the burial site and the subsurface location _of the i

waste will also determine,_to a great extent, the internal dose from crops and animal products derived from the site.

The rate at which the radionuclides enter the environment are also estimated, from information provided by the licensee.

The frequency of burials, the phys-ical form of the waste, and the type of container used to package the waste are some of the parameters considered when estimating a release rate to the sur-rounding medium.

The radioactivity released from the waste will also be limited by the amount of precipitation that infiltrates into the ground and comes into contact with the waste.

Information on precipitation and the quantity of water l

percolating down to the waste may be obtained from the water balance information provided by the licensee.

Radionuclide migration rates through the soil and in the groundwater strongly depend on the chemistry of the waste and the hydro-i logical and geochemical characteristics of the site and are estimated from site-specific parameters.

Several scenarios are appropriate for assessing onsite disposals by burial in soil.

These scenarios have incorporated the critical exposure pathways described l

above.

External Exposure Scenario - After the burial of waste occurs, an individual (intruder) visits, works, or resides on the disposal site for a number of hours throughout the year. While on the site, the intruder is subjected to I

whole-body radiation exposure from buried waste or from surface soil contamina-l tion.

The waste may be located at any depth and have various thickness and cross-sectional extent.

These geometric factors will influence the calculated dose rate.

Agriculture Scenario - An individual occupies the site soon after it is filled and uses the disposal site to raise vegetables, fruit, and livestock for meat l

NUREG-1101, Vol 2 2-5

and dairy products.

Use of the disposal site for raising livestock may be decoupled from raising fruit and vegetable crops depending on the particular situation.

The individual's diet from livestock on the site consist of eggs, milk, beef, pork, and poultry.

The individual's diet from raising fruit and vegetable crops consists of root, leafy, and other above ground vegetables as well as wheat, grain, and fruit.

During the growing season, the individual is assumed to derive part or all of a total diet, depending on the size of the site.

Production periods for most of the agriculture products are assumed to be 90 days.

Area yield rates for food products and annual rates of food con-sumption (Table 4.1) are used to estimate internal exposure due to food ingestion.

Inhalation Scenario - An individual is assumed to work or reside on the disposal area for a number of hours throughout the year.

Soil contamination resulting from an accidental spill during waste disposal operations or from an inadvertent waste exhumation due to agricultural activities is assumed to exist on the sur-face of the soil.

The individual is assumed to be subjected to an inhalation dose from airborne concentrations of radionuclides.

Airborne concentrations may be due to ongoing, time dependent processes (resuspension model) or pre-exist in a relatively static form (mass-loading model).

For the resuspension process, the top 1 centimeter of soil is assumed to be available for resuspen-sion.

A dilution factor may be applied to the soil concentration to correct for dilution due to excavation activities.

Airborne radionuclide concentrations due to resuspension processes will vary depending on the size of the disposal area.

Area correction factors may be applied to the dose calculation as described in Chapter 4 of this document.

Irrigation Scenario - An individual is assumed to use a contaminated water supply j

to irrigate an adjacent field at a rate of 150 liters per square meter per month during a 6-month growing season the site is assumed to be irrigated with con-taminated water from an onsite disposal area for 10 years prior to the beginning of the scenario.

The individual may then be exposed to external radiation from the contaminated soil, to internal radiation from inhalation of resuspended radionuclides, or to internal radiation from ingestion of food products derived from the irrigated field.

Boundary Well Scenario - Radionuclides are assumed to leach from the disposal area and migrate downward to the aquifer.

Uniform groundwater flow is assumed to transport the radionuclides in a horizontal direction toward a potable water well downgradient from the disposal area.

The well is assumed to be near the boundary of the licensee's property and is used either for irrigation or for drinking water purposes.

An individual is assumed to ingest 2 liters per day from this well for a period of 1 year.

Onsite Well Scenario - The onsite well refers to a potable water well constructed within the waste disposal area itself.

The well is assumed to be constructed by an individual after the disposal site is filled and no longer in use.

The radiological impact assessment requires an idealized but realistic descrip-tion of the physical system under consideration.

This description, referred to as a "conceptpal model," must include a sufficient number of physical parameters to permit both the mathematical characterization of the system and a simulation of its behavior in space and time.

Most of the physical parameter values needed NUREG-1101, Vol 2 2-6

to model the system may be derived from the information provided by the licensee.

Where information is not readily available or uncertainty exists on the part of the licensee, conservative but realistic estimates of parameter values will then be made based on the limited information provided and on the likely range of parameter < values applicable to the specific case.

In general, the conceptual model considers the (1) characteristics of the source term, (2) characteristics of the source medium, (3) transport mechanisms, (4) critical pathways, and (5) exposure scenarios.

Each of these five con-L siderations may be quantified, to the extent possible, in terms of site-specific data provided by the licensee.

2.6 Radiological Dose Assessment and Recommendations In general,.the committed effective dose equivalent for the critical exposure pathways will determine whether the licensee proposed disposal will or will not be authorized.

Experience has shown that for burial under prescribed conditions-along with imposed restrictions as to the proximity of nearby water wells on the licensee's property, conservatively estimated doses to the individual from any of the critical pathways generally do not exceed a few millirem for the radionuclides of concern.

As a minimum, prescribed conditions as described in Volume 1 of this NUREG [Neuder, 1986] would generally include:

(1) adequate site characteristics.for limiting radionuclide migration, (2) burials at least 4 feet beneath the soil surface, (3) burials at least 10 feet above the high point of the water table, (4) restricted access, and (5) an exclusion zone for potable water wells in the vicinity of the disposal area.

In keeping with the principle of ALARA, that is, that radiation exposure and release of radionuclide materials should be maintained as low as reasonably achievable [10 CFR 20.1(c)], dose levels conservatively estimated to be no greater than a few millirem per year to offsite individuals are considered to be acceptable levels without being unduly restrictive on the licensee.

It is generally agreed that these levels present a very low risk to public health and safety since they are a small fraction of the dose due to background levels of radiation. These levels do not necessarily apply to cleanup criteria following accidental releases of radioactive material.

Nor do they necessarily apply to residual contamination such as may exist at inactive uranium mining and milling facilities. These situations are evaluated on a case-by-case basis where the potential risk from radiation exposure is weighed against the cost and practi-cality of remedial action.

Recommendations to the licensee by the NRC are generally case specific.

For example, quantities of longer-lived radionuclides proposed for burials may have to be limited in order to keep projected doses within acceptable limits.

Another recommendation might be to not authorize burial of one of several radioisotopes proposed by the licensee for disposal because of excessive curie content of that one isotope.

A frequent recommendation would be to authorize burials with the condition that no drinking water wells exist within a prescribed distance from the burial area.

This presumes that the licensee ~ controls the site sur-rounding the burial area and is capable of providing appropriate surveillance.

j The NRC does not anticipate that disposals will be authorized which will require a period of long-term care after the site is filled.

It is expected, however, that control of the site be maintained for several years after it is filled.

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I NUREG-1101, Vol 2 2-7

3 THEORETICAL CONSIDERATIONS AND COMPUTATIONAL METHODS The committed effective radiation dose equivalent resulting from the primary exposure pathways described earlier may be estimated using the romputer programs and analytical methods discussed in this chapter.

These compu' ational methods and pathways are summarized in Table 3.1.

Table 3.1 Computational methods and associated p.'vays ComputationklMethodandPathways Quantity Calculated 1.

ONSITE/MAXII Code

• 1.

Calculates radiation exposure:

External pathway (a) external exposure from surface and Agriculture pathway buried waste (b) internal exposure Airborne pathway from ingestion of radioactive food products, (c) internal exposure from inhalation of resuspended radionuclides.

2.

M0CH0D84 Code 2.

Calculates radionuclide concentration in Groundwater pathway groundwater.

Area source, two-dimensional groundwater flow, and radionuclide trans-port, dispersion, retardation, and decay.

3.

Analytical Method 3.

Calculates radionuclide concentration in Groundwater pathway groundwater.

One-and two-dimensional analytical computational forms for radio-nuclide transport in flowing groundwater-point, line, area sources with instcnta-neous, exponential and continuous leach-ing functions.

• The ONSITE/MAXII code is not used in estimating the drinking-water dose via the groundwater pathway unless groundwater concentrations are known and can be directly entered as input data.

3.1 The ONSITE/MAXII Code The ONSITE/MAXII code [Napier et al., 1984; Kennedy et al., 1986, 1987] con-tains two primary computer programs, MAXII and ONSITE, and an extensive data base.

Complete listings of the computer programs and data files can be found in Kennedy et al. [:987].

ONSITE is an interactive program that allows the user to create and use the radiation-exposure scenarios. MAXII is then used, with the scenario information, to calculate dose to an individual exposed to radiation from several selected pathways resulting from onsite burial of radio-active wastes.

The data base (Section 3.2) contains data files of dose conver-sion factors for various pathways, master radionuclide and organ metabolic libraries, data, and reference generic environment data.

NUREG-1101, Vol 2 3-1

The ONSITE/MAXII code may be used to model the following pathways:

Direct external radiation exposure from surface contamination.

Direct external radiation exposure from subsurface waste.

Ingestion of crops grown in contaminated soil.

Ingestion of animal products derived from the contaminated site.

Ingestion of crops grown in soil irrigated with contaminated water.

Ingestion of crops contaminated by radionuclide airborne deposition.

+

Inhalation of resuspended radionuclides.

Ingestion of aquatic food products taken from a contaminated stream.

Ingestion of contaminated drinking water.*

Direct external radiation exposure from stored waste.

Site-spei:ific information will generally determine which of these pathways are critical pathways. The ONSITE program solicits scenario information from the user, controls parameter modification, selects the appropriate data files from the data base and constructs the input file for MAXII.

The user is thus given the option of selecting pathways ana parameters for a " user-defined" scenario.

These input files, the radionuclide library, and other appropriate files are then accessed by MAXIl.

For operation on a Vax or Cyber computer, two auxiliary programs, MAXI 2 and MAXI 3, are described in the ONSITE/MAXII documentation [Napier et al., 1984; Kennedy et al., 1986] that allow user modification of the reference environ-l ment data base, if desired.

Dose conversion factors, created by MAXI 2 for'the terrestrial food pathway and by MAXI 3 for the aquatic pathway, are stored in the data files where they can be accessed during the scenario simulations con-trolled by MAXI 1.

The MAXI 2 and MAXI 3 programs are not required for operations I

on a personal computer [ Kennedy et al., 1987).

In addition, the data files contain external and inhalation dose conversion factors.

These factors were calculated using codes external to the ONSITE/MAXII code.

The external dose conversion factors were calculated using the ISOSHLD code [Engel et al., 1966; Simmons et al., 1967] and are used for both the mainframe and IBM-PC versions of the code.

The inhalation dose conversion factors for estimating doses using the ICRP-2 methods and the Task Group Lung Model were calculated using the DACRIN computer program [ Houston et al., 1974].

For the external expcsure (direct gamma radiation) scenario, MAXIl uses the dose conversion factors calculated using the ISOSHLD computer program. These

• Not used for dose assessments via groundwater pathway.

The MOCM0084 code or the analytical methods described in Sections 3.7, 3.8, and 3.9 of this document are used for groundwater pathway analysis.

NUREG-1101, Vol 2 3-2

factors that relate the radionuclide source strength to the dose rate in tissue 1 meter above a finite " plane" of contamination, 15 cm thick, or above a finite slab of contamination, 1 meter thick.

External dose conversion factors are supplied for surface-soil contamination and for buried radioactive waste beneath either a 0.5-meter or 1.0-meter layer of soil.

These dose factors are intended i

to model doses to an individual standing above the contaminated slab for a given period of time.

3.2 Data Base for the GNSITE/MAXII Computer Program For the personal computer version of the ONSITE/ MAXI 1 computer program, the MAXI 2 and MAXI 3 auxiliary computer programs for calculating dose conversion factors for terrestrial and aquatic foods are not used.

Their calculations are done by a modified version of the MAXII computer program directly.

In addition, the personal computer version of MAXII includes the option to consider the International Commission on Radiological Protection ICRP Publication 30 dosimetry methods [ICRP, 1979-1982].

The data base for the personal computer version of the ONSITE/MAXIl computer program consists of [ Kennedy et al., 1987]:

(1) the radionuclide master data library, (2) the organ metabciic data library, (3) the food transfer coefficient library, (4) the aquatic bioaccumulation factor li-brary, and (5) the ICRP Publication 30 dose conversion factor library.

In ad-dition, external dose factors for various source geometries and inhalation dose factors are included as discussed in Sections 3.3 and 3.5, respectively.

The following sections briefly de:cribe the data basec that are used by the personal computer version of the 0NSITE/MAXII computer program.

3.2.1 The Radionuclide Data Library The Radionuclide Data Library contains all radiological data used by MAXII.

This includes data on radionuclides that are not members of decay chains as well as radionuclides that are members of decay chains, radiological half-lives, and translocation class for soluble and insoluble states of the radionuclide.

Translocation refers to the rate at which radionuclides are transported by body fluids from the lungs to the blood and gastrointestinal tract after inhalation.

l The Radionuclide Data Library contains 100 radionuclides for consideration in the radiological assessment of onsite waste disposals.

These are listed in Table 3.2.

These radionuclides are considered most likely to be encountered in institutional, industrial, and reactor low-activity, high-volume wastes -

wastes that are potential candidates for onsite burial.

Several of these radio-nuclides are listed with a +D (plus daughters) designation for the International Commission on Radiological Protection ICRP Publication 2 dosimetry calculations l

[ICRP, 1959].

For these radionuclides, the energies of the short-lived daughters in equilibrium with the parent radionuclides are included in the organ dose and external dose calculations.

For other radionuclides and for application of the ICRP Publication 30 dosimetry system, chain decay calculations are performed I

and daughters are permitted to reach their equilibrium values.

3.2.2 Master Organ Data Library The personal computer version of the ONSITE/ MAXI 1 computer program [ Kennedy et al., 1937] uses a master organ data library for calculating dose to specific organs from ingested radionuclides using the ICRP Publication 2 dosimetry system.

This library, designated as ORGLIB, is not needed when the ICRP 30 dose system NUREG-1101, Vol 2 3-3

Table 3.2 Radionuclides available for ONSITE/MAXII dose assessments Radionuclide R.ndionuclide Radionuclide i

3 144 137m H

Ce+D Ba

_14 152 141C

22h, 154fu 151 u

32 160

- 235 P

Tb 0

33 185 231 35h 191hs 231 36 192 227 C1 II Ac 40 203 227

45f, 210k+D 223fr 46 22G 223 Sc Ra+D Ra 51 228 237 Cr Th+D N

54 230 233 p Mn Th+D Pa 55 '

232 233 T+

59h, 233 229 UD h

57 234 225 Co U

Ra 60 235 +D 225 Co U

Ac 238 238[pD 59 237 I 234 63 -

h I

65 241 234m Zn Pu+D Pa 75 89 234 85f'r 89hh a

242pu 90Sr+D(a) 90 238 Sr N

93 90 238 p n

y 94k 99 244 g

106 99"Tc 244 Ru+D Pu 109 99 240 110k

+D 10kiu 240 Ag u

111 "

103n 243 I

Rh Cn 124 103 243 125$+0 129 243 125 +D 134 239 1

Cs N

131 135 239 p 137fs 241 "u 137 3

p 0

241Am

(*)Where +D means plus short-lived daughters in equilibrium for the ICRP Publicaticn 2 dose factors.

NUREG-1101, Vol 2 3-4 4

--,-n

.,-_.---n

is used.

In ORGLIB, data are arranged.in blocks by radionuclide and are ordered to be compatible with the master radionuclide order given in the radionuclide data library (see Section 3.2.1).

Organ data is included for up to 23 organs of the human body.

4 3.2.3 Food Transfer Coefficient Library The personal computer version of the ONSITE/MAXII computer program uses a master food transfer coefficient library (FTRANSLIB) for relating concentrations of elements in soil to concentrations in farm products produced on that soil. The factors in this library are also used to relate the concentrations in animal feed to concentrations in animal products.

The data contained in the library have entries for 63 elements arranged by increasing atomic number in a manner that is compatible with the radionuclide master data library.

3.2.4 Aquatic Bioaccumulation Factor Library The aquatic bioaccumulation library contains the factors used by the personal computer version of the ONSITE/ MAXI 1 computer program relating the concentration of radionuclides in aquatic biota to the concentration of the radionuclides in the water.

Separate factors are listed for fresh and salt war.er.

Also included l

is a factor for the cleanup of drinking water in water treatment plants.

Data

~

are included for 63 elements arranged by increasing atomic number in a manner that is compatible with the radienuclide master data library.

3.2.5 Internal Dose Conversion Factors Internal dose conversion factors for inhalation and ingestion based on ICRP Publication 30 [1979-1982] are contained in a separate data library.

The library contains data for the same radionuclides considered in the master data library; however, the listing does not require a +D designation to handle the ingrowth of daughter products.

This is because chain decay calculations are performed 4

and daughters are permitted to reach their equilibrium values. The internal dose conversion factors have units of Sv/Bq, and the committed effective dose equivalents reported by the ONSITE/MAXII computer program are automatically converted into units of rem.

f 3.3 External Exposure Pathway - the ONSITE/MAXIl External dose conversion factors (DCFs) for various waste di::posal geometries were calculated using the ISOSHLD code [Engel et al., 1966, Simmons et al.,

1967]. Waste disposal geometries include a 15-cm layer of surface soil conta-mination, a 1-meter-thick slab of contamination tangent to and immediately be-neath the soil surface, a 1-meter-thick slab with 0.5-meter soil overburden and a 1-meter-thick slab with 1.0-meter soil overburden.

Each set of external dose conversion factors contains data for three assumed waste densities.

The dose conversion factors relate radionuclide source strength (or concentration) to the dose rate in tissue at a point 1 meter above the soil surface.

3.3.1 General Source Geometrv The ISOSHLD code, an auxiliary program to the ONSITE/MAXIl codes, was used to create data files of dose conversion factors for use by MAXIl.

IS0SHLD calcu-lates flux and dose rate at an external point due to gamma photons origirating l

NUREG-1101, Vol 2 3-5

= _..- __. -

at all points within a radioactive volume source.

Details about the calculation of flux are given in references by Engel et al. [1966], Simmons et al. [1967],

and Taylor [cf Goldstein, 1971].

The region between the volume source and ex-ternal dose point may contain various layered media such as soil, concrete, water, paper, and air.

The geometry used to calculate total flux at the dose point is that of a truncated, circular cone of radioactive waste beneath shield-ing layers of soil and air.

The dose point is shielded by a 5-cm layer of tissue.

3.3.2 External Dose Conversion Factor Data Files Five external dose conversion factor data sets were created for use by MAXI 1.

These are:

(1) PLANS 0VRC, (2) VOLS0VRC, (3) BURIEDHF, (4) BURIED 1, and (5) STORED.

The first two sets correspond, respectively, to a 15-cm-thick layer and a 100-cm-thick layer of radioactive waste, with upper surfaces tangent to the surface of the soil. The third and fourth sets correspond to a 1-meter-thick subsurface layer of radioactive waste covered by 50 cm and 100 cm of soil, respectively.

The fifth data set simulates stored waste as a rectangular slab source, 10.0 meters by 30.0 meters with a waste thickness of 1.0 meter.

The dose factors relate the radionuclide source strength to the dose rate in human tissue at a point 1 meter above the surface of the soil.

This dose rate corre-sponds to the dose to an individual standing above the waste and exposed to gamma radiation from the subsurface source.

The 15-cm layer is referred to in NUREG/CR-3620 [Napier et al., 1984] as a piarie source.

Each of the five data sets consists of three files for three assumed waste densities; 1.8, 1.0, and 3

0.6 g/cm.

For situations in which the subsurface waste layer is thicker than 1 meter, sensitivity studies have shown that the contribution to the dose rate may be neglected (see Section 3.3.3).

Alternatively, using a 1-meter slab to model a subsurface layer of contaminant which is less than 1 meter thick may result in a conservative estimate (overestimation) of dose rate for the higher energy gamma emitters.

The dose conversion factors stored in the data files were calculated for a sub-tending half-angle of 90 degrees, corresponding to the case of a slab source with infinite cross-sectional area. Most onsite disposal areas are considerably smaller than 1 hectare and would be modeled by using the appropriate area correction factors (see Section 4.3).

3.3.3 External Exposure Sensitivity Studies In the development of the external exposure libraries used by the ONSITE/MAXII computer program, certain simplifying assumptions were made.

A determination of the sensitivity of the results on these assumptions is next discussed.

Two primary assumptions are evaluated in the sensitivity studies:

The effect of waste form density is determined by comparing the external exposure factors originally calculated for a waste form density of 1.8, 3

with factors calculated for densities of 1.0 and 0.6 g/cm.

The effect of the slab-source thickness is determined by comparing the external exposure factors calculated for 1-meter-thick and 3-meter-thick slabs.

NUREG-1101, Vol 2 3-6

The sensitivity studies are performed using the ISOSHLD computer program [Engel et al., 1966; Simmons et al., 1967] and modified input parameters.

A summary of the results generated is shown in Table 3.3 for selected radionuclides based 3 of each radionuclide.

These results on a constant source strength of 1 Ci/m show that there is a strong dependence on waste form density; that is, for a constant source strength, the less the waste density, the greater the calculated dose rate.

However, for each waste density there is little dependence upon slab thickness.

Table 3.3 Sensitivity study results for the external exposure pathway - mrem /hr per Ci/m3 for selected waste form densities and slab thicknesses 3

3 3

0.6 g/cm 1.8 g/cm 1.0 g/cm Slab Thickness:

Slab Thickness:

Slab Thickness:

Radionuclide 1m 3m 1m 3m 1m 3m 22 514 514 964 966 1600 1630 Na 51 12 12 28 28 470 471 54 327 327 684 684 1140 1150 57 21 21 70 70 117 117 Co 60 1003 1003 1930 1930 3200 3270 90fr+D 2

2 6

6 9

9 131 133 133 323 323 539 541 7

134 674 674 1500 1500 2470 2500 137 241 241 573 573 935 942 s+D Based on this result, the operation of the 0NSITE/MAXII computer program has been designed [ Kennedy et al., 1987] to permit the user to select the waste form density when performing external exposure calculations.

The possiole waste 3 (the approximate density form densities that the user may select are 1.8 g/cm 3 (the density of water), and 0.6 g/cm (the approximate deils-3 of soil), 1.0 g/cm ity of noncompacted trash waste forms).

Corrections for slab thickness are not, however, included in the external exposure data libraries.

3.4 Ingestion Pathway - the ONSITE/MAXII The ONSITE/MAXII computer program allows the user to select dose calculations based on the internal dosimetry models described by the International Commission on Radiological Protection (ICRP) in either Publication 2 or 30 [1959; 1979-1982].

The ability to produce ICRP Publication 2-based doses is included so that com-parisons with the Regulatory Guide 1.109 methods [U.S. Nuclear Regulatory Com-mission, 1977] can be obtained.

The ICRP Publication 30-based methods are included to be consistent with the newer dosimetry system considered by the NRC.

Details concerning the ingestion pathway doses calculated by these methods are found in references by the ICRP [1959; 1979-1982].

NUREG-1101, Vol 2 3-7

The NRC may adopt the use of the newer ICRP Publication 30 methods for estimat-ing public radiation doses.

These methods use a system for radiation protection that is based on limiting the total risk of health effects rather than on a controlling or " critical" organ risk. The dose equivalent, H, at a point in tissue is given by the equation [ICRP, 1977]:

H = DQN (3-1) where 1

H=

the dose equivalent at a point in tissue, D=

the absorbed dose, Q=

the quality factor to allow for the effect on the detriment of the microscopic distribution of absorbed energy, and N=

the product of all other modifying factors that might account for the absorbed dose rate and fractionation.

In addition to the basic dose equivalent, the ICRP defined the committed dose equivalent, H o, to a given organ or tissue from a single intake of radioactive 3

caterial.

This quantity is the dose equivalent that will be accumulated over 50 years following the intake [ICRP, 1977]:

t +50y g

Ho=

A(t) dt (3-2) 3 o

where Ho = the committed dose equivalent, 3

A(t) = the relevant dose-equivalent rate, and t = the time of intake.

g For stochastic effects, the ICRP recommended a dose limitation based on the principle that the risk to an individual should be equal whether the whole body is irradiated uniformly or whether there is nonuniform irradiation.

This condition is met provided that [ICRP, 1977]:

WHTT H

(3-3) wnere T = a weighting factor representing the proportion of the stochastic w

risk resulting from tissue (T) to the total risk when the body is irradiated uniformly, H = the annual dose equivalent in tissue (T), and T

NUREG-1101, Vol 2 3-8

"wb,L = the recommended annual dose-equivalent limit for uniform irradiation of the whole body.

For the ICRP Publication 30 dosimetry system, the standard dose that is calcu-lated is this weighted sum of individual tissue (or organ) committed dose equiva-lents known as the committed effective dose equivalent. This value is estimated by the ONSITE/MAXII computer program using the recommendations of the ICRI' [1977; 1979-1982].

3.5 Inhalation Dose-Conversion Factors - The Task Group Lung Model The MAXII computer program uses dose-conversion factors for inhalation that relate the dose to internal organs resulting from breathing air that contains resuspended radionuclides.

These inhalation dose-conversion factors are based on estimates of the distribution and retention of inhaled materials in the lung given by the Task Group on Lung Dynamics Model (TGLM), as developed by the Inter-national Commission on Radiological Protection [ICRP, 1966].

The dose-conversion factors include revisions to the metabolic data for plutonium and other actinides

[ICRP, 1972]. The TGLM model, with minor additional changes to the values of deposition and clearance for aaterials in the regions of the lung, has been applied in the calculation of " Annual Limits of Intake by Workers" in ICRP Pub-lication 30 [ICRP, 1979-1982].

The TGLM model takes into account the particle size distribution of the inhaled These material and defines three retention classes for materials in the lung.

retention classes are used to account for the chemical form of the inhaled mate-rial.

The clearance classes are defined as Class D, W, and Y, and relate to clearance half-times of material from the lung of 10 days or less,10-100 days, In the TGLM model, the respiratory system is divided and greater than 100 days.

into three regions including the nasal passages, the trachea and bronchial tree, and the pulmonary parenchyma [ICRP, 1979-1982].

The deposition of inhaled material is described by parameters that represent the fraction of the material deposited in each of the three regions of the lung.

Aerosols are assumed to contain a log-normal distribution of particle sizes, thus deposition is related to the A stan-activity median aerodynamic diameter (AMAD) of the airborne material.

dard particle size of 1 pm is assumed for estimating the inhalation dose-conversion factors used by the MAXII computer program.

The TGLM model assumes that daughter products behave metabolically like the inhaled parent radionuclides.

The linked, first-order differential equations that define the clearance of material from the lung are modified to include the radioactive decay constant of the daughter and are shown in ICRP Publication 30

[ICRP 1979-1982].

In a similar way, the ICRP Jescribes how the activities of a chain of parent and daughter radionuclides can be treated by the differential equations.

The ICRP TGLM model, in a version of the DACRIN [ Houston et al., 1974] computer program, was used in the calculation of ICRP Publication 2 inhalation dose-These factors can optionally be used by MAXII to estimate conversion factors.

the inhalation dose resulting from the resuspension of radioactive materials in The system of linear differential equations described by the surface soils.

the ICRP in the TGLM model was written in a modified form that would permit The OACRIN computer program easy data entry in the DACRIN computer program.

NUREG-1101, Vol 2 3-9

}

was selected for this application because hand calculations had previously been performed to verify the correct operation of the code.

This step of code veri-I fication is consistent with the quality assurance steps provided for the ONSITE/

MAXII computer software package [Napier et al., 1984].

3.6 Comparison of ICRP 2 and ICRP 26/30 Dose-Conversion Factors The regulations concerning radiation protection in the United States have been historically developed from the recommendations of the International Commission 3

of Radiological Protection (ICRP) and the National Council on Radiation Protec-tion and Measurements (NCRP).

The ICRP issued Publication 2 in 1959.

This publication contains specific recommendations on dose-rate limits, permissible i

body burdens of radionuclides, metabolic data for radionuclides, and maximum j

permissible concentrations (MPCs) in air and water.

New information became available over the next 20 years concerning the effects of radiation, the uptake and retention of radionuclides, and the radioactive decay schemes of parent radionuclides.

As a result, the ICRP issued Publication 26 in 1977 and Publica-tion 30 in 1979 to supersede Publication 2.

Because the two dosimetry systems rely on different metabolic data and organ-specific models, it is difficult to describe a comparison of the dose-conversion factors that result for the two systems on a general basis.

However, previous efforts have been made to provide a comparison of the inhalation comniitted dose-equivalent factors calculated by each method [ Kennedy and Watson, 1981].

This comparison was done for the critical organ for 67 radionuclides.

In the com-parison, committed dose-equivalent factors (over 50 years) were obtained from NUREG-0172 [Hoenes and Soldat, 1977] for the ICPR Publication 2 method and directly from ICRP Publication 30, Part I [1979].

For most radionuclides, the critical organ dose-equivalent factors calculated by each method agree within about a factor of 5 of each other.

A notable exception was found for the iso-topes of uranium where the Publication 30 lung factors are about 20 times higher than the Publication 2 factors.

This increase was determined to be partially due to the use of the Task Group on Lung Dynamics Model [ICRP, 1966] and to the use of updated metabolic data.

In addition to estimating doses for critical organs, the ICRP Publications 26/30 dosimetry system also contains a provision for estimating the effective dose equivalent.

The effective dose equivalent is defined so that stochastic effects, or those malignant and hereditary diseases for which the probability of occur-rence is regarded as a function of dose without threshold, can be determined using a single value for the estimated dose.

In determining the effective dose equivalent, a dose to the whole body is determined as the sum over all organs of the product of an organ-specific weighting factor (representing the tissue or organ stochastic risk to the total risk when the whole body is irradiated uniformly) times the committed dose equivalent received by that organ.

It should be noted that, for public exposure situations, there are some difficulties in determining the appropriate weighting factors and age-dependent dose-equivalent factors that represent the population exposed.

The 0NSITE/MAXII [ Kennedy et al.,1987] computer software package currently allows the user to select the dosimetry models of the ICRP Publication 2 system, or the ICRP Publication 30 system.

Thus, both critical organ and effective whole-body doses can be calculated.

For those radionuclides that are controlled i

a NUREG-1101, Vol 2 3-10

1

\\

c by internal dose resulting from the inhalation or ingestion pathways in the scenario analysis, the critical organ doses should generally be within about a factor of 5.

3. 7 Groundwater Pathway Modeling A rigorous solution to the complex partial differential equations governing groundwater flow and contaminant transport in inhomogeneous saturated porous media would provide a three-dimensional description of radionuclide migration as a function of time at all arbitrary locations throughout the aquifer.

In most cases, a rigorous three-dimensional description of contaminant migration is impractical as well as difficult to achieve because of the high degree of detail needed in the site-specific hydrogeologic and geochemical data and the complex techniques generally required in any three-dimensional modeling.

Groundwater pathway analysis may be simplified if certain conservative assump-tions can be applied in the modeling or if mathematical simulation can be per-formed in one or two dimensions.

For example, contaminant migration may be assumed to be primarily two-dimensional if the aquifer underlying the site is relatively thin.

In this approximation, concentrations in the aquifer are con-sidered to be uniformly distributed along the vertical direction and the mathe-matical description of flow and transport need only be formulated in two dimansinns In many situations where the data are not available, one may proceed toward a solution by selecting conservative parameter values and make simplifying, con-servative assumptions in the modeling.

The solution obtained in this way will overestimate the radionuclide concentrations in the groundwater.

Stated another way, where precise information does not exist, the uncertainties may be bounded with conservative assumptions.

This strategy may be employed for analyzing pro-posed disposal activities pursuant to g20.302 since the primary objective of the radiological impact assessment is to estimate the maximum potential dose to an individual member of the general public.

For example, if the depth of the underlying aquifer is unknown, a relatively small value for the vertical dimen-sion may be assumed and the mathematical analysis proceeds as in the case of the thin aquifer.

This approach will lead to an overestimation of the radio-nuclide concentration since the assumed thinness of the aquifer will underesti-mate the diluting volume of water.

As another example, where the source term cannot be determined, one may make a conservative simplifying assumption that the radioactive inventory leaches into the aquifer over a relatively short period of time.

The resulting calculation will overestimate the radionuclide concentration at points downgradient from the disposal site.

Release rates commonly assumed in the modeling include (1) instantaneous release of the entire inventory to the aquifer, (2) decreasing rate of release over an arbitrary period of time, and (3) constant rate of release over an arbitrary period of time.

The rates of radionuclide release into the aquifer or the period of time over which the release occurs should be chosen to provide conservative values (overestimates) of the radionuclide con-centrations at points downgradient from the site.

Simplifying assumptions in the modeling are acceptable when used with conserva-tively chosen parameters based on the fact that they generally lead to a con-servative estimate (i.e., overestimate) of the radionuclide concentration at NUREG-1101, Vol 2 3-11 i

__9

receptor points downgradient from the source.

The conservative concentration value provides guidance for determining the quantity of radioactivity acceptable for disposal.

The advantage of this modeling approach is that it usually reduces the amount of site-specific data required for a more rigorous pathway analysis and it simplifies the computational techniques needed for performing the radio-logical assessment.

Volume 3 of this NUREG discusses this kind of modeling approach for the groundwater pathway in greater detail.

Mathematical techniques for estimating radionuclide concentrations in ground-water include numerical methods and analytical methoas.

The M0CMOD84 code, described in the next section, is a computer code based on numerical methods for solving the groundwater flow and solute transport equations.

Section 3.9 describes the analytical method.

3.8 Groundwater Pathway - Numerical Model The M0CMOD84 computer code [Konikow and Bredehoeft, 1978; Tracy, 1982] numeri-cally simulates two-dimensional steady-state or transient groundwater flow and radionuclide transport in a nonhomogeneous saturated porous medium with complex geometry.

For two-dimerisional simulation, the conceptual model averages the concentration over the vertical and treats transport through the saturated zone by two-dimensional advection and dispersion.

Radionuclide injection into the aquifer may be modeled as a point, line, or area source with either steady or transient injection conditions.

Radioactive decay and retardation of contami-nant flow caused by radionuclide sorption are factored into the model.

Site-specific hydrogeological data needed as input to the model include porosity, dispersivity, soil solids density, distribution coefficient, and saturated thickness of the aquifer.

Other data input requirements inclu% radioactive decay rates, aquifer geometry, locatien of sources, disposal rates, and loca-tions of recharge and discharge points.

As mentioned above, if site-specific data are not well-defined, the licensee may need to make certain simplifying but conservative assumptions in order to perform the computer simulation.

For example, the aquifer may be approximated as a thin, homogr leous flow system with coastant cross-section.

This assumption will generally lead to a conser-vative prediction of concentration since dispersion and dilution effects will be reduced.

Other conservative simplifications may include the selection of low values for the distribution coefficient or for the perosity from the range of values published in the literature for that type of radionuclide-soil system.

Choosing low values for these parameters tends to reduce dilution of the con-taminant plume.

This type of conceptual modeling is discussed in greater detail in Volume 3 of this NUREG [Goode, Neuder, Pennifill, and Ginn, 1986].

For M0CM0084, with summation convention implied, the expression for two-dimensional areal flow may be written [Kooikow and Bredehoeft, 1978; Konikow and Grove, 1977]:

(T

)=S

+W, (i,j = 1,2)

(3-4) 1 1J J

where T is the transmissivity tensor, h is the hydraulic head, S is the stor-jj age coefficient, and W is the inflow / outflow volume flux per unit area repre-senting fluid sources and sinks.

NUREG-1101, Vol 2 3-12

Choosing the coordinate axes along the principal axes of the transmissivity tensor (T ) = 0, i / j), and using a rectangular, uniformly spaced, block-4 centered finite difference grid in which nodes are defined at the center of the rectangular cells, this equation is solved numerically for hydraulic head for each node at discrete time steps.

The average (macroscopic) groundwater flow velocity is related to the head gradient by Darcy's Law.

The directional components of the flow velocity may be written as:

V _

ij Bh (3-5) 4c 8x.

J where K is the hydraulic conductivity tensor and c is the porosity of the jj aquifer.

After computing the head distribution for a given time step, the flow velocity is computed at each node on the basis of the gradient of the alculated heads.

T M 4quation describing mass transport and dispersion in flowing groundwater Aey r derived from the principle of mass conservation.

For two-dimensional areal ifow, assuming time-independent head gradients and saturated thickness, the mass transport and dispersion in a saturated porous medium may be described [Konikow and Bredehoeft, 1978; Konikow and Grove, 1977] by:

BC_18 (bD.. E ) - V E + (C-C')W (3-6) i 8x cb at b 8x 1] 8x; j

j where C is the concentration of the solute mass (radionuclide concentration),

b is the thickness of the saturated medium, D is the coefficient of dispersion jj tensor, V is the flow velocity in the i-direction, and C' is the concentration j

of solute mass in the source or sink.

For an isotropic porous medium, the dis-persivity may be defined in terms of two constants - longitudinal dispersivity o and transverse dispersivity a.

Moreover, if one of the cartesian axes (say f

t x axis) is aligned with the direction of net flow velocity then the dispersion coefficient is reduced to just two components:

and D

=a O

"E x tx (3-7) xx To account for radioactive decay and the effects of solute sorption in the porous medium, the left-hand term in equation 3-6 is rewritten as R (BC/8t) d and an additional term of the form (-R AC) is added to the right-hand side d

[Tracy, 1982], where A is the radioactive decay rate and R is the retardation d

coefficient.

The retardation coefficient will generally be a non-linear function of solute concentration [Tracy, 1982].

In all cases considered here, R is taken as a constant, namely:

d NUREG-1101, Vol 2 3-13

Rd= 1 + ( b* ) pK (3-8) g d

where K is the distribution coefficient and p is the aquifer solids density.

d For no adsorption effect, Kd = 0 and Rd

• l' From the preceding equations, it is readily seen that any changes in solute concentration will be a function of the dispersion coefficient which in turn

.is related to the dispersivity of the medium and to the groundwater flow velocity.

The determination of flow velocities, in turn, requires simultaneous solution of the flow equation for hydraulic head and the computation of head gradients.

The solution of the transport equation (3-6) provides radionuclide concentra-tions at discrete points of the finite difference grid after each time step.

Volume 3 of this NUREG provides a detailed discussion of the applications of the M0CMOD84 computer code and a sample case study.

3.9 Groundwater Pathway - Analytical Model The differential equation which describes radionuclide migration in non-honogeneous saturated porous media is amenable to exact (analytical) solution if certain simplifying assumptions are made in the modeling.

A simplified analysis is acceptable if one is seeking a conservative estimate of radionuclide i

I concentration and demonstrably conservative parameter values are selected.

To demonstrate the procedure, consider an instantaneous release of radioactivity into a large aquifer with constant flow velocity, and negligible recharge in the vicinity of the disposal area.

From a mass balance analysis [Codell et al.,

1982], the governing equation describing the time varying concentration at an arbitrary point in the aquifer is:

(U

)~

- R AC (3-9)

R d

ax ij i

d j

g where the terms containing R account for radioactive decay and sorption in the d

saturated medium (see discussion in the text following equation 3-7).

As before, assuming an isotropic porous medium with unidirectional convective trans-port along the x direction and using equation 3-7, equation 3-9 becomes:

2 2

0C a2C aC aC 0 C + "t x 097 + "t x az R

7 Y

V

~Y

- R AC V

d 5t " "2 x 5xx z

x ax d

(3-10)

All parameters are as previously defined.

At downgradient distances which are large compared with the dimensions of the disposal area, one may treat an areal release into the aquifer as that of a point source in order to conservatively estimate the radionuclide concentration.

Assume a point source of unit activity instantaneously released into the aquifer with groundwater velocity V and choose the location of the release to be at the top of the aquifer (as illustrated in Figure 3.1).

NUREG-1101, Vol 2 3-14

I La

~- -

n!~ _...,

I Point at 8"'I'**

Source Point Top of Aquifer Figure 3.1 Conceptual Groundwater Model The concentration at a point in the aquifer at a depth I and horizontal distance x from the disposal site is [ Cod 511 et al., 1982]:

1 C(x,y,z,t) =

4cnVt Va a gt (x - Vt/R )

2

~

d y

exp

-At 4a Vt/R Vt/R g

d t

d.

u t

(3-11) f+f cos(]z)exp n=1 NUREG-1101, Vol 2 3-15

To simplify the calculations, make the conservative assumption that the aquifer is relatively thin so that at sufficient distances from the source the radio-nuclides may be considered to be uniformly distributed along the vertical direction. Averaging the concentration along the z direction will make the infinite series term vanish so that the last factor on the right-hand side of equation 3-11 reduces to 1/b.

Further, since peak concentrations will occur along the flow centerline and one is seeking a conservative (upper) estimate, a further simplification may be made by setting y = 0.

Equation 3-11 becomes:

1 (x - Vt/R )

d C(x,t) =

exp

-At 4a Vt/R (3-12) 4cnVtb Ja a E

d gt Equation 3-12 describes the radionuclide concentration C(x,t) in the ground-water at a later time t at an arbitrary distance x from the point of injection.

This expression will overestimate the radioactive concentration since it has been derived from a conservative set of assumptions.

The quantity of interest with respect to the licensee's request for disposal authorization is the mavimum concentration at some withdrawal point, such as a potable water well.

.ad at the distance x from the disposal location.

The peak in the concentration at a distance x will occur at a time t given by:

hy, d(G + w) 2 t = 4A + w,Y 4a[w

-1 (3-13) where w = V/a R.

Substituting this value of t into equation 3-12 will yield pd an expression for the maximum concentration at the water well located at a distance x downgradient from the source.

If site specif.ic data on geochemical retardation factors or hydrogeologic param-eters such as dispersivity, porosity, and groundwater velocity are not well defined, conservative but realistic values may be selected from a range of possible values, taking into account the type of soil, the radionuclides, the physical and chemical forms of the waste, and other pertinent site-specific characteristics.

A further simplication is possible if the half-life of the radionuclide is sufficiently long (i.e., the decay rate A is sufficiently small) so that A << w/4, and if the distance x to the water well is large when compared with the longitudinal dispersivity of (a likely condition) then equation 3-13 reduces to t = xR /V.

Substituting this expression for t into equation 3-12 d

gives:

[

Rd (3-14) 1 C,x(x) =

exp' y

4cnbxRd "E"t NUREG-1101, Vol 2 3-16

It should be emphasized that equations 3-12 and 3-14 will generally overestimate radionuclide concentrations if site-specific or conservatively estimated para-meter values are employed.

However, caution should be exercised when applying generic parameter values to a specific site.

In applying equation 3-14, it should be noted that the interval of time over which the concentration peak exists at the withdrawal point is not specified.

Since dose criteria (Chap-ter 4) are based on annual exposures, an appropriately adjusted time-averaged concentration should be considered when calculating the annual dose.

Analytical methods are recommended for use as a preliminary step in the impact assessment process.

Should the analytical solution (e.g., equation 3-12 or 3-14) predict concentrations in the groundwater which do not pose a threat to public health and safety, the licensee need not pursue a more complex, computer-assisted analysis for the groundwate? pathway.

The reader is referred to NUREG/CR-3332 [Till and Meyer, Ch. 4, 1983], NUREG-0868 [Codell et al., 1982]

and to Volume 3 of this NUREG for additional groundwater analytical models and their solutions, and for discussions on the selection of conservative parameter values.

l 3.10 Onsite Well After the sits has been filled or no longer used for disposal of waste, an in-truder constructs a well within the waste area itself.

The individual is assumed to ingest 2 liters of contaminated water per day, throughout the year.

To estimate the concentration of radionuclides in the groundwater, the diluting volume is calculated from the annual aquifer flow rate beneath the site and the aquifer cross-section taken to be the aquifer thickness times the width of the waste disposal area.

This is a conservative estimate of the volume of water which dilutes the radionuclides to a certain concentration and permits a dose rate determination.

The conservative assumption is made that the radioactivity remaining at the time of the final disposal will be relesed to the groundwater in a brief period of time.

i In the absence of site-specific data, a very conservative assumption may be made that the volume of water for dilution of contaminants is that quantity representing the annual needs of an individual living in a rural area.

This l

quantity of water is assumed to be the average annual per capita volume with-drawn from a rural domestic well.

The average per capita use in the United States is 104 gallons / day [Solley et al., 1983], so that the total dilution volume is the annual quantity of 3.8 x 10 4 gallons.

These conceptual models are useful for obtaining very conservative, upper-bound estimates of the radionuclide concentrations in the onsite well.

l r

I i

NUREG-1101, Vol 2 3-17

n 4 DOSE CALCULATIONS AND ADDITIONAL COMPUTATIONAL CONSIDERATIONS The basic equation for calculating radiation dose to the individual from any of the radionuclide pathways previously described is [Soldat et al., 1974; Napier ot al., 1984; Kennedy et al., 1987]:

R

=C U D (4-1) ipt

$p p ipr where Ripr = dose rate from radionuclide i via pathway p to organ r, C

= concentration of radionuclide i in the medium of pathway p, jp U

= usage parameter associated with exposure pathway p,and p

ipr = dose conversion factor for nuclide i, pathway p and organ r.

D Concentrations of radionuclides in soil, food, or air must either be prescribed or derived from other considcrations.

For external exposure due to buried radio-active waste or soil contamination, the radionuclide concentration in the waste or soil is usually determined from measurements or from dividing the curie inven-tory by the volume of waste or soil.

The following sections contain brief dis-tussions of the procedures used for estimating radiation doses from environmental pathways.

4 4.1 Usage Parameters For ingestion of agricultural products from a contaminated site, the concen-tration in crops or in animal products is derived from considerations of root transfer from soil, areal deposition onto vegetation, and animal consumption of contaminated forage and feed.

Concentrations of radionuclides in irrigated l

farm produce are derived from concentrations of radionuclides in the irrigation water.

Contaminated farm products may result from radioactive contaminants l

located on the soil surface, root penetration into the radioactive waste, irri-gation with contaminated water, or airborne deposition on leafy crops and pastureland.

Table 4.1 contains a listing of the terrestrial and aquatic foods which com-prise the entire diet of the intruder.

Concentration factors are calculated based on this diet.

The modeler's input will define the fraction and type of diet derived from the site.

Consideration was given to the radioactive decay time between harvest and consumption of the radionculides (holdup time) when L

calculating concentration factors for consumed foods.

i For ingestion of drinking water from a contaminated well, the concentration in the well downgradient from the site is calculated from a knowledge of the hydro-geologic parameters and modeling the groundwater flow and contaminant trans-port, as described in the previous chapter.

NUREG-1101, Vol 2 4-1

i 1

Table 4.1 Parameters used for calculation of radiation dose factors from consumption of foods Food (days)

(kg/m )

(days)g,)

Consumption (b)

Growing Period Yield Holdu 8

(kg/ year)

Leafy vegetables 90

1. 5 1

9.5 Other aboveground 60 0.70 1

9.5 vegetables Root vegetables 90 9.0 1

76 Fruit 90 1.7 10 42 Wheat and grain 90 0.72 10 51 Eggs 90 0.84(c) 2 19 Milk 30 1.3(c) 2 110(d)

Beef 90 0.84(C}

15 39 Pork 90 0.84(c) 15 29 Poultry 90 0.84(c) 2 8.5

?

6.9 Fish

(*) Time between harvest and consumption.

(b)These rates are obtained from Regulatory Guide 1.109 [U.S., NRC, 1977]

and prorated by food category using the fraction of total consumed by an average individual as calculated from Napier [ Table 8,1981].

(c) Yield of animal feed (i.e., grain or pasture grass).

(d) Units of liters / year.

I For inhalation of airborne constituents, the concentration of resuspended radio-nuclides is derived from the surface concentration by using a time-dependent resuspension factor or resuspension rate analysis [Anspaugh et al., 1975], or by using a mass-loading approach in the absence of data for the specific site.

The usage parameter, U, is an exposure rate or an intake rate associated with p

pathway p.

The usage parameter for the external exposure pathway would usually be specified in hours per year. Other units for U would be mass per unit time p

for food ingestion, liters per day for drinking water and volume per unit time for inhalation of airborne contaminants.

For example, the occupational expo-sure rate to a working adult would be 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year while the intake rate for drinking water for the average adult would be 2 liters per day.

The dose NUREG-1101, Vol 2 4-2

rate, R, is generally specified in millirem per year. - If R is an annual radia-tion dose equivalent or a committed radiation dose equivalent, then it would'be expressed in millirem.

For example,.if the soil concentration is in pCi/ge, the exposure rate in hr/yr and the dose conversion factor in mrem /hr per pCi/gm, then the dose rate will be in mrea/hr.

4.2 -Dose From External Radiation An individual standing on contaminated soil or above buried waste may be sub-jected to direct external beta and gamma photon radiation emitted by the radio-isotope sources. -The equation used for calculating the external exposure dose rate to-the organ of reference is:

n C

U D A

(4~2)

R

• j[y jp p ipr cp r

where'A is the area correction factor which is used for contaminated areal cp extents different from 1 hectare (see Section 4.3).

The sum is taken over j

n nuclides and subscript p refers to the external pathway.

The dose conversion factors were tabulated for unit surface-soil contamination and for subsurface or buried contamination.

The dose factors were calculated.

for absorption by 5-cm tissue at 1 meter above the soil surface.

Only whole-4 body dose is calculated.

The dose includes exposure to decay gammas ano to i

bremsstrahlung radiation generated in the source medium.

4.3 Area Correction Factor for External Exposure The area correction factor adjusts for the size of the waste burial area, based on fractions of a hectare.

The hectare is taken as the size of the reference i

site, being dose-wise equivalent to the infinite slab source.

Sensitivity studies show that the variation of exposure rate with source area is uniform for a wide range of beta gamma energies, so that a uniform correction.across the range of radioisotopes may be applied [Napier et al., 1984].

For all radionuclides of concern, the ratio of exposure rate for small area to exposure rate for large area as a function of fractional hectare of source area is shown'in Figure 4.1.

l This curve is approximated in the ONSITE computer program as the sum of four linear segments with different slopes.

The area correction factors are there-I fore a function of the fractional hectare (input parameter).

Further details may be found in NUREG/CR-3620 and its supplements [Napier et al., 1984; Kennedy et al., 1986, 1987].

l l

4.4 Dose From Ingestion of Food Products j

Contamination of crops grown on the site may be due to airborne radionuclide deposition on leafy vegetables and on soil, irrigation with contaminated water 1

or root uptake by crops grown in contaminated soil.

Concentration,of radio-nuclides in animal products such as meat, eggs, and milk is due to ingestion of contaminated forage, feed, or water by the animal.

A full description of i

the models used for calculating dose from ingestion of food products is given I

in Kennedy et al. [1987].

i' NUREG-1101, Vol 2 4-3 J

.r.

m 3.---w-a, n-

.m e w

.m--~.v--:=

-...----.-.-.,=-e,---.

s.--

i--4

---

.----------m

1.0 0.8 o

• ce E

0.6 e

lii m

2o 0.4 -

o E

ua 0.2 '

I i

0 0

0.06 0.10 0.15 Fraction of 1 ha l

FIGURE 4.1 Area Correction Factors NUNEG-1101, Vol 2 4-4

The deposition rate of airborne particulates may be described by:

dj=8.64x104 igj (4-3)

V where 3

dj= airborne deposition of radionuclide i, pCi/m -day, 3

5 = annual average air concentration of radionuclide i, pCi/m, and 9

V9= deposition velocity of radionuclide i, m/sec.

The coefficient 8.64 x 104 is a unit conversion factor of seconds / day.

The deposition velocity is assumed to have the value of 1 x 10 3 m/sec for all par-ticulates. This value was chosen to reflect a somewhat greater degree of deposi-tion than usually encountered fgr finer particulates in elevated stack releases.

The average air concentration, X, may be calculated either by the method of mass-loading or by resuspension analysis (user option).

4 The deposition rate for radionuclide i from irrigation water is given by:

I = C,I (4-4) d 9

g where d I=

deposition rate by irrigation, pCi/m -day, 2

9 C,=

radionuclide concentration in irrigation water, pCi/f., and 9

2 irrigation rate, 2/m -day.

I

=

Knowing the deposition rates will allow for the determination of the radionuclide concentrations in crops.

This concentration is made up of the sum of four terms:

(1) deposition on foliage due to resuspension and irrigation, (2) root uptake in contaminated soil due to deposition from resuspension and irrigation, (3) root uptake in waste disposed in the first 15 cm of soil, and (4) root uptake in waste buried below 15 cm.

The contribution of each term to the concentration in food products is described in Kennedy et al. [1987].

The sum of the radionuclide concentrations in crops, C, due to all sources y

described is given by:

4 n

1 C(ivk)

C

=I (4-5) y k=1 1=1 NUREG-1101, Vol 2 4-5

The radionuclide concentration in animal products such as meat, milk, and eggs due to ingestion of contaminated feed or forage derived from the site is de-scribed in Kennedy et al. [1987].

To account for the limited exposere potential from smaller disposal areas, site area correction factors are required in the ONSITE/MAXII computer program.

Alternatively, the fraction of the total diet grown on the site may be adjusted to account for limited quantities of agricultural products derived from a small disposal area.

The user will have the option to adjust. intake quantities by specifying either fractional hectare of disposal area 'oT fractional diet (but not both).

4.5 Dose From Inhalation An individual working or rr. siding on the disposal site may be subjected to an inhalation dose from contaninated airborne concentrations of radioactive parti-culates.

Airborne contaminant concentrations may be due to surface-soil contami-nation resulting from waste disposal operations, from excavation operations or from accidental spills. Air concentrations may be calculated using resuspension analysis [Anspaugh et al., 1975] or the average mass loading of the atmosphere.

The average airborne concentration i of resuspended contamination, in pCi/m3, is given by:

i(t)=S(t)S (4~0) f A

where 57 = resuspension factor, m 1, and 2

SA

• surface contamination deposited per unit area, pCi/m,

The resuspension factor, S, is defined as the resuspended air concentration f

per unit surface deposition.

The 0NSITE/MAXII computer program allows the user to select an empirically derived expression for the resuspension factor [Anspaugh, 1975].

Alternatively, the average airborne concentration of resuspended contaminant is the product of the soil concentration, S, in pCi/gm, and the A

average mass loading of the atmosphere.

In the absence of site-specific data, 3

the mass loading value may be taken as 100 pg/m as discussed in Napier et al.

[1984]. The user may select either the resuspension model or the mass-loading model when calculating the inhalation dose.

The ONSITE/MAXII computer program uses dose conversion factors for inhalation based on either the ICRP Publication 30 or ICRP Publication 2 dosimetry methods.

The ICRP Publication 2 inhalation dose conversion factors were calculated using the Task Group Lung Model as contained in the DACRIN [ Houston et al., 1974]

computer program.

The area correction factor for the inhalation pathway corrects for the decrease in resuspension with decreased site area.

The area correction factor for the inhalation pathway is taken to be the same as that for the external exposure pathway [ Kennedy et al., 1987], shown in Figure 4.1.

NUREG-1101, Vol 2 4-6

4.6 Dose From Drinking Water The concentration of radionuclides downgradient from the disposal site may be calculated from the MOCM0084 computer code or from the analytical solutions as described in Chapter 3.

The water ingestion rate is assumed to be 2 liters per day for an individual drinking water from either the boundary well or the onsite well.

Fifty year committed dose equivalents are calculated using internal dose conversion factors tabulated in the literature [0unning et al, 1981].

If the'radionuclide concentration is changing rapidly, either because of radioac-tive decay or groundwater flow dynamics, care should be taken in specifying the concentration level.

An annual average concentration should be calculated before calculating the committed dose equivalent.

The equation used for calculating the 50 year committe dose equivalent to an organ of reference is:

n (4~7)

C,Dir (Oose)7 = U, i =1 I

9 where the subscripts w and i refer to the water pathway and the organ of refer-ance, respectively.

4.7 Dose From an Onsite Well The radionuclide inventory that remains at the time the disposal area is filled or when the land is released for public use is taken as the source of potential radiological dose for the onsite well.

The concentration of radioactive contami-nants is calculated as described previously. Water ingestion rates are taken as 2 liters per day.

Fifty year dose commitments are calculated using the inges-tion dose conversion factors tabulated in the literature [ Dunning et al., 1981],

and equation 4-7 given previously.

4.8 Source Term Considerations Computational adjustments in the source term should be made before the curie quantity or the radior.uclide concentration is specified as an input to the com-puter programs.

In general, the input data for the source term would not be the same as the radionuclide quantity or concentration initially buried by the licensee.

This difference is due to radioactive decay prior to reaching the transport medium or to non-uniform spatial distributions at the burial loca-tion.

These computational adjustments are described below.

4.8.1 Temporal Averaging The ONSITE/MAXII computer program calculates dose on an annual basis.

Inherent in these calculations is the assumption that the concentration of the radio-nuclides in the soil is essentially constant over the 1 year time period.

That is, while radioactive decay is accounted for by the exponential function in the dose equations, the time step interval in the exponential function is no finer than 1 year.

In most cases of waste disposal by burial in soil, the frequency of burials may range from once or twice per month to once or twice per year.

Thus, because of radioactive decay, the concentration specified at the time of NUREG-1101, Vol 2 4-7

any one burial may be very different from the concentration found at the end of the 1 year period.

Adjustments should be made by averaging the concentrations over 1 year before entering the quantity for computation.

For each radionuclide in the waste, the concentration C(t) at any time t after disposal is:

-At C(t) = C(o) e (4-8) where C(o) = concentration at the time of disposal, and A = decay constant for that nuclide.

The time average concentration c, over period T, is obtained from:

c=ffC(o)e

-At dt (4-9) o For a 1 year period, the time-averaged radionucl,ide concentration is:

)

c=

[1 e-A]

(4-10)

If the half-life is short, as for example P-32 (half-life 14.3 days), the annual average will clearly be much smaller than the initial concentration.

For long-lived radionuclides, equation 4-10 reduces to c = C(o) as expected.

The terms "long-lived" and "short-lived" are both relative to the averaging period.

Tritium, for example, with half-life 12.3 years is considered long-lived for the purposes of this discussion.

Many of the radionuclides found in non-reactor waste streams are short-lived (half-lives less than 1 year) and their concentrations should therefore be time averaged over the 1 year period.

The same averaging considerations apply to the groundwater flow and contaminant transport scenario where analytical solutions are used. When using the M0CM00 code, however, one may account for the sequential injection of radioactive con-taminants directly in the user input.

That is, the M0CMOD input provides the flexibility for time sequencing injections as well as contaminant injection at different locations. Time averages are therefore not used with the MOCMOD code.

In addition, radioactive decay is accounted for by the M0CM00 code itself so that time-averaged corrections for decay need not be done.

4.8.2 Spatial Averaging The input to the ONSITE/MAXII program requires the specification of the contam-inated area in fractional hectare.

The assumption inherent in the MAXI calcu-lations is that the surface contamination area or contamination volume of the waste contains a homogeneous distribution of radionuclides.

Specifically, the dose conversion factors in the data files are based on homogeneous distribution of contaminants throughout the surface layer or volume of waste.

After a trench or pit is filled by the licensee's waste disposal operation, a new trench will be excavated in an adjacent area in order to accommodate NUREG-1101, Vol 2 4-8

i subsequent disposals by the licensee.

In general, trenches or pits will not be contiguous but will be separated by areas of undisturbed soil.

To model this geometry, it is assumed that the trenches have equally spaced centers, are of equal depth and are separated by uncontaminated areas of soil.

Because of this non-uniformity in radionuclide concentration across the disposal site, an average concentration is calculated using:

n

.I c) Aj j=1 (4-11) c

=

j n

I A 3

j=1 where 4 = spatially averaged concentration of i-th radionuclide, c

c) = concentration in the j-th disposal region, and A) = area of j-th disposal region.

when the disposal region is uncontaminated, c. is set equal to zero in the calculation.

J 4.8.3 Container Credit Waste packaged in strong, tight containers may preclude any radionuclide migra-tion for several years after burial.

Corrosion studies have shown, for example, that maximum pitting depths of carbon steel (frequently used waste drum material) at the end of 2 years in various kinds of soil do not exceed 1 or 2 millimeters of material [American Society For Metals, 1978].

Other ferrous metals show similar rates of pitting when buried in soil.

Minor changes in composition and structure of steel, for example, are not important to corrosion resistance.

Thus, a copper-bearing steel, a low-alloy steel, a mild steel, and wrought iron are found to corrode at approximately the same rate in any given soil.

When estimating the source term for dose calculations, credit is given for radio-active decay while still in the waste package.

4.8.4 Storage Credit Radioactive waste may sometimes be stored for partial decay over a period of several months prior to disposal by burial in soil.

In this situation, correc-tions are made to account for the radioactive decay of the source term before specifying the input data.

4.8.5 Unsaturated (Vadose) Zone Considerations Radionuclide migration time through the unsaturated zone may result in a sub-stantial reduction in the source term because of radioactive decay.

The many uncertainties which exist in attempting to model unsaturated flow allow only a NUREG-1101, Vol 2 4-9

rough estimate of the period of time for the nuclides to reach the water table.

The M0CMOD84 code simulates groundwater flow and contaminant transport only in the saturated zone so that the data input should reflect the reduced levels of contamination due to radioactive decay during the migration period in the unsaturated zone.

4.9 Comparison With 10 CFR Part 61 Methods and Results An independent derivation of the 10 CFR Part 61 [ Code of Federal Regulations, 1986] low-level waste disposal limits was performed using the MAXII computer program [ Kennedy and Napier, 1984] as a check of the modeling procedure.

The derivation used the intruder construction and intruder agriculture scenarios, as defined in the Draft Environmental Impact Statement (DEIS) in support of 10 CFR Part 61 [U.S. NRC, 1981].

The disposal limits shown in 10 CFR Part 61 [1986] are listed for three classes of commercial radioactive wastes:

Classes A, B, and C.

Class A wastes have minimum stability requirements and low activity levels, and reflect 100 years of radioactive decay that would occur during an institutional control period.

Class B wastes permit higher activity levels and must meet more rigorous waste-form requirements to ensure stability after disposal.

Class C wastes are re-quired to have a stable waste form and a package with higher integrity than required for Class A or B wastes, and reflect 500 years of radioactive decay.

Disposed Class C wastes are further assumed to provide 10 times more protection from intrusion than provided by disposed Class A wastes.

The results of the derivation of the Class A and Class C disposal limits using the MAXII computer code, based on maximum annual dose instead of the 50 year dose commitment, are shown in Table 4.2.

By carefully following the scenario descriptions given in the DEIS and correctly accounting for radioactive decay, the result generated by MAXII generally compare closely to the 10 CFR 61 dis-posal limits. The notable exceptions to this close agreement are the disposal limits for 99Tc and the Class C disposal limit for 137Cs where the calculated values are about 10 times the 10 CFR 61 value.

The difference in the 99Tc con-centrations is because a drinking-water scenario, not the intruder scenarios, controlled the 10 CFR 61 value.

The 137Cs difference for Class C wastes may be the result of radioactive decay calculational differences.

The general agreement of MAXII calculated values with the 10 CFR 61 disposal limits, accounting for minor modeling differences, indicates that the results from the two codes compare quite closely.

i I

1 i

i NUREG-1101, Vol 2 4-10

,y_,,,,.

._.7

._y,__

.,__.,,._-___.,,_._-____.,_,.,,_.v.-_

_.., _ -.., - _ _ _,... ~

Table 4.2 Comparison of calculated and 10 CFR Part 61 low-level waste disposal concentrations 10 CFR Part 61 Calculated Concentration Concentration-3 3

(Ci/m )

(Ci/m )

Radionuclide Class A Class C Class A Class C "C

0.8 8

0.8 8

--(a)

--(a) 400 GoCo 700 s9Ni 2.2 22 1

10 63Ni 3.5 700 1

200 90Sr+0(b) 0.04 7000 0.03 5000 99Tc 0.3 3

3 30 ID) 1 4500 0.3 30000 137Cs+D 839Pu 10(c) 100(c) 30(c) 300(c)

(a) Dashes indicate that no Class C limits are established (i.e., the concen-tration is limited only by practical considerations including the stability of the waste form, internal heat generation, and handling).

(b)

+D means plus short-lived daughters.

(c) Units for.239Pu are in nCi/g.

4 NUREG-1101, Vol 2 4-11

p i

L l

5 REFERENCES American Society for Metals (ASM), Metals Handbook Ninth Edition, Volume 1, 725-731, 1978.

Anspaugh, L. R., J. H. Shinn, P. L. Phelps, and N. C. Kennedy, "Resuspension and Redistribution of Plutonium in Soils," Health Physics, 29, 571-582, 1975.

^ Beck, T. J., L. R. Cooley, and M. R. McCampbell, " Institutional Radioactive Wastes," U.S. Nuclear Regulatory Commission, NUREG/CR-1137, 1979.

Brenneman, F. N., and S. M. Neuder, "0nsite Disposal of Radioactive Waste:

Guidance for Disposal of Reactor Wastes Both Onsite and by Other Modes," U.S.

Nuclear Regulatory Commission, NUREG 1101, Volume 4, in preparation.

Code of Federal Regulations, Title 10, " Energy" and Title 40, " Protection of Environment," U.5. Government Printing Office, Washington, DC, 1986.

Codell, R. B., K. T. Key, and G. Whelan, "A Collection of Mathematical Models for Dispersion in Surface Water and Groundwater," U.S. Nuclear Regulatory Commission, NUREG-0868, 1982.

Dunning, D. E., Jr., G. Killough, S. Bernard, J. Pleasant, and P. Walsh, "Esti-mates of Internal Dose Equivalent to 22 Target Organs for Radionuclides Occurring in Routir.e Releases From Nuclear Fuel-Cycle Facilities," Dak Ridge National Laboratory, NUREG/CR-0150, Volume 3,1981.

Engel, R.

L., J. Greenborg, and M. M. Hendrickson, "ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis," BNWL-236, Pacific Northwest Laboratory, Richland, Washington, 1966.

Goldstein, H., Fundamental Aspects of Reactor Shielding, Addison Wesley, Reading, Mass., 1959; Johnson Reprint Corp., NY, 1971.

Goode, D.

J., S. M. Neuder, R. Pennifill, and T. Ginn, "0nsite Disposal of Radio-active Waste:

Estimating Potential Groundwater Contamination," U.S. Nuclear l

Regulatory Commission, NUREG-1101, Volume 3, 1986.

Hoenes, G. R., and J. K. Soldat, " Age Specific Radiation Dose Commitment Factors l

for a One-Year Chronic Intake," Battelle Pacific Northwest Laboratory, NUREG-0172, 1977.

Houston, J. R., D. L. Strenge, and E. C. Waston, "DACRIN - A Computer Program for Calculating Organ Dose from Acute or Chronic Radionuclide Inhalation,"

BNWL-B-389, Pacific Northwest Laboratory, Richland, Washington,1974.

International Commission on Radiological Protection (ICRP), Report of ICRP Committee II on Permissible Dose for Internal Radiation.

ICRP Publication 2, Pergamon Press, New York, 1959.

NUREG-1101, Vol 2 5-1 e

v._

r

.-.-.,m.

" Task Group on Lung Dynamics for Committee II of the International Commission on Radiological Protection," Health Physics, 12, 173, 1966.

--, The Metabolism of Compounds of Plutonium and Other Actinides, ICRP Publication 19, Pergamon Press, Oxford, England, 1972.

--, Recommendations of the International Commission on Radiological Protection, ICRP Publication 26, Pergamon Press, Oxford, England, 1977.

--, Limits for Intakes of Radionuclides by Workers, ICRP Publication 30, Part 1-4 and Supplements, Pergamon Press, Oxford Eiigland, 1979-1982.

Kennedy, W. E., Jr. and B. A. Napier, "An Independent Derivation of the 10 CFR Part 61 Commercial Low-Level Waste Disposal Limits," PNL-SA-11983.

Presented at the Annual Meeting of the American Nuclear Society, New Orleans, Louisiana, 1984.

Kennedy W. E., Jr., and E. C. Watson, "MPC and ALI:

Their Basis and Their Com-parison," PNL-SA-9349.

Presented at the Joint Meeting of the Northern California and Columbia Chapter of the Health Physics Society, South Lake Tahoe, California, March, 1981.

Kennedy, W. E., Jr., R. A. Peloquin, B. A. Napier, and S. M. Neuder, " Intruder Dose Pathway Analysis for the Onsite Disposal of Radioactive Wastes:

The ONSITE/MAXII Computer Program," U.S. Nuclear Regulatory Commission, NUREG/CR-3620, Supplement No. 1, 1986.

l Kennedy, W. E., Jr., R. A. Peloquin, B. A. Napier, and S. M. Neuder, " Intruder Dose Pathway Analysis for the Onsite Disposal of Radioactive Wastes:

Operation on a Personal Computer," U.S. Nuclear Regulatory Commission, NUREG/CR-3620, Supplement No. 2, 1987.

Konikow, L. F., and D. B. Grove, " Derivation of Equations Describing Solute, l

Transport in Groundwater," U.S. Geological Survey Water Resources Investigation, l

77-19, 1977, Revised 1984.

l l

Konikow, L.

F., and J. D. Bredehoeft, " Computer Model of Two-dimensional Solute Transport and Dispersion in Ground-water," U.S. Geological Survey Techniques of Water Resources Investigations, Book 7, Chapter C2, 1978.

Napier, B.

A., Standardized Input for Hanford Environmental Impact Statements, PNL-3509, Part 1, Pacific Northwest Laboratory, Richland, Washington,1981.

--, R. A. Peloquin, W. E. Kennedy, Jr., and S. M. Neuder, " Intruder Dose Path-l way Analysis for the Onsite Disposal of Radioactive Wastes:

The ONSITE/MAXII Computer Program," U.S. Nuclear Regulatory Commission, NUREG/CR-3620,1984.

Neuder, S. M., "Onsite Disposal of Radioactive Waste:

Guidance for Disposal by Subsurface Burial," U.S. Nuclear Regulatory Commission, NUREG-1101, Volume 1, 1986.

NUREG-1101, Vol 2 5-2

4 Simmons, G. L., J. J. Regimbal, J. Greenborg, E. L. Kelly, Jr., and H. H. Van Tuyl, "ISOSHLD II Code Revision To Include Calculation of Dose Rate From Shielded Bremsstrahlung Sources," BNWL-236, Supplement 1, Pacific Northwest Laboratory, Richland, Washington, 1967.

Soldat, J. K., N. M. Robinson, and D. A. Baker, "Models and Computer Codes for Evaluating Environmental Radiation Doses," BNWL-1754, Pacific Northwest Laboratory, Richland, Washington, 1974.

Solley, W.

B., E. B. Chase, and W. B. Mann IV, " Estimated Use of Water in the United States in 1980," Geological Survey Circular 1001, United States Department of the Interior, 1983.

Till, J. E., and H. R. Meyer, Eds., " Radiological Assessment," U.S. Nuclear Regulatory Commission, NUREG/CR-3332,1983.

Tracy, J.

V., " Users Guide and Documentation for Adsorption and Decay Modifica-tions to the USGS Solute Transport Model," U.S. Nuclear Regulatory Commission, 4

NUREG/CR-2502, 1982.

1 U.S. Nuclear Regulatory Commission, " Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CFR Part 50, Appendix I," Regulatory Guide 1.109, Rev. 1, 1977.

--, " Draft Environmental Impact Statement on 10 CFR Part 61, Licensing Require-ments for Land Disposal of Radioactive Waste," NUREG-0782, Volume 1, 1981.

Wild, R. E., O. Oztunali, J. Clancy, C. Pitt, and E. Picazo, " Data Base for Radioactive Waste Management," U.S. Nuclear Regulatory Commission, NUREG/CR-1759, Volume 2, 1981.

I 9

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BIBLIOGRAPHIC DATA SHEET NUREG-1101 Vol. 2 Sit eNSTRUCTIONS Oas rHE R4 Vf RsE 3 TsTLE AND tustett J LE Abt BLANE Onsite Disposal c f Radioactive Waste Methodology.'or the Radiological Assessment of

. onpai oar w-avio DisposalbySubshrfaceBurial

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Ffbruary 1987 F v t f oaueNG oR3amilaisoN NAME A MA8DNG AoDats frac'vels Ceses aoJEcfiT Asa wore UNif NUMOER Division of Waste Man ement r

Office of Nuclear Mate 'al Safety and Safeguards P oa caa~' avva a U.S. Nuclear Regulatory ommission Washington, DC 20555

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13 SUPPLEMEN f aar NOTES 3

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Volume 1 of this NUREG provides gui for academic, medical, and industrial licensees seeking authorization to dispose o s 11 quantities of radioactive material by onsite subsurface disposal. Licensee re e s for such authorizations are made pursuant to Section 20.302 of 10 CFR Part 20 S 'nda s for Protection Against Radiation." This volume (volume 2) describes the r eria d technical methodology used by NRC staff to evaluate requests by licensees o 1'approva of onsite disposal by burial in soil. The technical methodology include /t ONSITE/M 11 code for calculating radiological exposure from various pathway, he M0 MOD 84 e, and analytical methods for calculating contaminant transoort and co tration of ra nuclides in flowing groundwater.

Radiological exposure analys sfinclude the foll ing pathways: (1) exposure to direct gamma from any surface cont l' nation or buried w te, (2) drinking water from a well contaminated by migration q adionuclides, (3) i stina agricultural products derived fromradionuclide-contaminf}dsoil,and(4)inhall radionuclides resuspended at the burial site. Licensee-proposed disposal activities e evaluated in terms of radiological l

impact on public health and safety and tne environmen The estimated committed l

effective dose equivalent fesulting from the technical valuation will usually be the determining factor in the authorization of the proposed isposal.

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