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UNITED STATES

* s - a 8 NUCLEAR REGULATORY COMMISSION

WASHINGTON, 0. C. C.5

MAY 1 1 1983

ATTENTION: Commission Licensees

SUBJECT: FINAL WASTE CLASSIFICATION AND WASTE FORM TECHNICAL POSITION PAPERS

By Federal Register Notice dated December 27, 1982 (47 FR 57446), NRC amended

its regulations to provide specific requirements for licensing facilities for

the land disposal of low-level radioactive waste. The majority of these requirements are contained in a new Part 61 to Title 10 of the Code of Federal Regulations (10 CFR Part 61) entitled "Licensing Requirements for Land Disposal of

Radioactive Waste." Some additional requirements directed primarily at waste

generators and handlers including certification and use of shipping manifests

were concurrently published as a new 20.311 of Part 20 ("Standards for

Protection Against Radiation").

As noted in the December 27 Federal R ster Notice, the effective date of

10 CFR Part 20, 20.311 is December 27, i193, while the effective date of

10 CFR Part 61 and all other amendments is January 26, 1983. Section 20.311

requires that any licensee who transfers radioactive waste to a land disposal

facility or to a licensed waste collector or processor must classify the waste

according to 61.55 of 10 CFR Part 61. Licensed waste processors who treat

or repackage radioactive waste for disposal into a land disposal facility must

also classify their waste according to 61.55. This section defines radioactive waste suitable for disposal as falling into one of three classes (Class A,

Class B, or Class C), and waste is determined to fall into one of the classes

by comparison to limiting concentrations of some particular listed radionuclides.

Class B and C wastes are subject to waste stability requirements which are set

forth in 61.56 of the rule. In addition, 20.311 also requires that waste

generators record on shipment manifests a description of the transferred waste

as well as a certification that the waste is properly classified anc that the

manifest is illed out correctly. Licensees must also conduct a quality control

program to assure compliance with the waste classification and waste stability

requirements.

NRC staff recognizes that the new requirements may result in some modifications

to existing licensee waste management practices, and furthermore believes that

it will be useful to licensees to begin planning for implementation of the new

requirements in advance of the December 27, 1983 effective date. At this time

NRC staff is preparing Regulatory Guides on both waste classification and waste

form. To provide immediate guidance to licensees, however, the NRC Low-Level

Waste Licensing Branch has prepared technical position papers on waste classification and waste form.

The waste classification technical position paper describes overall procedures

acceptable to NRC staff which may be used by licensees to determine the presence

and concentrations of the radionuclides listed in 61.55, and thereby classifying waste for near-surface disposal. This technical position paper also provides

guidance on the types of information which should be included in shipment manifests accompanying waste shipments to near-surface disposal facilities.
Commission Licensees 2

The technical position paper on waste form provides guidance to waste generators

on test methods and results acceptable to NRC staff for implementing the 10 CFR

Part 61 waste form requirements. It can be used as an acceptable approach for

demonstrating compliance with the 10 CFR Part 61 waste structural stability

criteria. This technical position paper includes guidance on processing waste

into an acceptable stable form, designing acceptable high-integrity containers,

packaging cartridge filters, and minimizing radiation effects on organic ionexchange resins.

The guidance in the waste form technical position paper may be used by licensees

as the basis for qualifying process control programs to meet the waste form stability requirements, including tests which can be used to demonstrate resistance

to degradation arising from the effects of compression, moisture, microbial activity, radiation, and chemical changes. Generic test data (e.g., topical reports

prepared by vendors who market solidification technology) may be used for process

control program qualification where such generic data is applicable to the

particular types of waste generated by a licensee.

While the NRC staff has not formally reviewed or aproved any products, NRC staff

believes that solidification processes and high-integrity containers currently

exist that can be qualified to meet the waste form stability requirements.

Licensees and vendors should continue their efforts to have qualified products

available in advance of the December 27, 1983 implementation deadline. NRC staff

will continue to work with licensees and vendors to meet the waste form requirements and implementation deadline. NRC staff will also continue to coordinate

their work with cognizant representatives of States which currently have licensed

low-level waste disposal sites.

Draft versions of both technical position papers have previously been made available to interested members of the public. Comments received on these drafts have

been considered during development of the technical position papers being published at this time. Further public comment on these technical positions is

welcomed, and any such comment received will be considered during preparation of

the waste classification and waste form Regulatory Guides. Comments on the technical position papers may be forwarded to myself (Address: U.S. Nuclear Regulatory Commission, Washington, DC, 20555, mailstop SS-623). Questions on the technical position papers may be referred to Mr. Paul H. Lohaus (301-427-4500), to

Mr. G. W. Roles (301-427-4593), or to Mr. Timothy C. Johnson (301-427-4697) of my

staff.

The waste classification and waste form technical position papers are included

as attachments to this letter. The information collections contained in these

technical positions have been approved under OMS number 3150-0014.

Leo B. Higg o am. Chief

Low-Level Waste Licensing Branch

Division of Waste Management
May 1983

Rev. 

LOW-LEVEL WASTE LICENSING BRANCH

TCHNICAL PSITION ON

RADIOACTIVE WASTE CLASSIFCATION

A. Introduction

Section 20.311 ("Transfer for isposal and Manifests") of 10 CFR Part 20

("Standards for Protection Against Radiation") requires that any licensee who

transfers radioactive waste to a land disposal facility or to a licensed waste

collector or processor must classify the waste according to Section 61.55

("Waste Classification") of 10 CFR Part 61 ("Licensing Requirements for Land

Disposal of Radioactive Waste"). Section 20-311 also requires that any licensed

waste processor who treats or repackages radioactive waste for disposal into a

land disposal facility also classify their waste according to Section 61.55.

Section 61.55 defines radioactive waste suitable for land disposal as falling

into one of three categories--i.e., Class A waste, Class B waste, and Class C

waste. Wastes are determined to fall into one of the classes by comparison to

limiting concentrations of particular radionuclides which are set forth in

Table 1 and Table 2 of Section 61.55. Wastes determined to fall into one of

the classes must be labeled as such in accordance with Section 61.57 ("Labeling").

Waste generators and waste processors must record on shipment manifests a

description of the transferred waste, and must also carry out a quality control

program to assure that classification of waste is carried out in a proper manner.

All three classes of waste are required to meet certain minimum

requirements as set forth in paragraph 61.56(a) of Section 61.56 ("Waste

Characteristics") which are intended to facilitate handling of waste at the

disposal site and provide protection of public health and safety. Class B and

Class C wastes, however, are required to meet more rigorous requirements on

waste stability. These stability requirements are set forth in paragraph

61.56(b) of Section 61.56. Class C waste must be also identified to allow for

additional disposal procedures to be carried out at the disposal site to provide

protection to a potential inadvertent intruder. Finally, wastes having concentrations of particular radionuclides exceeding those allowed for Class C waste

are generally considered unacceptable for near-surface disposal.

This technical position describes overall procedures acceptable to the

regulatory staff which may be used by licensees to determine the presence and

concentrations of radionuclides listed in Section 61.55, and thereby classifying

waste for near-surface disposal. The technical position also provides guidance

and clarification on the minimum types of information which should be included

on shipment manifests.

B. Discussion

Each shipment of radioactive waste by a waste generator to a licensed

collector, processor, or operator of a land disposal facility must ensure that

a shipment manifest accompany the waste. Section 20.311 states that the manifest must include information on waste characteristics including (as a minimum)
a physical description of the waste, the volume, the radionuclide identity and

quantity, the total radioactivity, the principal chemical form, the solidification agent (if any), and the waste class. As a minimum, the total quantity of

the radionuclides H-3, C-14, Tc-99, and I-129 must be listed. These radionuclides, as well as the other radionuclides listed in Section 61.55, are used

to determine the classification status of radioactive waste. Controlled disposal of wastes containing these radionuclides is considered important in

assuring that the performance objectives of Subpart C of the Part 61 regulation

are met. The manifest must also identify waste containing more than 0.1 percent

by weight chelating agents, as well as provide an estimate of the weight percentage of the chelating agent. Additional information may be required for

shipment to a particular disposal facility depending upon facility-specific

license conditions.

To classify waste for disposal and fill out shioment manifests, a licensee

must make two basic determinations: (1) whether the waste i~ ?:ceptable for

near-surface disposal, and (2) if acceptable for near-surface disposal, whether

the waste is classified as Class A, Class B, or Class C waste. Another determination is whether the waste complies with any additional waste form, package,

or content requirement which may be in place at the particular disposal facility

to which the waste is to be shipped.

Waste is determined to be generally unacceptable for near-surface disposal

if it contains any of the radionuclides listed in Tables 1 and 2 of Section

61.55 in concentrations exceeding the limits established for the radionuclides.

If determined to be acceptable for near-surface disposal, waste is determined

to be Class A, Class B, or Class C based upon the lists of radionuclide concentration limits set forth in Tables 1 and 2.

C. Regulatory Position

All licensees must carry out a compliance program to assure proper

classification of waste. Licensee programs to determine radionuclide concentrations and waste classes may, depending upon the particular operations at the

licensee's facility, range from simple programs to more complex ones. In

general, more sophisticated programs would be required for licensees generating

Class or Class C waste, for licensees generating waste for which minor process

variations may cause a change in classification, or for licensees generating

waste for which there is a reasonable possibility of the waste containing

concentrations of radionuclides which exceed limiting concentration limits for

near-surface disposal.

The regulatory staff is prepared to be flexible in the adaptation of a

particular program to a particular waste generating facility. A principal

consideration for the acceptability of a particular program will be whether a

reasonable effort has been made to ensure a realistic representation of the

distribution of radionuclides within waste, given physical limitations, and to

classify waste in a consistent manner. Example "physical limitations" can

include difficulties in obtaining and measuring representative samples at

reasonable costs and acceptable occupational exposures. The staff considers a

reasonable target for determining measured or inferred radionuclide concentrations is that the concentrations are accurate to within a factor of 10. The

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staff recognizes, however, that this target may be difficult to achieve for

some waste types and forms.

A licensee's program should be specific to a particular facility, and

should consider the different radiological and other characteristics of the

different waste streams generated by the facility. There are at least four

basic methods which may be potentially used either individually or in combination by licensees:

- materials accountability;

- classification by source;

- gross radioactivity measurements; or - direct measurement of individual radionuclides.

The following discussion outlines instances and conditions whereby each basic

method or combinations thereof would be acceptable to regulatory staff as a

program for demonstrating compliance with the waste classification requirement.

Some licensees, such as nuclear power facilities, are expected to employ a

combination of methods. Appendix A to this technical position outlines an

example program for nuclear power facilities which the regulatory staff would

find acceptable.

1. General Criteria

a. Compliance Through Materials Accountability

One method which the regulatory staff would find acceptable to determine

radionuclide concentrations and demonstrate compliance with the waste classification requirements is through a program of materials accountability. That is,

a given quantity (and resulting concentration) of radioactive material may be

known to be contained within a given waste or may be inferred through determining the difference between the quantity of radioactive material entering and

exiting a given process. This procedure is expected to be most useful for

licensees who receive and possess only a limited number of different radioisotopes in known concentrations and activities (e.g., holders of source

material, special nuclear material, or specific byproduct material licenses).

An example use of this method would be at a biomedical research facility

at which known amounts of a radioisotope are injected into research animals,

the carcasses of which are ultimately disposed of as radioactive waste. Another

example would be a research or test facility performing activation analysis

experiments. In this case, the quantity of radioactive material within a given

waste stream may be inferred through calculation. A third example would be a

power plant in which the radionuclide content of a particular process vessel

(e.g., a resin bed) is determined on the basis of measurements of influent and

effluent streams.

This method may also be used to determine the absence of particular

radionuclides. That is, for most licensees, the absence of particular radionuclide may be determined through a knowledge of the types of radioisotopes

received and possessed, as well as the process producing the waste. For example.

if a licensee receives, possesses, and uses only tritium, there is no need to

measure the waste stream for other isotopes such as iodine-129 or cesium-13 7.

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b. Classification by Source

This method is similar to the above method of materials accountability and

involves determining the radionuclide content and classification of waste

through knowledge and control of the source of the waste. This method is

expected to be useful for occasions when the radionuclide concentrations within

waste generated by a particular process are relatively constant and unaffected

by minor variations in the process.

This method is also expected to be frequently useful for determining the

absense of particular radionuclides from a given waste stream. For example,

within a given licensed facility there may be a number of separate controlled

areas within which only a limited number of radioisotopes are possessed and

used (e.g., Cs-137 may be used in one area and tritium in another). As long

as facility operations are conducted so that transfer of radioactive material

from one controlled area to another cannot occur, waste generated from a

particular area may be readily classified by source. An example of a licensee

for which this method is expected to be useful is a large university which

holds a broad license for byproduct material.

c. Gross Radioactivity Measurements

Measurement of gross radioactivity is an acceptable method for all classes

of waste provided that:

- the gross radioactivity measurements are correlated on a consistent

basis with the distribution of radionuclides within the particular

waste stream analyzed, and

- the radionuclide distributions are initially determined and

periodically verified by direct measurement techniques.

Licensees carrying out gross radioactivity measurements to assure

compliance with the waste classification requirement must establish a program

to correlate and calibrate measured radioactivity levels with radionuclide

concentrations in wastes prepared for shipment. This program must at a minimum

take into account waste package and detector geometry, shielding and attenuaticeffects, the effective gamma energies of the emitted photons, and the number or

photons per decay. The accuracy of the correlation must be initially established by and periodically checked through more detailed sample analysis

involving measurement of specific radionuclides. The accuracy of the correlation should also be checked whenever there is reason to believe that process

changes may have significantly altered previously determined correlations.

d. Measurement of Specific Radionuclides

Another method acceptable to the regulatory staff for determining

radionuclide concentrations in waste is by direct measurement. In using this

method, licensees may directly measure individual radionuclides or may establish an inferential measurement program whereby concentrations of radioisotopes

which cannot be readily measured (through techniques such as gamma-spectral

analysis) are projected trough ratioing to concentrations of radioisotopes

which can be readily measured. An example would be the practice of scaling

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transuranic concentrations to concentrations of the isotope Ce-144. Scaling

factors should be developed on a facility and waste stream specific basis, and

should be initially determined and periodically confirmed through direct

measurements.

2. Determination and Verification of Radionuclide Concentrations and

Correlations

Radionuclide concentrations should be determined based upon the volume or

weight of the final waste form. Samples may be taken for analysis either from

the final waste form or from the waste prior to processing into a final waste

form (e.g., from any intermediate process step). Samples taken prior to final

processing should enable the results of the sample analysis to be directly

translated to the final waste form.

The lower limit of detection of a asurement technique for direct measurement of a particular radionuclide should be no more than 0.01 times the concentration for that radionuclide listed in Table 1, and 0.01 times the smallest

concentration for that radionuclide listed in Table 2. For this technical

position, the lower limit of detection for a particular measurement may be

assumed by licensees to be consistent with the definitions for lower limit of

detection (LLD) provided in references 1 and 2.

The radionuclides listed in Section 61.55 may be roughly organized into

two groups: (1) those that are amenable to routine quantification by direct

measurement techniques (e.g., gamma spectral analysis of isotopes such as Co-60

or Cs-137), and (2) those that require more costly and time consuming analysis

frequently removed from the waste generator's facility (e.g., alpha/beta

analysis).

For the first group of radionuclides, measurements to identify and quantify

specific radionuclides within final waste forms should be performed:

- routinely for Class 8 and Class C wastes; and

- routinely for wastes for which minor process variations could cause

an upward change in wast. lassification.

In this case, "routine" measurements would involve a limited number (e.g., one

or two) of samples out of a batch. If radionuclide distributions are shown to

be reasonably consistent from one batch to the next, however, consideration

will be given to decreasing the frequency of routine measurements.

A more detailed analysis for the second group of radionuclides should be

performed:

- periodically to confirm the correlation of measurements made from

gross radioactivity measurements; and

- whenever there is reason to believe that facility or process changes

may have significantly altered (e.g., by a factor of 10) previously

determined correlations of gross radioactivity measurements.

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The staff believes that for most facilities and for most Class and C waste

types, this confirmatory analysis should be performed on at least an annual

basis. Confirmatory analyses for Class A wastes should be performed on at

least a biannual basis. However, these frequencies may be raised or lowered

based upon consideration of particular facility, waste stream, or radionuclide

characteristics. Factors which would influence this consideration include the

frequency of process vessel changeout or waste shipment, the difficulty (e.g.,

costs, occupational exposures) in obtaining a representative sample of a

particular waste stream, the variability of the radionuclide distribution

within the waste stream over time, and the availability of analytical capability

for particular radionuclides.

It is recognized that it is sometimes difficult to obtain a truly

representative sample of some waste streams and that some judgment will be

necessary to determine sampling adequacy and representative radionuclide

distributions. One example could include Class A dry active waste such as

miscellaneous trash. In this case, an estimate of the radionuclide distribution

within the waste could be made based upon distributions determined from other

waste streams associated with generation of the trash waste. Alternatively,

radionuclide distributions could be potentially estimated from smear samples

obtained from locations in which the waste is generated. Another example could

include activated metals. In many cases, radionuclide concentrations within

activated metals will be difficult to directly measure, and NRC staff will in

such cases accept estimates based on consideration of activation analysis

calculations for similar material types.

3. Concentration Volumes and Masses

Paragraph 61.55(a)(8) states that the concentration of a radionuclide may

be averaged over the volume of the waste, or the weight of the waste if the

concentration units are expressed as nanocuries per gram. This requirement

needs to be interpreted in terms of the variety of different types and forms

of low-level waste. Principal considerations include: () whether the distribution of radionuclides within the waste can be considered to be reasonably

homogeneous, and (2) whether the volume of the waste container is significantly

larger than the volume of the waste itself, and the differential volume consists

largely of void space.

Most waste streams may be considered to be homogeneous for purposes of

waste classification. Such waste streams would include, for example, spent

ion-exchange resins, filter media, solidified liquids, or contaminated dirt.

Contaminated trash waste streams, which are composed of a variety of miscellaneous materials, may be considered homogeneous for purposes of waste classification when placed and compacted within shipping containers. The activity of

small concentrated sources within the trash, such as small check sources or

gauges, may be generally averaged over the trash volume.

In many cases the volume used for waste classification purposes may be

considered to correspond to the volume of the waste container. This would be

the case, for example, for trash waste streams which are compacted into shipping

containers. The waste classification volume of large unpackaged components

such as contaminated pumps, heat exchangers, or other machinery may be taken to

be the overall volume of the component.

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Care needs to be taken, however, to differentiate between the volume of

the waste form and the volume of the waste container if the latter is signifi- cantly larger (e.g., greater than 10%). For example, for wastes such as ion-exchange resins or filter media contained within a disposable demineralizer

or liner, the volume used for waste classification should be the volume of the contained waste rather than the gross volume of the container. Waste classi- fication volumes of cartridge filters stabilized by emplacement within high integrity containers should be determined as calculated over the volume of the cartridge filter itself rather than the gross volume of the container. Simi- larly, the volume and mass considered for purposes of waste classification of dewatered ion-exchange resins and filter media placed into high integrity containers should be the volume and mass of the contained waste. Classification of absorbed liquids should be based on the volume and mass of the liquids prior to absorption.

An exception to the above would be a situation in which a particular waste

type is stabilized within a waste container using a solidification media. For example, assume that a cartridge filter or large sealed source is solidified with a 55-gallon drum using a binder such as cement or bitumen. In this case, the waste and binder forms a solid mass within the container and the waste classification volume may be considered to be the volume of the solidified mass. Similarly, classification of solidified liquids would be based on the volume and mass of the solidified waste mass.

4. Reporting on Manifests

Section 20.311 of 10 CFR Part 20 requires that each shipment of radioactive

waste to a land disposal facility be accompanied by a manifest which describes the shipment contents. This manifest may be shipping papers used to meet regulations promulgated by the Oepartment of Transportation or the Environmental

Protection Agency, provided that the information required by Section 20.311 is included. The waste shipment receiver (e.g., the disposal facility operator)

may also require specific additional information. In addition to shipper identification requirements and a certification, the manifests required by Section 20.311 must include the following information as a minimum:

- the waste class; - a radiological description; and - a physical and chemical description.

Waste class

Identifying the waste class of-the shipped waste is required, since certain disposal requirements are imposed for each waste class and the waste disposal facility operator must be able to identify the waste in order to carry out these disposal requirements. The individual waste containers must be labeled as being Class A, Class B, or Class C, and the waste class of each container must also be indicated on the manifest. The format of the shipment labels (or markings) is at the discretion of the disposal facility operator. Unpackaged Class A waste (eq.. bulk shipments of contaminated dirt) do not need to be labeled provided tat the waste class is recorded on the manifest.

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The shipment manifest should also record the date for which the

classification determination is valid. This can be, for example, the date of

transfer of the waste package from the site of generation to the disposal site.

In no case should the date chosen for decay correction be beyond the date on

which the waste is transported to the disposal site.

Radiological description

The requirements in Section 20.311 include a general requirement to list

radionuclide identities and quantities, a general requirement to list the total

radioactivity in the waste, and a specific requirement to list four individual

radionuclides: H-3, C-14, Tc-99, and 1-129. These requirements need some

further guidance, however, since a wide range of radionuclides over a wide

range of concentrations may be contained in a particular waste package.

The regulatory staff has determined three criteria for determining specific

radionuclides which should be listed in manifests:

1. any radionuclide specifically required to be listed by Section 20.311,

or by license conditions at the disposal facility to which the waste

is shipped;

2. any radionuclide which is listed in Section 61.55 and forms a

significant part of the total activity which determines the waste

class; and

3. any radionuclide which is contained in significant quantities within

a waste container or shipment.

Currently, only the isotopes H-3, C-14, Tc-99, and I-129 are required in

Section 20.311 to be specifically identified and their quantities listed in

manifests. In the manifests, if a particular one of these four radioisotopes

is known to be not present within a waste stream (e.g., through material

accountability), the quantity of the radionuclide should be recorded as "not

present." If the radionuclide is determined through material accountability,

direct measurement, or inference through direct measurement or gross radioactivity measurement, this quantity should be reported as determined. If the

radionuclide is known or suspected to be contained within the waste but is in

quantities less than the lower limit of detection for the analyzed sample, the

quantity of the radionuclide should be recorded as being less than the minimum

detectable, with the minimum detectable amount included alongside in parentheses.

The total quantities of these four nuclides may be reported on a waste shipment

rather than an individual waste container basis. In the case of Tc-99, care

should be taken to distinguish between this nuclide and its short-lived precursor.

Tc-99m.

Other radionuclides listed in Section 61.55 should be specifically

identified and the quantities reported if they are significant for purposes of

classification. A radionuclide shall be determined to be "significant for

purposes of classification" if it is contained in waste in concentrations

greater than 0.01 times the concentration of that nuclide listed in Table 1 or

0.01 times the smallest concentration of that nuclide listed in Table 2. This

criterion does not include isotopes identified in Table 2 as having half-lives
less than 5 years. An isotope (other than Cm-242) having a half-life less than

5 years is considered significant for the purposes of waste classification if

it is contained in the waste in concentrations greater than 7 Ci/cm3 (0.01

times the Table 2, Column 1 value).

Radionuclides not listed in Section 61.55 should also be specifically

identified and the quantities reported if they are contained in significant

quantities within a waste container or shipment. In general, a radionuclide

shall be deemed to be "contained in significant quantity" if it is in concentrations greater than 7 Ci/cm3. In addition, the total quantity of source or

special nuclear material should be reported, if the waste contains such material.

Otherwise, radionuclides should be listed in shipment manifests in

compliance with Oepartment of Transportation requirements in 49 CFR Part 172,

Section 172.203.

Physical and chemical description

Items to be included in the physical and chemical description include, as

a minimum, the following: a physical description of the waste; the volume; the

principal chemical form; and the solidification agent used (if any). Waste

containing more than 0.1% chelating agents by weight must be identified, and

the weight percentage of the chelating agent estimated. Amplification of NRC's

intent regarding these requirements is provided below.

A physical description of the waste is needed in order to facilitate safe

handling at the disposal facility and to better predict long-term environmental

impacts. The description need only be a few words but should be as specific as

possible. For example, a description such as "solidified resins" or "solidified

evaporator bottoms" should be used rather than the. description "solidified

radwaste." Similarly, the description "scintillation vials" is preferable to

the description "laboratory waste."

The volume listed in the manifest should be the volume of the waste

container, if any, or the volume of the waste itself if shipped unpackaged

(e.g., a bulk quantity of contaminated soil).

The principal chemical form of the waste also needs to be provided as an

aid to waste handling safety and to improve prediction of long-term environmental impacts. This should be the principal chemical form in which the

radioactivity is contained (e.g., calcium fluoride, toluene, etc.). There is

no need to list trace chemical contaminants, however.

The solidification agent need only be provided in general terms (e.g.,

cement, asphalt, vinyl ester styrene). The type of solidification agent used

may be combined with the physical description of the waste (e.g., "resins

solidified in cement").

The intent of the requirement to identify waste containing chelating

agents in quantities greater than 0.1% is to identify waste containing large

quantities of such agents. Large quantities of such agents may be segregated

from other waste at a disposal facility and/or disposed through some special

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disposal method. Disposal facility operators need to have such waste identified in order for them to perform these additional disposal operations. For

purposes of this requirement, chelating agents include the following: amine

polycarboxylic acids (e.g., EDTA, TPA), hydroxy-carboxylic acids, and

polycarboxylic acids (e.g., citric acid, carbolic acid, and glucinic acid).

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APPENDIX A

GENERAL PROGRAM FOR CLASSIFYING WASTES

AT NUCLEAR POWER FACILITI-ES

In order to meet the requirements in 10 CFR Part 61 to classify radioactive

wastes at nuclear power plants, NRC staff has prepared a general program for

implementing the waste classification system. This implementation program

consists of a three-tiered approach which includes:

1. periodic analysis for all nuclides listed in Table 1 of Section 61.55;

?. gamma spectroscopy of certain nuclides from which waste classification

nuclides are correlated; and

3. gross radioactivity measurements which correlate activity levels of

wastes from similar batches to the gamma-spectroscopy measurements.

The periodic sampling for listed nuclides would be performed on various

waste streams in the plant. These periodic analyses should be the basis for

establishing correlation factors between the waste classification nuclides and

nuclides which can be more easily measured using gamma spectroscopy techniques.

Samples should be taken nominally on an annual basis from individual waste

streams such as boric acid evaporator bottoms, primary system cleanup resins,

chemical regenerative evaporator bottoms, etc., which are likely to be Class 

or C wastes. If unit operations or plant conditions are modified such that the

radionuclide distribution for any of the individual waste streams changes by a

factor of 10, a reanalysis should be performed. Plant operational changes would

include changes in the failed fuel fraction or a crud burst. If operations

remain consistent, consideration can be given to performing reanalysis on a

less frequent basis. In addition, consideration should be given to increasing

the frequency of analysis depending upon individual facility, waste stream, and

radionuclide characteristics. Factors which would influence this consideration

include the frequency of process vessel changeout or waste shipment, the difficulty in obtaining a representative sample of a particular waste stream, the

variability of the radionuclide distribution within the waste stream, and the

available analytical capacity for particular radionuclides.

The gamma spectroscopy measurements should be performed on a limited

number of samples obtained from individual waste batches. This can be performed

by analyzing waste samples prior to or after volume reduction and/or solidification, analyzing waste drums or liners by any of the commercial devices

designed for this task, or by analyzing influent and effluent samples from the

process stream. Other methods which provide reasonable analysis will also be

considered. Efforts should be made to obtain reasonably representative samples

for analysis. The results of the gamma spectroscopy measurements should be

applied with the correlation factors to obtain concentrations for those nuclides

listed in the waste classification table.

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Gross radioactivity measurements may also be performed on individual waste

packages from similar waste batches for which gamma-spectroscopy results are

available. Gross radioactivity measurements should include corrections for

attenuation and container size and configuration. The gross activity measurements should be used to scale the nuclide concentrations obtained from the

gamma spectroscopy data and correlations.

For Class A wastes such as contaminated trash, gross radioactivity

measurements may be performed as the basis for waste classification provided

that these measurements can be correlated to the concentrations of the radionuclides listed in Section 61.55. Confirmatory reanalysis of the correlation

factors should be performed on at least a biannual basis.

The NRC staff believes that the above approach presents a workable and

enforceable program for implementing the waste classification system. This

approach should minimize the administrative and operational burdens on plant

personnel, but still provide reasonably accurate data for use in quantifying

disposal site nuclide concentrations and inventories.

12
REFERENCES

(1) U.S. Nuclear Regulatory Commission, "Radiological Effluent Technical

Specifications for PWR's," NUREG-0472 (as revised), July 1979.

(2) U.S. Nuclear Regulatory Commission, "Radiological Effluent Technical

Specifications for WR's," NUREG-0473 (as revised), July 1979.

13
May 1983

Rev. 0

Technical Position on

Waste Frn

A. Introduction

The regulation, "Licensiny Requirements for Land Disposal of Radioactive

Waste," 10 CFR Part 61, establishes a waste classification system based

on the radionuclide concentrations in the wastes. Class B and C waste

are required to be stabilized. Class A waste have lower concentrations,

and may be segregatea without stabilization. Class A wastes may also be

stabilized and disposed of with the Class B wastes. All Class A liquid

wastes, however, require solidification or absorption to meet the free

liquid requirements. Structural stability is intended to ensure that the

waste does not degrade and promote slumping, collapse, or other failure

of the cap or cover over the disposal trench and thereby lead to water

infiltration. Stability is also a factor in limiting exposure to an

inadvertent intruder since it provides greater assurance that the waste

form will be recognizable and nondispersable during its hazardous

lifetime. Structural stability of a waste form can be provided by the

waste form itself (as with lrge activated stainless steel components),

by processing the waste to a stable form (e.g., solidification), or by

empiacing the waste in a container or structure that provides stability

(e.g., high integrity container).

This technical position on waste form has been developed to provide

guidance to both fuel-cycle and non-fuel-cycle waste generators on waste

form test methods and results acceptable to the NRC staff for

implementing the 10 CFR Part 61 waste form requirements. It can be used

as an acceptable approach for demonstrating compliance with the 10 CFR

Part 61 waste stability criteria. This position includes guidance on the

processing of wastes into an acceptable, stable waste form, the design of

acceptable high integrity containers, the packaging of filter cartridges,

and minimizing the radiation effects on organic ion-exchange resins.

It is the intent of the RC staff to add other guidance on waste form in

additional technical positions as is necessary to address other pertinent

waste form issues.
2

B. Background

Historically, waste form and container properties were considered of

secondary importance to good site selection; the combination of a

properly operated site having good geologic and hydrologic

characteristics were considered the only barriers necessary to isolate

low-level radioactive wastes from the environment. Experience in

operating low-level waste disposal sites indicated that the waste form

should play a major role in the overall plan for managing these wastes.

The regulation for near-surface dispcsal of radioactive wastes, 10 CFR

Part 6i, includes requirements which must be met by a waste form to be

acceptable for near-surface disposal. The regulation includes a waste

classification system which divides waste into three general classes: A,

B, dnd C.

The classification system is based on the overall disposal hazards of the

wastes. Certain minimum requirements must be met by all wastes. These

minimum requirements are presented in Section 61.56(a) and involve basic

packaging criteria, prohibitions against the disposal of pyrophoric,

explosive, toxic and infectious materials, and requirements to solidify

or aDsorb liquids.

In addition to the minimum requirements, Class and C wastes are

required to have stability. As defined in Section 61.56(b) of the rule,

stability requires that the waste form maintain its structural integrity

under the expected disposal conditions. Structural stability is

necessary to inhibit slumping, collapse, or other failure of the disposal

trench resulting from degraded wastes which could lead to water

infiltration, radionuclide migration, and costly remedial care programs.

Stability is also considered in the intruder pathways where t is assumed

that after the active control period wastes are recognizable and,

therefore, continued inadvertent intrusion is unlikely. To the extent

practical Class B and C waste forms should maintain gross physical

properties and identity over a 300 year period.

In order to ensure that Class B and C waste or its container will

maintain its stability, the following conditions need to be met:

a. The waste should be a solid form or in a container or structure

that provides stability after disposal.
3

b. The waste should not contain free standing and corrosive

liquids. That is, the wastes should contain only trace amounts

of drainable liquid, and in no case may the volume of free

liquid exceed one percent of the waste volume when wastes are

disposed of in containers designed to provide stability, or 0.5

percent of the waste volume for solidified wastes.

c. The waste or container should be resistant to degradation

caused by radiation effects.

d. The wdste or container should be resistant to biodegradation.

e. The waste or container should remain stable under the

compressive lodds innerent in the disposal environment.

f. The waste or container should remain stable if exposed to

moisture or water after disposal.

g. The ds-generated waste should be compatible with the

solidification media or container.

A large portion of the waste produced in the nuclear industry is in a

form which is either liquid or in a wet solid form (e.g., resins, filter

sludge., etc.) and requires processing to achieve an acceptable solid,

monolithic form for burial. The liquid wastes, irregardless of its

classification, re required to be either absorbed or solidified. In

order to assure that the solidification process will consistently produce

a product which is acceptable for disposal and will meet disposal site

license conditions a process control program should be used. General

requirements for process control programs are provided in the NRC

Standard Reyiew Plan 11.4, "Solid Waste Management Systems,"

(NUREG-0800 ) ana its accompanying Branch Technical Position ETSB 11-3,

"Design Guidance for Solia Waste Management Systems Installed in

Light-Water-Cooled Nuclear Power Reactor Plants," (revised in July 1981).

These documents may also be used as the basis for individual

solidification process control programs by other fuel-cycle and by

non-fuel-cycle waste generators who would solidify wastes. The guidance

in this technical position should be the basis for qualifying process

control programs for Class B nd C wastes. The use of applicable generic

test data (e.g., topical eporti) may be used for process control program

qualification. Process cntrol programs for solidified Class A waste

products, which re sesrejdted trom Class B and C wastes, need only
4

demonstrate that the product is a free standing monolith with no more

than 0.5 percent of the waste volume as free liquid.

An alternative to processing some Class B and C waste streams,

particularly ion exchange resins and filter sludges, is the use of a high

integrity container. The high integrity container would be used to

provide the long-term stability required to meet the stability

requirements in 10 CFR Part 61. The design of the high integrity

container should be based on its specific intended use in order to ensure

that the waste contents, as well as interim storage and ultimate disposal

environments, will not compromise its integrity over the long-term. As

with waste solidification, d process control program for dewatering wet

solids should be developed and utilized to ensure that the free liquid

requirements in 10 CFR Part 61 are being met.

C. Regulatory Position

1. Solidified Class A Waste Products

a. Solidified Class A waste products which are segregated from

Class B and C wastes should be free standing monoliths and have

no more than 0.5 percent of the waste volume as ree liquids as

measured using the method described in ANS 55.1.

b. Solidified Class A waste products which are not segregated from

Class B and C wastes should meet the stability guidance for

Class B and C wastes provided below.

2. Stability Guidance for Processed (i.e., Solidified) Class B and C

Wastes

a. The stability guidance in this technical position for processed

wastes should be implemented through the qualification of the

individual licensee's process control program. Generic test

data may be used for qualifying process control programs.

Through the use of a well designed and implemented process

control program, frequent requalification to demonstrate

stability is expected to be unnecessary. However, process

control programs should include provisions to periodically

demonstrate that the solidificdtion system is functioning

properly and waste products continue to meet the 10 CFR Part 61

stability requirements. Waste specimens should be prepared
5

based on the proposed waste streams to be solidified and based

on the range of waste stream chemistries expected. The tests

identified may be performed on radioactive or non-radioactive

samples.

b. Solidified waste specimens should have compressive strengths of

at least 50 psi when tested in accordance with ASTM C39 .

Compressive strength tests for bituminous products should be

performed in accordance with ASTM 1074 .

Many solidification agents will be easily capable of meeting

the 0 psi limit for properly solidified wastes. For these

cases, process control parameters should be developed to

achieve te maximum practical compressive strengths, not simply

to achieve the minimum acceptable compressive strength.

c. The specimens for each proposed waste stream formulation should

remain stable after being exposed in a radiation field

equivalent to the maximum level of exposure expected from the

proposed wastes to be solidified. Specimens for each proposes

waste stream formulation should be exposed to a minimum of 10

Rdds in a gamma irradiator or equivalent. If the maximum level

of exposure is expected to exceed 10 Rads, testing should be

performed at the expected maximum accumulated dose. The

irradiated specimens should have a minimum compressive strength

of 50 psi following irradiation as tested in accordance with

ASTM C39 or ASTM 01074.

d. Specimens for each proposed waste stream formulation should be

tested fr resistance o biodegradation in accordance with both

ASTM G21 and ASTM G22 . No indication of culture growth should

be visible. Specimens should be suitable for compression

testing in accordance with ASTM C39 or ASTM 01074. Following

the biodegradation testing, specimens should have compressive

strengths greater thdn 50 psi as tested using ASTM C39 or ASTM

D1074.

For polymeric or bitumen products, some visible culture growth

from contamination, additives or biodegradable components on

the specimen surface which do not relate to overall substrate

integrity my be present. For these cases, additional testing

should be performed. If culture growth is observed upon
6

completion of the biodegradation test for polymeric or bitumen

products, remove the test specimens from the culture, wash them

free of dll culture and growth with water and only light

scrubbing. An organic solvent compatible with the substrate

may be used to extract surface contaminants. Air dry the

specimen at room temperature ana repeat the test. Specimens

should have observed culture growths rated no greater than 1 in

the repeated ASTM G21 test, and compressive strengths greater

than 50 psi. The specimens should have no observed growth in

the repeated ASTM G22 test, and a compressive strength greater

than 50 psi. Compression esting should be performed in

accordance with ASTM j or ASTM 01074.

If growth is observed following the extraction procedure,

longer term testing of at least six months should be performed

to determine biodegradation rates. The Bartha-Pramer Method

is acceptdble for this testing. Soils used should be

representative of those at burial grounds. Biodegradation

extrapolated or full-size waste forms to 300 years should

produce less than a 10 percent loss of the total carbon in the

waste form.

e. Leach testing should be performed for a mnimum of 90 days in

accordance with the procedure in ANS 16.1 . Specimen sizes

should be consistent with the samples prepared for the ASTM C39

or ASTM 01074 compressive strength tests. In addition to the

demineralized water test specified in ANS 16.1, additional

testing using other leachants specified in ANS 16.1 should also

be performed to confirm the solidification agents leach

resistance in other leachant media. It is preferred that the

synthesized sea water leachant also be tested. In addition, it

is preferable that radioactive tracers be utilized in

performing the leach tests. The leachability index, as

calculated in accordance with ANS 16.1, should be greater than

6.

f. Waste specimens should maintain a minimum compressive strength

of 50 psi as tested using ASTM C39 or ASTM D1074, following

immersion for a minimum period of 90 days. Immersion testing

may be performed in conjunction with the leach testing.

g. Waste specimens should e resistant to thermal degradation.

The heating ana cooling chambers used for the thermal
7

degradation testing should conform to the description given in

ASTM B553, Section 3. Samples suitable for performing

compressive strength tests in accordance with ASTM C39 or ASTM

01074 should be used. Samples should be placed in the test

chamber an a series of 30 thermal cycles carried out in

accordance witn Section 5.4.1 through 5.4.4 of ASTM 8553. The

high temperature limit should be 60C and the low temperature

limit -40C. Following testing the waste specimens should nave

compressive strengths greater than 50 psi as tested using ASTM

C39 or ASTM 1074.

h. Waste specimens should have less than 0.5 percent by volume of

the waste specimen as free liquids as measured using the method

described in AS 55.1. Free liquids should have a pH between 4

and 11.

i. If small, simulated laboratory size specimens are used for the

above testing, test data from sections or cores of the

anticipated full-scale products should be obtained to correlate

the characteristics of actual size products with those of

simulated laboratory size specimens. This testing may be

performed on non-radioactive specimens. The full-scale

specimens should be fabricated using ctual or comparable

solidification equipment.

j. Waste samples from full-scale specimens should be destructively

analyzed to ensure that the product produced is homogeneous to

the extent that all regions in the product can expect to have

compressive strengths of at least 50 psi. Full-scale specimens

may be fabricated using simulated non-radioactive products, but

should be fabricated using actual solidification equipment.

3. Radiation Stability of Organic Ion-Exchdnge Resins

In order to ensure that organic ion exchange resins will not produce

adverse radiation degradation effects, resins should got be generated

that have loadings which will produce greater than 10 Rads total

accumulated dose. For Cs-137 and Sr-90 a total 3accumulated dose of 10

Rads is approximately equivalent to an 10 Ci/ft concentration. This

position is applicable to resins in the unsolidified, as-generated form.

In the event that the waste generator considers it necessary to load

resins higher than LO Rads, it should be demonstrated that the specific
8

resin will not undergo radiation degradation at the proposed higher

loading. The test method should adequately simulate the chemical and

racioloyic conditions expected. A gamma irradiator or equivalent should

be utilized for these tests. There should be no adverse swelling, acid

formation or gas generation which will be detrimental to the proposed

final waste product.

4. High Integrity Containers

a. The maximum allowable free liquid in a high integrity container

should be less than one percent of the waste volume as measured

using the method described in ANS 55.1. A process control

program should be developed and qualified to ensure that the

free liquid requirements in 10 CFR Part 61 will be met upon

delivery of the wet solid material to the disposal facility.

This process control program qualification should consider the

effects of transportation on the amount of drainable liquid

which might e present.

b. High integrity containers should have as a design goal a

minimum lifetime of 300 years. The high integrity container

should be designed to maintain its structural integrity over

this period.

c. The high integrity container design should consider the

corrosive and chemical effects of both the waste contents and

the disposal trench environment. Corrosion and chemical tests

should be performed to confirm the suitability of the proposed

container materials to meet the design lifetime goal.

d. The high integrity container should be designed to have

sufficient mechanical strength to withstand horizontal and

vertical loads on the container equivalent to the depth of

proposed burial assuming a cover material density of 120

lbs/ft . The high integrity container should also be designed

to withstand the routine loads and effects from the waste

contents, waste preparation, transportation, handling and

disposal site operations, such as trench compaction procedures.

This mechanical cosign strength should be justified by

conservative design dyses.

e. For polymeric material, design mechanical strengths should be

conservatively extrapulated from creep test data.
9

f. The design should consider the thermal loads from processing,

storage, transportation and burial. Proposed container

materials should be tested in accordance with ASTM 8553 in the

manner described in Section C2(g) of this technical position.

No significant changes in material design properties should

result from this thermal cycling.

g. The high integrity container design should consider the

radiation stability of the proposed container materials d well

as the radiation egradation effects of the wastes.

Radiation degradation testing should be performed on proposed

container materials using a gamna irradiator or equivalent. No

significant changes in material design properties shguld result

following exposure to a total accumulated dose of 10 Rads. If

it is proposed to design the high integrity container to

greater accumulated doses, testing should be performed to

confirm te adequdcy of the proposed materials. Test specimens

should be prepared using the proposed fabrication techniques.

Polymeric high integrity container designs should also consider

the effects of ultra-violet radiation. Testing should be

performed on proposed materials to show that no significant

changes in material design properties occur following expected

ultra-violet radiation exposure.

h. The high integrity container design should consider the

biodegradation properties of the proposed materials and any

biodegradation of wastes and disposal media. Biodegradation

testing should be performed on proposed container materials in

accordance with ASTM G21 and ASTM G22. No indication of

culture growth should be visible. The extraction procedure

described in Section C2(d) of this technical position may be

performed where indications of visible culture growth can be

attributable to contamination, additives, or biodegradable

components on the specimen surface that do not affect the

overall integrity of the substrate. It is also acceptable to

determine biodegradation rates using the Bathta-Pramer Method

described in Section C (d). The rate of biodegradation should

produce less than a 10 percent loss of the total carbon in the

container material after 300 years. Test specimens should be

prepared using the proposed material fabrication techniques.
10

1. The high integrity container should be capable of meeting the

requirements for a Type A package as specified in 49 CFR

173.398(b). The free drop test may be performed in accordance

with 10 CFR 7, Appendix A, Section 6.

j. The high integrity container and the associated lifting devices

should be designed to withstand the forces applied during

lifting operations. As a minimum the container should be

designed to withstand a 3g vertical lifting load.

k. The high integrity container should be designed to avoid the

collection or retention of water on its top surfaces in order

to minimize accumulation of trench liquids which could result

in corrosive or degrading chemical effects.

1. High integrity container closures should be designed to provide

a positive seal for the design lifetime of the container. The

closure should also be designed to allow inspections of the

contents to be conducted without damaging the integrity of the

container. Passive vent designs may be utilized if needed to

relieve internal pressure. Passive vent systems should be

designed to minimize the entry of moisture and the passage of

waste materials from the container.

m. Prototype testing should be performed on high integrity

container designs to demonstrate the container's ability to

withstand the proposed conditions of waste preparation,

hdndling, transportation and disposal.

n. High integrity containers should be fabricated, tested,

inspected, prepared for use, filled, stored, handled,

transported and disposed of in accordance with a quality

assurance program. The quality assurance program should also

address how wastes which are detrimental to high integrity

container materials will be precluded from being placed into

the container. Special emphasis should be placed on

fabrication process control for those high integrity containers

which utilize fbrication techniques such as polymer molding

processes.

5. Filter Cartridge Wastes
11

For Class B and C wastes in the form of filter cartridges, the waste

generator should demonstrate that the selected approach for providing

stability will meet the requirements in 10 CFR Part 61. Encapsulation of

the filter cartridge in a solidification binder or the use of d high

integrity container are acceptable options for providing stability. When

high integrity containers are used, waste generators should demonstrate

that protective means are provided to preclude container damage during

packaging handling and transportation.

D. Implementation

This technical position reflects the current NRC staff position on

acceptable means for meeting the 10 CFR Part 61 waste stability

requirements. Therefore, except in those cases in which the waste

generator proposes an acceptable alternative method for complying with

the stability requirements of 10 CFR Part 61, the guidance described

herein will be used in the evaluation of the acceptability of waste forms

for disposal at near-surface disposal facilities.
12

References:

1. NUREG-C800, Standard Review Plan

2. ANS 55.1, "American National Standard for Solid Radioactive Waste

Processing System for Light Water Cooled Reactor Plants," American

Nuclear Society, 1979

3. ASTM C39, "Compressive Strength of Cylindrical Concrete Specimens,"

American Society for Testing and Materials, 1979

4. ASTM 01074, "Compression Strength of Bituminous Mixtures," American

Society for Testing and Materials, 1980

5. ASTM G21, "Determining Resistance of Synthetic Polymeric Materials to

Fungi," American Society for Testing and Materials, 1970

6. ASTM G22, "Determining Resistance of Plastics to Bacteria," American

Society for Testing and Materials, 1976

7. R. Bartha, 0. Pramer, "Features of a Flask and Method for Measuring

the Persistance and Biological Effects of Pesticides in Soils,"

Soil Science 100 (1), pp-68-70, 1965

8. ANS 16.1, "Measurement of the Leachability of Solidified Low-Level

Radioactive Wastes," American Nuclear Society Oraft Standard, April 1981

9. ASTM 8553, "Thermal Cycling of Electroplated Plastics," American

Society for Testing and Materials, 1979