UNITED STATES 

0• NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D. C. 20555 

SJAN 241991 

ATTENTION: Commission Licensees 

SUBJECT: Waste Form Technical Position, Revision 1 

The regulation, "Licensing 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 wastes have lower concentrations and may be segregated without stabilization. Class A wastes may also be stabilized and disposed of with stabilized Class B and C 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 (a) promote slumping, collapse, or other failure of the 

cap or cover over a near-surface disposal unit and thereby lead to water infiltration, or (b) impart a substantial increase in surface area of the 

waste form that could lead to an increase in leach rate. 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 'rovided by the waste form itself (as with activated stainless steel components), by processing the waste to a stable form (e.g., solidification), or by emplacing the waste in a container or structure that provides stability (e.g., high integrity container or engineered structure). 

This technical position on waste form was initially developed in 1983 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 has been used as an acceptable approach for demonstrating compliance with the 10 CFR Part 61 waste stability criteria. This position includes guidance on (1) the processing of wastes into 

an acceptable, stable waste form, (2) the design of acceptable high integrity containers, (3) the packaging of filter cartridges, and (4) minimization of radiation effects on organic ion-exchange resins. The regulation, 10 CFR 20.311 (d)(1), requires waste generators and processors to prepare wastes that meet the waste characteristics requirements of Part 61 (including the requirements for structural stability). The recommendations and guidance provided in this technical position are an acceptable method to demonstrate 

waste stability. One way of demonstrating conformance with the general recommendations contained in this technical position is to reference an approved Topical Report, because such reports are reviewed and approved in accordance with the acceptance criteria contained in this technical position. However, additional actions (e.g., plant-specific process control procedures.) by waste generators will be needed to demonstrate that a stabilized plant-specific waste stream satisfies Part 61 waste form requirements.
Since the initial issuance of the Technical Position, it has been the intent 

of the NRC staff to provide additional guidance on waste form as it became 

necessary to address other pertinent waste form issues. One such issue 

involves the use of cement to stabilize low-level wastes. Field experience 

and laboratory testing of cement-solidified low-level radioactive waste has 

indicated that some unique chemical and physical interactions can occur between 

the cement constituents and the chemicals and compounds that can exist in the 

waste materials. Therefore, an appendix (Appendix "A") dealing with the 

qualification testing, performance confirmation and reporting of mishaps 

involving cement-stabilized waste forms has been included in this revision to 

the Technical Position. 

To provide more comprehensive guidance on cement stabilization of low-level 

radioactive waste, Appendix A addressed several areas of concern that were not 

considered in the May 1983, Revision 0, version of this Technical Position. 

Thus, information and guidance on cement waste form specimen preparation, 

statistical sampling and analysis, waste characterization, process control 

program (PCP) specimen preparation and examination, surveillance specimens and 

reporting of mishaps are provided in Appendix A. The guidance provided in 

Appendix A is the culmination of an extended period of study and information 

gathering and exchange between the NRC staff and representatives of various 

organizations, including government laboratories, the Advisory Committee on 

Nuclear Waste (ACNW), cement processing vendors, other waste form vendors, 

nuclear utilities, and state regulatory agencies. Especially useful in the 

development of the guidance in Appendix A was the information exchanged in a 

Workshop on Cement Stabilization of Low-Level Radioactive Waste held in June 

1989. The Workshop proceedings have been published as an NRC report, 

NUREG/CP-0103, which is available from the following sources: 

Superintendent of Documents 

U.S. Government Printing Office 

P.O. Box 37082 

Washington, DC 20013-7082 

and 

National Technical Information Service 

Springfield, VA 22161 

The waste form technical position is attached to this letter. Questions on the 

technical position paper may be referred to Dr. M. Tokar (301-492-0590), or to 

Ms. M. T. Adams (301-492-0505) of my staff. The information collections 

contained in the technical position have been approved under OMB number 

3150-0014. 

Paul H. Lohaus, Chief 

Low-Level Waste Management Branch 

Division of Low-Level Waste Management 

and Decommissioning, NMSS

Enclosure: As stated
[7590-01]

U. S. NUCLEAR REGULATORY COMMISSION 

Availability of Revised Staff Technical Position 

on Waste Form 

AGENCY: Nuclear Regulatory Commission 

ACTION: Notice of Availability 

SUMMARY: The Nuclear Regulatory Commission (NRC) is announcing the 

availability of a revised Staff Technical Position entitled "Technical 

Position on Waste Form (Revision 1)." 

The Position provides guidance on acceptable methods for demonstrating 

compliance with the waste form structural stability requirements of 

10 CFR Part 61 and for supporting the waste generator and processor 

certification requirements of 10 CFR 20.311.

i
[7590-01]

The Technical Position on Waste Form was initia~lly developed in 1983 to 

provide guidance to low-level radioactive waste generators on waste form 

test methods and results acceptable to the NRC staff for implementing 

the 10 CFR Part 61 waste form requirements. Since the initial 

issuance of the technical position, field experience and laboratory 

testing of cement-solidified low-level waste have indicated that some 

unique chemical and physical interactions can occur between the cement 

and the waste constituents, interactions that can affect the waste form 

stability. Therefore, an appendix (Appendix "A") dealing with 

cement-stabilized waste forms has been included in this revision to the 

Technical Position. 

To provide more comprehensive guidance on cement stabilization of 

low-level radioactive waste, Appendix A addresses several areas of 

concern that were not considered in the May 1983, Revision 0, version of 

this Technical Position. Information and guidance on cement waste form 

specimen preparation, statistical sampling and analysis, waste 

characterization, process control program (PCP) specimen preparation and 

examination, surveillance specimens and reporting of mishaps are 

provided In Appendix A. 

The guidance provided in the revised Technical Position is the 

culmination of an extended period of study and information gathering and 

exchange between the NRC staff and representatives of various 

organizations including government laboratories, the Advisory Committee on '7

2
£7590-01l

Nuclear Waste (ACNW), cement processing vendors, other waste form vendors, 

nuclear utilities, and state regulatory agencies. Especially useful in 

the development of the guidance in Appendix A was the information 

exchanged in a Workshop on Cement Stabilization of Low-Level Radioactive 

Waste held in June 1989. 

The Workshop proceedings have been published as an NRC report, 

NUREG/CP-0103, which is available from the following sources: 

Superintendent of Documents 

U.S. Government Printing Office 

P.O. Box 37082 

WASHINGTON, DC 20013-7082 

and 

National Technical Information Service 

Springfield, VA 22161 

Copies of the revised Technical Position are being distributed (under 

seperate cover) to licensees. Copies are also being distributed 

(seperately) by State Programs to the Agreement States, Non-Agreement 

States, State Liaison Officers, and others who are on the NRC's Compact 

Distribution List. 

ADDRESSES: Copies of the Staff Technical Position may be obtained by 

writing to M. T. Adams at Mail Stop 5E-2 OWFN, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555.

A 3
[7590-01]

FOR FURTHER INFORMATION CONTACT: M. T Adams, Division of Low-Level Waste 

Management and Decommissioning, Office of Nuclear Material Safety and 

Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 

Telephone (301) 492-0505. 

Dated at Rockville, Maryland, this 121 day of 1989. 

For the Nuclear Regulatory Commission 

Paul H. Lohaus, Chief 

Low-Level Waste Management Branch 

Division of Low-Level Waste Management 

and Decommissioning, NMSS

4
'PAREG&4 

United States Nuclear Regulatory Commission 

Office of Nuclear Material Safety and Safeguards 

Washington, D.C. 20555 

TECHNICAL POSITION 

ON 

WASTE FORM 

(Revision 1) 

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

Prepared by: Low-Level Waste Management Branch 

Division of Low-Level Waste Management 

and Decommissioning 

January 1991

i
TABLE OF CONTENTS 

Technical Position on Waste Form 

A. INTRODUCTION .................................................. 1 

B. BACKGROUND .................................................... 2 

C. REGULATORY POSITION .................................. ........ 4 

1. Solidified Class A Waste Products ....................... 4 

2. Stability Guidance for Processed Class B and C Wastes .... 4 

3. Radiation Stability of Organic Ion-Exchange Resins ....... 7 

4. High Integrity Containers ................................ 7 

5. Filter Cartridge Wastes ................................. 10 

6. Reporting of Mishaps .................................... 10 

D. IMPLEMENTATION ............................................... 11 

E. REFERENCES ................................................... 12 

Appendix A - Cement Stabilization 

I. INTRODUCTION ................................................. A-1 

II. WASTE FORM QUALIFICATION TESTING ............................. A-2 

A. General ................................................. A-2 

B. Compression .............................................. A-2 

C. Thermal Cycling .......................................... A-4 

D. Irradiation ............................................. A-5 

E. Biodegradation ........................................... A-6 

F. Leach Testing .............. ............................. A-6 

G. Immersion Testing ....................................... A-8 

H. Free Standing Liquids ................................... A-8 

I. Full-scale Testing ...................................... A-9 

III. QUALIFICATION TEST SPECIMEN PREPARATION ...................... A-9 

A. Mixing .................................................. A-9 

B. Curing .................................................. A-9 

C. Storage ................................................. A-10 

IV. STATISTICAL SAMPLING AND ANALYSIS ............................ A-1O 

V. WASTE CHARACTERIZATION ...................................... A-11 

VI. PCP SPECIMEN PREPARATION AND EXAMINATION ..................... A-12 

A. General ................................................. A-12 

B. Preparation of PCP Specimens ............................ A-14 

C. PCP Specimen Examinations and Testing ................... A-14 

1. Short-term Specimens ............................... A-14 

2. Long-term Surveillance Specimens ................... A-15 

VII. SURVEILLANCE SPECIMENS ....................................... A-15 

VIII. REPORTING OF MISHAPS ......................................... A-16 

IX. IMPLEMENTATION ............................................... A-17 

X. REFERENCES ................................................... A-18

i
Technical Position on Waste Form

A. INTRODUCTION 

The regulation, "Licensing 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 wastes have lower concentrations and may be segregated 

without stabilization. Class A wastes may also be stabilized and disposed of 

with stabilized Class 8 and C 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 

(a) promote slumping, collapse, or other failure of the cap or cover over a 

near-surface disposal trench and thereby lead to water infiltration, or (b) 

impart a substantial increase in surface area of the waste form that could lead 

to an increase in leach rate. Stability is also a factor in limiting exposure 

to an inadvertent intruder since it provides greater assurance that the waste 

form will remain in a recognizable and nondispersable state. Structural 

stability of a waste form can be provided by the waste form itself (as with 

activated stainless steel components), by processing the waste to a stable form 

(e.g., solidification), or by emplacing the waste in a container or structure 

that provides stability (e.g., high integrity container or engineered 

structure). 

This technical position on waste form was initially developed in 1983 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 has been used as 

an acceptable approach for demonstrating compliance with the 10 CFR Part 61 

waste stability criteria. This position includes guidance on (1) the 

processing of wastes into an acceptable, stable waste form, (2) the design of 

acceptable high integrity containers, (3) the packaging of filter cartridges, 

and (4) minimization of radiation effects on organic ion-exchange resins. The 

regulation, 10 CFR 20.311, requires waste generators and processors to certify 

that their waste forms meet the requirements of Part 61 (including the 

requirements for structural stability). The recommendations and guidance 

provided in this technical position are an acceptable method upon which to base 

such certification by waste generators. One way of demonstrating conformance 

with the general recommendations contained in this technical position is to 

reference an approved Topical Report, because such reports are reviewed and 

approved in accordance with the acceptance criteria contained in this technical 

position. Additional actions (e.g., plant-specific process control procedures) 

by waste generators, however, to demonstrate that a stabilized plant-specific 

waste stream satisfies Part 61 waste form requirements, will be needed. 

Since the initial issuance of the Technical Position, it has been the intent of 

the NRC staff to provide additional guidance on waste form as it became 

necessary to address other pertinent waste form issues. One such issue 

involves the use of cement to stabilize low-level wastes. Field experience and 

laboratory testing of cement-solidified low-level radioactive waste has 

indicated that some unique chemical and physical interactions can occur between 

the cement constituents and the chemicals and compounds that can exist in the

1
waste materials. Therefore, an appendix (Appendix "A") dealing with the 

qualification testing, performance confirmation and reporting of mishaps 

involving cement-stabilized waste forms has been included in this revision to 

the Technical Position. (Reporting of mishaps is addressed for other types of 

waste forms in Section C.6 of the main body of this Technical Position). 

To provide more comprehensive guidance on cement stabilization of low-level 

radioactive waste, Appendix A addresses several areas of concern that were not 

considered in the May 1983, Revision 0, version of this Technical Position. 

Thus, information and guidance on cement waste form specimen preparation, 

statistical sampling and analysis, waste characterization, process control 

program (PCP) specimen preparation and examination, surveillance specimens and 

reporting of mishaps are provided in Appendix A. The guidance provided in 

Appendix A is the culmination of an extended period of study and information 

gathering and exchange between the NRC staff and representatives of various 

sectors of the nuclear industry, including government laboratories, cement 

processing vendors, other waste form vendors, nuclear utilities, state 

regulatory agencies, and industry representative organizations such as the 

Nuclear Management Resources Council (NUMARC) and the Electric Power Research 

Institute (EPRI). Especially useful in the development of the guidance in 

Appendix A was the information exchanged in a Workshop on Cement Stabilization 

of Low-Level Radioactive Waste (Ref. 1). 

B. BACKGROUND 

Historically, waste form and container properties were considered of secondary 

importance to good site selection; a properly operated site having good 

geologic and hydrologic characteristics was considered the only barrier 

necessary to isolate low-level radioactive wastes from the environment. As 

experience in operating low-level waste disposal sites was acquired, however, 

it became apparent that the waste form should play a significant role in the 

overall plan for managing these wastes. 

The regulation for near-surface disposal of radioactive wastes, 10 CFR Part 61, 

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, and 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 absorb liquids. 

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

have structural stability. As stated 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 

(a) slumping, collapse, or other failure of the disposal unit (if an engineered 

structure is not used) resulting from degraded wastes which could lead to water 

infiltration, radionuclide migration, and costly remedial care programs and (b) 

radionuclide release from the waste form that might ensue due to increases in

2
leaching that could be caused by premature disintegration of the waste form. 

Stability is also considered in the intruder pathways where it is assumed that 

wastes are recognizable after the active control period, and that, therefore, 

continued inadvertent intrusion would be unlikely. To the extent practical, Class B and C waste forms should maintain gross physical properties and 

identity over a 300 year period. 

To ensure that Class B and C wastes will maintain stability, the following conditions should be met: 

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

provides stability after disposal. 

b. The waste should not contain free standing and corrosive liquids. 

That is, the wastes should contain only trace amounts of drainable 

liquid, and, as required by 10 CFR 61.56(b)(2), 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 waste or container should be resistant to biodegradation. 

e. The waste or container should remain stable under the compressive 

loads inherent in the disposal environment. 

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

or water after disposal. 

g. The as-generated waste should be compatible with the solidification 

medium or container. 

A large portion of the waste produced in the nuclear industry, including waste 

from nuclear power plants, 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 form for burial. The wet wastes, regardless of their 

classification, are required to be either absorbed or solidified. To assure 

that this processing will consistently produce a product which is acceptable 

for disposal and will meet disposal site license conditions, nuclear power plant licensees are required to process their wastes in accordance with a 

plant-specific process control program (PCP). Guidance for such PCPs was 

provided in NRC Standard Review Plan Section 11.4, "Solid Waste Management 

Systems," NUREG-0800 (Ref. 2) and its accompanying Branch Technical Position 

ETSB 11-3, "Design Guidance for Solid Waste Management Systems Installed in 

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

However, 10 CFR Part 61 became effective in January 1983, providing requirements regarding waste form, and superseding certain of the guidance previously provided in NUREG-0800. Licensee's PCPs provide assurance that the 

processing of wet radioactive wastes will result in waste forms that meet the 

requirements of 10 CFR Part 61 and low-level waste disposal sites licenses.

3
Plant-specific PCPs developed and approved without consideration of Part 61 

should be revised to provide assurance that applicable Part 61 requirements 

will be satisfied. In many cases, licensee PCPs are based on generally 

applicable (generic) PCPs contained in vendor-submitted topical reports that 

are reviewed by the NRC for referencing in licensing actions. 

The guidance in this technical position may also serve as the basis for 

ualifying generic PCPs for Class B and C wastes. Applicable generic test data 

e.g., topical reports) may be used for generic PCP qualification, and may be 

used in part as the basis for a plant-specific PCP. PCPs for solidified Class 

A wdste products that are to be segregated from Class B and C wastes need only 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 

(HIC). The high integrity container would be used to provide the long-term 

stability required to meet the structural 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, a PCP for dewatering wet 

solids in HICs or liners 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 free liquids as measured using 

the method described in ANS 55.1 (Ref. 4). 

b. 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 

The stability guidance in this technical position for processed wastes 

should be implemented through the qualification of the individual 

licensee's PCP. Generic test data may be used for qualifying generic 

PCPs, and incorporated as part of the individual licensee's (i.e., 

plant-specific) PCP. Tests to demonstrate waste form stability through a 

generic testing program include the following: 

a. Solidified waste specimens should have compressive strengths of at 

least 60 psi when tested in accordance with ASTM C39 (Ref. 5). 

Compressive strength tests for bituminous products should be 

performed in accordance with ASTM D1074 (Ref. 6).

4
Many solidification agents (such as cement) will be easily capable 

of meeting the 60 psi limit for properly solidified wastes. For 

such cases, process control parameters should be developed to achieve 

maximum practical compressive strengths, not simply to achieve the 

minimum acceptable compressive strength; (see Section II.B of 

Appendix A for further guidance on cement-stabilized wastes). 

b. Waste specimens should be resistant to thermal degradation. The 

heating and cooling chambers used for the thermal degradation 

testing should conform to the description given in ASTM B553, 

Section 3 (Ref. 7). Samples suitable for performing compressive 

strength tests in accordance with ASTM C39 or ASTM D1074 should be 

used. Samples should be placed in the test chamber and a series of 

30 thermal cycles carried out in accordance with Section 5.4.1 

through 5.4.4 of ASTM 8553. The high temperature limit should be 

600C and the low temperature limit -40°C. Following testing the 

waste specimens should have the maximum practical compressive 

strengths; (a minimum compressive strength of 60 psi as tested using 

ASTM D1074 is acceptable for bituminized waste forms--for cement

stabilized wastes see Section II.C of Appendix A). 

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 proposed waste stream formulation 

should be exposed to a minimum of 10E+8 Rads in a gamma irradiator 

or equivalent. If the maximum level of exposure is expected to 

exceed 10E+8 Rads, testing should be performed at the expected 

maximum accumulated dose. Following irradiation the irradiated 

specimens should have the maximum practical compressive strengths (a 

minimum compressive strength Of 60 psi as tested using ASTM D1074 is 

acceptable for bituminized waste forms--for cement-stabilized wastes 

see Appendix A). 

d. Specimens for each proposed waste stream formulation should be 

tested for resistance to biodegradation in accordance with both ASTM 

G21 and ASTM G22 (Refs. 8 & 9, respectively). No indication of 

culture growth should be visible. Specimens should be suitable for 

compression testing in accordance with ASTM C39 or ASTM D1074, as 

applicable. Following the biodegradation testing, specimens should 

have the maximum practical compressive strengths (a minimum 

compressive strength of 60 psi as tested using ASTM D1074 is 

acceptable for bituminized waste forms--see Section II.E of Appendix 

A for guidance on biodegradation testing of cement-stabilized 

wastes). 

For polymeric or bitumen products, some visible culture growth from 

contamination, additives, or biodegradable components on the 

specimen surface that does not relate to overall substrate integrity

5
may be present. For these cases, additional testing should be 

performed. If culture growth is observed upon completion of the 

biodegradation test for polymeric or bitumen products, the test 

specimens should be removed from the culture and washed free of all 

culture and growth with water, with only light scrubbing. An 

organic solvent compatible with the substrate may be used to extract 

surface contaminants. The specimen should be air dried at room 

temperature and the test repeated. Specimens should have observed 

culture growths rated no greater than I in the repeated ASTM G21 

test. The specimens should have no observed growth in the repeated ASTM G22 test. Compression testing should be performed in 

accordance with ASTM C39 or ASTM D1074, as applicable, following the 

repeated G21 and G22 tests. The minimum acceptable compressive strength for bituminized waste forms is 60 psi. Maximum practical compressive strengths should be established for other media. 

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 (Ref. 10) is 

acceptable for this testing. Soils used should be representative of 

those at disposal facilities. Biodegradation extrapolated for 

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 minimum of 90 days (5 days for cement-stabilized waste forms--see Section II.F of Appendix A 

for cement-stabilized wastes) in accordance with the procedure in 

ANS 16.1 (Ref. 11). Specimen sizes should be consistent with the 

samples prepared for the ASTM C39 or ASTM D1074 compressive strength 

tests. In addition to the demineralized water test specified in ANS 

16.1, additional testing using other leachants specified in the 

Standard 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. For proposed nuclear power station 

waste streams, cobalt, cesium, and strontium should be used as 

tracers. The leachability index, as calculated in accordance with 

ANS 16.1, should be greater than 6.0. 

f. Waste specimens should maintain maximum practical compressive strengths 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; (see Section II.G 

of Appendix A for guidance on cement-stabilized wastes). 

g. 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 ANS 55.1. Free liquids should have a pH between 4 and 

11; (for cement-solidified water, free liquids should have a minimum 

pH of 9--see Section II.H of Appendix A). 

6
h. 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. Correlation testing should be performed 

using 90-day immersion (including post-immersion compression) tests 

on the most conservative waste stream(s) intended for use for the 

particular solidification medium; i.e, the waste stream that presents 

the most difficulty in consistently producing a stable product(s). 

For cement-solidified waste forms, the mixed bead resin waste stream 

is expected to be the most conservative. For bituminized wastes, the 

sodium sulfate waste stream should be used. The full-scale specimens 

should be fabricated using solidification equipment the same as or 

comparable to that used for processing actual low-level radioactive 

wastes in the field.. 

i. 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 representative of the compressive strength as determined by 

testing lab-scale specimens (i.e., that meet the criteria called out 

in Section C2.a. above). Full-scale specimens may be fabricated 

using simulated non-radioactive products; however, the specimens 

should be fabricated using solidification equipment that is the same 

as or comparable to that used in the field for actual low-level 

radioactive wastes. 

3. Radiation Stability of Organic Ion-Exchange Resins 

To ensure that organic ion exchange resins will not undergo adverse 

degradation effects from radiation, resins should not be generated having 

loadings that will produce greater than 1OE+8 Rads total accumulated dose. 

For Cs-137 and Sr-90 a total accumulated dose of 10E+8 Rads is 

approximately equivalent to a 10 Ci/ft concentration in resins in the 

unsolidified, as-generated form. In the event that the waste generator 

considers it necessary to load resins higher than 10E+8 Rads, it should be 

demonstrated that the specific resin will not undergo radiation 

degradation at the proposed higher loading. The test method should 

adequately simulate the chemical and radiologic conditions expected. A 

gamma irradiator or equivalent should be utilized for these tests. There 

should be no adverse swelling, acid formation or gas generation that 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 

7
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 be 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 

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 deptý 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 design strength should be justified by 

conservative design analyses. 

e. For polymeric material, design mechanical strengths should be 

conservatively extrapolated from creep test data. It should be 

demonstrated for high integrity containers fabricated from polymeric 

materials that the containers will not undergo tertiary creep, creep 

buckling, or ductile-to-brittle failure over the design life of the 

containers. 

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

storage, transportation and burial. Proposed container materials 

should be tested in accordance with ASTM B553 in the manner 

described in Section C2(b) 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 as well as the 

radiation degradation effects of the wastes. Radiation degradation 

testing should be performed on proposed container materials using a 

gamma irradiator or equivalent. No significant changes in material 

design properties should result following exposure to a total 

accumulated dose of 10 E+8 Rads. If it is proposed to design the

8
high integrity-container to greater accumulated doses, testing 

should be performed to confirm the adequacy of the proposed 

materials. Test specimens should be prepared using the proposed 

fabrication techniques. 

High integrity container designs using polymeric materials 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 

Bartha-Pramer Method described in Section C2(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. 

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

requirements for a Type A package as specified in 49 CFR 173.411 and 

173.412. Conditions that may be encountered during transport or 

movement are to be addressed by meeting the requirements of 

10 CFR 71.71. 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.

9
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, handling, transportation 

and disposal. 

n. High integrity containers should be designed, fabricated, and 

used in accordance with a quality assurance program. The quality 

assurance program should address the following topics concerning 

the high integrity container: fabrication, testing, inspection, 

preparation for use, filling, storage, handling, transportation, 

and disposal. 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 fabrication 

techniques such as polymer molding processes. 

5. Filter Cartridge Wastes 

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 a 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. 

6. Reporting of Mishaps 

In all future reviews and approvals of stabilization media and high 

integrity containers, waste generators, vendors and processors will, as a 

condition of approval, be asked to commit to reporting any knowledge they 

may have of misuse or failure of their waste forms and containers. Such 

mishaps include, but are not necessarily limited to, the following: 

a. The failure of high integrity containers used to ensure structural 

stability. Such failure may be evidenced by changed container 

dimensions, cracking, or injury from mishandling.(e.g., dropping or 

impacting against another object). 

b. The misuse of high integrity containers, as evidenced by a quantity 

of free liquid greater than one percent of container volume, or an 

excessive void space within the container; (such use is in violation 

of 10 CFR 61.56(a)). 

c. The production of a solidified Class B or C waste form that has any 

of the following characteristics; 

1. greater than 0.5 percent volume of free liquid.

10

I
2. concentrations of radionuclides greater than the 

concentrations demonstrated to be stable in the waste form 

in qualification testing accepted by the regulatory 

agency. 

3. greater or lessor amounts of solidification media than 

were used in qualification testing accepted by the 

regulatory agency. 

4. contains chemical ingredients not present or accounted in 

qualification testing accepted by the regulatory agency. 

5. shows instability evidenced by crumbling, cracking, 

spalling, voids, softening, disintegration, 

nonhomogeneity, or change in dimensions. 

6. evidences processing phenomena that exceed the limiting 

processing conditions identified in applicable topical 

reports or process control programs, such as foaming, 

excessive temperature, premature or slow hardening, 

production of volatile material, etc. 

Waste form mishaps should be reported to the NRC's Director of the 

Division of Low-Level Waste Management and Decommissioning and the 

designated State disposal site regultory authority within 30 days of 

knowledge of the incident. For any such waste form mishap occurrence, the 

affected waste form should not be shipped off-site until approval is 

obtained from the disposal site regulatory authority. The reason for this 

is that the low-level waste generators and processors are required by 10 

CFR 20.311 to certify that their waste forms meet all applicable 

requirements of 10 CFR Part 61, and waste forms that are subject to the 

types of mishaps mentioned above may not possess the required long-term 

structural stability. When mishaps of the nature described above occur, it 

is expected that, before the waste form is shipped to a disposal facility, 

either adequate mitigation of the potential effects on the waste form or 

an acceptable justification concerning the lack of any potential 

significant effects of the affected waste form on the overall performance 

of the disposal facility would be provided. 

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, vendor, and/or processor 

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.

11 A
E. REFERENCES 

1. "Proceedings of the Workshop on Cement Stabilization of Low-Level 

Radioactive Waste," NUREG/CP-0103, October 1989. 

2. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear 

Power Plants," (LWR Edition), NUREG-0800, July, 1981. 

3. "Update on Waste Form and High Integrity Container Topical Report Review 

Status, Identification of Problems with Cement Solidification, and Reporting of 

Waste Mishaps," NRC Information Notice No. 90-xx, (in preparation). 

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

Processing System for Light Water Cooled Reactor Plants," American Nuclear 

Society, 1979. 

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

American Society for Testing and Materials, 1979. 

6. ASTM D1074, "Compression Strength of Bituminous Mixtures," American Society for Testing and Materials, 1980. 7. ASTM B553, "Thermal Cycling of 

Electroplated Plastics," American Society for Testing and Materials, 1979. 

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

Fungi," American Society for Testing and Materials, 1970. 

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

Society for Testing and Materials, 1976. 

10. R. Bartha, D. Pramer, "Features of a Flask and Method for Measuring the 

Persistence and Biological Effects of Pesticides in Soils," Soil Science 100 

(1), pp. 68-70, 1965. 

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

Radioactive Wastes," American Nuclear Society Draft Standard, April 1981. 

12
Appendix A 

Cement Stabilization 

I. INTRODUCTION 

This Appendix to the Technical Position on Waste Form provides guidance to 

waste generators and processors who intend to use cementitious materials such 

as Portland and pozzolonic-type cements to solidify and stabilize low-level 

radioactive wastes in accordance with the requirements of 10 CFR Part 61 (Ref. 

Al(a)). This guidance is applicable for cementious waste forms destined for 

disposal in shallow-land disposal sites and engineered structures where the 

regulatory authorities require stable waste forms. It is expected that the 

guidance described herein would be used by NRC staff in any Topical Report evaluation of the acceptability of cement waste forms for disposal at 

near-surface disposal facilities. Waste generators using cement solidification 

systems and media not approved generically through the Topical Report review 

process may use this guidance to conduct testing to demonstrate that waste 

forms satisfy the requirements of Part 61. NRC regulation 10 CFR 20.311 (Ref. A1(b)) requires waste generators to certify that their waste forms meet the 

requirements of Part 61 (including the requirements for structural stability). Waste generators whose cement waste formulations meet the provisions of this 

Technical Position will be able to certify that the formulations meet the 

requirements of Part 61. The disposal site regulatory authorities, however, have the ultimate reponsibility for accepting or rejecting the waste. 

Portland and pozzolonic cements have been observed to exhibit unique chemical 

and physical interactive behavior when used with certain materials and 

chemicals encountered in some low-level radioactive waste streams. Therefore, 

this Appendix specifically addresses cement waste form qualification only and 

is not intended to be applied generically to all stabilization agents (although 

many of the provisions discussed are, in principle, applicable to other media). This Appendix thus complements, and does not replace, the main body of the 

Technical Position on Waste Form. 

Included in this Appendix are descriptions of methods that may be used in 

cement waste form qualification testing. Associated acceptance criteria that 

may be used by NRC staff or others to evaluate the acceptability of the test 

results are also provided. Included in this waste form testing guidance are 

descriptions of acceptable procedures for sample preparation and statistical 

treatment of data. In addition, this Appendix provides guidance on waste 

stream characterization, process control program (PCP) recipe qualification and 

specimen examination, surveillance specimen preparation and testing, and 

procedures for reporting of cement waste form preparation mishaps. This 

guidance on cement waste forms is intended to provide the best available 

information on an acceptable approach for demonstrating that a 

cement-solidified low-level radioactive waste form will possess the long-term (300-year) structural stability that is required by Part 61 for Class B and 

Class C wastes. 

A - 1
Linkage between the waste form qualification test recommendations in this 

Technical Position and the requirements of Part 61 is provided in 10 CFR 

61.56(b)(1), where it is stated that "a structurally stable waste form will 

generally maintain its physical dimensions and form, under the expected 

disposal conditions such as weight of overburden and compaction equipment, the 

presence of moisture and microbial activity, and internal factors such as 

radiation effects and chemical changes." The discussion provided in Section II 

of this Appendix addresses the details of the testprocedures and acceptance 

criteria recommended for cement-stabilized wastes. Further information on test 

specimen preparation and analysis of data is provided in Section III and 

Section IV, respectively. 

II. WASTE FORM QUALIFICATION TESTING 

A. General 

As indicated in Section C.2 of the main body of this Technical Position, 

generic test data may be used "for qualifying process control programs." That 

is, a low-level radioactive waste generator/processor may perform qualification 

testing, as described in the following subsections of this Appendix, to qualify 

recipes for a range of waste compositions (concentrations and loadings) for a 

given type of waste stream. It is incumbent upon the party providing 10 CFR 

20.311 certification, however, to show that the composition(s) of the waste 

form specimens used in the qualification testing adequately covers the range of 

waste compositions that will be encountered in the field. An acceptable 

approach to qualification testing is to perform the tests not only at the 

maximum waste loading but also at lower loadings (at least one), with 

appropriate variations in water/cement ratios and proportions of additives. It 

should not be necessary to perform all the qualification tests for all of the 

waste loadings, but adequate justifications should be provided for any 

omissions. 

Each individual waste stream should be qualified with test data obtained for 

that specific waste stream. In cases where two or more waste streams are 

combined, it should be demonstrated that the specimen compositions used in the 

qualification testing adequately cover the range of compositions that are 

intended to be stabilized in the field. This may be accomplished by performing 

the full series of qualification tests on the "worst-case" composition only, 

along with one or more tests on alternate compositions, sufficient to show that 

the selected "worst-case" was chosen correctly. 

B. Compression 

It is stated in 10 CFR 61.56(b)(1) that "a structurally stable waste form will 

generally maintain its physical dimensions and form under expected disposal 

conditions such as weight of overburden and compaction equipment...." Assuming 

a cover material density of 120 lbs./cu.ft., a minimum compressive strength 

criterion of 50 psi was established in section C.2.b. of the 1983 Revision 0 

portion of this Technical Position. To reflect the increase in burial depth 

(from 45 to 55 feet) at Hanford, Washington, the minimum compressive strength 

criterion for generic waste forms was later increased from 50 to 60 psi.

A-2
However, as further noted in the above-cited section C.2.a., for solidification 

agents that are easily capable of meeting the 50 (now 60) psi minimum 

compressive strength, the waste forms should achieve "maximum practical compressive strengths," not just the "minimum acceptable compressive strength." This provision was included in the Rev. 0, 1983 Technical Position in 

recognition of the fact that mere resistance to deformation under burial loads 

is, in itself, inadequate evidence that the waste form microconstituents are 

bonded together sufficiently well to ensure that the waste form will not over 

time fall apart due to internal stresses that are chemically, physically, or 

irradiation induced. 

Portland cement mortars, which are comprised of mixtures of cement, lime, 

silica sand and water, are readily capable of achieving compressive strengths of 5000 to 6000 psi; that is approximately two orders of magnitude greater than 

the minimum compressive strength required to resist deformation under load in 

current low-level waste burial trenches. Therefore, to provide greater 

assurance that there will be sufficient cementitious material present in the 

waste form to not only withstand the burial loads, but also to maintain general 

"dimensions and form" (i.e., to not disintegrate) over time, it is recommended 

that cement-stabilized waste forms possess compressive strengths that are 

representative of the values that are reasonably achievable with current cement 

solidification processes. Taking into consideration the fact that low-level 

radioactive waste material constituents are not in most cases capable of 

providing the physical and chemical functions of silica sand in a cement 

mortar, a mean compressive strength equal to or greater than 500 psi is 

recommended for waste form specimens cured for a minimum of 28days (see 

Section III.B of Appendix A). This value of compressive strength is 

recommended as a practical strength value that is representative of the quality 

of cementitious material that should be used in the waste form to provide 

assurance that it will maintain integrity and thus possess the long term 

structural capability required by Part 61. 

Compressive strengths of cement-stabilized waste forms should be determined in 

accordance with procedures described in ASTM Standard C39: Compressive Strength 

of Cylindrical Concrete Specimens (Ref. A2). It is recommended that the 

compressive strength test specimens be right circular cylinders, 2 to 3 inches 

in diameter, with a length-to-diameter (L/D) ratio of approximately two. 

Because hydrated cement solids are brittle ceramic materials that fail in 

tension or shear rather than compression, and at regions of localized stress 

concentration or microstructural flaw, there tends to be considerable scatter 

in the strength test data even if all processing variables are kept relatively 

constant. Therefore, sufficient specimens should be tested to determine the 

mean compressive strength and standard deviation. Because of the many 

variables involved, a decision regarding the specific number of specimens to be 

tested is left to the judgement of the waste processor/qualifier; in no case, 

however, should the number of as-cured (pre-environmental test) compressive 

strength test specimens be less than ten. This approach should continue until 

there are sufficient data available to permit judgements to be made regarding what is reasonably achievable, from a statistical standpoint, in compressive 

strength testing of low-level waste test specimens. No precision criterion, in 

the form of an acceptable variance or standard deviation, is recommended at 

this time. 

A - 3

I
For the purposes of verification of Process Control Program (PCP) parameters 

see discussion in Section VI of Appendix A), compressive strength tests and/or 

penetrometer hardness tests should be performed after the qualification test 

specimens have been allowed to cure for approximately 24 hours. The results of 

these tests should be retained and made available for comparison with the 

results of similar tests that should be performed on PCP specimens fabricated 

from actual radioactive wastes in the field; (see Appendix A, Section VI.C for 

details).] 

C. Thermal Cycling 

Though thermal effects are not called out specifically as an item of concern in 

10 CFR 61.56(b)(1), as other factors are, cement-stabilized low-level 

radioactive waste forms should be demonstrated to be resistant to thermal 

degradation. There are three basic reasons for this: (1) Section 61.56(b)(1) 

of Part 61 lists "internal factors" as a condition that must be considered in 

assuring that a waste form will retain structural stability, and temperature 

and thermal effects are internal factors; (2) thermal cycling of the waste form 

will occur, particularly during the storage and transport phase of the waste 

form's performance "life;" and (3), experience has shown that the thermal 

cycling test has served well in distinguishing between "strong" and "weak" 

solidified waste forms. The thermal cycling test imposes a stress (due to 

differential thermal expansion) between the various microconstituents of the 

waste form and between different regions of the waste form. By cycling between 

the maximum and minimum temperatures called for in the test, any cracks 

initiated in the test specimen may propagate and eventually measurably weaken 

the waste form. The extent of any degradation that might occur will be a 

function of various factors such as the amount of cementitious material in the 

waste form, the bond strength between the materials present, and the morphology 

of the microconstituents in the waste form microstructure. Thus, the thermal 

cycling test, by subjecting the waste form specimens to a short-term cyclic 

thermal stress, challenges the structural capability of the specimens and thus 

serves as a very useful vehicle for screening out unfavorable "weak" 

formulations. 

The heating and cooling chambers used in determining the thermal cycling 

resistance of cement-stabilized waste forms should, as stated in Section C.2.b. 

of the main body of this Technical Position, conform to the description given 

in ASTM Standard Test Method B553 (Ref. A3). However, because that test method 

addresses thermal cycling of electroplated plastics, not cement-solidified 

waste materials, some modifications to the test procedure are necessary. Test 

specimens suitable for performing compressive strength tests in accordance with 

ASTM C39 should be used. The specimens should be tested "bare;" i.e., not in a 

container. Specimens should be placed in the test chamber, and a series of 30 

thermal cycles should be carried out in accordance with Section 5.4.1 through 

5.4.4 of ASTM B553, with the additional proviso that the specimens should be 

allowed to come to thermal equilibrium at the high (60 degrees C) and low (-40 

degrees C) temperature limits. Thermal equilibrium should be confirmed by 

measurements of the center temperature of at least one specimen (per test 

group). A minimum of three specimens for each waste formulation should be 

subjected to the thermal cycling tests.

A- 4
Following exposure to 30 thermal cycles the specimens should be examined 

visually and should be free of any evidence of significant cracking, spalling, 

or bulk disintegration; i.e., visible evidence of significant degradation would 

be indicative of failure of the test. Because it is not possible to provide an 

a priori assessment of the significance of visible defects, taking into 

consideration the wide range of possible defect configurations, no definition 

of "significant degradation" is provided here. The organization performing the 

tests should (1) assess whether visible defects are significant, and (2) obtain 

and retain photographic evidence of any defects that are judged to be 

insignificant for future reference. If there are no significant visible 

defects, the test specimens should be subjected to compression strength testing 

in accordance with ASTM C39 and should have mean compressive strengths that are 

equal to or greater than 500 psi. 

D. Irradiation 

In accordance with the requirements of 10 CFR61.56(b)(1), and as indicated in 

Section C.2.c. of the main body of this Technical Position, irradiation testing 

of solidified waste forms should be conducted on specimens exposed to a minimum 

dose of 10E+8 rads. The 10E+8 rads radiation dose is approximately equivalent 

to the dose that would be acquired by a waste form over a 300-year period, if 

the waste form were loaded to a Cesium-137 or Strontium-90 concentration of 10 

Ci/cu.ft. This is the recommended (Ref. A3) maximum activity level for organic 

resins based on evidence that while a measurable amount of damage to the resin 

will occur at 10E+8 rads, the amount of damage will have negligible effect on 

power plant or disposal site safety. However, cementitious materials are not 

affected by gamma radiation to relatively high cumulative doses (e.g., greater 

than 1OE+9 rads--Ref. A4) considerably in excess of 10E+8 rads. Therefore, for 

cement-stabilized waste forms, irradiation qualification testing need not be 

conducted unless (1) the waste forms contain ion exchange resins or other 

organic media or (2) the expected cumulative dose on waste forms containing 

other materials is greater than 10E+9 rads. Testing should be performed on 

specimens exposed to (1) 10E+8 rads or the expected maximum dose greater than 

IOE+8 rads for waste forms that contain ion exchange resins or other organic 

media or (2) the expected maximum dose greater than 10E+9 rads for other waste 

forms. In cases where irradiation testing is warranted, a minimum of three 

specimens should be tested for each waste formulation being qualified. 

Following the irradiation exposure the specimens should be examined visually 

and should be free of any evidence of significant cracking, spalling, or bulk 

disintegration; i.e., visible evidence of significant degradation would be 

indicative of failure of the irradiation test. If there are no significant 

visible defects (see Section II.C for discussion of "significant degradation"), 

the test specimens should be subjected to compressive strength testing in 

accordance with ASTM C39 and should have mean compressive strengths that are 

equal to or greater than 500 psi.

A- 5
E. Biodegradation 

As indicated in 10 CFR 61.56(b)(1), a structurally stable waste form is one 

that will be relatively unaffected by "microbial activity." Generic (not 

specific to type of waste form) recommendations for biodegradation testing 

provided in Section C.2.e. of the main body of this Technical Position indicate 

that ASTM Standard Practice G21 (Ref. A5) and G22 (Ref. A6) are suitable 

methods of test for determining susceptibility to fungi and bacteria, 

respectively. Experience in biodegradation testing of cement-stabilized waste 

forms has shown (Refs. A7-A9), however, that they generally do not support 

fungal or bacterial growth. The principal reason for this appears to be that 

the fungi and microbes used in the G21 and G22 tests require a source of carbon 

for growth, and in the absence of any carbonaceous materials in the waste 

stream, there is no internal food source available for culture growth. 

Consequently, for cement-stabilized waste forms, biodegradation qualification 

testing need not be conducted unless the waste forms contain carbonaceous 

materials (e.g., ion exchange resins or oils). 

For cement-stabilized waste forms containing carbonaceous materials, there 

should be no evidence of culture growth during the G21 and G22 tests. The test 

specimens (at least three for each organic waste stream formulation being 

qualified) should also be free of any evidence of significant cracking, 

spalling or bulk disintegration; i.e., visible evidence of significant 

degradation would be indicative of failure of the test. If there are no 

significant visable defects following the test exposures (see Section II.C of 

this Appendix for discussion of "significant degradation"), the test specimens 

should be subjected to compression strength testing in accordance with ASTM C39 

and should be shown to have mean compressive strengths equal to or greater than 

500 psi. 

F. Leach Testing 

Resistance to leaching of radionuclides is not specifically mentioned in Part 

61, nor is radionuclide containment called out as a specific requirement for 

low-level waste packages. Minimization of contact of waste by water is a 

fundamental concern of Part 61, however, as evidenced by the statement in 

Section 61.7 that "...a cornerstone of the system is stability...so that 

access of water to the waste can be minimized (emphasis added). Migration of 

radionuclides is thus minimized..." In addition, there are several statements 

in Section 61.51 that address minimization of contact of water with waste. 

These statements are in recognition of the fact that contact of waste with 

water is the first step in a potentially major pathway for radionuclide release 

and migration off-site. Thus, "leaching," or release of radionuclides from a 

waste form through contact with water is a first step in subsequent migration 

of the radionuclides from the waste through the groundwater and off the site. 

Therefore, leaching is a phenomenon that is of fundamental interest in waste 

disposal.

A-6
The leach testing procedure specified in Section C.2.e. of the main body of this Technical Position is ANSI/ANS 16.1: Measurement of the Leachability of 

Solidified Low-Level Radioactive Wastes by a Short-Term Test Procedure (Ref-. A1O). In the ANS/ANSI 16.1 test, a test specimen is completely immersed in a measured volume of water, which is changed on a prescribed schedule. Upon removal, the leachant is analyzed for the radionuclides (or elements) of interest. The data obtained by this procedure are expressed as a material 

parameter of the leachability of each leached species. This parameter is called the "Leachability Index" (L), which is the arithmetic mean of the L values obtained for each leaching interval (where the L value is the logarithm of the inverse of the effective diffusivity). The leachability index, as calculated in accordance with ANSI/ANS 16.1, should be greater than 6.0. 

The period of time specified for the leach test in the above-cited Section C.2.e. of this Technical Position is a minimum of 90 days, and the test period called out in the Standard corresponds to 90 days. This time period was selected as a means of determining whether there might be a change in leach mechanism with time; (as explained in the Standard, early leach rates observed 

with solidified waste forms are most often explained by diffusion--other mechanisms, such as erosion, dissolution, or corrosion, would generally be discernible only after longer leaching times). However, any leaching that involves other mechanisms such as erosion, dissolution, corrosion or other chemical or physical phenomena would most likely be readily observed visually and through mechanical testing. Such observations would be made as part of the immersion test, which is a 90-day test. These facts, coupled with comparisons of 5-day and 90-day data (Ref. All) on cement waste forms that showed that the percentage differences between 5-day and 90-day leach indices were relatively small for most specimens, indicate that a 5-day leach testing period is 

sufficient for cement-solidified wastes. 

The leachant specified in ANSI/ANS 16.1 is deionized water. It is stated in the above-cited Section C.2.e. of this Technical Position that additional testing using other leachants should also be performed to confirm the solidification agents leach resistance in other leachant media. Synthesized sea water leachant is listed as a preferred alternate leachant. The basis for this is, that while leachability indices are generally lower (i.e., leach rates 

are higher) for tests conducted in demineralized water than in sea water (Ref. All), this is not true in all cases for all waste streams. For reasons of 

economy, however, it is desirable to limit the bulk of the testing to one 

leachant. If it can be shown that the chosen leachant is the most aggressive 

one, testing with one leachant is appropriate. Since it is not possible to initially predict (Ref. A9) which leachant (deionized water or synthesized seas water) would be most aggressive, sufficient preliminary testing should be conducted to identify the most aggressive leachant for each waste form formulation being qualified, and that leachant should be used for the balance of the testing (if only one is used). An acceptable method of identifying the 

most aggressive leachant is to perform 24 hour (or longer) leaching measurements on both leachants and to use the leachant that resulted in the lowest leach indices (i.e., highest leach rate) for the remaining days of 

testing.

A- 7
G. Immersion Testing 

No "Standard Method of Test" for immersion testing has been adopted for 

low-level radioactive waste, but as indicated in Section C.2.f. of the main 

body of this Technical Position, immersion testing may be performed in 

conjunction with the leach testing (which is to be performed in accordance with 

ANSI/ANS 16.1). However, in contrast with the period of time (5 days) 

necessary for leach testing of cement-stabilized wastes, immersion testing 

should be performed for a minimum period of 90 days. The immersion testing 

should be performed in either deionized water or synthesized sea water. The 

immersion liquid should be selected on the basis of short-term (24-hour or 

longer) leach tests that identify the most aggressive immersion medium (see 

discussion of leach testing). 

The test specimens (at least three for each waste stream formulation being 

qualified) should be cured for a minimum cure time of 28 days (see Section III, 

"Specimen Preparation," of Appendix A for details) prior to being immersed. 

Following immersion, the specimens should be examined visually and should be 

free of any evidence of significant cracking, spalling, or bulk disintegration. 

If there are no significant visible defects (see Section II.C of this Appendix 

for discussion of "significant degradation"), the specimens should be subjected 

to compressive strength testing in accordance with ASTM C39 and should have 

post-immersion mean compressive strengths that are equal to or greater than 500 

psi and not less than 75 percent of the pre-immersion test (i.e., as-cured) 

mean compressive strength. If the post-immersion mean compressive strength is 

less than 75 percent of the as-cured specimens' pre-immersion mean compressive 

streigth, (but not less than 500 psi) the immersion testing interval should be 

extended (using additional specimens) to a minimum of 180 days. For these 

cases, sufficient compressive strength testing should be conducted (for 

example, after 120, 150, and 180 days of immersion) to establish that the 

compressive strengths level off and do not continue to decline with time. 

For certain waste streams (viz., bead resins, chelates, filter sludges, and 

floor drain wastes) that have been found to exhibit complex relationships of 

cure time and immersion resistance (Ref. A12), additional immersion testing 

should be performed on specimens that have been cured (in sealed containers) 

for a minimum of 180 days. The immersion period should be for a minimum of 7 

days, followed by a drying period of 7 days in ambient air at a minimum 

temperature of 20 degrees Celsius. After the specimens are dried, they should 

meet the post-immersion test visual and compressive strength criteria specified 

above. 

H. Free Standing Liquids 

It is stated in 10 CFR 61.56(b)(2) that "...liquid wastes, or wastes containing 

liquid, must be converted into a form that contains as little free standing or 

noncorrosive liquid as is reasonably achievable, but in no case shall the 

liquid exceed...0.5% of the volume of the waste for waste processed to a stable 

form." Correspondingly, waste test specimens should have less than 0.5 percent 

by volume of the waste specimen volume as free liquids as measured using the 

method described in Appendix 2 of ANSI/ANS 55.1 (Ref. A13). Inasmuch as cement

A-8
is an alkaline material, evidence of acidic free liquids is indicative of 

improper waste form preparation or curing. Therefore, any free liquid from 

cement-stabilized waste forms should have a minimum pH of 9. 

I. Full-scale Testing 

It is expected that the testing performed in accordance with the guidance 

provided in Sections A through H above will be carried out on small, laboratory 

scale specimens. As indicated in Section C.2.h. of the main body of this 

Technical Position, therefore, it is necessary to correlate the characteristics 

of full-size products with those of laboratory size specimens. The full-scale 

specimens should be fabricated using solidification equipment that is the same 

as or comparable to that used in processing real low-level waste forms in the 

field. The correlation of full-scale product characteristics should be 

accomplished by performing (1) compressive strength tests on as-cured material 

(cured for a minimum of 28 days), and (2) 90-day immersion tests that include 

post-immersion compressive strength tests (See Section II.G above) for the most 

conservative waste stream(s) being qualified. 

Test specimens obtained from the full-scale waste forms by coring or sectioning 

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 that meet the criteria called out in Section II.B above. 

III. QUALIFICATION TEST SPECIMEN PREPARATION 

A. Mixing 

Experience in preparation of lab-scale and full-scale cement-solidified waste 

forms (Ref. A9) has shown that the method employed in mixing the ingredients 

can have a dramatic influence on the reactivity of the materials, the structure 

of the solidified waste form, and the resultant properties and characteristics 

of the waste form. Important parameters include type of equipment and mixing 

time because they will determine the amount of energy imparted to the 

ingredients used in the solidification recipe. This is especially important in 

cases where properties and characteristics of small, lab-scale specimens are 

used to predict the behavior of large, full-scale products. In preparing 

laboratory-sized qualification test specimens, it should be shown by analysis 

and/or testing that the type of equipment used, the mixing time, the speed of 

the mixer, etc. will, in combination, impart the same degree of mixing to the 

laboratory specimens as the full-scale mixing equipment and procedure will 

impart to full-scale waste forms and that the degree of mixing is sufficient to 

ensure production of homogeneous waste forms. 

B. Curing 

The curing conditions for small, laboratory-scale qualification test specimens, 

should, to the extent practical, be the same as the conditions obtained with 

full-scale products. Inasmuch as cement constituents exhibit a significant 

exothermic heat of hydration, while possessing low thermal conductivity, the 

interior temperature of large, full-scale cement waste forms may be. elevated

A-9
significantly (approaching even the boiling point of water). To ensure that 

the laboratory specimens endure curing conditions that are reasonably similar 

to those of full-size products, the waste form centerline temperature profile 

as a function of time should be obtained for the largest full-sized waste form 

to be qualified for each waste stream. That profile should be duplicated, to 

the extent practical, in the laboratory specimens. An acceptable method is to 

cure the specimens in a suitable oven for a period of time equivalent to the 

peak heat of hydration period. For the purposes of this Technical Position 

that period of time is taken to be that required for the centerline temperature of a full-scale waste form to decrease to a near-ambient (30 degrees Celsius or 

lower) temperature level. 

Care should be taken to ensure that the waste loadings and cement concentrations in the full-scale waste forms provide sufficient margin to 

preclude reaching the boiling point of the pre-solidification mix. This is 

necessary to ensure that the waste form formulations will not be subject to 

uncontrolled variations due to water losses caused by evaporation during set. 

Uncontrolled porosities due to vapor bubble formation and rapid set due to elevated temperatures will also be avoided by limiting the maximum temperatures in the cement-solidified waste forms. 

The compressive strength of hydrated cement and concrete solids increases 

asymptotically as the mixtures cure. Normally, the strength at 28 days approaches seventy-five percent or more of the "peak" value, though when 

pozzolonic cements are used the time required to reach peak strength may be extended. Sufficient test specimens should be prepared to determine the 

compressive strength increase with time to ensure that the specimens have 

attained sufficient (i.e., greater than 75% of the projected peak) strength prior to subjecting the remaining specimens to the qualification testing called 

out in Sections II.C through II.G. of this Appendix. 

C. Storage 

Test specimens that will be subjected to the qualification testing described in 

Section II of this Appendix should be kept in sealed containers during curing and storage. This is intended to simulate the environment that would be 

obtained in a typical full-scale waste form liner and will prevent loss of 

water that might affect the performance of the waste form specimens during subsequent testing. 

IV. STATISTICAL SAMPLING AND ANALYSIS 

As noted in the discussion of compressive strength testing (see Section II.B 

above), there tends to be considerable scatter in the compressive strength data 

obtained on brittle ceramic materials such as cement. Therefore, sufficient 

specimens should be tested in the as-cured condition to provide enough data to 

establish a mean and standard deviation, though for reasons discussed in 

Appendix A Section II.B, the number of as-cured specimens to be tested is left 

to the judgement of the waste formulation qualifier. For statistical purposes, i however, the number of as-cured (pre-environmental test) compressive strength 6 specimens should be ten or greater for a given formulation. Further discussion 

A - 10
of the rationale for this provision is provided in Section II.B of this 

Appendix. For the minimum quantities of test specimens recommended in the 

respective subsections of this Appendix, the specimens tested should have a 

post-test mean compressive strength that is equal to or greater than 500 psi. 

Note that for the immersion tests, a slightly different acceptance criterion is 

identified, in subsection II.G of this Appendix. Variations in individual 

specimen compression strength need not be considered. 

Other than the determinations of compressive strength, the only other parameter 

of interest in qualification testing of low-level waste forms that lends itself 

to statistical treatment is the leachability index. ANSI/ANS 16.1 (Ref. A10) 

uses the confidence range and correlation coefficient as measures of 

discrepancies in the measurements of leachability. The Standard requires that 

the confidence range and correlation coefficient be reported with the 

Leachability Index. As is the case of the ASTM C39 Compressive Strength 

standard, however, no precision criterion has been established yet for the 

ANSI/ANS 16.1 leach test. 

V. WASTE CHARACTERIZATION 

The importance of waste characterization was extensively discussed at the 

May/June Workshop on Cement Stabilization of Low-Level Radioactive Waste that 

was held in Gaithersburg, MD. The Proceedings (Ref. A9) of the Workshop, 

particularly the efforts of Working Group 4, record the discussions and provide 

useful information on the routine characterization of typical waste streams. 

Waste characterization would typically be expected to include as a minimum the 

identification of major constituents in the waste (including primary ions and 

salts or other solids), density, pH, temperature, radioactive isotopes, and a 

check for the presence of secondary ingredients that could significantly affect 

the hydration of the cement. 

Some waste streams, such as pressurized water reactor (PWR) primary coolant 

system borated water, are relatively well-characterized and free of secondary 

ingredients. There are other waste streams, however, such as ion exchange 

resins, filter sludges and floor drain liquids, that may contain chemicals that 

can significantly retard or accelerate the hydration of cement or in other ways 

adversely affect cement waste form performance (Ref. A9). It is impractical 

for a waste processor to perform qualification testing on every possible 

combination and concentration of secondary constituents in a given type of 

waste stream. Nor is it considered practical or necessary for a waste 

generator to perform a complete quantitative chemical analysis on every batch 

of waste that is produced. It is, however, incumbent on radwaste system 

managers and processors to be cognizant of the types of chemicals that may 

produce problems in using cement in the solidification and stabilization of 

low-level radioactive waste. The introduction of such chemicals into waste 

treatment systems that utilize cement stabilization media should be avoided or 

specifically compensated for in the formula used for stabilizing that waste 

stream. If the waste processor is a vendor or is otherwise not the generator 

of the waste, it is incumbent on all parties to be in adequate communication 

with each other with regard to the types and quantities of chemical ingredients 

in the waste and the capability of the waste formulation to provide long-term

A - 11
structural stability to the waste form. As a part of process control, mixing 

of different wastes in holding tanks and transfer of liquid wastes without 

adequate flushing of lines should be generally avoided, because such mixing 

might introduce ingredients into the waste that were not present in the 

qualification test program that was conducted for the waste stream in question. 

To assist waste generators and processors in developing a sense of greater 

awareness of low-level radioactive waste stream ingredients that may adversely 

affect the setting and stability of cement-solidified waste forms, a list of 

such chemicals is provided in Table I. This list is not intended to be all

inclusive. Moreover, some of the constituents listed may be considered 

hazardous materials, as defined by Environmental Protection Agency (EPA) 

criteria, and which thus, if mixed with radioactive material, could be 

classified as a "mixed waste." Any questions about low-level radioactive wastes 

that might be classified as mixed wastes should be directed to the EPA. 

Low-level radioactive waste generators and processors who intend to stabilize 

Class 8 and Class C waste with cement should either (a) prevent the 

contamination of, (b) limit to the extent practical, or (c) pre-treat as 

appropriate, waste streams that may contain the chemicals and constituents in 

Table I. It is the responsibility of the waste generator and processor to 

ensure that the cement formulation used for a given waste stream is qualified 

for the waste stream chemical constituents and concentrations in question. 

VI. PCP SPECIMEN PREPARATION AND EXAMINATION 

A. General 

The purpose of a Process Control Program (PCP) is to describe the envelope 

within which processing and packaging of low-level radioactive wastes will be 

accomplished to provide reasonable assurance of compliance with low-level waste 

requirements. All commercial nuclear power plants have plant specific PCPs. 

The guidance provided in this section of this Appendix is not, however, 

intended to address facility-specific PCPs, which, in addition to containing a 

general description of the methods for controlling the processing and packaging 

of radioactive waste, may also contain a description of the system and 

operating procedures, instructions on manifest preparation, and a discussion of 

administrative controls. Rather, this guidance addresses only the recipe 

portion of cement stabilization of low-level waste; that is, the guidance 

addresses the nature of the information that should be provided in a generic 

PCP concerning the type and quantity of ingredients used in the cement waste 

form formulation, the order of addition, and the method, process, and time 

required for mixing the ingredients in the preparation of verification and 

surveillance specimens as well as the full-scale waste forms. Also provided is 

guidance on the preparation of PCP "verificationu and surveillance specimens 

and the type of examinations and testing that should be performed on those 

specimens. 

'a

A - 12
This information on verification specimens is intended to provide assurance 

that the formulations used in the qualification testing program correspond to 

those actually used in the field. The surveillance specimen program, described 

in Section VII of this Appendix, is intended to provide verification that the 

waste forms are remaining stable with time. 

For each low-level radioactive waste formulation, the generic PCP should 

address the boundary conditions (i.e., bounding process parameters) for 

processing the waste to provide reasonable assurance that the final waste form 

will meet 10 CFR Part 61 stability requirements. The process parameters will 

be influenced by (a) the characteristics of the waste prior to processing, (b) the qualities of the solidification medium, as influenced by additives, and (c) the physical/chemical process of preparing the waste into a final waste form. 

Variables that influence the process and have an effect on the product, and 

that should be, therefore, be identified and restricted within acceptable bounds for each waste form include the following: 

1. Type of waste (e.g., bead resin, including type--anion/cation/mixed/ manufacturer/weak acid/strong acid, percent depleted, powdered resins, 

boric acid, sludges); 

2. Waste characteristics having influence on the final waste form (e.g., pH, oil content, chelating agents, water content, maximum concentration of 

secondary ingredients); 

3. Additives (e.g., type of cement, water, lime, silica fume, fly ash, 

furnace slag,) and the order of addition; 

4. Physical process parameters (e.g. maximum temperature, mixing equipment required, mixing and curing timesi. 

The generic PCP should indicate how representative samples of the feed waste 

are to be obtained for preparing PCP verification and surveillance specimens. The PCP should identify typical and maximum batch sizes and the number of PCP 

specimens to be taken for each batch. The PCP should describe where 

adjustments could be made to the feed waste material, in the event that certain 

feed material parameters that may be encountered in the field fall outside of 

the acceptable range for processing. These adjustments should not be 

undertaken if the resultant waste stream feed material and stabilized waste 

form were to be chemically or physically different from that qualified in 

laboratory testing. 

If, during the course of full-scale waste form preparation at a nuclear power plant, it should become necessary to effect an ad hoc, impromptu change in the 

approved recipe or procedure to avoid an incomplete or otherwise unsatisfactory solidification condition, the change should be reviewed and approved by the 

facility licensee pursuant to the provisions of 10 CFR 50.59. This process should be followed in all such cases where ad hoc changes are necessary whether 

or not a generic PCP has received approval as part of a Topical Report review 

process. Inasmuch as the affected waste form would lack assurance of long-term 

A - 13

I
structural stability (because it was produced under conditions that were 

outside of the envelope of the conditions used in the qualification tests), it 

is anticipated that the resultant waste form would not be accepted for disposal 

at a disposal site without the expressed approval of the disposal site 

regulatory authorities. It is also anticipated that, prior to accepting the 

waste, the regulatory authority would require either (1) adequate mitigation of 

any potential adverse effects on the long-term structural stability of the 

waste form or (2) an acceptable justification concerning the lack of any 

potential significant effect of the affected waste form on the overall 

performance of the facility. Alternatively, the disposal site regulatory 

authority could accept the affected waste for disposal with the provision that 

the required structural stability would be provided at the disposal facility by 

means of an engineered structure. 

After the generic PCP has been reviewed and approved by the NRC, the PCP 

parameters and procedures should be followed as described in the Topical Report 

(or other documentation) so that the 10 CFR 20.311 certification can be made 

without the need for additional justification that the cement-solidified waste 

meets the requirements of 10 CFR Part 61. Once a generic PCP has been approved 

by the NRC any subsequent changes to the generic PCP should be reviewed and 

approved by the NRC. Any incomplete or otherwise unsatisfactory solidification 

condition known to waste generators and processors is requested to be reported 

to the NRC (Director, Division of Low-Level Waste Management and 

Decommissioning) within 30 days after such an occurrence is known (see Section 

VIII). The actions taken to produce an acceptable waste form after the initial 

unsatisfactory solidification condition was identified should be described. 

B. Preparation of PCP Specimens 

Prior to plant-specific solidification of full-scale waste forms, 

representative samples of the feed waste should be obtained in sufficient 

quantity to prepare the desired number of PCP specimens. The feed waste 

material should be solidified using the recipe that has been qualified in 

laboratory testing for the given waste stream. Mixing of the waste materials 

with the cement and additives should be accomplished in a manner that 

duplicates, to the extent practical, the mixing conditions that are obtained 

with full-scale mixing. The specimens should be cured under conditions similar 

to those used in the laboratory qualification test program. PCP specimens 

should be prepared for each batch of waste that is required to meet the 10 CFR 

Part 61 structural stability criteria. For the purposes of the guidance 

provided in this Technical Position, a "batch" is herein defined as any 

quantity of waste stream feed material that is from a single source (e.g., a 

holding tank), that is processed as a single batch (even though it maybe 

subdivided in more than one unit waste form; e.g., liner), and that, 

therefore, possesses unvaried, single operation, batch characteristics. 

C. PCP Specimen Examinations and Testing 

1. Short-term (24-hour PCP Verification) Specimens 

Prior to solidifying full-scale waste forms, plant-specific PCP verification 

specimens should be prepared, in accordance with procedures described above,

A - 14
for examination and compressive strength testing. The specimens should be free of significant visible defects, such as cracking, spalling or disintegration and should exhibit less than 0.5% by volume of the specimen as free liquid. As 

a measure of process control, the specimens should, within a 24-hour period after preparation, be subjected to an ASTM C39 compressive strength test; (penetrometer measurements may be substituted, as described below). The 

compressive strength values should be within two standard deviations of the 

mean compressive strength values obtained at 24 hours for test specimens prepared and tested as part of the associated laboratory generic qualification 

test program for the waste formulation. Alternatively, penetrometer tests can 

be used in lieu of C39 compressive strength measurements if acceptable correlation data demonstrating the relationship between the compressive strength values and penetrometer values have been obtained for the waste stream 

formulation in question. If penetrometer tests are used, the mean penetrometer hardness values obtained on the verification specimens should be within two 

standard deviations of the mean obtained on the qualification test specimens for that formulation. If the compressive strength or penetrometer measurements 

do not meet the above criteria, a second set of PCP specimens should be 

prepared and retested. The second set of PCP specimens should be fabricated 

using either the same formula or an adjusted one that falls within the 

compositional envelope of the qualification tests conducted for that waste 

stream. 

2. Long-term Surveillance Specimens 

The guidance herein addressing long-term surveillance specimens is directly applicable to waste generators and to vendors processing wastes at licensed facilities who intend to certify, in accordance with the provisions of 10 CFR 

20.311, that the cement-solidified waste meets the structural stability requirements of 10 CFR Part 61. Sufficient PCP specimens should be prepared to 

permit the retention, examination and testing of surveillance specimens. The surveillance specimens should be stored in sealed containers at normal room 

temperatures. The examination and testing of surveillance specimens is 

described in Section VII of this Appendix. 

VII.'SURVEILLANCE SPECIMENS 

The purpose of the surveillance specimens is to provide confirmation that the 

waste forms prepared for certain waste streams, (in particular bead resins, chelates, filter sludges, and floor drain wastes) are performing as expected. At periods of time equal to 6 months and 12 months after preparation, the 

surveillance specimens should be examined visually and should be free of 

evidence of significant cracking, spalling or bulk disintegration (see Section 

II.C of Appendix A for discussion of "significant degradation"). At least one 

specimen should be subjected to an ASTM C39 compressive strength (or penetrometer) test at the 6 and 12 month periods. The mean compression strength (or penetrometer) value(s) obtained should be not more than two 

standard deviations below the mean of the as-cured strength or penetrometer values obtained with the qualification test specimens cured for an equivalent 

period of time. 

A - 15

i
At 12 months after preparation, one or more PCP surveillance specimens should 

be subjected to an immersion test. The duration of the immersion test should 

be a minimum of 14 days. Upon removal from the immersion liquid, which should 

be either deionized water or synthesized sea water (see Section II.F of this 

Appendix) the specimens should be allowed to dry in ambient air for a minimum 

of 48 hours. The specimens should then be examined visually and should be free 

of significant surface or bulk defects such as cracking, spalling, or bulk 

disintegration. Following the immersion test, the specimen(s) should be 

subjected to an ASTM C39 compressive strength (or penetrometer) test. The test 

results should meet the criteria discussed above. 

If the PCP surveillance specimens tested either by the vendor of an 

NRC-approved Topical Report or by a utility or other licensee, should fail any 

of the above tests, the wastes previously solidified may not meet the stability 

requirements of 10 CFR Part 61. Therefore, the NRC (Director, Division of 

Waste Management and Decommissioning) and licensee (if other than the waste 

processor that shipped the suspect waste to the disposal facility) should be 

notified in writing within 30 days. In turn, the licensee should notify the 

disposal facility operator and regulatory authority if the 10 CFR 20.311 

certification as to waste stability was invalidated by this finding. The 

licensee's report should satisfy the information needs of the regulatory 

authority and should describe the waste stream solidified, the waste 

formulation used, the number of full-scale waste forms that had been produced, 

date of shipment, manifest numbers, and the results of the tests. The report 

should also contain a discussion of the significance of the test results and 

proposed changes, if any, that might have to be made to the waste formulation 

to ensure that, for the waste stream in question, future waste forms would be 

stable. 

For all waste processors (including utility licensees and vendors of 

NRC-approved Topical Reports), it is recommended that a summary report that 

addresses the results of PCP surveillance specimen preparations and 

examinations should be prepared annually by the waste processor and submitted 

to the NRC (Director, Division of Waste Management and Decommissioning). The 

report should document the results of all visual examinations and immersion, 

compression, and/or penetometer tests performed on the cement-stabilized waste 

form surveillance specimens during the calendar year. The annual report should 

be submitted within 90 days of the end of each calendar year. A commitment to 

provide this information will be made a condition of approval for all future 

license applications, topical report submittals or other regulatory actions 

that deal with cement waste forms, where the waste generators and/or processors 

desire NRC endorsement of their 10 CFR 20.311 certifications. 

VIII. REPORTING OF MISHAPS 

Known cement waste form processing mishaps, including but not restricted to, 

cement waste forms that have not solidified completely, waste forms that have 

swelled and/or disintegrated, waste forms that were not prepared in accordance 

with an approved PCP, and waste form preparations that resulted in unusual 

exothermic reactions, should be reported by the cognizant waste processor to 

the NRC (Director of the Division of Waste Management and Decommissioning)

A - 16
within 30 days of the time that the vendor becomes aware of the incident. 

Licensees should also report such mishaps to the disposal site regulatory 

authority since such an event may indicate the waste form will or does not 

satisfy the stability requirements of 10 CFR Part 61. If the mishap becomes 

known to the waste generator and/or processor before the waste forms are 

shipped off-site, the affected waste form(s) should not be shipped until 

approval is obtained from the disposal site regulatory authority. A commitment 

to report and deal with waste form mishaps as discussed above will be made a 

condition of approval for all future license applications, topical report 

submittals, or other regulatory actions that deal with cement waste forms, 

where the waste generators and/or processors desire NRC endorsement of their 10 

CFR 20.311 certifications. 

IX. IMPLEMENTATION 

This Appendix to the Technical Position on Waste Form reflects the current NRC 

staff position on an acceptable means for meeting the 10 CFR Part 61 structural 

stability requirements for cement waste forms. Therefore, except in those 

cases in which the waste generator, vendor, and/or processor proposes an 

acceptable alternative method for complying with the stability requirements of 

10 CFR Part 61, the guidance described herein will be used by the NRC staff in 

all future evaluations of the acceptability of cement waste forms for disposal 

at near-surface disposal facilities. 

A - 17
X. REFERENCES 

Al(a). Part 61 - Licensing Requirements for Land Disposal of Radioactive Waste, 

Code of Federal Regulations, Title 10: Energy. 

Al(b). "Method for Obtaining Approval of Proposed Disposal Procedures," 

Subsection 311 of Part 20 (20.302), Code of Federal Regulations, Title 10: 

Energy. 

A2. American Society for Testing and Materials Standard Test Method for 

Compressive Strength of Cylindrical Concrete Specimens, ASTM C39, October 1984. 

A3. D.R. MacKenzie, M. Lin, and R.E. Barletta, "Permissible Radionuclide 

Loading for Organic Ion Exchange Resins for Nuclear Power Plants," Brookhaven 

National Laboratory Draft Report, BNL-NUREG-30668, January 1982. 

A4. P. Soo and L. W. Milian, "Sulfate-Attack Resistance and Gamma-Irradiation 

Resistance of Some Portland Cement Based Mortars," Brookhaven National 

Laboratory Report, NUREG/CR-5279, March 1989. 

A5. American Society for Testing and Materials Standard Practice for 

Determining Resistance of Synthetic Polymeric Materials to Fungi, ASTM G21, 

1985. 

A6. American Society for Testing and Materials Standard Practice for 

Determining Resistance of Plastics to Bacteria, ASTM G22, 1985. 

A7. P.L. Piciulo, C.E. Shea, and R.E. Barletta, "Biodegradation Testing of 

Solidified Low-Level Waste Streams," Brookhaven National Laboratory Report, 

NUREG/CR-4200 (BNL-NUREG-51868), May 1985. 

A8. B.S. Bowerman, et al., "An Evaluation of the Stability Tests Recommended in 

the Branch Technical Position on Waste Forms and Container Materials," 

Brookhaven National Laboratory Report, NUREG/CR-3289 (BNL-NUREG-51784), March 

1985. 

A9. Proceedings of the Workshop on Cement Stabilization of Low-Level 

Radioactive Waste, U.S. Regulatory Commission Report, NUREG/CP-0103, October 

1989. 

AlO. American National Standards Institute/American Nuclear Society American 

National Standard for Measurement of the Leachability of Solidified Low-Level 

Radioactive Wastes by a Short-Term Test Procedure, ANSI/ANS 16.1-1986, April 

14, 1986. 

All. W. Chang, L. Skoski, R. Eng, and P.T. Tuite, "A Technical Basis for 

Meeting the Waste Form Stability Requirements of 10 CFR 61," Nuclear Management 

and Resources Council, Inc. Report, NUMARC/NESP-002, April 1988.

A - 18

I
A12. P. L. Piciulo, J. W. Adams, J. H. Clinton, and B. Siskind, "The Effect of 

Cure Conditions on the Stability of Cement waste Forms after Immersion in 

Water," Brookhaven National Laboratory Report, WM-3171-4, August 1987. 

A13. American National Standards Institute/American Nuclear Society American 

National Standard for Solid Radioactive Waste Processing System for Light Water 

Cooled Reactor Plants, Appendix 2, March 1979. 

A - 19
Table I 

LIST OF WASTE CONSTITUENTS THAT MAY CAUSE PROBLEMS WITH CEMENT SOLIDIFICATION 

POTENTIAL PROBLEM CONSTITUENTS WHICH MAY BE EXPECTED IN THE WASTE STREAM

Inorganic Constituents 

Borates [1] 

Phosphates [1] 

Lead salts [2] 

Zinc salts 

Ammonia and ammonium salts 

Ferric salts 

"Oxidizing agents" [1] 

(often proprietary) 

Permanganates [1] 

Chromates [2] 

Nitrates [1] 

Sulfates [1]

POTENTIAL PROBLEM CONSTITUENTS THAT MAY 

[4] 

Generic Problem Constituents 

L5] 

Oil [1] and grease 

"Aromatic oils" [1] 

"Organic solvents" [1],[2] 

Dry-cleaning solvents [1],[2] 

"Industrial cleaners" [1],[2] 

Paint thinners [1],[2] 

"Decon solutions" [i] 

Soaps and detergents [1]

Organic Constituents - Aqueous Solutions 

Organic acids [1] 

Formic acid (and formates) 

"Chelates" [1],[3 

Oxalic acid (and oxalates) 

Citric acid (and citrates) 

Picolinic acid (and picolinates) 

EDTA (and its salts) 

NTA (and its salts) 

"Decon solutions"[1] 

Soaps and detergents [1] 

Organic Constituents - Oily Wastes 

Benzene ,2 

Toluene 1] 

Hexane [1] 

Miscellaneous hydrocarbons 

Vegetable oil additives 

BE AVOIDED BY HOUSEKEEPING OR PRETREATMENT 

Specific Problem Constituents - Organic 

Acetone [1],[2] 

Methyl ethyl ketone [2] 

Trichloroethane [2] 

Trichlorotrifluoroethane [2] 

Xylene [2] 

Dichlorobenzene [2] 

Specific Problem Constituents - Inorganic

Sodium hypochlorite [1]

NOTES:

[I] These constituents have been specifically identified by vendors as having 

the potential to cause problems with cement solidification of low-level 

wastes. 

[2] The presence of these constituents may result in the generation of mixed 

wastes. The Environmental Protection Agency should be contacted for 

more information. 

3 All of these chelating agents could also be identified as "organic acids." 

4] Good housekeeping and pretreatment could also be effective in 

preventing problems with cement solidification for many of the 

constituents listed in the top list. 

[5] These specific constituents also fall into several of the "generic" 

problem constituents "categories" listed at the left.

A - 20

ii)