United States         Office of          Teehnie«i Note
            Environmental Protection     Radiation Programs       OHP/TAD-78-2
            Agency           Washington DC 20460      November 1978
            Radiation
&EPA      A Survey Of The Available
            Methods Of Solidification
            For Radioactive Wastes

-------
                                   Technical Note
                                   ORP/TAD-78-2
     A SURVEY OF THE AVAILABLE
     METHODS OF SOLIDIFICATION
       FOR RADIOACTIVE WASTES
                 by

         William F. Holcomb
    Office of Radiation Programs
U.S. Environmental Protection Agency
      Washington, D.C.  20460
           November 1978

-------
                           EPA REVIEW NOTICE
    This report has been reviewed by the Office of Radiation Programs,
U. S. Environmental Protection Agency (EPA) and approved for
publication.  Approval does not signify that the contents necessarily
reflect the views and policies of the EPA.  Neither the United States
nor the EPA makes any warranty, expressed or implied, or assumes any
legal liability or responsibility of any information, apparatus,
product or process disclosed, or represents that its use would not
infringe privately owned rights.
                                   11

-------
                                PREFACE
    The Office of Radiation Programs of the U. S. Environmental
Protection Agency carries out a national program designed to evaluate
population exposure to radiation, and to promote development of
controls necessary to protect the environment and public health.  The
mission of th£ Technology Assessment Division is to provide the primary
assessments of the technologies currently utilized, or being proposed
which have a potential radiation impact on man or his environment.
This report was prepared as a survey and technology assessment of the
currently available and proposed methods of solidification for
radioactive wastes.

    Readers of this report are encouraged to inform the Office of
Radiation Programs of any omissions or errors.  Comments or requests
for futher information are also invited.
                                  ^..'-
                                  David S. Smith
                                     Director
                     Technology Assessment Division (ANR-459)
                           Office of Radiation Programs
                                  iii

-------
                                ABSTRACT
    This report reviews the numerous solidification techniques and
related matrix materials that are presently being offered or proposed
for incorporating the radionuclides into an immobile material.  Both
high- and low-level waste solidification processes are covered.  Key
features of the equipment used in individual solidification processes
are discribed.  At present the high-level waste solidification methods
are being developed by the Government, while the low-level waste
solidification methods are being developed commercially.
                                  iv

-------
                           TABLE OF CONTENTS
EPA Review Notice                                             ii
Preface                                                      iii
Abstract                                                      iv
Table of Contents                                              v
List of Tables                                               vii
List of Figures                                             viii

I. Introduction                                               1

II. Background                                                2

    1.0 Solidification                                        3
    2.0 Waste Classification                                  4
    3.0 Cited References                                      5


III. High-Level Radioactive Waste Solidification Techniques   6

    1.0 Calcination - Solidification Processes                6

         1.1  Fluidized-Bed calcination                       6
         1.2  Spray Calcination                               9
         1.3  Pot Calcination                                 9
         1.4  Rotary Kiln Calcination                        12

    2.0  Cement Solidification                               12

    3.0  Classification - Solidification Processes           14

         3.1  In-Can Melting                                 14
         3.2  Continuous Melting                             15

    4.0  Other Solidification Processes                      18

    5.0  Cited References                                    18

    6.0  Additional References                               22

-------
IV.  Low-Level Radioactive Waste Solidification Techniques   30

     1.0  Solidification of Low-Level Radioactive Wastes
         in Bitumen                                          33

         1.1  Stirred - Evaporator Batch Process             35
         1.2  Emulsified Bitumen Process                     35
         1.3  Film Evaporation Process                       38
         1-4  Screw Extrusion Process                        40
         1.5  Other Research                                 42
         1.6  Advantages and Disadvantages                   42

     2.0  Solidification of Low Level Radioactive Wastes
         in Cement                                           44

         2.1  In-drum Mixer Process                          49
         2.2  External Mixer Process                         52
         2.3  Other Processes                                54
         2.4  Advantages and Disadvantages                   54

     3.0  Polymeric Solidification Processes for Low Level
         Radioactive Wastes                                  56

         3.1  Urea Formaldehyde Method                       56
         3.2  Advantages and Disadvantages                   61
         3-3  Other Polymeric Processes                      62

     4.0  Use of Absorbents for Solidification of Low-Level
Radioactive Wastes                                           64

     5.0  Cited References                                    64

     6.0  Additional References                               70

V.   Leaching Studies                                         72

     1.0  References                                          74

VI.  Conclusions                                              76

Appendix A - EPA Presentation Before National Academy
         of Sciences, February 1, 1977.                      78

Appendix B - President Carter's Statement on Nuclear
         Power Policy, April 7, 1977.                        82

Appendix C - Status of Fuel Reprocessing in the U. S.        85
                                   VI

-------
                             LIST OF TABLES



I.    Solidification Agents and Vendors                      31

II.   Cementation Practices at Various Establishments        4?

III.  Comparisons of Leach Rates for Various Solidified
      Waste Products                                         73
                                  vii

-------
                            LIST OF FIGURES


1.  Fluidized-Bed Calciner                                    8

2.  Spray Calciner                                           10

3.  Pot Calciner                                             11

4.  Rotary KiJ,n Calciner                                     13

5.  In-Can Melter                                            16

6.  Continuous Melter                                        17

7.  Low-Level Waste Processing Steps                         3^

8.  Stirred-Evaporator Batch Process                         36

9.  Emulsified Bitumen Process                               37

10.  Turbulent-Film Evaporation Process                      39

11.  Screw Extruder Evaporation Process                      41

12.  Flow Diagram For Cement Incorporation Processes         50

13.  Example of In-Drum Mixing Process                       51

11.  Example of External Mixing Process                      53

15.  Flow Diagram for UF Incorporation Processes             57

16.  Urea Formaldehyde Incorporation Process - External
     Mixing                                                  58
                                  viii

-------
I.  INTRODUCTION

    It is the intent of this report to review the various
solidification systems and matrix materials that are presently
employed, proposed or under development; to include operational
descriptions, where possible, and also to indicate how these materials
and methods are affected by characteristics of the wastes.  Both high -
and low-level waste solidification processes are reviewed and described.

    This review is being done in accordance with the Environmental
Protection Agency's mission of assessment of the technologies currently
being utilized, or proposed for use in activities which may have a
potential radiation impact on man or his environment.  The EPA has for
many years been involved in radioactive waste management activities,
and in particular the waste characteristics which include inventories,
sources and waste forms.  The Office of Radiation Programs have
undertaken many in-house efforts to address these problems and also to
interact with other agencies and groups to effect efficient solutions
to radioactive waste disposal problems.  Appendix A is the presentation
EPA made before the National Academy of Sciences on February 1, 1977
concerning solidification.

-------
II. BACKGROUND

    The nuclear power and radioisotope industry that has developed in
this century has presented a challenging problem in the management of
radioactive waste materials.  The treatment and disposal of radioactive
wastes involve both environmental considerations and technical
processing problems which are complex and potentially far-reaching
because of the long effective half-lives of certain radionuclides.
Waste management as used here encompasses the treatment and packaging
as well as the disposal (i.e., burial) of the wastes.  Unlike many
industrial wastes, radioactive wastes are not susceptible to
neutralization techniques.  Natural decay is the only means of
destroying radioactivity (the process of transmutation is not
considered viable at this time).  Each radionuclide decays at its own
particular rate regardless of outside influences.  Since the various
waste radionuclides have decay rates ranging from days to thousands of
years,-treatment and processing such as solidification techniques,
become an important factor in radioactive waste management.  As it is
virtually impossible with present technology to destroy the
radioactivity associated with waste materials, containment or dispersal
have been the only two ways of dealing with these unwanted products (1).

    The Environmental Protection Agency (EPA) considers a high degree
of permanent containment of radioactive materials to be the primary
requirement for environmentally adequate radioactive waste management;
i.e., the assurance of isolation of the waste from the environment
(biosphere) for the duration of the wastes hazardous lifetime (2).  As
a result EPA promotes the policy of containment rather than the concept
of planned release or dispersion as currently used in the regulation of
gaseous emissions from nuclear facilities (3).

    EPA recognizes the need for environmental criteria and standards
for waste management.  The Agency was directed to establish
environmental radiation protection standards for the disposal of
high-level radioactive waste as a result of the Presidential Nuclear
Policy Statement of October, 1976 (H).  In implementing EPA's program
for preparing environmental criteria and establishing generally
applicable radioactive waste disposal standards to protect the
environment and public health the Agency's programs center on several
areas,  one of which is the assessment of the performance of engineering
barriers such as solidified waste matrices.  In reviewing the available
information EPA has relied on both the available and future
technologies developed by the Department of Energy and the commerical
industry.

-------
1.0 Solidifioation

    The solidification process has become an important step in the
scheme of waste management and the philosophy of environmental
containment.  This solidification step converts the waste, usually a
liquid or semi-liquid, into a physical form which can be handled,
stored and disposed of more safely and conveniently.  It also reduces
the volume of the waste by removing liquids; and reduces the potential
for movement of the incorporated radionuclides through the geosphere
after disposal.

    There are many varieties of solidification materials and techniques
available.  In all cases the method of solidification used should not
be a reversible process which can allow the solid to return to a liquid
form.  Estimation of the rate of leaching from a solidified matrix
during disposal is one of the important considerations in the
assessment of a solidification method as it will strongly influence the
amount of treatment, containment, and surveillance that will be
needed.  Low matrix solubility will improve the safety of waste
management implemented through isolation by reducing further the
likelihood of an unplanned release.

    In any solidification or stabilization process where radioactive
materials are used, the process and operational conditions are
complicated by the difficulties of remotely operating and maintaining
equipment.  Attention must be given to reliability, and rugged
equipment may have advantages over complex or sensitive equipment.

    Present regulations require that commercially-generated (as opposed
to those generated by the Federal Government) high-level liquid
radioactive wastes be solidified within five years and shipped to a
Federal repository within ten years (5).  The necessity of solidifying
the low-level wastes is not the result of burial considerations, but of
transportation considerations.  The transport of radioactive wastes is
done in compliance with regulations issued by the Department of
Transportation and by the Nuclear Regulatory Commission for the safe
transport of radioactive materials (6, 7).

-------
2.0 Waste Classification

    There are  two classes of radioactive wastes for which
solidification techniques have been developed:  high-level and
low-level.

    The high-level, high-activity radioactive wastes are those liquid
wastes, or  the solidified products of those wastes, which are
associated  with  the first cycle solvent extraction or equivalent in a
facility for processing irradiated reactor fuels.  Research and
development on solidification processes have been underway for
high-level  wastes since the late 1950's.  A number of methods for
processing  these wastes have been proposed and investigated.  These
include conversion of the liquid wastes to a stable calcined powder or
granules, and  conversion of the calcine or liquids to a glass form.
Other methods  include fixation within solids such as clays, cement,
ceramics or similar materials.

    The low-level radioactive wastes include those radioactively
contaminated wastes generated from the nuclear fuel cycle operations
and facilities not specifically designated high-level.  The majority of
these wastes are generated by nuclear power plants and are usually in
the form (prior  to solidification) of processed waste water, evaporator
concentrates,  sludges, filter-aids, demineralizer resins, miscellaneous
scrap material,  etc.  These wastes encompass a broad spectrum of
materials varying widely in chemical and radioactive content.  Various
solidification techniques for these wastes have been demonstrated in
operating nuclear facilities throughout the world for many years.   The
various solidification agents include portland cement, concrete,
plaster of  paris, asphalt (or bitumen), polymers, and a blend of
absorbent material and cement or plaster.

    It should  be noted throughout the following sections that the
solidifcation  processes and techniques developed for the two classes of
wastes, (high  and low-level), are entirely different.  The major
reasons are that the high-level wastes have higher thermal activity,
larger quantities of radioacitivty per volume, and higher radiation
fields than the  low-level wastes.  As a result of these differences the
techniques  for high-level waste require entirely remote operations and
much greater shielding considerations than the low-level waste
techniques; although, in both cases there are common considerations in
the techniques for assessing the potential environmental impact  from
the solidified waste form.  They include such properties as:   (a)
leachability;   (b) thermal conductivity; (c) chemical stability;  (d)
radiation resistence; and (e) mechanical ruggedness.

-------
3.0 Cited References

1.  P.N. Cheremisinoff and W.F. Holoomb, "Management of Hazardous and
Toxic Wastes," Pollution Engineering. Vol. 8, No. 4, page 24, April 1976

2.  W. D. Rowe and W. F. Holcomb, "The Hidden Commitment of Nuclear
Wastes," Nuclear Technology, Vol. 24, No. 3, page 286, December 1974.

3.  Statement -of David G. Hawkins, Assistant Administrator for Air and
Waste Management, Environmental Protection Agency, before the
Subcommittee on Nuclear Regulation of the Committee on Environmental
and Public Works, U.S. Senate, March 22, 1978.

4.  "The White House Fact Sheet — President's Nuclear Waste Management
Plan," Office of the White House Press Secretary, Washington, D.C.,
October 27, 1976.

5.  "Licensing of Production and Utilization Facilities — Policy
Relating to the Siting of Fuel Reprocessing Plants and Related Waste
Management Faciliites," Code of Federal Regulation, Title 10, Chap. I.,
Part 50, Appendix F, U.S. Government Printing Office, Washington,  D.C.,
1976.

6.  "Shipper - General Requirements for Shipments and Packaging," Code
of Federal Regulations, Title 49, Chap. I, Part 173, U.S. Government
Printing Office, Washington, D.C., 1976.

7.  "Packaging of Radioacitve Material for Transport and Transportation
of Radioactive Material Under Certain Conditions," Code of Federal
Regulations, Title  10, Chap. I, Part 71, U.S. Government Printing
Office, Washington, D.C.  1976.

-------
III.  HIGH-LEVEL RADIOACTIVE WASTE SOLIDIFICATION TECHNIQUES

    High-level liquid radioactive wastes are produced during spent-fuel
recovery operations at reprocessing plants.  These wastes are the
highly radioactive solutions that remain after the separation of the
unused (or unfissioned) uranium from the fission product waste produced
during reactor operation.  Reprocessing of spent fuels from the
commercial nuclear power industry is not being done in the United
States at present (1).  See Appendix B for text of statement.  However,
a substantial amount of these wastes are on hand from past reprocessing
of  defense and commercial materials (2-10).  See Appendix C for status
of  fuel reprocessing.

    Extensive research and development has taken place throughout the
world for the past 20 years on possible alternatives for processing
these high-activity wastes (11-14).  Many of these techniques have been
successfully demonstrated on either pilot-plant or full scale
operation.  These alternatives have resulted in a number of
solidification technologies, of which two seem to have emerged as the
most prominent - calcination and glassification.  While both
solidification technologies are discussed as possible alternatives it
is  expected that they will complement one another by using the
calcination step as a conversion process from liquid to solid (with
volume reduction included) and the glassification step will provide a
vitrified solid for terminal disposal operations.

1.0  Calcination-Solidification Processes

    Calcination is a process that heats a material to a temperature
which will drive off most volatile material, but not hot enough to
cause fusing of the material.  Four major U. S. and foreign
technologies relate to the calcination process; they are fluidized-bed,
spray, pot, and rotary kiln calcination (15-20).

1.1  Fluidized-Bed Calcination

    The fluidized bed calcination process was the earliest technique
investigated for the conversion of radioactive waste solutions to
solids (15, 21).  Development of this process was sponsored under an
Atomic Energy Commission (now the Department of Energy) program which
started in 1955.  A Waste Calcining Facility (WCF) was built at the
Idaho National Engineering Laboratory near Idaho Falls, Idaho, as part

-------
of the Idaho Chemical Processing Plant.  It is the world's only plant
scale unit which converts high-level radioactive liquids to solids
using fluidized bed calcination (3, 11, 14, 15, 21-24).  From 1963
through 1977 the WCF converted about 12 thousand cubic meters of
high-level radioactive liquid wastes to about 1500 cubic meters of
granular calcine solids (5).  Solidification of the nitric acid fuel
dissolution wastes began at a net rate of 227 liters per hour and has
been increased to about 3^0 liters per hour.   Wastes from sulfuric
acid and hydrofluoric acid fuel dissolutions also have been solidified.

    The liquid waste is pneumatically atomized and sprayed into a
fluidized bed of granular solids operating at a temperature of
400-500 C.  Here the liquids are vaporized and the gaseous materials
enter an off-gas system, while the metallic salts are converted to
their corresponding oxides or fluorides and deposited layerwise on the
spherical bed particles.  The solid particles are removed continuously
from the fluidized bed and transported pneumatically to storage
facilities.  The waste solutions calcine to such compounds as
A1203, CaF2, ZK>2, Fe203, Al2(8011)3, etc and also
certain amounts of amorphous materials; while water vapor, oxides of
nitrogen, and other volatiles leave with the calciner off-gas.  During
calcination of hydrofluoric acid-type waste, the feed also includes the
additon of calcium nitrate to complex the fluoride ion thus preventing
the formation of corrosive fluoride gases.

    The volumetric ratio of atomizing air to liquid feed in the waste
nozzle is varied as necessary to cause sufficient breakup of bed
particles to control the average bed particle size at a desired value.
Heat for the endothermic calcination process is supplied by in-bed
combustion of kerosene injected and atomized by oxygen through fuel
nozzles located in the walls of the calciner vessel (22, 23).  Startup
of the process requires heating the fluidized bed to greater than 360 C
using heated fluidizing air.  The liquid wastes containing nitrate are
then injected through a waste atomizing nozzle, followed immediately by
the injection of the fuel-oxygen mixture through the fuel atomizing
nozzle.  Ignition of the fuel-oxygen mixture is spontaneous at
temperatures above 270 C in the presence of nitrates.  Off-gas from the
calciner is cleaned by successive passage through a dry cyclone, a wet
venturi scrubbing system, silica gel absorbers, and high efficiency
particulate filters to remove contaminants before the gas is exhausted
to the atmosphere.  Figure 1 shows a fluidized bed calciner as used at
the Idaho Facility.

-------
   WASTE FEED
    NOZZLE
ATOMIZING AIR

      OXYGEN
 KEROSENE TO
 FUEL NOZZLE
 FLUIDIZING
    AIR
               CALCINER
                VESSEL
•*:*:•»:•:•:•:•:
                                                TO OFF-GAS CLEANUP
                                                     SYSTEM
                                                  TO CYCLONE
                                                   FOR FINES
                                                  REMOVAL TO
                                                PRODUCT STORAGE
           PRODUCT OVER FLOW
              TO STORAGE
                            FIGURE 1

                    FLUIDIZED-BED CALCINER

-------
1.2  Spray Calcination

    This process has been under development at the Department of
Energy's Hanford Reservation, Richland, Washington, for over 15 years
and is presently undergoing tests by the Battelle Pacific Northwest
Laboratories using simulated wastes (11, 15, 18-20, 25).

    The liquid waste is fed to a pneumatic air atomizing nozzle in the
top of a heated (700-800 C) spray calciner vessel.  Convective and
radiant heat transfer flash dries and calcines the atomized droplets as
they fall through the chamber.  Over half the calcine falls into an
outlet cone while the balance collects as a coating on sintered
stainless steel filters located in the vessel gas outlet.
Periodically the filter coating,is blown back to dislodge the powder
which falls into the cone.  Figure 2 illustrates a spray calciner
vessel.  The material in the cone is discharged to either canisters or
into a close-coupled melter.  The process has operated at feed rates
from 75 to 300 liters per hour.  The gaseous and liquid effluents can
be treated by the usual off-gas purification type equipment, i.e., wet
scrubbers, quenching spray tower, venturi scrubber, cyclone, condenser,
knockout pot, silica gel sorption towers etc.  A variation of the spray
calcining process has been under devlopment and in use at the Karlsruhe
(W. Germany) Nuclear Research Center since the early 1970s (26, 27).

1.3  Pot Calcination

    This process was developed at the Department of Energy's Oak Ridge
National Laboratory and was demonstrated at the Hanford Reservation
facilities (11, 15, 18, 19, 25).

    The liquid waste is fed to a heated pot located in a multiple-zone
heating and cooling furnace.  The pot also serves as the final storage
vessel.  As the liquid waste concentrates, it begins to form a scale on
the inside walls of the pot.  As calcination continues, the scale grows
in thickness while the feed rate decreases and finally is stopped.  At
this point, the scale has grown inward and upward to fill the pot,
except for the cone-shaped void at the top.  Heating is continued until
the liquid is boiled to dryness and all the waste has been calcined at
temperatures up to 900 C.  The pot is then cooled, removed, sealed and
stored.  The total time cycle for filling, cooling and changeout of an
20 cm. dia pot with 183 cm of calcine is about U4 hours at an average
processing rate of 11 liters per hour.  Figure 3 shows the pot calciner.

-------
                     10
                           ATOMIZING AIR

                                 AND
                           WASTE FEED NOZZLE
  FURNACE
               CALCINER
               CHAMBER
                                             OFF-GAS
                                             OUT
VIBRATOR
                                   SINTERED STAINLESS
                                   STEEL FILTERS
               CALCINE POWDER OUT


                     FIGURE 2

                 SPRAY CALCINER

-------
                             11
WASTE, PLUS ADDITIVES
    IF REQUIRED
OFF GAS
ZONED FURNACE
                      POT
                            FIGURE 3

                         POT CALCINER
                                               PROCESSING AND
                                               FINAL STORAGE CANISTER
                                                 BOILING LIQUID
                                                 CALCINE SCALE
                                                 FORMATION

-------
                                   12

1.4  Rotary Kiln Calcination

    This process was developed in the early  '60s at the Department of
Energy's Brookhaven National Laboratory.  In the rotary-ball kiln
calciner the waste solution is dripped or sprayed onto a bed of moving
metal balls in a slowly rotating cylinder, electrically-heated to a
temperature between 600  and 800 C.  The product formed from deposition
of the oxides of the metallic salts on the balls is pulverized during
kiln operation.  The capacity of a  rotary-ball kiln calciner is based
on the surface to bed heat transfer coefficient obtainable and
effectively dispersing heat throughout the length of the calciner
vessel.  It is estimated that the capacity could be as high as 760-1130
liters per hour (11, 15, 20). Figure 4 illustrates a rotary kiln
calciner.

    The French Commissariat a 1'Energie Atomique facility at Marcoule
is using a variation of the rotary kiln process in connection with a
continuous melter process for the conversion of high-level liquid
wastes to a glass.  The feed rate is expected to be 30-40 liters/hr
(28).  All recent development and commercialization activities
concerning this process has been accomplished at the Marcoule facility.

2.0  Cement Solidification

    The use of concrete as a matrix for solidification of the
high-level radioactive wastes at DOE's Savannah River Plant (SRP) has
been studied on an experimental, laboratory-scale program using both
simulated and actual radioactive waste sludges (6, 29-33).  Because of
dilution and aging, the SRP high-level waste is suitable for fixation
in cement.

    The studies indicated that formulations of cement with washed and
dried sludge powders are easy to prepare, and the waste is chemically
compatible with concrete.  Sludge loadings up to 40 vit% give waste
forms with generally excellent properties.  The concrete matrix
properties studied were: (a) compressive strength; (b) cesium,
strontium and alpha emitter leachability; (c) radiation and thermal
stability; and (d) types of cement and sludge compositions used.  The
initial results indicated compressive strengths of 2000-3000 psi for
formulations with 40$ sludge content: reasonable long-term thermal and
radiation stability; and strontuim ~90 and alpha emitter
leachabilities typically in the 10~5 g/cm^-d or greater range.

-------
                               3% SLOPE
     WASTE
  PLUS ADDITIVES
   IF REQUIRED
DRIVE GEAR
         ZONED FURNACE
                                     FIGURE 4

                             ROTARY KILN CALCINER
OFF GAS
                                                                    TO PRODUCT
                                                                    RECEIVER

-------
                                    14

    Some problems encountered were  high cesium - 137 leachibility  (less
 than  10~3 g/cm^-d); steam generation if heated in closed
 containers, and excessive cement set times; although some methods  were
 developed to  overcome  these problems.

 3.0   Classification -  Solidification Processes

    In  connection with the calcination/solidification technique  of
 immobilizing  the high-level liquid  radioactive wastes, the U.S.  and  the
 other countries using  nuclear power have directed their research and
 development toward the goal of glassification (or use of similar
 vitreous forms) of these wastes for disposal in geological
 repositories.

    The use of glass or glasslike materials provides a highly  immobile
 solid form for fixing  the wastes.   In addition, it reduces air-borne
 releases, provides a less leachable material, is chemically, thermally,
 and radiolytically stable, and provides a suitable heat dissipating
 solid for geologic isolation.  Borosilicate glass appears to be  the
 preferred form in most glass programs.  It has relatively low
 temperatures  of formation, low corrosiveness to container material
 during  formation, high retention of volatile species during processing,
 and low leach rates (11, 14, 16-18, 34).  There are two processes  which
 have  been developed in the U.S. as  candidates for commercial use:
 in-can  melting and continuous melting.

 3.1   In-Can Melting

    The in-can melting batch process is being developed by Battelle's
 Pacific Northwest Laboratories for  the Department of Energy.   The
 process was developed  as a tandem system coupled to the spray  or
 fluidized bed calcination process.

    The calcine powder falls directly into a close-coupled melter
 canister along with specially-formulated frit.  The frit and the
 calcine are melted together in a metallic canister at processing
 temperatures of 1000-1100 C using a multizone furnace.

    The advantages include:  (a) minimum process steps and equipment;
 (b) non-transfer of melt; (c) everything entering the melter  (except
some volatile species)  is fixed in  the glass; (d) disposable melter;
and (e)  the melting step is flexible enough to accommodate a wide  range

-------
                                    15

of calcine characteristics and compositions, i.e., it could also use
the calcine product from other processes such as spray or  rotary kiln
calcination.  Melting rates up to  100 kg/hr have been achieved using
nonradioactive simulated high-level waste.  Figure 5 illustrates the
in-can melting process.

    Even through the melting is a  batch process and the calcination
step is a continuous process, by using a diverter valve and multiple
melting furnace-canisters, the systems are compatible (11, 19, 25,
35-38).

3.2  Continuous Melter

    Battelle's Pacific Northwest Laboratories also have under
development a continuous melter process (also called joule heated
melter).  This process is closely  allied with the commercial electric
glass melter processes and its flexibility allows it to be coupled to
different kinds of waste calciners and even receive liquid wastes
directly.

    The system has a long life time while producing a high quality
glass with low off-gas effluents.   The process is carried  out at
temperatures ranging from 1000 to  1200 C in a refactory-lined melter
with internal electrodes; the molten glass acts as its own electrical
resistance heating element.  Prototypes have operated at feed rates up
to  100 kg/hr.  The system also allows for flexibility in glass
composition and for the controlled draining of glass from  the melter
which can permit changes in the final waste form package (11, 18, 25,
36, 38).  Figure 6 shows the continuous melter.

    The French are developing a system using the rotary calciner
coupled to a melting furnace which is scheduled for full scale
operation in 1978.  In addition, the West German Government is
developing systems using the spray calciner coupled to a continuous
melter (27, 28).

-------
                          16
                                z
CALCINE
                                      GLASS FRIT ADDITION
                           DIVERTER VALVE
•

=

•
•
•
•
•
•

(/
-x-
mmmmmmmmm^
—
=

•H
STORAGE
CANISTER
IN-CAN
FURNACE
                        FIGURE 5
                    IN-CAN MELTER

-------
                   CALCINE OR LIQUID WASTE
                            AND
                         GLASS FRIT
                                     OFF-GAS
                       MOLTEN GLASS
                           AND
                       WASTE MIXTURE
ELECTRODES
MOLTEN GLASS
     TO
  STORAGE
  CANISTER
                          FIGURE 6

                    CONTINUOUS MELTER

-------
                                    18

^•0  Other Solidifioation Processes

    There are numerous other processes and methods being  investigated
as alternatives to calcination and glassification to reduce costs,
simplify processing operations, increase waste form compatibility,  and
to increase  the reliability of the inertness and low dispersibility of
waste  form.  Many of  these processes are under laboratory-scale
development  while others are variations of the present techniques  (3,
11, 13, 1U).  The various alternatives have not been under development
as long as the existing glassification and calcination processes,  and
thus the best combinations and options have not been optimized.  They
include:

    (1) Supercalcine  - a process to produce an assemblage of
    thermodynamically stable crystalline ceramic waste form using
    additives.

    (2) Sintered calcine - a process to produce a sintered glass and
    crystalline phase product, using fluxes and binders.

    (3) Metal Matrices - a process to produce a high thermal
    conductivity monolithic form of calcine, vitreous beads, or pellets
    cast in  molten metal or embedded in sintered metal.

    (1) Glass/Ceramic - a process producing a glass subjected to
    controlled crystallization, thus ending up with a fine grained
    crystalline body  product.

    (5) Coated Pellets - a process to take calcine powders and coat
    them with a durable material such as A^Og or glass and sinter
    them to  produce stable pellets.

    (6) Aqueous silicate - a process to convert wastes to solid,
    relatively insoluble, aluminosilicate minerals.  Powdered clays
    such as  kaolin or bentonite are mixed with, and allowed to react
    with,  aqueous solutious or slurries of caustic wastes at
    temperatures ranging from 30-100 C.

5.0  Cited References

    1.   Statement by  President J. E. Carter on "Nuclear Power Policy,"
    April 7,  1977,  White House, Washington, D. C. (see also Nucleonics
    Week,  Vol. 18,  No. 15,  April lU, 1977).                         '

-------
                               19

2.  Report of Task Force for Review of Nuclear Waste Management -
Draft. Report DOE/ER-0004/D, U. S. Department of Energy,
Washington, D. C., February 1978.

3.  Alternatives  for Long-Term Management of Defense High-Level
Radioactive Waste, Idaho Chemical Processing plant, Report ERDA
77-43, U. S. Energy Research and Development Administration,
Washington, D. C., September 1977.

4.  Final Environmental Impact Statement, Waste Management
Operations, Idaho National Engineering Laboratory, Idaho, Report
ERDA-1536, U. S.  Energy Research and Development Administration,
Washington, D. C., September 1977.

5.  INEL Waste Management Plan for FY-1978 Idaho National
Engineering Laboratory, Report IDO-10051, U. S. Energy Research and
Development Administration, Idaho Falls, Idaho, August 1977

6. (a)   Alternatives for Long-Term Management of Defense
High-Level Radioactive Waste, Savannah River Plant, Report ERDA
77-42, U. S. Energy Research and Development Administration,
Washington, D. C., May 1977.  (b)  Environmental Statement, Waste
Management Operations, Savannah River Plant, Aiken, South Carolina,
Report ERDA-1537, U. S. Energy Research and Development
Administration, Washington, D. C., September 1977

7.  Alternatives  for Long-Term Management of Defense High-Level
Radioactive Waste, Hanford Reservation, Report ERDA 77-44, U. S.
Energy Research and Development Administration, Washington, D. C.,
September 1977.

8.  Final Environmental Statement, Waste Management Operations,
Hanford Reservation, Richland, Washington, Report ERDA-1538, 2
Volumes, U. S. Energy Research and Development Administration,
Washington, D. C., December 1975.

9.  Alternative Processes for Managing Existing Commerciaj.
High-Level Radioactive Wastes, Report NUREG-0043, U. S. Nuclear
Regulatory Commission, Washington, D. C., April 1976.

10.  G. J. Dau and R. F. Williams, Status of Commercial Nuclear
High-Level Waste  Disposal, Report EPRI-NP-44-SR, Electric Power
Research Institute, Palo Alto, California, September 1976.

-------
                               20

11.  Alternatives for Managing Wastes from Reactors and Post
Fission Operations in the LWR Fuel Cycle, Volume 2;  Alternatives
for Waste Treatment, Report No. ERDA-76-43, U. S. Energy Research
and Development Administration, Washington, D. C., May 1976.

12.  High-Level Radioactive Waste Management Alternatives, Report
No. WASH-1297, U. S. Atomic Energy Commission, Washington, D. C.,
May 1974.

13.  W. J. George, "Treating and Disposing of Radioactive Wastes,"
Chemical Engineering, page 151, December 14, 1959-

14.  "Management of Radioactive Wastes," Atomics, page 16,
May/June, 1965.

15.  B. R. Wheeler, et al., A Comparison of Various Calcination
Processes For Processing High-Level Radioactive Wastes, Report
IDO-1422, U. S. Atomic Energy Commission, Idaho Falls, Idaho, April
1964.

16.  R. E. Tomlinson (Ed), Radioactive Wastes From the Nuclear Fuel
Cycle, Symposium Series 154, Volume 72, Americal Institute of
Chemical Engineers, New York, N. Y., 1976.

17.  Management of Radioactive Wastes from Fuel Reprocessing,
Proceedings of a Symposium organized jointly by the OECD Nuclear
Energy Agency and-the International Atomic Energy Agency, Paris,
November 27-December 1, 1972.

18.  A. M. Platt and C. R. Cooley, "Waste Solidifiation Prototype
Program," Atomics, page 22, July/August 1965.

19.  K. J. Schnieder, "Solidification of Radioactive Waste,"
Chemical Engineering Progress, Vol. 66, No. 2, page 35, February
1970.

20.  B. R. Dickey, et al., "High-Level Waste Solidification:
Applicability of Fluidization-Bed Calcination to Commercial
Wastes," Nuclear Technology, Vol, 24, No. 3, p. 371, December 1977.

21.  W. F. Holcomb, "Uses of the Fluidization Bed Process,"
Combustion,  Vol 48, No. 10, page 31, April 1977.

-------
                               21

22.  J. A. Wielang and W. A. Freeby, The Fifth Processing Campaign
in the Waste Calcining Facility-FY 1972. USAEC Report No. ICP-1021,
Idaho National Engineering Laboratory, Idaho Falls, Idaho, June
1973.

23.  W. F. Holcomb and J. A. Wielang, "02 for Calcification of
Radioactive Wastes," Cryogenics and Industrial Gases, Vol. 8 No. 4,
page 24, July/August 1973.

24.  L. T. Lakey and B. R. Wheeler, "Solidification of High-Level
Radioactive Wastes at the Idaho Chemical Processing Plant," in
reference 17, page 731.

25.  A. G. Blasewitz et al., "The High-Level Waste Solidification
Program:, in reference 17, page 615.

26.  F. Kaufmann, et al., "Recent Experience with a Steam-Heated
Spray Calcining Unit for High-Level Waste Solidification," in
reference 16, page 132.

27.  W. Guber et al., "Pilot Plant Experience on High-Level Waste
Solidification and Design of the Engineering Prototype VERA," in
reference 17, page 489-

28.  R. Bonniaud et al., "French Industrial Plant AVM for
Continuous Vitrification of High-Level Radioactive Wastes", in
reference 16, page 145.

29.  J. A. Stone, Evaluation of Concrete as a Matrix for
Solidification of Savannah River Plant Waste, USERDA Report
DP-1448, Savannah River Laboratory, Aiken, South Carolina, June
1977.

30.  M. J. Plodinec, Evaluation of Cs-137 Sorbents for Fixation in
Concrete, USERDA Report DP-1444, Savannah River Laboratory, Aiken,
South Carolina, February 1977.

31.  N. E. Bibler, Radiolytic Gas Production From Concrete
Containing Savannah River Plant Waste, USDOE Report DP-1464,
Savannah River Laboratory, Aiken, South Carolina, January 1978.

32.  J. A. Stone and P. D. d'Entremont, Measurement and  Control of
Cement Set Times in Waste Solidification, USERDA Report  DP-1404,
Savannah River Laboratory, Aiken, South Carolina, September 1976.

-------
                                   22

    33.  J. A. Stone, "Evaluation of Concrete as a Matrix for
    Solidification of Savannah River Plant Waste," presented at the
    80th Annual meeting of the American Ceramic Society,  Detroit,
    Michigan, May 6-11, 1978.

    34.  Ceramic and Glass Radioactive Waste Forms, Workshop Summary,
    Report No. CONF-770102, U. S. Energy Research and Development
    Administration, Washington, D. C., January U-5, 1977.

    35.  R. B. Keely et al., "Technology Status of Spray
    Calcination/Vitrification of High-Level Liquid Waste for Full-Scale
    Application," Presented at the 70th Annual Meeting of the American
    Institute of Chemical Engineers, New York, November 13-17, 1977.

    36.  C. C. Chapman et al., "Experience with Waste Vitrification
    Systems at Battelle-Northwest," in reference 16, page 151.

    37.  D. E. Larson and W. F. Bonner, High-Level Waste Vitrification
    by Spray Calcination/In-Can Melting, USAEC Report BNWL-2092,
    Battelle Pacific Northwest Laboratories, Richland, Washington,
    November 1976.

    38.  J. E. Mendel, "Glass as a Waste Form:  Overview and Processing
    Considerations," in reference 29, page 29.

    39.  R. N. Rickles, "The Encapsulation of High-Level Nuclear
    Wastes," Atomics, page 50, March/April 196U.

6.0  Additional References

    I.  J. E. Mendel, et al, Annual Report on the Characteristics of
    High-Level Waste Glasses, USERDA Report BNWL-2252, Battelle Pacific
    Northwest Laboratories, Richland, Washington, June 1977.

    2.  W. J. Gray, Volatility of a Zinc Borosilicate Glass Containing
    Simulated High-Level Radioactive Waste, USERDA Report BNWL-2111,
    Battelle Pacific Northwest Laboratories, Richland, Washington,
    October 1976.

    3.  H. T. Blair, Vitrification of Nuclear Waste Calcines by In-Can
    Melting. USERDA Report BNWL-2061, Battelle Pacific Northwest
    Laboratories, Richland, Washington, May 1976.

-------
                               23

4.  J. E. Mendel, "High-Level Waste Glass,"  Nuclear Technology,
Vol. 32, No. 1, page 72, January 1977.

5.  J. A. Kelley, Evaluation of Glass as a Matrix for
Solidification of Savannah River Plant Waste, USERDA Report
DP-1397, Savannah River Laboratory, Aiken, South Carolina, October
1975.

6.  Management of Radioactive Wastes From the Nuclear Fuel Cycle,
Proceedings of a Symposium jointly organized by the IAEA and NEA
(OECD), Vienna, Austria, March 22-26, 1976.

7.  C. C. Chapman and J. L. Buelt, "Vitrification of High-Level
Waste  in a Joule Heated Ceramic Melter," Presented at the 70th
Annual Meeting of the American Institute of Chemical Engineers, New
York,  November 13-17, 1977.

8.  J. R. La Riviere and E. L. Moore, Preliminary Canister Size
Criteria for Commercial Solidified High-Level Waste, USERDA Report
ARH-ST-151, Atlantic Richfield Hanford Company, Rlchland,
Washington, April 1977.

9.  R. P. Turcotte, Radiation Effects in Solidified High-Level
Wastes. Part 2-Helium Behavior, USERDA Report BNWL-2051, Battelle
Pacific Northwest Laboratories, Richland, Washington, May 1976.

10.  G. J. McCarthy, "High-Level Waste Ceramics: Materials
Considerations, Process Simulation, and Product Characterization,"
Nuclear Technology, Vol, 32, No. 1, p 92, January 1977.

11.  W. K. Eister,  "Materials Considerations in Radioactive Waste
Storage," Nuclear Technology, Vol 32, No. 1, page 6, January 1977.

12.  S. C. Slate and W. A. Ross, "High-Level Radioactive Waste
Glass  and Storage Canister Design," Presented at the 70th Annual
Meeting of the American Institute of Chemical Engineers, New York,
November 13-17, 1977.

13.  R. E. Schindler, et al, Development of a Fluidized-Bed
Caloiner and Post-Treatment Processes for Solidification of
Commercial Fuel-Reprocessing Liquid Wastes, USDOE Report ICP-1136,
Idaho  National Engineering Laboratory, Idaho Falls, Idaho, December
1977.

-------
                               24

14.  J. A. Wielang, et al, The Fourth Processing Campaign in the
Waste Calcining Facility FY 1971. USAEC Report ICP-1004, Idaho
National Engineering Laboratory, Idaho Falls, Idaho, March 1972.

15.  W. A. Freeby, Interim Report; Fluidized-Bed Calcination of
Simulated High-Level Commercial Wastes, USERDA Report ICP-1075,
Idaho National Engineering Laboratory, Idaho Falls, Idaho, June
1975.

16.  W. J. Bjorklund, Fluidized Bed Calcination Experience with
Simulated Commercial High-Level Nuclear Waste. USERDA Report
BNWL-2138, Battelle Pacific Northwest Laboratories, Richland,
Washington, November 1976.

17.  G. L. Tingey and W. D. Felix, Radiolytic Gas Generation from
Calcined Nuclear Waste, USERDA Report BNWL-2381, Battelle Pacific
Northwest Laboratories, Richland, Washington, August 1977.

18.  J. R. Berreth, "Stabilization and Storage of Solidified
High-Level Radioactive Wastes," Nuclear Technology, Vol. 32, page
16, January 1977.

19.  Current Activities in DOE's Commercial Waste Management
Program, Report DOE/ET-0042, U. S. Department of Energy,
Washington, D. C., March 1978.

20.  W. P. Bishop and F. J. Miraglia, Jr., Eds., Environmental
Survey of the Reprocessing and Waste Management Portions of the LWR
Fuel Cycle, Report NUREG-0116 (Supp. 1 to WASH-1248), U. S. Nuclear
Regulatory Commission, Washington, D. C., October 1976.

21.  The Waste Calcining Facility at the Idaho Chemical Processing
Plant, U. S. Atomic Energy Commission, Idaho Falls, Idaho, 1970.

22.  W. F. Bonner, et al, Spray Solidification of Nuclear Waste,
USERDA Report BNWL-2059, Battelle Pacific Northwest Laboratories,
Richland, Washington, August 1976.

23.  W. A. Ross, Development of Glass Formulations Containing
High-Level Nuclear Waste. Report PNL-2481, Battelle Pacific
Northwest Laboratories, Richland, Washington, February 1978.

-------
                               25

24.  T. H. Smith and W. A. Ross, Impact Testing of Vitneous
Simulated High-Level Waste in Canisters Report BNWL-1903, Battelle
Pacific Northwest Laboratories, Richland, Washington, May 1975.

25.  "A New Glass Container for the Safe Encapsulation of
Radioactive Wastes from Nuclear Power Plants Has Been Invented by
Three Catholic University Professors,"  Atomic Energy Clearing
House. Vol. 23, No. 43, page 9, Washington, D. C., October 24, 1977.

26.  The Management and Storage of Commercial Power Reactor Wastes,
Report ERDA-76-162, U. S. Energy Research and Development
Administration, Washington, D. C., 1976.

27.  G. Malow and W. Lutze, "Sodium diffusion in borosilicate
glasses.  Part 1. Influence of glass composition," and "Part 2. Ion
dipole interactions," Physics and Chemistry of Glasses, Vol. 16,
No. 6, December 1975.

28.  W. Heimerl, et al, Research on Glasses for Fission Product
Fixation, Summary Report January 1968-June 1971, Report No.
HMI-B-109, Hahn-Mertner-Institute Berlin GmbH, Federal Republic of
Germany, September 1971.

29.  D. W. Clelland, et al, "Design of a Plant for the
Incorporation of Highly Active Wastes Into Glass," presented at the
60th Annual Meeting of the American Institute of Chemical
Engineers, New York, N.Y., November 26-30, 1967.

30.  F. Laude, "Pot Vitrification of Concentrated Fission Product
Solutious," symposium proceedings on Solidification and Long-Term
Storage of Highly Radioactive Wastes, U. S. Atomic Energy
Commission, Richland, Washington, 1966.

31.  P. R. Savage, "Nuclear Waste disposal: politics closed
prospects," Chemical Engineering, Vol. 84, page 72, June 20, 1977.

32.  M. D. Alford, Fluidized Bed Calcination of Simulated Purex
High-Activity Level Wastes, Interim USAEC Report No. HW-66384 RD,
General Electric Hanford Atomic Products Operation, Richland,
Washington, September I960.

33.  M. D. Alford and K. J. Schneider, Prototype Waste Calciner.
USAEC Report No. HW-62631, General Electric Hanford Atomic Products
Operation, Richland, Washington, November 1959.

-------
                               26

34.  W. A. Ross and J. E. Mendel, "Glass Waste Forms for
Radioactive Waste Containment," Presented at the 70th Annual
Meeting of the American Institute of Chemical Engineers, New York,
November 13-17, 1977.

35.  K. M. Lamb, Final Report; Development of a Metal Matrix for
Incorporating High-Level Commercial Waste, USDOE Report ICP-1144,
Allied Chemical Corporation, Idaho Chemical programs, Idaho Falls,
Idaho, March 1978.

36.  "Waste Solidification Gains Major Attention," Nucleonics, Vol,
21, No. 2, page 58, February 1963.

37.  J. E. Mendel, et al., A Program Plan for Comprehensive
Characterization of Solidified High-Level Wastes, USAEC Report
BNWL-1940, Battelle Pacific Northwest Laboratories, Richland,
Washington, December 1975.

38.  A. K. De, et al., "Development of Glass Ceramics for the
Incorporation of Fission Products."  Ceramic Bulletin, Vol. 55, No.
5, page 500, 1976.

39.  G. H. Thompson, Interim Solidification of SRP Waste With
Silica, Bentonite. or Phosphoric Acid, OSERDA Report DP-1403,
Savannah River Laboratory, Aiken, South Carolina, March 1976.

40.  C. M. Slansky, Ed., Technical Division Quarterly Progress
Report, October-December 1977, USDOE Report ICP-1141, Idaho
National Engineering Laboratory, Idaho Falls, Idaho, February 1978.

41.  A. M. Platt (Compiler), Nuclear Waste Management Quarterly
Progress Report, July through September 1977.  USDOE Report
PNL-2377-3, Battelle Pacific Northwest Laboratories, Richland,
Washington, January 1978.

42.  J. L. McElroy, Quarterly Progress Report, REsearch and
Development Activities Waste Fixation Program, January through
March 1977. USDOE Report PNL-2265-1. Battelle Pacific Northwest
Laboratories, Richland, Washington November 1977.

-------
                               27

43-  Management of Wastes from the LWR Fuel Cycle, Proceedings of
the International Symposium, Denver, Colorado, Report CONF-76-0701.
U. S. Energy Research and Development Administration, Washington,
D. C.,  July 11-16, 1976.

44.  T. English, et al, An Analysis of the Technical Status of    /
High-Level Radioactive Waste and Spent Fuel Management Systems, V
Report JPL-77-69, Jet Propulsion Laboratory, Pasadena, California,
December 1977.

45.  "Waste Management:  Technological Advances and Attitudes of
Safety," Nuclear News, Vol 7, No. 10, p 94, October 1964.

46.  Radioactive Waste Management - A Bibliography of Publicly
Available Literature Pertaining to the USAEC's Hanford, Washington
Production Site, Report TID-3340, U. S. Atomic Energy Commission,
Oak Ridge, Tennessee, August 1973.

47.  Radioactive Waste Management - A Bibliography of Publicly
Available Literature Pertaining To the USAEC's Savannah River
Plant, South Carolina, Report TID-3341, U. S. Atomic Energy
Commission, Oak Ridge, Tennessee, August 1973.

48.  Radioactive Waste Management - A Bibliography of Publicly
Available Literature Pertaining to the USAEC's National Reactor
Testing Station, Idaho, Report TID-3342, U. S. Atomic Energy
Commission, Oak Ridge, Tennessee, August 1973.

49.  W. G. Belter and D. W. Pearce, "Radioactive Waste Management,"
Reactor Technology - Selected Reviews - 1965, Report TID-8541, U.
S. Atomic Energy Commission, Oak Ridge, Tennessee, January 1966.

50-  C. L. Bendixsen, et al, The Third Processing Campaign in the
Waste Calcining Facility, USAEC Report IN-1474, Idaho National
Engineering Laboratory, Idaho Falls, Idaho, May 1971.

51.  Siting of Fuel Reprocessing Plants and Waste Management
Facilities, USAEC Report ORNL-4451, Oak Ridge National Laboratory,
Oak Ridge, Tennessee, July 1970.

52.  G. E. Lohse, et al, "Calcination of Zirconium Fluoride Wastes
in Pilot Scale and Plant Scale Fluidised Beds," International
Chemical Engineering Symposium Series No. 30 London, 1968.

-------
                               28

53-  K. J. Schneider, Editor, Waste Solidification Program,
     Volume  1            Report BNWL-1073          August 1969
     Volume  2            Report BNWL-968           February 1969
     Volume  3            Report BNWL-832           December 1968
     Volume  4            Report BNWL-814           December 1968
     Volume  5            Report BNWL-1185          January 1970
     Volume  6            Report BNWL-1391          August 1970
     Volume  7            Report BNWL-1541          January 1971
     Volume  8            Report BNWL-1583          June 1971
     Volume  9            Report BNWL-1628          January 1972
Battelle Pacific Northwest Laboratories, Richland, Washington.

54.  J.N.C.  van Geel, et al, "Conditioning High-Level Radioactive
Waste," Chemical Engineering Progress, Vol 72, No. 3, P 49, March
1976.

55.  R. E. Blanco, et al, "Solving the Waste-Disposal Problem,"
Nucleonics,  Vol 25, No. 2, p 58, February 1967.

56.  "Calcining Techniques to Ease Nuclear-Waste Woes," Chemical
Engineering, p 26, April 1, 1963.

57.  "A Glass Trap for Nuclear Wastes," Chemical Engineering, p 84,
July 22, 1963.

58.  E. D. Cooper et al, "Pilot - and Plant-Scale Fluidized Bed
Calciners",  Chemical Engineering Progress, Vol 61, No. 7, P 89,
July 1965.

59.  "AEC Well Along on Waste Disposal," Nucleonics, p 78, February
1964.
           /
60.  B. L. Cohen, "The Disposal of Radioactive Wastes from Fission
Reactors", Scientific American, Vol 236, No. 6, p 21, June 1977.

61.  K. M. Harmon, Summary of National and International
Radioactive Waste Management Programs, USDOE Report PNL-2598,
Battelle Pacific Northwest Laboratories, Richland, Washington,
March 1978.

-------
                               29

62.  J. R. Berreth, et al., Status Report; Development and
Evaluation of Alternative Treatment Methods for Commercial and ICPP
High-Level Solidified Wastes. USERDA Report ICP-1089, Idaho
National Engineering Laboratory, Idaho Falls, Idaho, May 1976.

63.  J. A. Powell, (Ed), Bibliography of PNL Publications in
Managment of Radioactive Wastes, USERDA Report BNWL-2201, Battelle
Pacific Northwest Laboratories, Richland Washington, July 1976.

64.  W. A. Ross, et al, Annual Report on the Characterization of
high-Level Waste Glasses USDOE Report PNL-2625, Pacific Northwest
Laboratory, Richland, Washington, June 1978.

65.  R. 0. Lokken, A Review of Radioactive Waste Immobilization in
Concrete, USDOE Report PNL-2654, Battelle Pacific Northwest
Laboratory, Richland, Washington, June 1978

-------
                                   30

IV.  LOW-LEVEL RADIOACTIVE WASTE SOLIDIFICATION TECHNIQUES

    Due to the increasing amount of waste being generated both in
volume and activity by the nuclear industry, considerable interest has
been shown during the last several years concerning various methods and
systems for the solidification of liquid and solid low-level
radioactive wastes from nuclear power plants; and, to a varying degree,
from other fuel cycle facilities.  The present U.S. method for the
disposal of low-level radioactive wastes is by burial in shallow
trenches dug in the earth's surface.

    There are many available varieties of solidification materials and
techniques.  Solidification agents include Portland cement, concrete,
plaster of paris, asphalt (or bitumen), polymers, and a blend of
absorbent material and cement or plaster (1-7).  Presently, most of the
solidification systems used in the United States utilize either cement
or an organic polymer, such as urea formaldehyde, as the basic
solidification matrix material.

    Table I lists the various solidification agents and the process
vendor or developer.

-------
                               31

                            Table I

              .. SOLIDIFICATION _AGENTS

     Vendor or Developer

Aerojet Energy Conversion Co.
Sacramento, California
ANEFCO, Inc.
White Plains, New York

ATCOR, Inc. (a)
Peekskill, New York

Brookhaven National Laboratory (b)
Upton, Long Island, New York

Chem-Nuclear Systems, Inc.
Bellevue, Washington

Delaware Custom Materials
Cleveland, Ohio

Dow Chemical Company
Midland, Michigan

Energy , Inc .
Idaho Falls, Idaho
Newport News Industrial Corp.
Newport News, Virginia

General Electric Company
San Jose, California

Hittman Nuclear & Development Corp
Columbia, Maryland
Los Alamos Scientific Laboratory
Los Alamos, New Mexico
AND VENDORS

  Solidification Agent

  Cement
  Urea Formaldehyde
  Bitumen

  Urea Formaldehyde


  Cement


  Cement with organic polymers
  Cement
  Urea Formaldehyde

  Cement with shale or silcates
  Organic Polymer
  Cement
  Urea Formaldehyde and Bitumen
  Cement
  Cement
  Urea Formaldehyde

  Cement

-------
                                   32
    Oak Ridge National Laboratory (b)
    Oak Ridge, Tennessee

    Protective Packaging, Inc.
    Jeffersontown, Kentucky

    Stock Equipment Company
    Cleveland, Ohio

    Todd Research and Technical Div.
    Galveston, Texas

    United Nuclear Industries
    Richland, Washington

    United Technologies
    Sunnyvale, California

    Werner and Pfleiderer Corp.
    Waldwick, New Jersey

    Westinghouse Electric Corp.
    Pittsburg, Pennsylvania
Cement with additives
Urea Formaldehyde
Cement
Urea Formaldehyde

Organic polymer
Cement with sodium silicate
Urea Formaldehyde

Organic polymer
Bitumen
Cement with vermiculite
a/ A division of Chem-Nuclear Systems, Inc.,
   See Nucleonics Week, August 31, 1978

JV Non-commercial applications

-------
                                   33

    Wastes from the nuclear fuel cycle and particularly 'from power
plants can be solid, liquid or gaseous with varying chemical, physical,
and radiological characteristics with many systems processing special
liquids which produce wet solids such as evaporator concentrates,
resins, etc.

    The radioactive liquid wastes from the power reactors, such as,
primary system blowdown, equipment drains, resin sluicing water,
evaporator condensates, decontamination solutions, demineralizer
regenerative solutions, laundry and laboratory wastes, evaporator
concentrates, spent ion exchange resins, filter precoat and cake
materials (powdex and solka-floc), cartidge filter units, and
diatomaceous earth are suitable for immobilization.  These wastes
contain the bulk of the volume and radioactivity of the solidified
wastes sent to the commercial low-level burial facilities.

    The processing of these wastes can be broken down into five steps:
(a) waste collection; (b) pretreatment; (c) the solidification process;
(d) mixing and packaging; and (e) final handling (Figure 7) (3,8).
Collection usually takes place in sumps or tanks; the contents are then
processed on batch or semi-continuous basis.  The solids pretreatment
operation consists of reducing the volume of the wet solids by using an
evaporator or other volume reduction device.  The solidification and
mixing steps involve the use of an agent, such as cement or an organic
polymer with additives or a catalyst, to produce an immobilized,
monolithic, inert matrix.  The container handling operations include
inspection to ascertain that solidification took place, capping the
container and adding appropriate shielding, decontamination, marking
and labeling, container testing, and storage awaiting shipment to a
burial facility.

1.0  Solidification Of Low-Level Radioactive Wastes In Bitumen

    The use of bitumen to solidify low-level radioactive wastes has
been successfully applied on an industrial scale for many years in
different countries (1-5, 9-19).  Bitumen or asphalt, a mixture of
high-molecular weight hydrocarbons, is a by-product residue from the
petroleum refining processes.  Various grades of bitumen are
commercially available with a wide range of physical properties.
Bitumen processes generally operate in the range of 150 to 230° C, at
which temperature water originally present in the waste potentially can
be volatilized.

-------
  WASTE
COLLECTION
PRETREATMENT
SOLIDIFICATION
    AGENT
MIXING/PACKAGING
    SYSTEMS
CONTAINER
 HANDLING
 SPENT RESIN
   TANKS
 FILTER SLUDGE
   TANKS
  LIQUIDS FOR
 EVAPORATION
MISCELLANEOUS
 LIQUID TANKS
                      DECANT
                     CENTRIFUGE
                     DEWATERING
   EVAPORATION
   CALCINATION
                     ABSORBENTS/
                       BLENDS
                      PARAFFIN/
                     POLYETHYLENE
                                         CEMENT
                                         ASPHALT/
                                         BITUMEN
                                         UREA
                                      FORMALDEHYDE
                                                            I
                    IN-CONTAINER
                     MIXING
                                                            ROLLER
                                                            MIXERS
                                                           TUMBLER
                                                            MIXERS
                                          PADDLE
                                          MIXERS
                                                            T
                                                           IN-LINE
                                                           MIXING
                                          DYNAMIC
                                          MIXER
                                           BATCH
                                           MIXER
                                                            SHIPPING
                                                           CONTAINERS
INSPECTION
MONITORING,
AND LABELING
i
1
STORAGE
i
r
TRUCK
LOADING
i
r
MONITORING
1
r
SHIPMENT TO
BURIAL SITE
                                                                                               co
                                                                                               .£»
                                     FIGURE 7
                LOW-LEVEL WASTE PROCESSING STEPS

-------
                                   35

    Basically, all of the bitumen processes consist of mixing the waste
solution, slurry, or solids with commercial emulsified asphalt or
molten base-asphalt and raising the temperature to evaporate the
fluid.  The contaminated solids remain intimately dispersed in the
asphalt and the product flows out of an evaporator into a receptacle.
The process has been successfully demonstrated in both continuous and
batch mixing operations.  Four different processes have been developed
for the bitumen-waste incorporation process:  (a) stirred evaporation;
(b) film evaporation; (c) the emulsified bitumen process; and (d) the
screw extrusion.  Several versions of the bitumen processes for
incorporating radioactive wastes for disposal have been utilized on an
industrial scale in Europe and on a laboratory development basis in the
U. S.

1.1  Stirred-Evaporator Batch Process

    In this process the bitumen is preheated and charged to an
evaporator with waste being added in with the bitumen.  Electric heat
is provided by immersion heaters and through the vessel wall.  An
agitator with adjustable blades is used for mixing and homogenizing the
waste-bitumen charge.  The charge is kept in the evaporator for several
hours, prior to discharging to a disposal container. Figure 8
illustrates this process.  The first plant scale bitumen process, using
the stirred evaporator method was started in 1964 with an evaporation
rate of 100 liters per hour at Mol, Belgium (I,1*).  The initial
operation was directed at the bitumen incorporation of radioactive
chemical sludges.  Subsequently, concentrated solutions, incinerator
ash, vermiculite, ion-exchange material and sand have been processed
using a slow-mixing device, particle separation, and resin
incineration.   Certain oxidizing salts, particularly nitrates, produce
an undesirable hardness in the final product; however, this difficulty
is overcome by use of reducing agents or different bitumen mixes.
Boric acid solutions can be incorporated into bitumen if the solutions
are first neutralized to prevent acid volatilization and water leaching
in the final product.

1.2  Emulsified Bitumen Process

    In this process, as shown in Figure 9, the waste material is
concentrated by filtration on a rotary filter then fed to a mixing
station where the bitumen and surface-active agents are added.  Here
the mixing and coating of the solid waste particles takes place with
separation of the water.  Upon removal of the water, the mix goes to a
drier.  The remaining water is reduced and the product discharged to
drums.

-------
                     36
CHEMICALS

 WASTE
           LL
BITUMEN
HEATER
                     AGITATOR
  TO
OFF-GAS
SYSTEM
                            STIRRED
                           EVAPORATOR
                      BITUMEN/WASTE PRODUCT
                         TO STORAGE DRUM
                 FIGURE 8
   STIRRED-EVAPORATOR BATCH PROCESS

-------
             FILTER
SLUDGE
                                     SURFACE-ACTIVE
                                        AGENTS
                                                               MIXING
                                                               STATION
              RELEASED
               WATER
                                      BITUMEN/WASTE PRODUCT
                                        TO STORAGE DRUM
                             FIGURE 9

               EMULSIFIED BITUMEN PROCESS

-------
                                   38

    This process was pioneered at the Marcoule Centre, France for
radioactive sludges and requires a much higher bitumen content.
Various surface-active agents (a surface-active agent is a soluble
compound that reduces the surface tension of liquids, or reduces
interfacial tension between two liquids or between a liquid and a
solid), variable reaction times, and different bitumen varieties are
used depending on the sludge to be treated.

    Experimental work on this process was also done at the Department
of Energy's Oak Ridge National Laboratory (ORNL), Tennessee (4, 10).
In this process the initial mixing of liquid waste and bitumen could be
effected readily at any convenient temperature below the boiling point
of the waste solution with the water and/or volatiles being removed by
heating.  The liquid wastes of special interest to ORNL were the
evaporator concentrates and solutions of sodium metaborate, nitrate and
nitrite.  No significant difficulities were experienced in the
incorporation of 60 weight per cent (w/o) of solids in bitumen from
evaporator concentrates; however, the best final products were produced
at solids contents of 45-50 w/o.  Boron compounds required higher
temperatures, neutralization, and low sodium to boron ratios,
especially with tetraborates which harden the bitumen, making stirring
impossible.

1.3  Film Evaporation Process

    This process has been developed both in the U.S. and in foreign
countries, with an industrial-size unit operating in France.  It
consists of a vertical tube  evaporator with a steam-heated jacket and
internal rotor that forms a thin film of the waste-bitumen mixture on
the heated surface.  The waste plus preheated bitumen is continuously
fed to the unit.  The water is evaporated and removed from the top,
while the bitumen-waste is discharged from the bottom.  The residence
time in the evaporator is only a few minutes.  Figure 10 illustrates
this process.  ORNL has also investigated the film evaporator process,
along with Marcoule, for incorporation of industrial, urban and
radioactive wastes in asphalt (10, 17).

-------
                      39
     CHEMICALS
WASTE
          MIXER
                       I i ' i,
                       Mill
                       JJLU
 VERTICAL TUBE
 TURBULENT-FILM
  EVAPORATOR
                              TO OFF-GAS SYSTEM
                  J
/
                          c
         BITUMEN
                          BITUMEN/WASTE PRODUCT
                            TO STORAGE DRUM
                   FIGURE 10
   TURBULENT-FILM EVAPORATION PROCESS

-------
                                   40

1.4  Screw Extrusion Process

    The screw extruder evaporator concept, as shown in Figure 11, was
developed at the Eurocheraic Plant at Mol, Begium and at the Karlsruhe
Nuclear Research Center in West Germany with good success at both
facilities (11, 16).  Industrial bituminization plants for
solidification of evaporator concentrates, utilizing the
screw-extruder-evaporator, have been sucessfully operated at an
evaporation capacity of 140 kg of water per hour with the final product
containing 50% salts (11-16).

    This process consists of two or four screw extruders operating in a
steel barrel that is divided into a number of stages heated
individually by steam.  Waste and preheated bitumen are charged
directly into the barrel.  While the screws mix the waste-bitumen
mixture and transport it forward, the water is evaporated and removed.
The product is discharged into drums and allowed to cool before capping.

    Werner & Pfleiderer Corporation (WPC) offers a waste solidification
system that includes a volume reduction system (3, 11, 14-16, 18, 19).
With asphalt as the solidification agent, WPC indicates the end product
is stable, particularly against leaching.  Although it has only been
recently introduced in North and South America, and Asia, it has been
operating routinely in Europe since 1965.  The WPC radwaste
solidification process yields a liquid-free (less than 0.5%} solid
using a continuous, fully automatic process with a multi-screw
compounding extruder.   The extruder-evaporator simultaneously provides
homogenous mixing (including reagent additives), liquid evaporation and
solidification in one machine.  The extruder evaporator normally
discharges the asphalt/salts mix into standard DOT 208 liter drums at a
rate from 1 to 114 liters per hour, depending on the size and speed of
the extruder and the concentration of the feed stream.  The entire
process, complete with interlocks, can be controlled remotely.

    The extruder-evaporator system can also evaporate the excess water
away from powdered resin sludge and bead resins in addition to
evaporator concentrates.  This system can also be used, to some extent,
for volume reduction and solidification of diatomaceous earth filter
sludge,  depending on the quantity.

    The WPC extruders, originally developed for the plastics industry,
are designed and built to operate a full year without maintenance.  The
operating record established in Eurpoe bears this out and is quite
impressive (over 134,00 hours operation at Marcoule without mechanical
failure).

-------
           'CHEMICALS
WASTE
BITUMEN-|
       1   /
       HEATER
                          SCREW-EXTRUDER
                                       TO OFF-GAS
                                        SYSTEM
BITUMEN/WASTE PRODUCT
  TO STORAGE DRUM
                      FIGURE 11

      SCREW EXTRUDER EVAPORATION PROCESS

-------
1.5  Other Research

    Additional research and operational work involving radioactive
wastes in bitumen has been done by other countries and nuclear
facilities; specifically, the USSR, Bulgaria, Japan, Hungary, and
Austria  (4, 13).  The Hungarian Mineral, Oil and Natural Gas Research
Institute developed a technique for the incorporation of
nitrate-containing wastes.  The technique utilizes an inert gas
introduced during the process which eliminates the danger of fires and
explosions; the gas also ensures adequate mixing which results in a
more satisfactory dispersion of the salts in the bitumen.  The Soviet
Union has a great deal of interest in the solidification of radioactive
concentrates by incorporation in bitumens; they have had a pilot plant
in operation since 1969 to evaluate the conditions and feasibility of
incorporatiang wastes enriched in sodium nitrate.  Also, both the
British  and Russians have developed methods for immobilizing
radioactive wastes attached to natural and synthetic sorbents such as
vermiculites, zeolites, clinoptilolite, and ion exchange resins, into a
bitumen  matrix.
1.6  Advantages and Disadvantages

    The main reported advantages and disadvantages of using bitumen  for
the insolubilization of radioactive wastes are as follows (3, 4, 9,  13,
18):

Advantages:

(a) The leach rate of the final product can be expected to be between
100 and 1000 times lower than cement mixes with a similar concentration
of radioactivity.

(b) There are a large number of types of bitumen with a wide variety of
properties; thus it is usually possible to obtain a suitable material
for any waste.

(c) Bitumen has good coating characteristics and good adhesion  to
incorporated material.

(d) The solubility of bitumen in water is negligible.

(e) Bitumen possesses a degree of plasticity and elasticity which  are
of benefit during the incorporation process.

-------
                                   43

(f) The volume reduction of bitumen compared to cement can be much
greater than 5 to 1, depending upon the percentage of radioactive salts
or percentage of water in the feed.

(g) Bitumen coats the inside surfaces of the drum, providing resistence
to internal corrosion.

(h) Attack by microorganisms on bitumen and bitumen compounds has been
regarded as generally insignificant.

Disadvantages

(a) There is always an inherent risk in working with organic material
at elevated temperatures; however, there is little evidence that
incorporation of the waste material into bitumen increases the risk of
fire or explosion.  (Tests indicate that the ignition temperature for
normal bitumen/salt ratios is above 350°C.)

(b) There is some evidence that the presence of nitrates or nitrites
and other oxidizing agents increases the risk of explosions or
combustibility of bitumen; however, in laboratory tests neither the
high thermal loads nor detonating  charges led to explosion-like
reactions.

(c) It is obvious that no substance should be added to bitumen which
decomposes at the working temperature.  Difficulties may be experienced
with certain plastics and compounds such as sodium citrate.

(d) Heating of bitumen mixes can result in the releases of oils, fumes
and mercaptans.

(e) It is necessary to work at high temperatures to obtain efficient
mixing.

(f) To obtain the best final product it is necessary to remove as much
as possible of the water present in the waste.  Leachability increases
significantly wth increasing amounts of retained water.

(g) Strict temperature control is  required in the bituminization
process.

-------
(h) Mixing tetraborates and iron and aluminium salts with bitumen
causes hardening to an extent which can interfere with discharge of the
final product from the equipment.

(i) Irradiation of bitumen modifies the chemical and physical
properties.  In some cases the effect is negligible and in other
considerable.  Irradiation up to an integrated dose of lO^O rad by an
external source or up to 10° rad due to the incorporation of
radionuclides can be accepted based on the type of bitumen chosen.

(j) Phase separation in bitumen mixes is likely to occur more readily
than in cement-waste products, particularly during accidental fires.

(k) Experiments at Marcoule and Karlsruhe have shown that swelling of
certain bitumen products can occur in water.

2.0  Solidification Of Low-Level Radioactive Wastes In Cement

    The cement solidification process, with and without additives, for
radioactive waste treatment has been in common practice on an
industrial scale at nuclear facilities for many years in different
countries (1-4, 8, 9, 12, 19, 20).

    Cement has been described as an adhesive substance, lime being the
principal constituent, capable of uniting fragments of solid matter
into a compact monolithic structure.  The most commonly used cement for
the incorporation of radioactive wastes is the "Portland" variety.  It
is obtained by intimately mixing silica-, alumina- and ferric
oxide-bearing materials to the lime and burning these materials to a
very hard brick and grinding the resulting brick.  There are various
types of portland cement depending on the fineness of the grinding and
on the addition of certain additives or the amount of various
constituents (22-24).  Portland cement Type I has been most commonly
used, but other Types, such as Type V which is resistant to sulfate
salt deterioration, have also been employed.

    The cement solidification processes basically consist of mixing the
cement with waste solution, slurry, or solids either within the
receptacle or just prior to being placed into the receptacle.  The
actual kinetic process leading to the curing of cement is not well

-------
                                   45

known.  However, the mechanism for the setting of cement, according to
the literature, has been postulated to include a reaction between water
and cement which forms solid particles and causes qrystallization of
the calcium hydroaluminate, hydroferrite, and hydrosilicate with the
crystals giving the strength to the hardened cement.  If the cement
product is to be in satisfactory condition for transportation or burial
it must have adequate compressive strength.  It is common practice to
neutralize the acid wastes before the cementation process and to
control the salt content.  Poorly cured cement will crack and spall,
causing more surface area to be exposed to leaching conditions (22, 23).

    The mixing of the cement with the various radwaste forms, i.e.,
sludges, resin beads, etc., affects the properties of the product.  The
strength of the cement will be a function of the total salt content in
the sludges,resins, etc., where there is a narrow range in the
acceptable values for the ratio of basic and acidic oxides in the final
product.  In this regard, waste to cement ratios recommended for proper
curing vary significantly among both foreign and U.S. users as shown in
Table II.  USSR studies have shown that ih order to produce cement of
acceptable structural strength the concentration of sodium nitrate salt
should perferably not exceed 130 g per kg of cement (4).

    The presence of water, nitrates, sulfates, borates, and other
unstable (in a radiation environment) compounds in the cement could
give rise to gaseous radiolytic products (12).  Gases also could result
from volatilization of compounds by elevated temperature in the
cement-waste mixture causing voids to form within the crystalline
structure.

    Some of the U.S. electric utility companies employ a combination of
vermiculite and cement to solidify their radwaste.  The expanded
vermiculite is porous, permitting the infiltration of dry cement into
the vermiculite structure.  This would act like a sponge absorbing the
liquid and giving a better final product than  when cement is used
alone.  Two cement solidification vendors, United Nuclear Industries
and Delaware Custom Materials, have developed a process which use
sodium silicate as an additive with Portland cement (3, 25, 26).
However, the addition of sodium silicate to cement-waste mixtures
increases the volume of waste per volume of solid formed.

-------
                                   46

    In the non-commercial area of using cement for solidification,
several of the Department of Energy Laboratories use a variety of
techniques.  Oak Ridge National Laboratory blends their
intermediate-level radioactive wastes with a dry mixture of
cementitious materials (Portland Cement-Type 1 and fly ash) and clays.
Then using a hydrofracturing technique, the resulting grout is pumped
into an isolated shale formation and allowed to set.  The clays are
used as an additive to retard the movement of radioactive nuclides.
The dry solids mix is usually blended with the liquid waste solution at
a solids to liquid ratio of between 0.7 and 1 Kg per liter.  The volume
ratio of grout to liquid waste is approximately 1.4 (27-30).

    Brookhaven National Laboratory found that the leachability
properties of cement could be improved by developing a polymer-
impregnated concrete (PIC) matrix (31-34).  PIC composites containing
tritiated aqueous waste, solid calcine, incinerator ash, aqueous and
solid sodium nitrates, reactor wastes, acidic and neutralized fuel
reprocessing wastes, and ion-exchange and sludge materials have been
produced.  The PIC process utilizes a soak impregnation technique to
fill the pores with a styrene monomer, which is then polymerized
in-situ by heating to 50-70° C.  In addition, the mechanical
properties of these cements are significantly improved as a result of
the styrene impregnation.  PIC also possesses improved radiation
stability and resistance to chemical attack.

    The Los Alamos Scientific Laboratory mixes transuranic sludge
materials and non-transuranic neutralized liquid wastes with cement for
storage and burial operations (4, 35, 36).

-------
                                    Table II
Establishment
France
CEMENTATION PRACTICES AT VARIOUS ESTABLISHMENTS

             Nature of Waste           Composition of Mixture
F. R. Germany
USA
             (a)  Evaporator
             concentrate
             (UOO g/liter)

             (b)  sludge
             Evaporator
              concentrate

             (a)  Evaporator
                  concentrate
                  (20% solids)

             (b) Neutralized
                  Concentrate
Czechoslovakia
                              (c)  Evaporator
                              Concentrates

                              (d)  Evaporator
                               Concentrates
             (e)  Ion Exchange
             Resins

             (a)  Sludge with solids
              content of 20 to
              25%
250 liters sludge
300 kg cement
40 kg vermiculite

83 kg sludge
55 kg cement
10 liters water

100 to 110 liters waste
150 to 200 kg cement

Vermiculite (2.7 M3)
 and Portland cement
 (0.68 m3)

(i) 75 liters of concen-
  trate, 128 kg cement,
  4 kg vermiculite

(ii) 20 to 35 liters
concentrate, 60 to
65 kg of cement

1 to 2 ratio waste
to cement

2 to 3 ratio waste to
cement with a sodium
silicate additive

1 to 1 ratio resin
slurry to cement

 35 liters of sludge
110 kg of cement

-------
                                   48

                             (b)  Evaporator concen-    10 kg of sludge
                               rates neutralized to     5 kg of evaporator
                               pH 6 to 8                concentrate
                               (200 g/liter)           22 kg of cement

USSR                         Evaporator concentrates   130 g of salt of the
                             (Max.  150 g/liter)         sodium nitrate-type
                                                       per kg of cement

-------
                                   49

    As shown in Figure 12, the cement and radwaste could be mixed
either within the shipping container or prior to loading the shipping
container.  Basically there are two mixing techniques used today:
in-drum mixers and external mixers (1-7).

2.1  In-Drum Mixer Process

    This technique is probably the simplest version.  There are two
methods:  (a) one is to add weights to a drum along with the cement and
waste slurry, then the drum is capped and transferred to a tumbling or
rolling station where the contents are throughly mixed; (b) another way
is to have an external agitator lowered into the drum to blend the
waste and cement, either after or during filling.  The mixing blade may
be a disposable type which is left in the container or may be a
removable type.  Figure 13 illustrates this process.

    Another system that can be characterized as an in-drum mixing
process has been developed by Stock Equipment Company (S-E-Co).  Since
transport of fresh cement has historically presented difficulties due
to the premature hardening and resultant incomplete curing of the
waste, S-E-Co. has developed a process to overcome this difficulty by
mixing in the final storage drum at the rate of 50 to 200 kg of waste
per hour.  S-E-Co. has concluded that the  quantity of cement and/or
additive in each 208 liter drum averages about 91 kg and the amount of
radwaste averages about 106 liters.  For this system, cap removal,
filling, cap replacement, and mixing is an automatic operation.

    The S-E-Co. Solid Radwaste System is designed and manufactured as a
completely integrated system utilizing components which are designed
specifically for the service expected rather than attemping to modify
standard equipment.  The S-E-Co. System is furnished complete for
placement into a building and interfacing wih liquid system piping,
utilities, etc.  The S-E-Co. design packages the solidified radwaste
into standard DOT, 208 liter durms.  The S-E-Co. waste system consists
of:  a cement storage hopper; a storage tank to hold liquid wastes that
contain concentrated solutions of dissolved solids; a decant tank for
filter media, resins, and/or the solid waste slurry from the storage
tanks; a drum processing unit which is fully automatic for uncapping
the drums, filling the drums with cement, filling from the decant tank,
reinsertion of the cap, and for the mixing/tumbling operation (2, 3,
37).

-------
                                     50
                                   SYSTEM FOR ENCAPSULATING
                                          RADWASTE
                                          CEMENT
               MIXES RADWASTE & CEMENT
             WITHIN THE SHIPPING CONTAINER
                                        MIXES RADWASTE & CEMENT PRIOR
                                        TO LOADING SHIPPING CONTAINER
ROLLER
MIXERS
TUMBLER
 MIXERS
PADDLE
MIXERS
  IN-LINE
MECHANICALLY
DRIVEN MIXER
BATCH
MIXER
                         REMOVEABLE
                        PADDLE MIXER
                          DISPOSABLE
                         PADDLE MIXER
   BASICALLY LIMITED TO
     SMALL SHIPPING
     CONTAINERS
   (eg.-55 GAL DRUMS)
              BASICALLY LIMITED TO
            RELATIVELY SMALL SHIPPING
             CONTAINERS WITH SPECIFIC
            CONFIGURATIONS & FEATURES.

                FIGURE  12
                        SMALL OR LARGE SHIPPING
                        CONTAINERS OF VARIOUS
                       CONFIGURATIONS & WEIGHTS
                       (eg. SHIELDED OR UNSHIELDED)
                            CAN BE USED.
                       FLOW DIAGRAM FOR
           CEMENT INCORPORATION  PROCESSES

-------
            DRY
            CEMENT
            HOPPER
                      . MIXING
                      
-------
                                   52

    The present General Electric Solid Radwaste Systems use cement as
the solidification agent with a disposable mixer and large disposal
containers  (2, 3, 38, 39).

2.2  External Mixer Process

    This process technique involves both continuous in-line dynamic and
static mixers and a batch mixing blade method (1-7).  In-line dynamic
mixing systems usually have small volume holdup times and utilize
either ribbon mixers, pug mills, or positive displacement pumps.  The
waste and the cement are charged into one end of the mixer and the
homogenous  mix is discharged into a container and capped or sealed.
The static  mixer consists of a section of pipe containing stationary
helical vanes which mixes the waste and cement as they flow through the
mixer.  The mixing blade system consists of introducing the liquid
waste and cement into a conically shaped batch mixer.  A mixing blade
then blends the constituents and the product is drained into a
container through the bottom of the mixer.  Figure 1U illustrates this
process.  For example, ATCOR Inc. performs all its mixing with an
in-line dynamic or mechanically driven mixer.  The ATCOR process system
mixes liquid and solid wastes with cement to produce a solidified
product within a disposable receptacle.  Basically, the system mixes
separate feeds of moist radioactive waste or evaporator concentrates
and dry cement in a small volume continuous mixer.  Solid waste
materials are preconditioned within the radioactive waste feed tank to
provide sufficient moisture when mixed with the dry cement to achieve
an acceptable cement mixture.  The system not only solidifies resins,
sludges and evaporator concentrates, but it can also be used to fix
spent filter cartridges within a solidified matrix.  In this case drums
or large volume liners containing spent filter cartridges could be
filled with a cement mixture that contains radioactive wastes.  The
cement waste mixture can be loaded directly into standard 208 liter
drums or larger receptacles.  Where waste is to be packaged in drums,
drum capping and decontamination can also be provided.  There is no
preparation of drums required prior to filling.

    The United Nuclear Industries (UNI) and Delaware Custom Material
(DCM) systems for solidifying radwaste use an in-line batch mixer for
waste and cement which is then mixed with sodium silicate in the
shipping container.  The UNI and DCM systems provide:  (a) proportional
pumps for metering waste feed; (b) in-line mixer to assure homogeneity;
(c) single  fill port for wastes; and  (d) in-container solidification
(3, 25,  26,  HO,

-------
    SLURRY
    WASTE


      OR


LIQUID WASTE
       MIX TANK
                      DEWATERING
                        TANK
                                        CEMENT
                                         SILO
  SODIUM
SILICATE TANK
                                ±1
                                     FEEDER
                              MIXING PUMP

                                       rt—'
                                           CEMENT/WASTE PRODUCT TO
                                           STORAGE DRUM
                          FIGURE 14
         EXAMPLE OF EXTERNAL MIXING PROCESS
                                                                 Ui
                                                                 OJ

-------
                                   5'4

2.3  Other Processes

    There are other variations of the two systems such as the
Westinghouse Waste Encapsulation System which is basically a vacuum
packaging process in which spent radioactive resins and waste
evaporator bottoms are encapsulated using a cement vermiculite mixture
in standard DOT 17 H drums (42).  Also, the Delaware Custom Materials
company utilizes the Chemfix, Inc., process which offers a complete
service of equipment and chemicals for solidification of radioactive
wastes, including a variety of inorganic and organic sludges.  The
process uses a combination of cements, shales and silicates as the
solidification agent (26).

2.4  Advantanges and Disadvantages

    The main reported advantages and disadvantages of using cement for
insolubilization are as follows (1-9, 12):

    Advantages

    (a)  No complex equipment; it is often possible to carry out the
    incorporation in the disposal receptacle.

    (b)  Low capital investment and low running costs; power
    requirements minimal.

    (c)  No applied heat required; low operating temperature means no
    fire risk and eliminates difficulties with off-gas purification.

    (d)  Most systems fully automatic; and therefore, operators can be
    trained easily.

    (e)  Waste-cement mixes are not grossly affected by pH.

    (f)  Cement is relatively cheap, but this is often off-set by the
    greater quantity required.

    (g)  Chemical and physical properties of cement are well known.

    (h)  Cement imparts good shielding properties.

-------
                               55

(i)  Natural alkalinity of cement is useful in helping to
neutralize acidity in waste solutions.

(j)  Little reported trouble with phase separation in the mix.

(k)  Water is required for setting the mix so there is no need for
extensive dewatering provided a satisfactory water/cement ratio is
maintained.

(1)  Presence of nitrates and nitrites and other oxidizing agents
do not have the same detrimental effects as they can have when
mixed with an organic material such as bitumen.

(m)  Less subject to irradiation damage than bitumen.

Disadvantages

(a)  The concentration of certain salts, such as borates, may cause
the cement-waste matrix some difficulty in curing, and may cause
deterioration and leaching over time at an abnormally high rate.
However, progress is being made to solve this problem by using
selected types of cement.

(b)  The weight and volume of the final product will normally be as
much as twice that for other corresponding solidification
processes.  The weight and volume increase is mainly due to the
amount of cement which must be added to react with the residual
water in the waste.  (In ocean disposal, this is an advantage which
overcomes the effect of buoyancy on the container volume.)

(c)  If mixing equipment experiences operational trouble and break
downs, this could require frequent cleaning of the equipment,
particularly the blades.

(d)  Nonautomated systems require several manual operations during
the solidification process.

(e)  Most studies have shown that when buried, and after the
container rusts away, the cement will leach if in contact with
ground water.

-------
                                    5'6

 3.0  Polymeric  Solidification  Processes  for  Low-Level  Radioactive Wastes

     Incorporation  of  radioactive  wastes  into polymeric fixation  agents
 is a relatively new solidification  process when  compared  to
 incorporation in cement  or  bitumen.  The solidification process  can
 take place  either  at  ambient temperatures or with hot  evaporator
 concentrates (up to 60°C).  Presently, several U.S.  companies  sell
 urea-formaldehyde  (UP) solidification systems (as shown in Table I).
 All  the  organic processes are  essentially batch  processes where  a
 catalyst is generally mixed with  the wastes  and  polymer either in a
 premixer vessel or in the receptacle itself.  The polymeric  processes
 do not really solidify the  wastes;  the long  chain molecules  of the
 organic  polymer are linked  together to form  a multi-voided sponge that
 "traps"  the waste.  (1-3, 12)

     Paraffin and polyethylene  based solidification  agents can  also be
 used to  solidify wastes.  These agents must  be liquified  by  heating
 prior to mixing with  the wastes (1, 2, 17, 43, 44).

     The  only industrial  experience  with  polymeric solidification
 systems  to  date has been with  the UF process.  The  process description,
 advantages  and  disadvantages are  based on systems using UF,  even though
 there are other organic  polymer processes, such  as  the Dow Chemical  and
 the  Todd Research  processes (45,  46), which  are  either not operational
 or have  not been in operation  long  enough to provide operational
 information comparisons.

 3.1  Urea Formaldehyde

     The  physical method  of  organic  polymeric mixing depends  upon the
 type of  solidification agent and  receptacle  used.   In  general, there
 are  three types of UF mixing:

     1.   In-container  disposable paddle mixer.
     2.   In-line static mixer.
     3.   In-line mechanically driven mixer.

     The  in-container  mixer  is  generally  used for mixing resin  beads
with UF.   The in-line dynamic  mixer is used  by UNI  to  mix liquid waste
and UF prior to discharging into  a  receptacle.   UF  systems generally
employ an in-line  static mixer which contains stationary  helical vanes
to mix the  fluids  as  they flow through the mixer.   Just as the mixed
polymer and waste  are injected into a container, the acidic  catalyst is
added to  initiate  solidification.   Figure 15 is  a system  diagram for UF
indicating  the  incorporation steps.  Figure  16 is a flow  diagram of the
static mixer technique.

-------
                  57
                       UF
MIXES RADWASTE, UF & CATALYST
  PRIOR TO LOADING SHIPPING
       CONTAINER
                MIXES RADWASTE, UF & CATALYST
                WITHIN THE SHIPPING CONTAINER
IN-LINE STATIC
  MIXERS
   IN-LINE
MECHANICALLY
DRIVEN MIXER
 DISPOSABLE
PADDLE MIXER
   SMALL OR LARGE SHIPPING
   CONTAINERS OF VARIOUS
  CONFIGURATIONS & WEIGHTS
  (eg. SHIELDED OR UNSHIELDED)
       CAN BE USED.
                  FIGURE 15
           FLOW DIAGRAM  FOR
    UF INCORPORATION PROCESSES

-------
                            UREA FORMALDEHYDE
                                  TANK
  SLURRY OR
  LIQUID WASTE
DEW ATE RING
TANK
          1
                          , , STATIC MIXER
                                       I
CATALYST TANK
                                                               00
                                           U-F/WASTE
                                           PRODUCT TO
                                           STORAGE DRUM
                          FIGURE 16

     UREA FORMALDEHYDE INCORPORATION PROCESS -
                     EXTERNAL MIXING

-------
                                   59

    Protective Packaging, Inc.  (PPI), a wholly owned subsidiary of
Nuclear Engineering Co. , developed and was the first company to design
and sell, a system using a chemical solidification agent other then
cement.  Since then, they have  filed several patent applications on  the
system and trademarked  the name "TIGER-LOCK"R.  Their patent
applications cover both the use of the liquid solidification agent
(TIGER-LOCKR, a type of urea  formaldehyde polymer), and also all the
related hardware that makes up  a TIGER-LOCKR Radwaste Solidification
System.  This includes  the process equipment, control panel, power
panel, and associated material  handling equipment.  The key aspects  of
the PPI design are:  (a) three  separate tanks, pumps, different size
liners for radwaste, TIGER-LOCKR, and catalyst respectively; (b) a
premixer for the radwaste and UF to homogenize the slurry prior to
contact with the catalyst; (c)  a manual decoupling device to seal the
liners that contain the cured waste; and (d) an automatic level
detector to indicate filling  to 90% volume (1-3, *»7,
    TIGER LOCKR is a proprietary augmented agent of urea formaldehyde
resin that is manufactured to strict physical and chemical
specifications.  The PPI system currently being sold includes TIGER
LOCKR and catalyst (usually  sodium bisulphate-NaHSCty) , associated
processing equipment, and container liners for use in the
transportation and burial of the solidified radwaste.

    PPI suggests that the desired ratio of TIGER LOCKR to radwaste is
2:1 by volume.  For this system the operator has the task of estimating
the correct amount of catalyst for solidification which is highly
dependent upon the quantity  and type of radwaste that would be
solidified by the PPI system.

    Gel time of the product  can be adjusted from minutes to hours by
the catalyst concentration (normally about 1 to 3% by volume) .  If UF
is used after its shelf life has been exceeded or at low temperatures
or low viscosity, the "cottage cheese" effect will occur, i.e., little
solidifying and essentially  a settling of materials of different
density within the container.

    ANEFCO, Inc., offers a waste solidification system using urea
formaldehyde and sulfuric acid, or an equivalent catalyst, as a
solidifying agent.  Their process system uses a 3785 liter batch tank,
a static mixer and a disposable polyethylene liner in the disposal
container (2, 3).
R-Registered Trademark, Protective Packaging, Inc., Jeffersontown,
Kentucky

-------
                                   60

    Hittman Nuclear & Development Corporation offers radioactive waste
solidification systems using cement or a polymer such as urea
formaldehyde as the solidifying agent.  Chemical additives are used
with both agents to enhance the efficiency, i.e., volume of waste per
unit volume of solidified product.  The disposable containers used to
package radwaste vary in size from a standard 208 liter drum up to
5.7 UP capacity (2, 3, 21, 49).

    The United Nuclear Industries also offers a radwaste solidification
system utilizing urea formaldehyde (UF) as the solidification agent
(1-3, 25, 40, 41).   The use of the UF material permits the use of
in-line static mixers with no moving parts.  Solidification of the
waste - UF mixture is accomplished using either a sodium bisulfate
catalyst (pH range of 3 to 7.5) or a phosphoric acid  catalyst (pH
range of 3 to 10).

    Chem-Nuclear Systems Inc., offers either portions of or a complete
waste system design, component selection, procurement, fabrication,
construction, installation and operation of solidification systems
using either cement or urea formaldehyde as the solidification agent.
Chem-Nuclear also has available a mobile solidification unit using the
UF system (2, 3, 50).

    Energy Incorporated and Newport News Industrial Corporation have
jointly developed a radioactive waste reduction (designated RWR-l™)
system, to convert low and medium level liquid and solid combustible
radioactive wastes to solids by a fluidized bed calcining and
incineration process.  The RWR-1™ system reduces the volume of both
liquid and solid radwastes, e.g., concentrated chemical wastes, filter
sludges, spent resin beads, rags and other similar materials, and
produces a granular, anhydrous solid which may be placed directly in
burial containers or incorporated into matrices such as concrete, urea
formaldehyde or bitumen for burial (2, 51).  The RWR-1™ system has
indicated volume reduction factors of 5-80:1, depending on the
material; and the main components are designed for remote operation.
The operating parameters include process temperatures of 400 C for the
calciner and 1000 C for the incinerator, with waste feed rates of 45-90
Kg per hour and 132 liters per hour.

    The Aerojet Energy Conversion Company has marketed a VR-20
Radioactive Waste Management System which reduces the volume and
encapsulates the waste.  The volume reduction is achieved by conversion
of all liquid wastes into anhydrous calcined solids and drying of
dewatered spent resins and sludges.  These solid wastes can  then be

-------
                                   61

encapsulated in cement, UF or bitumen as the solidification agent for
subsequent shipment and burial.  Using the VR-20 process, the volume of
liquid waste could be reduced by a factor of 10-20., while the volume of
liquid spent resins and sludges could be reduced by a factor of 2-4
when compared with other solidification methods.

    The main feature of this system is a fluidized bed calciner which
receives the radioactive liquid waste feed containing the dissolved
chemical solids and processes these aqueous solutions into free-flowing
anhydrous particles.  Concentrated radioactive liquid waste (evaporator
bottoms, etc.) is pumped from the concentrated liquid waste storage
tank to a heated fluidized bed calciner concentrator.  The volatiles
exit with the water vapor at the top of the fluidized bed concentrator,
leaving behind the dissolved solids.  The granular solid produced can
then be encapsulated (2, 3, 52-56).

3.2  Advantages and Disadvantages

    The main reported advantages and disadvantages of using
urea-formaldehyde for insolubilization are as follows (1-3, 9, 12).

    Advantages

    (a)  The amount of waste capable of incorporation in a receptacle
    wih UF is about 30% by volume more than with cement.

    (b)  For shipping not requiring radiation shields, shipping cost
    with UF or polymerics is less than for cement and bitumen due to
    the ability to put more waste in a given container and a lower
    density for the mixture.

    (c)  Mixture of UF and radwaste are not combustible.  Further, no
    detectable exothermic reaction occurs with UF.

    Disadvantages

    (a)  For shipments requiring radiation shields, UF or polymeric
    solidified materials, due to its lower bulk density and higher
    activity, requires more shielding than materials solidified with
    cement.

    (b)  It appears that routine attainment of the complete elimination
    of free standing water is a problem with encapsulated UF radwaste,
    particularly those having lower concentrations of the polymer
    (volume ratios of 1 to 3 or less).

-------
                                   62

     (c)  According to the utility operators, it is difficult to work
     with UF because of the relatively low viscosity of the mixture,
     which permits settling or floating  (segregation) of materials of
     different  densities.

     (d)  Solidification time is affected by both the pH of the mixture,
     which is regulated by the amount of catalyst, and the temperature
     of the mixture.

     (e)  The UF  shelf life is limited and is dependent upon storage
     conditions.

     (f)  Equipment must be designed to  eliminate fume problems with UF;
     the odor is  disagreeable even in small concentrations.

     (g)  Some  manufacturers of UF have  stated that this product, is
     biodegradable; also the catalyst is corrosive to most metals.

     (h)  During  the solidification process when the UF-radwaste mixture
     is exposed to air, water evaporates from the mass, but if the
     matrix remains in an air-tight container, the mixture will remain
     semi-liquid.

3.3  Other Polymeric Processes

     Polyethylene agents are not used commercially in the U.S.
Polyethylene is a superior solidifying  agent for most organic liquid
wastes.  The waste is combined with molten polyethylene inside a heated
chamber in which the water and other volatile constituents are
evaporated.  The mixed and dehydrated liquid product is discharged  to  a
container where it solidifies upon cooling.  The final product is a
solid  plastic  block, which is relatively inert at room temperature  and
is insoluble in water.  It has good freeze-thaw characteristics and a
storage life of several years.  Polyethylene is completely combustible
and  can be incinerated.  It is flammable with a flash point of 250  C
(1,  2, 17, 43, W).

     United Technologies-Chemical Systems Division offers the  Inert
Carrier Process which handles ingredients in an inert organic liquid.
The  operating  concept is based on dispersal of the reactants  in an
inert  carrier  to provide maximum surface area for solid-liquid reaction
mechanism.   In addition, the process provides for a clean  separation  of
the  reaction product from the inert carrier.  The waste materials are

-------
                                   63

low viscosity dispersions in an inert carrier.  The inert carrier is a
fluid selected so that neither the starting materials nor the products
are soluble in it or chemically reactive with it.  The process has
particular advantages in operations which require (a) preparation of
compositions which are too viscous to mix by ordinary methods; (b)
extremely intimate mixing of solids with small quantities of liquids;
(c) safe control of highly exothermic chemical reactions; or (d) a
closed system and/or remote controlled processing of hazardous, toxic,
or explosive materials (57).

    Dow Chemical Company has developed a radwaste solidification system
using organic polymers that produces a radwaste product free of liquid,
reasonably hard, and free standing.  The solidification system is
usable for all anticipated chemical decontamination solvents and
regular wastes from nuclear power stations.  To simulate disposal
conditions, Dow evaluated the solidification product for the following:
(a) compressive strength; (b) temperature cycling; (c) radiation
stability; (d) leachability; (e)~ impact testing; (f) heat exposure; and
(g) free liquid.

    To date, Dow has solidified the following simulated wastes in the
laboratory and in 208 liter drums, containing, no detectable free
liquid:  (a) spent decontamination solutions at pH's of 3 to 5 and 9 to
10 with HQ% solids; (b) filter aid and slurries, 90/10 by volume; (c)
ion exchange resins, 90/10 by volume; (d) PWR evaporator bottoms with a
pH of 2.5 and 7% solids; and (e) BWR evaporator bottoms with a pH of
10.6 and 6% solids.  After casting, the drums solidified within one
hour.  The radwaste to solidification agent ratio is a least 1.25 to 1
and as high as 2.5 to 1.  A field demonstration was carried out by
successfully solidifying 3^00 liters of radioactive decontamination
solvent at a nuclear power plant (45).

    Todd Research and Technical Division is marketing a solidification
agent called SAFE-T-SETR, which is a long chain linkage organic
polymer.  The agent can be used with concentrated low-level liquid
radioactive wastes from filtration, precipitation, ion-exchange or
evaporation.  The set-up time varies from one minute to several hours
depending on the amount used in proportion to the volume of waste.
One-half kilogram of SAFE-T-SETR will solidify 3.8 liters of liquid
material.  The agent can be tailored to any particular system or
circumstances including pumping the waste and SAFE-T-SET" mixture and
can be adapted to molds of any type.  The solidified matrix remains
stable under conditions of freezing, high temperature and leaching (M6),
R-Registered Trademark, Todd Research and Technical Division,
Galveston, Texas.

-------
                                   64

4.0  Use of Absorbents for Solidification of Low-Level Radioactive
Wastes

    Absorbents are used to eliminate free standing liquids by virtue of
their ability to hold water molecules within their pores.  The
absorbent is, however, not chemically bound to the waste, nor does it
represent a free standing monolithic solid; therefore, the absorbents
should not be considered as solidification agents.  Further, they do
not provide or enhance resistance to leaching, if water comes in
contact with the absorbed radioactive materials.  The absorbents are
stored in a dry environment and are placed in the shipping container
prior to adding radioactive liquids.  Some commonly used absorbents are
vermiculite, clays, silica gel, plaster of paris, microcell and/or
diatomaceous earth filter aid.

    Vermiculite, dehydrated clay granules, and diatomite absorbents
have been routinely used for liquid wastes, with perhaps vermiculite
the most widely used.  The absorbent method, when properly applied will
physically entrap the waste liquid so that no appreciable free liquid
will leak out if the container is breached.  However, in most cases,
the liquids can be displaced readily by the addition of water.

    In preparation, the receptable is filled with vermiculite and
liquid waste equivalent to about 1/3 to 1/2 of the volume.  For some
materials such as the diatomaceous earth, physical mixing of the liquid
and absorbent may be necessary.  Care must be taken with all absorbent
to avoid supersaturation (1, 2, 12).

5.0  Cited References

    1.  Alternatives for Managing Wastes from Reactors and Post Fission
    Operations in the LWR Fuel Cycle, Volume 2:  Alternatives for Waste
    Treatment, Report No. ERDA-76-43, U. S. Energy Research and
    Development Administration, Washington, D. C.  May 1976.

    2.  W.  F. Holcomb and S. M. Goldberg, Available Methods of
    Solidification for Low-Level Radioactive Wastes in the United
    States,  Report No. ORP/TAD-76-4, U. S. Environmental Protection
    Agency,  Office of Radiation Programs, Washington, D. C., December
    1976.

-------
                               65

3-  Radioactive Waste Management for Nuclear Power Reactors, UCLA
Extension Course, Engineering 821.7, Los Angeles, California,
October 20-23, 1975.

4.  R. H. Burns, "Solidification of Low and Intermediate Level
Wastes", Atomic Energy Review, 9 (3), 1971.

5.  Treatment of Low and Intermediate-Level Radioactive Waste
Concentrates, Technical Report Series No. 82, STI/DOC/10 82,
International Atomic Energy Agency, Vienna, Austria 1968.

6-  Management of Low and Intermediate-Level Radioactive,
STI/PUB/264, International Atomic Energy Agency, Vienna, 1970.

7.  Management of Radioactive Wastes at Nuclear Power Stations ,
Safety Series No. 28, International Atomic Energy Agency, Vienna
1968.

8.  L. H. Barrett,  "Solid Waste Treatment at Nuclear Stations",
presented at 1975 Annual Conference, Southeastern Electric
Exchange, April 17-18, 1975.

9.  P. Columbo and  R. M. Neilson, Jr., Critical Review of the
Properties of Solidified Radioactive Waste Packages Generated at
Nuclear Power Reactors. USNRC Report BNL-NUREG-50591, Brookhaven
National Laboratory, Upton, New York, December 1976.

10.  R. E. Blanco,  H. W. Godbee, and E. J. Frederick, "Radioactive
Wastes...Incorporating Industrial Wastes in Insoluble Media,
"Chemical Engineering Progress, 66(2), pages 50-56, February 1970.

11.  W. Hild, W. Kluger, and H. Krause, "Bituminization of
Radioactive Wastes  at the Nuclear Research Center," Transactions of
ANS, Vol. 19, 1971*  Winter meeting of the American Nuclear Society,
Washington, D. C.,  October 27-31, 1974.

12.  J. P. Duckworth, M. J. Jump, and B. E. Knight, Low-Level
Radioactive Waste Management Research Project Final Report. Nuclear
Fuel Services, Inc., West Valley, New York, September 15, 1974.

13.  Bituminization of Radioactive Wastes, Technical Reports Series
No. 116, International Atomic Energy Agency, Vienna, 1970.

-------
                               66

 14.   J.  E.  Stewart, and  R. Herter,  "Solid Radwaste Experience  in
 Europe Using  Asphalt," ASME-IEEE Joint Power Generation Conference,
 Portland, Oregon,  September 28-October 1, 1974.

 15.   Radioactive Residues, Their Origin and Elimination,
 KS-Information Brief  Report No. 20, Werner and Pfleiderer
 Engineers,  Stuttgart, Federal Republic of Germany, 1975.

 16.   G.  Meier, and W. Bahr, The Incorporation of Radioactive Wastes
 into  Bitumen, Part 1;  The Bituminization Plant for Radioactive
 Evaporator  Concentrates  at the Karlsruhe National Research  Center,
 Report KFK-2104, Karlsruhe, Federal Republic of Germany, April 1975.

 17.   G.  L.  Fitzgerald, H. W. Godbee, R. E. Blanco, and W. Davis,
 Jr.,  "The Feasibility of Incorporating Radioactive Wastes in
 Asphalt  or  Polyethylene,'! Nucl. Appl. Technol. , Vol 9, p. 821, 1970.

 18.   Topical  Report;  Radwaste Volume Reduction and Solidification
 System,  Report No. WPC-VRS-001, Werner & Pfleiderer Corporation,
 Waldwick, N.  J., November 1976 (Revision 1-May 1978).

 19.   J.  E.  Stewart, "European Experience with Asphalt Packaging of
 Radioactive Wastes," presented at the Fifth International Symposium
 on Packaging  and Transportation of Radioactive Materials, Las
 Vegas, Nevada, May 7-12, 1978.

 20.   R.  J.  Stouky, "Operating Costs Nuances of Nuclear Power Plant
 Radioactive Waste  Disposal," presented at The Symposium on  the
 Management  of Low-Level  Radioactive Waste, Atlanta, Georgia, May
 23-27, 1977.

 21.   R. B.  Wilson, "Low-Level Radwaste Packaging - Why Not  Cement,"
 presented at Fifth International Symposium Packaging and
 Transportation of  Radioactive Materials, Las Vegas, Nevada, May
 7-12,  1978.

 22.  W. F.  Holcomb, Causes of Concrete Cracking and Methods of
 Control,  USAEC Report NVO-38-26, Fenix and Scisson, Inc., Las
 Vegas, Nevada, July 1970.

 23.  Concrete Manual, 7th Edition, U. S. Department of Interior,
Bureau of Reclamation, U. S. Government Printing Office,
Washington,  D. C., 1966.

-------
                               67

24.  S. Brunauer and L. E. Copeland, "The Chemistry of Concrete,"
Soientific American, p 80, April 1964.

25.  W. H. Heacock, "Alternative Nuclear Waste Solidification
Processes," Waste Management 75 Symposium, R. G. Post (Editor),
University of Arizona, Tucson, Arizona, March 1975.

26.  John Hays, Personal Communication, Delaware Custom Material,
State College, Pennsylvania 1976.

27.  Final Environmental Impact Statement, Management of
Intermediate Level Radioactive Waste, Oak Ridge National
Laboratory, Oak Ridge, Tennessee, Report ERDA-1553, 0. S. Energy
Research and Development Administration, Washington, D. C.,
September 1977.

28.  Environmental Statement, Radioactive Waste Facilities, Oak
Ridge National Laboratory, Oak Ridge, Tennessee, Report WASH-1532,
U. S. Atomic Energy Commission, Washington, D. C., August 1974.

29.  W. de Laguna, "Radioactive Waste Disposal By Hydraulic
Fracturing", Industrial Water Engineering, page 32, October 1970.

30.  J. G. Moore and E. W. McDaniel, "Fixation of Intermediate -
Level Radioactive Waste in Hydrofracture Grout", presented at the
80th Annual Meeting of the American Ceramic Society, Detroit,
Michigan, May 6-11, 1978.

31.  P. Colombo and R. M. Neilson, Jr., "Some Techniques  for the
Solidification of Radioactive Wastes in Concrete", Nuclear
Technology, Vol. 32, p. 30, January 1977.

32.  Development of Durable Long-Term Radioactive Waste Composite
Materials, USAEC Progress Reports No. 1-10, Brookhaven National
Laboratory, Upton, N.Y., July 1972-April 1975.

33.  L. E. Kukacka, "Production Methods and Applications  For
Concrete Polymer Materials," Presented at the 70th Annual Meeting
of the American Institute of Chemical Engineers, New York, November
13-17, 1977.

34.  M. Steinberg, et al, "Concrete-Polymer Materials Development,"
Nuclear News, p. 48, September 1970.

-------
                               68

 35.  Draft Environmental Impact Statement, Las Alamos, Scientific
 Laboratory Site. Las Alamos. New Mexico. Report DOE-EIS-0018 D, U.
 S. Department of Energy, Washington, D. C., May 1978.

 36.  R. E. McLin, et al, "Experience with  Cement Fixation of
 Nuclear Waste at LASL," presented at the 80th Annual Meeting of the
 American Ceramic Society, Detroit, Michigan, May 6-11, 1978.

 37.  J. Stock, "A Radwaste Disposal System Type VI", Stock
 Equipment Company, Cleveland, Ohio, May 1972.
                                                              /
 38.  J. M. Smith, and J. E. Kjemtrup, "BWR Development in Nuclear
 Plant Effluent Management," presented at American Power Conference,
 Illinois Institute of Technology, Chicago, Illinois, April 18-20,
 1972.

 39.  H. L. Loy and D. C. Saxena, "Processing and Packaging of Solid
 Wastes From BWR's," Proceedings 3rd International Symposium,
 Packaging and Transportation of Radioactive Materials, Richland,
 Washington, August 16-20, 1971, Report CONF 710801  (Vol 1), U. S.
 Atomic Energy Commission.

 40.  Summary of Qualifications for Furnishing of Nuclear Power
 Plant Radwaste Solidifixation Systems, Douglas United Nuclear,
 Inc., Richland Washington, January 1972

 41.  Radwaste Solidification Septems, United Nuclear Industries,
 Inc., 1975

 42.  Personal Communication, Westinghouse  Electric  Corporation
 (PWR  Systems Division) Pittsburg, Pennsylvania, 1976.

 43.  R. V. Subramanian, and N. Raff, "Polymeric Immobilization of
 Low-Level Radioactive Wastes," presented at American Institute of
 Chemical Engineers 80th National Meeting,  Boston, Mass., September
 7-10, 1975.

44.  R. V. Subramanian and R. Mahalingam,  Immobilization of
Hazardous Residuals by Encapsulation, Annual Letter Technical
Report to National Science Foundation for  Grant AEN-75-06583,
Washington State University, Pullman, Washington, April 30, 1976.

-------
                               69

45.  H. E. Filter, The Dow System for Solidification of Low-Level
Radioactive Waste from Nuclear Power Plants, The Dow Chemical
Company, Midland, Michigan, October 1976.

46.  The Use of "SAFE-T-SET" As a Radioactive Liquid Waste
Solidification Medium, Todd Shipyards Corporation, Galveston,
Texas, May 1967.

47.  K. A. Gablin, and J. H. Leonard, "Leachability Evaluation of
Radwastes Solidified with Various Agents," The American Society of
Mechanical Engineers, United Engineering Center, New York, N.Y.,
74-WA/NE-8, 1974.

48.  S. K. Cosgrove, K. M. Emmerich, and J. H. Leonard, Interim
Report on Evaluation of Solidification Techniques for Low-Level
Nuclear Waste Materials, J. H. Leonard Associates, Cincinnati,
Ohio, August 1974.

49.  P. T. Tuite, S. R. Zimmerman, and G. K. Bolat, "A System for
Solidification and Packaging of Radioactive Waste at a PWR,"
Presented at the American Power Conference, Illinois Institute of
Technology, Chicago, Illinois 1972.

50.  C. D. Johnson, "Modern Radwaste System: An Overview," Chemical
Engineering Progress, Vol. 72, No. 3, page 43, March 1976.

51.   (a) Energy Incorporated Sales Brochure, "RWR-1:  Radioactive
Waste Reduction," Energy Inc., Idaho Falls, Idaho, 1975;
(b) Newport News Industrial Corporation Sales Brochure,"Radwaste
Volume Reduction, RWR-1™," Newport News Industrial Corp.,
Newport News, Virginia, 1976;
(c) News Release Technical Bulletin, "RWR-1™," Newport News
Industrial Corp., Newport News, Virginia, October 1, 1976.

52.  L. E. White, and R. Garcia, "Use and Economic Advantages of
Fluid Bed Calciners for Volume Reduction of Liquid Radwaste," The
American Society of Mechanical Engineers for presentation at the
Winter Annual Meeting, New York, New York, November 17-22, 1974.

53.  L. E. White and R. Garcia, "Environmental Survey of
Transportation of Radioactive Wastes to the Burial Site," Fourth
National Symposium on Radioecology, Oregon State University, May
12-14, 1975.

-------
                                   70

    54.  L.  E. White and R.  Garcia, "Environmental Impact of
    Radioactive Waste Solidification Process on Burial Site
    Operations," Fourth National Symposium on Radioecology, Oregon
    State University, May 12-14, 1975.

    55.  L.  E. Tokerud and R.  Garcia, "Iodine Decontamination Factor
    for Liquid Radioactive Waste Volume Reduction System," Trans.  Am.
    Nucl. Soc.. 23, 264, 1976.

    56  Aerojet Energy Conversion Company, Topical Report;  Fluid Bed
    Dryer, Report No. AECC-l-A,  Sacramento,  California,  February 21,
    1975.

    57.  Randall D. Sheeline,  Personal Communication,  United
    Technologies, Sunnyvale, California, September 1976.

6.0  Additional References

    1.  S. E. Pihlajavaara and J-P. Aittola, "Radwaste Concrete:
    Solidification of Nuclear Wastes with Portland Cement," presented
    at the 80th Annual Meeting of the American Ceramic Society,  May
    6-11, 1978, Detroit, Michigan.

    2.  H. E. Flora, "Radioactive Waste Processing for Nuclear Power
    Plants," presented at the llth Annual Liberty Bell Corrosion
    Course,  September 13, 1973,  Philadelphia, Pennsylvania.

    3.  Draft Report:  The Conditioning of Residues from the Treatment
    of Low-  and Intermediate-Level Radioactive Wastes and Criteria for
    Their Storage or Disposal on Land,  International Atomic Energy
    Agency,  Moscow, USSR, December 9-13, 1974.

    4.  T. B. Mullarkey et al (NUS Corporation), A Survey and
    Evaluation of Handling and Disposing of Solid Low-Level Nuclear
    Fuel Cycle Wastes, Report AIF/NESP-008,  Atomic Industiral Forum,
    Inc., Washington, D. C., October 1976.

    5.  P. Colombo and R. M. Neilson, Jr.,  Properties of Radioactive
    Wastes and Waste Containers Progress Reports:

         BNL-NUREG-50571                April-June 1976
         BNL-NUREG-50617     '           July-September 1976
         BNL-NUREG-50664                October-December  1976
         BNL-NUREG-50692                January-March 1977
         BNL-NUREG-50763                April-June 1977
         BNL-NUREG-50774                July-September 1977
         BNL-NUREG-50837                October-December  1977

    Brookhaven National Laboratory, Upton, New York.

-------
                               71

6.  H. L. Freese and W. T. Gregory III, "Volume Reduction of Liquid
Radioactive Wastes Using Mechanically Agitated Thin-Film
Evaporators," presented at the 85th National Meeting of the
American Institute of Chemical Engineers, June 4-8, 1978,
Philadelphia, Pennsylvania.

7.  A. H. Kibbey and H. W. Godbee, "Solid-Radioactive-Waste
Practices at Nuclear Power Plants," Nuclear Safety, Vol 16, No. 5,
p. 581, September-October 1975.

-------
                                   72

V.   LEACHING STUDIES

     The  most likely mechanism of  radionuclide release  to  the
surroundings is by solution  in the water  existing in the  environs  of
the  burial  site.  Therefore,  measurements are usually  attempted  to
indicate the rate at which radionuclides  are leached from the
solidified  products.   The leachability properties of a radwaste
solidification matrix  will strongly influence the amount  of treatment,
containment and surveillance that will be required.

     A  review of the various  leach rate tests and results  indicates a
wide variety of testing methods in use by industry and government
facilities  with little attempt at standardized techniques.  In
addition, the tests are not  performed to  simulate actual  ground  media
and  leachate conditions available at the  various facilities and
potential sites around the U.  S.  This along with the  many other
factors  involved, e.g., radionuclide measured, pH, temperature,
leachant composition,  have resulted in different test  procedures and
results  reported in many different units.  Also, in connection with
leaching, consideration must be given to  the corrosion of the various
containers, both internally  and externally.  Then there is also  the
deterioration of the matrices by  such things as radiation and
bacteriological attack.

     As a result many studies have been made on leach rates and several
studies  are still underway.   The  International Atomic  Energy Agency has
proposed a  leach test  method (1)  which is presently under evaluation,
in a modified form, as a standardized leach test procedure by the
Brookhaven  National Laboratory (2,3) for  low-level solidified
radioactive wastes.

     Studies have been  underway for many years to provide  information  on
the  leach resistance of various solidified waste products (4-16).
Leach  rates for alkali and alkaline earth, rare earth, and actinide
elements  from various  waste  matrices are  compared in Table III  (17).   A
comparison  shows: (a)  cement has  wide ranging leach rates; (b) calcines
are  extremely leachable; (c)  that for a given waste matrix the  leach
rates  for rare-earth and actinide elements are about a factor of 1,000
less than those for alkali and alkaline earth elements; and  (d)  the
leach  rates for rare-earth and actinide elements from  cements and
grouts are as low as those from glasses.  Additional leachability work
needs  to be accomplished in  the near future to permit  a more  complete
understanding of the environmental impact of both high- and  low-level
solidified radioactive waste.

-------
                                   73

                               Table III

 COMPARISONS OF LEACH RATES FOR VARIOUS SOLIDIFIED WASTE PRODUCTS (17)

                      Leach Rates, Grams/Cm^-Day
Calcines
Ceramics
  Phosphate
  Devitrified
  Phosphate glass

Glasses
  Borosilicate
  Phosphate
  Aluminosilicate

Bitumens
Cements
Grouts
Alkali and
Alkaline-Earth

ID-1-!

10-5-10-2

10-^-10-2
10-7-10-5
10~8-10-5
10-8-10-7

10-7-10-1*
10-6-10-1
10-7-10-1*
Rare
Earth

10-^-10-3

10-9-10-6
10-9-10-7
10-9-10-6
                                                           Actinide
                  10-8-10-7
                  10-9_io-7
                  10-7

-------
                                   74

1.0  References

    1.  E. D. Hespe (Ed),  "Leach Testing of Immobilized Radioactive
    Waste Solids," Atomic Energy Review, Vol.  9,  No.  1, 1971.

    2.  P. Colombo and R.  M.  Neilson, Jr.,  Critical Review of the
    Properties of Solidified Radioactive Waste Packages Generated at
    Nuclear Power Reactors,  Report BNL-NUREG-50591, Brookhaven National
    Laboratory, Upton, New York, December 1976.

    3.  P. Colombo and R.  M.  Neilson, Jr.,  Properties of Radioactive
    Wastes and Waste Containers, Quarterly Progress Report, April-June
    1976, Report BNL-NUREG-50571, Brookhaven National Laboratory,
    Upton, New York, October 1976.

    4.  J. E. Mendel,  A Review of Leaching Test Methods and the
    Leachability of Various Solid Media Containing Radioactive Waste,
    USAEC Report BNWL-1765,  Battelle Pacific Northwest Laboratories,
    Richland, Washington,  July 1973.

    5.  J. G. Moore, et al,  Development of Cementitious Grouts for the
    Incorporation of Radioactive Wastes, Part 1;   Leach Studies, USERDA
    Report ORNL-4962 April 1975 and Part 2:  Continuation of Cesium and
    Strontium Leach Studies,  USERDA Report ORNL-5142 September 1976,
    Oak Ridge National Laboratory, Oak Ridge,  Tennessee.

    6.  J. G. Moore, et al,  "Leach Behavior of Hydrofracture Grout
    Incorporating Radioactive Wastes," Nuclear Technology, Vol 32, p.
    39, January 1977.

    7.  H. W. Godbee and D.  S. Joy, Assessment of the Loss of
    Radioactive Isotopes From Waste Solids to the Environment, Part 1;
    Background and Theory, USAEC Report ORNL-TM-4333, Oak Ridge
    National Laboratory, Oak Ridge, Tennessee, February 1974.

    8.  J. A. Kelley and R.  M. Wallace, "Procedure for Determining
    Leachabilities of Radioactive Waste Forms," Nuclear Technology,
    Vol.  30,  p. 47,  July 1976.

    9.  W. F. Merritt, "High-Level Waste Glass: Field Leach Test,"
    Nuclear Technology,  Vol.  32, p. 88, January 1977.

-------
                               75

10.  J. A. Kelley and W. N. Rankin,  Correlation of Radionuclide
Leachabilities with Microstruotures of Glasses Containing Savannah
River Plant Waste, USERDA Report DP-1411, Savannah River
Laboratory, Aiken, South Carolina, May 1976.

11.  M. W. Wilding and D. W. Rhodes, Leachability of Zirconia
Calcine Produced in the Idaho Waste Calcining Facility, USAEC
Report IN-1298, National Reactor Testing Station, Idaho Falls,
Idaho, June 1969.

12.   B. E. Paige, Leachability of Alumina Calcine Produced in the
Idaho Waste Calcining Facility, USAEC Report IN-1011, National
Reactor Testing Station, Idaho Falls, Idaho, July 1966.

13.  R. H. Burns, "Solidification of Low- and Intermediate-Level
Wastes," Atomic Energy Review, Vol. 9, No. 3, 1971.

14.  J. H. Leonard and K. A. Gablin "Leachability Evaluation of
Radwaste Solidified with Various Agents," 74-WA/NE-8, The American
Society of Mechanical Engineers, New York, 1974.

15.  S. L. Cosgrove, et al, Interim Report on Evaluation of
Solidification Techniques For Low-Level Nuclear Waste Materials, J.
H. Leonard Associates, Cincinnati, Ohio, August 1974.

16.  H. E. Filter, The Dow System For Solidification of Low-Level
Radioactive Wastes From Nuclear Power Plants, The Dow Chemical
Company, Midland, Michigan, October 1976.

17.  Alternatives For Managing Wastes From Reactors And
Post-Fission Operations In The LWR Fuel Cycle., Volume 2;
Alternatives For Waste Treatment, Report ERDA-76-43 (Vol. 2 of 5),
U. S. Energy Research and Development Administration, Washington,
D. C., May 1976.

-------
                                   76

VI.  CONCLUSIONS

    In the foregoing sections brief descriptions have been presented of
several established and proposed processes for the solidification of
high- and low-level radioactive wastes and the glassification of
high-level wastes.  Each of the processes, as well as each of the
solidified waste products, have a number of advantages and
disadvantages.

    The safe and final disposition of radioactive wastes produced in
the nuclear fuel cycle is dependent on proper management, which
includes the solidification of these wastes.  Solidification is
desirable to provide protection against dispersal in adverse events.
Thermal and chemical stability, insolubility, and capability to
withstand impact are the major advantages.

    The importance of the solidification processes is seen in the
emphasis put on the development not only in the U.S., but in other
countries as well.  In the development of criteria, standards, and
regulations, the Environmental Protection Agency, the Nuclear
Regulatory Commission, and the Department of Energy have recognized the
need for containment and for solidification as a necessary requirement
(1-M).  Also, the National Academy of Sciences/National Research
Council Panel on Waste Solidification recognized, in its 1978 Report on
Solidification of High-Level Radioactive Wastes, that the first
requirement of the solid waste form, is that it be capable of
furnishing a major barrier to the migration of radionuclides from the
waste.

    As these processes and techniques receive more attention and
regulatory requirements surface, the best combinations and options will
also surface.  This implies that further improvements are possible and
that the limitations and potential problems have not necessarily been
identified.

1.0  REFERENCES

    1.  J.J. Cohen, et al; Determination of Performance Criteria for
    High-Level Solidified Nuclear Waste, USNRC Report NUREG-0279,
    Lawrence Livermore Laboratory, Livermore, California, July 1977.

-------
                               77

2.  Background Report;  Considerations of Environmental Protection
Criteria for Radioactive Waste, U.S. Environmental Protection
Agency, Office of Radiation Programs, Washington, D.'C., February
1978.

3.  E. L. Moore, Interim Report;  Commercial Waste Packaging
Studies, Task 2, High-Level Waste Package Acceptance Criteria
Study, USDOE Report RHO-ST-10, Rockwell Hanford Operations,
Richland, Washington, March 1978.

4.  P. Colombo and R. M. Neilson, Jr., Critical Review of the
Properties of Solidified Radioactive Waste Packages Generated at
Nuclear Power Reactors, Report BNL-NUREG-50591, Brookhaven National
Laboratory, Upton, New York, December 1976.

-------
                                   78

                               APPENDIX A

Presentation by William F. Holcomb, Office of Radiation Programs,
Environmental Protection Agency Before the Panel on Waste
Solidification Formed by the Committee on Radioactive Waste Management,
under  the  Sponsorship of the National Academy of Sciences/National
Research Council - February 1, 1977


     I  have been asked to present a few points concerning EPA's role  in
Radioactive Waste Management and our interest in waste solidification.

     First,  from the environmental protection standpoint, EPA is
concerned  with proper management of all types of radioactive wastes.
Second, our objective is one of containment rather than planned release
and  dispersion.  Third, EPA recognizes the need for environmental
criteria and guidance for waste management.  Fourth, EPA has been
directed to establish environmental radiation protection standards for
the  terminal storage of high-level radioactive waste by 1978, as a
result of  the review by the Presidential Nuclear Policy Task Force,
which  led  to the Presidential Nuclear Policy Statement of October 27,
1976.  We  are implementing a program to carry out these objectives.
Consideration will be given to the many key issues involved in making
these  decisions.  Philsophically, EPA's goal for the management of
radioactive wastes is to assure that no unwarranted risks are imposed
upon present or future generations.

     Pragmatically, EPA believes this goal can be achieved, through the
combined use of techniques for waste processing, containerization,
engineering controls in site selection, construction and operation;  in
judicious  use of carefully selected geologic environmental barriers;
and  in pre-planned emergency response procedures.  The intent of this
multi-faceted approach is to contain radioactive wastes within the
earth  until they have decayed to innocuous levels; and to protect
cost-effectively, the environment and public health from any
uncontrollable routine or accidental impacts, now and in the future.

    The Federal Government recognizes that efficient solutions to
radioactive waste disposal problems will require close coordination  and
cooperation among all agencies involved.  In this regard, EPA has
worked with NRC, ERDA,* USGS, CEQ, FEA,* ERC,* and OMB to assure the
development of coordinated radioactive waste disposal plans.  To
*  Now part of the Department of Energy

-------
                                   79

fulfill SPA's obligations we have developed a three-step waste
management program.  The first step involves the development of
pertinent environmental criteria.  The second step leads to the
establishment of generally applicable environmental standards.  The
third step consists of the development of standards and regulations in
areas where EPA has specific regulatory authority; such as under the
Ocean Dumping Act, the Federal Water Pollution Control Act, and the
Safe Drinking Water Act.*  In carrying out these steps, EPA's objective
is to assure adequate radiation protection for public health and the
environment over the period of time that radioactive wastes remain a
potential hazard.  In all three steps, the requirements of the National
Environmental Policy Act of 1969 (NEPA) will be followed; and full
participation by all interested parties will be afforded in public
hearings.

Generally Applicable Environmental Standards for High-Level Waste

    The work EPA has planned in the preparation of quantitative,
generally applicable, high-level radioactive waste disposal standards
includes broad participation by Government, public and private
organizations, as  specified in NEPA.  Also, these standards will not be
for regulation of  specific radwaste management practices, facilities,
or sites, (which is NEC's responsibility); instead, they will consist
of numercial standards to define present and future containment levels
that will be protective of the environment.

    In order to provide technical support for these environmental
standards, EPA has established a priority project as a major focal
point to help quantify potential long-term environmental impacts from
high-level waste.  This project will center on the following five areas:

    1.  The quantification of the high-level waste problem (i.e., the
magnitude of the source term), and comparison with other categories of
radioactive wastes;

    2.  The projected performance of engineering barriers (such as
solidified waste matrices and packaging containers);

    3.  Assessment of the effectiveness of environmental barriers to
mitigate radionuclide transport through geological strata and
formations;
•Additionally the Clean Air Act, as amended August 1977.

-------
                                   80

    U.  Assessment of the potential risk and resultant impacts
resulting from accidental releases;

    5.  Compilation and evaluation of all factors involved in
implementing high-level numerical standards.

    In carrying out steps 1 through 3, EPA does not intend to undertake
independent engineering programs, but to use available information from
industry, ERDA, NRC, USGS, and other sources to the maximum extent
possible.

    Due to the critical need for the timely completion of these tasks,
EPA has established a condensed time schedule to compile, analyze, and
publish the technical support information from this contract.

    As noted in the above project areas, we are particularly interested
in the effectiveness of engineering barriers.  The purpose of this task
is to assess both available and future technologies which provide
engineering controls to high-level waste.  There are already several
programs both in this country and abroad which are mature and are being
vigorously pursued.  However, the commercial application is still in
its infancy.  In addition, we must look at the capabilities of the
spent fuel itself for retaining waste for the so called throw away
option.

    We intend to assess the current process development programs in the
U. S., U. K., Germany and France concerning (a) solidification systems
for aqueous raffinate wastes, including such processes as calcination,
vitrification, sintering, pelletization, and encapsulation within a
metal matrix; (b) encapsulation of transuranic contaminated wastes; and
(c) container utilization for use at potential terminal disposal
sites.  Calcination and vitrification systems have been developed to  an
industrial stage in France and Germany, and; therefore, the
capabilities of these foreign technologies will be considered.

    It is our intent to rely on the information from ERDA and NRC
regarding the prime candidates for the solidification matrix processes
for high-level wastes.  We want to emphasize the need for early and
frequent information interchange.  It is our belief that the matrix and
container to be used is a important and fundamental part of the
disposal process.  One should keep in mind that a reliable matrix

-------
                                   81

and container will provide added assurance during the early years
(i.e., the first 20-30 years)  of the repositories validation phase,
that in the event of any problems arising, the container and/or matrix
will still be retrievable.

    The application of engineering controls,  in  terms of packaging,
.i.e., solidification, has always been  part of EPA's development plan
for attaining the necessary  information  for preparing the generally
applicable environmental standards.

-------
                                   82

                               APPENDIX B
          PRESIDENT CARTER'S STATEMENT ON NUCLEAR POWER POLICY
                             April 7, 1977
    There is no dilemma today more difficult to resolve than that
connected with the use of nuclear power.  Many countries see nuclear
power as the only real opportunity, at least in this century, to reduce
the dependence of their economic well-being on foreign oil - an energy
source of uncertain availability, growing price, and ultimate
exhaustion.  The U. S., by contrast, has a major domestic energy source
- coal - but its use is not without penalties and our plans also call
for the use of nuclear power as a share in our energy production.

    The benefits of nuclear power are thus very real and practical.
But a serious risk accompanies world-wide use of nuclear power - the
risk that components of the nuclear power process will be turned to
providing atomic weapons.

    We took an important step in reducing the risk of expanding
possession of atomic weapons through the Non-Proliferation Treaty,
whereby more than 100 nations have agreed not to develop such
explosives.  But we must go.further.  The U. S. is deeply concerned
about the consequences for all nations of a further spread of nuclear
weapons or explosive capabilities.  We believe that these risks would
be vastly increased by the further spread of sensitive technologies
which entail direct access to plutonium, highly enriched uranium or
other  weapons usable material.  The question I have had under review
from my first day in office is how can that be accomplished without
foregoing the tangible benefits of nuclear power.

    We are now completing an extremely thorough review of all the
issues that bear on the use of nuclear power.  We have concluded that
the serious consequences of proliferation and direct implications  for
peace and security - as well as strong scientific and economic evidence
- require
    - a major change in U. S. domestic nuclear energy policies and
programs; and
    - a concerted effort among all nations to find better answers  to
the problems and risks accompanying the increased use of nuclear power.

-------
                                   83

    I am announcing today some of my decisions resulting from that
review.

    First, we will defer indefinitely the commercial reprocessing and
recycling of the plutonium produced in the U. S. nuclear power
programs.  From our own experience we have concluded that a viable and
economic nuclear power program can be sustained without such
reprocessing and recycling.  The plant at Barnwell, South Carolina,
will receive neither federal encouragement nor funding for its
completion as reprocessing facility.

    Second, we will restructure the U. S. breeder reactor program to
give greater priority to alternative designs of the breeder, and to
defer the date when breeder reactors would be put into commercial use.

    Third, we will redirect funding of U. S. nuclear research and
development programs to accelerate our research into alternative
nuclear fuel cycles which do not involve direct access to materials
useable in nuclear weapons.

    Fourth, we will increase U. S. production capacity for enriched
uranium to provide adequate and timely supply of nuclear fuels for
domestic and foreign needs.

    Fifth, we will propose the necessary legislative steps to permit
the U. S. to offer nuclear fuel supply contracts and guarantee delivery
of such nuclear fuel to other countries.

    Sixth, we will continue to embargo the export of equipment or
technology that would permit uranium enrichment and chemical
reprocessing.

    Seventh, we will continue discussions with supplying and recipient
countries alike, of a wide range of international approaches and
frameworks that will permit all nations to achieve their energy
objectives while reducing the spread of nuclear explosive capability.
Among other things, we will explore the establishment of an
international nuclear fuel cycle evaluation program aimed at developing
alternative fuel cycles and a variety of international and U. S.
measures to assure access to nuclear fuel supplies and spent fuel
storage for nations sharing common non-proliferation objectives.

-------
                                   8U

    We will continue to consult very closely with a number of
governments regarding the most desirable multilateral and bilateral
arrangements for assuming that nuclear energy is creatively harnessed
for peaceful economic purposes.  Our intent to is develop wide
international cooperation in regard to this vital issue through
systematic and thorough interntional consultations.

-------
                                   85

                               APPENDIX C

                STATUS OF FUEL REPROCESSING IN THE U. S.
    There are more than 90 million gallons of high-level radioactive
wastes from Federal defense reprocessing plants presently in storage in
the United States.  Although some of this material has been reduced to
solids, the bulk remains in liquid form.  Most of this material, mainly
the byproduct of nuclear weapons production and related research, is
stored by the Department of Energy (DOE) at its Hanford Reservation in
Washington, the Savannah River Plant in South Carolina, and the Idaho
National Engineering Laboratory (INEL) in Idaho (1, 2).

    Also in storage are about 600,000 gallons of high-level wastes from
the reprocessing of fuels used in the operation of commercial nuclear
plants.  This waste is stored at the Nuclear Fuel Services' (NFS) spent
fuel reprocessing plant near West Valley, New York.  The NFS plant,
after operating for six years, was shut down in 1972 to expand from a
capacity of 300 to 750 metric tons per year and to modernize, bringing
the plant to current standards in all respects.  However, because of
the substantial scope of the changes in regulatory requirements NFS in
September 1976 announced that they no longer intended to reprocess fuel
at the West Valley Plant (3, 4).

    The General Electric Company built and tested the 300 metric tons
per year capacity Midwest Fuel Recovery Plant at Morris, Illinois; but
in 1971* the plant experienced technical problesm in starting up and was
not operable in its current configuration.  The plant did not reprocess
any spent fuel and has never been licensed to operate as a reprocessing
plant before being shutdown.  A final decision regarding the plant
remains to be made (5-8).

    Allied-General Nuclear Services  (AGNS) has under construction a
plant at Barnwell, South Carolina.   The AGNS plant is designed to
reprocess 1600 metric tons per year  using the conventional Purex
-process.  The construction permit for this plant was granted in 1970
with the central part of the plant,  the separations facility completed.

    However, completion and licensing of the plant is now delayed by a
complex situation including the general consideration of plutonium
recycle and the deferral of commercial reprocessing in the U. S.  The
plant still lacks the facilities needed for the solidification of
high-level waste and plutonium nitrate.  Even the existing separations
facility cannot be licensed to operate until the plutonium recycle

-------
                                   86

 issue  is resolved.  Meanwhile, licensing reviews and hearings on safety
 and  environmental issues are continuing (5-8).

     Finally, Exxon Nuclear Company, Inc., has filed a partial
 application in connection with their plans to construct and operate a
 fuel reprocessing plant in the Oak Ridge, Tennessee area.  The Exxon
 plant  design has an initial capacity of 1500 metric tons per year with
 a  growth capacity to 2100 metric tons per year.  The plant would use
 the  conventional Purex process for reprocessing low enriched uranium
 oxide  fuel.  Safety and environmental reviews have been by NRC, but a
 construction permit cannot be granted until the plutonium recycle
 question is resolved (5-9).

     The volumes of radioactive wastes will continue to grow as nuclear
 power  production increases in the years ahead.  If commercial spent
 fuel reprocessing were allowed, high-level radioactive wastes generated
 from these operations would require storage, reduction in volume, and
 finally disposal.  These are activities that are under continuing
 investigation.  They have not been fully resolved.  Because of the
 relatively long life of many of the radioactive materials involved,
 isolation of these wastes from the environment will be required for
 many thousands of years.

     The current approach to the management of high-level radioactive
 waste  has two principal steps:  (1) converting these wastes to a less
 mobile, solid form, thereby reducing the potential for accidental
 leakage to the environment, and (2) finding a suitable place to put
 them that will provide adequate isolation and protection to the public
 for  extremely long periods of time.

     In addition to reducing the potential for accidental leaks to the
 environment, solidification will also reduce the volumes of wastes
 requiring storage by some 80 percent and will facilitate transport of
 these materials to a permanent disposal site.  At the present time, DOE
 has  not yet announced the location or design of the Federal repository
 to which the solidified wastes will eventually be shipped.

     DOE is now evaporating its own high-level waste at the Hanford
Reservation, converting part of it into solid salt cakes which are
stored in underground tanks; and part into other solids which are
stored in massive concrete structures above the ground.  Similar
practices are being followed at the Savannah River Plant.  The DOE
facility at INEL converts liquid high-level wastes to a granular solid
form which is being stored in specially designed underground vaults
(10).

-------
                                   87

    Final plans for the ultimate disposal of the solidified high-level
wastes have not been developed, and there are many complex factors
which must be evaluated before a final selection is made.  DOE is
investigating several interim storage and ultimate disposal systems:
underground storage or disposal in salt formations, caverns mined in
bedrock, surface storage in massive concrete structures; and other more
exotic concepts such as seabed, ice cap, or space disposal (11).

    A major review of the technical alternative methods of managing
radioactive wastes from the commercial nuclear reactor fuel cycle
including the ultimate storage of this waste was conducted by the
Energy Research and Development Administration (which was reorganized
into DOE) in 1977.  The review document provides a comprehensive
compilation and description of alternative waste management operations,
concepts, and characterizes the technologies in terms of
state-or-readiness for use and development programs (12).

    The treatment and disposal of radioactive wastes involve both
environmental considerations and technical processing problems which
are complex and potentially far-reaching because of the long effective
half-lives of certain radionuclides.  Moreover, the legal and
administrative problems in radioactive waste management are more
involved and difficult than in most general industrial operations.

REFERENCES

1.  "Nuclear Power Fuel Reprocessing - Part I," page 61 December 28,
    1959; "Hanford: Atoms for Peace - Part II," page 113, January 11,
    I960; Idaho Falls: Atoms in the Desert - Part III," page 105,
    January 25, I960; "Savannah River Does It Cheaper - Part IV," page
    135, February 22, I960; Oak Ridge: Home of The Bomb - Part V," page
    137, March 7, I960, Chemical Engineering.

2.  G. W. Hogg, W. F. Holcomb, L. T. Lakey, L. H. Jones and D. D.
    Coward, A Survey of NRTS Wasts Management Practices, Volume I and
    II, USAEC Report ICP-10U2-I & II, Idaho National Engineering
    Laboratory, Idaho Falls, Idaho, September 1971.

3.  "Private Firm Gets Set to Reprocess Atomic Fuels," Chemical
    Engineering, page 68, April 29, 1963.

4.  "NFS Quits Reprocessing, But Waste Solidification Problem Remains,"
    Nucleonics Week, Vol. 17, No. 39, September 23, 1976.

-------
                                   88

5.  "Nuclear Reprocessing at Standstill Despite Demand," Chemical
    Engineering, page 68, January 6, 1975.

6.  L. J. Colby, Jr., "Fuel Reprocessing in the United States:   A
    Review of Problems and Some Solutions," Nuclear News,  page  68,
    January 1976.

7.  "Nuclear-power prospects soured by oxide - fuel - reprocessing
    stall," Chemical Engineering, page 123, February 28, 1977.

8.  Nuclear Fuel Reprocessing and High-Level Waste Management;
    Informational Hearings - Volume V, Reprocessing Part 2,  California
    Energy Resources Conservation and Development Commission,
    Sacramento, California, March 8, 1977.

9-  "Exxon Nuclear Tells Plan for Oak Ridge Facility," Nuclear  News,
    page 59, March 1976.

10. W. L. Lennemann, "Management of Radioactive Aqueous Wastes  from AEC
    Fuel-Reprocessing Operations," Nuclear Safety, Vol. 14,  No.  5,  page
    482, September-October 1973.

11. High-Level Radioactive Waste Management Alternatives,  USAEC Report
    WASH-1297, U. S. Atomic Energy Commission, Washington, D. C., May
    1974.

12. Alternatives for Managing Wastes From Reactors and Past  Fission
    Operations in The LWR Fuel Cycle, 5 Volumes, Report No.  ERDA-76-43,
    U. S. Energy Research and Development Administration,  Washington,
    D. C.,  May 1976.
                                            U. S. GOVERNMENT PRINTING OFFICE :  1978 — 620-007/3732

-------