Department of Energy

Washington, DC 20585

March 8, 2011

Mr. Larry Camper, Director

Division of Waste Management and

Environmental Protection

U. S. Nuclear Regulatory Commission

11555 Rockville Pike

Rockville, Maryland 20852

Dear Mr. Camper:

The Department of Energy (DOE) has prepared a draft evaluation to analyze whether the West

Valley Demonstration Project (WVDP) vitrification melter is waste incidental to reprocessing,

not high-level waste, and may be managed as low-level waste pursuant to DOE Order 435.1,

Radioactive Waste Management. Enclosed for your consultative review and comment is the

WVDP Draft Waste Incidental to Reprocessing Evaluation for the Vitrification Melter, and a CD

which contains the key references cited in the evaluation. Your consultative review of the draft

evaluation will provide valuable technical input. Please conduct your consultative review in

accordance with.the Interagency Agreement DE-EM0000284 between the Nuclear Regulatory

Commission (NRC) and DOE. We would appreciate your comments within 45 days of receipt of

this letter.

This draft evaluation for the melter is a pre-decisional document. In addition to consulting with

NRC, DOE is providing the opportunity for public and state review and comment. DOE will

consider NRC, public and state comments in its final waste determination.

If you have any questions or need additional information regarding the review of this Draft WIR

Evaluation, please contact me at (202) 586-0370 or Mr. William Levitan, Director, Office of

Environmental Compliance, at (301) 903-3339.

Sincer

Fra Marcinowski

Deputy Assistant Secretary for

Technical and Regulatory Support

Office of Environmental Management

Enclosures

®• Printed with soy ink on recycled paper
West Valley Demonstration Project

Draft Waste Incidental to Reprocessing Evaluation

for the

Vitrification Melter

March 2011

Prepared by the

U.S. Department of Energy

West Valley, New York

DRAFT FOR U.S. NUCLEAR REGULATORY

COMMISSION CONSULTATION REVIEW

AND PUBLIC REVIEW
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DRAFTWASTE INCIDENTAL TO REPROCESSING EVALUATION FOR THE WVDP VITRIFICATION MELTER

CONTENTS

NOTATION (Acronyms, Abbreviations, and Units) .............................................. v

1.0 INTRODUCTION ...................................................................................................... 1

1.1 Purpose ................................................................................................................... 2

1.2 Scope .................................................................................................................... 2

1.3 Technical Basis for the Draft Evaluation .............................................................. 3

1.4 Background .................................................................................................... 4

1.4.1 The W estern New York Nuclear Service Center ......................................... 4

1.4.2 Spent Nuclear Fuel Processing ............................................................... 6

1.4.3 The W est Valley Demonstration Project ........................................................ 7

1.4.4 Characterization of the Vitrification Melter ................................................. 7

1.4.5 Incorporation into a Solid Physical Form .................................................... 7

1.4.6 Potential Waste Disposal Facilities .............................. 8

1.4.7 Previous NRC Staff Review ...................................................................... 9

1.5 Organization of this Draft Evaluation ....................................... 9

2.0 BACKGROUND .................................................................................................. 11

2.1 Introduction .......................................................................................................... 12

2.2 Nuclear Fuel Reprocessing .................................................................................. 12

2.2.1 The Basic Process .................................................................................. 12

2.2.2 Contents of the W aste Storage Tanks ..................................................... 13

2.3 The Beginning of the West Valley Demonstration Project .................. 16

2.3.1 The W VDP Act ..................................................................................... 16

2.3.2 W aste Treatment Preparations ............................................................... 17

2.4 HLW Processing .................................................................................................... 17

2.4.1 Pretreatment of the W aste ..................................................................... 18

2.4.2 Vitrification of the HLW ......................................................................... 19

2.5 Vitrification Melter Description, Operation, and Characterization ........................... 20

2.5.1 Vitrification Melter Description ................................................................ 20

2.5.2 Operational History ............................................................................... 22

2.5.3 Characterization ................................................................................... 24

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CONTENTS

2.6 WVDP W aste Management Plans ....................................................................... 26

3.0 WASTE DETERMINATION CRITERIA ................................................................. 28

3.1 Waste Determination Criteria Background ......................................................... 28

3.2 Applicable Waste Determination Criteria ............................................................ 29

4.0 THE WASTE HAS BEEN PROCESSED TO REMOVE KEY RADIONUCLIDES TO THE

MAXIMUM EXTENT THAT IS TECHNICALLY AND ECONOMICALLY PRACTICAL .... 30

4.1 Key Radionuclides ........................................................................................... 31

4.1.1 Introduction ........................................................................................ 31

4.1.2 DOE Guidance on Key Radionuclides .................................................... 31

4.1.3 Requirements of 10 CFR 61.55 .............................................................. 32

4.1.4 Radionuclides in W est Valley HLW ........................................................ 33

4.1.5 Radionuclides Important to the Disposal Site Performance Assessments ...... 34

4.1.6 Conclusions About Key Radionuclides in the Vitrification Melter ............... 34

4.2 Removal to the Maximum Extent Technically and Economically Practical .............. 36

4.2.1 Technical Practicality Assessment .......................................................... 36

4.2.2 Methods Considered ............................................................................. 37

4.2.3 Flushing With Decontamination Solution .............................................. 38

4.2.4 Evacuated Canisters ...................................... 39

4.2.5 Mechanical Decontam ination ...................................................................... 40

4.2.6 Dism antlem ent .................................................................................... 41

4.2.7 Other Potential Decontamination Methods ............................................ 41

4.2.8 Summary and Conclusions .................................................................. 42

4.3 Economic Practicality Assessment ....................... .................................... 42

4.3.1 Evaluation of Additional Flushing .......................................................... 43

4.3.2 Evaluation of Vitrification Melter Dismantlement .................................... 45

4.3.3 Summary and Conclusions ................................................................... 46

5.0 THE WASTE WILL BE MANAGED TO MEET SAFETY REQUIREMENTS

COMPARABLE TO THE PERFORMANCE OBJECTIVES OF 10 CFR 61, SUBPART C..47

5.1 Introduction ................................................................................................... 48

5.2 DOE Safety Requirements ............................................................................... 49

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5.2.1 General Safety Requirement ................................................................ 50

5.2.2 Protection of the General Population from Releases of Radioactivity ...... 50

5.2.3 Protection of Individuals from Inadvertent Intrusion ............................... 53

5.2.4 Protection of Individuals During Operations ...................... 55

5.2.5 Stability of the Disposal Site After Closure ............................................. 57

5.3 The Vitrification Melter Meets Disposal Site Waste Acceptance Criteria ................. 58

5.3.1 Nevada National Security Site Waste Acceptance Criteria ........................ 59

5.3.2 The Vitrification Melter Waste Package ................................................ 60

5.3.3 The Vitrification Melter Meets Nevada National Security Site

W aste Disposal Criteria ........................................................................ 60

5.4 Meeting WCS Waste Acceptance Criteria .............................. 62

6.0 THE WASTE WILL NOT EXCEED CLASS C CONCENTRATION LIMITS AND WILL

BE MANAGED IN ACCORDANCE WITH DOE REQUIREMENTS AS LLW .................. 63

7.0 CONSULTATION WITH NRC AND OPPORTUNITY FOR PUBLIC COMMENT ........... 67

8.0 CONCLUSIONS .................................................................................................. 68

9.0 REFERENCES ... ........................................................ 69

APPENDICES

A Drawings of the Vitrification Melter and Its Shipping Container ................................... A-1

B Management Controls to Ensure Quality in this Draft Evaluation ................................. B-1

C Comparability of DOE, NRC, and Texas Requirements for LLW Disposal ...................... C-1

D Comparability of DOE, NRC, and Texas Dose Standards ............................................. D-1

E Consideration of the Criteria in Section 3116 of the Ronald W. Reagan

National Defense Authorization Act for Fiscal Year 2005 ............................................. E-1

FIGURES

1-1 The Western New York Nuclear Service Center and the WVDP .................... 5

1-2 The WVDP Area of the Center in 2006 ......................................................................... 6

2-1 The Process Building, the Vitrification Facility, and Ancillary facilities in 2006 ................. 13

2-2 Spent Fuel Reprocessing Diagram (PUREX Process) ...................................................... 14

2-3 Vitrification Facility General Arrangement ................................................................... 19

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FIGURES (continued)

2-4 Vitrification Process Flow Diagram ............................................................................. 20

2-5 Vitrification Melter Design Features ........................................................................... 20

2-6 Melter Installed in Vitrification Cell ............................................................................. 21

2-7 Vitrification Melter and Turntable Assemblies During Vitrification .................................. 22

2-8 Vitrification Melter Shipping Container ...................................................................... 23

4-1 Evacuated Canister Receiving Residual Glass From the Vitrification Melter ................... 40

4-2 Effectiveness of Actual Flushes and Hypothetical Additional Flushes ............................. 44

5-1 General Process Used to Ensure Performance Objectives are Achieved .............. 50

TABLES

2-1 Estimated Radionuclide Content (in Curies) of Tanks 8D-2 and 8D-4 at the

Com pletion of Fuel Reprocessing ................................................................................. 14

2-2 Vitrification Melter Total Activity Estimates .................................................................. 25

4-1 10 CFR 61.55, Table 1 (Long-Lived Radionuclides) ........................................................ 32

4-2 10 CFR 61.55, Table 2 (Short-Lived Radionuclides) ....................................................... 32

4-3 Key Radionuclides for this Draft Evaluation ................................. 34

4-4 Vitrification Melter Flushing Effectiveness ................................................................... 38

4-5 Estimated Effectiveness of Continued Vitrification Melter Flushing .................................. 43

4-6 Vitrification Melter Dismantlement Cost Analysis ......................................................... 45

5-1 Estimated Maximum Dose Impacts to a Representative Member of the Public

Associated With Vitrification Melter Disposal (mrem/y) .............................................. 52

5-2 Estimated Acute Dose Impacts to an Inadvertent Intruder Associated With Vitrification

Melter Disposal (m rem /y) ............................................................................................ 54

5-3 Vitrification Melter Radionuclide Concentrations in Bq/m3 ..................... . . . .. . . . . . . . .. . . . . . . . . . 61

5-4 Key WCS Federal Facility Waste Disposal Facility License Requirements ............................. 62

6-1 Vitrification Melter Waste Concentration Results .............................. 64

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NOTATION

Acronyms and Abbreviations

ALARA as low as reasonably achievable

Am americium

CFR Code of Federal Regulations

Cm curium

Co cobalt

Cs cesium

DOE Department of Energy

FR Federal Register

HLW high-level waste

i iodine

IP industrial package

K 1000

K potassium

LLW low-level waste

Ni nickel

Mn manganese

N north

NDA NRC-Licensed Disposal Area

Np neptunium

NRC Nuclear Regulatory Commission

.NYSERDA New York State Energy Research and Development Authority

PE-g plutonium equivalent grams

Pu plutonium

PUREX plutonium uranium extraction [process]

S south

SDA State-Licensed Disposal Area

Sr strontium

Tc technetium

THOREX thorium uranium extraction [process]

U uranium

WCS Waste Control Specialists

WIR waste incidental to reprocessing

WSMS Washington Safety Management Solutions

WVDP West Valley Demonstration Project

WVES West Valley Environmental Services

WVNSCO West Valley Nuclear Services Company

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NOTATION

Units

Bq Becquerel

Ci curie

cm centimeter

cm2 square centimeter

cm 3 cubic centimeter

g gram [mass]

h hour

kg kilogram, 1000 grams

L liter

m meter

m2 square meter

m 3 cubic meter

mg milligram, 0.001 gram

mrem millirem, 0.001 Roentgen equivalent man

millirem 0.001 Roentgen equivalent man

mSv millisievert, 0.001 sievert

mL milliliter, 0.001 liter

mrem millirem, 0.001 rem

pCi microcurie, 0.000001 curie

pL microliter, 0.000001 liter

nCi nanocurie, 10.9 curie

pCi picocurie, 10-12 curie

R Roentgen

rem Roentgen equivalent man

s second

Sv sievert

y year

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1.0 INTRODUCTION

Section Purpose

The purpose of this section is to provide introductory information that lays the

foundation for detailed discussions in later sections.

Section Contents

This section describes the purpose and scope of this draft evaluation, identifies the

technical requirements on which it is based, summarizes the background, and

outlines the contents of the rest of the evaluation.

Key Points

* The Department of Energy is evaluating whether the Vitrification Melter at the

West Valley Demonstration Project in New York meets the waste incidental to

reprocessing criteria of Department of Energy Manual 435.1-1, Radioactive Waste

Management

* Reprocessing of irradiated nuclear fuel at the West Valley site produced

approximately 600,000 gallons of high-level radioactive waste.

" The Department of Energy's West Valley Demonstration Project pretreated this

waste to partition it into a high activity waste stream that was vitrified into

borosilicate glass, and low activity waste streams that were solidified for disposal

as low-level waste.

* The Vitrification Melter was used in the vitrification process to turn high-level

waste slurry and glass formers into homogenized molten glass.

" The Department is responsible for disposal of low-level radioactive waste

produced by the solidification of high-level waste under the West Valley

Demonstration Project, as part of the Department's obligations under the West

Valley Demonstration Project Act.

" The Vitrification Melter has been characterized for radioactivity, determined to

have radionuclide concentrations that do not exceed limits for Class C low-level

radioactive waste, and will be grouted for stability purposes and packaged for

shipment to an offsite low-level waste disposal facility.

" To dispose of the used Vitrification Melter offsite as low-level radioactive waste,

the Department must first demonstrate it meets criteria of Department of Energy

Manual 435.1-1, Radioactive Waste Management, which it has accomplished by

the evaluations described in this draft evaluation.

" This draft evaluation is being issued initially in draft form to facilitate consultation

with the U.S. Nuclear Regulatory Commission, state, and public review and

comment, consistent with the Department's policy.

" The Department will make its final determination of whether the Vitrification

Melter is or is not high level waste after consideration of the Commission's

comments and any state and public comments on this draft evaluation.

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1.1 Purpose

The purpose of this draft waste incidental to reprocessing (WIR) evaluation is to evaluate

whether the Vitrification Melter, which was wetted by high-level waste (HLW) at the West Valley

Demonstration Project (WVDP) site in Western New York, meets the waste incidental to

reprocessing criteria, is not HLW, and may be managed as low-level waste (LLW) pursuant to

Department of Energy (DOE) Manual 435.1-1, Radioactive Waste Management.

This equipment was used as part of the process to vitrify liquid HLW that was generated by

commercial reprocessing of spent nuclear fuel by Nuclear Fuel Services, Inc. from 1966 to 1972

and stored in underground waste tanks at the West Valley site. HLW is the highly radioactive waste

material resulting from the reprocessing of spent nuclear fuel, including liquid waste produced

directly in reprocessing and any solid material derived from such liquid waste that. contains fission

products in sufficient concentrations, and other highly radioactive material that is determined,

consistent with existing law, to require permanent isolation (DOE Manual 435.1-1).

DOE plans to dispose of this equipment offsite to meet its obligations under the West Valley

Demonstration ProjectAct, Public Law 96-368, 42 U.S.C 2010a, which is described below.

Waste Incidental to Reprocessing Requirements

The term waste incidental to reprocessing refers The citation method applies to radioactive wastes

not to a type of waste but rather to a "process",

whereby "[c]ertain waste streams produced

during the generation of high-level waste may be

determined to. be non-high-level waste through

the waste incidental to reprocessing determination

process." (DOE Guide 435.1-1). DOE Manual

435.1-1 provides two methods for determining

whether waste associated with spent nuclear fuel

reprocessing can be determined to be incidental

to reprocessing and managed as LLW: the citation

method and the evaluation method.

such as, but not limited to, contaminated job

wastes including laboratory items such as clothing,

tools, and equipment which have been found to

routinely meet the criteria for management as LLW.

Consistent with DOE Guide 435.1-1, the equipment

addressed in this determination does not fall within

a category of materials to which DOE considers the

citation process can be applied. Therefore, the

evaluation method was used for the subject

equipment as described in this draft WIR evaluation.

1.2 Scope

This draft WIR evaluation applies only to

equipment.

the WVDP Vitrification Melter and to no other

The Vitrification Melter was previously decontaminated and characterized. Such considerations

are discussed further in Section 4.2 below.

DOE considers this draft evaluation to be a predecisional draft because DOE is consulting with

the U.S. Nuclear Regulatory Commission (NRC) and making the draft evaluation available for state

and public review before making a final determination'. Making this draft evaluation available to

the public at the time it is being provided ta NRC for review provides stakeholders an opportunity

to review it and submit comments to the Department before the final determination is made.

'This consultation is consistent with guidance in DOE Guide 435.1-1, Implementation Guide For Use With DOE

Manual 435.1-1, which states that while formal involvement by NRC in making incidental waste determinations is not

required, NRC involvement as a consultant to Field Offices and Programs on technical issues is recommended.

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After consideration of NRC, state, and public comments related to this draft evaluation, DOE

will issue its final determination on whether the Vitrification Melter is or is not HLW and may be

managed and disposed of as LLW.

1.3 Technical Basis for the Draft Evaluation

This draft evaluation has been prepared in accordance with DOE Manual 435.1-1, Radioactive

Waste Management, following guidance in DOE Guide 435.1-1, Implementation Guide For Use With

DOE M 435.1-1.

The method used involves evaluating whether the used Vitrification Melter is incidental to

reprocessing and can be managed under DOE's authority in accordance with requirements for LLW

waste. Criteria in Section II.B(2)(a) of DOE Manual 435.1-1 for determining that waste is incidental

to reprocessing, is not HLW, and can be managed as LLW are that the wastes:

(1) Have been processed, or will be processed, to remove key radionuclides to the maximum

extent that is technically and economically practical;

(2) Will be. managed to meet safety requirements comparable to the performance

objectives set out in 10 CFR Part 61, Subpart C, Performance Objectives, and

(3) Are to be managed, pursuant to DOE's authority under the Atomic Energy Act of 1954, as

amended, and in accordance with the provisions of Chapter IV of this Manual, provided the

waste will be incorporated in a solid physical form at a concentration that does not exceed

the applicable concentration limits for Class C low-level waste as set out in [Code of

Federal Regulations] 10 CFR 61.55, Waste Classification; or will meet alternative

requirements for waste classification and characterization as DOE may authorize.

This draft evaluation focuses on the criteria of DOE Manual 435.1-1, Section II.B(2)(a)

summarized above, which are discussed in Section 3 of this draft evaluation and addressed in detail

in Sections 4, 5, and 6, respectively.

Although.criteria in DOE Manual 435.1-1 for managing evaluated waste or equipment as LLW

are generally similar to the provisions in Section- 3116(a) of the Ronald W. Reagan National

Defense Authorization Act for Fiscal Year 2005, Public Law 108-375, that Act does not apply to the

West Valley site. Nevertheless, as a matter of policy, DOE has considered the Section 3116(a)(1)

criteria for perspective and information. This matter is addressed in detail in Appendix E.

WIR criteria were also established by NRC as part of the decommissioning criteria for the

WVDP in accordance with the WVDP Act (NRC 2002 and NRC 2003). However, these criteria were

issued "for the classification of reprocessing wastes that will likely remain in tanks at the site after

the HLW is vitrified" and "to clarify the status of and classify any residual waste present after

cleaning of the HLW tanks." These statements, which appear in Section 6.4 of the NRC

Implementation Plan (NRC 2003), indicate that these criteria apply to the HLW tanks themselves.

NUREG-1854, NRC Staff Guidance for Activities Related to U.S. Department of Energy Waste

Determinations (NRC 2007), reiterates that these NRC criteria apply to "any residual materials

remaining at the [West Valley] site, including any incidental [to reprocessing] waste." NRC also

acknowledges in NUREG-1854 that "For [West Valley] offsite waste disposal, it is DOE's

responsibility to determine which [WIR] criteria are applicable; for example, DOE may decide to

apply DOE Order 435.1."

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DOE's consultation with NRC on this draft WIR evaluation is consistent with NRC's role in

providing consultation on DOE activities related to the WVDP as expressed in the DOE-NRC

Memorandum of Understanding on this project (DOE and NRC 1981).

1.4 Background

The following general information is provided to put the draft evaluation into context. Section 2

provides more detailed background information on reprocessing of spent nuclear fuel, HLW waste

pretreatment, HLW vitrification, the Vitrification Melter design, residual radioactivity in the Melter,

and WVDP waste management plans.

1.4.1 The Western New York Nuclear Service Center

The WVDP is located at the Western New York Nuclear Service Center. The center is a 3,340-

acre site located approximately 30 miles south of Buffalo, New York. It is owned by the New York

State Energy Research and Development Authority (NYSERDA) on behalf of the State of New York,

the original owner. Figure 1-1 shows the location of the Center and the WVDP.

The Center was established in the early 1960s as a nuclear industrial complex that would

include spent nuclear fuel reprocessing and waste disposal facilities. The reprocessing facilities

were constructed and operated by a private company, Nuclear Fuel Services, Inc. Nuclear Fuel

Services also operated the two onsite radioactive waste disposal facilities - the NRC-Licensed

Disposal Area (NDA) and the State-Licensed Disposal Area (SDA).

The major facilities at the Center are located within a central area of approximately 200 acres.

Among these facilities are the reprocessing plant itself, also referred to as the Process Building, and

four underground liquid waste storage tanks. Figure 1-2 shows the WVDP area of the Center.

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Figure 1-1. The Western New York Nuclear Service Center and the WVDP

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Figure 1-2. The WVDP Area of the Center in 2006 (WVDP photo)

1.4.2 Spent Nuclear Fuel Reprocessing

Reprocessing operations at the West Valley site began in 1966 and were performed under

license from the U. S. Atomic Energy Commission; licensing and related regulatory functions of the

Atomic Energy Commission were transferred to the NRC in 1974. During six years of operation, the

plant reprocessed spent nuclear fuel containing approximately 626 metric tons of uranium (DOE

1996). Nuclear Fuel Services used the PUREX (plutonium uranium extraction) chemical separations

process for most of the irradiated fuel and the similar THOREX (thorium uranium extraction)

process for a single fuel lot enriched in uranium and thorium.

Approximately 600,000 gallons of liquid HLW were produced during reprocessing and stored in

Tanks 8D-2 and 8D-4. The approximately 560,000 gallons of neutralized PUREX waste inside Tank

8D-2 consisted of a bottom sludge layer containing insoluble hydroxides and other salts that

precipitated out of solution, covered by liquid (supernatant) rich in sodium nitrate and sodium

nitrite. Additionally, approximately 12,000 gallons of acidic THOREX waste commingled with

recovered thorium was stored in Tank 8D-4. Tanks 8D-1 and 8D-3 served as standby spares and

were not used by Nuclear Fuel Services for HLW storage, as indicated previously. (Rykken 1986)

In 1972, Nuclear Fuel Services shut the reprocessing plant down for expansion, modifications,

and additions. However, reprocessing never resumed.

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The HLW produced during plant operation and stored in the underground waste storage tanks

contained an estimated 31 million curies of radioactivity. This estimate, adjusted for radioactive

decay and in-growth to July 1987, included approximately 15 million curies of cesium 137 and its

short-lived progeny barium 137m, 14.8 million curies of strontium 90 and its short-lived progeny

yttrium 90, and approximately 196,000 curies of transuranic radionuclides, as well as lesser

amounts of other radionuclides including but not limited to carbon 14, iron 55, cobalt 60, and nickel

63 (Rykken 1986).

1.4.3 The West Valley Demonstration Project

Federal legislation was enacted in 1980 in the form of the WVDP Act to provide for solidification

of the high-level liquid radioactive waste generated by reprocessing, followed by clean-up of

related areas and wastes.

In 1982, DOE assumed control, but not ownership, of a 156-acre portion of the central area of

the Center in order to carry out its responsibilities under the WVDP Act. The NRC license technical

specifications were effectively suspended for the duration of the DOE project.

To meet the objective of solidifying HLW at the site, the WVDP developed and built the

Integrated Radwaste Treatment System and built the Vitrification Facility. The Integrated Radwaste

Treatment System was used to separate the waste into high activity and low activity radioactive

constituents and to solidify and store the low activity portion. Its primary component was the

Supematant. Treatment System, which decontaminated solutions from the underground storage

tanks through an ion exchange and removal process.

The Vitrification Facility was designed for the solidification of high-activity sludge and spent ion

removal media (zeolite) generated from Supernatant Treatment System operations. The former

reprocessing facilities were modified to accommodate the vitrification system and ancillary waste

treatment and storage systems, and some new facilities were constructed by the WVDP for this

purpose. For example, the new Supernatant Treatment System was installed by DOE adjacent to

and inside of Tank 8D-1.

DOE completed vitrification of the treated HLW in 2002. Since then, the WVDP has focused on

decontaminating and deactivating facilities and shipping LLW offsite. Alternatives for

decommissioning of the WVDP and the rest of the Center were evaluated in an Environmental

Impact Statement (DOE and NYSERDA 2010).2

1.4.4 Characterization of the Vitrification Melter

The Vitrification Melter has been characterized for radioactivity based on measured gamma

radiation levels and sample analytical data and found to not exceed NRC limits for Class C LLW

under 10 CFR 61.55 (WVES 2010). Section 2 of this draft evaluation provides more detail on the

characterization process.

1.4.5 Incorporation into a Solid Physical Form

The Vitrification Melter is in a solid physical form. However, void spaces in the Vitrification

Melter and its waste container will be filled with grout consisting of low-density cellular concrete,

2The Final Environmental Impact Statement for Decommissioning and/or Long-Term Stewardship at the West Valley

Demonstration Project and Western New York Nuclear Service Center(DOE and NYSERDA 2010) and the associated

Record of Decision (75 FR 75) were issued in 2010.

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mainly to help stabilize the component within the shipping container during transport and to

encapsulate surface contamination. This grout will not increase the waste disposal volume and was

not considered in the classification of the waste.

1.4.6 Potential Waste Disposal Facilities

DOE plans to ship the packaged Vitrification Melter to a suitable offsite LLW disposal facility,

either the Nevada National Security Site (formerly the Nevada Test Site) in Nevada or the Waste

Control Specialists (WCS) facility in Texas for disposal.

The DOE's Nevada National Security Site maintains two separate LLW disposal facilities known

as the Area 3 and the Area 5 Radioactive Waste Management Sites.

The Area 3 site, which is presently inactive, is located in the Yucca Flat area on the east side of

the 1,350-square-mile Nevada National Security Site. It has three disposal units located in shallow

depressions in the ground created by underground nuclear weapons detonations hundred of

meters below the surface. Waste is placed in these unlined craters and covered with soil. The Area

5 site is located in the Frenchman Flat area about 12 miles south of the Area 3 site. Waste in the

Area 5 site is generally disposed of in trenches approximately 22 feet deep and covered with eight

feet of soil. If DOE decides to dispose of the Vitrification Melter at the Nevada National Security

Site, it would be disposed of in the Area 5 site.

The commercial WCS radioactive waste disposal facility is located near Andrews, Texas on a

semi-arid, isolated 1,338-acre site. It is licensed by the State of Texas 3 for near-surface disposal of

Class A, B, and C LLW from Texas Compact 4 waste generators as well as Class A, B, and C and

mixed low-level Federal facility wastes. Federal facility waste, which includes LLW owned or

generated by DOE, would be disposed of in a separate landfill disposal unit called the Federal

Facility Waste Disposal Facility that is planned to be constructed and scheduled to open in 2011. If

DOE decides to dispose of the Vitrification Melter at the WCS facility, the Vitrification Melter waste

package would be disposed of as Class C LLW in the Federal Facility Waste Disposal Facility.6

3 Texas became an NRC Agreement State in 1963, and as an NRC Agreement State, regulates and licenses certain

radioactive materials within its borders, including the disposal of certain LLW. The Texas program is periodically

reviewed by the NRC; under the NRC Agreement State Program, NRC evaluates technical licensing and inspection

issues from Agreement States, and periodically evaluates State rules for health and safety and compatibility with NRC

requirements. Pursuant to applicable law, including Title 30 of the Texas Administrative Code, WCS was issued a

license, with conditions, by the Texas Commission on Environmental Quality in 2009 for a compact waste disposal

facility and a federal waste disposal facility. The license was subsequently amended, and in 2010, WCS filed a

request for an additional minor license amendment, which is being reviewed by the Texas Commission on

Environmental Quality.

4 The Texas compact consists of the states of Texas and Vermont. Waste generators in these states are authorized to

dispose of LLW in the WCS Texas Compact disposal facility.

5 The license does not authorize disposal of greater-than-Class C LLW.

6 Texas law requires, among other things, that the licensee submit a written agreement, signed by the Secretary of

Energy, to the Texas Commission on Environmental Quality, stating that the Federal government will assume all

right, title, and interest in land and buildings for the disposal of Federal facility waste (Texas Administrative Code,

Title 30, Part 1, §336.909). The DOE and the Texas Commission on Environmental Quality entered into a nonbinding Memorandum of Agreement in January 2010, concerning the requirements in §336.909, and recognizing that

DOE, in its sole discretion, will decide whether to award a contract for waste disposal to WCS and whether to dispose

of LLW in the WCS Federal Facility Waste Disposal Facility. Should DOE decide to dispose of the Vitrification Melter in

the WCS Federal Facility Waste Disposal Facility, such disposal would be in accordance with the license, as may be

amended, and the WCS waste acceptance procedures and plans approved by the Texas Commission on

Environmental Quality.

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Offsite disposal of WVDP LLW at the Nevada National Security Site is consistent with DOE's

February 25, 2000 Record of Decision for the Department of Energy's Waste Management

Program: Treatment and Disposal of Low-Level Waste and Mixed Low-Level Waste; Amendment of

the Record of Decision for the Nevada Test Site (Federal Register 65 FR 38) related to the Waste

Management Programmatic Environmental Impact Statement (DOE 1997). DOE observed in this

Record of Decision that the arid Nevada Test Site (now named the Nevada National Security Site)

provides environmental benefits for waste disposal, such as geology, that greatly restrict the

potential for any contamination movement into groundwater.

DOE selected offsite disposal of WVDP LLW at DOE facilities or commercial facilities in its

Record of Decision on the Final West Valley Demonstration Project Waste Management

Environmental Impact Statement, DOE/EIS-0337 (70 FR 115). Therefore, disposal of the

Vitrification Melter at either the Nevada National Security Site or the WCS facility is consistent with

this Record of Decision.

As discussed in subsequent sections of this draft evaluation, the requirements for disposal of

LLW such as the Vitrification Melter waste package at the commercial WCS facility are similar to the

requirements for disposal at DOE's Nevada National Security Site. The State of Texas regulations

that apply to the WCS facility mirror the NRC regulations for LLW and are comparable to DOE LLW

disposal requirements. For example, the provisions concerning performance objectives, solid waste

form, waste stability, and Class C concentration limits would be comparable regardless of the site

selected by DOE for disposal of the waste.

DOE's decision on the disposal site to be used is not within the scope of this draft evaluation.

Any DOE decision on the facility .to which the Vitrification Melter waste package would be sent

would be made after the final WIR evaluation and determination, following consideration of NRC

and public comments on this draft evaluation, and after DOE confers with appropriate State

officials in the state where the waste package may be disposed.

1.4.7 Previous NRC Staff Review

Nuclear Regulatory Commission staff members make routine visits to the WVDP to monitor

DOE activities pursuant to the Commission's review responsibilities under the WVDP Act. Two such

visits in 2004 focused on vitrification equipment. During these visits, NRC staff members reviewed

information on the characterization, classification, and packaging for the Vitrification Melter and

concluded that all applicable regulatory requirements had been met (NRC 2004).

1.5 Organization of this Draft Evaluation

Information in the remainder of this draft evaluation is presented as follows:

Section 2 describes the waste stored in Tanks 8D-2 and 8D-4 at the conclusion of

reprocessing and the HLW pretreatment process. It also describes the Vitrification Melter,

including the characterization that has been performed.

Section 3 describes DOE Manual 435.1-1 waste incidental to reprocessing waste determination

criteria.

Section 4 describes how key radionuclides have been removed from the Vitrification Melter to

the maximum extent technically and economically practical.

Section 5 discusses how safety requirements comparable to NRC performance objectives in 10

CFR 61, Subpart C, and how waste acceptance criteria for the potential disposal sites (the

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Nevada National Security Site Area 5 Radioactive Waste Management Site or the WCS site) will

be achieved.

Section 6 explains that the radionuclide concentrations in the packaged Vitrification Melter are

less than Class C concentration limits, and that the Vitrification Melter will be managed in

accordance with Chapter IV of DOE Manual 435.1-1.

Section 7 describes consultation with the NRC and the opportunity for public comment.

Section 8 summarizes DOE's preliminary conclusions related to the draft WIR evaluation.

Section 9 identifies the references cited in the draft WIR evaluation.

Appendix A provides copies of drawings for the Vitrification Melter and its shipping container.

Appendix B describes management controls used to ensure quality in this draft evaluation.

Appendix C discusses the comparability of DOE, NRC, and State of Texas requirements for

LLW disposal.

Appendix D discusses the comparability of DOE, NRC, and State of Texas radiation dose

standards.

Appendix E discusses the criteria in Section 3116 of the Ronald W. Reagan National Defense

Authorization Act for Fiscal Year 2005.

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2.0 BACKGROUND

Section Purpose

The purpose of this section is to provide detailed background information to support

the discussions in the sections that follow.

Section Contents

This section describes nuclear fuel reprocessing, the contents of the underground

waste storage tanks at the conclusion of spent fuel reprocessing, initial West Valley

Demonstration Project activities, the subject Vitrification Melter equipment,

radiological characterization of the equipment, and waste management plans.

Key Points

" A salt/sludge separation process was used to treat the high-level waste in Tanks

8D-2 and 8D-4 to produce a high-activity waste mixture to be stabilized by

vitrification into a borosilicate glass waste form suitable for geologic disposal.

" The Melter - the last component in the vitrification system - consists of an

electrically heated box structure, approximately 10 feet on each side, with a

stainless steel outer shell and an interior lined with refractory materials.

" The Vitrification Melter was used from 1996 through 2002 to heat the highactivity waste mixture and glass formers toturn them into homogenized molten

glass.

" In September 2002, after completion of vitrification operations, the Vitrification

Melter was flushed, emptied, and shut down.

" The Vitrification Melter has been characterized for residual radioactivity using

measured dose rates, dose conversion factors calculated by two different

geometric computer models, and radionuclide scaling factors based on glass

sample analytical data.

* Based on the characterization results, the Vitrification Melter meets concentration

limits for Class C low-level waste.

" The Vitrification Melter has been loaded in a custom-built shielded steel shipping

container in preparation for offsite disposal.

" The Department plans to ship the packaged Vitrification Melter to an offsite waste

disposal facility. For purposes of this draft evaluation, this facility is assumed to be

either the Nevada National Security Site or the licensed WCS Federal waste facility

in Texas.

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2.1 Introduction

This section establishes the context for the evaluations of the Vitrification Melter that are

described in Sections 4, 5, and 6 by providing the following information:

Section 2.2 provides a brief review of nuclear fuel reprocessing, with emphasis on

management of the liquid HLW stream that impacted the equipment that is the subject of

this draft evaluation.

* Section 2.3 provides summary information on the WVDP and on preparations for waste

treatment.

Section 2.4 summarizes how the waste was pretreated and describes how the HLW was

stabilized into a vitrified glass form for transport to an appropriate Federal repository for

permanent disposal.

* Section 2.5 describes the Vitrification Melter, explains how it was used, and describes how

it was characterized for residual radioactivity, providing information important in the

evaluations described in Sections 5, 6, and 7.

* Section 2.6 describes DOE plans for disposing of the Vitrification Melter.

Note that the brief descriptions of removal of HLW from the underground storage tanks and its

pretreatment and vitrification are provided here solely for information purposes; the effectiveness

of removal of key radionuclides from the underground waste tanks is not being evaluated in this

draft evaluation. Section 4 addresses removal of key radionuclides from the Vitrification Melter.

2.2 Nuclear Fuel Reprocessing

Spent nuclear fuel began arriving at the Western New York Nuclear Service Center in 1965.

Reprocessing was accomplished in 27 campaigns, 11 of which involved fuel from the N-Reactor at

the U.S. Atomic Energy Commission's Hanford, Washington site. The other spent nuclear fuel came

from commercial nuclear reactors. Reprocessing recovered both uranium and plutonium from the

fuel, and produced the approximately 600,000 gallons of liquid HLW mentioned previously.

2.1.1 The Basic Process

Reprocessing operations were conducted in the Process Building. Figure 2-1 shows this

building, the Vitrification Facility, and other nearby facilities. (The Vitrification Facility, which is

described in Section 2.4.2, was built by the WVDP for solidification of the HLW.)

The first step in reprocessing operations involved disassembling fuel assemblies and chopping

them into pieces. The small pieces of fuel were transported to vessels where they were dissolved in

concentrated nitric acid, which transformed them into *an aqueous stream containing uranium

nitrate, plutonium nitrate, and fission products.

As noted in Section 1, Nuclear Fuel Services mainly used the PUREX process. This five-stage

solvent extraction process used tributyl phosphate/n-dodecane solution to separate the fission

products from the uranium and plutonium. (The tributyl phosphate process step generated the

majority of the HLW produced in the reprocessing.)

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Following initial separation, the uranium-bearing and plutonium-bearing solutions underwent

additional purification. The purified product solutions were then concentrated, packaged, stored,

and shipped offsite. A simplified diagram representing the PUREX fuel reprocessing operation

appears in Figure 2-2. (Note that the West Valley plant did not produce oxide products - U03

product and PuO 3 product as shown on the diagram - but rather uranyl nitrate and plutonium

nitrate, materials that could be converted to the oxide products.)

2.2.2 Contents of the Waste Storage Tanks

The largest volume of waste (approximately 560,000 gallons) remaining from the normal

operation of the plant in reprocessing uranium fuel was neutralized by the addition of sodium

hydroxide before transfer to Tank 8D-2. Neutralizing the initially acidic HLW prior to transfer caused

most of the fission product elements (the major exception was cesium) to precipitate out and form

sludge at the bottom of Tank 8D-2. Therefore, the HLW was not homogeneous but was comprised

of supernatant (liquid) and sludge (solids).

The approximately 12,000 gallons of acidic high-level radioactive liquid waste produced in

reprocessing thorium-enriched uranium fuel using the THOREX process was stored in Tank 8D-4

without being neutralized.

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Nitric Acid

Solvent tributyl

phosphate and

n-dodecane

Irradiated Fuel

Dislto+ nNti

Aci

+SeaainoEsin

Off-gases and

Cladding (Hulls)

- Off-Gases

Hig-Level

Waste

U03 Product PuO2 Product

Figure 2-2. Spent Fuel Reprocessing Diagram (PUREX Process)

Table 2-1 shows the estimated radionuclide inventory in Tank 8D-2 and Tank 8D-4 at the

completion of reprocessing, adjusted for decay and in-growth to July 1987. Information in this

table is based on analytical data from samples collected by the WVDP in the initial project waste

characterization program begun shortly after DOE assumed control of the project premises (Rykken

1986 and Eisensatt 1986).

Table 2-1. Estimated Radionuclide Content (in Curies) of Tanks 8D-2 and 8D-4 at the

Completion of Fuel Reprocessing (from Eisenstatt 1986 Table 6, fission and activation products

decay-corrected to July 1987)

Radionuclide SupernatantI Tank 8D-2 Tank 8D-2 Sludge Tank 8D-4 j Total

-3 . 9.5E+1 .i. <2.OE+O0 :' <9..7E+01.

C-14 1.4E+02 0() 1.4E+02

.e-5 ._ _ __ __._. ,,. ... • . 1.OE+03.

Ni-59 8.2E+01 8.2E+01

C.0-60: ... 47EO0A 11.2E-i03 :.2+03..

Ni-63 8.9E+02 6.4E+03 (1) 7.3E+03

Se-79 3.7E+01 ,,0 3.7E+01

Sr-90 2.9E+03 6.9E+06 5.0E+05 7.4E+06

y_90(2) .2.9E+03 6.9E+06 5.0E+05 7.4E+06

Zr-93 (1) 2.3E+02 (1) 2.3E+02

Nb-93m 2.3E+02 . )2.3E+02

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Table 2-1. Estimated Radionuclide Content (in Curies) of Tanks 8D-2 and 8D-4 at

the Completion of Fuel Reprocessing (Continued) (from Eisenstatt 1986 Table 6,

fission and activation products decay-corrected to July 1987)

Radionclide Tank 8D-2

Radionuclide Supernatant Tank 8D-2 Sludge Tank 8D-4 Total

Tc-99 1.6E+03 (1) 8.0E+01 1.7E+03

Ru-106 1.3E+02 <3.1E-01 1.3E+02

Rh-106 1.3E+02 <3.1E-01 1.3E+02

Pd-107 1.2E+00 (1) 1.2E+00

Sb-125 4.8E+01 4.5E+03 (1) 4.5E+03

.Te-125m 1.1E+01 • 1,05÷03 () . , .OE,+03.

Sn-126 4.OE+01 4.OE+01

Sb-126m . 4.0E+01 (1) 4.0E+01

Sb-126 * ) 5.6E+01 (1) 5.6E+01

1-129: .2.1E-01 . (1)..*.. <1.5E&01 <3.65&01

Cs-134 1.4E+04 (1) 2.9E+02 1.4E+04

Cs-135 1.6E+02 * )(1) 1.6E+02

Cs-137 7.3E+06 (1) 5.1E+05 7.8E+06

Ba-137m(2) 6.8E+06 (1) 4.8E+-05 7.3E+06

Ce-144 2.9E-05 1.4E+01 <2.OE-02 1.4E+01

Pr-144 2.9E-05 1.4E+01 <2.OE-02 1.4E+01

Pm-147 1.7E+02 3.1E+05 4.5E+03 3.1E+05

Sm-151 1.1E+00 2.1E+05 1.5E+01 2.1E+05

Eu-152 4.2E-02 4.2E+02 5.8E+00 4.3E+02

Eu-154 1.4E+01 1.3E+05 2.6E+03 1.3E+05

Eu-155 2.3E+00 2.3E+04 3.1E+02 2.3E+04

Th-232 (1) 1.6E+00 1.6E+00

U-233 4.9E-01 6.9E+00 2.6E+00 1.0E+01

U-23 .... 2.9E- 4.OE+00 3.0E-01 4.6E+00,O

U-235 6.4E-03 8.9E-02 4.9E-03 1.OE-01

U-236 1.9E-02 2.7E-01 1,0E-02 3.OE-01

U-238 5.7E-02 7.9E-01 - 6.1E-04 8.5E-01

Np-237 1 1.11E+o1 (l) 11E0l

Np-239 (1) 2.4E+03 (1) 2.4E+03

Pu-238 1.3E+02 6.5E+03 5.3E+02 7.2E+03

Pu-239 2.5E+01 1.7E+03 1.7E+01 1.7E+03

Pu-240 1.9E+01 1.3E+03 9.O0+00o 1.3E+503

Pu-241 1.5E+03 8.5E+04 9.3E+02 8.7E+04

Pu-242 2.5E-02 i.7E+00 - 1.3E-02 1.7E+00. .

Am-241 7.2E+04 2.7E+02 7.2E+04

Am-242 * ( .2.1E+01 (1) 2.iE+01

Am-242m 2.1E+01 2.1E+01

Am-243. . -. 2.4E03': .8.8E+00 . 2.4E-03ý..:,.

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Table 2-1. Estimated Radionuclide Content (in Curies) of Tanks 8D-2 and 8D-4 at

the Completion of Fuel Reprocessing (Continued) (from Eisenstatt 1986 Table 6,

fission and activation products decay-corrected to July 1987)

Tank 8D-2 Radionuclide Supernatant Tank 8D-2 Sludge Tank 8D-4 Total

Cm-242 (1) 2.2E+00 <1.1E-03 2.2E+00

Cm-243 (_ _ _1.7E+02 5.OE-02 1.7E+02

Cm-244 _ _ ) _ 2.2E+04 1.6E+01 2.2E+04

Cm-245 (1) 1.OE+01 1.2E-03. 1.OE+01

Cm-246 (1) 4.3E+00 (1) 4.3E+00

NOTES: (1) Not present or undetectable.

(2) The progeny of Sr-90. and Cs-137 are included here because they were reported in Table 6 of Eisenstatt 1986.

2.3 The Beginning of the West Valley Demonstration Project

This brief summary begins with a discussion of key points in the WVDP Act, then summarizes

the WVDP preparations for waste treatment.

2.3.1 The WVDP Act

After conducting studies and hearings related to dealing with the radioactivity from

reprocessing activities that remained at the West Valley facility, the U.S. Congress enacted the

WVDP Act. The WVDP Act directed the DOE to carry out the following activities:

(1) Solidify the HLW;

(2) Develop containers suitable for permanent disposal of the solidified HLW waste;

(3) Transport the waste to an appropriate Federal repository for permanent disposal;

(4) Dispose of LLW and transuranic waste produced in the solidification of the HLW; and

(5) Decontaminate and decommission the tanks, facilities, materials, and hardware used in the

project in accordance with requirements prescribed by the NRC.

The Act directed DOE to enter into a cooperative agreement with the State (NYSERDA) for the

State to make available the facilities and HLW necessary to carry out the project, without transfer

of title, with DOE providing technical assistance in securing required license amendments. This

cooperative agreement became effective on October 1, 1980 (DOE and NYSERDA 1981).

The Act also directed DOE to enter into an agreement with the NRC for informal review and

consultation on the project by NRC and to afford NRC access to the site to monitor activities under

the project to assist DOE in protecting health and safety. This agreement was formalized in a

memorandum of understanding signed in September 1981 (DOE and NRC 1981).

In addition, pursuant to the WVDP Act, applicable decontamination and decommissioning

activities shall be in accordance with such requirements as the NRC may prescribe. The review and

consultation by the NRC shall not include or require formal procedures or activities by the NRC

pursuant to the Atomic Energy Act or any other law.

In accordance with the WVDP Act, and under the cooperative agreement with DOE, NYSERDA

made available to DOE, without transfer of title, the 156.4-acre area known as the Project

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Premises 7

. DOE assumed operational responsibility for this area in February 1982 and employed

the West Valley Nuclear Services Company (WVNSCO) as the managing and operating contractor

for the WVDP8.

2.3.2 Waste Treatment Preparations

To manage the HLW, DOE selected onsite processing using a salt/sludge 9 separation process

(47 FR 179). This approach involved use of a chemical pretreatment method to:

(1) Separate the major radioactive species (i.e., cesium 137) from the liquids held in Tanks . 8D-2 and 8D-4,

(2) Combine the separated cesium 137 with the sludge to produce a high activity waste

mixture, and

(3) Vitrify the resulting high activity waste mixture into an approved glass waste form.

DOE refined the approach as details of plans for separation of the waste streams and

vitrification of the HLW were developed. Preparations for these activities included:

" Constructing the Supernatant Treatment System, including a new building to house a valve

aisle and other equipment;

* Constructing the HLW Transfer Trench and associated piping to transport waste from the

underground tanks to the Vitrification Facility;

" Modifying Tank 8D-1 for use as a treatment tank, including installation of ion exchange'°

columns containing zeolite;

* Adapting existing tanks within the Process Building for use with the Integrated Radwaste

Treatment System;

" Removing equipment from other areas of the Process Building to provide room for

Integrated Radwaste Treatment System equipment;

• Removing equipment from the Chemical Process Cell in the Process Building and setting up

this shielded area for interim storage of the HLW canisters;

* Installing the waste mobilization pumps, waste transfer pumps, and other necessary

hardware in Tanks 8D-1, 8D-2, and 8D-4 to facilitate removal of tank contents; and

" Constructing the Vitrification Facility and installing the vitrification equipment, including the

Melter.

2.4 HLW Processing

Processing of HLW involved two major programs: pretreatment, followed by vitrification.

Two other small parcels of land were transferred to DOE in 1986, bringing the actual total to approximately 167

acres. The Project Premises is commonly referred to as being 200 acres in size.

8 In October 2007, West Valley Environmental Services LLC (WVES) superseded WVNSCO as DOE's site contractor.

9 Salt in this context means the liquid portion of the stored waste, i.e., the supernatant.

10 Although the zeolite-loaded columns are commonly referred to as ion exchange columns, the zeolite actually

functions as a molecular sieve characterized by pores and crystalline cavities of uniform dimensions that adsorb

certain molecules.

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2.4.1 Pretreabnent of the Waste

The pretreatment program consisted of four major tasks: (1) supernatant processing, (2)

PUREX sludge washing, (3) PUREX/TFHOREX sludge washing, and (4) zeolite transfer to Tank 8D-2.

DOE consulted with NRC on the treatment processes, consistent with provisions of the DOE/NRC

Memorandum of Understanding.

The major steps involved: (1) decontaminating PUREX supernatant from Tank 8D-2 in the

Supernatant Treatment System columns inside Tank 8D-1, (2) transferring the decontaminated

liquid to the Liquid Waste Treatment System evaporator, and (3) transferring the evaporator

concentrates to the Cement Solidification System set up in the 01-14 Building, where they were

solidified in cement in 71-gallon steel drums.

The Integrated Radwaste Treatment System was operated from May 1988 until November

1990, pretreating approximately 600,000 gallons of PUREX supernatant. Cesium 137 was removed

from this liquid at a decontamination effectiveness of greater than 99.99 percent and adsorbed on

zeolite, which was stored under liquid in Tank 8D-1. Some Pu removal was also accomplished (Kelly

and Meess 1997).

The PUREX sludge in Tank 8D-2 was washed from October 1991 to January 1992. Washing

consisted of adding a sodium hydroxide solution to increase the alkalinity of the liquid waste and

adding additional water.

The washing process dissolved the hard layer of sludge present in the tank, solubilized the

sulfate and other undissolved salts present in the sludge, and mixed the interstitial liquid trapped in

the sludge with the wash solution. This sludge washing was performed in conjunction with

sequential operation and lowering of the five mobilization pumps in Tank 8D-2 to thoroughly mix

the contents.

A second wash of the PUREX sludge was performed from May to June 1994 to further reduce

the amount of sulfates in the high activity waste prior to vitrification. As with the first sludge wash,

sodium hydroxide and water were added to Tank 8D-2 while the mobilization pumps mixed the

contents of the tank. Following the second wash, the wash solution was again processed through

the Integrated Radwaste Treatment System from June to August of 1994.

Following the completion of sludge washing, final preparations were made to complete the

installation of the HLW transfer system which links all three underground waste storage tanks that

contained HLW (Tanks 8D-1, 8D-2, and 8D-4) to the Vitrification Facility using double-contained

piping run in underground concrete trenches and pits. To facilitate waste removal, waste transfer

pumps were installed in Tanks 8D-1, 8D-2, and 8D-4. Tank 8D-2 was prepared for the acidic

THOREX addition from Tank 8D-4 during November and December 1994 by increasing its alkalinity

with sodium hydroxide. The acidic THOREX was transferred from Tank 8D-4 to Tank 8D-2 and

neutralized during January 1995. (Kelly and Meess 1997)

Following neutralization, sodium nitrite was added to Tank 8D-2 to minimize pitting corrosion

that could result from the large amount of nitrates in the THOREX solution (Kelly and Meess 1997).

After mixing the contents of Tank 8D-2 - which included washed PUREX sludge, sludge wash liquid,

THOREX precipitates, and THOREX solution - using the waste mobilization pumps, the THOREX/

PUREX wash liquid was processed through the Integrated Radwaste Treatment System.

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2.4.2 Vitrification of the HLW

The Vitrification Facility was designed and used to stabilize the following waste streams in a

borosilicate glass matrix: (1) the radioactive high activity sludge that had been generated during

PUREX reprocessing of spent uranium fuel, (2) THOREX waste that resulted from the reprocessing

of thorium-uranium fuel, and (3) contaminated cesium-loaded zeolite generated during

Supernatant Treatment System operations. Figure 2-3 shows the general arrangements in the

facility.

In Cell Operating Equipment Aisle Crane

Maintenance Room / oCMR pe

Aisli

Figure 2-3. Vitrification Facility General Arrangement (The Equipment Decontamination Room and the

Chemical Process Cell, where the HL W canisters are temporarily stored, are part of the original Process Building.)

The Vitrification Facility building houses the Vitrification Cell, operating aisles, and a control

room. The shielded Vitrification Cell contained the equipment used to concentrate the high activity

waste slurry, mix it with glass formers (oxide additives), melt this mixture to form borosilicate

glass, pour the molten glass into the stainless steel canisters, seal the canisters, and

decontaminate the canister exteriors. Among this equipment were the Concentrator Feed Makeup

Tank, the Melter Feed Hold Tank, and the Vitrification Melter.

Figure 2-4 illustrates the general process flow, and shows the location of the Melter in the

vitrification process.

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Zeodite Siarage

ED-I

CESI UM Waste

mI

r upratant

Sludge

81-4 PUREX Waste

THOREX

Waste

Waste Tank

Farm

.. . ..- .-.- 

Canister

Canister Cbosure

Slaae Inspection

rrangser Mt Tunnel

Vitrification Facility

Figure 2-4. vitrification Process Flow Diagram (for information, not to scale)

Between 1996 and 2002, the WVDP retrieved the high activity waste from the tanks and

stabilized, it by vitrification. Deactivation of the Vitrification Facility, which included removal of all of

the process equipment, was completed in July 2005. The WVDP is currently focusing on facility.

decontamination and deactivation, waste management, and plans for decommissioning. DOE plans

for waste shipment are summarized in Section 2.6 below.

2.5 Vitrification Melter Description, Operation, and Characterization

This subsection describes the Vitrification Melter, summarizes the radiological Heating Bars Not Shown for Clarnty Off-Gas

characterization process, and provides the -. Melter Feed West Electrode

characterization results. 15_ 4Z. Melter Cav4y

2.5.1 Vitrification Melter Description • Electrode Wall

Also known as the Slurry-Fed Ceramic lPenetroalo

Melter, the Vitrification Melter consists of

an electrically heated box structure Floor Electrode

approximately 10 feet on each side.

The outer shell is formed of stainless J

steel. The interior is lined with a composite Overow Discharge Passage

of various refractory materials" to with- Chamber [Not to Scale]

stand high temperatures. The sides and Figure 2-5. Vitrification Melter Design Features

l The surface in contact with the molten glass is composed of MonofraxTM K-3. This material is a chrome alumina

fusion cast refractory with high corrosion resistance used in applications such as fiberglass furnaces and nuclear

encapsulation melters (RHI 2010).

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bottom of the outer shell are covered with a cooling water jacket. The Melter weighs approximately

106,000 pounds (Vance, et al. 1997). Figures 2-5 and 2-6 show the Melter. 12 Appendix A provides

copies of Melter drawings.

: Turnrtable.

Fioure 2-6. Melter Installed in Vitrification Cell (WVDP ohoto)

The Vitrification Melter is divided into two sections. The main section contains the Melter cavity,

which has an overall height of approximately 4.5 feet. The upper part of the cavity is rectangular in

shape, with the lower part in the form of an inverted truncated rectangular pyramid. During normal

operation, the Vitrification Melter would accommodate 227 gallons (approximately 30 cubic feet) of

slurry. The slurry was heated with three electrodes, one of which served as the floor of the vessel.

The discharge section of the Vitrification Melter contains a primary and a secondary pour

chamber, each with spouts and silicone carbide radiant heaters. The Vitrification Melter was

mounted on a track system in the northeast corner of the Vitrification Cell.

12 Note that the unit contained three external electrode assembles - rather than two as indicated in Figure 2-5 -

which were removed flush with the side of the outer box surface before the unit was placed into its shipping

container, which is shown in Figure 2-8 below.

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2.5.2 Operational History

The Melter was installed by the WVDP along with the other vitrification equipment and used

throughout vitrification operations. During operation, batches of HLW slurry feed material were

transferred from the Melter Feed Hold Tank to the Melter. Inside the Vitrification Melter, calcined

wastes and glass formers were melted and fused into a glass pool where they homogenized.

Homogenized molten glass in the Vitrification Melter was periodically transferred into a

stainless steel canister held in position beneath the Melter by the canister turntable. After the glass

was allowed to cool, the loaded canister was transferred from the turntable to the weld station,

where it was sealed with a welded lid. Figure 2-7 illustrates how the Vitrification Melter was

positioned over the turntable to fill a HLW canister.

Off-Gas Cleaner

PassiveCoo~led

Off-Gas

Film Cool.,--

To SBS"L

Aidift --

Overflow Heaters

ý;hernio,

Melter Viewing

, "•System

ouples

Melter Electrodes

-North Cell Wall

Canister

Access Po I

Y

- - Filing Canister

Load Cell ---

-Turntable Frame

Figure 2-7. Melter and Turntable Assemblies During Vitrification.

During the glass melting process, volatile elements evaporating from the glass pool were

vented to the off-gas treatment system, along with feed particles entrained in the process off-gas.

The silicon carbide heaters used in the discharge section of the Vitrification Melter were

expected to have limited service life based on system testing, and two heater assemblies failed

during use. Another operating problem was encountered when the glass exit port plugged with

glass near the end of vitrification operations. This condition was not unexpected and this possibility

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led to the twin pour chamber design. The secondary pour chamber was brought on line after this

event. (Petkus, et al. 2003)

In September 2002, after completion of HLW vitrification, the Vitrification Melter was flushed

with decontamination solutions, emptied using two evacuated canisters, and shut down, as

described in Section 4. Based on recorded data, approximately 2200 kg of molten residual glass -

about 88 percent of the estimated amount present - were removed from the Vitrification Melter

during this process.

The residual material which could not be removed by these processes - approximately 12

percent of the original amount - consists of the glass in the plugged discharge port and

approximately eight inches of glass in the bottom of the Melter cavity, along with a thin layer on

the sides of the Melter cavity.

It was estimated that approximately 300 kg of glass remained in the Melter cavity, a heel of

approximately eight inches. This estimate was based on a combination of level detector responses,

preliminary canister weights, and thermocouple responses during the evacuation evolution. Visual

observation indicated that the residual glass was present in portions of the Melter at the 10-inch

level. (Lachapelle 2003)

Section 4 discusses the efficiency of the decontamination processes.

In 2004, the Vitrification Melter was removed from the Vitrification Cell and loaded into its

shipping container, which is shown in Figure 2-8. Container drawings appear in Appendix A.

This Industrial Package (IP)-2 type shipping container is formed of steel that is six inches thick

on the sides and four inches thick on the top and bottom. The empty container weighs

approximately 210,000 pounds. The total package weight with the Vitrification Melter and lowdensity cellular concrete, which will be used to fill both the Melter cavity and the container, will be

approximately 360,000 pounds. (WMG 2004b)

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2.5.3 CHARACTERIZATION

Details of the Vitrification Melter characterization appear in the characterization report (WMG

2004a). The characterization process took into account three different sources of residual

radioactivity:

" The Vitrification Melter cavity;

* The plugged discharge port, which is separated from the cavity by 18 inches or more of

refractory material; and

* The exterior of the Vitrification Melter.

Data Used in Characterization

The characterization for residual activity inside the Vitrification Melter was based on measured

gamma dose rates and analytical data from two samples of residual glass.

For the Vitrification Melter cavity, gamma dose rates were measured in January 2003 with an

unshielded Ludlum Model 133-7 Geiger-Mueller detector probe lowered into nozzles located above

the residual glass. These dose rates ranged from 330 R/h to >1000 R/h, with levels of 748 and 749

R/h measured through a nozzle located directly above the residual glass. A level of 2.1 R/h

measured one foot above the lid of the Vitrification Melter assembly was also used in the

calculations. A dose rate of 40 R/h measured at the southernmost lip of the west discharge cavity

was used in calculations for the discharge cavity source. (Lachapelle 2003)

The two glass samples were obtained from the evacuated canisters. . They were analyzed in

the WVDP Analytical and Process Chemistry Laboratory. To promote accuracy, three separate

analyses (the original and two replicates) were performed on material from each canister.

Copies of both the radiation survey records and the laboratory analysis reports appear in the

characterization report (WMG 2004a).

The exterior surfaces of the Vitrification Melter were characterized based on measured surface

contamination levels. The maximum beta-gamma contamination level on portions of the surface

impacted by accidental spills of slurry was determined to be 16.7 pICi/cm 2

; the maximum alpha

surface activity from airborne contamination was determined to be 0.0443 pCi/cm 2 (WMG 2004a).

Both types of contamination are now considered to be fixed because Polymeric Barrier System, a

latex fixative coating, was applied to the outside of the Vitrification Melter before it was placed in

the shipping container.

Models Used in Characterization of the Vitrification Melter Interior

Two geometry models were used to calculate dose conversion factors, which were combined

with measured dose rates to estimate residual cesium activity in the two areas of the Vitrification

Melter.

A model prepared using the QAD-CGGP-A computer code' 3 was used to represent the complex

geometry of the Vitrification Melter cavity. This model accounted for a layer of contamination

0.125-inch thick on the cavity surfaces to its full height of 28 inches, as well as the eight inches of

solidified glass in the bottom of the cavity. The basis for the thickness of the contamination layer is

13QAD-CGGP-A is a point-kernel computer code for calculating fast neutron and gamma ray penetration through

various shield configurations. It was developed by Atomic Energy of Canada, Ltd.

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previous WVDP experience in Vitrification Melter testing described in the Slurry Fed Ceramic Melter

Disassembly Report (Brooks 1993)'1.

It was also assumed that glass had penetrated fissures in the refractory material to a depth of

approximately 0.5 inch, based on experience in Vitrification Melter testing (Brooks 1993). This

factor was taken into account in the model (WMG 2004a).

A model prepared with the Megashield TM code was used for the plugged discharge port.

Solidified glass was assumed to completely fill this discharge tube and the discharge cavity itself to

a level of four inches.

Assumptions Used in Characterization of the Vitrification Melter Interior

Two assumptions added conservatism to the calculations:

* It was assumed that the measured gamma radiation levels were due entirely to Cs-137,

even though other gamma emitters such as Am-241 are present and contribute to the

measured gamma radiation levels' 5; and

" It was assumed that radioactivity in the Vitrification Melter cavity did not contribute to the

40 R/h dose rate used to characterize radioactivity in the discharge port, although the

detector was positioned such that radiation from the cavity influenced this measurement.

Results for the Vitrification Melter Interior

These calculations - which were performed using the data, models, and assumptions identified

above - produced an estimate of 4062 curies of cesium 137 in the Vitrification Melter cavity and an

additional 252 curies of cesium 137 in the plugged discharge tube. Scaling factors based on the

glass sample analytical data were used with the RADMANTM computer code to estimate the

amounts of other radionuclides in the Vitrification Melter. Table 2-2 shows the estimates.

Table 2-2. Vitrification Melter Total Activity Estimates(')

Nuclide Activity (Ci) Nuclide Activity (Ci)

C-14 2.12E-02 U-232 5.01E-04

K-40 8.19E-02 U-234 9.81E-03

Mn-54 8.57E-02 U-235 3.76E-04

Co-60 8.33E-02 U-238 2.25E-03

Sr-90 2.47E+02 Np-237 6.20E-03

Zr-95 1.65E+00 Pu-238 6.84E-01

Tc-99 1.1IE-02 Pu-239 1.59E-01

Cs-137 4.31E+03 Pu-241 3.12E+00

Eu-154 1.21E+00 Am-241 3.OOE+00

Th-228 4.09E-02 Am-243 3.50E-02

Th-230 3.65E-04 Cm-242 7.33E-02

Th-232 4.01E-04 Cm-243 1.68E-02

NOTES: (1) From WMG 2004a, Table 3, as of October 1, 2004. Note that Table 5-3 identifies the concentrations

of additional radionuclides. However, these additional radionuclides are present in such small

amounts that their contributions to the total activity are negligible.

14 The WVDP conducted a five-year test program producing glass from simulated waste slurry with an actual

vitrification melter that was disassembled and examined after completion of the testing (Brooks 1993).

15 This approach is conservative because less than 100 percent of the gamma radiation levels used in the calculation

were produced by cesium 137, a condition that results in slightly over estimating the amount of cesium 137 present.

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The estimates in Table 2-2 total approximately 4570 Ci as of October 1, 2004. The estimated

amounts of residual glass in the Vitrification Melter are as follows:

* Melter heel, 300 kg;

0 On Melter cavity surfaces and in refractory material fissures, 26 kg; and

* The plug in the discharge port, 99 kg.

These amounts total approximately 425 kg. These estimates were based on the west discharge

port containing a solid plug of glass, the discharge tube being completely filled, and the discharge

cavity being filled to a height of four inches. (WMG 2004a)

Radioactivity on the Outside Surfaces

The amount of contamination on the outside surfaces of the Vitrification Melter was estimated

by using the maximum measured surface activity. The total surface area used in the calculations

was 9.08E+05 cm 2 (approximately 980 ft2). The slurry contamination level of 16 pCi/cm 2 was used

for 10 percent of the surface area and the maximum measured removable activity of 1.OE+08

dpm/100 cm2 was used for the rest of the surface area, with the slurry and airborne contamination

radionuclide distributions used in the calculations. The calculations yielded a total estimate of 5.2 Ci

as of October 1, 2004, with approximately 4.1 Ci of this amount associated with Cs-137. This

activity represents approximately 0.11 percent of the total estimated residual radioactivity

associated with the Vitrification Melter. (WMG 2004a)

NRC Staff Monitoring Visit Assessment

As explained in Section 1.4.7, NRC representatives concluded in an assessment performed in

connection with two 2004 monitoring visits that the Vitrification Melter was appropriately

characterized, packaged, and prepared for offsite disposal in accordance with regulatory

requirements. The purpose of these visits was to evaluate preparations for packaging and disposal

of three vitrification process components, including the Melter. (NRC 2004)

The NRC representatives evaluated the characterization and waste profile methodologies, the

design and fabrication of the waste packages, and verification that the packages were prepared for

shipment and disposal in accordance with applicable requirements. They reviewed the

characterization data, the methods used to determine activity amounts, and the sample analytical

data used to develop radionuclide scaling factors, and interviewed cognizant site personnel. (NRC

2004)

The NRC conclusion about characterization described in Section 1.4.7 independently confirms

the validity of the characterization process used by DOE for the Vitrification Melter.

2.6 WVDP Waste Management Plans

This section briefly summarizes DOE plans for managing WVDP LLW, including the Vitrification

Melter.

The Department evaluated management of radioactive waste at West Valley in its WVDP Waste

Management Environmental Impact Statement (DOE 2003). In its Record of Decision (70 FR 115),

DOE decided that, for WVDP LLW and mixed LLW that is currently in storage at the site or that will

be generated at the site over the next ten years (i.e., through 2015), DOE will to ship such WVDP

LLW and mixed LLW offsite for disposal, in accordance with all applicable requirements, at

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commercial sites (such as EnergySolutions [formerly known as Envirocare], a commercial

radioactive waste disposal site in Clive, Utah), one or both of two DOE sites, the Nevada Test Site

[now called the Nevada National Security Site] in Mercury, Nevada, or the Hanford Site in Richland,

Washington, or a combination of commercial and DOE sites. This Record of Decision included

wastes that DOE may determine in the future to be LLW or mixed LLW pursuant to a WIR

determination using the evaluation process [this draft evaluation, for example].

In June 2006, DOE issued a Supplement Analysis (DOE 2006) to its WVDP Waste Management

Environmental Impact Statement to address shipment of components from the Vitrification Facility

and shipment of an increased volume of LLW. This Supplement Analysis specifically addressed the

Vitrification Melter. The analysis noted that the Vitrification Melter may be shipped to one of four

sites that can accept Class C LLW, including the Nevada Test Site (now called the Nevada National

Security Site) and the WCS site.

In 2001, after completing the required approval process, the WVDP received approval to ship

LLW to the Nevada Test Site (now called the Nevada National Security Site) and has been shipping

LLW to that facility since that time (WVES and URS 2009). DOE plans to ship the Vitrification Melter

waste package to an offsite LLW disposal facility. For the purposes of this draft WIR evaluation,

this facility is assumed to be either the Nevada National Security Site or the WCS disposal facility

for Federal LLW in Texas. A final decision on the facility to which the Vitrification Melter waste

package will be sent will be made after consideration of NRC and public comments on this

predecisional draft evaluation, and after DOE confers with appropriate State officials for the states

where the waste package may be disposed.16

Section 6 demonstrates that the Vitrification Melter waste package does not exceed

concentration limits for Class C LLW.

16 DOE also will comply with the provisions in DOE Manual 435.1-1, Section I.2.F(4), concerning approval of

exemptions for use of non-DOE disposal facilities, should DOE decide to dispose of the Vitrification Melter in the WCS

facility.

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3.0 WASTE DETERMINATION CRITERIA

Section Purpose

The purpose of this section is to describe the criteria applicable to this draft waste

incidental to reprocessing evaluation.

Section Contents

This section provides background information on Department of Energy criteria that

apply to this draft waste incidental to reprocessing evaluation that have been

considered, providing more detailed information on this matter than is provided in

Section 1 of this draft evaluation. It then describes the Department's criteria that

apply to management of the Vitrification Melter.

Key Points

Applicable criteria appear in Department of Energy Manual 435.1-1, Radioactive

Waste Management.

3.1. Waste Determination Criteria Background

The WVDP is required to comply with two separate and distinct sets of criteria to determine

whether waste from reprocessing is incidental to reprocessing, is not HLW and may be managed as

other than HLW through a demonstration of compliance with the appropriate waste determination

criteria:

* DOE Manual 435.1-1, Radioactive Waste Management, applies to wastes that DOE disposes

of offsite.

" The NRC's Final Policy Statement on Decommissioning Criteria for the West Valley

Demonstration Project (M-32) at the West Valley Site (NRC 2002) describes criteria for

classification of "any residual wastes present after cleaning of the high-level radioactive

waste (HLW) tanks at West Valley."

Because the NRC West Valley decommissioning criteria policy statement (NRC 2002) does not

apply to waste shipped offsite for disposal, as explained in Section 1.3.1, this draft WIR evaluation

for the Vitrification Melter is being performed in accordance with DOE Manual 435.1-1. DOE's waste

determination criteria are described in Section 3.2.17

17 Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 sets forth criteria

which apply only to certain waste from reprocessing of spent nuclear fuel that under certain conditions can be

disposed of in Idaho or South Carolina. The WVDP is not within a covered state included in the scope of Section

3116. Although these criteria do not apply to the WVDP, DOE has, for perspective and information, considered

whether disposal of the Vitrification Melter would be consistent with the relevant provisions of Section 3116. This

matter is addressed in detail in Appendix E of this draft evaluation.

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3.2 Applicable Waste Determination Criteria

Section I.l.C of DOE Manual 435.1-1 provides that all radioactive waste subject to DOE Order

435.1 be managed as HLW, transuranic waste, LLW, or mixed LLW. DOE Manual 435.1-1, Section

II.B, also states that waste resulting from reprocessing spent nuclear fuel determined to be

incidental to reprocessing is not HLW and shall be managed in accordance with the requirements

for transuranic waste or LLW, as appropriate. The determination that waste is incidental to spent

nuclear fuel reprocessing, and therefore not HLW, is called a "waste incidental to reprocessing

determination," which is also referred to in this draft evaluation as a WIR determination or waste

determination.

DOE Manual. 435.1-1, Section II.B.2(a), lists three criteria to demonstrate that wastes related

to spent nuclear fuel reprocessing is not HLW and should be managed as LLW:

(1) Criterion 1 - the wastes have been processed, or will be processed, to remove key

radionuclides to the maximum extent that is technically and economically practical;

(2) Criterion 2 - the wastes will be managed to meet safety requirements comparable to the

performance objectives set out in 10 CFR Part 61, Subpart C, Performance Objectives, and

(3) Criterion 3 - the wastes are to be managed, pursuant to DOE's authority under the Atomic

Energy Act of 1954, as amended, and in accordance with the provisions of Chapter IV of

DOE Manual 435.1-1, provided the waste will be incorporated in a solid physical form at a

concentration that does not exceed the applicable concentration limits for Class C low-level

waste as set out in 10 CFR 61.55, Waste Classification; or will meet alternative

requirements for waste classification and characterization as DOE may authorize. 18

As will be demonstrated in the next three sections of this draft evaluation, DOE has evaluated

the Vitrification Melter against these criteria, and, for the reasons presented, the draft evaluation

shows that the Vitrification Melter meets the applicable criteria and can be managed and disposed

of as LLW.

18 DOE did not authorize alternative requirements for the Vitrification Melter.

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4.0 THE WASTE HAS BEEN PROCESSED TO REMOVE KEY

RADIONUCLIDES TO THE MAXIMUM EXTENT THAT IS

TECHNICALLY AND ECONOMICALLY PRACTICAL

Section Purpose

The purpose of this section is to evaluate whether the waste (i.e., the

Vitrification Melter) has been processed to remove key radionuclides to the.

maximum extent that is technically and economically practical.

Section Contents

This section describes the process used in determining the key radionuclides in

the Vitrification Melter and identifies those radionuclides. It then describes the

technical and economic practicality evaluations that have been performed and

their results.

Key Points

" The evaluations show that key radionuclides have been removed from the

Vitrification Melter to the maximum extent that is technically and

economically practical.

" The key radionuclides in the Vitrification Melter are those long-lived and

short-lived radionuclides listed in Tables 1 and 2 of the Nuclear Regulatory

Commission's regulations in 10 CFR 61.55, three of which are important to

the performance assessment of the WCS low-level waste disposal facility,

along with four other radionuclides that are important to the results of the

performance assessment of the Nevada National Security Site low-level waste

disposal facility.

" Evaluation of representative potential methods of removing key radionuclides

showed that flushing the Vitrification Melter with decontamination solutions,

using the Evacuated Canister System, and dismantling the Melter were the

only methods technically practical.

" Flushing with decontamination solutions and the Evacuated Canister System

were used and proved to be effective in removing key radionuclides.

• The economic practicality assessment evaluated additional flushing while the

vitrification process was still operational and concluded that this approach

would not have been economically practical.

" This assessment also evaluated Vitrification Melter dismantlement and

concluded that this approach also would not have been economically practical

because of increased worker radiation dose and other factors.

" The economic practicality assessment compared the impacts of additional

flushing and Vitrification Melter dismantlement expressed as estimated

monetary costs with the potential benefits expressed in terms of the

monetary value of the reduction in collective radiation dose.

" This assessment demonstrated that further efforts to remove key

radionuclides would have increased worker radiation dose without resulting

benefits; disposal of the Vitrification Melter would have an insignificant

impact on the performance of the disposal area in protecting the health and

safety of workers and the public.

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The criterion of DOE Manual 435.1-1, Section II.B.2(a) to be evaluated is the first criterion:

"[The subject wastes] have been processed, or will be processed, to remove key radionuclides

to the maximum extent that is technically and economically practical."

4.1 Key Radionuclides

This section begins with a brief introduction that describes the various factors considered,

provides additional information on these factors, discusses their relevance to key radionuclide

selection, and concludes with the identification of key radionuclides for this draft WIR evaluation.

4.1.1 Introduction

The key radionuclides in this draft WIR evaluation are based on consideration of the following

information:

" Guidance in DOE Guide 435.1-1 on identification of key radionuclides;

* NRC requirements for classification of radioactive waste for near-surface disposal that

appear in 10 CFR 61.55;

• Radionuclides known to be.present in the West Valley HLW;

" The relationship between DOE disposal site waste acceptance criteria and the performance

of DOE LLW disposal sites in meeting objectives for protecting individuals and the

environment;

* The radionuclides of importance in the performance assessment of the Nevada National

Security Site Area 5 LLW disposal area, although such consideration is not required by DOE

Manual 435-1 or DOE Guide 435.1-1;

* The State of Texas requirements for classification of radioactive waste in the Texas

Administrative Code, which mirror the NRC requirements in 10 CFR 61.5519;

* Radionuclides specifically limited in the WCS radioactive material license; and

* Radionuclides important to meeting the State of Texas performance objectives - which

mirror the NRC performance objectives in 10 CFR Part 61, Subpart C - based on the

radionuclides of importance in the performance assessment of the WCS LLW disposal

facility.

Consideration of this information will ensure that those radionuclides in the Vitrification Melter

that could contribute significantly to radiological risks to workers, the public, and the environment

are identified and taken into account.

4.1.2 DOE Guidance on Key Radionuclides

DOE guidance on selection of key radionuclides is provided in Section II.B of DOE Guide 435.1-

1, with the applicable portion reading as follows:

"... it is generally understood that [the term] key radionuclides applies to those

radionuclides that are controlled by concentration limits in 10 CFR 61.55. Specifically these

are: long-lived radionuclides, C-14, Ni-59, Nb-94, Tc-99, 1-129, Pu-241, Cm-242, and

alpha emitting transuranic nuclides with half-lives greater than five years and; short-lived

radionuclides, H-3, Co-60, Ni-63, Sr-90, and Cs-137. In addition, key radionuclides are

19 The Texas requirements, license limits, and performance assessment information are considered for completeness

and additional information, although such consideration is not specifically required by DOE Manual 435-1 or DOE

Guide 435.1-1.

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those that are important to satisfying the performance objectives of 10 CFR Part 61,

Subpart C [for near-surface radioactive waste disposal facilities]."

This guidance considers both the waste classification requirements in 10 CFR 61.5520 for

radioactive waste destined for near-surface disposal and achieving the waste disposal site

performance objectives.

4.1.3 Requirements of 10 CFR 61.55

The radionuclides listed in the guidance found in DOE Guide 435.1 appear in 10 CFR 61.55 in

the form of two tables, which are reproduced here as follows.

Table 4-1. 10 CFR 61.55, Table 1 (Long-Lived Radionuclides)

Radionuclides Concentration (Ci/m 3)

C-14 8

C-14 in activated metal 80

Ni-59 in activated metal 220.

Nb-94 in activated metal 0.2

TC:-99 3

1-129 0.08

.Alpha Emitting Transuranic (TRU) nuclides with half-life 100(M)

greater than 5 years

Pu-241 3,500(l)

Cm-242 201000(l)

NOTES: (1) These values are in units of nanocuries per gram.

Table 4-2. 10 CFR 61.55, Table 2 (Short-Lived Radionuclides)

Concentration (Ci/m 3)

Radionuclides Column 1 Column 2 Column 3

[Class A] [Class B] [Class C]

Total of all nuclides with less than 5 year half- 700 (1) (1)

life 700_______

H-3 40 (1) (1)

Co-60 700 C')

Ni-63 3.5 70 700

Ni-63 in activated metal 35 700 7000

Sr-90 0.04 150 7000

Cs-137 1 44 4600

NOTE: (1) There are no limits established for these radionudides in Class B or C wastes. Practical

considerations such as the effects of external radiation and internal heat generation on

transportation, handling, and disposal will limit the concentrations for these wastes. These

wastes shall be Class B unless the concentrations of other nuclides in the table determine the

waste to be Class C independent of these nuclides.

The concentrations given in these tables are used for waste classification purposes.

Classification is determined by concentrations of long-lived radionuclides, by concentrations of

short-lived radionuclides, or by both in those cases where the waste contains both types of

20 Title 30 of the Texas Administrative Code has similar requirements (Rule §336.362, Appendix E).

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radionuclides. The tables in the Texas Administrative Code mirror the 10 CFR 61.55 tables in Table

4-1 and 4-2 (Rule §336.362, Appendix E, Table I and 1I).

4.1.4 Radionuclides in the West Valley HLW

The West Valley HLW contained a mixture of both long-lived and short-lived radionuclides.

Table 2-1 of this draft evaluation includes, for example, long-lived radionuclides listed in Table 4-1

such as technetium 99 and short-lived radionuclides listed in Table 4-2 such as strontium 90 and

cesium 13721.

The classification requirements of 10 CFR 61.55 for waste containing both long-lived and shortlived radionuclides. are as follows:

(1) If the concentration of a nuclide listed in Table 1 does not exceed 0.1 times the value listed

in Table 1, the class shall be that determined by the concentration of nuclides listed in

Table 2.

(2) If the concentration of a nuclide listed in Table I exceeds 0.1 times the value listed in Table

I but does not exceed the value in Table 1, the waste shall be Class C, provided the

concentration of nuclides listed in Table 2 does not exceed the value shown in Column 3 of

Table 2.

For mixtures of radionuclides, 10 CFR 61.55 specifies that the sum of fractions rule will be used

in determining waste classification. This rule entails dividing each radionuclide's concentration by

the appropriate limit, adding the resulting fractions, and comparing their sum to 1.0. A sum of

fractions less than 1.0 indicates compliance of the radionuclide mixture with the relevant

classification criteria.

As noted previously, DOE Guide 435.1-1 indicates that one criterion for determining key

radionuclides in waste is their importance in satisfying safety requirements comparable to the

performance objectives of 10 CFR Part 61, Subpart C for the waste disposal facility. These

performance objectives are described in Section 5.2.2 below.22

21 The approximate half-lives of these radionuclides are as follows: strontium 90, 28 years; cesium 137, 30 years;

and technetium 99, 212,000 years (BRH 1970).

22 In practice, meeting the waste acceptance criteria for the disposal facility ensures that the facility performance

objectives will be achieved. The rationale for this conclusion for a DOE LLW disposal facility such as the Nevada

National Security Site may be briefly summarized as follows:

" DOE performance objectives for its LLW disposal facilities are comparable with those of 10 CFR 61, Subpart

C;

* Disposal site performance in compliance with the performance objectives is determined by a performance

assessment of the facility and by a composite analysis that considers other radioactivity sources in the area

along with the radioactivity in the disposal site;

* These analyses are based on a projected total radionuclide inventory for the full, closed disposal site;

* This projected total inventory is based on the waste acceptance criteria, thus linking these criteria directly to

the calculated disposal site performance;

* The subject LLW stream (the Vitrification Melter) will meet the waste acceptance criteria; and

* Meeting the waste acceptance criteria will therefore ensure that the performance objectives will be

achieved, because waste meeting these criteria would not increase the assumed waste inventory used in the

performance assessment analyses.

These matters are addressed in more detail in Section 5.2. The link between waste acceptance criteria and disposal

site performance described in this footnote is similar for the commercial WCS LLW waste facility. Appendix C shows

that the State of Texas performance objectives in Texas Administrative Code, Title 30, Part 1, Chapter 336,

Subchapter H, Rules §336.723-727 mirror the NRC performance objectives.

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4.1.5 Radionuclides Important to the Disposal Site Performance Assessments

Because meeting the waste acceptance criteria for a given disposal facility ensures that the

facility performance objectives will be achieved, those radionuclides that are of particular

importance in the disposal site performance analyses are considered in identifying key

radionuclides. These radionuclides are Tc-99, Th-229, U-233, U-23423, and U-238 for the Nevada

National Security Site Area 5 waste disposal area (DOE 2010b). The results of the latest

performance assessments of the Nevada National Security Site LLW disposal areas are discussed in

Section 5.2 of this draft evaluation.

For the commercial WCS LLW disposal facility, the Radioactive Material License (TCEQ 2010)

identifies total radioactivity limits for three radionuclides for disposal in the Federal Facility Waste

Disposal Facility: C-14, Tc-99, and 1-129. These radionuclides (which also are included in Tables 1

and 2 of 10 CFR 61.55) contribute most to predicted dose according to the WCS performance

assessment submitted with the WCS license application and are therefore important to meeting the

performance objectives for the WCS facility. The WCS performance assessment is discussed in

Section 5.224. (The Texas Administrative Code in §336.723-727 sets forth performance objectives

for LLW disposal facilities, which track the NRC performance objectives in 10 CFR Part 61, Subpart

C, as further discussed in Section 5.2 and Appendix C of this draft WIR evaluation.)

4.1.6 Conclusions About Key Radionuclides in the Vitrification Melter

Based on consideration of the factors discussed above, DOE considers all radionuclides listed in

tables 1 and 2 of 10 CFR 61.55 to be key radionuclides for the purposes of this draft WIR

evaluation, with the caveat that some are of lesser importance due to their low concentrations in

the waste, their small dose conversion factors, or both. Table 4-3 shows these radionuclides.

Table 4-3. Key Radionuclides for this Draft Evaluation

10 CFR 61.55 Long- 10 CFR 61.55 Short- Radionuclides

Lived Radionuclides Lived Radionuclides Important to PA

H-3 X

C-14 X X(

Co-60 X

Ni-59 X

Ni-63 X

Sr-90 X

Nb-94 X

Tc-99 X X(l)(2)

1-129 X X(

Cs-137 X

Th-229 X(2)

23

U-234 present at the time of disposal is the predominant source of Pb-210 (DOE 2010b). Pb-210 was not identified

as a key radionuclide because its presence at the time of estimated maximum dose is due to U-234 in the disposed

of waste, rather than Pb-210 in the waste.

24 WCS is required to have a performance assessment maintenance plan and to update the performance assessment,

consistent with this plan, prior to receipt of waste and annually thereafter (License condition 87, TCEQ 2010).

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Table 4-3. Key Radionuclides for this Draft Evaluation (Continued)

10 CFR 61.55 Long- 10 CFR 61.55 Short- Radionuclides Radionuclide Lived Radionuclides Lived Radionuclides Important to PA

U-233 X(2)

U-234 X(2)

U-238 X(2)

Np-237(3 ) X

Pu-238(3 ) X

Pu-239(3) X

Pu-240(3) X

Pu-241 X

Pu-242(3) X

Am-241(3) X

Am-243 X

Cm-242 X

Cm-243(3) X

Cm-244(3) X

NOTES: (1) Radionuclides important to the performance assessment of the WCS Federal Facility Waste Disposal Facility

(WCS 2007).

(2) Radionuclides important to the performance assessment of the Area 5 Radioactive Waste Management Site.

(3) Alpha emitting transuranic radionuclides with half-life greater than five years (NRC 1982, Table 4.2).

Performance of the Area 5 Radioactive Waste Management Site was last evaluated at the end

of Fiscal Year 2009 as described in the 2009 Annual Summary Report for the Area 3 and Area 5

Radioactive Waste Management Sites at the Nevada Test Site Nye County, Nevada.: Review of the

Performance Assessments and Composite Analyses (NST 2010), using the estimated closure

inventory. This evaluation showed that:

* The maximum member of public all-pathways dose was predicted to occur at 1,000 years

for the resident farmer scenario. The resident farmer dose was predominately due to Tc-99

(78 percent) and Pb-210 (13 percent). Lead-210 present at 1,000 years would be

produced predominately by radioactive decay of U-234 present at the time of disposal.

(DOE 2010b)

* The maximum acute intruder dose was predicted to occur at 1,000 years for the shallow

land burial disposal units under the acute construction scenario. The acute intruder dose is

caused by multiple radionuclides including Th-229 (30 percent), U-238 (29 percent), U-233

(16 percent), and U-234 (10 percent). The maximum chronic intruder dose occurred at

1,000 years for the intruder-agriculture scenario. The intruder-agriculture dose is due

predominantly to Tc-99 (72 percent) and U-238 (14 percent). (DOE 2010b)

Performance of the WCS Federal Facility Waste Disposal Facility was evaluated by WCS in

connection with the WCS license application (WCS 2007). The three radionuclides limited in

quantity in the radioactive material license (TCEQ 2010) - C-14, Tc-99, and 1-129 - are the

radionuclides that contribute most significantly to estimated dose in the performance assessment.

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4.2 Removal to the Maximum Extent Technically and Economically Practical

Removal to the maximum extent "technically and economically practical" is not removal to the

extent "practicable" or theoretically "possible."25 Nor does the criterion connote removal which may

be notionally capable of being done.26 Rather, the adverbs "technically" and "economically" modify

and add important context to that which is contemplated by the criterion. Moreover, a "practical"

approach as specified in the criterion is one that is "adapted to actual conditions" (Fowler 1930);

"adapted or designed for actual use" (Random House 1997); "useful" (Random House 1997);

selected "mindful of the results, usefulness, advantages or disadvantages, etc., of [the] action or

procedure" (Random House 1997); fitted to "the needs of a particular situation in a helpful way"

(Cambridge 2004); "effective or suitable" (Cambridge 2004). Therefore, the evaluation as to

whether a particular key radionuclide has been or will be removed to the "maximum extent that is

technically and* economically practical" will vary from situation to situation, based not only on

reasonably available technologies but also on the overall costs and benefits of deploying a

technology with respect to a particular waste stream. The "maximum extent that is technically and

economically practical" standard contemplates, among other things: consideration of expert

judgment and opinion; environmental, health, timing, or other exigencies; the risks and benefits to

public health, safety, and the environment arising from further radionuclide removal as compared

with countervailing considerations that may ensue from not removing or delaying removal; life

cycle costs; net social value; the cost (monetary as well as environmental and human health and

safety costs) per curie removed; radiological removal efficiency; the point at which removal costs

increase significantly in relationship to removal efficiency; the service life of equipment; the

reasonable availability of proven technologies; the limitations of such technologies; the usefulness

of such technologies; and the sensibleness of using such technologies. What may be removal to the

maximum extent technically and economically practical in a particular situation or at one point in

time may not be that which is technically and economically practical, feasible, or sensible in another

situation or at a prior or later point in time. In this regard, it may not be technically and

economically practical to undertake further removal of certain radionuclides because further

removal is not sensible or useful in light of the overall benefit to human health and the

environment.27

4.2.1 Technical Practicality Assessment

This section identifies different methods to remove key radionuclides from the Vitrification

Melter that were considered before the Melter was shut down and describes the evaluations

performed for four representative removal methods to determine whether they would have been

25 In evaluating' whether key radionuclides have been removed to the maximum extent that is "technically and

economically practical", DOE has considered the guidance in DOE Guide 435.1-1 as well as the plain meaning of the

phrase "technically and economically practical." DOE's evaluation also reflects a risk-based approach, and is

consistent with the NRC Policy Statement concerning WVDP decommissioning criteria for waste to remain at the

WVDP (NRC 2002), NRC staff guidance for NRC consultation activities related to DOE waste determinations (NRC

2007), and the approach taken pursuant to the similar criterion in Section 3116(a) of the Ronald W. Reagan National

Defense Authorization Act for Fiscal Year 2005 (see e.g., Basis for Section 3116 Determination for the Idaho Nuclear

Technology and Engineering Center Tank Farm Facility (DOE 2006)). 26 Compare the meaning of practical with the meaning of practicable. As commonly understood, practicable refers to

something that can be put into effect. Practicalrefers to something that is also sensible and worthwhile. Thus, it

might be practicable to transport children to school by balloon, but it would not be practical. (Webster's 1984) 27

As a general matter, such a situation may arise if certain radionuclides are present in such extremely low quantities

that they make an insignificant contribution to potential dose to workers, the public, and the hypothetical human

intruder.

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technically practical. It also addresses the technical practicality of dismantling the Vitrification

Melter in its current condition.

As noted above, HLW vitrification was completed in September 2002. At this point, increasingly

dilute liquids were being retrieved from the underground storage tanks and the Vitrification Melter's

design life was nearing its end. Before shutting down the vitrification process - which would

eliminate the only available means of effectively managing additional removed radionuclides - DOE

evaluated various methods of removing key radionuclides and utilized two effective methods as

described below.

4.2.2 Methods Considered

DOE developed a vacuum extraction system using evacuated canisters to remove the majority

of residual dilute molten glass from the Vitrification Melter prior to shutdown. This system could

only be utilized before the contents of the Vitrification Melter were allowed to solidify after the

heaters were secured and the Melter allowed to cool.

The WVDP also considered other available, proven decontamination technologies that might be

suitable for removing key radionuclides from the Vitrification Melter. These included those

discussed in DOE's Decommissioning Handbook (DOE 1994). This handbook describes a wide range

of technologies used for decontamination at DOE sites. Some of these methods - such as

vacuuming, flushing with water, grinding, grit blasting, and milling - are widely used industrial

technologies. Others are innovative technologies developed with the support of the DOE Office of

Science as part of a continuing program to improve methods used in decontamination and

decommissioning work. Some innovative decontamination technologies have been tested in field

demonstration projects sponsored by DOE as described in innovative technology summary reports

issued by the DOE Office of Science and Technology, and these were considered. However, none

of these technologies was identified as specifically applicable to a vitrification melter.

DOE's Decommissioning Handbook identifies advantages and disadvantages of

decontamination technologies in various applications. The most relevant application for

decontaminating the Vitrification Melter internals was "Embedded Material and Some Oxide

Surfaces." The DOE Handbook considered five technologies to be highly effective in this

application: (1) hydroblasting, (2) ultra high pressure water grit blasting, (3) drill and spall, (4)

paving breaker/chipping hammer, and (5) expansive grout.

After considering available options, the WVDP determined that three fundamental approaches

existed for potential removal of additional key radionuclides from the Vitrification Melter: (1)

flushing, (2) mechanical means (including the five technologies the DOE Handbook identified as

high effective in the application just discussed), and (3) dismantlement.

The methods evaluated in detail were:

* Flushing - decontamination solutions;

• Mechanical - evacuated canisters, ball milling; and

* Dismantlement - size reduction and waste segregation.

The objective of each of these potential methods was to remove residual material in the

equipment, including key radionuclides, to the extent that was technically and economically

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practical. The remaining available options were considered to be experimental2 8 on such an

application as a vitrification melter. In 2008, DOE reconsidered use of a dismantlement approach

to determine whether using this approach with present-day technologies would be technically

practical as discussed in Section 4.2.6.

4.2.3 Flushing with Decontamination Solutions

A technically and economically practical method for removing key radionuclides from the

Vitrification Melter involved flushing the internals with decontamination solutions from other

vessels. This method was used successfully while the Vitrification Facility was still operational 29.

Prior to introducing the flush solutions, vitrified glass had a typical radioactivity concentration

of 3E+04 pCi/cm 3 Cs-137. Using a typical Vitrification Melter molten glass level of 26 inches, this

concentration corresponded to a volume of 1000 liters of glass in the Melter, or a total of

approximately 30,000 curies of Cs-137. After flushing, the remaining contents of the Vitrification

Melter were removed by the evacuated canisters (evacuated canisters are discussed below).

Because the remaining activity in the Vitrification Melter cavity was determined to be approximately

4,120 curies of Cs-137 after flushing and deployment of the evacuated canisters (WMG 2004a), a

total of approximately 25,900 curies of Cs-137 were determined to have been removed by these

efforts (30,000 curies - 4,120 curies).

Another measure of the flushing effectiveness was the reduced Cs-137 concentration in the

glass inside the Vitrification Melter. The samples of glass from the evacuated canisters showed an

average Cs-137 concentration of approximately 2E+03 pCi/g. Using an average density of 2.5

g/cm 3, this value corresponds to a concentration of 5E+03 pCi/cm 3

, well below the original

concentration of 3E+04 pCi/cm 3.

The evacuated canisters removed 8.9E+05 cm 3 of glass, which corresponds to a total of 4,500

curies of Cs-137. Therefore, the total number of curies of Cs-137 removed from the Vitrification

Melter by flushing with three rounds of decontamination solutions was 25,900 curies - 4,500 curies

= 21,400 curies, with 8,600 curies remaining in the Melter after flushing but before Evacuated

Canister System removal.

The following table summarizes the effectiveness of curie removal from the Vitrification Melter

using decontamination solution flushing.

TABLE 4-4. Vitrification Melter Flushing Effectiveness(l)

Curies

Flush Number Remaining Melter Melter Inventory Removed Per

Inventory in Curies Reduction Factor Melter Fls

Melter Flush

Starting Point 30,000.00 _

1 16,126.60 46.24% 13,873.34

2 8,668.91 71.10% 7,457.69

3 4,660.00(2) 84.47% 4,008.91

NOTEs: (1) From Purdue 2004, Table 11.

(2) The evacuated canister removed another 4480 curies from the Vitrification Melter cavity. This left the 4660

curies cited (4120 cauity plus 540 ex-cavity).

28 DOE considered it necessary to use a proven, mature technology to avoid unnecessary development costs and the

inherent uncertainties that would have been associated with an experimental technology.

29 Nitric acid and demineralized water were used in the flushes.

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While the effectiveness of the. flushes was determined by reductions in concentrations of Cs137, which makes up approximately 94 percent of the total activity, other key radionuclides were

removed from the Vitrification Melter with approximately the same efficiency as Cs-137. The HLW

slurry had been mixed in the Melter Feed Hold Tank before entering the Vitrification Melter, making

it relatively homogeneous. The homogeneity of the mixture was further increased in the

Vitrification Melter as the molten state was achieved. Since flushing removed molten glass, all

radionuclides in the mixture were removed in approximately equal proportions.

The economic practicality analysis discussed below evaluated the probable effectiveness of

further Vitrification Melter flushing beyond these three cycles while the Melter was still operational.

4.2.4 Evacuated Canisters

The WVDP developed the Evacuated Canister System in parallel with the overall Vitrification

Facility design to enable removal of the residual dilute molten glass expected to remain in the

Vitrification Melter prior to its shutdown and cooling. The Evacuated Canister System consisted of

two stainless steel canisters the same size as a standard HLW canister but equipped with a special

L-shaped "snorkel" assembly, both of which were designed to be positioned within a steel cage and

positioned over the Vitrification Melter. The residual glass was retrieved into the canisters under

vacuum after the plug within the snorkel melted from the heat of the Vitrification Melter after its

insertion.

Figure 4-1 shows the snorkel portion of the evacuated canister in the process of removing

molten glass from the Vitrification Melter. The metal has been heated by the molten glass to the

point where it produced the orange glow shown in the figure.

During deployment, approximately 88 percent of the residual molten glass volume was

removed using this system. This technology was technically practical only for removal of key

radionuclides from the Vitrification Melter while they were contained in the molten glass. Because

the Evacuated Canister System already existed and had been deployed before the time of this

analysis, no further economic practicality assessment of this technology was performed.

As with flushing, the Evacuated Canister System removed all key radionuclides from the

Vitrification Melter with approximately equal efficiency.

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Figure 4-1. Evacuated Canister Removing Residual

Glass from the Vitrification Melter

4.2.5 Mechanical Decontamination

This process would have used a combination of grinding and impact in a ball mill to separate

radioactive contamination from the base material. 30 This process was evaluated as representative

of decontamination by mechanical means.

This process was initially evaluated for potential use in decontamination of small pieces of

equipment and other materials contaminated with HLW during the vitrification process. In that

application, it was envisioned that equipment pieces would tumble against each other within a

closed system, possibly using zirconia as an abrasive to promote grinding action. This process

would dislodge the contamination, which would be returned to the vitrification process for

solidification as HLW. (WVNSCO 2001)

30 In a commercial ball mill, balls tumble around within a rotating cylinder breaking material that becomes

sandwiched between the balls and the cylinder walls. Such systems typically use media 1/2 inch or larger as the

grinding instruments. A laboratory jar mill is a small version of a ball mill.

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The WVDP evaluation of this process included small-scale tests performed using a laboratory

jar mill. Tests were performed on specimens consisting of Inconel and stainless steel plates with

glass annealed to their surfaces, short lengths of pipe containing dried slurry, and glass chunks.

These tests showed that the process could decontaminate metal and reduce the size of glass

pieces.

Disadvantages evident from the laboratory tests included production of fine glass powder,

which could cause problems in the process, and a propensity for embedding contamination in the

metal. These characteristics would have limited the effectiveness of the process.

DOE concluded that the process could be optimized, but that development for the Vitrification

Melter application would be costly and the results uncertain, resulting in another disadvantage

since DOE was seeking a proven, mature technology to avoid the uncertainties inherent in

unproven technologies. The WVDP determined that this process was not technically practical

because of such disadvantages. Given this conclusion, details of how to use the process on the

Vitrification Melter and how to dispose of the resulting waste stream were not developed.

4.2.6 Dismantlement

Another mechanical option considered involved dismantling the Vitrification Melter.

Dismantling the Vitrification Melter would involve cutting it up, removing and segregating residual

glass and refractory brick; characterizing and packaging the waste into disposal-ready containers,

storing the wastes in an interim onsite location, and eventually shipping the containers offsite for

disposal. Because of the radiation levels associated with the Vitrification Melter, such

dismantlement would need to be performed remotely in an area such as the Vitrification Cell.

Three disadvantages in this approach are evident:

It would- not remove key radionuclides, but simply segregate the residual glass into a

separate waste stream;

* Dismantlement of the Vitrification Melter would result in the pieces being placed in multiple

containers for disposal, rather than the intact Vitrification Melter being placed in a single

shipping and disposal container, thus requiring more handling and increased worker

radiation exposure; and

* The potential generation of orphan waste because it is unlikely that the separated residual

glass could be disposed of as LLW and it would not be in a form (e.g., encased in stainless

steel) suitable for disposal as HLW.

Despite these disadvantages and the practical difficulties that would be associated with

Vitrification Melter dismantlement, this process was considered in 2002 to be technically practical.

It would still be technically practical today, albeit with the same major disadvantages.

4.2.7 Other Potential Decontamination Methods

DOE considered whether other decontamination methods developed since 2002 could be

technically practical for removing key radionuclides from the Vitrification Melter in its present

condition with residual radioactivity in the hardened glass. Recent compilations of decontamination

methods such as those in The Decommissioning Handbook (ASME 2004) were considered, but DOE

identified no new methods that warranted evaluation for possible application to the Vitrification

Melter.

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4.2.8 Summary and Conclusions

In the technical practicality assessment, numerous methods that might have been used to

remove key radionuclides from the Vitrification Melter were considered and four representative

methods evaluated. Of these four methods, flushing with decontamination solutions, the Evacuated

Canister System, and Vitrification Melter dismantlement were determined to be technically

practical.

4.3 Economic Practicality Assessment

Economic practicality includes consideration of total lifecycle costs, the cost per curie removed,

the relationship between costs and removal of the key radionuclides, and the point in this

relationship at which removal costs increase significantly and thus become impractical (DOE Guide

435.1). In. this regard, removal of key radionuclides to the "niaximum extent . . . economically

practical" includes consideration of net social benefit, the remaining service life of equipment,

expert judgment, and whether the benefits to health and safety outweigh the disadvantages, that

is, whether further radionuclide removal would be useful and sensible in light of the overall benefit

to human health and safety.

The evaluation of the economic practicality of additional radionuclide removal from the

Vitrification Melter focused on:

(1) Performing additional flushing before shutdown of the Vitrification Melter, considering the

effectiveness of the flushing already performed; and

(2) Vitrification Melter dismantlement, in comparison with intact removal and shipment of the

intact Vitrification Melter for offsite disposal.

In each case, the evaluation compared the potential benefits in improved worker and public health

and safety (i.e., reduced worker and public risk from radiation exposure) with the expected

impacts.

This cost-benefit analysis was performed using methodology consistent with guidance issued

by the U.S. Office of Management and Budget (OMB 1992) and by NRC in NUREG/BR-0058,

Regulatory Analysis Guidelines of the US. Nuclear Regulatory Commission (NRC 1995) and the

companion handbook, NUREG/BR-0184, Regulatory Analysis Technical Evaluation Handbook, (NRC

1997). It is documented in a detailed report (Purdue 2004). The following summary is drawn from

that report 31 .

31 The Purdue report uses the term net social benefitfor the results of this analysis, with this term defined as

incremental gross benefits minus incremental gross costs. Gross benefits are measured by the collective radiation

dose and costs avoided for the particular approach being analyzed. Gross costs are the total monetary costs of

implementing the approach, which includes the monetized costs of estimated radiation doses to workers and the

public.

NRC in NUREG/BR-0058 (NRC 1995) uses the term value-impact analysis for this type of evaluation. Values are

defined as "the beneficial aspects anticipated from a proposed regulatory action such as, but not limited to, the (1)

enhancement of health and safety, (2) protection of the natural environment, (3) promotion of the efficient

functioning of the economy and private markets, and (4) elimination or reduction of discrimination or bias." Impacts

are defined as "the costs anticipated from a proposed regulatory action such as, but not limited to, the (1) direct

costs to the NRC and Agreement States in administering the proposed action and to licensees and others in

complying with the proposed action, (2) adverse effects on health, safety, and the natural environment, and (3)

adverse effects on the efficient functioning of the economy or private markets."

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4.3.1 Evaluation of Additional Flushing

The economic practicality assessment used a Monte Carlo simulation computer model to

calculate estimates of value in terms of expected incremental monetary benefits that represented

improvements to worker and public health and safety. These benefits were defined as the

monetized value of the reduction in collective person-rems combined with the costs avoided by the

approach considered.

The Estimated Effectiveness of Additional Flushing

For continued Vitrification Melter flushing, the assessment considered a series of 12 additional

flushes using the Concentrator Feed Makeup Tank and Melter Feed Hold Tank in series to feed the

Melter. The results of the three actual flushes performed were used as the base case for

comparison purposes.

Each hypothetical flushing round was defined as filling the vessels to their working volume by

water spraying and subsequent processing of the waste stream with glass formers through the

Vitrification Melter into HLW canisters. Table 4-5 shows the estimated effectiveness of the 12

additional flushes considered in the analysis as calculated by the vessel cleaning module of the

computer model.

TABLE 4-5. Estimated Effectiveness of Continued Vitrification Melter Flushing(1)

Remaining Melter Melter Inventory Curies Removed

Flush Number Inventory Reduction Factor Per Melter Flush(2 )

(Curies) (0

/%)

0 30,000.00 _

1 16,126.60 46.24 13,873.34

t•

< 2 8,668.91 71.10 7,457.69

3 4,660.00 84.47 4,008.91

1 4,637.97 84.54 22.03

2 4,541.60 84.86 96.37

3 4,435.83 85.21 105.77

4 4,346.08 85.51 85.75

5 4,273.48 85.76 72.60

0

6 4,214.82 85.95 58.66

7 4i166.91 86.11 47.91

8 4,127.24 86.24 39.67

0-9 4,093.95 86.35 33.29

10 4,065.67 86.45 28.28

11 4,041.37 86.53 24.30

12 4,020.28 86.66 21.09

NOTE: (1) From Table 11 of Purdue 2004.

(2) Some Vitrification Melter inventory variations are related to the flushing effectiveness of upstream

vessels (the Concentrator Feed Makeup Tank and the Melter Feed Hold Tank).

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Table 4-5 shows that the amount of radioactivity removed per flush drops rapidly with

succeeding flushes to the point where additional flushing removes very little additional residual

radioactivity. This pattern is consistent with typical industry experience with repeated flushing of a

radioactively contaminated system.

For perspective, the first three hypothetical additional flushes would have removed a combined

total of approximately 224 additional curies of Cs-137, less than the reduction from radioactive

decay between October 2004 and August 2010, which amounted to approximately• 540 curies. The

first hypothetical additional flush would have removed only approximately 22 curies.

Another factor that was taken into account in considering hypothetical additional flushes was

the limited 'Vitrification Melter service life. Analysis indicated a 35 percent probability of Melter

failure occurring within six months of additional operation (six additional hypothetical flushes) and

a 98 percent probability of Melter failure in 12 months of additional hypothetical operation. (Purdue

2004)

The Estimated Costs of Additional Flushing

• Monetary costs for continued flushing were also evaluated on a per flush basis and on a

cumulative basis. The following graph depicts the results of this cost comparison.

16

(-

oa

(-

*0

U)

0

0

0

50,000,00

45,000.00

40,000.00

35,000.00

30,000.00

25,000.00

20,000.00

15,000.00

10,000.00

5,000.00

0.00

* Curies remaining In Melter

* Curies removed by flush

* Cumulative cost of additional flushes in thousands of $

0 1 2 3 1 2 3 4 5 6 7 8 9 10 11 12

Actual Flushes Hypothetical Additional Flushes

Figure 4-2. Effectiveness of Actual Flushes and Hypothetical Additional Flushes

Figure 4-2 shows that the three flushes performed proved to be effective in removing key

radionuclides to the maximum extent economically practical. It also shows that additional flushes

would have produced minimal results with costs that would have far outweighed the benefits.

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The costs of the hypothetical additional flushes reflect the cumulative cost of operating the

Vitrification Facility. These estimated costs account for such factors as the projected duration of

each flush cycle (one month), equipment repairs and maintenance, and the estimated cost of

additional canisters to contain the removed material. Estimated radiation exposure to workers and

the public was also taken into account and valued at $2000 per person-rem consistent with NRC

guidance in NUREG/BR-0058 (NRC 1995).

The analysis results can also be considered in terms of the costs per curie of Cs-137 removed

by the additional flushes. According to the model, the first additional flush would have cost

approximately $1,900,000 and removed approximately 22 curies for approximately $86,000 per

curie. All 12 .additional flushes would have cost approximately $47,500,000 and would have

removed a total of approximately 636 curies for approximately $74,000 per curie. The minimal

benefit of the estimated 636 curie radioactivity reduction would have been greatly overshadowed

by the costs involved, which would have included increased radiation exposure and resulting risk to

workers and to the public as well as direct monetary costs. (Purdue 2004)

Conclusions

The economic practicality analysis therefore concluded with reasonable certainty that further

flushing of the Vitrification Melter and vitrification of resulting waste would have imposed significant

costs without producing any meaningful benefits in improved worker-or public health and safety.

The practical limits, of this technology had been reached with the three flushes that had been

performed.

4.3.2 Evaluation of Vitrification Melter Dismantlement

Vitrification Melter dismantlement was evaluated in comparison to the costs associated with

intact removal of the Melter, which was considered to be the base case in the analysis.

Specifically, the cost of intact removal and shipment of the Vitrification Melter for disposal was

compared to the costs for dismantlement with respect to the sum of the following:

* Process labor to remotely dismantle and size-reduce the Vitrification Melter;

• Process labor required to put either the intact Vitrification Melter or its dismantled pieces

into disposal-ready containers, working remotely, and either ship the Melter for disposal or

place it in interim storage,

* Transportation costs to interim off-site storage or final disposal, and

* Disposal costs.

The following table summarizes those costs and illustrates cost savings from shipping the

Vitrification Melter intact versus dismantling the Vitrification Melter.

Table 4-6. Vitrification Melter Dismantlement Cost Analysis(l)

Cost Intact Melter Dismantled Melter Cost Savings (Intact

Category Option ($K) Option ($K) minus Dismantle) ($K)

Process Labor 354 3,368 -3,032

Transportation 3,258 1,692 1,566

Disposal 29 32 -3

TOTAL 3,641 5,110 -1,468

NOTE: (1) From Purdue 2004, Table 13.

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The process labor costs for Vitrification Melter dismantlement are much larger than for leaving

the Melter intact. It should be noted that this cost comparison assumed that the dismantled

Vitrification Melter pieces could be disposed of as LLW. However, as noted above, it is likely that

some Vitrification Melter components such as residual glass would be considered orphan waste

and, as such, have no identified pathway for disposal. This cost comparison does not consider the

implications of continued indefinite storage of orphan waste, another factor that adds conservatism

to the results since the potential costs of evaluation of orphan waste, its onsite storage, and its

ultimate disposal are not taken into account.

The analysis described above was completed in 2004. DOE considered whether improvements

in technologies since that time would change the assessment results, with the Vitrification Melter

now located in its shipping container and the low-density cellular concrete not yet poured into the

Melter or the container. The conclusions were: (1) the Vitrification Melter would still have to be

dismantled remotely in a shielded cell because of its radiation levels, which, considering radioactive

decay, are about 90 percent of the levels in 2004, and (2) no improved cutting or dismantlement

technologies developed since 2002 would make a significant difference in this process.

The estimates in Table 4-6 for Vitrification Melter dismantlement remain valid today with one

exception: There would be additional costs associated with the Vitrification Melter dismantlement

work to decontaminate the shielded cell where the work was done. Shielded cells that might be

used for this purpose have been decontaminated since 2002. The work would result in spreading

radioactive contamination in the work area, which would have to be cleaned up with the resulting

additional radioactive waste placed in containers and disposed of offsite.

4.3.3 Summary and Conclusions

The technical practicality assessment showed that flushing the Vitrification Melter with

decontamination solutions, removing molten residual glass using the Evacuated Canister System,

and dismantling the Vitrification Melter were the technically practical options for removing key

radionuclides from the Melter to the extent practical. Economic practicality analysis determined that

flushing was economically practical for the first three flushes but that continued flushing of the

Vitrification Melter was not practical because of high cost and minimal curies removed. In addition,

the probability of Melter failure increased with additional flushing. Dismantlement was considered

economically impractical due to the high cost associated with dismantlement and uncertainties

associated with management of the orphan waste that might be generated.

Estimated radiation doses to workers and the public were accounted for in the analyses. The

hypothetical additional flushes and dismantlement of the Vitrification Melter would have resulted in

higher radiation exposure (valued at $2000 per person-rem, as noted previously) and the resulting

risks both to workers and the public, so neither of these methods would have been beneficial from

a health and safety standpoint. (Purdue 2004)

In summary, both flushing and the Evacuated Canister System were successfully deployed and

removed key radionuclides to the maximum extent that is technically and economically practical.

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5.0 THE WASTE WILL BE MANAGED TO MEET SAFETY REQUIREMENTS

COMPARABLE TO THE PERFORMANCE OBJECTIVES OF 10 CFR 61, SUBPART C

Section Purpose

The purpose of this section is to evaluate whether the Vitrification Melter waste

package will be managed to meet safety requirements comparable to the

performance objectives of 10 CFR 61, Subpart C for disposal of low-level

radioactive waste.

Section Contents

This section addresses whether the Vitrification Melter waste package will meet

safety requirements comparable to the performance objectives of 10 CFR 61,

Subpart C for disposal of low-level radioactive waste and explains how the

Vitrification Melter waste package will meet criteria for disposal as low-level.

radioactive waste.

Key Points

" Management of the Vitrification Melter waste package will meet safety

requirements comparable to the performance objectives of 10 CFR 61,

Subpart C.

" The Vitrification Melter waste package will meet the waste acceptance criteria

for the Area 5 Radioactive Waste Management Site at the Nevada National

Security Site.

" The performance objectives in the Texas Administrative Code applicable to

the commercial WCS low-level waste disposal facility for Federal waste

mirror the performance objectives in 10 CFR 61, Subpart C and the facility

must be operated to provide reasonable assurance that those performance

objectives will be met; consequently, disposal of the Vitrification Melter waste

package at the WSC Federal waste disposal facility will meet safety

requirements comparable to the performance objectives of 10 CFR 61,

Subpart C.

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5.1 Introduction

Section 2 describes the design of the Vitrification Melter and its operational history. Section 2

also explains how the Vitrification Melter was characterized for residual radioactivity, with Table 2-2

providing the total activity estimates in the final waste form. As noted previously, the Vitrification

Melter has been placed in a custom made shipping container and the void spaces in the Vitrification

Melter and the shipping container will be filled with low-density cellular concrete in preparation for

transport to an offsite disposal facility.

Because the packaged Vitrification Melter will not exceed concentration limits for Class C LLW

as explained in Section 6, the second criterion of section II.B.2(a) of DOE Manual 435.1-1 applies

to this evaluation:

"[The waste] will be managed to meet safety requirements comparable to the performance

objectives set out in 10 CFR Part 61, Subpart C, Performance Objectives."

This section addresses this criterion as follows:

Section 5.2 begins by summarizing key DOE safety requirements related to disposal of LLW.

Section 5.2.1 describes DOE's general safety requirement and compares it with the similar

general safety requirements promulgated by NRC and the State of Texas;

Section 5.2.2 provides the following information regarding requirements for the protection of

the general population from releases of radioactivity:

* A description of the DOE requirements,

* A comparison between these requirements and the similar requirements promulgated

by NRC and the State of Texas,

* A summary of the results of the most recent performance assessment for Nevada

National Security Site Area 5 facility related to protection of the general population,.

* An estimate of the impact on the general population of disposal of the Vitrification

Melter in the Nevada National Security Site Area 5 facility, and

* A summary of the results of a performance assessment of the WCS Federal Facility

Waste Disposal Facility related to protection of the general population.

Section 5.2.3 provides similar information related to protection of individuals from inadvertent

intrusion into the closed LLW disposal facilities.

Section 5.2.4 discusses protection of individuals during operations at the WVDP, at the

Nevada National Security Site, and at the WCSLLW disposal facility.

Section 5.2.5 compares DOE requirements for stability of the disposal site after closure with

the requirements of the NRC and the State of Texas and briefly discusses preliminary closure

plans for the Nevada National Security Site Area 5 facility and the WCS facility.

Section 5.3 begins with a discussion of DOE waste acceptance criteria.

Section 5.3.1 discusses waste acceptance criteria for the Nevada National Security Site.

Section 5.3.2 summarizes how it is determined that a waste package meets the Nevada

National Security Site waste acceptance criteria.

Section 5.3.3 demonstrates that the Vitrification Melter Waste package meets the Nevada

National Security Site waste acceptance criteria.

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Section 5.4 discusses waste acceptance criteria for the WCS Federal Facility Waste Disposal

Facility and how it would be established that the Vitrification Melter Waste package meets

these criteria if the waste package were to be shipped to that facility for disposal.

5.2 DOE Safety Requirements

DOE has established requirements for management of radioactive waste to ensure protection

of workers, the public, and the environment, and complies with applicable Federal, State, and local

laws and regulations (DOE Order 435.1). DOE has also established specific requirements for its

radioactive waste disposal facilities, including the Area 5 Radioactive Waste Management Site at

the Nevada National Security Site. These requirements include:

(1) Performance objectives set forth in Chapter IV of DOE Manual 435.1-1, which include

maximum dose limits;

(2) DOE regulations at 10 CFR Part 835 and DOE Order 5400.5, cross referenced in Chapters I

and IV of DOE Manual 435.1-1;

(3) Waste acceptance requirements, which, among other things, establish limits on

radionuclides that may be disposed of based on a performance assessment of the facility;

(4) A performance assessment (and updates) of the facility, to provide reasonable expectation

that DOE's performance objectives will not be exceeded;

(5) A composite analysis that considers other radioactivity sources in the area as well as the

disposal facility;

(6) A performance assessment and composite analysis maintenance plan,

(7) A preliminary closure plan; and

(8) A monitoring plan (DOE Manual 435.1-1).

For wastes to be disposed of at DOE facilities, DOE establishes waste acceptance criteria,

based upon an independently reviewed and accepted LLW performance assessment, which also

includes provisions for maintenance and updating. Acceptability of the LLW performance

assessment is verified against the performance objectives of Section IV.P of DOE Manual 435.1-

1, as well as other requirements in DOE Manual 435.1-1, through an independent review process.

This review serves as the basis for DOE to issue a Disposal Authorization Statement, which

specifies any additional conditions that the site may need to impose to ensure the performance

objectives of DOE Manual 435.1-1, IV.P are met.

Figure 5-1 illustrates the general process used to provide reasonable expectation that disposal

site performance objectives are achieved, which is in addition to the use of formal waste

acceptance requirements.

The following subsections address the specific DOE performance objectives, and relevant DOE

regulations and Orders, for DOE LLW disposal sites. These performance objectives, regulations,

and Orders are set forth or cross referenced in DOE Manual 435.1-1, and provide safety

requirements comparable to the NRC performance objectives of 10 CFR 61, Subpart C, as

discussed further in Appendix C. As explained in Appendix C, the performance objectives in the

State of Texas regulations mirror the NRC performance objectives at 10 CFR 61, Subpart C, i.e.,

they are essentially identical except for the use of difference section numbers.

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SPerformance Assessment (PA)• Composite Analysis (CA)"".

Assures that disposal of LLW will II|Assesses impact to a, hypothetical member of

meet DOE performance objectives j] l |the public from all radioactive sources that I

.. . . .. . . . .. • • • L m ay inte ra ct w ith LLW d ispo sa l' fa cil'itie s . . .

Disposal Authorization Statement

Key elements:

* DOE-Headquarters review and approval of PAs and CAs

* Satisfactory site response to all approval conditions

_ Monitorinq, Maintenance, and Closure Plans

PA and CA Maintenance

Key elements:

* Annual review of disposal activities

* Research and testing

* Unreviewed disposal question evaluations

e Special Analyses

* Revisions as necessary

Figure 5-1. General Process Used to Ensure Performance Objectives Are Achieved

5.2.1 General Safety Requirement

The general requirement in DOE Manual 435.1-1, Section IV.P(1), is expressed as follows:

"Low-level waste disposal facilities shall be sited, designed, operated, maintained, and closed

so that a reasonable expectation exists that the following performance objectives will be met

for waste disposed of after September 26, 1988."

As discussed further in Appendix C, the general requirement in NRC's performance objectives

for licensed LLW disposal facilities at 10 CFR 61.40 sets forth a nearly identical, comparable

requirement:

"Land disposal facilities must be sited, designed, operated, closed, and controlled after closure

so that reasonable assurance exists that exposures to humans are within the limits established

in the performance objectives in Sections 61.41 through 61.44."

The general safety requirement of the State of Texas echoes the NRC general safety requirement

as shown in Appendix C.

The four relevant DOE performance objectives are addressed in Subsections 5.2.2 through 5.2.5.

5.2.2 Protection of the General Population from Releases of Radioactivity

DOE requirements in DOE Manual 435.1-1, Section IV.P(1), read as follows:

"(a) Dose to representative members of the public shall not exceed 25 millirem (0.25 mSv) in a

year total effective dose equivalent from all exposure pathways, excluding the dose from

radon and its progeny in air.

(b) Dose to representative members of the public via the air pathway shall not exceed 10

millirem (0.10 mSv) in a year total effective dose equivalent, excluding the dose from

radon and its progeny.

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(c) Release of radon shall be less than an average flux of 20 pCi/m 2

/s (0.74 Bq/m 2/s) at the

surface of the disposal facility. Alternatively, a limit of 0.5 pCi/L (0.0185 Bq/L) of air may

be applied at the boundary of the facility."

As discussed further in Appendix D, DOE's dose limits are comparable to those in the NRC

performance objectives at 10 CFR 61.41, although DOE uses more current radiation protection

methodology32. The NRC performance objective at 10 CFR 61.41 provides:

"Concentrations of radioactive material which may be released to the general environment in

ground water, surface water, air, soil, plants, or animals must not result in an annual dose

exceeding an equivalent of 25 millirems to the whole body, 75 millirems to the thyroid, and 25

millirems to any other organ of any member of the public. Reasonable effort should be made to

maintain releases of radioactivity in effluents to the general environment as low as is

reasonably achievable."

The requirement of the State of Texas for protection of the general population is virtually identical

to the NRC requirement as shown in Appendix C.

Assessment of Area 5 Performance

The report of the most recent annual review of performance assessments and composite

analyses for the Area 3 and Area 5 waste disposal facilities at the Nevada National Security Site

was issued in 2010 (NST 2010). This report addresses matters such as new or revised waste

streams, monitoring results, research and development, the inventory estimates at planned

closure, updated performance assessment results, and updated composite analysis results. It also

33 identifies special analyses that were performed in the previous year

As explained in the report of the annual review (NST 2010), the updated Area 5 performance

assessment provides reasonable expectation that DOE's performance objectives will be achieved.

This report summarizes the results of probabilistic analyses for Area 5 for the following scenarios:

32 NRC recommends in NUREG-1573, A Performance Assessment Methodology for Low-Level Radioactive Waste

Disposa/ Facidites, Recommendations of NRC's Performance Assessment Working Group (NRC 2000), use of the more

current radiation protection methodology, making NRC standards comparable to those of DOE in this regard.

33 A LLW disposal facility performance assessment involves detailed analyses of potential radiation doses to those

who may be affected in future years to ensure that the closed facility will meet its performance objectives. These

performance objectives include dose limits for a member of the public and for a hypothetical person who, unaware of

the buried radioactivity, might drill a well into the buried waste, referred to as the post-drilling scenario, or establish

a farm on the site, known as the intruder-agriculture scenario. A LLW disposal facility performance assessment

makes use of two basic models.

A conceptual model describes all of the relevant properties of the disposal site. Area 5 is scheduled for closure in

2028. The estimated radionuclide inventory at closure is made up of two components: the known activity in the

buried waste and the projected activity in waste to be disposed of in the future, which is based primarily on the

waste acceptance criteria and the types and amounts of radioactivity in the waste already disposed on in the facility.

The closure date and the estimated radionuclide inventory at closure are two examples of the many elements which

make up the conceptual model

A mathematical model is used with the conceptual model to calculate potential doses under different scenarios.

The Nevada National Security Site uses the GoldSim mathematical model, a widely-used software package that

simulates the future behavior of the closed disposal site in a probabilistic manner, providing a range of resultsý with

different probabilities. The Nevada National Security Site typically expresses key performance assessment results as

mean values and 95th percentile values.

Special analyses and composite analyses use similar methodologies, with the focus on the waste stream of

interest and all relevant radioactivity sources at the site, respectively. Special analyses are performed for waste

streams with a sum of fractions greater than one or where preliminary screening indicates that disposal of a new

waste stream has a potential to alter performance assessment assumptions or conceptual models. A composite

analysis is required for all DOE sites that manage radioactive waste; these analyses are updated annually.

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* All pathways dose for members of the public, with the predicted peak annual dose

estimated to occur at 1000 years after planned closure (i.e., in 3028, the end of the

compliance period) for the resident farmer scenario;

* The air pathway dose for members of the public, with the predicted peak annual dose to a

resident farmer at 1000 years after facility closure; and

* The Radon 222 flux density at the surface of the disposal units, which is predicted to reach

a peak 1000 years after facility closure.

This report shows that the predicted potential doses to representative members of the public to be

much less than the performance objective dose limits.

Estimated Impact of the Vitrification Melter

Even though the Vitrification Melter waste package meets the waste acceptance criteria for the

Nevada National Security Site Area 5 Radioactive Waste Management Site, a dose assessment was

performed to quantify the potential dose impact of disposal of this equipment in Area 5 for

perspective and additional information (DOE 2010b)3 4. The impact was assessed by evaluating the

Area 5 shallow land burial inventory disposed of though Fiscal Year 2009 with and without the

Vitrification Melter Inventory. 35 Table 5-1 shows the results for members of the public. The limiting

pathways/scenarios are those among the various combinations of pathways and scenarios analyzed

that are estimated to produce the highest annual all-pathways dose to representative members of

the public.

Table 5-1. Estimated Maximum Dose Impacts to a Representative Member of the Public

Associated With Vitrification Melter Disposal (mrem/y)(1)

Estimated Maximum Annual Dose

Limiting Pathway/Scenario (Probability) Limit Area 5 Area 5 Difference

(Without (With Due to

Melter) Melter) Melter

Member of Public: All Pathways/Resident Farmer (Mean) 25 1.4 1.4 negligible

Member of Public: All Pathways/ Resident Farmer 25 4.5 4.5 negligible

(95P Percentile)

NOTES: (1) From DOE 2010b.

The information in Table 5-1 shows that disposal of the Vitrification Melter in the Area 5 facility

is estimated to have no significant impact on facility performance insofar as members of the public

are concerned..

The Waste Profile for the Vitrification Melter, which describes the characteristics of the waste

as required by the waste acceptance criteria for the Nevada National Security Site Area 5

Radioactive Waste Management Site, was submitted by the WVDP in May 2010 (WVES 2010). The

Waste Profile was reviewed by the Nevada National Security Site Waste Acceptance Review Panel,

34 This dose assessment was performed using the same methodology as a special analysis even though no special

analysis was required for the Vitrification Melter. A special analysis was not required for the Vitrification Melter waste

package even though the waste stream sum of fractions exceeded one because the screening evaluation indicated

that the available capacity sum of fractions (the fraction of the available capacity that the entire waste stream

represents) was less than the criterion that would trigger a special analysis. 31 These analyses were performed without accounting for the projected future inventory, as noted above, the normal

practice used for special analyses. As such, they produce slightly different results than the performance assessments

reported in the 2009 annual review (NST 2010) because the performance assessments reported in the annual review

account for projected future inventory.

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a group of waste management specialists who review new and revised waste streams planned for

disposal in the Area 5 Radioactive Waste Management Site. The panel's review resulted in formal

acceptance of the Vitrification Melter waste package for disposal, conditional upon it meeting WIR

criteria in accordance with DOE Order 435.1 [that is, this draft WIR evaluation resulting in a

determination that it can be managed as LLW] (DOE 2010a). Section 5.3 below discusses the waste

acceptance process in more detail.

Assessment of WCS Federal Facility Waste Disposal Facility Performance

Unlike the Nevada National Security Site Area 5 LLW disposal facility, which has been in

operation for decades, the WCS Federal Facility Waste Disposal Facility is not yet in operation.

However, WCS included a performance assessment with its license application (WCS 2007). The

performance assessment of the Federal Facility Waste Disposal Facility considered a total inventory

at closure of 57 million cubic feet of waste with 16.4 million curies of radioactivity 36

. It included the

following estimated dose for the post-institutional control period: a maximum annual dose of 3.4

millirem per year to an adjacent resident, weli under the limit of 25 millirem per year37 . In addition,

a license condition requires that WCS prepare an updated performance assessment prior to

accepting waste for disposal and annually thereafter to demonstrate that performance objectives

will be met (TCEQ 2010).38

5.2.3 Protection of Individuals from Inadvertent Intrusion

DOE requirements of DOE Manual 435.1-1, Section IV.P(2)(h), for protection of individuals

from inadvertent intrusion read as follows:

"For purposes of establishing limits on the concentration of radionuclides that may be disposed

of near-surface, the performance assessment shall include an assessment of impacts calculated

for a hypothetical person assumed to inadvertently intrude for a temporary period into the lowlevel waste disposal facility. For intruder analyses, institutional controls shall be assumed to be

effective in deterring intrusion for at least 100 years following closure. The intruder analyses

shall use performance measures for chronic and acute exposure scenarios, respectively, of 100

millirem (1 mSv) in a year and 500 millirem (5 mSv) total effective dose equivalent excluding

radon in air."

As discussed in Appendix C, NRC in 10 CFR 61.42 sets forth the following requirements:

"Design, operation, and closure of the land disposal facility must ensure protection of any

individual inadvertently intruding into the disposal site and occupying the site or contacting the

waste at any time after active institutional controls over the disposal site are removed."

As discussed further in Appendix C, DOE's dose limits for the hypothetical human intruder are

more stringent that the dose limit used for NRC's performance objective at 10 CFR 61.42. Pursuant

36 The radioactive material license provides for a total LLW volume of 26,000,000 cubic feet with 5,600,000 curies of

total activity; the limits for Class A containerized, Class B, and Class C waste are 8,100,000 cubic feet and 5,500,000

curies of total activity (TCEQ 2010). As shown in Table 5-4 below, the Vitrification Melter waste package would

constitute small fractions of these limits. The use of the higher total inventory - 57,000,000 cubic feet and

16,400,000 curies - in the performance assessment is conservative compared to the lower license limits.

37 This estimate was for the Federal Facility Waste Disposal Facility Non-Canister Disposal Unit. The estimated

maximum annualdose for the Canister Disposal Unit where the Vitrification Melter waste package would be disposed

of was 1.1 millirem per year due to Tc-99 (95 percent) and 1-129 (5 percent) occurring at year 36,400. (WCS

calculated doses to 100,000 years). (WCS 2007, Volume 31, Appendix 8.0-6, Table 8.0-6.10-2) 38 Because the WCS facility is not yet in operation, this draft waste incidental to reprocessing evaluation does not

quantitatively estimate the potential additional dose impact from disposal of the Vitrification Melter in the WCS facility

in the same manner as was done for the Nevada National Security Site Area 5 Radioactive Waste Management Site

(i.e., by evaluating the inventory to be disposed of, with and without the Vitrification Melter.)

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to NRC guidance, NRC licensees for LLW disposal facilities should demonstrate that the annual dose

to the inadvertent human intruder will not exceed 500 mrem per year after removal of institutional

control, whereas DOE imposes a 100 mrem per year and 500 mrem per year total effective dose

equivalent (excluding radon in air) for chronic and acute inadvertent human intruder exposures,

respectively.

The State of Texas requirement for protection of individuals from inadvertent intrusion tracks

the NRC requirement as shown in Appendix C.

Assessment of Area 5 Performance

The report of the most recent (2009) annual review of performance assessments and

composite analyses for the Area 3 and Area 5 radioactive waste management sites at the Nevada

National Security Site (NST 2010) demonstrates that there is a reasonable expectation that the

Area 5 Radioactive Waste Management Site will meet the DOE intruder dose criteria. The scenarios

evaluated as described in this report were as follows:

" The drilling worker intruder scenario, with the predicted peak annual acute dose 1000

years after facility closure, the end of the compliance period; and

• The home construction intruder scenario, with the predicted peak annual acute dose 1000

years after facility closure.

Chronic intruder scenarios are no longer reported in the Annual Summary Report for the Area 5

Radioactive Waste Management Site becausechronic intrusion would be unlikely due to a change

in the institutional control policy made in 2008. The planned land-use restrictions will prohibit public

access to groundwater for 1,000 years within the compliance boundary negotiated with the State of

Nevada, which is to include the Area 5 Radioactive Waste Management Site.39 (NST 2010)

Estimated Impact of the Vitrification Melter

Table 5-2 shows the estimated impacts of disposal of the Vitrification Melter in the Nevada

National Security Site Area 5 Radioactive Waste Management Site to a hypothetical human

intruder.

Table 5-2. Estimated Acute Dose Impacts to an Inadvertent Intruder Associated With

Vitrification Melter Disposal (mrem/y)(1)

Estimated Maximum

Annual Limit

Limiting Pathway/Scenario (Probability) (Acute Area 5 Area 5 Difference

Exposure) (Without (With Due to

Melter) Melter) Melter

Inadvertent Intruder: Construction (mean)(2) 500 280 280 negligible

Inadvertent Intruder: Construction (95t Percentile) 500 600 600 negligible

NOTES: (1) From DOE 2010b. The limiting pathways/scenarios are those estimated to produce the highest dose. In each

case, the peak annual doses were estimated to occur 1000 years after facility closure.

(2) The mean estimate is used for DOE Manual 435.1-1 compliance purposes4o. The 95t" percentile estimate

provides additional information.

39 Chronic intruder doses continue to be calculated by the performance assessment model but are no longer reported

in the Annual Summary Report for reasons specified in that report (DOE 2010b). This practice is consistent with

Section IV.P(2) of DOE Manual 435.1-1, which provides for considering the likelihood of inadvertent intruder

scenarios in interpreting the results of the analyses if adequate justification is provided. 40

The use of the mean also is consistent with NRC guidance in NUREG-1573, A Performance Assessment

Methodology for Low-Level Radioactive Waste Disposal, Recommendations of NRC's Performance Assessment

Working Group (NRC 2000).

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The information in Table 5-2 shows that disposal of the Vitrification Melter in the Area 5 facility

is estimated to have no significant impact on facility performance insofar as inadvertent intruders

are concerned.

Assessment of WCS Federal Facility Waste Disposal Facility Performance

The performance assessment included with the WCS license application (WCS 2007) provides

the following estimated doses to inadvertent intruders:

* A maximum acute dose of 6.7 millirem per year to the intruder driller, and

* A maximum chronic dose of 4.6 millirem per year to the intruder resident farmer.

These estimated doses are well below the 500 millirem annual limit41.

5.2.4 Protection of Individuals During Operations

The DOE requirements in DOE Manual 435.1-1, Section I.E(13), for protection of individual

during operations read as follows:

"Radioactive waste management facilities, operations, and activities shall meet the

requirements of 10 CFR Part 835, Occupational Radiation Protection, and DOE 5400.5,

Radiation Protection of the Public and the Environment"

As discussed in Appendix C, NRC in 10 CFR 61.43 provides similar, comparable requirements:

"Operations at the land disposal facility must be conducted in compliance with the

standards for radiation protection set out in Part 20 of this chapter, except for releases of

radioactivity in effluents from the land disposal, facility, which shall be governed by Section

61.41 of this part. Every reasonable effort shall be made to maintain radiation exposures as

low as is reasonably achievable."

The State of Texas requirements track the NRC requirements, as discussed in Appendix D.

Comparability of DOE, NRC, and State of Texas Requirements

DOE's requirements and dose limits for protection of individuals during operations in 10 CFR

Part 835 and DOE Order 5400.5, cross referenced in Section I.E(13) of DOE Manual 435.1-1, are

comparable to the relevant NRC standards for radiation protection in 10 CFR Part 20, as cross

referenced in the NRC performance objective at 10 CFR 61.43. For example, both DOE and NRC

limit occupational dose to a total effective dose equivalent of 5 rem per year and doses to the

public from operations to 0.1 rem per. year, as further discussed in Appendices C and D. As

explained in Appendix D, the State of Texas requirements and dose limits for protection of

individuals during operations mirror the NRC requirements and dose limits.

DOE's regulatory and contract requirements for DOE facilities and activities ensure compliance

with DOE's regulations at 10 CFR Part 835 and relevant DOE Orders that establish dose limits for

the public and the workers during operations. Appendix D provides additional details concerning

the DOE requirements for protecting individuals during operations.

in addition, DOE's regulation at 10 CFR 835.101(c) requires that each radiation protection

program include formal plans and measures for applying the ALARA (as low as is reasonably

41 These estimates were for the Commercial Waste Facility. The estimated maximum annual dose for the Federal

Facility Waste Disposal Facility Canister Disposal Unit where the Vitrification Melter waste package would be disposed

of was 1.8 millirem per year for the intruder driller and 0.79 millirem per year for the intruder resident (WCS 2007,

Volume 31, Appendix 8.0-6, Table 8.0-6.13-2)

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achievable) approach to occupational exposures. Similarly, DOE's policy statement (DOE Policy

441.1) expresses the Department's policy to ensure that radiation exposure to workers and the

public, as well as releases of radioactivity to the environment, are maintained below regulatory

limits, and that deliberate efforts are taken to further reduce exposures and releases to as low as

reasonably achievable.

Protection of Individuals During Operations at the WVDP and the Nevada National

Security Site

The DOE requirements apply to the workers at the WVDP who will be involved with grouting

and preparing the Vitrification Melter and its waste package for disposal, as well as to the public at

the site. The DOE performance requirements also apply to the workers at the Nevada National

Security Site who would handle disposal of the Vitrification Melter waste package and to the public

at that site.

Both the WVDP and the Nevada National Security Site maintain radiation protection programs

based on the requirements of 10 CFR 835. These programs also comply with various DOE

directives (including DOE Order 5400.5, other Orders, policies, guides, and manuals), and

supplemental technical standards.

The WVDP radiological protection program and these measures are described in the WVDP

Radiological Controls Manual (WVNSCO 2006). The Nevada National Security Site radiological

protection program and ALARA measures are described in the Nevada National Security Site

Radiological Control Manual (NST 2009b)

Gamma radiation levels on the sides of the Vitrification Melter waste package range from

approximately 2 to 6 millirem per hour (WVNSCO 2004); levels will be lower after the waste

package is grouted. Workers involved with handling of the waste package have received doses

below the WVDP administrative control level of 500 mrem per year, which is 10 percent of the

annual DOE occupational dose limit of 5000 mrem per year in 10 CFR 835, Subpart C. The radiation

doses to workers to be involved with preparation of the Vitrification Melter waste package for

shipment will be minimized by compliance with the WVDP radiological control program and the

associated ALARA processes.

Compliance with the radiological control program requirements and the ALARA processes will

provide reasonable expectation that WVDP worker doses will be well below the 500 mrem per year

limit, especially considering the low radiation levels on the outside of the Vitrification Melter waste

package and the short duration of the work to prepare the waste package for shipment.

Furthermore, the work associated with preparing the waste package for shipment is similar in

nature to other WVDP waste management work for which worker doses have been maintained

ALARA and well below the 500 mrem annual limit.

Compliance with the WVDP radiological control program requirements and the associated

ALARA processes will also ensure that potential exposures to the public from onsite work related to

preparing the Vitrification Melter for shipment are well below the applicable limit 42

. This work will

be performed within a radiologically controlled area within the WVDP security fence. Past WVDP

experience with similar waste management work indicates that potential doses to the public will be

42 The applicable limit is 10 mrem per year for exposure to a member of the public from air emissions (DOE Order

5400.5), which is the same as the U.S. Environmental Protection Agency requirement in 40 CFR 61.92, with which

DOE complies.

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very low. In 2008, for example, a year in which similar waste management work was performed by

the WVDP, the estimated dose to a maximally exposed offsite individual from WVDP airborne

radioactivity emissions was 0.0011 mrem (WVES and URS 2009). The airborne pathway is the only

pathway of interest for potential exposure to a member of the public from onsite work to prepare

the Vitrification Melter for shipment. Such factors provide reasonable expectation that doses to the

public from preparing the waste package for shipment will be far below the applicable limit.

Doses to workers at the Nevada National Security Site who would be involved with handling the

Vitrification Melter waste package to dispose of it in the Area 5 Radioactive Waste Management Site

would be minimized by compliance with that site's radiological control program and the associated

ALARA processes. Compliance with the radiological control program requirements, following ALARA

processes, the low radiation levels on the waste package, and the short duration of the work to

place it in the disposal facility provides reasonable expectation that worker doses will be ALARA.

Potential exposures to members of the public associated with onsite handling of the Vitrification

Melter at the Nevada National Security Site are also expected to be very low. Operations to dispose

of the shielded Vitrification Melter waste package would be of short duration, would take place in a

radiologically controlled area with no routine public access, and would take place at the isolated

government-controlled Nevada National Security Site.

Protection of Individuals During Operations at the WCS Facility

. If DOE were to transport the Vitrification Melter waste package to the WCS facility for disposal,

individuals would be protected during operations in manner similar to the Nevada National Security

Site. As noted above, the applicable State of Texas dose standards mirror those of NRC. WCS is

required to comply with the requirements of Texas Administrative Code, Title 30, Part 1, Chapter

336, Subchapter D, Standards for Protection Against Radiation, which provide for a comprehensive

* program to protect individuals and the public during waste disposal site operations.

5.2.5 Stability of the Disposal Site After Closure

The DOE requirements in DOE Manual 435.1-1, Sections IV.Q(1)(a) and (b) and IV.Q(2)(c), for

stability of the disposal site after closure are expressed as follows:

"Disposal Site Stability (DOE Manual 435.1, Section IV.Q(1)(a) and (b)). A preliminary closure

plan shall be developed and submitted to Headquarters for review with the performance

assessment and composite'analysis. The closure plan shall be updated following issuance of

the disposal authorization statement to incorporate conditions specified in the disposal

authorization statement. Closure plans shall:

(a) Be updated as required during the operational life of the facility.

(b) Include a description of how the disposal facility will be closed to achieve long-term

stability and minimize the need for active maintenance following closure and to ensure

compliance with the requirements of DOE 5400.5, Radiation Protection of the Public

and the Environment."

"Disposal Facility Closure (DOE Manual 435.1, Section IV.Q(2)(c)). Institutional control

measures shall be integrated into land use and stewardship plans and programs, and shall

continue until the facility can be released pursuant to DOE Order 5400.5, Radiation Protection

of the Public and the Environment."

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As discussed in Appendix C, NRC requirements in 10 CFR 61.44 sets forth similar, comparable

requirements:

"The disposal facility must be sited, designed, used, operated, and closed to achieve long-term

stability of the disposal site and to eliminate to the extent practicable the need for ongoing

active maintenance of the disposal site following closure so that only surveillance, monitoring,

or minor custodial care are required."

As explained in Appendix C, the State of Texas requirements for the stability of the disposal site

after closure mirror the NRC requirements.

DOE has developed a preliminary closure plan for the Area 5 Radioactive Waste Management

Site in accordance with the DOE requirements. This plan entails use of 2.5-meter (8.2-foot) thick

closure cover, consistent with assumptions used in the performance assessment (NST 2010),

although the closure cover thickness over the top of the buried Vitrification Melter waste package

would more than 10 feet (three meters)43 . The plan will ensure that the applicable requirements of

DOE Order 5400.5 will be met following closure of the Area 5 Radioactive Waste Management Site,

which is currently planned for 2028. The applicable requirements of DOE Order 5400.5 include the

public dose limit of 100 mrem per year effective dose equivalent (Chapter II, Section 1.a(1)) and

the airborne emissions limit of 10 mrem per year effective dose equivalent (Chapter II, Section 1.b

and 10 CFR Part 61).

The WCS license application (WCS 2007) - in Volume 2, Section 6, Closure - describes features

of the planned closure system for that facility to meet the State requirements for stability of the

disposal site after closure. These features include a depth of disposal significantly greater than five

meters (16.4 feet) for all waste.

5.3 The Vitrification Melter Meets Disposal Site Waste Acceptance Criteria

As noted previously, the WVDP has been accepted as an approved waste generator by the

Nevada National Security Site and has shipped LLW there for disposal on numerous occasions.

To protect workers, the public, and the environment, DOE establishes waste acceptance criteria

for its LLW disposal facilities, which, among other things, provide limits on the radionuclides that

may be disposed of at the facility, based on a performance assessment for the facility. As discussed

in Section 5.2, the performance assessment (and updates) for each LLW disposal facility provides

reasonable expectation that DOE's performance objectives in Chapter IV of DOE manual 435.1-1

will not be exceeded. Accordingly, disposal of the Vitrification Melter in compliance with the waste

41 NRC at 10 CFR 61.52(a)(2) provides that "Wastes designated as Class C pursuant to § 61.55, must be disposed of

so that the top of the waste is a minimum of 5 meters below the top surface of the cover or must be disposed of

with intruder barriers that are designed to protect against an inadvertent intrusion for at least 500 years." This

requirement, which does not apply to DOE LLW waste disposal facilities, is among the NRC rules at 10 CFR 61 that

were developed for nationwide application at a wide variety of facilities with geological and environmental settings

that are largely unknown until a specific facility is proposed.

DOE uses a different approach because its established disposal sites have well known geological and

environmental characteristics. This approach involves setting basic performance objectives, which are comparable to

those of NRC as demonstrated in Appendix C, and demonstrating with a site-specific performance assessment that

the total disposal system (site, design features, waste form, radionuclide content operating practices, and closure

plans) provides reasonable assurance that the performance objectives will be met. (DOE Guide 435.1-1, Appendix A,

Basis for Regulation of Low-Level Waste)

Table 5-1 and Table 5-2 show that disposal of the Vitrification Melter waste package at the Nevada National

Security Site Area 5 facility would have a negligible impact on disposal site performance.

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acceptance criteria for the Nevada National Security Site Radioactive Waste Management Site will

provide reasonable expectation that disposal will not exceed the DOE performance objectives in

Chapter IV of DOE Manual 435.1-1.

To help establish the relationship between the waste acceptance criteria and performance

assessments of the waste disposal sites, this subsection provides a summary of disposal site waste

acceptance criteria and explains how the Vitrification Melter meets these criteria. It also addresses

meeting the WCS waste acceptance criteria.

5.3.1 Nevada National Security Site Waste Acceptance Criteria

For its LLW disposal facilities, DOE provides formal waste acceptance criteria that comprise the

technical and administrative requirements that a waste must meet in order for it to be accepted at

the disposal facility (DOE Manual 435.1-1, Attachment 2). The Nevada National Security Site

provides specific radionuclide waste acceptance criteria for LLW (DOE 2009) that are expressed

primarily in terms of waste package activity limitations based on plutonium 239 equivalent grams

(PE-g).

This quantity relates the amount of a particular radionuclide to plutonium 239. Appendix B to

the criteria document (DOE 2009) contains a table of PE-g radionuclide conversion factors. These

conversion factors relate amounts of an individual radionuclide to plutonium 239. For example, the

conversion factor for cesium 137 is 2.72E-14 PE-g/Bq or approximately 1.OE-03 PE-g/Ci.

The Nevada National Security Site waste package limit for a single Department of

Transportation Type A drum is 300 PE-g total. The limit for a strong-tight container such as an

intermodal shipping container is also 300 PE-g total. An additional limitation of 2000 PE-g per

individual shipment also applies, except for Type B shipping containers in cases where the

containers themselves are to be disposed of. (DOE 2009)

Action levels for individual radionuclides are also provided in the Nevada National Security Site

waste acceptance criteria to identify radionuclides that must be reported on two key documents to

be submitted by the waste generator: the Waste Profile and the Package Storage and Disposal

Request. These action levels are used to. identify waste streams that may require special

consideration with regard to meeting the waste acceptance requirements."'4 5

4Radionuclides whose concentrations exceed one percent of the action level are required to be specifically reported

on the Package Storage and Disposal Request and the Waste Profile and require rigorous characterization (DOE

2009).

45 Other requirements address transuranic activity (the concentrations of alpha-emitting transuranic radionuclides

with half-lives over 20 years which must not exceed 100 nCi/g) and the amounts of fissile material present. The DOE

Manual 435.1-1 definition of transuranic waste includes alpha-emitting transuranic radionuclides with half-lives

greater than 20 years. In the 10 CFR 61.55 definition of transuranic waste, NRC includes alpha-emitting transuranic

radionuclides with half-lives greater than five years. In practice, Cm-244 (with its 18.1 year half-life) is the only

radionuclide covered by 10 CFR 61.55 that is not addressed as a transuranic radionuclide by DOE Manual 435.1-1.

The 10 CFR 61.55 requirements also include specific limits for Pu-241 and Cm-242, because these two radionuclides

decay to alpha-emitting transuranic isotopes with half-lives greater than five years (i.e., Am-241 and Pu-238,

respectively). The definition of transuranic waste in DOE Manual 435.1-1 is based on 10 CFR 149 and the Waste

Isolation Pilot Plant Land Withdrawal Act (U.S. Congress 1996), which identify transuranic radionuclides based on

concentrations and half-lives greater than 20 years. Consequently, these three radionuclides (Pu-241, Cm-242, and

Cm-244) are not transuranic radionuclides under concentration limits of Table E-1 of the Nevada National Security

Site Waste Acceptance Criteria (DOE 2009). Table E-1 provides radionuclide action levels for waste characterization

and reporting purposes.

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The criteria require radionuclides known or reasonably expected to be present in a waste

stream to be reported in the Package Storage and Disposal Request and the Waste Profile as

follows:

(1) When the activity concentration in the final waste form exceeds one percent of a specified

reporting action level specified in Table E-1 of the Waste Acceptance Criteria (DOE 2009),

(2) Any alpha-emitting. transuranic radionuclide with half-life over 20 years that exceeds 10

pCi/g,

(3) Any radionuclide whose concentration exceeds one percent of the total activity

concentration.

The Nevada National Security Site Waste Acceptance Criteria were developed to ensure

protection of public health and the environment both during ongoing operations of the waste

disposal sites and after these sites are closed. The acceptability of radionuclide concentration limits

specified to these ends is verified by DOE in a comprehensive performance assessment program,

as noted previously.

5.3.2 The Vitrification Melter Waste Package

Determining whether a waste package will meet the Nevada National Security Site radionuclide

limits involves: (1) identifying the activity of each reportable radionuclide present, (2) converting

this activity to PE-g, (3) summing all the individual PE-g values, and (4) comparing this total to the

Nevada National Security Site waste acceptance criteria individual package limit in PE-g. DOE has

compared characteristics of the Vitrification Melter waste package to the Nevada National Security

Site waste acceptance criteria and determined that it will meet these criteria. For example, the

Vitrification Melter contained an estimated 78.5 PE-g as of October 1, 2004 based on the

radionuclide estimates in Table 2-2, well below the individual package limit of 300 PE-g (WSMS

2008)6.

5.3.3 The Vitrification Melter Meets the Nevada National Security Site Waste

Acceptance Criteria

As noted previously, the WVDP has been accepted as an approved waste generator by the

Nevada National Security Site and has shipped LLW there for, disposal on numerous occasions.

Because of the established relationship between the waste acceptance criteria and performance

assessments of the waste disposal sites, satisfying the waste acceptance criteria indicates

compliance with the disposal site performance assessment and, hence, with the DOE performance

objectives.

For its LLW disposal facilities, DOE provides formal waste acceptance criteria that comprise the

technical and administrative requirements that a waste must meet in order for it to be accepted at

the disposal facility (DOE Manual 435.1-1, Attachment 2). These criteria for the Nevada National

Security Site are contained in its Waste Acceptance Criteria document (DOE 2009).

As described in Section 2, DOE has packaged the Vitrification Melter in a robust Type IP-2

container. Prior to shipment for disposal, internal void spaces will be filled with low-density cellular

concrete. The low-density cellular concrete will be compatible with the waste acceptance criteria

46 The waste profile shows a higher estimate of 97.7 PE-g, which was developed using the "high activity" radionuclide

concentration estimates, which provide the upper bound of the estimated inventory.

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and will help reduce the possibility of disposal cell subsidence in the area of the buried waste

container by eliminating the container void spaces.

A Waste Profile Sheet for disposal of the Vitrification Melter waste package was submitted to

the Nevada National Security Site for formal review and approved as specified in the waste

acceptance criteria, as noted previously. Table 5-3 compares the radionuclide concentrations in the

Vitrification Melter with the radionuclide action levels for waste characterization and reporting

provided in the Nevada National Security Site Waste Acceptance Criteria (DOE 2009). Radionuclides

with concentrations exceeding one percent of the action level are highlighted in the table.47

Table 5-3. Vitrification Melter Radionuclide Concentrations in Bq/m 3(l)

Nuclide Action Level(2 ) Melter(3 ) Nuclide Action Level(2) Melter(3)

H-3 6.2E+11 1.38E+04 U-232 4.3E+10 8.80E+07

C-14 5.4E+15 3.97E+07 U-233 8.2E+10 3.59E+07

K-40 9.4E+10 1.43E+08 U-234 1.3E+10 1.71e+07

Mn-54 NA 1.49E+08 U-235 1.1E+10 6.56E+05

Fe-55 NA 6.69E+06 U-236 2.8E+11 1.97E+06

Ni-59 1.7E+14 3.64E+05 U-238 3.5E+11 3.93E+06

Co-60 1.6E+12 1.46E+08 Np-237 3.4E+10 1.08E+07

Ni-63 3.2E+ 14 1.77E+09 Pu-238 1.8E+12 1.19E+09

Sr-90 4.3E+11 4.32E+11 Pu-239 5.1E+11 2.78E+08

Zr-95 NA 2.89E+09 Pu-240 5.2E+11 2.11E+08

Tc-99 3.2E+09 1.94E+07 Pu-241 5.8E+12 5.46E+09

1-129 3.4E+09 2.77E+05 Pu-242 3.7E+11 3.97E+04

Cs-137 2.5E+11 . 7.52E+12 Am-241 1.7E+11 5.25E+09

Pm-147 NA 4.95E+07 Am-243 5.8E+10 6.30E+07

Eu-154 1.7E+12 2.12E+09 Cm-242 NA 1.28E+08

Th-228 4.3E+13 7.13E+07 Cm-243 8.3E+11 2.93E+07

Th-229 2.8E+10 2.8E+02(4) Cm-244 3.4E+12 7.67E+08

T 306:0E+O7'`_ý- 6 3E+05'-r,,, Cm-245 4.6E+10 1.92E+07

Th-232 8.1E+09 6.99E+05 Cm-246 9.2E+10 3.12E+06

NOTES: (1) To convert Bq/m 3 to pCi/L multiply by 0.027.

(2) From the Table E-1 of the Waste Acceptance Criteria (DOE 2009).

(3) From the Vitrification Melter waste profile technical basis document (WVES 2010). The waste profile provides

a shorter list of radionuclides focused on those that exceed one percent of the action level.

(4) Th-229 concentration estimated using Th-229 to Th-230 ratio from high-level waste sludge solids

characterization (Rykken 1986, Table 22).

Table 5-4 shows that the estimated concentration of Cs-137 in the Vitrification Melter exceeds

the action level for this radionuclide. However, this would have a negligible impact on performance

47 As noted previously, the waste acceptance criteria document (DOE 2009) requires that activity concentrations of

the radionuclides in the final waste form exceeding one percent of the action level in Table E-1 of that document

receive rigorous waste characterization and be reported on the package storage and disposal request and the waste

profile. The Vitrification Melter was rigorously characterized by the WVDP.

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of the Nevada National Security Site Area 5 Radioactive Waste Disposal Site as shown by the

results of the dose impact analyses provided in Tables 5-1 and 5-2 (DOE 2010b).

5.4 Meeting WCS Waste Acceptance Criteria

The WCS waste acceptance criteria document (WCS 2008) addresses matters such as

operations and regulatory parameters, pre-shipment requirements, 'documentation, and

transportation. It provides various forms including a waste profile sheet. Unlike the Nevada

National Security Site waste acceptance criteria, the WCS waste acceptance criteria document does

not provide numerical radionuclide concentration action levels. However, the separate WCS Waste

Acceptance Plan (WCS 2009) provides additional information related to the waste acceptance

process, including waste form requirements and a description of the generator and waste approval

processes.

The WCS license (TCEQ 2010) contains additional requirements related to waste disposal,

including total waste volume limitations and total activity limitations for certain radionuclides. Table

5-4 shows representative requirements compared to the related parameters for the Vitrification

Melter waste package.

TABLE 5-4. Key WCS Federal Facility Waste Disposal Facility License Requirements

License Vitrification Melter %/o Requirement (Section)(') LimitM) Melter of Limit

Total waste volume, ft3 (§7.B) 8,100,000 2331(2) 0.029

Total activity, curies (§7.B) 5,500,000 4570(3) 0.083

Total C-14, curies (§5.D) 180 0.0212(3) 0.012

Total Tc-99, curies (§5.D) 35 0.0111(3) 0.032

Total 1-129, curies (§5.D) 0.15 0.00049(2) 0.33

NOTES: (1) From the WCS license (TCEQ 2010) with the associated section numbers and limits. The §7.B limits are for

Class A containerized, Class B, and Class C LLW, collectively.

(2) From the Vitrification Melter Waste Profile Sheet for the Nevada National Security Site (WVES 2010).

(3) From Table 2-2.

Table 5-4 shows the volume of the Vitrification Melter waste package is a small fraction of the

WCS federal facility waste disposal facility capacity limit and that the total activity and the activity

of the license-limited radionuclides in the waste package are small fractions of the WCS limits.

If DOE were to elect to dispose of the Vitrification Melter at the WCS federal facility waste

disposal facility, DOE would confirm that the waste package meets the waste acceptance criteria

for that facility prior to shipment. DOE would follow the WCS process and submit all of the

necessary supporting information; such as the Waste Profile Form48.

48 Because the WCS facility is licensed to accept Class C LLW, DOE would expect that the Vitrification Melter waste

package would be approved for disposal. However, because the Vitrification Melter is not among the planned waste

streams identified in the documents supporting the WCS license application, a license amendment may be necessary

to obtain approval from the regulator for disposal of the Vitrification Melter waste package.

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6.0 The Waste Does Not Exceed Class C Concentration Limits and

Will Be Managed in Accordance With DOE Requirements as LLW

Section Purpose

The purpose of this section is to demonstrate that the Vitrification Melter waste

package will be in a solid physical form, will not exceed Class C concentration

limits, and will be managed in accordance with DOE requirements as low-level

radioactive waste as applicable.

Section Contents

This section provides information showing that the grouted Vitrification Melter

waste package will be in a solid physical form, will not exceed the concentration

limits for Class C low-level waste in 10 CFR 61.55, and will be managed and

disposed of as low-level waste in accordance with DOE requirements.

Key Points

" The grouted Vitrification Melter waste package will be in a solid physical

form.

" The radioactivity in the Vitrification Melter waste package will not exceed

Class C concentration limits.

• The Vitrification Melter waste package will be managed and disposed of at an

offsite low-level radioactive waste disposal facility in accordance with

applicable requirements for low-level waste.

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The third and final criterion of DOE Manual 435.1-1, Section II.B(2)(a) to be demonstrated is:

"[The wastes] are to be managed, pursuant to DOE's authority under the Atomic Energy Act of

1954, as amended, and in accordance with the provisions of Chapter IV of DOE Manual 435.1-

1, provided the waste will be incorporated in a solid physical form at a concentration that does

not exceed the applicable concentration limits for Class C low-level waste as set out in 10 CFR

61.55, Waste Classification; or will meet alternative requirements for waste classification and

characterization as DOE may authorize."

As explained previously, the Vitrification Melter has been packaged in a shielded IP-2 shipping

container and voids in the Vitrification Melter and the space between the Vitrification Melter and the

inside of the container will be filled with low-density cellular concrete prior to shipment for

disposal. Hence, the Vitrification Melter waste package will be in a solid physical form. (The Melter

itself is already in a solid physical form, as noted previously.)

Because the Vitrification Melter contains a mixture of radionuclides, the total concentration is

determined by the sum of the fractions rule, as specified in NRC's regulations at 10 CFR

61.55(a)(7) (§336.362(a)(7) of the Texas Administrative Code parallels the NRC's regulations).

Additionally, because the radionuclide mixture contains some long-lived radionuclides that are listed

on Table 1 of 10 CFR 61.55 (reproduced in Table 4-1 of this draft waste incidental to reprocessing

evaluation), and some short-lived radionuclides that are listed on Table 2 of 10 CFR 61.55

(reproduced in Table 4-2 of this draft waste incidental to reprocessing evaluation), waste

classification would be determined as specified in 10 CFR 61.55(a)(5), which states:

"If radioactive waste contains a mixture of radionuclides, some of which are listed in Table 1,

and some of which are listed in Table 2, classification shall be determined as follows:

(i) If the concentration of a nuclide listed in Table 1 does not exceed 0.1 times the value

listed in Table 1, the class shall be that determined by the concentration of nuclides

listed in Table 2.

(ii) If the concentration of a nuclide listed in Table 1 exceeds 0.1 times the value listed in

Table 1 but does not exceed the value in Table 1, the waste shall be Class C, provided

the concentration of nuclides listed in Table 2 does not exceed the value shown in

Column 3 of Table 2."

Radiological characterization of the Vitrification Melter before packaging was as described in

Section 2.5.3. Using the results of that characterization, Table 6-1 below shows that the Melter

does not exceed Class C concentration limits.

Table 6-1. Vitrification Melter Waste Concentration Results(')

Estimated Class C Class C Melter Melter Table 1(2) Table 2(2)

Isotope Activity Limit Limit Concentration Concentration

(Ci) (Ci/m 3

) (nCi/g) (Ci/m3) (nCi/g) Fraction Fraction

C0-4 2.12E-02 8.00E+00 1.OOE-03 1.25E-04

Ni-63 1.01 E+O0 7.OOE+02 4.76E-02 6.81E-05

Sr-90 2.47E+02 7.00E+03 1.17E+01 1.66E-03

Tc-99 1.11 E-02 3.OOE+O0 5.24E-04 1.75E-04

1-129 5.64E-03 8.OOE-02 2.66E-04 3.33E-03

Cs-137 4.31E+03 4.60E+03 2.03E+02 4.42E-02

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Table 6-1. Vitrification Melter Waste Concentration Results(') (Continued)

Estimated Class C Class C Melter Melter Table 1(2) Table 2(2)

Isotope Activity Limit Limit Concentration Concentration

(Ci) (Ci/m 3 ) (nCi/g) (Ci/m3) (nCi/g) Fraction Fraction

Np-237 6.20E-03 1.OOE+02 1.29E-01 1.29E-03

Pu-238 6.84E-01 1.OOE+02 1.42E+01 1.42E-01

Pu-239 1.59E-01 1.OOE+02 3.31 E+00 3.31 E-02

Pu-240 1.21 E-01 1.00E+02 2.52E+00 2.52E-02

Pu-241 3.12E+00 3.50E+03 6.49E+01 1.85E-02

Am-241 3.OOE+00 1.0DE+02 6.24E+01 6.24E-01

Am-243 3.50E-02 1 .OOE+02 7.28E-01. 7.28E-03

Cm-242 7.33E-02 2.OOE+04 1.52E+00 7.62E-05

Cm-243 1.68E-02 1 .OOE+02 3.49E-01 3.49E-03

Cm-244 4.35E-01 1.OOE+02 9.04E+00 9.04E-02

Sums of fractions(3) 9.49E-01 4.59E-02

NOTES: (1) From WSMS 2008, with the calculations based on the Vitrification Melter weight and size. The activity estimates

used were as of October 1, 2004; the activities are now somewhat lower due to radioactive decay. The weight

used in the calculation was 106,000 pounds and the volume used was 750 cubic feet.

(2) Table numbers refer to 10 CFR 61.55, Tables 1 and 2. (Table I and Table II to Appendix E to Rule §336.362 of

the Texas Administrative Code are identical to.NRC's Table I and Table 2.)

(3) As a sensitivity analysis related to the Vitrification Melter volume, another calculation was performed with the

volume reduced by 50 percent to 375 cubic feet. The resulting sums for fractions were 9.52E-01 for Table 1

and 9.19E-02 for Table 2 (WSMS 2008).

This, table shows that the sum of fractions for both the 10 CFR 61.55 Table 1 (long-lived) and

Table 2 (short-lived) radionuclides is below 1.0, and thus the Vitrification Melter does not exceed.

concentration limits for Class C LLW. The calculations were performed using the weight and size of

the Vitrification Melter itself; neither the grout nor the shipping container was considered.

As discussed previously, this waste may be transported to the Nevada National Security Site

Area 5 Radioactive Waste Management Site for disposal. At the Nevada National Security Site, the

Vitrification Melter waste package would be disposed of as LLW and managed in accordance with

DOE requirements for LLW disposal in Chapter IV of DOE Manual 435.1-1. The required Waste

.Profile has been developed by DOE in accordance with the Nevada National Security Site Waste

Acceptance Criteria (WVES 2010). This Waste Profile has been formally approved by the Nevada

National Security Site, as noted previously (DOE 2010a), in case a final decision is made to send

the Vitrification Melter waste package to the Nevada National Security Site for disposal.

As noted previously, DOE may elect to send the Vitrification Melter to the commerciallyoperated WCS federal facility waste disposal facility in Texas. As demonstrated earlier in this draft

evaluation, the Vitrification Melter waste package will be in a solid physical form (with voids in the

Vitrification Melter and the space between the Vitrification Melter and the inside of the container.

filled with low-density cellular concrete), and will not exceed Class C concentration limits in 10 CFR

61.55. In this regard, the Texas Administrative Code has similar requirements concerning waste

stability and as little free standing liquid as possible. 49 In addition, the State of Texas Class C

49 Texas Administrative Code, Title 30, Part 1, §336.362, Appendix E.

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concentration limits mirror the concentration limits in 10 CFR 61.55; consequently, disposal of the

Vitrification Melter waste package in the WCS Federal Waste Disposal Facility would not exceed

Class C LLW concentration limits set forth in the Texas Administrative Code.

DOE Order 435.1, Radioactive Waste Management, provides that requirements in the Order

that duplicate or conflict with requirements of an applicable Agreement State do not apply to

facilities and activities licensed by the Agreement State. Therefore, the provisions in Chapter IV of

DOE Manual 435.1-1 concerning matters such as monitoring, waste acceptance criteria,

performance assessments, composite analysis, disposal facility operations, disposal authorizations,

institutional control, and disposal facility closure do not apply to the WCS facility; instead, these

matters are governed by the State of Texas requirements and license conditions.

Accordingly, as demonstrated above, disposal of the Vitrification Melter waste package at the

WCS Federal Waste Disposal Facility would meet the third criterion of DOE Manual 435.1-1, Section

II.B.2(a).

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DRAFT WASTE INCIDENTAL TO REPROCESSING EVALUATION FOR THE WVDP VITRIFICATION MELTER

7.0 CONSULTATION WITH NRC AND OPPORTUNITY FOR PUBLIC COMMENT

As explained previously, DOE is consulting with NRC concerning this draft evaluation, and is

making this draft evaluation available for public review and comment -including the States of

Nevada and Texas where the Vitrification Melter waste package might be disposed of as LLW.

DOE will consider NRC comments, as well as comments from the public and the States of

Nevada and Texas, before finalizing the evaluation and before making any final determination as to

whether the Vitrification Melter meets the criteria in DOE Manual 435.1-1 for waste incidental to

reprocessing, and thus is not high-level waste and is to be managed and disposed of as low-level

waste pursuant to DOE's regulatory authority under the Atomic Energy Act of 1954, as amended.

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8.0 CONCLUSIONS

Based on information provided in the preceding sections of this draft evaluation, DOE has reached the

preliminary conclusion that the Vitrification Melter is not HLW based on the criteria of DOE Manual 435.1-1 and

may be managed as LLW.

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9.0 REFERENCES

Federal Statutes

West Valley Demonstration Project Act, Public Law 96-368 (S. 2443), of October 1, 1980.

Low-Level Radioactivity Waste Policy Act of 1985.

Waste Isolation Pilot Plant Land Withdrawal Act of 1992 as amended in 1996, Public Law 102-579. Public

Law 104-201, 110 Stat. 2422, 1996.

Ronald W Reagan National Defense Authorization Act for Fiscal Year 2005, Public Law 108-375, 118 Stat.

1811. 108th United States Congress, October 28, 2004.

Code of Federal Regulations and Federal Register Notices

10 CFR 20, Standards for Protection Against Radiation.

10 CFR 61.55, Waste Classification.

10 CFR 61, Subpart C, Licensing Requirements for Land Disposal of Radioactive Waste, Performance

Objectives.

10 CFR 835, Occupational Radiation Protection.

40 CFR 61, National Emission Standards for Hazardous Air Pollutants.

47 FR 179, Long-Term Management of Liquid High-Level Radioactive Wastes Stored at the Western New

York Nuclear Service Center, West Valley; Record of Dedsion, U.S. Department of Energy,

Washington, D.C., September 15, 1982.

65 FR 38, Record of Decision for the Department of Energy's Waste Management Program: Treatment

and Disposal of Low-Level Waste and Mixed Low-Level Waste; Amendment of Decision for the

Nevada TestSite, U.S. Department of Energy, West Valley, New York, February 25, 2000.

70 FR 115, West Valley Demonstration Project Waste Management Environmental Impact Statement,

Record of Decision, U.S. Department of Energy, West Valley, New York, June 16, 2005.

75 FR 75, Record of Decision: Final Envfronmental Impact Statement for Decommissioning and/or LongTerm Stewardship at the West Valley Demonstration Project and Western New York Nuclear

Service Center. U.S. Department of Energy, West Valley, New York, April 20, 2010.

DOE Orders, Policies, Manuals, and Standards

DOE Order 435.1, Radioactive Waste Management, Change 1. U.S. Department of Energy, Washington,

D.C., August 28, 2001.

DOE Order 5400.5, Radiation Protection of the Public and the Environment, Change 2. U.S. Department of

Energy, Washington, D.C., January 7, 1993.

DOE Policy 441.1, Department of Energy Radiological Health and Safety Policy. U.S. Department of

Energy, Washington, D.C., April 26, 1996.

DOE Manual 435.1-1, Radioactive Waste Management Manual, Change 1. U.S. Department of Energy,

Washington, D.C., June 19, 2001.

DOE Guide 435.1-1, Implementation Guide For Use With DOE M 435.1-1. U.S. Department of Energy,

Washington, D.C., July 9, 1999.

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State Regulations

Texas Administrative Code, Title 30, Part 1, Chapter 336, Radioactive Substance Rules (http://

info.sos.state.tx.us/pls/pub/readtac$ext.ViewTACtacview=4&ti =30&pt= l&ch =336)

Other References

ASME 2004, The Decommissioning Handbook, Toboas, A., A, Moghissi, and T. LaGuardia, Ed., American

. Society of Mechanical Engineers and the American Nuclear Society, New York, New York, 2004.

BRH 1970, Radiological Health Handbook, Revised Edition. Bureau of Radiological Health, U.S.

Department of Health, Education, and Welfare, Rockville, Maryland, January 1970.

Brooks 1993, Slurry Fed Ceramic Melter Disassembly Report, Brooks, R., West Valley Nuclear Services,

West Valley, New York, April 23, 1993.

Cambridge 2004, Cambridge Dictionary of American English, Cambridge University Press, 2004 (http://

dictionary.cambridge.org/define.asp?key=practical*2+0&dict=A).

CLI-05-05, 2005, In the Matter of Louisiana Energy Services, L.P. (National Enrichment Services), slip op.

at 11, January 18, 2005.

Cole, et al. 1995, Comparison of Selected DOE and Non-DOE Requirements, Standards, and Practices for

Low-Level Radioactive Waste Disposal, DOE/LLW-225, Revision 0. Cole, L., D. Kudera, and W.

Newberry, Idaho National Engineering Laboratory, Lockheed Idaho Technologies Company, Idaho

Fall, Idaho, December 1995.

DOE 1994, Decommissioning Handbook, DOE/EM-0142P. U. S. Department of Energy, Washington, D.C.,

March 1994.

DOE 1996, Plutonium Recovery from Spent Fuel Reprocessing by Nuclear Fuel Services at West Valley,

New York from 1966 to 1972. U. S. Department of Energy, Washington, D.C., February 1996.

DOE 1997, Final Waste Management Programmatic Environmental Impact Statement for Managing

Treatment, Storage, and Disposal of Radioactive and Hazardous Waste, DOE/EIS-0200-F. U.S.

Department of Energy, Washington, D.C., May 1997.

DOE 2003, Final West Valley Demonstration Project Waste Management Environmental Impact Statement,

DOE/EIS-0337. DOE-West Valley Office, West Valley, New York, December 2003.

DOE 2006, West Valley Demonstration Project Waste Management Environmental Impact Statement,

Supplement Analysis, DOE/EIS-0337-SA-01, Revised Final. U.S. Department of Energy, West

Valley, New York, June 7, 2006.

DOE 2009, Nevada Test Site Waste Acceptance Criteria, DOE/NV-325-Rev. 7-01. -U.S. Department of

Energy, National Nuclear Security Administration, Nevada Site Office, Waste Management Project,

Las Vegas, Nevada, May 2009.

DOE 2010a, Letter from DOE-Nevada Site Office to DOE-West Valley on Approval to Ship West Valley

Environmental Services (WVES) Low-Level Radioactive Waste [the Vitrification Melter Waste

Package] to the Nevada Test Site, dated July 6, 2010.

DOE 2010b, Letter from DOE-Nevada Site Office to DOE-West Valley on Special Performance Assessment

(PA) Results Summary of Impacts of the Disposal of the West Valley Glass Melter in Area 5

Radioactive Waste Management Site (RWMS), WMP:6655:JC, dated July 29, 2010.

DOE and NRC 1981, West Valley Demonstration Project Memorandum of Understanding Between the U.S.

Department of Energy and the U.S. Nuclear Regulatory Commission. September 23, 1981.

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DOE and NYSERDA 1981, Cooperative Agreement between the United States Department of Energy and

New York State Energy Research and Development Authority on the Western New York Nuclear

Service Center at West Valley, New York. Signed November 3, 1980, amended September 18,

1981.

DOE and NYSERDA 2010, Final Environmental Impact Statement for Decommissioning and/or Long-Term

Stewardship at the West Valley Demonstration Project and Western New York Nuclear Service

Center, DOE/EIS-0226-D. U.S. Department of Energy and the New York State Energy Research

and Development Authority, West Valley, New York, January 2010.

Eisenstatt 1986, Description of the West Valley Demonstration Project Reference High-Level Waste Form

and Canister, DOE/NE/44139-26. Eisenstatt, L. R., U.S. Department of Energy, West Valley, New

York, 1986.

Fowler 1930, A Dictionary of Modern English Usage. Fowler, H., 1930.

Kelly and Meess 1997, THOREX Processing and Zeolite Transfer for High-Level Waste Stream Processing

Blending, Topical Report DOE/NE/44139-82. Kelly, S. and D.C. Meess, West Valley Nuclear

Services Company, West Valley, New York, July 1997.

Lachapelle 2003, Vitrification Facility (VF) Radioisotope Inventory Report, RIR-403-010, Revision 1.

Lachapelle, E., West Valley Nuclear Services Company, West Valley, New York, June 17, 2003.

Merriam-Webster 2010, Merriam-Webster On-Line Dictionary, www.merriam-webster.com/netdict/.

NRC 1982, Final Environmental Impact Statement on 10 CFR 61 "Licensing Requirements for Land

Disposal of Radioactive Waste," NUREG-0945. U.S. Nuclear Regulatory Commission, Washington,

D.C., November 1982.

NRC 1995, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission, NUREG/BR-0058,

Revision 2. U.S. Nuclear Regulatory Commission, Washington, D.C., November 1995.

NRC 1997, Regulatory Analysis Technical Evaluation Handbook, NUREG/BR-0184. U.S. Nuclear Regulatory

• Commission, Washington, D.C., January 1997.

NRC 2000, A Performance Assessment Methodology for Low-Level Radioactive Waste Disposal Facilities,

Recommendations of NRC's Performance Assessment Working Group, NUREG-1573. U.S. Nuclear

Regulatory Commission, Washington, D.C., October 2000.

NRC 2002, Decommissioning Criteria for the West Valley Demonstration Project (M-32) at the West Valley

Site; Final Policy Statement U.S. Nuclear Regulatory Commission, Washington, D.C., Federal

Register, Vol. 67, No. 22, February 1, 2002.

NRC 2003, U.S. Nuclear Regulatory Commission Implementation Plan for the Final Policy Statement on

Decommissioning Criteria for the West Valley Demonstration Project at the West Valley Site. U.S.

Nuclear Regulatory Commission, Division of Waste Management, Office of Nuclear Materials Safety

and Safeguards, Washington, D.C., May 2003.

NRC 2004, Letter from R.R. Bellamy (NRC) to T.J. Jackson (DOE-West Valley), U.S. Nuclear Regulatory

Commission Monitoring Visit 2004-02. U.S. Nuclear Regulatory Commission Region, 1 King of

Prussia, Pennsylvania, December 9, 2004.

NRC 2005, U.S. Nuclear Regulatory Commission Technical Evaluation Report for the U.S. Department of

Energy Savannah River Site Draft Section 3116 Waste Determination for Salt Waste Disposal. U.S.

Nuclear Regulatory Commission, Washington, D.C., December 2005.

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NRC 2007, NRC Staff Guidance for Activities Related to U.S. Department of Energy Waste Determinations,

NUREG-1854. Draft Final Report, U.S. Nuclear Regulatory Commission, Washington, D.C., August

2007.

NST 2009a, Nevada Test Site Environmental Report 2008, DOE/NV/25946 - 790. National Security

Technologies, LLC, Las Vegas, Nevada, September 2009.

NST 2009b, Nevada Test Site Radiological Control Manual, DOE/NV/25946-801, Revision 0. National

Security Technologies LLC, Las Vegas, Nevada, October 2009.

NST 2010, 2009 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites

.at the Nevada Test Site Nye County, Nevada: Review of the Performance Assessments and

Composite Analyses, DOE/NV/25946-6893. National Security Technologies, LLC, North Las

Vegas, Nevada, March 2010.

OMB 1992, Regulatory-Impact Analysis Guidance, Appendix V in "Regulatory Program of The United

States Government." U.S. Office of Management and Budget, Washington, D.C., April 1991 -

March 31, 1992.

Petkus, et al. 2003, A Complete History of the High-Level Waste Plant at the West Valley Demonstration

Project. Petkus, L.L, J. Paul, P. Valenti, H. Houston, and J. May, Waste Management 03

Conference, Tucson, Arizona, February 23-27, 2003.

Purdue 2004, Economic Practicality Analysis for the West Valley Demonstration Project's Vittification

Facility Melter. Purdue, R.K., Westinghouse Electric Company Science and Technology

Department, Pittsburgh, Pennsylvania, November 8, 2004.

Random House 1997, Random House Unabridged Dictionary, Random House, Inc. (http://www.infoplease

.com/dictionary.html, Web page visited March 15, 2005).

RHI 2010, RHI Monofrax website, www.rhi-ag.com/internet/localchannels/monofrax/en/default.html.

Rykken 1986, High-Level Waste Characterization at West Valley, DOE/NE/44139-14. Rykken, L.E., West

Valley Nuclear Services Co., West Valley, New York, June 2, 1986.

TCEQ 2010, Radioactive Material License R04100 Issued to Waste Control Specials, LLC, Amendment 02.

Texas Commission on Environmental Quality, Austin, Texas, March 5, 2010.

Vance, et al. 1997, Design of Equipment Used for High-Level Waste Vitrification at the West Valley

Demonstration Project, Topical Report DOE/NE/44139-73. Vance, R.F, B.A. Brill, D.E. Carl, H.S.

Dhingra, R.A. Meigs, and M.G. Studd, West Valley Nuclear Services Company, West Valley, New

York, June 1997.

WCS 2007, Application for License to Authorize Near-Surface Land Disposal for Low-Level Radioactive

Waste, Section 8." Performance assessment, Revision 12a. Waste Control Specialists LLC,

Rosebud, Texas, March 16, 2007. (http://www.urs-slc.com/wcs/)

WCS 2008, Waste Acceptance Criteria. Waste Control Specialists LLC, Rosebud, Texas, June 18, 2008.

WCS 2009, Waste Acceptance Plan for Low-Level Radioactive Waste Disposal Facilities, Waste Control

Specialists, LLC, Andrews, Texas Facility, Revision 1. Waste Control Specialists LLC, Rosebud,

Texas, September 3, 2009.

Webster's 1984, Webster's II New Riverside University Dictionary, Riverside Publishing Co., 1984.

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Wilhite 2001, Compafison of LLW Disposal Performance Objectives, 10 CFR 61 and DOE 435.1, WSRC-RP2001-00341. Wilhite, E.L., Westinghouse Savannah River Company, Aiken, South Carolina, 2001.

WMG 2004a, West Valley Nuclear Services Company (WVNSCO) Melter Characterization Results, Report

4005-RE-024, Revision 3. WMG, Inc. Peekskill, New York, May 2004.

WMG 2004b, WVDP Package Description Report, Report 4005-RE-030, Revision 2. WMG, Inc. Peekskill,

New York, May 2004.

WSMS 2008, Vitrification Melter Waste Classification, WSMS-WV-08-CALO02, Washington Safety

Management Solutions, West Valley New York, March 2008.

WVES 2010, Waste Profile Sheet for WVDP Vitrification Melter, WVDP-000000021, Revision 1 (and

associated documentation). West Valley Environmental Services LLC, West Valley, New York, May

24, 2010.

WVES and URS 2009, West Valley Demonstration Project Annual Site Environmental Report for Calendar

Year 2008. West Valley Environmental Services LLC and URS - Washington Division, West Valley,

New York, September 2009.

WVNSCO 2001, Waste Incidental to Reprocessing (WIR) Evaluation for Vitrification Facillty Expended

Materials, Revision 1. West Valley Nuclear Services Company, West Valley, New York, October

25, 2001.

WVNSCO 2004, Radiation and Contamination Survey Report 127046, West Valley Nuclear Services

Company, West Valley, New York, November 15, 2004.

WVNSCO 2006, WVDP Radiological Controls Manual, WVDP-010, Revision 27. Lachapelle, E.B., West

Valley Nuclear Services Company, West Valley, New York and URS Corporation, Buffalo, New

York, November 8, 2006.

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APPENDIX A

Drawings for the Vitrification Melter and Its Shipping Container

Appendix Purpose

The purpose of this appendix is to provide copies of the drawings of the Vitrification

Melter and its shipping container so this information will be readily available.

Appendix Content

Five of these drawings are included.

Key Points

The drawings show the configuration and key dimensions of the Vitrification Melter

and its waste container.

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DRAFT WASTE INCIDENTAL TO REPROCESSING EVALUATION FOR THE WVDP VITRIFICATION MELTER

_________ 8 I 7 I 6 5 4 4 I 3 I 0I-27 A ,

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Figure A -1. Vitrification Melter Main Assembly Drawing

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DRAFT WASTE INCIDENTAL TO REPROCESSING EVALUATION FOR THE WVDP VITRIFICATION MELTER

NOZZLE LEGEND

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Figure 1-2. Vitrification Melter Plan and Nozzle Location Drawing

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0

C

DRAFT WASTE INCIDENTAL TO REPROCESSING EVALUATION FOR THE WVDP VITRIFICATION MELTER

ELEVATION A-A - ao

FOR NOTES SEE DRAWING 900D-2780 SH.I

•, I•, . !t (- • fIf MAIN FACILITY ASSEMBLY - MELTER

EAST ELEVAiON

- r"r -""--" -- " .2 .. . i 7 6 5 4

B

D

Figure A-3. Vitrification Melter Elevation Drawing

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A-4

0 A-4
DRAFT WASTE INCIDENTAL TO REPROCESSING EVALUATION FOR THE WVDP VITRIFICATION MELTER

'rooo

Container Top 4-in Thick Steel

Lifting Attachments r. - ` Void•Sp;c6: ip`itn`ict ion4

and Corner Bracin

Not Shown dr

C1)

Container Sides 6-in Thick Steel,

Bottom 4-in Thick

12 ft

Figure A-4. Vitrification Melter Shipping Container

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APPENDIX B

Management Controls to Assure Quality in this Draft Evaluation

Appendix Purpose

The purpose of this appendix is to describe management control systems used to

ensure quality in the process of evaluating whether the Vitrification Melter is

incidental to reprocessing so it can be managed as low-level waste.

Appendix Content

This appendix identifies applicable Department of Energy and contractor

requirements, describes the contractor quality assurance program, explains how

quality was assured in data evaluation and engineering calculations, describes

qualifications of the waste evaluation team, and outlines how documents and records

related to the draft evaluation are controlled.

Key Points

* The process used was consistent with Department of Energy quality standards in

10 CFR 830, Subpart A, Quality Assurance Requirements, and DOE Order 414.1B,

Quality Assurance.

* The process used was also consistent with contractor quality standards included in

the Washington Safety Management Solutions Quality Assurance Plan and related

implementing procedures.

* Engineering calculations prepared in support of this draft evaluation were

independently reviewed and formally approved.

* This draft evaluation was prepared by a contractor team made up of waste

management professionals experienced in waste incidental to reprocessing

determinations, under the direction and oversight of the Department's West Valley

Project Office.

* This draft evaluation received several reviews.

1.0 Introduction

The purpose of this appendix is to describe management controls used to ensure that this draft

evaluation was prepared with quality commensurate with its importance in relation to safe

management of the Vitrification Melter to protect worker and public health and safety and the

environment.

The management control systems used in this draft evaluation are formalized in a series of

related standards and written procedures. They are designed to ensure that primary project

objectives are met and that an optimum margin of safety for protection of personnel, the public,

and the environment is achieved. Implementation of these systems by DOE and contractor

managers and engineers provides assurance that information and engineering analyses which form

the basis for this draft waste evaluation are accurate and supportable, and that the proper

conclusions were reached.

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.This ,appendix addresses the following matters: (1) requirements and procedures, (2) quality

assurance, (3) data evaluation and engineering calculations, (4) report preparation and review, (5)

personnel qualification, (6) oversight and independent review, and (7) document and record

control.

2.0 Requirements and Procedures

The process followed in preparation of this draft evaluation was consistent with applicable

requirements and guidance promulgated by DOE and the two contractors involved - West Valley

Environmental Services (WVES) and Washington Safety Management Solutions (WSMS)50 -

including the following:

DOE

* 10 CFR 830, Subpart A, QualityAssurance Requirements

* DOE Order 414.1B, QualityAssurance

* DOE Order 435.1, Radioactive Waste Management

* DOE Manual 435.1-1, Radioactive Waste Management Manual

* DOE Guide 435.1-1, Implementation Guide

* DOE Order 5400.5, Radiation Protection of the Public and the Environment

Contractor Manuals and Procedures

* WSMS QA 100, Washington Safety Management Solutions Quality Assurance Plan (WSMS

2008)

" WV-929, Waste Incidental to Reprocessing-Determination (WVNSCO 2007)

3.0 Quality Assurance

The quality assurance process followed in this draft evaluation was based on DOE requirements

contained in (1) .10 CFR Part 830, Subpart A which establishes quality requirements for DOE

contractors conducting activities including providing items and services that affect, or may affect,

nuclear safety of DOE facilities and (2) DOE Order 414. 11, Quality Assurance and its associated

manuals.

The WSMS Quality Assurance Plan (WSMS 2009) establishes WSMS policy related to quality and

provides procedures to implement this policy, consistent with applicable DOE requirements in 10

CFR 830 and DOE Order 414.1B. It also establishes responsibilities for achieving and verifying

quality in company activities, including engineering evaluations such as the one reflected in this

document. American Society of Mechanical Engineers NQA-1, Quality Assurance Requirements for

Nuclear Facilities (ASME 2000) and other consensus quality standards were used in the

development of the WSMS quality assurance program.

The primary applicable guidance at the contractor level appears in WV-929, (WVNSCO 2007).

This procedure outlines the process of identifying waste streams that may be managed in

accordance with DOE requirements for LLW if the requirements for waste incidental to reprocessing

are met. It incorporates applicable DOE requirements for waste incidental to reprocessing

50 This company is now known as URS Safety Management Solutions.

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determinations, such as those in DOE Manual 435.1-1. Guidance in WV-929 was considered and

applied as appropriate in preparation of this evaluation.

4.0 Data Evaluation and Engineering Calculations

This draft evaluation did not entail collecting data by field measurements or by obtaining

samples and analyzing them in a laboratory. Rather, it utilized existing data from prior field

measurements and sample analyses. All calculations Were performed by other contractor staff with

two exceptions: the sum-of-fraction calculation (Table 6-1) and the PE-g estimate.

To assure quality, these engineering calculations were prepared, reviewed, and approved in

accordance with the following process:

* Calculations were formally documented by compiling all relevant information into a

calculation package, which included compilations of the calculations and all related

information on which they were based. This calculation package describes input data and

their sources, software used, assumptions made, and calculation results. Sufficient detail is

included to allow a reviewer to independently duplicate the results.

* This calculation package was reviewed by a technically competent person who had not

been involved in its development to independently check the calculations and their

accuracy. This check included verifying matters such as (1) use of appropriate data

sources, (2) correct utilization of input data, (3) use of appropriate methods, (4) use of

valid assumptions, (5) accuracy of results by representative independent calculations, and

(6) reasonableness of the results in considering the range of expectations.

* This calculation package was also reviewed and approved by the WSMS Project Manager,

who verified that it had received a thorough technical review and that review comments

were appropriately resolved.

. This calculation package includes a coversheet signed by the preparer, the reviewer, and

the WSMS Project Manager signifying that it was prepared, reviewed, and approved using

this basic process.

5.0 Draft Evaluation Preparation and Review

This draft evaluation was prepared by the WSMS team described below in Section 6.

Draft Evaluation Preparation

The WSMS Project Manager assigned particular team members to assume the lead on

individual parts of the draft evaluation. These parts were prepared in accordance with requirements

and guidance described above, factoring in the results of supporting engineering calculations and

results, as applicable.

Appropriate precedents were considered. These included waste determinations prepared by the

DOE pursuant to the Section 3116 of the Ronald W. Reagan National Defense Authorization Act for

Fiscal Year 2005, even though that act does not apply to the WVDP (DOE 2006a and DOE 2006b).

As individual parts of the document were completed, they were integrated into a single

document to form the complete draft evaluation, and the draft evaluation was then edited for

consistency.

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Draft Evaluation Review

As individual parts of the draft evaluation were completed, they received one or more peer

reviews by members of the WSMS team. Review comments were incorporated or otherwise

resolved.

Th e complete draft evaluation received additional review by the WSMS team. This detailed

review addressed adequacy, completeness, correctness, and compliance with applicable

requirements. Review comments were incorporated or resolved.

The completed draft evaluation was reviewed by the WVES Project Manager and WVES

technical specialists, as assigned, to ensure that it met all DOE requirements and was technically

accurate and complete. Comments from this review were incorporated before the draft evaluation

was submitted to DOE.

Draft Evaluation Approval

DOE reviewed the draft evaluation and approved it after written resolution of the DOE

comments and incorporation of those comments.

6.0 Personnel Qualification

Members of the WSMS team dedicated to the evaluation and preparation of this draft

evaluation report are experienced regulatory analysts. They brought to this project extensive

knowledge and experience in radioactive waste management and related technical standards such

as:

0 DOE Order 435.1, Radioactive Waste Management,

0 DOE Manual 435.1-1, Radioactive Waste Management Manual,

0 DOE Guide 435. 1-1, Implementation Guide for Use With DOEM 435.1-1,

* 10 CFR Part 61, Licensing Requirements for Land Disposal of Radioactive Waste, and

* NUREG-0782, Draft Environmental Impact Statement on 10 CFR Part 61 Licensing

Requirements for Land Disposal of Radioactive Waste (NRC 1981).

Team members included engineers and scientists who are W.SMS employees, along with an

independent consultant. All are knowledgeable of the Vitrification Melter and have previous

experience with waste incidental to reprocessing determinations.

7.0 Oversight and Independent Review

A key element in implementation of the quality standards and procedures outlined above is that

individuals are responsible for their own work. WSMS management also provided direction and

oversight to ensure that these standards and procedures were closely followed, and regularly

assessed compliance with quality requirements. In addition, WVES management and DOE

personnel in the WVDP Field Office provided oversight and review to this end.

8.0 Document and Record Control

Documents and records associated with this draft evaluation are controlled in accordance with

the WVDP Records Management System (WVES 2008). This system is used to ensure that records

important to safety and quality are generated, reviewed, approved, collected, and maintained so

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they accurately reflect completed work and facility conditions, and comply with applicable statutory

or contractual requirements.

The WVDP Records Management System incorporates the requirements of DOE Order 200.1,

Information Management Program, and DOE Order 414.1B, Quality Assurance. Schedules for

records retention and disposition comply with the General Records Schedule of the National

Archives and Records Administration and other approved records schedules.

The Records Management procedures include instructions for retention, protection,

preservation, changes, traceability, accountability, and retrievability of records. They also provide

controls to ensure records are legible, accurate, complete, retrievable, and validated by authorized

personnel. Records are stored and maintained to minimize the risk of damage, larceny, vandalism,

or deterioration. Active records are not sent to records holding facilities, but are stored in a facility

where the records may be readily accessed.

9.0 References

Federal Statutes

Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005, Public Law 108-375,

118 Stat. 1811. 108'h United States Congress, October 28, 2004.

Code of Federal Regulations and Federal Register Notices

10 CFR 830, Subpart A, Quality Assurance Requirements.

DOE Orders, Policies, Manuals, and Standards

DOE Order 414.1B, Quality Assurance. U. S. Department of Energy, Washington, D.C., April 2004.

DOE Order 435.1, Radioactive Waste Management, Change 1. U. S. Department of Energy,

Washington, D.C., August 28, 2001.

DOE Manual 435.1-1, Radioactive Waste Management Manual, Change 1. U. S. Department of

Energy, Washington, D.C., June 19, 2001.

DOE G 435.1-1, Implementation Guide for Use with DOE M-435.1-1, U.S. Department of Energy,

Washington, D.C., July 1999.

Other References

ASME 2000, Quality Assurance Requirements For Nuclear Facilities, NQA-1. American Society of

Mechanical Engineers, New York, 2000.

DOE 2006a, Basis for Section 3116 Determination, Salt Waste Disposal, Savannah River Site, DOEWD-2005-001. U.S. Department of Energy - Savannah River, Aiken, South Carolina,

January 2006.

DOE 2006b, Basis for Section 3116 Determination for the Idaho Technology and Engineering

Center Tank Farm Facility, DOE/NE-ID-11226, Revision 0. U.S. Department of Energy -

Idaho National Laboratory, Idaho Falls, Idaho, November 2006.

NRC 1981, Draft Environmental Impact Statement on 10 CFR Part 61 Licensing Requirements for

Land Disposal of Radioactive Waste, NUREG-0782. U.S. Nuclear Regulatory Commission,

Washington, D.C., 1981.

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WSMS 2009, Washington Safety Management Solutions LLC Quality Assurance Plan, WSMS QA 100,

Revision 1. Washington Safety Management Solutions, Aiken, South Carolina, August 10,

2009.

WVES 2008, WVNSCO Manual for Records Management and Storage. WVDP-262, Revision 9. West

Valley Environmental Services LLC, West Valley, New York, September 4, 2008.

WVNSCO 2007, Waste Incidental to Reprocessing Determination, WV-929, Revision 7. West Valley

Nuclear Services Company, West Valley, New York, March 27, 2007.

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APPENDIX C

Comparability of DOE, NRC and Texas Requirements for LLW Disposal

Appendix Purpose

The purpose of this appendix is to show that Department of Energy, Nuclear

Regulatory Commission, and State of Texas requirements for disposal of low-level

waste are comparable.

Appendix Content

This appendix identifies applicable Department of Energy performance objectives

and the similar Nuclear Regulatory Commission and State of Texas performance

objectives and discusses their comparability.

Key Points

" Requirements for low-level waste disposal are embodied in sets of pefformance

objectives for the waste disposal facility.

" The Department of Energy performance objectives are described in DOE Manual

435.1-1, Radioactive Waste Management.

* The Nuclear Regulatory Commission performance objectives are described in

Subpart C, Performance Objectives, of 10 CFR Part 61,, Licensing Requirements

for Land Disposal of Radioactive Waste.

" The performance objectives in the Texas Administrative Code that apply to the

WCS low-level waste disposal facility - which are included in the Title 30, Part 1,

Chapter 336, Subchapter H, Rule §336.723-727 - mirror the Nuclear Regulatory

Commission performance objectives.

• Department of Energy, Nuclear Regulatory Commission, and State of Texas

performance objectives for low-level waste disposal are comparable.

1.0 Introduction

This appendix identifies performance objectives for disposal of LLW by the DOE, the NRC, and

the State of Texas. It then compares these performance objectives. As noted previously, the

performance objectives in the State of Texas regulations mirror the NRC performance objectives at

10 CFR 61, Part C, i.e., they are essentially identical except for the use of difference section

numbers.

Information in this appendix is based in part on previous detailed comparison studies of DOE

and NRC performance objectives for LLW disposal (Cole, et al. 1995 and Wilhite 2001).

2.0 Applicable Performance Objectives

DOE Manual 435.1-1, Radioactive Waste Management, describes DOE requirements for

disposal of LLW. The comparable NRC requirements appear in Subpart C of 10 CFR Part 61, which

lists one general requirement and four performance objectives, which are reproduced below.

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Section 61.40, General Requirement

"Land disposal facilities must be sited, designed, operated, closed, and controlled after

closure so that reasonable assurance exists that exposures to humans are within the limits

established in the performance objectives in Sections 61.41 through 61.44."

Section 61.41, Protection of the General Population from Releases of Radioactivity

"Concentrations of radioactive material which may be released to the general environment

in ground water, surface water, air, soil, plants, or animals must not result in an annual

dose exceeding an equivalent of 25 millirem to the whole body, 75 millirem to the thyroid,

and 25 millirem to any other organ of any member of the public. Reasonable effort should

be made to maintain releases-of radioactivity in effluents to the general environment as

low as is reasonably achievable."

Section 61.42, Protection of Individuals from Inadvertent Intrusion

"Design, operation, and closure of the land disposal facility must ensure protection of any

individual inadvertently intruding into the disposal site and occupying the site or contacting

the waste at any time after active institutional controls over the disposal site are removed."

Section 61.43, Protection of Individuals During Operations

"Operations at the land disposal facility must be conducted in compliance with the

standards for radiation protection set out in Part 20 of this chapter, except for releases of

radioactivity in effluents from the land disposal facility, which shall be governed by Section

61.41 of this part. Every reasonable effort shall be made to maintain radiation exposures as

low as is reasonably achievable."

Section 61.44, Stability of the Disposal Site After Closure

"The disposal facility must be sited, designed, used, operated, and closed to achieve longterm stability of the disposal site and to eliminate to the extent practicable the need for

ongoing active maintenance of the disposal site following closure so that only surveillance,

monitoring, or minor custodial care are required."

The State of Texas requirements for LLW disposal at .Texas Administrative Code, Title 30,

Part 1, Chapter 336, Subchapter H, Rule §336.723-777 are based on the NRC requirements at

Subpart C of 10 CFR Part 61 and are exactly the same except for minor wording differences

identified below.

3.0 Comparability of the General Requirements

3.1 DOE

The general requirement in DOE Manual 435.1-1, Section IV.P(1), is expressed as follows:

"Low-level waste disposal facilities shall be sited, designed, operated, maintained, and closed

so that a reasonable expectation exists that the following performance objectives will be met

for waste disposed of after September 26, 1988."

3.2 NRC

The NRC regulations in 10 CFR 61.40 provide in relevant part:

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"Land disposal facilities must be sited, designed, operated, closed, and controlled after closure

so that reasonable assurance exists that exposures to humans are within the limits established

in the performance objectives in Sections 61.41 through 61.44."

3.3 State of Texas

The State of Texas regulations (Rule §336.723) mirror the NRC regulations in 10 CFR 61.40.

3.4 Discussion

The statement of NRC requirements in 10 CFR 61.40 is nearly identical to that of the DOE

general requirement. The DOE requirement adds the concept of maintenance, which is imp!icit in

the NRC requirement. The DOE requirement does not mention control after closure, but this

concept is embodied in the DOE requirements for closure, specifically DOE Manual 435.1, Section

IV.Q (2)(c), which requires DOE control until it can be shown that release of the disposal site for

unrestricted use will not compromise DOE requirements for radiological protection of the public.

The DOE general requirement for LLW disposal, the NRC general requirement of 10 CFR 61.40,

and the State of Texas general requirement are therefore comparable.

4.0 Comparability Regarding Protection of the General Population from Releases of

Radioactivity

4.1 DOE

DOE requirements of DOE Manual 435.1-1, Section IV.P(1), read as follows:

"(a) Dose to representative members of the public shall not exceed 25.millirem in a year total

effective dose equivalent from all exposure pathways, excluding the dose from radon and

its progeny in air.

(b) Dose to representative members of the public via the air pathway shall not exceed 10

millirem in a year total effective dose equivalent, excluding the dose from radon and its

progeny.

(c) Release of radon shall be less than an average flux of 20 pCi/m 2

/s at the surface of the

disposal facility. Alternatively, a limit of 0.5 pCi/L of air may be applied at-the boundary of

the facility."

4.2 NRC

NRC regulations in 10 CFR 61.41 are expressed as follows:

"Concentrations of radioactive material which may be released to the general environment

in ground water, surface water, air, soil,. plants, or animals must not result in an annual

dose exceeding an equivalent of 25 millirem to the whole body, 75 millirem to the thyroid,

and 25 millirem to any other organ of any member of the public. Reasonable effort should

be .made to maintain releases of radioactivity in effluents to the general environment as

low as is reasonably achievable."

4.3 State of Texas

The State of Texas regulations (Rule §336.724) mirror the NRC regulations in 10 CFR 61.41

with two minor wording differences. The Texas rule uses the phrase "annual dose above

background" instead of "annual dose." In the second sentence, the Texas rule uses the phrase

"Effort shall be made" instead of "Reasonable effort should be made."

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4.4 Discussion

DOE uses more current radiation protection methodology, consistent with that used in NRC's

radiation protection standards in NRC's 10 CFR 20, Standards for Protection Against Radiation.

Because NRC has not revised 10 CFR 61.41 to reflect the more current methodology in 10 CFR 20,

DOE's requirements and those in 10 CFR 20 differ slightly from those in 10 CFR 61.41. However,

the resulting allowable doses are comparable, as NRC has acknowledged (NRC 2005). NRC has

indicated that it expects DOE to use the newer methodology in 10 CFR 20 and DOE Manual 435.1-1

for the WVDP decommissioning (NRC 2002). Both NRC and DOE use a performance assessment to

assess whether the dose limit will be met.

The DOE requirements go beyond this NRC performance objective by specifying an assessment

of the impacts of LLW disposal on water resources (i.e., DOE Manual 435.1, Section IV.P (2)(g)).

The NRC requirement includes maintaining releases to the environment ALARA. Although this

requirement is not included in the DOE performance objective, it is included in the performance

assessment requirements (i.e., DOE Manual 435.1-1, Section IV.P (2)(f)).

Because the State of Texas regulations are essentially the same as the NRC regulations, the

conclusions about the comparability of the DOE and NRC requirements also apply to the

comparability of the State of Texas requirements.

5.0 Comparability Regarding Protection of Individuals from Inadvertent Intrusion

5.1 DOE

DOE requirements of DOE Manual 435.1-1, Section IV.P(2)(h), for protection of individuals

from inadvertent intrusion read as follows:

• "For purposes of establishing limits on the concentration of radionuclides that may be disposed

of near-surface, the performance assessment shall include an assessment of impacts calculated

for a hypothetical person assumed to inadvertently intrude for a temporary period into the lowlevel waste disposal facility. For intruder analyses, institutional controls shall be assumed to be

effective in deterring intrusion for at least 100 years following closure. The intruder analyses

shall use performance measures for chronic and acute exposure scenarios, respectively, of 100

millirem (1 mSv) in a year and 500 millirem (5 mSv) total effective dose equivalent excluding

radon in air."

5.2 NRC

NRC requirements of 10 CFR 61.42 are expressed as follows:

"Design, operation, and closure of the land disposal facility must ensure protection of any

individual inadvertently intruding into the disposal site and occupying the site or contacting

the waste at any time after active institutional controls over the disposal site are removed."

5.3 State of Texas

The State of Texas regulations (Rule §336.725) mirror the NRC regulations in 10 CFR

61.42.

5.4 Discussion

The DOE LLW disposal requirement that the performance assessment include an assessment of

the impacts on a person inadvertently intruding into the disposal facility is more stringent than the

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NRC requirement. The NRC waste classification system is based on intruder calculations using a

500 millirem per year dose limit (NRC 1982). The DOE requirement uses a 100 millirem per year

limit for chronic exposures and a 500 millirem limit for acute exposures.

Because the State of Texas regulations mirror the NRC regulations, the conclusions about the

comparability of the DOE and NRC requirements also apply to the comparability of the State of

Texas requirements.

6.0 Comparability Regarding Protection of Individuals During Operations

6.1 DOE

The DOE requirements in DOE Manual 435.1-1, Section I.E(13), for protection of individual

during operations read as follows:

"Radioactive waste management facilities, operations, and activities shall meet the

requirements of 10 CFR Part 835, Occupational Radiation Protection, and DOE 5400.5,

Radiation Protection of the Public and the Environment."

6.2 NRC

The NRC requirements of 10 CFR 61.43 are expressed as follows:

"Operations at the land disposal facility must be conducted in compliance with the

standards for radiation protection set .out in Part 20 of this chapter, except for releases of

radioactivity in effluents from the land disposal facility, which shall be governed by Section

61.41 of this part. Every reasonable effort shall be made to maintain radiation exposures as

low as is reasonably achievable."

6.3 State of Texas

The State of Texas regulations (Rule §336.726) mirror the NRC regulations in 10 CFR

61.43.

6.4 Discussion

. The ALARA concept is an integral part of DOE radiation and environmental protection

programs, as expressed in DOE Policy 441.1, Department of Energy Radiological Health and Safety

Policy DOE requirements for occupational radiological protection are addressed in 10 CFR 835, and

similar requirements for radiological protection of the public and the environment are addressed in

DOE Order 5400.5. The NRC 10 CFR 61.43 requirement references 10 CFR 20, Standards for

Protection Against Radiation, which contains similar radiological protection standards for workers

and the public.

Appendix D provides additional information on the comparability of DOE and NRC radiation

dose standards that apply to protection of individuals during operations. The State of Texas

radiation dose standards mirror the NRC dose standards as explained in Appendix D.

7.0 Comparability Regarding Stability of the Disposal Site After Closure

7.1 DOE

The DOE requirements of DOE Manual 435.1-1, Sections IV.Q(1)(a) and (b) and IV.Q(2)(c), for

stability of the disposal site after closure are expressed as follows:

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"Disposal Site Stability (DOE Manual 435.1. Section IV.Q(1)(a) and (b)). A preliminary closure

plan shall be developed and submitted to Headquarters for review with the performance

assessment and composite analysis. The closure plan shall be updated following issuance of

the disposal authorization statement to incorporate conditions specified in the disposal

authorization statement. Closure plans shall:

(a) Be updated as required during the operational life of the facility.

(b) Include a description of how the disposal facility will be closed to achieve long-term

stability and minimize the need for active maintenance following closure and to ensure

compliance with the requirements of DOE 5400.5, Radiation Protection of the Public

and the Environment"

"Disposal Facility Closure (DOE Manual 435.1, Section IV.O(2)( c). Institutional control

measures shall be integrated into land use and stewardship plans and programs, and shall

continue until the facility can be released pursuant to DOE Order 5400.5, Radiation Protection

of the Public and the Environment"

7.2 NRC

The NRC requirements of 10 CFR 61.44 state that:

"The disposal facility must be sited, designed, used, operated, and closed to achieve longterm stability of the disposal site and to eliminate to the extent practicable the need for

ongoing active maintenance of the disposal site following closure so that only surveillance,

• monitoring, or minor custodial care are required."

7.3 State of Texas

The State of Texas regulations (Rule §336.727) mirror as the NRC regulations in 10 CFR

61.43.

7.3 Discussion

The DOE LLW disposal requirements address long-term stability of the site by requiring a

description of how closure will achieve stability in the closure plan, and by a description of how

closure will minimize the need for active maintenance following closure (DOE Manual 435.1, Section

IV.Q (1)(b)). Additionally, one of the performance assessment requirements (DOE Manual 435.1,

Section IV.P (2)(c)) states: "Performance assessments shall address reasonably foreseeable natural

processes that might disrupt barriers against release and transport of radioactive materials." Thus,

the performance assessment will include a projection of the long-term stability of the site,

considering reasonably foreseeable natural processes such as erosion, degradation of waste

packages, etc.

8.0 References

Code of Federal Regulations

10 CFR 20, Standards for Protection Against Radiation.

10 CFR 61, Subpart C, Licensing Requirements for Land Disposal of Radioactive Waste, Performance Objectives.

10 CFR 835, Occupational Radiation Protection.

DOE Orders, Policies, and Manuals

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DOE Order 5400.5, Radiation Protection of the Public and the Environment, Change 2. U.S.

Department of Energy, Washington, D.C., January 7, 1993.

DOE Policy 441.1, Department of Energy Radiological Health and Safety Policy. U.S. Department of

Energy, Washington, D.C., April 26, 1996.

DOE Manual 435.1-1, Radioactive Waste Management Manual, Change 1. U. S. Department of

Energy, Washington, D.C., June 19, 2001.

State Regulations

Texas Administrative Code, Title 30, Part 1, Chapter 336, Radioactive Substance Rules (http://

info.sos.state.tx. us/pls/pub/readtac$ext.ViewTAC?tacview=4&ti =30&pt= 1&ch =336)

Other References

Cole, et al. 1995, Comparison of Selected DOE and Non-DOE Requirements, Standards, and

Practices for Low-Level Radioactive Waste Disposal, DOE/LLW-225, Revision 0. Cole, L., D.

Kudera, and W. Newberry, Idaho National Engineering Laboratory, Lockheed Idaho

technologies Company, Idaho Fall, Idaho, December 1995.

NRC 1982, Final Environmental Impact Statement on 10 CFR Part 61 "Licensing Requirements for

Land Disposal of Radioactive Waste, "Summary and Main Report, NUREG-0945, Volume 1.

U.S. Nuclear Regulatory Commission, Washington, D.C., November 1982.

NRC 2002, Decommissioning Criteria for the West Valley Demonstration Project (M-32) at the West

Valley Site; Final Policy Statement. U.S. Nuclear Regulatory Commission, Washington, D.C.,

Federal Register, Vol. 67, No. 22, February 1, 2002.

NRC 2005, U.S. Nuclear Regulatory Commission Technical Evaluation Report for the U.S.

Department of Energy Savannah River Site Draft Section 3116 Waste Determination for

Salt Waste Disposal. U.S. Nuclear Regulatory Commission, Washington, D.C., December

2005.

Wilhite 2001, Comparison of LLW Disposal Performance Objectives, 10 CFR 61 and DOE 435.1,

WSRC-RP-2001-00341. Wilhite, E.L., Westinghouse Savannah River Company, Aiken, South

Carolina, 2001.

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APPENDIX D

Comparability of DOE, NRC, and Texas Dose Standards

Appendix Purpose

The purpose of this appendix is to compare Department of Energy, Nuclear

Regulatory Commission, and State of Texas radiation dose standards that apply to

individual workers and to members of the public.

Appendix Content

This appendix identifies applicable Department of Energy dose standards and the

similar Nuclear Regulatory Commission and State of Texas dose standards and

discusses their comparability.

Key Points

* The Department of Energy radiation dose standards appear in 10 CFR 835,

Occupational Radiation Protection, and in DOE Orders.

* The Nuclear Regulatory Commission radiation dose standards appear in 10 CFR

20, Standards for Protection Against Radiation.

" Department of Energy and Nuclear Regulatory Commission radiation dose

standards are comparable.

" The State of Texas dose standards that apply to the WCS low-level waste

disposal facility - which are included in the Texas Administrative Code, Title. 30,

Part 1, Chapter 336, Subchapter D - mirrors the Nuclear Regulatory Commission

dose standards.

1.0 Introduction

The purpose of this appendix is to compare the DOE, NRC, and State of Texas dose standards

that apply to protection of the public and the workers from radiation during operations associated

with preparing the Vitrification Melter for shipment at the WVDP and handling of the Melter when it

is received at either the Nevada National Security Site or the WCS LLW disposal facility in Texas for

disposal, assuming that it will be sent to one of those facilities.

Section 5.2.4 of the body of this draft evaluation briefly addressed protection of individuals

during these operations at the WVDP, the Nevada National Security Site, and the WCS LLW disposal

facility. Appendix C also addressed this matter2 This appendix provides a more detailed treatment of

the dose standards used.

Requirements in NRC's regulations at 10 CFR 61.43 state:

"[O]perations at the land disposal facility must be conducted in compliance with

the standards for radiation protection set out in part 20 of this chapter [10 CFR], except

for releases of radioactivity in effluents from the land disposal facility, which shall be

governed by §61.41 of this part. Every reasonable effort shall be made to maintain

radiation exposures as low as is reasonably achievable."

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This requirement references 10 CFR 20, Standards for Protection Against Radiation, which

contains radiological protection standards for workers and the public. The DOE requirements for

occupational radiological protection are provided in 10 CFR 835, Occupational Radiation Protection,

and those for radiological protection of the public and the environment are provided in DOE Order

5400.5, Radiation Protection of the Public and the Environment. The State of Texas radiation

protection standards appear in the Texas Administrative Code, Title 30, Part 1, Chapter 336,

Subchapter D.

The NRC standards for radiation protection in 10 CFR 20 that are considered in detail in this

draft evaluation are the dose limits for the public and the workers during disposal operations

set forth in 10 CFR 20.1101(d), 20.1201(a)(1)(i), 20.120 1(a)(1)(ii), 20.120 1(a)(2)(i),

20.120 1(a)(2)(ii), 20.1201(e), 20.1208(a), 20. 1301(a)(1), 20.1301(a)(2), and 20.1301(b)."'

These NRC dose limits correspond to the DOE dose limits in 10 CFR 835 and relevant DOE orders

that establish DOE regulatory and contractual requirements for DOE facilities and activities. As

discussed in Section 5.2.4 of this draft evaluation, operations related to disposal of the Vitrification

Melter will meet these dose limits and doses will be maintained ALARA. As explained below, the

State of Texas radiation protection standards mirror the NRC radiation protection standards.

2.0 Air Emissions Limit for Individual Member of the Public

2.1 DOE

DOE limits doses from air emissions to the public to 10 mrem/y in DOE Order 5400.5 (1993).

The DOE is also subject to and complies with the U.S. Environmental Protection Agency requirement in

40 CFR 61.92, which has the same limit.5 2

2.2 NRC

The NRC regulation in 10 CFR 20.1101(d) provides in relevant part:

"[A] constraint on air emissions of radioactive material to the environment, excluding

radon-222 and its daughters, shall be established ... such that the individual member of

the public likely to receive the highest dose will not be expected to receive a total

effective dose equivalent in excess of 10 mrem (0.1 mSv)/y from these emissions."

5' The "standards for radiation protection" in 10 CFR 20 (as cross-referenced in the performance objective in 10

CFR 61.43), which are relevant to this draft evaluation, are the dose limits for radiation protection of the public and

the workers during disposal operations, and not those which address general licensing, administrative, programmatic,

or enforcement matters administered by NRC for NRC licensees. Accordingly, this draft evaluation addresses in detail

the radiation dose limits for the public and the workers during disposal operations that are contained in the

provisions of 10 CFR 20 referenced above. Although 10 CFR 20.1206(e) contains limits for planned special exposures

for adult workers, there will not be any such planned special exposures for work related to the Vitrification Melter.

Therefore, this limit is not discussed further in this draft evaluation. Likewise, 10 CFR 20.1207 specifies

occupational dose limits for minors. However, there will not be minors working at the WVDP or the Nevada National

Security Site who would receive an occupational dose. Therefore, this limit is not discussed further in this draft

evaluation.

52 40 CFR 61.92 provides as follows: "Emissions of radionudides to the ambient air from DOE facilities shall not exceed

those amounts that would cause any member of the public to receive in any year an effective dose equivalent of 10

mrem/y. It is assumed that the individual is an adult living at the site perimeter that is exposed to the maximum yearly

radioactive atmospheric release and maximum radiation concentration in food for 365 days per year. For the airborne

pathway,. the dose is developed by the input of atmospheric release data, vegetation consumption data, milk

consumption data, and beef consumption data."

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2.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.304

mirrors the NRC regulation.

2.4 Discussion

The DOE, NRC, and State of Texas requirements are comparable.

3.0 Total Effective Dose Equivalent Limit for Adult Workers

3.1 DOE

DOE's regulation in 10 CFR 835.202(a)(1) requires that the occupational dose per year for

general employees shall not exceed a total effective dose equivalent of 5 remi5 3.

3.2 NRC

The NRC regulation in 10 CFR 20.1201(a), concerning occupational dose limits for adults, provides

in relevant part:

"(a) [C]ontrol the occupational dose to individual adults, except for planned special

exposures...to the following dose limits.

(1) An annual limit, which is the more limiting of -

(i) The total effective dose equivalent being equal to 5 reins (0.05 Sv)."

3.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.305

is mirrors the NRC regulation.

3.4 Discussion

The DOE, NRC, and State of Texas requirements are comparable.

4.0 Any Individual Organ or Tissue Dose Limit for Adult Workers

4.1 DOE

The DOE regulation in 10 CFR 835.202(a)(2) provides in relevant part:

"... the occupational dose received by general employees shall be controlled such that

the following limits are not exceeded in a year:

(2) The sum of the deep dose equivalent for external exposures and the committed

dose equivalent to any organ or tissue other than the lens of the eye of 50 rems

(0.5 sievert)"

53 The DOE's regulations at 10 CFR 835.202(a)(1) and (a)(2) require that the occupational dose per year for general

employees shall not exceed both a total effective dose equivalent of 5 rem and the sum of the deep-dose equivalent

for external exposures and the committed dose equivalent to any other organ or tissue other than the lens of the eye

of 50 rem. The NRC's regulation specifies that either of these two limits shall be met by NRC licensees, whichever is

more limiting. Thus, DOE imposes stricter, separate requirements. The provisions of DOE's requirements at 10 CFR

835.202(a)(1) and (a)(2), which correlate to NRC requirements at 10 CFR 20.1201(a)(1) and (a)(2), are discussed

in separate subsections in this draft evaluation.

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4.2 NRC

The NRC regulation in 10 CFR 20.1201(a), concerning occupational dose limits for adults, provides

in relevant part:

"(a) [C]ontrol the occupational dose to individual adults, except for planned special

exposures ... to the following dose limits.

(1) An annual limit, which is the more limiting of

(ii) The sum of the deep-dose equivalent and the committed dose equivalent to

any individual organ or tissue other than the lens of the eye being equal to

50 rems (0.5 Sv)."

4.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.305

mirrors the NRC regulation.

4.4 Discussion

The DOE, NRC, and State of Texas requirements are comparable.

5.0 Annual Dose Limit to the Lens of the Eye for Adult Workers

5.1 DOE

The DOE regulation in 10 CFR 835.202(a)(3) provides in relevant part:

the occupational dose received by general employees shall be controlled such that

the following limits are not exceeded in a year:

(3) A lens of the eye dose equivalent of 15 rems (0.15 sievert)"

5.2 NRC

The NRC regulation in 10 CFR 20.1201(a), concerning occupational dose limits for adults, provides

in relevant part:

(a) [C]ontrol the occupational dose to individual adults, except for

planned special exposure the following dose limits.

(2) The annual limits to the lens of the eye, to the skin of the whole

body or to the skin of the extremities, which are:

(i) A lens dose equivalent of 15 reins (0.15 Sv).

5.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.305

mirrors the NRC regulation.

5.4 Discussion

The DOE, NRC and State of Texas requirements are comparable.

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6.0 Annual Dose Limit to the Skin of the Whole Body and to the Skin of the

Extremities for Adult Workers

6.1 DOE

The DOE regulation in 10 CFR 835.202(a)(4) provides in relevant part:

the occupational dose received by general employees shall be controlled such that

the following limits are not exceeded in a year:

(4) A shallow dose equivalent of 50 rems (0.5 sievert) to the skin or any extremity."

6.2 NRC

The NRC regulation in 10 CFR 20.1201(a), concerning occupational dose limits for adults,

provides in relevant part:

(a) [C]ontrol the occupational dose to individual adults, except for

planned special exposures...to the following dose limits.

(2) The annual limits to the lens of the eye, the skin of the whole

body, or to the skin of the extremities, which are: ...

(ii) A shallow-dose equivalent of 50 rem (0.5 Sv) to the skin of the

whole body or to the skin of any extremity.

6.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.305

mirrors the NRC regulation.

6.4 Discussion

The DOE, NRC, and State of Texas requirements are comparable.

7.0 Limit on Soluble Uranium Intake

7.1 DOE

Requirements in DOE Order 440.1A for soluble uranium intake are the more restrictive of the

concentrations in the American Conference of Governmental Industrial Hygienists threshold limit

values (0.2 mg/m 3, which is the same as noted in 10 CFR 20, Appendix B) or the Occupational

Safety and Health Administration permissible exposure limit (0.05 mg/m 3). The permissible

exposure limit for soluble uranium, which equates to a soluble uranium intake of 2.4

mg/week, is the more restrictive of the two.

7.2 NRC

The NRC regulation in 10 CFR 20.1201(e), concerning occupational dose limits for adults,

provides in relevant part: "in addition to the annual dose limits, ... limit the soluble uranium intake by

an individual to 10 milligrams in a week in consideration of chemical toxicity."

7.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.305

mirrors the NRC regulation.

7.4 Discussion

The DOE requirements are more restrictive.

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8.0 Dose Equivalent to an Embryo/Fetus

8.1 DOE

The DOE regulation in 10 CFR 835.206(a) provides in relevant part:

"The dose equivalent limit for the embryo/fetus from the period of conception to birth, as a

result of occupational exposure of a declared pregnant worker, is 0.5 rem (0.005 sievert)."

After declaration of pregnancy, DOE provides the option of a mutually agreeable assignment of

work tasks, without loss of pay or promotional opportunity, such that further occupational radiation

exposure during the remainder of the gestation period is unlikely. In addition, personnel dosimetrys

is provided and used to track exposure carefully.

8.2 NRC

The NRC regulation in 10 CFR 20.1208(a), concerning the dose equivalent to an embryo/fetus,

provides in relevant part:

"ensure that the dose equivalent to the embryo/fetus during the entire pregnancy, due to

the occupational exposure of a declared pregnant woman, does not exceed 0.5 rem (5

mSv)."

8.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.312

mirrors the NRC regulation.

8.4 Discussion

The DOE, NRC, and State of Texas requirements are comparable.

9.0 Dose Limits for Individual Members of the Public (Total Annual Dose)

9.1 DOE

Provisions in DOE Order 5400.5 limit public doses to 0.1 rem per year.

9.2 NRC

The NRC regulation in 10 CFR 20.1301(a), concerning dose limits for individual members of the

public, provides in relevant part:

(a) "[C]onduct operations so that -

(1) The total effective dose equivalent to individual members of the public ...does not

exceed 0.1 rem (1 mSv) in a year, exclusive of the dose contributions from background

radiation, from any medical administration the individual has received, from exposure

to individuals administered radioactive material and released..., from voluntary

participation in medical research programs, and from the ... disposal of radioactive

material into sanitary sewerage."

'The term dosimetryor personnel dosimetry refers to a device carried or worn by an individual working

near radiation for measuring the amount of radiation to which he or she is exposed.

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9.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.313

mirrors the NRC regulation.

9.4 Discussion

The DOE, NRC, and State of Texas requirements are comparable.

10.0 Dose Limits for Individual Members of the Public (Dose Rate in Unrestricted Areas)

10.1 DOE

DOE's regulation in 10 CFR 835.602 establishes the expectation that the total effective dose

equivalent in controlled areas will be less than 0.1 rem per year. In accordance with 10 CFR

835.602, radioactive material areas have been established for accumulations of radioactive material

within controlled areas that could result in a radiation dose of 100 millirem per year or greater.

Averaged over a work year, this yields a constant average dose rate of 0.00005 rem per hour. In

addition, training and dosimetry are required for individual members of the public for entry into

controlled areas, as well as signs at each access point to a controlled area.

10.2 NRC

The NRC regulation in 10 CFR 20.1301(a), concerning dose limits for individual members of the

public, provides in relevant part:

(a) "[C]onduct operations so that -

(1) The total effective dose equivalent to individual members of the public from the

licensed operation does not exceed 0.1 rem (1 mSv) in a year, exclusive of the

dose contributions from background radiation, from any administration the

individual has received, from exposure to individuals administered radioactive

material and released under § 35.75, from voluntary participation in medical

research programs, and from the licensee's disposal of radioactive material into

sanitary sewerage in accordance with § 20.2003, and

(2) The dose in any unrestricted area from external sources, exclusive of the dose

contributions from patients administered radioactive material and released in

accordance with § 35.75, does not exceed 0.002 rem (0.02 millisievert) in any

one hour.

(b) If the licensee permits members of the public to have access to controlled areas,

• the limits for members of the public continue to apply to those individuals."

10.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.313

mirrors the NRC regulation.

10.4 Discussion

The DOE, NRC and State of Texas requirements are comparable.

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11.0 Dose Limits for Individual Members of the Public With Access to Controlled Areas 55

11.1 DOE

The DOE regulation in 10 CFR 835.208 provides:

"The total effective dose equivalent limit for members of the public exposed to radiation

and/or radioactive material during access to a controlled area is 0.1 rem (0.001 sievert) in

a year.-

DOE requires training for individual members of the public before entry into controlled areas.

In addition, to ensure no member of the publicexceeds radiation exposure limits, use of dosimetry

is required if a member of the public is expected to enter a controlled area and receive a dose that

may exceed 0.05 rem in a year

11.2 NRC

The NRC regulation in 10 CFR 20.1301(b), concerning dose limits for individual members of the

public, provides in relevant part:

"if ... members of the public [are permitted] to have access to controlled areas, the limits for

members of the public [0.1 rem (1 mSv)] continue to apply to those individuals."5 6

11.3 State of Texas

The State of Texas regulation in the Texas Administrative Code, Title 30, Part 1, Rule §336.313

mirrors the NRC regulation.

11.4 Discussion

The DOE, NRC, and State of Texas requirements in this area are comparable.

12.0 As Low As Reasonably Achievable

12.1 DOE

The DOE regulation in 10 CFR 835.2 defines ALARA as "the approach to radiation protection to

manage and control exposures (both individual and collective) to the work force and to the general

public to as low as is reasonable, taking into account social, technical, economic, practical, and

public policy considerations." The DOE regulation in 10 CFR 835.2 also specifies: "ALARA is not

a dose limit but a process which has the objective of attaining doses as far below the applicable

limits as is reasonably achievable."

12.2 NRC

The NRC regulation in 10 CFR 20.1003 defines ALARA in relevant part: "ALARA . . . means

making every reasonable effort to maintain exposures to radiation as far below the dose limits ... as

is practical consistent with the purpose for which the ... activity is undertaken."

55 DOE defines a controlled area in 10 CFR 835.2 as "any area to which access is managed by or for DOE to protect

individuals from exposure to radiation and/or radioactive material." NRC in 10 CFR 20.1003 defines restricted areas as "an area, access to which is limited ... for the purpose of protecting individuals against undue risks from exposure to

radiation and radioactive materials." The two definitions are essentially the same. 56

NRC in 10 CFR 20.1301(d) allows licensees to request NRC authorization to allow an individual member of the

public to receive up to an annual dose limit of 0.5 rem. DOE in 10 CFR 835 is more restrictive for the dose to an

individual member of the public with a limit of 0.1 rem maximum annual dose.

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12.3 State of Texas

The State of Texas in the Texas Administrative Code, Title 30, Part 1, Rule §336.2 defines

ALARA as follows:

"Making every reasonable effort to maintain exposures to radiation as far below the dose

limits in this chapter as is practical, consistent with1 the purpose for which the licensed

activity is undertaken, taking into account the state of technology, the economics of

improvements in relation to the state of technology, the economics of improvements in

relation to benefits to the public health and safety, and other societal and socioeconomic

considerations, and in relation to utilization of ionizing radiation and licensed radioactive

materials in the public interest."

12.4 Discussion

The DOE, NRC, and State of Texas definitions of ALARA are comparable.

12.0 References

Code of Federal Regulations

10 CFR 20, Standards for Protection Against Radiation.

10 CFR 835, Occupational Radiation Protection.

40 CFR 61, National Emission Standards for Hazardous Air Pollutants.

DOE Orders and Policies

DOE Order 440. 1A, Worker Protection Management for DOE Federal and Contractor Employees.

U.S. Department of Energy, Washington, D.C., March 27, 1998.

DOE Order 5400.5, Radiation Protection of the Public and the Environment, Change 2. U.S.

Department of Energy, Washington, D.C., January 7, 1993.

DOE Policy 441.1, Department of Energy Radiological Health and Safety Policy. U.S. Department of

Energy, Washington, D.C., April 26, 1996.

State Regulations

Texas Administrative Code, Title 30, Part 1, Chapter 336, Radiation Substance Rules

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APPENDIX E

Consideration of the Criteria in Section 3116 of the Ronald W. Reagan

National Defense Authorization Act for Fiscal Year 2005

Appendix Purpose

The purpose of this appendix is to discuss the criteria in Section 3116 of the Ronald

W. Reagan National Defense Authorization Act for Fiscal Year 2005 with respect to

this draft evaluation.

Appendix Content

This appendix describes the subject criteria in relation to the Department's plans for

disposal of the Vitrification Melter.

Key Points

* Section 3116 of the Ronald W. Reagan National Defense Authorization Act for

Fiscal Year 2005 does not apply to the Vitrification Melter.

* However, disposal of the Vitrification Melter at the Nevada National Security Site

or the WCS facility as low-level radioactive waste would be consistent with the

criteria of Section 3116 of the Ronald W. Reagan National Defense Authorization

Act for Fiscal Year 2005.

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1.0 Introduction

Sections 4 through 6 of this draft evaluation demonstrate that the Vitrification Melter waste

package meets the criteria of DOE Manual 435.1-1 for determining that the waste is incidental to

reprocessing and is not HLW, and will be managed and disposed of as LLW under DOE's

regulatory authority as applicable pursuant to the Atomic Energy Act of 1954, as amended. Section

3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 contains

similar criteria, and provides that the Secretary of Energy, in consultation with NRC, may determine

that waste resulting from reprocessing of spent nuclear fuel at DOE facilities in South Carolina and

Idaho, that is to be disposed of within those states, is not HLW where the criteria in section

3116(a)(1)-(3) are met.5 7

57 The criteria appear in Subsection (a) of Section 3116. Section 3116(a) provides:

"IN GENERAL-Notwithstanding the provisions of the Nuclear Waste Policy Act of 1982, the requirements of

section 202 of the Energy Reorganization Act of 1974, and other laws that define classes of radioactive waste,

with respect to material stored at a Department of Energy site at which activities are regulated by a covered

State pursuant to approved closure plans or permits issued by the State, the term 'high-level radioactive waste'

does not include radioactive waste resulting from the reprocessing of spent nuclear fuel that the Secretary of

Energy (in this section referred to as the 'Secretary'), in consultation with the Nuclear Regulatory Commission (in

this section referred to as the 'Commission'), determines-

(1) does not require permanent isolation in a deep geologic repository for spent fuel or high-level

radioactive waste;

(2) has had highly radioactive radionuclides removed to the maximum extent practical; and

(3) (A) does not exceed concentration limits for Class C low-level waste as set out in Section 61.55 of title

10, Code of Federal Regulations, and will be disposed of-

(i) in compliance with the performance objectives set out in subpart C of part 61 of title 10,

Code of Federal Regulations; and

(ii) pursuant to a State-approved closure plan or State-issued permit, authority for the approval

or issuance of which is conferred on the State outside of this section; or

(B) exceeds concentration limits for Class C low-level waste as set out in section 61.55 of title 10, Code

of Federal Regulations, but will be disposed of -

(i) in compliance with the performance objectives set out in subpart C of part 61 of title 10,

Code of Federal Regulations;

(ii) pursuant to a State-approved closure plan or State-issued permit, authority for which is

conferred on the State outside of this section; and

(iii) pursuant to plans developed by the Secretary in consultation with the Commission."

Subsection (b) of Section 3116 addresses monitoring by NRC. Subsection (c) addresses inapplicability to certain

materials (i.e., materials transported from the covered State). Subsection (d) identifies the covered States (South

Carolina and Idaho.) Subsection (e) addresses certain matters concerning construction of section 3116, and provides

that the section does not establish any precedent in any State other than South Carolina and Idaho, and does not

amend the West Valley Demonstration Act. Subsection (f) provides for judicial review of determinations made

pursuant to section 3116 and of any failure by NRC to carry out its monitoring responsibilities. (U.S. Congress 2004)

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Although Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal

Year 2005 does not apply to the Vitrification Melter,5 8 the following discussion addresses the

relevant criteria in 3116(a)(1)-(3) for perspective and information, and, because it may be of

interest to stakeholders, shows that disposal of the Vitrification Melter waste package as LLW at the

Nevada National Security Site or the WCS facility would be consistent with relevant criteria in

Section 3116(a)(1)-(3) of the National Defense Authorization Act for Fiscal Year 2005.

2.0 Consideration of Whether the Vitrification Melter Requires Permanent Isolation in a

Deep Geologic Repository

The first criterion or clause in Section 3116(a), as set forth in Section 3116(a)(1), provides that

the waste "does not require permanent isolation in a deep geologic repository for spent fuel or

high-level radioactive waste." DOE Manual 435.1-1 does not contain an identical consideration, but

similarly provides in relevant part in ChapterII.B.(2)(a) that the waste "will be managed as lowlevel waste" and meet the criteria in Section II.B.(2)(a).

With respect to the first criterion or clause, as provided in Section 3116(a)(1), the DOE, in

consultation with the NRC, has explained:

"Clause (1), noted above, is a broader criterion for the Secretary, in consultation with

the NRC, to consider whether, notwithstanding that waste from reprocessing meets the

other two criteria, there are other considerations that, in the Secretary's judgment, require

its disposal in a deep geologic repository. Generally, such considerations would be an

unusual case because waste that meets the third criterion would be waste that will be

disposed of in a manner that meets the 10 CFR 61, Subpart C performance objectives and

either falls within one of the classes set out in 10 CFR 61.55 that the NRC has specified are

considered "generally acceptable for near-surface disposal" or for which the Secretary has

consulted with NRC concerning DOE's disposal plans. As the NRC explained in In the Matter

of Louisiana Energy Services, L.P. (National Enrichment Services) (NRC 2005), the 10 CFR

Part 61, Subpart C performance objectives in turn "set forth the ultimate standards and

radiation limits for (1) protection of the general population from releases of radioactivity;

(2) protection of individuals from inadvertent intrusion; (3) protection of individuals during

operations; and (4) stability of the disposal site after closure." It follows that if disposal of a

waste stream in a facility that is not a deep geologic repository will meet these objectives,

58 That Section 3116(a) applies only to waste from reprocessing at DOE facilities in South Carolina and Idaho, which

is to be disposed of in those states, is made clear by the language used, which includes the following:

"(c) INAPPLICABILITY TO CERTAIN MATERIALS. - Subsection (a) shall not apply to any material otherwise

covered by that subsection that is transported from the covered State.

(d) COVERED STATES.-- For purposes of this section, the following States are covered States:

(1) the State of South Carolina.

(2) the State of Idaho."

(e) CONSTRUCTION. -

(2) Nothing in this section establishes any precedent or is binding on the State of Washington, the State of

Oregon, or any other State not covered by subsection (d) for the management, storage, treatment, and

disposition of radioactive and hazardous materials.

(5) Nothing in this section amends the West Valley Demonstration Act (42 U.S.C.2121a note)."

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in the ordinary case that waste stream does not "require disposal in a deep geologic

repository" because non-repository disposal will be protective of public health and safety.

It is possible that in rare circumstances a waste stream that meets the third criterion

might have some other unique radiological characteristic or may raise unique policy

considerations that warrant its disposal in a deep geologic repository. Clause (1) is an

acknowledgement by Congress of that possibility. For example, the waste stream could

contain material that, while not presenting a health and safety danger if disposed of at

near- or intermediate-surface, nevertheless presents non-proliferation risks that the

Secretary concludes cannot be adequately guarded against absent deep geologic disposal.

Clause (1) gives the Secretary, in consultation with NRC, the authority to consider such

factors in determining whether waste that meets the other two criteria needs disposal in a

deep geologic repository in light of such considerations." 5 9

That is not the case here. As demonstrated in Section 4 of this draft evaluation, key

radionuclides have been removed from the Vitrification Melter to the maximum extent technically

and economically practical. Moreover, the Vitrification Melter waste package will be in a solid

physical form and will not exceed the concentration limits for Class C low-level waste in 10 CFR

61.55, as described in Section 6. As explained in Section 5, management and disposal of the

Vitrification Melter as LLW at the Nevada National Security Site or the WCS facility also would meet

safety requirements comparable to the NRC performance objectives in 10 CFR 61, Subpart C, so as

to provide for the protection of human health and safety and the environment. As such, the

disposal of the Vitrification Melter waste package as LLW does not present a danger to human

health and safety, such that disposal in a deep geologic repository would be warranted.

Furthermore, the Vitrification Melter does not present unique radiological characteristics, or raise

non-proliferation risks or other unique policy considerations, which, while not manifesting a danger

to human health, nevertheless would command deep geologic disposal. Accordingly, the planned

disposal of the Vitrification Melter as LLW at the Nevada National Security Site or the WCS facility

meets DOE criteria and would be consistent with the first criterion of Section 3116(a).

3.0 Consideration of Removal of Highly Radioactive Radionuclides

The second criterion of Section 3116(a) specifies that the waste "has had highly radioactive

radionuclides removed to the maximum extent practical." DOE Manual 435.1-1, Chapter

II.B.(2)(a)l, contains a similar provision, which specifies that such wastes "[h]ave been processed,

or will be processed, to remove key radionuclides to the maximum extent that is technically and

economically practical."60

Section 4, Table 4-3, of this draft evaluation identifies key radionuclides for the Vitrification

Melter. As can be seen in this table, all radionuclides in Tables 1 and 2 of 10 CFR 61.55 were

considered. Furthermore, Section 4 of this draft evaluation describes how key radionuclides in the

Vitrification Melter have been removed to the maximum extent technically and economically

59 Basis for Section 3116 Determination for the Idaho Nuclear Technology and Engineenng Center Tank Farm

Facility(DOE 2006).

60 In this regard, NRC staff considers key radionuclides and highly-radioactive radionuclides - which are those

radionuclides that contribute most significantly to risk to the public, workers, and the environment - to be equivalent

for the purpose of evaluating waste determinations (NRC 2007).

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practical, thus satisfying the DOE criterion and evincing consistency with the second criterion of

3116(a).

5.0 Consideration of Radionuclide Concentration Limits and Waste Disposal Performance

Objectives

The third criterion in sc tion 3116(a)(3) ..... e..........." '

The hircrierin.i sectio If6()() concerns whether the waste mneet the concentration

limits for Class C LLW in 10 CFR 61.55 and whether the waste will be disposed of in accordance

with the performance objectives at 10 CFR 61, Subpart C.61 The criteria in DOE Manual 435.1-1,

Chapter II (B)(2)(a)2 and (a)3 similarly provide that waste "[w]ill be managed to meet safety

requirements comparable to the performance objectives set out in 10 CFR Part 61, Subpart C" and

"will be incorporated in a solid physical form at a concentration that does not exceed the applicable

concentration limits for Class C low-level waste as set out in 10 CFR 61.55", respectively.

Table 6-1 of this draft evaluation demonstrates that the Vitrification Melter waste package does

not exceed the Class C concentration limits in 10 CFR 61.55 (which are mirrored in the Texas

Administrative Code, Rule §336.362, Appendix E). In addition, the Vitrification Melter has been

packaged in a shielded shipping container which will be filled with low-density cellular concrete,

and thus will be in a solid physical form as discussed in Section 6. Section 4 of this draft evaluation

further shows that management and disposal of the waste will meet safety requirements

comparable to NRC performance objectives in 10 CFR Part 61, Subpart C. Given these

considerations, management and disposal of the Vitrification Melter as planned meets the abovereferenced DOE criteria and would be consistent with the third criterion of Section 3116(a).

6.0 Consultation with NRC

Section 3116(a) also provides for consultation with the NRC. As explained previously, DOE is

consulting with NRC concerning this draft evaluation, as well as making this draft evaluation

available for public review and comment. DOE will consider NRC comments, as well as comments

from the public and the States of Nevada and Texas, before finalizing the evaluation and before

making any final determination as to whether the Vitrification Melter is or is not high level waste.

Accordingly, such consultation is consistent with the provision for NRC consultation in section 3116

(a) of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005.

7.0 References

Federal Statutes

Ronald W Reagan National Defense Authorization Act for Fiscal Year 2005, Public Law 108-375,

118 Stat. 1811. 108th United States Congress, October 28, 2004.

Code of Federal Regulations

10 CFR 61.55, Waste Classifica'ion.

10 CFR 61, Subpart C, Licensing Requirements for Land Disposal of Radioactive Waste,

Performance Objectives.

DOE Manuals

DOE Manual 435.1-1, Radioactive Waste Management Manual, Change 1. U.S. Department of

Energy, Washington, D.C., June 19, 2001.

61 Although not germane here, section 3116(a)(3) also provides that the waste be disposed of "pursuant to a Stateapproved closure plan or State issuedpermit" for activities regulated by South Carolina or Idaho.

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State Regulations

Texas Administrative Code, Title 30, Part 1, Chapter 336, Radioactive Substance Rules (http://

info.sos.state.tx. us/pls/pub/readtac$ext.ViewTAC?tacview=4&ti =30&pt= l&ch =336)

Other References

NRC 2005, In the Matter of Lou-,iana - ergy Services, L.P. (Uational Eiiment Serwrces), CLI-05-

05. U.S. Nuclear Regulatory Commission, Washington, D.C., January 18, 2005.

NRC 2007, NRC Staff Guidance for Activities Related to U.S. Department of Energy Waste

Determinations, NUREG-1854. Draft Final Report, U.S. Nuclear Regulatory Commission,

Washington, D.C., August 2007.

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