West Valley Demonstration Project 

Draft Waste-Incidental-to-Reprocessing Evaluation

for the Concentrator Feed Makeup Tank

and the Melter Feed Hold Tank

June 2012

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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DRAFT WASTE-INCIDENTAL-TO-REPROCESSING EVALUATION FOR THE WVDP CFMT AND MFHT 

CONTENTS

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

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

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

1.2 Scope ............................................................................................................... 3

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

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

1.4.1 The Western New York Nuclear Service Center............................................ 4

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

1.4.3 The West Valley Demonstration Project ...................................................... 7

1.4.4 Characterization of the Two Vessels ........................................................... 7

1.4.5 Incorporation into a Solid Physical Form ..................................................... 8

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

1.4.7 Previous NRC Staff Review...................................................................... 10

1.5 Organization of this Draft Evaluation ................................................................... 10

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 Waste Storage Tanks....................................................... 13

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

2.3.1 The WVDP Act ....................................................................................... 16

2.3.2 Waste Treatment Preparations................................................................. 17

2.4 HLW Processing ................................................................................................ 18

2.4.1 Pretreatment of the Waste ...................................................................... 18

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

2.5 Concentrator Feed Makeup Tank Description, Operation, and Characterization ......... 20

2.5.1 Description............................................................................................ 20

2.5.2 Operational History................................................................................. 21

2.5.3 Characterization..................................................................................... 22

2.5.4 Characterization Quality Assurance........................................................... 23
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2.5.5 Consideration of Data Variability and Results Uncertainty............................ 23

2.5.6 NRC Independent Assessment ................................................................. 26

2.6 Melter Feed Hold Tank Description, Operation, and Characterization ....................... 26

2.6.1 Description............................................................................................ 26

2.6.2 Operational History................................................................................. 26

2.6.3 Characterization..................................................................................... 26

2.6.4 Characterization Quality Assurance........................................................... 27

2.6.5 Consideration of Data Variability and Results Uncertainty............................ 27

2.6.6 NRC Independent Assessment ................................................................. 29

2.7 WVDP Waste Management Plans......................................................................... 29

3.0 WASTE DETERMINATION CRITERIA.....................................................................31

3.1 Waste Determination Criteria Background ........................................................... 31

3.2 Applicable Waste Determination Criteria.............................................................. 31

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

MAXIMUM EXTENT THAT IS TECHNICALLY AND ECONOMICALLY PRACTICAL....33

4.1 Key Radionuclides ............................................................................................ 34

4.1.1 Introduction......................................................................................... 34

4.1.2 DOE Guidance on Key Radionuclides ....................................................... 34

4.1.3 Requirements of 10 CFR 61.55 ............................................................... 35

4.1.4 Radionuclides in West Valley HLW .......................................................... 36

4.1.5 Radionuclides Important to the Disposal Site Performance Assessments...... 37

4.1.6 Conclusions About Key Radionuclides in the Equipment............................. 37

4.2 Removal to the Maximum Extent Technically and Economically Practical................. 38

4.2.1 Technical Practicality Assessment ........................................................... 40

4.2.2 In-Place Decontamination By Flushing With Water Using 

High-Pressure Spray.............................................................................. 40

4.2.3 Mechanical Decontamination Using a Ball Mill........................................... 44

4.2.4 Chemical Decontamination..................................................................... 45

4.3 Economic Practicality Assessment....................................................................... 45

4.3.1 Economic Practicality of Additional Flushing ............................................. 45
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4.3.2 Summary and Conclusions ..................................................................... 51

5.0 THE WASTE WILL BE MANAGED TO MEET SAFETY REQUIREMENTS 

COMPARABLE TO THE PERFORMANCE OBJECTIVES OF 10 CFR 61, SUBPART C..52

5.1 Introduction .................................................................................................... 53

5.2 DOE Safety Requirements ................................................................................. 54

5.2.1 General Safety Requirement................................................................... 55

5.2.2 Protection of the General Population from Releases of Radioactivity............ 55

5.2.3 Protection of Individuals from Inadvertent Intrusion ................................. 58

5.2.4 Protection of Individuals During Operations ............................................. 60

5.2.5 Stability of the Disposal Site After Closure................................................ 62

5.3 The Vessels Will Meet Disposal Site Waste Acceptance Criteria .............................. 63

5.3.1 Nevada National Security Site Waste Acceptance Criteria........................... 64

5.3.2 The Vessel Waste Packages ................................................................... 65

5.3.3 The Vessels Meet Nevada National Security Site Waste Disposal Criteria ..... 65

5.4 Meeting WCS Waste Acceptance Criteria............................................................. 66

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

BE MANAGED IN ACCORDANCE WITH DOE REQUIREMENTS AS LLW..................68

7.0 CONSULTATION WITH NRC AND OPPORTUNITY FOR PUBLIC COMMENT...........71

8.0 CONCLUSIONS ......................................................................................................72

9.0 REFERENCES .........................................................................................................73

APPENDICES

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

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

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

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

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

2-5 The Concentrator Feed Makeup Tank (left) and the Melter Feed Hold Tank Installed 

in the Vitrification Cell ............................................................................................... 21

2-6 The Concentrator Feed Makeup Tank Shipping Container .............................................. 22

4-1 Melter Feed Hold Tank Head Showing Potential Spray Nozzle Conditions ......................... 41

4-2 Before and After Flushing Images of the Two Vessels ................................................... 42

4-3 High-Pressure Spray of the Concentrator Feed Makeup Tank and the Melter Feed 

Hold Tank ................................................................................................................ 48

4-4 Variations in Concentrator Feed Makeup Tank Liquid Levels........................................... 49

4-5 Estimating Flushing Effectiveness ............................................................................... 50

5-1 General Process Used to Ensure Performance Objectives are Achieved............................ 55

TABLES

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

Completion of Fuel Reprocessing.................................................................................. 14

2-2 Concentrator Feed Makeup Tank Activity Estimates ........................................................ 22

2-3 Measured Radionuclide Concentrations and Associated Uncertainty .................................. 25

2-4 Melter Feed Hold Tank Activity Estimates ...................................................................... 27

2-5 Measured Radionuclide Concentrations and Associated Uncertainty .................................. 29

4-1 10 CFR 61.55, Table 1 (Long-Lived Radionuclides) ......................................................... 35

4-2 10 CFR 61.55, Table 2 (Short-Lived Radionuclides)......................................................... 35

4-3 Key Radionuclides for this Draft Evaluation.................................................................... 38

4-4 Vessel Flushing Effectiveness in Terms of Estimated Cs-137 Removal............................... 43

4-5 Summary of Direct and Indirect Flushes Performed ........................................................ 46

5-1 Vessel Radionuclide Concentrations in Bq/m3................................................................. 66

5-2 Key WCS Federal Facility Waste Disposal Facility License Requirements ............................ 67

6-1 Vessel Waste Classification Results With Respect to Class C limits .................................... 69
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NOTATION

Revision 0 v

Acronyms and Abbreviations 

ALARA as low as reasonably achievable

Am americium

CFMT concentrator feed makeup tank

CFR Code of Federal Regulations

CHBWV CH2M Hill B&W West Valley

Cm curium

Co cobalt

Cs cesium

DOE Department of Energy

FR Federal Register

HLW high-level waste

I iodine 

IP industrial package

K potassium 

LLW low-level waste

MFHT melter feed hold tank

Mn manganese 

Ni nickel 

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]

SA supplement analysis

Sr strontium

Tc technetium 

THOREX thorium uranium extraction [process]

U uranium

WCS Waste Control Specialists

WIR waste incidental to reprocessing

WVDP West Valley Demonstration Project 

WVES West Valley Environmental Services

WVNSCO West Valley Nuclear Services Company
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NOTATION

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Units 

Bq Becquerel 

Ci curie

cm centimeter

cm3cubic centimeter 

g gram [mass]

h hour

kg kilogram, 1,000 grams

L liter

m meter

m

2

square meter

m

3

cubic meter

mrem millirem, 0.001 Roentgen equivalent man (rem)

millirem 0.001 Roentgen equivalent man

mSv millisievert, 0.001 sievert 

mL milliliter, 0.001 liter

μCi microcurie, 0.000001 curie

nCi nanocurie, 10-9curie

pCi picocurie, 10-12 curie

R Roentgen

rem Roentgen equivalent man 

s second

Sv sievert
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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 concentrator feed makeup tank 

and the melter feed hold tank from the West Valley Demonstration Project in New York 

meet the waste-incidental-to-reprocessing criteria of Department of Energy Manual 

435.1-1, Radioactive Waste Management Manual.

 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 concentrator feed makeup tank and the melter feed hold tank are vessels used in 

the vitrification process to prepare and temporarily store, respectively, slurry consisting 

of pre-treated high-level waste and glass formers that was supplied to the vitrification 

melter. 

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

during 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 concentrator feed makeup tank and the melter feed hold tank have been 

characterized for radioactivity, determined to have radionuclide concentrations that do 

not exceed limits for Class C low-level radioactive waste, and packaged for shipment to 

an offsite low-level waste disposal facility.

 To dispose of the these vessels offsite as low-level radioactive waste, the Department 

must first demonstrate that they meet criteria of Department of Energy Manual 435.1-1, 

Radioactive Waste Management Manual, 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 as well as state and public review and 

comment, consistent with the Department’s policy. 

 The Department will make its final determination of whether the concentrator feed 

makeup tank and the melter feed hold tank are or are 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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Waste Incidental to Reprocessing Requirements and this Draft Evaluation

The term waste incidental to reprocessing refers 

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.

The citation process is intended for those waste 

streams for which it can be easily determined up 

front (that is, without detailed analysis) that they 

do not pose the long-term hazards associated with 

HLW (DOE Guide 435.1-1). The concentrator feed 

makeup tank and the melter feed hold tank do not 

fall into this category. 

In the West Valley Demonstration Project Waste 

Management Environmental Statement Supplement Analysis (DOE 2006), DOE stated:

“At this point [in 2006], DOE intends to prepare 

draft waste-incidental-to-reprocessing (WIR)

determinations in accordance with DOE Order 

435.1 for the components of the Vitrification 

Facility included in this SA, as those components 

[including the subject vessels] have been in 

direct proximity to HLW in the vitrification 

process and require a WlR determination to be 

classified as LLW or another waste type. DOE 

intends to issue the draft WIR determination for 

publication in the Federal Register for a 45-day 

comment period. In the same timeframe, DOE 

will forward the draft WIR determination to the 

Nuclear Regulatory Commission for their review 

in accordance with their responsibilities under the 

West Valley Demonstration Project Act. At such 

time as their review is completed, DOE may issue 

a final WIR determination.”

Consistent with this statement, DOE policy, and 

guidance in DOE Guide 435.1-1, Implementation

Guide for Use with DOE Manual 435.1-1, this draft 

evaluation is being issued for NRC review and public 

comment.

1.1 Purpose

The purpose of this draft waste-incidental-to-reprocessing evaluation is to evaluate whether 

the concentrator feed makeup tank and the melter feed hold tank – which contained pretreated 

high-level waste (HLW) at the West Valley Demonstration Project (WVDP) site in Western New York 

– meet the waste-incidental-to-reprocessing criteria, are not HLW, and may be managed as lowlevel waste (LLW) pursuant to Department of Energy (DOE Manual 435.1-1, Radioactive Waste 

Management Manual. 

These vessels were used in DOE’s process to vitrify liquid HLW, which 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 the two vessels offsite to meet its obligations under the West Valley 

Demonstration Project Act of 1980, which is described below. Note that the concentrator feed 

makeup tank and the melter feed hold tank are frequently referred to as the vessels or the subject 

vessels in the discussions that follow.
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1.2 Scope

This draft waste-incidental-to-reprocessing evaluation applies only to the WVDP concentrator 

feed makeup tank and the melter feed hold tank and to no other equipment. These vessels were 

previously decontaminated, characterized, and prepared for shipment as discussed further in 

Section 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 to NRC for review provides stakeholders an opportunity to 

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

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

will issue its final determination on whether the concentrator feed makeup tank and the melter 

feed hold tank are or are 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 Manual, 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 concentrator feed makeup tank and the 

melter feed hold tank are 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; and

(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, that Act does not 

apply to the West Valley site.

1

 

1 DOE considered the Section 3116(a)(1) criteria for perspective and information in this waste-incidental-toreprocessing evaluation as explained in Appendix C.
DRAFT WASTE-INCIDENTAL-TO-REPROCESSING EVALUATION FOR THE WVDP CFMT AND MFHT 

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Waste-incidental-to-reprocessing 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 [waste-incidental-to-reprocessing] criteria are applicable; for 

example, DOE may decide to apply DOE Order 435.1.”

This evaluation was prepared in accordance with the WVDP procedure for waste-incidental-toreprocessing determinations (WVES 2011a), which is based on requirements in DOE Manual 435.-1 

and the associated guidance in DOE Guide 435.1-1, by a consultant experienced in DOE radioactive 

waste management and waste-incidental-to-reprocessing evaluations under the direction and 

oversight of a manager and a senior engineer from West Valley Environmental Services with similar 

experience. The quality assurance process followed was consistent with the DOE requirements at 

10 CFR Part 830, Subpart A, Quality Assurance Requirements, and DOE Order 414.1D, Quality 

Assurance. 

Data used in vessel waste package characterization were validated as discussed in Section 2. 

Calculations performed in support of this draft evaluation were formally documented and peer 

reviewed. This draft evaluation also underwent detailed reviews by West Valley Environmental 

Services, CH2M Hill B&W West Valley (CHBWV)2, and the DOE Office of Environmental 

Management for technical adequacy, completeness, correctness, and compliance with applicable 

requirements. These reviews were formally documented and all review comments incorporated or 

otherwise resolved.

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 vessel design, residual radioactivity in the vessels, 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. 

 

2 CHBWV replaced West Valley Environmental Services as DOE’s site contractor in September 2011 after a 60-day 

transition period.
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Figure 1-1. The Western New York Nuclear Service 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 and the State-Licensed Disposal Area. 

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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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, recovering approximately 620 metric tons of uranium and 

approximately 1,926 kilograms of uranium (DOE 1999). 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. (Rykken 1986)

Rock Springs Road

Approximate Location of WVDP Premises Boundary

Process Building

SDA NDA

Area of HLW Tanks

S

Figure 1-2. The WVDP Area of the Center in 2006 (WVDP photo) 
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In 1972, Nuclear Fuel Services shut down the reprocessing plant for expansion, modifications,

and additions. However, reprocessing never resumed. 

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

contained an estimated 30 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 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 

Supernatant 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).3

1.4.4 Characterization of the Two Vessels

The concentrator feed makeup tank and the melter feed hold tank have been characterized for 

radioactivity based on measured gamma radiation levels and sample analytical data and found to 

 

3

The 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 20582 (April 20, 2010)) were issued in 2010. 
DRAFT WASTE-INCIDENTAL-TO-REPROCESSING EVALUATION FOR THE WVDP CFMT AND MFHT 

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not exceed NRC limits for Class C LLW under 10 CFR 61.55 (WMG 2011)

4

. Section 2 of this draft 

evaluation provides more detail on the characterization process. 

1.4.5 Incorporation into a Solid Physical Form

The concentrator feed makeup tank and the melter feed hold tank are both in a solid physical 

form. Void spaces in both vessels and their waste containers have been filled with grout consisting 

of low-density cellular concrete to stabilize the vessels within the shipping containers during 

transport and to encapsulate surface contamination. This grout does not increase the waste 

disposal volume (which is based on the exterior of the entire disposal package) and was not 

considered in the classification of the waste.

1.4.6 Potential Waste Disposal Facilities

DOE plans to ship the concentrator feed makeup tank and the melter feed hold tank waste 

packages 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.5

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 is presently inactive. 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 vessels at the Nevada National Security Site, they 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 Texas6for near-surface disposal of 

Class A, B, and C LLW from Texas Compact7 waste generators and limited quantities of nonparty 

compact waste imported from other states, as well as Class A, B, and C and mixed low-level

Federal facility waste8. 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. If DOE decides to dispose of the concentrator feed makeup tank and the melter feed hold 

 

4 Owing to the concentrations of Cs-137, both vessel waste packages are Class C waste. 

5

The two vessels constitute a unique waste stream. While it would be physically possible to ship one vessel to one 

disposal site and the other vessel to the other disposal site, such an arrangement would be costly and inefficient due 

to factors such as the extensive waste acceptance processes and transportation logistics. 

6

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 initially 

issued a license, with conditions, by the Texas Commission on Environmental Quality in 2009, which subsequently 

has been amended several times, for a compact waste disposal facility and a Federal waste disposal facility. In April 

2012, the Texas Commission on Environmental Quality issued an authorization to accept and begin waste disposal 

under the license. A request for a contested hearing was granted by the 261st District Civil Court of Travis County 

Texas in May 2012.

7

The Texas Compact consists of the states of Texas and Vermont. Waste generators in these states, as well as 

generators in other states (upon approval of applicable import petitions), are authorized to dispose of LLW in the 

WCS Texas Compact disposal facility. 

8

The license does not authorize disposal of greater-than-Class C LLW.
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tank at the WCS facility, the waste packages would be disposed of as LLW in the Federal Facility 

Waste Disposal Facility.9 

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 (65 FR 10061 (February 25, 2000)) 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 35073 (June 16, 2005). Therefore, 

disposal of the vessels at either the Nevada National Security Site or the WCS facility is consistent

with this Record of Decision.

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 

concentrator feed makeup tank and the melter feed hold tank. The analysis noted that these 

vessels 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. 

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

LLW such as the vessel waste packages 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 vessel waste packages would be sent would be made 

after the final waste-incidental-to-reprocessing evaluation and determination, following 

consideration of NRC and public comments on this draft evaluation, and after DOE confers with 

appropriate State officials in the states where the waste packages may be disposed. Prior to 

disposal, DOE will post notice of its decision concerning the disposal location on DOE’s WVDP 

website (www.wv.doe.gov) and DOE’s Office of Environmental Management website 

(www.em.doe.gov).

 

9

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 vessels 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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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 concentrator feed makeup 

tank and the melter feed hold tank 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 concentrator feed 

makeup tank and the melter feed hold tank, 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 concentrator feed 

makeup tank and the melter feed hold tank 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 

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 concentrator feed 

makeup tank and the melter feed hold tank are less than Class C concentration limits, and that 

these vessels will be managed in accordance with Chapter IV of DOE Manual 435.1-1.

Section 7 describes the opportunity for NRC and public comments. 

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

Section 9 identifies the references cited in the draft evaluation.

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

LLW disposal.

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

standards.

Appendix C 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 vessels, radiological characterization of 

the vessels, 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 concentrator feed makeup tank is a 6,000-gallon capacity Hastelloy-22 vessel 

used from 1996 through 2002 to prepare the high-activity waste and glass 

formers mixture for vitrification.

 The melter feed hold tank is a 5,000-gallon capacity stainless steel vessel used 

from 1996 through 2002 to supply feed material from the concentrator feed 

makeup tank to the vitrification melter.

 In 2002, both vessels were flushed, emptied, and shut down. 

 The two vessels have been characterized for residual radioactivity using measured 

dose rates and radionuclide scaling factors based on sample analytical data. 

 Based on the characterization results, the radioactivity concentrations in the two 

vessels are below Class C concentrations limits and the two vessels are Class C 

low-level waste.

 Each vessel has been loaded in a custom-built steel shipping container in 

preparation for offsite disposal and the vessel and container filled with grout for 

stabilization purposes.

 The Department plans to ship the packaged vessels 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 subject vessels 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 concentrator feed makeup tank, 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 provides similar information for the melter feed hold tank.

 Section 2.7 describes DOE plans for disposing of the subject vessels. 

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 in the subject vessels.

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-Reactor10 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.)

 

10 N-Reactor was a dual-purpose reactor that produced electricity as well as plutonium.
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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 – UO3

product and PuO3 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. 

Process Building

01-14 Building

Plant Office Building

Utility Room

Main Plant Stack

Vitrification Facility

Utility Room Expansion

Laundry Room

Fuel Receiving Lagoon 2

and Storage

Figure 2-1. The Process Building, the Vitrification Facility, and Ancillary Facilities in 2006 (WVDP photo)
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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 Tank 8D-2 

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

H-3 9.5E+01 ~0 <2.0E+00 <9.7E+01

C-14 1.4E+02 ~0 (1) 1.4E+02

Fe-55 (1) 1.0E+03 (1) 1.0E+03

Ni-59 (1) 8.2E+01 (1) 8.2E+01

Co-60 ~0 4.7E+00 1.2E+03 1.2E+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

Irradiated Fuel

Fuel Preparation

Dissolution in Nitric 

Acid

Separation of Fission 

Products from 

Uranium-Plutonium

Uranium-Plutonium 

Partition

Uranyl Nitrate Plutonium Nitrate

Purification and 

Conversion to Oxide

Purification and 

Conversion to Oxide

UO3 Product PuO2 Product

Nitric Acid

Off-Gases

Off-gases and 

Cladding (Hulls)

Nitric Acid 

Recovery

High-Level 

Waste

Solvent tributyl

phosphate and 

n-dodecane

Figure 2-2. Spent Fuel Reprocessing Diagram (PUREX Process)
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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 (from Eisenstatt 1986 Table 6, fission and activation products 

decay-corrected to July 1987) (Continued)

Radionuclide Tank 8D-2 

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

Nb-93m (1) 2.3E+02 (1) 2.3E+02

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

Ru-106 (1) 1.3E+02 <3.1E-01 1.3E+02

Rh-106 (1) 1.3E+02 <3.1E-01 1.3E+02

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

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

Te-125m 1.1E+01 1.0E+03 (1) 1.0E+03

Sn-126 (1) 4.0E+01 (1) 4.0E+01

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

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

I-129 2.1E-01

(1) <1.5E-01 <3.6E-01

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

Cs-135 1.6E+02 (1) (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 <2.0E-02 1.4E+01

Pr-144 2.9E-05 1.4E+01 <2.0E-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) 1.6E+00 1.6E+00

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

U-234 2.9E-01 4.0E+00 3.0E-01 4.6E+00

U-235 6.4E-03 8.9E-02 4.9E-03 1.0E-01

U-236 1.9E-02 2.7E-01 1.0E-02 3.0E-01

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

Np-237 (1) 1.1E+01 (1) 1.1E+01

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

Pu-238 1.3E+02 6.5E+03 5.3E+02 7.2E+03
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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 (from Eisenstatt 1986 Table 6, fission and activation products 

decay-corrected to July 1987) (Continued)

Radionuclide Tank 8D-2 

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

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

Pu-240 1.9E+01 1.3E+03 9.0E+00 1.3E+03

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

Pu-242 2.5E-02 1.7E+00 1.3E-02 1.7E+00

Am-241 (1) 7.2E+04 2.7E+02 7.2E+04

Am-242 (1)2.1E+01 (1) 2.1E+01

Am-242m (1) 2.1E+01 (1) 2.1E+01

Am-243 (1) 2.4E+03 8.8E+00 2.4E+03

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

Cm-243 (1) 1.7E+02 5.0E-02 1.7E+02

Cm-244 (1) 2.2E+04 1.6E+01 2.2E+04

Cm-245 (1) 1.0E+01 1.2E-03 1.0E+01

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

NOTES: (1) Not present or undetectable. 

(2) The progeny of Sr-90 (Y-90) and Cs-137 (Ba-137m) 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). 
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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 

Premises11. 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 WVDP12. 

2.3.2 Waste Treatment Preparations

To manage the HLW, DOE selected onsite processing using a salt/sludge13 separation process 

(47 FR 40705 (September 15, 1982)). 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 exchange14

columns containing zeolite; 

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

Treatment System;

 

11 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.

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

In September 2011, CH2M Hill-B&W West Valley LLC became the site contractor for Phase 1 decommissioning and 

facility disposition activities. 

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

14 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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 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 

subject vessels.

2.4 HLW Processing

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

2.4.1 Pretreatment of the Waste

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

PUREX sludge washing, (3) PUREX/THOREX 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 
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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. 

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.

The Vitrification Facility building housed 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 subject vessels in 

the vitrification process.

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

Chemical Process Cell, where the HLW canisters were temporarily stored, were part of the original Process 

Building.) 
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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 preparations for Phase 1 of the 

decommissioning. DOE plans for waste shipment are summarized in Section 2.7 below.

2.5 Concentration Feed Makeup Tank Description, Operation, and Characterization 

Figure 2-5 shows the concentrator feed makeup tank in position before the start of the 

vitrification program.

2.5.1 Description 

The concentrator feed makeup tank is a cylindrical vessel approximately 13.5 feet long and 10 

feet in diameter with a nominal capacity of approximately 6,000 gallons. Constructed of Hastelloy 

C-22, it contains an agitation system used to stir its contents and four baffles. The lower part of the 

vessel exterior is covered with heat transfer coils formed of half-sections of 3.5 inch stainless steel 

pipe covered with fiberglass insulation and stainless steel sheet. The vessel was supported by a 

lower skirt attached to a base plate. 

Figure 2-4. Vitrification Process Flow Diagram (for information, not to scale)
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2.5.2 Operational History

During use, the concentrator feed makeup tank received pretreated liquid HLW. This waste was 

combined with the heel remaining from the previous batch and with vitrification process recycle 

streams. The vessel contents were then sampled and analyzed, and chemicals were added to 

achieve the required waste form composition. Excess water was removed by evaporation. After 

verification that the mixture met specifications, the feed slurry batch was transferred to the melter 

feed hold tank.

Following completion of vitrification in 2002, the concentrator feed makeup tank was 

extensively flushed as described in Section 4. Preparations for removal of the vessel from the 

Vitrification Cell included removal of external hardware such as the agitation system motor. 

In 2004, the concentrator feed makeup tank was loaded into its shipping container inside the 

Process Building Equipment Decontamination Room, which adjoins the Vitrification Facility. Figure 

2-6 shows the shipping container.

Figure 2-5. The Concentrator Feed Makeup Tank (left) and the Melter Feed Hold Tank Installed 

in the Vitrification Cell (WVDP photo)
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The vessel and the container 

were then filled with cement grout 

and relocated to an onsite rail 

storage area. The loaded shipping 

container, which is steel shielded 

and meets Department of 

Transportation Industrial Package 

2 (IP-2) requirements (WMG 

2004d) , is approxi-mately 19 feet 

long, 13 feet wide, and 14 feet 

high and weighs approximately 

322,000 pounds. The vessel itself, 

before grout was added, weighed 

approximately 18,206 pounds. 

(CHBWV 2011a) 

2.5.3 Characterization

Details of the waste package characterization appear in the waste profile prepared for disposal 

at the Nevada National Security Site (CHBWV 2011a) and the associated characterization report

(WMG 2011). The characterization process made use of sample analytical data and the average 

measured dose rate of collimated readings taken with a shielded radiation probe one foot from the 

sides of the installed vessel (1.62 R/h). 

A QAD15 geometry model was used to calculate a dose-to-curie conversion factor for cesium 

137, the amount of cesium 137 estimated from the measured dose rate, and the amounts of other 

radionuclides estimated using radionuclide scaling factors based on sample analytical data. The 

RADMANTM13 and MegashieldTM13 computer codes were used in the calculations. 

Table 2-2 shows the estimated residual radioactivity in the concentrator feed makeup tank, 

which totaled 96.5 Ci as of October 1, 2004.

Table 2-2. Concentrator Feed Makeup Tank Total Activity Estimate(1)

 Nuclide Activity (Ci) Nuclide Activity (Ci)

C-14 NA U-238 6.39E-06

K-40 NA Np-237 6.66E-05

Mn-54 NA Pu-238 5.19E-03

Co-60 1.85E-03 Pu-239 1.40E-03

Sr-90 1.12E+00 Pu-240 1.07E-03

Zr-95 NA Pu-241 1.54E-02

Tc-99 4.17E-03 Pu-242 NA

Cs-137 9.53E+01 Am-241 2.48E-02

Eu-154 5.17E-02 Am-242m NA

 

15 The QAD, RADMANTM and MegashieldTM software are computer codes commonly used in evaluation of 

radioactive waste packages and associated shielding.

Figure 2-6. The Concentrator Feed Makeup Tank Shipping 

Container (WVDP photo)
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Table 2-2. Concentrator Feed Makeup Tank Total Activity Estimate(1) (continued)

 Nuclide Activity (Ci) Nuclide Activity (Ci)

Th-228 NA Am-243 2.55E-04

Th-230 NA Cm-242 9.78E-05

Th-232 2.53E-06 Cm-243 1.55E-04

U-232 1.41E-04 Cm-244 4.02E-03

U-233 5.84E-05 Cm-245 NA

U-234 2.79E-05 Cm-246 NA

U-235 NA

LEGEND: NA = not available. 

NOTES: (1) From WMG 2011 as of October 1, 2004. 

The Nevada National Security Site Waste Acceptance Criteria (DOE 2012) includes limits on the 

amount of plutonium 239 equivalent grams that can be disposed in an individual package. 

Calculations included in the waste profile show that the plutonium 239 equivalent gram estimate in 

the concentrator feed makeup tank disposal package falls well below the individual package limit of 

300 (CHBWV 2011a)

16

.

2.5.4 Characterization Quality Assurance 

The characterization performed was consistent with the WVDP characterization process 

described in the Characterization Management Plan for the Facility Characterization Project

(Michalczak 2004). This plan provides data quality objectives and describes how data were 

validated. 

The data used in the characterization were supplied by WVNSCO, the site contractor. The dose 

rate data used were collected for this purpose in February 2004 by WVNSCO using a calibrated 

instrument – a Geiger-Mueller detector with a Ludlum model 2241 scaler/ratemeter, with the 

detector shielded to provide collimated measurements (WVNSCO 2004c). The analytical data came 

from analysis of five samples as shown in the characterization report (WMG 2011). These samples 

were collected during the 2001 – 2003 period and analyzed by the onsite Analytical and Process 

Chemistry Laboratory, which was subject to periodic assessments to ensure that analytical 

requirements were being satisfied (Michalczak 2004). The WMG characterization report was also 

independently reviewed by the site contractor and changes from this review were incorporated

(WMG 2011). 

2.5.5 Consideration of Data Variability and Results Uncertainty

The characterization made use of the average value of 10 dose rate measurements to calculate 

the amount of Cs-137 present in the vessel, and geometric means of sample analytical data to 

calculate scaling factors used to estimate the amounts of the other radionuclides present. To 

account for uncertainties in the radionuclide activity estimates, the Nevada National Security Site 

waste profile radiological technical basis document (CHBWV 2011a) identifies high and low activity 

ranges that are plus 20 percent and minus 20 percent, respectively, of the final waste form activity 

concentrations, which are based on the estimates in the characterization report (WMG 2011) and

shown in Table 2-2. 

 

16 The waste profile (CHBWV 2011a) indicates that the maximum plutonium 239 equivalent gram value for the waste 

stream is 13. The WCS LLW disposal facility does not have plutonium 239 equivalent gram limits. 
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Consideration of Data Variability 

The dose rates measured on the melter feed hold tank varied with location, ranging from 1.17 

R/h to 2.25 R/h (WVNSCO 2004c). The average value of 1.62 R/h was used in the calculation, as 

noted previously. Some variation in dose rates in different locations on the outside of the vessel 

would be expected due to varying amounts of contamination in the vessel interior and the 1.17 R/h 

to 2.25 R/h range is consistent with this expectation. The use of an average dose rate value in the 

calculation is considered to be appropriate since it accounts for the observed variations. 

Consideration of the variability in the sample analytical data is more complex because of the 

multiple data sets used in the calculation and the multiple analyses of the five samples. As 

discussed later in this draft evaluation, a total of 66 batches of HLW slurry numbered 10 through 

75 were prepared for vitrification in the concentrator feed makeup tank, along with two batches of 

lower-activity vitrification system decontamination solutions identified as batches 76 and 77.

Analytical data used in characterization came from two samples of batch 72 taken at different 

times, one sample from batch 74, one sample from batch 75, and one sample of residual liquid 

collected from the vessel after completion of vitrification (WMG 2011)17. Each of the batch samples 

was analyzed nine times for most radionuclides of interest (WVNSCO 2002j) and the average 

values used in the calculations. The July 2003 sample was analyzed three times as shown in the 

characterization report (WMG 2011). The geometric means of the measured radionuclide

concentrations in these samples were used to produce the scaling factors that were used in the 

characterization (WMG 2011). 

The radionuclide concentrations and distributions in the different samples are somewhat 

different, as can be seen from the table in the report, a condition attributed mainly to differences 

among the various waste batches18. However, it is appropriate to consider different waste batches 

because residual contamination inside the vessel likely built up over time as vitrification proceeded. 

The use of the geometric means of the concentrations from the different samples is considered to 

be reasonable for estimating the scaling factors because this approach takes into account the 

differing radionuclide distributions in the samples. 

Consideration of Data Uncertainty

Uncertainty in a typical field dose rate measurement made using a calibrated instrument 

depends on various factors. However, the use of the average value of the nine measurements 

taken in various locations on the concentrator feed makeup tank (WVNSCO 2004c) minimizes the 

uncertainty in the Cs-137 activity estimate. 

Regarding the analytical data used to develop scaling factors, the spreadsheet showing 

analytical data from samples of the various concentrator feed makeup tank batches (WVNSCO 

2002j) identifies the uncertainty in the individual measurements. Table 2-3 shows the estimated 

uncertainty for Cs-137 and Am-241, the two radionuclides most important in classification of the 

 

17 This information appears in the report (WMG 2011) in a table labeled “Analysis of Multiple Sample Data Sets 

(SCAL)” marked CFMT. 

18 This table also appears in Attachment A to the Nevada National Security Site waste profile (CHBWV 2011a). It 

shows measured radionuclide concentrations from analysis of seven samples. The radionuclide concentration data

shows some variation in radionuclide distributions. For example, the Sr-90 to Cs-137 ratios in the seven sets of 

analytical data vary from 1.53E-03 to 7.13E-02. As noted in the text, geometric means of the measured radionuclide 

concentrations in these samples were used to produce the scaling factors. 
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concentration feed makeup tank waste package19, for sets of nine measurements of one sample. 

For Cs-137, sample 00-1534, the first batch 72 sample is used. For Am-241, sample 01-2498, the 

batch 75 sample, is used.20

Table 2-3. Measured Radionuclide Concentrations and Associated Uncertainty(1) Cs-137, Sample 00-15

34

Analysis Concentration (Ci/g) Uncertainty (Ci/g) Percent Uncertainty

1 4.70E+02 7.20E+01 15%

2 4.98E+02 7.63E+01 15%

3 4.85E+02 3.00E-01 <1%

4 5.32E+02 8.15E+01 15%

5 4.68E+02 2.90E+00 1%

6 4.81E+02 2.90E+00 1%

7 4.48E+02 2.90E+00 1%

8 4.81E+02 2.90E+00 1%

9 5.00E+02 3.10E+00 1%

Am-241, Sample 01-2498

Analysis Concentration (Ci/g) Uncertainty (Ci/g) Percent Uncertainty

1 3.88E+00 2.33E-01 6%

2 3.87E+00 2.34E-01 6%

3 3.72E+00 2.24E-01 6%

4 4.17E+00 2.51E-01 6%

5 3.66E+00 2.21E-01 6%

6 4.04E+00 2.43E-01 6%

7 3.87E+00 2.33E-01 6%

8 3.75E+00 2.26E-01 6%

9 3.80E+00 2.29E-01 6%

NOTE: (1) From WVNSC 2002j. 

In these representative examples, the maximum amount of uncertainty in the individual 

measurements is 15 percent for Cs-137 and six percent for Am-241. The use of the average of 

multiple measurements minimizes the uncertainty associated with the analytical data used in 

calculating the scaling factors related to Cs-137.

Conclusions

Given the approach used in the characterization, and the uncertainty in the analytical data, 

DOE concludes that uncertainty associated with radionuclide estimates is likely bounded by the 20 

percent concentration range estimates included in the Nevada National Security Site waste profile 

(CHBWV 2011a). 

 

19The Class C calculations on page 14 of the waste characterization report (WMG 2011) show Am-241 to dominate 

the Table 1 sum of fractions and Cs-137 to dominate the Table 2 sum of fractions. 

20 These samples were considered representative in that they were from two different batches and the radionuclides 

were selected because they are important to waste classification as discussed in the previous footnote. 
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2.5.6 NRC Independent Assessment

In October and November 2004, NRC representatives made monitoring visits to evaluate 

activities associated with packaging and eventual offsite disposal of the three vitrification process 

components: the vitrification melter, the concentrator feed makeup tank, and the melter feed hold 

tank. The NRC representatives determined that the three components were appropriately 

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

requirements. 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 independently confirmed the validity of the 

characterization process used by DOE for the concentrator feed makeup tank. 

2.6 The Melter Feed Hold Tank Description, Operation, and Characterization

2.6.1 Description

The melter feed hold tank is similar to the concentrator feed makeup tank in many respects, 

although slightly smaller as can be seen on Figure 2-5. It is a cylindrical vessel approximately 10

feet long and 10 feet in diameter with a nominal capacity of approximately 5,000 gallons. 

Constructed of stainless steel, it contains four baffles and an agitation system used to stir its 

contents. The vessel exterior is partially covered by a cooling jacket. The agitation system was used 

to maintain homogeneity of the slurry. The vessel was supported by four trunnions.

2.6.2 Operational History

During use, the melter feed hold tank held and mixed HLW slurry feed for delivery to the 

melter. Following completion of vitrification in 2002, the vessel was extensively flushed as 

described in Section 4. In 2004, the melter feed hold tank was loaded into its shipping container 

inside the Process Building Equipment Decontamination Room. The vessel and the container were 

then filled with cement grout and relocated to an onsite rail storage area. 

The loaded melter feed hold tank IP-2 shipping container (WMG 2004e), which is a shorter 

version of the concentrator feed makeup tank IP-2 container, is approximately 15 feet long, 13 feet 

wide, and 14 feet high and weighs approximately 272,000 pounds. The vessel itself, before grout 

was added, weighed approximately 23,486 pounds. (CHBWV 2011a) 

2.6.3 Characterization

As with the concentrator feed makeup tank, details of the waste package characterization 

appear in the waste profile prepared for disposal at the Nevada National Security Site (WVES

2011a) and the associated characterization package (WMG 2011). The characterization process, 

like that for the concentrator feed makeup tank, made use of sample analytical data and the 

average measured dose rate of collimated readings taken with a shielded radiation probe one foot 

from the sides of the installed vessel (1.64 R/h). 

The radioactivity in the melter feed hold tank package was estimated using the same process

as with the concentrator feed makeup tank, using a QAD geometry model and the RADMANTM and 

MegashieldTM computer codes. Table 2-4 shows the estimated residual radioactivity in the melter 

feed hold tank, which totaled 103 Ci as of October 1, 2004. 
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Table 2-4. Melter Feed Hold Tank Total Activity Estimate(1)

Nuclide Activity (Ci) Nuclide Activity (Ci)

C-14 3.98E-04 U-236 2.12E-05

K-40 1.54E-03 U-238 4.23E-05

Mn-54 1.67E-03 Np-237 7.26E-05

Co-60 1.58E-03 Pu-238 9.19E-03

Ni-63 1.89E-02 Pu-239 2.28E-03

Sr-90 5.34E+00 Pu-240 1.74E-03

Zr-95 3.72E-02 Pu-241 5.88E-02

Tc-99 8.34E-04 Pu-242 NA

Cs-137 9.71E+01 Am-241 4.33E-02

Eu-154 3.18E-02 Am-242m NA

Th-228 7.79E-04 Am-243 3.93E-04

Th-230 6.84E-06 Cm-242 3.42E-04

Th-232 7.53E-06 Cm-243 2.84E-04

U-232 9.40E-04 Cm-244 7.36E-03

U-233 3.86E-04 Cm-245 NA

U-234 1.84E-04 Cm-246 NA

U-235 7.07E-06

LEGEND: NA = not available. 

NOTES: (1) From WMG 2011 as of October 1, 2004. 

As with the concentrator feed makeup tank, calculations included in the waste profile show that 

the plutonium 239 equivalent gram estimate in the melter feed hold tank disposal package falls 

well below the individual package limit of 300 (CHBWV 2011a). 

2.6.4 Characterization Quality Assurance 

The characterization of the melter feed hold tank waste package, like the characterization of 

the concentrator feed makeup tank waste package, was consistent with the WVDP characterization 

process described in the Characterization Management Plan for the Facility Characterization Project

(Michalczak 2004). 

The data used in the characterization were supplied by WVNSCO. The dose rate data used 

were collected for this purpose in February 2004 by WVNSCO using the same instrument used for 

the concentrator feed makeup tank (WVNSCO 2004c). The analytical data came from analysis of six 

samples (WMG 2011), which were analyzed by the onsite Analytical and Process Chemistry 

Laboratory. And the information on the melter feed hold tank characterization in the WMG 

characterization report was also independently reviewed by the site contractor and changes from 

this review were incorporated (WMG 2011). 

2.6.5 Consideration of Data Variability and Results Uncertainty

As with the concentrator feed makeup tank, the characterization made use of the average 

value of dose rate measurements to calculate the amount of Cs-137 present in the vessel, and 

geometric means of sample analytical data to calculate scaling factors used to estimate the 

amounts of the other radionuclides present. To account for uncertainties in the radionuclide activity 

estimates, the Nevada National Security Site waste profile radiological technical basis document 
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(CHBWV 2011a) identifies high and low activity ranges that that are plus 20 percent and minus 20 

percent, respectively, of the final waste form activity concentrations, which are based on the 

estimates in the characterization report (WMG 2011) and shown in Table 2-4

21

. 

A brief discussion of data variability and data uncertainty follows.

Consideration of Data Variability 

The dose rates measured on the concentrator feed makeup tank varied with location, ranging 

from 1.35 R/h to 2.39 R/h (WVNSCO 2004c). An average value of 1.64 R/h was used in the 

calculation, as noted previously. As with the concentrator feed makeup tank, the variations in 

measured dose rates are consistent with expectations. The use of an average dose rate value in 

the calculation is considered to be appropriate since it accounts for the observed variations. 

Analytical data used in characterization came from four of the samples used in characterization 

of the concentrator feed makeup tank – two samples of batch 72 taken at different times, one 

sample from batch 74, one sample from batch 77 – along with two glass shard samples taken from 

the two evacuated canisters used to remove molten glass from the vitrification melter (WMG 

2011)22. The batch samples were the same ones used in the concentrator feed makeup tank 

characterization. The geometric means of the measured radionuclide concentrations in these 

samples were used to produce the scaling factors related to Cs-137 that were used in the 

characterization (WMG 2011). 

The radionuclide concentrations and distributions in the different samples are somewhat 

different, as can be seen from the table in the report, a condition attributed mainly to differences 

among the various waste batches23. However, as with the concentrator makeup tank, it is 

appropriate to consider different waste batches and the use of the geometric means of the 

concentrations from the different samples for estimating the scaling factors is considered to be 

appropriate because this approach takes into account the differing radionuclide distributions. 

Consideration of Data Uncertainty

The use of the average value of nine dose rate measurements taken in various locations on the 

concentrator feed makeup tank (WVNSCO 2004c) minimizes the uncertainty associated with the 

Cs-137 calculation. 

Regarding the analytical data used to develop scaling factors, the uncertainties associated with 

the batch sample results were relatively small as shown in Table 2-3 as discussed previously. Table 

2-5 shows uncertainties in Cs-137 and Am-241 for the three analyses of each of the glass shard 

samples from the evacuated canisters. 

 

21 DOE would expect to use the same uncertainty range in the WCS waste profile if the vessel waste packages were 

to be sent to that facility for disposal. 

22 This information appears in the report (WMG 2011) in a table labeled “Analysis of Multiple Sample Data Sets 

(SCAL)” marked MFHT. Two evacuated canisters were used to remove molten glass from the melter cavity at the 

conclusion of the vitrification program as discussed later in this draft evaluation.

23 This table also appears in Attachment B to the Nevada National Security Site waste profile (CHBWV 2011a). It 

shows measured radionuclide concentrations from analysis of seven samples. As with the concentrator feed makeup 

tank, the radionuclide concentration data shows obvious variations in radionuclide distributions.
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Table 2-5. Measured Radionuclide Concentrations and Associated Uncertainty(1) Sample 04-0073

Analysis Concentration (Ci/g) Uncertainty (Ci/g) % Uncertainty

Cs-137

1 2.19E+03 5.58E+01 3

2 2.26E+03 6.08E+01 3

3 2.46E+03 6.78E+01 3

Am-241

1 1.49E+00 8.98E-02 6

2 1.51E+00 9.08E-02 6

3 1.48E+00 8.90E-02 6

Sample 04-0074

Cs-137

1 2.47E+03 6.66E+01 3

2 2.50E+03 6.69E+01 3

3 2.40E+03 6.62E+01 3

Am-241

1 2.22E+00 1.33E-01 6

2 1.97E+00 1.18E-01 6

3 1.27E+00 7.64E-02 6

NOTE: (1) From WVNSCO 2004a (sample 04-0073) and WVNSCO 2004b (sample 00-0074). 

As can be seen from these representative examples, the maximum amount of uncertainty in 

the individual measurements is three percent for Cs-137 and six percent for Am-241. The use of 

the average of the three measurements minimizes the uncertainty associated with the analytical 

data used in calculating the scaling factors related to Cs-137. 

Conclusions

Given the approach used in the characterization, and the uncertainty in the analytical data, 

DOE concludes that uncertainty associated with radionuclide estimates is likely bounded by the 20 

percent concentration range estimates included in the Nevada National Security Site waste profile 

(CHBWV 2011a). 

2.6.6 NRC Independent Assessment

As explained in previously, NRC representatives concluded in an assessment performed in 

connection with two 2004 monitoring visits that the melter feed hold tank was appropriately 

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

requirements (NRC 2004). The NRC representatives evaluated the characterization and waste 

profile methodologies, the design and fabrication of this waste package, and verification that the 

package was 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 independently confirmed the validity of the 

characterization process used by DOE for the melter feed hold tank. 

2.7 WVDP Waste Management Plans

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

concentrator feed makeup tank and the melter feed hold tank.

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 35073

(June 16, 2005)), DOE decided that, for WVDP LLW and mixed LLW that is currently in storage at 
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the site or that will be generated at the site over the next ten years (i.e., through 2015), DOE will 

ship such WVDP LLW and mixed LLW offsite for disposal, in accordance with all applicable 

requirements, at 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 waste-incidental-to-reprocessing determination using the evaluation process (this 

draft evaluation, for example). 

As noted previously, 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 concentrator feed makeup tank and the melter feed hold tank. The 

analysis noted that the these vessels 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 2010). DOE plans to ship the concentrator feed 

makeup tank and the melter feed hold tank waste packages to an offsite LLW disposal facility. For 

the purposes of this draft 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 vessel waste packages 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.24 

Section 6 demonstrates that the concentrator feed makeup tank and the melter feed hold tank 

waste packages do not exceed concentration limits for Class C LLW. 

 

24 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 subject vessels in the WCS 

facility. 
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3.0 WASTE DETERMINATION CRITERIA

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 Manual, applies to WVDP 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 evaluation for 

the subject vessels is being performed in accordance with DOE Manual 435.1-1. DOE’s waste 

determination criteria are described in Section 3.2.

3.2 Applicable Waste Determination Criteria 

Section I.1.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 waste determination. 

Section Purpose

The purpose of this section is to describe the criteria applicable to this draft wasteincidental-to-reprocessing evaluation.

Section Contents 

This section provides brief background information on Department of Energy criteria

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

considered, then describes the Department’s criteria that apply to management of the 

concentrator feed makeup tank and the melter feed hold tank. 

Key Points

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

Waste Management Manual. 
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DOE Manual 435.1-1, Section II.B.2(a), lists three criteria to demonstrate, using the evaluation 

method, that wastes resulting from spent nuclear fuel reprocessing are not HLW and should be 

managed as LLW: 

“(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) [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) [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.”

25

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

the subject vessels against these criteria, and, for the reasons presented, the draft evaluation 

shows that the vessel waste packages meet the applicable criteria and can be managed and 

disposed of as LLW.

 

25 DOE did not authorize alternative requirements for the subject vessels.
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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 

concentrator feed makeup tank and the melter feed hold tank) have 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 

these vessels 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 

subject vessels to the maximum extent that is technically and economically 

practical. 

 The key radionuclides in the concentrator feed makeup tank and the melter 

feed hold tank are those long-lived and short-lived radionuclides listed in 

Tables 1 and 2 of the Nuclear Regulatory Commission’s regulations at 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 subject vessels with demineralized water under high 

pressure was the only method technically practical. 

 The two vessels were flushed with demineralized water under high pressure, 

which proved to be effective in removing key radionuclides.

 Other flush solutions that passed through the vessels also likely removed 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 demonstrated that further efforts to remove key 

radionuclides would have increased costs and worker radiation dose without 

resulting benefits. 
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The first criterion of DOE Manual 435.1-1, Section II.B.2(a) is evaluated in this section. It states:

“[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 evaluation. 

4.1.1 Introduction

The key radionuclides in this draft waste-incidental-to-reprocessing 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 435.1-1 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 61.5526; 

 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 subject vessels 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, I-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 

 

26 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 61.5527 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 435.1-1 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/m3)

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

I–129 0.08

Alpha Emitting Transuranic (TRU) nuclides with half-life 

greater than 5 years 100(1)

Pu–241 3,500(1)

Cm–242 20,000(1)

NOTES: (1) These values are in units of nanocuries per gram.

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

Radionuclides 

Concentration (Ci/m3)

Column 1

[Class A]

Column 2

[Class B]

Column 3

[Class C]

Total of all nuclides with less than 5 y half-life 700 (1) (1)

H–3 40 (1) (1)

Co–60 700 (1) (1)

Ni–63 3.5 70 700

Ni–63 in activated metal 35 700 7,000

Sr–90 0.04 150 7,000

Cs–137 1 44 4,600

NOTE: (1) There are no limits established for these radionuclides 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.

 

27 Title 30 of the Texas Administrative Code has similar requirements (Rule §336.362, Appendix E).
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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 

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 II).

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 13728. 

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 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.

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.

29

 

28 The approximate half-lives of these radionuclides are as follows: strontium 90, 28 years; cesium 137, 30 years; 

and technetium 99, 212,000 years (HEW 1970). 

29 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 two vessels) 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. 
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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-23430, U-238, and Pu-239 for the 

Nevada National Security Site Area 5 waste disposal area, based on the following analyses (NST 

2012b):

 All-pathway analyses which show that the radionuclide that would contribute most 

significantly to dose to a member of the public in the controlling scenario (resident farmer) 

are Tc-99 (79 percent), Pb-210 (13 percent), and U-238 (3 percent), with Pb-210 produced 

predominately by decay of U-234 in the waste at the time of disposal;

 Intruder analyses which show that the radionuclides that would contribute most 

significantly to dose to an inadvertent intruder in the chronic agriculture scenario are Tc-99 

(70 percent), U-238 (13 percent); and Th-229 (5 percent);

 Intruder analyses which show that the radionuclides that would contribute most 

significantly to dose to an inadvertent intruder in the acute construction scenario are U-238 

(36 percent), Th-229 (25 percent), Pu-239 (8 percent), U-233 (7 percent), and U-234 (7 

percent).

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 2012) 

identifies total radioactivity limits for three radionuclides for disposal in the Federal Facility Waste 

Disposal Facility: C-14, Tc-99, and I-129. These radionuclides (which also are included in Table 1 of 

10 CFR 61.55) contribute most to predicted dose according to the WCS performance assessment 

submitted with the WCS license application (WCS 2007) and are therefore important to meeting the 

performance objectives for the WCS facility. The WCS performance assessment is discussed in 

Section 5.231. 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.

4.1.6 Conclusions About Key Radionuclides in the Subject Vessels

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 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.

 

30U-234 present at the time of disposal is the predominant source of Pb-210 (NST 2012b). 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. 

31 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 89, TCEQ 2012).
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Table 4-3. Key Radionuclides for this Draft Evaluation

Radionuclide 10 CFR 61.55 LongLived Radionuclides

10 CFR 61.55 ShortLived RadionuclidesRadionuclides 

Important to PA 

H-3 X

C-14 X X

(1)

Co-60 X

Ni-59 X

Ni-63 X

Sr-90 X

Nb-94 X

Tc-99 X X

(1)(2)

I-129 X X

(1)

Cs-137 X

Th-229 X

(2)

U-233 X

(2)

U-234 X

(2)

U-238 X

(2)

Np-237(3) X

Pu-238(3) X

Pu-239(3) X X

(2)

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

(NST 2012b). 

(3) Alpha emitting transuranic radionuclides with half-life greater than five years (NRC 1982, Table 4.2).

4.2 Removal to the Maximum Extent Technically and Economically Practical

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 (DOE 2006).
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Because of the close relationship between the two vessels, this section addresses both 

together. Section 4.2.1 describes the technical practicality assessment of methods that might have 

been used to remove key radionuclides from the vessels. Section 4.2.2 discusses the method 

actually used and its effectiveness. Sections 4.2.3 and 4.2.4 discuss two other methods considered 

that were not technically practical. Much of the information in this section is drawn from three

technical documents: 

 A 2002 report on flushing of the melter feed hold tank (WVNSCO 2002b),

 A 2002 report describing the equipment and methods used and the results of this 

decontamination program (WSNSCO 2002c), and

 The HLW Processing System Flushing Operations Run Plan (WVNSCO 2002a).

The Meaning of the Phrase “Technically and Economically Practical” 

Removal to the maximum extent “technically and 

economically practical” is not removal to the extent 

“practicable” or theoretically “possible.” Nor does the 

criterion connote removal which may be notionally 

capable of being done. 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. Such a

situation may arise if certain radionuclides are 

present in such extremely small quantities that 

they make an insignificant contribution to 

potential doses to workers, the public, and the 

hypothetical human intruder. 
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4.2.1 Technical Practicality Assessment

The WVDP considered various proven decontamination technologies that might be suitable for 

removing key radionuclides from the two vessels, including those discussed in DOE’s 1994 

Decommissioning Handbook (DOE 1994). This handbook discusses 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 were 

developed with the support of the DOE Office of Science and Technology (now the DOE Office of 

Science) as part of a continuing program to improve methods used in decontamination and 

decommissioning work. 

DOE’s 1994 Decommissioning Handbook identifies advantages and disadvantages of 

decontamination technologies in various applications. For the subject vessels, the technologies 

proven to be effective in decontaminating pipes and tanks were the ones of primary interest. Table 

9.2 of the Handbook showed the following technologies to be highly effective in this application: 

(1) ultra-high pressure water, (2) grit blasting, (3) flushing with water, and (4) hydroblasting32. 

Grit blasting did not show promise in this application owing to potential for the grit material 

interfering with the vitrification process and resulting in disposal difficulties. The other three 

methods are generally similar except for the water pressure used. High-pressure spray, a

combination of the other three methods, was selected for evaluation and actually used as the 

primary decontamination method, as discussed below. 

The methods evaluated in detail were: 

 Flushing vessel internals with water using high-pressure spray, 

 Mechanical decontamination using a ball mill, and

 Chemical decontamination. 

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 practical. 

Prior to decontamination, the insides of the two vessels were expected to be relatively clean in 

the lower two thirds, with the upper portions coated with residual dried HLW slurry. Dead spaces 

inside the nozzles of the concentrator feed makeup tank had been observed to be coated with 

slurry. Dried slurry was also expected on the stiffeners of the head of the melter feed hold tank, 

which are illustrated in Figure 4-1. (WVNSCO 2002a) 

The high-pressure spray method is discussed first, followed by the other alternatives that were 

evaluated but not used. While the high-pressure spray method was the primary method used to 

removal key radionuclides from the vessels, it was not the only method that was used as discussed 

in Section 4.3. 

4.2.2 In-Place Decontamination By Flushing With Water Using High-Pressure Spray

The concentrator feed makeup tank and the melter feed hold tank were decontaminated by 

flushing with demineralized water before they were removed from the Vitrification Facility. This 

flushing was performed in connection with flushing of other HLW processing systems (WVNSCO 

2002a). The process involved use of high-pressure water spray to clean the inner surfaces of the 

 

32Other decontamination processes described in The Decommissioning Handbook (ASME 2004) were considered in a 

retrospective fashion when this evaluation was prepared. It is unlikely that any of these other processes – which 

include soft-media blasting and steam vacuum cleaning – would have been more effective than the high-pressure 

water spray actually used, which was very effective based on the visual inspection results and dose rate reductions 

that are discussed later in this subsection and shown in Figure 4-2.
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two vessels. It was completed in 2002 prior to shutdown of the vitrification melter so the material 

collected during the flushing could be vitrified for disposal as HLW, which was done to the extent 

practical.

33

Flushing Equipment Design and Flushing Performed

Substantial development work was performed to determine the optimum process for spraying 

the vessels interiors with high-pressure water (WVNSCO 2002f). This flushing would focus on the 

upper third of each vessel.

Special fixtures to position the spray nozzles inside the upper portions of the vessels were 

developed and tested on a mockup. Figure 4-1 shows spray nozzle positions considered. A separate 

spray wand for decontamination of the vessel exteriors was designed, built, and tested. Tooling 

was also developed to clean the vessel nozzles using rotary driven wire brushes. However, these 

tools were not used because the water spray proved to be effective in nozzle areas, as well as in 

other parts of the vessels. (WVNSCO 2002f)

Pressurized water at approximately 1,000 

pounds per square inch was delivered at 25 to 

35 gallons per minute though the Gamajet®

rotary spray nozzle. The nozzle was inserted 

through one or more openings in the vessel 

heads to ensure complete coverage of internal 

surfaces. Each complete spray cycle took 

approximately 24 minutes. (WVNSCO 2002f)

The concentrator feed makeup tank was

flushed three times using this process and the 

melter feed hold tank was also flushed three 

times using this process. Each such flush 

involved two complete spray cycles that 

together took 55-60 minutes to accomplish. 

(WVNSCO 2002f)

The resulting flush solutions were concentrated in the concentrator feed makeup tank, 

combined with other flush solutions that were also concentrated and then diluted with glass 

formers and, when this mixture met specifications, transferred to the melter feed hold tank for 

feeding to the melter. For flushing the melter feed hold tank, that vessel was used as its own catch 

tank, after which the flush solution was transferred to the concentrator feed makeup tank for 

evaporation/concentration and subsequent sampling. After this material was combined with other 

flush solutions and glass formers, it was transferred back to the melter feed hold tank for feeding 

to the melter. (WVNSCO 2002a)

The various flush solutions that were used for other equipment in the Vitrification Cell and the 

waste tank farm, including nitric acid as well as water, were transferred into the concentrator feed 

makeup tank and the melter feed hold tank, with these liquids also having a decontamination effect 

by loosening deposits (WVNSCO 2002a). In addition, the external surfaces of the two vessels were 

washed down using high-pressure demineralized water from a spray wand, which was suspended 

 

33 In late 2000, DOE commissioned a Vitrification Completion Team composed of representatives from DOE, 

NYSERDA, West Valley Nuclear Services, and NRC to review issues surrounding the ability to complete vitrification 

operations (VCT 2001). This team developed an approach to retrieving waste from the underground waste tanks, 

washing and characterizing the residual tank materials, and flushing the vitrification system, including the subject 

vessels, prior to completing a controlled shutdown of the melter. 

Figure 4-1. Melter Feed Hold Tank Head Showing 

Potential Spray Nozzle Positions. (Two positions were 

actually used in the flushing.) (From WVNSCO 2002c)
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from the cell crane. Flushing results were monitored primarily by visual inspections and dose rate

measurements. (WVNSCO 2002f)

Visual Inspection Results

A camera with underwater lighting was used to obtain still images of the vessel internal 

surfaces to determine the effectiveness of the flushing. Figure 4-2 shows before and after flushing 

conditions in the two vessels.

Figure 4-2. Before and After Flushing Images of the Two Vessels. (The shiny 

surfaces in the after images illustrate the effectiveness of the high-pressure washing 

in these WVDP photos taken with the underwater camera.) (from WVNSCO 2002c)
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Inspection before flushing showed substantial amounts of dried slurry in both vessels as can be 

seen in Figure 4-2. Inspection after flushing completion showed essentially no visible deposits, with 

visibility sufficient to show fabrication weld beads and threads on bolts.

Dose Rate Reduction and Reduction in Residual Radioactivity

Radiation detectors positioned using special fixtures to monitor decontamination progress 

showed that flushing reduced dose rates, although the amount of dose rate reduction was masked 

to some degree by other radiation sources in the cell, which included filled HLW canisters. Dose 

rates measured near the head of the concentrator feed makeup tank dropped from 200 to 8 R/h. 

Dose rates near the melter feed hold tank dropped from 250 to 22 R/h34. 

An estimate of the flushing effectiveness can also be made by comparing the estimated 

residual activity in each vessel before flushing and after all of the flushing was completed. Table 4-

4 shows estimated residual cesium 137 in the two vessels before flushing and after completion of 

flushing, when the vessels were drained to the extent practical.

Table 4-4. Vessel Flushing Effectiveness in Terms of Estimated Cs-137 Removal

Condition

CFMT 

Remaining 

Inventory (Ci)

CFMT 

Reduction 

Factor (%)

MFHT 

Remaining 

Inventory (Ci)

MFHT 

Reduction 

Factor (%)

Before Flushing(1) 1,500 NA 1,200 NA

After All Flushes 96.5(2) 93 103(3) 91

LEGEND: CFMT = concentrator feed makeup tank, MFHT = melter feed hold tank, NA = not applicable

NOTES: (1) The activity in each vessel before flushing began was estimated in the following manner: (a) the dried slurry 

coating observed on the vessel interior surfaces before flushing was assumed to average 0.250-inch 

thickness over the upper one-third of the vessels, a conservatively low value based on pre-flush visual 

inspection results; and (b) the Cs-137 concentration in this material was assumed to the same as that in the 

last batch of slurry sent to the concentrator feed makeup tank before the vitrification system flushing began

(batch 75), which had the second highest Cs-137 concentration (1.16E+04 Ci/cm3) among the feed material 

batches (Kurasch 2011). 

(2) From Table 2-2 above.

(3) From Table 2-3 above. 

The estimates in Table 4-4 should be considered to be order-of-magnitude estimates. As can 

be seen in Figure 4-2, the sludge buildup in some areas was much greater than 0.250 inch. The 

 

34The pre-flush dose rates were recorded on May 1, 2002 (WVNSCO 2002b). The post-flush dose rates were recorded 

on July 15, 2002 after all of the high-pressure spray flushing had been completed (WVNSCO 2002d). 

NOTE

Additional dose rate measurements made in February 2004 (WVNSCO 2004c) 

showed lower levels, with a maximum of 2.25 R/h on the concentrator feed 

makeup tank and a maximum of 2.39 R/h on the melter feed hold tank. The 2004 

data, which were used in vessel characterization (WMG 2011), are indicative of a 

decontamination factor for the flushing of 89 for the concentrator feed makeup 

tank and 105 for the melter feed hold tank. (For the concentrator feed makeup 

tank, the decontamination factor is based on the initial measurement of 200 R/h 

divided by the final measurement of 2.25 R/h and, for the melter feed hold tank, 

the initial measurement of 250 R/h was divided by the final 2.39 R/h.) 
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area values used in the estimates did not include the areas of the internal baffles. These factors 

suggest that the before-flushing estimates in Table 4-4 are low. 

Evaluation of conditions before and after flushing and considering the measured dose rates 

and the estimates on Table 4-4 indicate that flushing removed more than 90 percent of the residual 

cesium 137 inside the concentrator feed makeup tank and the melter feed hold tank. That is, the 

overall decontamination factor for the flushing performed – the “direct” flushes using the highpressure spray apparatus and the “indirect” flushes associated with other vitrification flush solutions 

passing through the two vessels – was of the order of 10 based the estimated residual sludge 

removal or 100 based on the reduction in measured dose rates, with the latter considered to be a 

more accurate estimate due to the uncertainty in the sludge removal estimates. 

The flushing Plan (WVNSCO 2002a) identified the expected conditions after the flushes for both 

vessels as “Dried slurry deposits are expected to be removed from the surfaces accessible to the 

spray head, with some removal from protected areas.” The significant differences between the 

before and after surface conditions inside the vessels, combined with the dose rate reductions and 

the residual radioactivity reductions summarized in Table 4-4, demonstrate that these objectives 

were achieved and that the flushes were effective in removing key radionuclides. 

Additional information on the overall effectiveness of the various flushes in reducing residual 

radioactivity in the concentrator feed makeup tank and the melter feed hold tank is provided in 

Section 4.3 below, including a table that summarizes all of the flushes performed – the “direct 

flushes” using the high-pressure spray, the “indirect flushes” associated with other vitrification 

facility flush solutions that passed through the vessels, and additional flushes that were performed 

prior to removal of the vessels in 2004. 

4.2.3 Mechanical Decontamination Using a Ball Mill 

This process uses a combination of grinding and impact in a ball mill to separate radioactive 

contamination from the base material.35 It 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) 

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. (WVNSCO 2001) 

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. The WVDP concluded that the process could be optimized, but that this effort would change 

the proven technology to an experimental one, resulting in another disadvantage. The WVDP 

determined that this process was not technically practical because of these disadvantages

 

35 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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(WVNSCO 2001). Given this conclusion, details of how to use the process on the two vessels and 

how to dispose of the resulting waste stream were not developed. 

4.2.4 Chemical Decontamination

This method uses a chemical solution to dissolve the slurry material or leach radionuclides from 

the glass or slurry. 

 Like ball milling, this process was evaluated for potential use in decontamination of small 

pieces of equipment and other materials contaminated with HLW during the vitrification process. 

Evaluation included laboratory tests where specimens consisting of Inconel and stainless steel 

plates with glass annealed to their surfaces and pieces of pipe containing dried slurry were placed 

in various chemical solutions at different temperatures. Solutions tested included nitric acid, 

hydrofluoric acid, oxalic acid, sodium hydroxide, and water. (WVNSCO 2001) 

These tests demonstrated that chemical dissolution could be an effective decontamination 

process for the equipment. However, the presence of the chemicals could have been incompatible 

with the recipe for producing an acceptable borosilicate glass matrix in the vitrification process. If 

this were to happen, the resulting out-of-specification HLW product would have had no approved 

pathway for disposal. This overriding disadvantage made chemical decontamination an 

unacceptable approach from a technical standpoint.36

4.3 Economic Practicality Assessment

The assessment of the economic practicality of further radionuclide removal focused on the 

costs and benefits of performing additional flushing before shutdown of the vitrification system 

because the flushing process used was the only method among those evaluated for removal of key 

radionuclides that was determined to be technically practical. Section 4.2.2 showed that the

flushing process performed:

 Removed more than 90 percent of the estimated residual radioactivity in the concentrator 

feed makeup tank, leaving approximately 96.5 curies; 

 Removed more than 90 percent of the estimated residual radioactivity in the melter feed 

hold tank, leaving about 103 curies; and

 Removed visible dried slurry from the surfaces of both vessels.

4.3.1 Economic Practicality of Additional Flushing 

The assessment considered the benefits and costs of one additional flushing cycle for each 

vessel. For conservatism, it assumed that this additional flushing cycle would have removed 90

percent of the activity remaining in that vessel at the conclusion of the flushing that was actually 

performed37. To help establish the context for the cost-benefit assessment, this section begins with 

a more detailed discussion of the flushes performed and their effectiveness in removing key 

radionuclides. 

Direct and Indirect Flushes

Table 4-5 summarizes the flushing of the two vessels, both direct flushes using the highpressure spray and indirect flushes that involved temporary storage and agitation of vitrification 

 

36 Note that the concentrator feed makeup tank was actually flushed with sodium hydroxide solution before removal 

as discussed in Section 4.3 below, but the flush solution was sent to Tank 8D-4 since the vitrification system had 

already been shut down. 

37 Ninety percent effectiveness would have been very unlikely in these circumstances, given the extensive previous 

flushing which removed all of the visible sludge from both vessels.
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system flush liquid of decreasing radioactive concentrations or additions of uncontaminated water. 

The table also includes a final flush of the concentrator feed makeup tank that was performed just 

prior to removal of the vessels from the vitrification facility in 2004. The batch numbers refer to 

batches of material made up in the concentrator feed makeup tank. 

A total of 66 batches of HLW slurry numbered 10 through 75 were sent to the concentrator 

feed makeup tank to be prepared for vitrification, along with two batches of lower-activity 

vitrification system decontamination solutions identified as batches 76 and 77. Batch 76 contained 

about 51 percent of the radioactivity in batch 75 and batch 77 contained about 11 percent of the 

radioactivity in batch 75 (Kurasch 2011). 

Table 4-5. Summary of Direct and Indirect Flushes Performed (1)

Flush 1

(indirect)

Batch 76, which included material from acid flush of Tank 8D-4, stored in the vessel and 

periodically stirred for five months between January 2 and June 1 of 2002. After the 

prepared batch 76 material was transferred from the CFMT to the MFHT, 2,000 liters 

(530 gallons, at least 10 inches in the vessel) of water were added to the CFMT and 

transferred to the MFHT.

Flush 2

(direct)

Spray nozzle deployed inside CFMT on May 2, 2002 per WO 63499 through Nozzle F, 

operated for two cycles (55-60 minutes).(2) Dislodged material becomes part of batch 76 

material.

Flush 3

(indirect)

After transfer of batch 76 from the CFMT to the MFHT, at least 530 gallons (2,000 liters) 

of uncontaminated water transferred to CFMT and on May 31, 2002 was sent to 

MFHT.(4)

Flush 4

(indirect)

Batch 77 periodically stirred by the CFMT’s agitator for two months (June 5, 2002 

through July 31, 2002).

Flush 5

(direct)

Spray nozzle deployed inside CFMT on June 14, 2002 per WO 65670 through Nozzle F, 

operated for two cycles.(5) Dislodged material became part of batch 77 material.

Flush 6

(direct)

Spray nozzle deployed inside CFMT on June 18, 2002 through Nozzle F, operated for two 

cycles. Dislodged material became part of batch 77 material.

Flush 7

(indirect)

Water-diluted MFHT Heel after airlift(6) 16 from the final production canister WV-412 and 

the MFHT water flush from September 10, 2002 transferred to the CFMT on September 

10, 2002, raising the level in the CFMT from 58 inches to 86 inches. This dilute liquid 

stirred by the CFMT agitator periodically between September and December 2002, while 

the liquid level in the CFMT is reduced to about 34 inches.

Flush 8

(direct)

In late December 2003, with vessel level at about 34 inches, about 55 gallons of 50% 

sodium hydroxide solution and 30 gallons of non-radioactive water added to CFMT. 

Agitator operated for 30 minutes and this material remained in vessel for about one 

month(7)

Flush 9

(direct)

In late January 2004, the CFMT agitator was run for at least one hour and the CMFT 

contents transferred to Tank 8D-4. About 200 gallons of utility water transferred to the 

CFMT, the agitator operated for at least one hour, and the vessel contents transferred to 

tank 8D-4, leaving a heel of about 55 gallons.(8) 

Concentrator Feed Makeup Tank
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Table 4-5. Summary of Direct and Indirect Flushes Performed (1) (continued)

Flush 1

(indirect)

Batch 76 transferred from the CFMT to the MFHT, filling the MFHT to the 101.7 inch 

level. Batch 76 material periodically stirred and fed to the melter for 16 days until June 

14, 2002. 

Flush 2 

(direct)

Spray nozzle deployed inside MFHT Nozzle E on July 8, 2002 per WO-65708 and 

operated for two cycles. Decontamination solution collected and transferred to the CFMT 

to become part of batch 77.(3)

Flush 3

(direct)

Spray nozzle deployed inside MFHT Nozzle E on July 11, 2002 per WO-65708 and 

operated for two cycles.(3) The resulting decontamination solution transferred to CFMT 

became part of Batch 77.

Flush 4 

(direct)

Spray nozzle deployed inside MFHT Nozzle R on July 11, 2002 per WO-65708 and 

operated for two cycles.(3) Resulting decontamination solution transferred to CFMT 

became part of batch 77. 

Flush 5

(indirect)

Batch 77 material transferred from CFMT to MFHT on July 31, 2002, filling MFHT to the 

98 inch level. Batch 77 material periodically stirred and fed to the melter for 10 days 

(August 5 through August 14, 2002.) 

Flush 6

(indirect)

Water added continuously to the MFHT after airlift 8 of the last production canister, 

diluting the batch 77 material for final airlifts 9 through 16. These final 8 airlifts take 30 

hours during which water is added and the contents are periodically stirred by the MFHT 

agitator. This was completed on August 14, 2002.(9)

Flush 7

(direct)

On September 10, 2002, MFHT heel transferred to the CFMT. MFHT subsequently filled 

with 22 inches of water and periodically stirred for a minimum of two hours, after which 

contents transferred to the CFMT that same day. Afterwards, the MFHT level instrument 

showed that the vessel was empty.

LEGEND: WO = work order

NOTES: (1) This table was compiled from vitrification system records, including the completed work orders listed in notes

(2), (3), (4), and (6).

(2) WVNSCO 2002b. 

(3) WVNSCO 2002d. 

(4) WVNSCO 2002i.

(5) WVNSCO 2002c.

(6) Air pressure was used to raise the molten glass level in the vitrification melter discharge cavity to send the 

molten glass into the waiting HLW canister. This process was called airlifting. 

(7) WVNSCO 2003.

(8) WVNSCO 2004d.

(9) WVNSCO 2002h.

(10) Water and dislodged waste from spraying the inside of the concentration feed makeup tank became part of 

the material batch inside the vessel at the time the spray was performed. 

As can be seen in Table 4-5, the concentration feed makeup tank received three separate, twocycle flushes with high-pressure spray, was indirectly flushed four times, was flushed with small 

amounts of caustic solution and water, and, finally, flushed with approximately 200 gallons of 

water in early 2004. The melter feed hold tank received three separate, two-cycle flushes with 

high-pressure spray, was indirectly flushed three times, and, finally, flushed with uncontaminated 

water and the heel from the concentrator feed makeup tank. Figure 4-3 illustrates the direct 

flushes of the two vessels using the high-pressure stray apparatus. Melter Feed Hold Tank
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Figure 4-3. High-Pressure Spray of the Concentrator Feed Makeup Tank and the Melter Feed Hold Tank
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Figure 4-4 illustrates how liquid levels in the concentrator feed makeup tank varied in the first 

seven months of 2002 when batches 76 and 77 were in the vessel and the direct flushes and some 

of the indirect flushes were taking place. The figure shows wide variations in the liquid levels over 

this period. Note that the increases in the figure were from incoming material transfers and the 

decreases were from evaporation, except for the identified transfers to the melter feed hold tank. 

Based on the results of the visual inspections, the reductions in dose rates, and the reductions 

in estimated residual radioactivity discussed previously, it is clear that the direct and indirect 

flushes effectively removed key radionuclides from the two vessels. The high-pressure spray had 

the greatest effect. The indirect flushes also had some effect.

The overall results show major decreases in residual radioactivity. Figure 4-5 illustrates the 

overall reductions in the maximum measured dose rates and the estimated cesium 137 activity 

resulting from the direct and indirect flushes shown in Table 4-5.

Although the figure plots the reduction in Cs-137, all key radionuclides were removed in 

approximately the same proportions based on the similarities of the radionuclide distribution in the 

material removed from the vitrification melter in the two evacuated canisters38 to the radionuclide 

distribution in batches 74 and 75, the last batches of HLW slurry sent to the concentrator feed 

makeup tank. The measured dose rates are primarily from cesium 137 since it is the dominant

gamma emitting radionuclide in the vessels as can be seen from Table 2-2 and Table 2-3. 

 

38As noted previously, two evacuated canisters were used to remove molten glass from the melter cavity at the 

conclusion of the vitrification program. Comparing glass sample scaling factors to the average of scaling factors from 

batch 74 and batch 75 shows that Sr-90 was removed in essentially the same proportion as Cs-137 and other key 

radionuclides were removed in slightly higher proportions than Cs-137.

 Figure 4-4. Variations in Concentrator Feed Makeup Tank Liquid Levels (WVNSCO 2002h)
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Potential Benefits

The benefits from the hypothetical additional flushes would have been limited for the following 

reasons:

 The dose rates on the outside of the waste package in its present condition are low – a 

maximum of 16 mR/h on contact with the side of the concentrator feed make up tank and 

a maximum of 5 mR/h on contact with side of the melter feed hold tank (WVES 2011b).

Compliance with radiological control program requirements in handing of the waste 

package at the WVDP and the LLW disposal facility will ensure the protection of individuals 

during operations related to disposal as discussed in Section 5.2.4.

 Worker radiation doses would not have been significantly reduced. The external dose rates 

with a reduced amount of residual radioactivity and less shielding would have been 

approximately the same. 

 The flushes actually performed left both vessels in a condition suitable for disposal as LLW 

as discussed in Section 5.3. 

 The potential impacts to the general population from disposal of the vessel waste packages 

at the LLW disposal facility without further decontamination will be negligible as discussed 

in Section 5.2.2, so a further reduction in residual radioactivity would not have been 

beneficial from the standpoint of potential doses to members of the public.

 The potential impacts to an inadvertent intruder from disposal of the vessel waste 

packages at the LLW disposal facility without further decontamination would be negligible 

Figure 4-5. Vessel Flushing Effectiveness 
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as discussed in Section 5.2.3, so a further reduction in residual radioactivity would not have 

been beneficial from the standpoint of an inadvertent human intruder.

A single monetary benefit would have been realized had the hypothetical additional flushes 

been performed: The shielded containers for the vessels could have been designed and constructed 

of lighter weight steel, which would have reduced costs associated with materials, fabrication, and 

transportation by approximately $200,000.39

Costs of Additional Flushing

A total of approximately $1 million in additional direct costs (in 2002 dollars) would have been 

involved for each additional direct flush. 

One additional direct flush cycle for each vessel would have produced approximately 1,400 

gallons of additional liquid to be processed through the vitrification melter. Even though this 

volume would have been reduced by evaporation, processing of the remaining liquid would still 

have required one additional processing cycle which would have produced one more canister of 

vitrification waste from the vitrification melter. One additional flushing and processing cycle would 

have taken about two weeks to complete at a cost of approximately $1 million, based on 

vitrification system operating costs that were running $25 million to $30 million per year40.

Another factor in considering the costs of processing of additional flush solutions was the 

limited vitrification melter service life. By the time the flushes had been completed, the melter was 

near the end of its useful service life. A failure of the melter would have, for all practical purposes, 

stranded radionuclides within the melter.

41

In the interest of conservatism, no attempt was made to quantify other costs associated with 

processing of the additional flush solutions, such as the monetary value of additional worker 

radiation dose that would have been necessary to continue vitrification melter operations.

4.3.2 Summary and Conclusions

The technical practicality assessment showed that flushing the vessels with water using highpressure spray was the only method among the three methods evaluated that would be technically 

practical. This method proved to be effective. The discussion of the economic practicality of 

additional flushing while the vitrification process was still operational showed that the cost of an 

additional high-pressure spray flush would have far outweighed the benefits. 

 

39 This estimate is for a one-time savings. It is a conservative, order-of-magnitude estimate in 2002 dollars for 

savings in raw material, handling, and transportation costs. (Kurasch 2011)

40 In addition, there would be indirect life-cycle costs. Interim onsite storage of one additional canister would have 

cost around $35,000 per year based on actual and predicted annual costs of interim storage, including surveillance 

and maintenance. The estimate does not include the cost of ultimate disposition. (Kurasch 2011)

41 As discussed previously, the WVDP used processing of decontamination solutions and the evacuated canister 

system to remove residual radioactive material from the melter. Those processes were effective in removing residual 

waste from the melter, but could only be used for waste in a molten glass form, before the glass solidified. It would 

not have been feasible to process decontamination solutions or use the evacuated canister system in the event of a 

failure of the melter, because upon melter failure, the residual glass in the melter cavity would solidify, thereby 

precluding use of these processes to clean the melter and effectively stranding the solidified waste in a solid form 

within the melter.
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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 concentrator feed 

makeup tank and the melter feed hold tank waste packages 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 vessel waste packages 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 vessel waste 

packages will meet criteria for disposal as low-level radioactive waste. 

Key Points

 Management of the concentrator feed makeup tank and the melter feed hold 

tank waste packages will meet safety requirements comparable to the 

performance objectives of 10 CFR 61, Subpart C. 

 The concentrator feed makeup tank and the melter feed hold tank waste 

packages 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 concentrator feed makeup tank 

and the melter feed hold tank waste packages 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

The second criterion of Section II.B.2(a) of DOE Manual 435.1-1 is evaluated in this section. 

This criterion reads as follows: 

“[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.”

Section 2 describes the design of the concentrator feed makeup tank and the melter feed hold 

tank and their operational histories. Section 2 also explains how these vessels were characterized 

for residual radioactivity, with Tables 2-2 and 2-3 providing the total activity estimates in the final 

waste form. As noted previously, each vessel has been placed in its own custom made shipping 

container and the void spaces in the vessels and in the shipping containers filled with low-density 

cellular concrete in preparation for transport to an offsite disposal facility. 

This section addresses the second criterion in the following subsections:

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,

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 WCS LLW 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 vessel waste packages meet the Nevada National 

Security Site waste acceptance criteria.

Section 5.4 discusses waste acceptance criteria for the WCS Federal Facility Waste Disposal 

Facility and how it would be established that the vessel waste packages meet these criteria if 

the waste packages were to be shipped to that facility for disposal. 
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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 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 458.1, Radiation Protection of the 

Public and the Environment,

42 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 of the disposal facility, with updates, 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 that 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.

43

 

42 DOE Order 458.1 has cancelled and superseded DOE Order 5400.5 of the same name, which is cross referenced in 

DOE Manual 435.1-1. The technical purpose and scope of the order remain the same. 

43 Appendix A of this draft evaluation demonstrates that the DOE performance objectives provide safety requirements 

comparable to the NRC performance objectives at 10 CFR 61, Subpart C and that 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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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.” 

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 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.

(c) Release of radon shall be less than an average flux of 20 pCi/m2/s (0.74 Bq/m2/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.”

Performance Assessment (PA)

Assures that disposal of LLW will 

meet DOE performance objectives

Composite Analysis (CA)

Assesses impact to a hypothetical member of 

the public from all radioactive sources that 

may interact with LLW disposal facilities

PA and CA Maintenance

Key elements:

 Annual review of disposal activities

 Research and testing

 Special Analyses

 Revisions as necessary

Disposal Authorization Statement

Key elements:

 DOE-Headquarters review and approval of PAs and CAs

 Satisfactory site response to all approval conditions

 Monitoring, Maintenance, and Closure Plans
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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 methodology44. 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.”

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 2012 (NST 2012a). 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 

identifies special analyses that were performed in the previous year45.

As explained in the report of the annual review (NST 2012a), the updated Area 5 performance 

analyses provide 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:

 

44 NRC recommends in NUREG-1573, A Performance Assessment Methodology for Low-Level Radioactive Waste 

Disposal Facilities, 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. 

Appendix B demonstrates that the DOE and NRC dose standards are comparable and shows that the State of Texas 

dose standards mirror those of NRC.

45 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 results 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 of 1.9

millirem estimated to occur at 1,000 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 of 

0.045 millirem to a resident farmer at 1,000 years after facility closure; and

 The average Radon 222 flux density at the surface of the disposal units, which is predicted 

to reach a peak of 4.3 pCi/m2/s 1,000 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. Note that the estimates given are mean 

values. The 95th percentile values are also below the performance objective dose limits (NST 

2012a).

Estimated Impact of Disposal of the Subject Vessels

Disposal of the concentrator feed makeup tank and the melter feed hold tank at the Nevada 

National Security Site Area 5 facility would have a negligible impact, if any, on facility performance

for the following reasons:

 The vessel waste packages meet the waste acceptance criteria, with no radionuclide 

concentrations exceeding the waste acceptance criteria action levels as shown in Table 5-1 

below. 

 Screening of the waste profile (CHBWV 2011a) by the Nevada National Security Site has 

shown that no special analysis is required for this waste stream, confirming that the vessel 

waste packages do not have the potential for altering disposal site performance 

assessment assumptions, conceptual models, or performance assessment results.

46

A waste profile package for the subject vessels, which describes their characteristics 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 September 2011 (CHBWV 2011a). This

waste profile was reviewed by the Nevada National Security Site Waste Acceptance Review Panel, 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 vessel waste packages for disposal, conditioned upon a determination that the 

waste packages are LLW in accordance with the waste-incidental-to-reprocessing criteria in DOE 

Manual 435.1-1 (that is, this draft evaluation resulting in a final evaluation and determination that 

the vessel waste packages can be managed as LLW) (DOE 2011). Section 5.3 below discusses the 

waste acceptance process in more detail.

 

46Note that a special performance assessment was performed for disposal of the vitrification melter waste package at 

the Nevada National Security Site Area 5 facility, which showed that the melter waste package would have a 

negligible impact on performance of the closed facility (DOE 2010). The vitrification melter waste package has been 

estimated to contain approximately 4,570 curies as of October 2004 (WMG 2004a). The concentrator feed makeup 

tank and melter feed hold tank waste packages are each about the same size as the vitrification melter waste 

package and each contains only about two percent of the activity in the melter waste package, with similar 

radionuclide distributions. Given this situation, it is evident that the impact of disposal of the two vessel waste 

packages will have negligible, if any, impact on performance of the Nevada National Security Site Area 5 facility since 

the vitrification melter waste package that contains much more radioactivity has been determined to have a 

negligible impact on performance of this disposal facility. 
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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) and 

completed an updated performance assessment in September 2011 (WCS 2011). The updated 

performance assessment of the Federal Facility Waste Disposal Facility considered a total inventory 

at closure of 26 million cubic feet of waste with 5.6 million curies of radioactivity. It included the 

following estimated dose for the post-institutional control period: a maximum annual dose of 

0.0064 millirem per year to an adjacent resident, a small fraction of the 25 millirem per year limit47. 

The updated performance assessment (WCS 2011) was developed in compliance with a license 

condition that requires WCS to prepare an updated performance assessment prior to accepting 

waste for disposal and annually thereafter to demonstrate that performance objectives will be met 

(TCEQ 2011).

Consideration of the license limitations and the available performance assessment results for 

the WCS Federal Facility Waste Disposal Facility suggests that disposal of the concentrator feed 

makeup tank and melter feed hold tank waste packages at that facility would have negligible, if 

any, impact on disposal facility performance for the following reasons:

 As shown in Table 5-2 below, the activity in each of the vessel waste packages – 96.5 

curies for the concentrator feed make up tank and 103 curies for the melter feed hold tank 

– would amount to only about 0.002 percent of the license limit for the Federal Facility 

Waste Disposal Facility.

 The 5.6 million curies in 26 million cubic feet of waste considered in the performance 

assessment would have an average activity density of 0.22 Ci per cubic foot, much higher 

than the activity density of the concentrator feed makeup tank waste package 

(approximately 0.03 Ci per cubic foot) and the melter feed hold tank waste package 

(approximately 0.04 Ci per cubic foot).

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.”

NRC in 10 CFR 61.42 sets forth the following requirements:

 

47 This estimate is for the Federal Facility Waste Disposal Facility Canister Disposal Unit, where the vessel waste 

packages would be disposed of.
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“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.”

DOE’s dose limits for the hypothetical human intruder are more stringent than the dose limit 

used for NRC’s performance objective at 10 CFR 61.42. Typically, NRC applies a whole-body dose

equivalent limit of 500 mrem per year to assess compliance with the requirement at 10 CFR 51.42 

(NRC 2007), 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.48

Assessment of Area 5 Performance 

The report of the most recent (2011) 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 2012a) 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 of 3.4

millirem 1,000 years after facility closure, the end of the compliance period; and

 The home construction intruder scenario, with the predicted peak annual acute dose of 130

millirem 1,000 years after facility closure.

Chronic intruder scenarios are no longer reported in the Annual Summary Report for the Area 5 

Radioactive Waste Management Site because chronic 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.49 (NST 2012a)

Estimated Impact of Disposal of the Subject Vessels

Disposal of the concentrator feed makeup tank and the melter feed hold tank at the Area 5 

facility will have a negligible impact, if any, on facility performance with respect to inadvertent 

intruders, for the same reasons that disposal of these vessels would have a negligible impact, if 

any, on facility performance with respect to protecting members of the public. As discussed 

previously, this conclusion is based on the following factors:

 The vessel waste packages meet the waste acceptance criteria, with no radionuclide

concentrations exceeding the waste acceptance criteria action levels as shown in Table 5-1 

below. 

 Screening of the waste profile (CHBWV 2011a) by the Nevada National Security Site has 

shown that no special analysis is required for this waste stream.

 

48 Texas imposed on WCS a 25 mrem per year limit for intruder doses (WCS 2011). This matter and the comparability 

of DOE, NRC, and State of Texas regulatory provisions for imposing additional requirements on LLW disposal is 

discussed further in Appendix A.

49 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 (NST 2012a). 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.
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Assessment of WCS Federal Facility Waste Disposal Facility Performance

The updated WCS performance (WCS 2011) provides the following estimated doses to 

inadvertent intruders:

 A maximum acute dose of 1.4 millirem per year to the intruder driller, and

 A maximum chronic dose of 0.62 millirem per year to the intruder resident farmer. 

These estimated doses are well below the 25 millirem WCS annual limit50.

Consideration of the license limitations and the available performance assessment results for 

the WCS Federal Facility Waste Disposal Facility suggests that disposal of the concentrator feed 

makeup tank and melter feed hold tank waste packages at that facility would have negligible, if 

any, impact on dose to the inadvertent intruder for the following reasons:

 As shown in Table 5-2 below, the activity in each of the vessel waste packages – 96.5 

curies for the concentrator feed make up tank and 103 curies for the melter feed hold tank 

– would amount to only about 0.002 percent of the license limit for the Federal Facility 

Waste Disposal Facility.

 The 5.6 million curies in 26 million cubic feet of waste considered in the performance 

assessment would have an average activity density of 0.22 Ci per cubic foot, much higher 

than the activity density of the concentrator feed makeup tank waste package 

(approximately 0.03 Ci per cubic foot) and the melter feed hold tank waste package 

(approximately 0.04 Ci per cubic foot).

 5.2.4 Protection of Individuals During Operations

The DOE requirements in DOE Manual 435.1-1, Section I.E(13), for protection of individuals

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 [Order] 5400.5

[now DOE Order 458.1], Radiation Protection of the Public and the Environment.”

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 B. 

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 458.151 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 

 

50 These estimates are for the Federal Facility Waste Disposal Facility Canister Disposal Unit where the vessel waste 

packages would be disposed of. 

51 Section I, 1. E (13) of DOE Manual 435.1-1 cross-references 10 CFR Part 835 as well as prior DOE Order 5400.5. 

Prior DOE Order 5400.5 has been cancelled and replaced by DOE Order 458.1.
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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. 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. 

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 reasonably 

achievable) approach to occupational exposures.

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 preparing

the vessel waste packages 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 vessel waste packages 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 Part 835. These programs also comply with various DOE 

directives (including DOE Order 458.1, 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 (WVES 2010). The Nevada National Security Site radiological 

protection program and ALARA measures are described in the Nevada National Security Site

Radiological Control Manual (NST 2010) 

Gamma radiation levels one foot from the sides of the vessel waste packages range from 0.1 to 

10 millirem per hour for the concentrator feed makeup tank and from 0.1 to 3.5 millirem per hour 

for the melter feed hold tank (WVES 2011b). Workers involved with handling of the waste 

packages 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 5,000 mrem per year in 10 CFR 

835, Subpart C. The radiation doses to workers to be involved with preparation of the vessel waste 

packages 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 vessel waste packages

and the short duration of the work to prepare the waste packages for shipment. Furthermore, the 

work associated with preparing the waste packages 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 vessel waste packages for shipment are well below the applicable limit52. This work 

will be performed within a radiologically controlled area within the WVDP security fence. Past 

 

52 The applicable limit is 10 mrem per year for exposure to a member of the public from air emissions as specified in

the U.S. Environmental Protection Agency requirement in 40 CFR 61.92, with which DOE complies. The two vessels 

waste are essentially ready for shipment, except for loading them on the transport vehicles. 
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WVDP experience with similar waste management work indicates that potential doses to the public 

will be very low. In 2009, 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 0.0017 mrem (CHBWV 2011b). The airborne pathway is the 

only pathway of interest for potential exposure to a member of the public from onsite work to 

prepare the subject vessels for shipment. Such factors provide reasonable expectation that doses 

to the public from preparing the waste package for shipment will continue to be far below the 

applicable limit. 

Doses to workers at the Nevada National Security Site who would be involved with handling the 

vessel waste packages 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 vessel 

waste packages at the Nevada National Security Site are also expected to be very low. Operations 

to dispose of the vessel waste packages 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 vessel waste packages to the WCS facility for disposal, individuals 

would be protected during operations in a 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.53 

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 Facility Closure Plans (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 [now DOE Order 458.1].”

“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 

 

53 The updated WCS performance assessment (WCS 2011) provides an estimated average dose to workers during 

normal operations at the Federal Waste Facility of 95 mrem per year. 
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continue until the facility can be released pursuant to DOE Order 5400.5, Radiation Protection 

of the Public and the Environment [now DOE Order 458.1].”

As discussed in Appendix A, NRC requirements in 10 CFR 61.44 set 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.”

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 2012a). The 

plan will ensure that the applicable requirements of DOE Order 458.1 (which replaced prior DOE 

Order 5400.5 referenced in DOE Manual 435.1-1) 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 458.1 include the public dose limit of 100 mrem per year effective dose 

equivalent. The airborne emissions limit of 10 mrem per year effective dose equivalent (40 CFR 

61.92) also applies to emissions of radionuclides to the ambient air from DOE facilities.

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 Vessels Will Meet 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 vessel waste packages in compliance with the 

waste 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 vessels meet 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 2012) that are expressed 
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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 2012) 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 1.0E-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 2,000 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 2012)

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.54,55 

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 2012), 

(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 

 

54 Radionuclides 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 

2012). 

55 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 10 CFR 61.55, concerning classification of low-level 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 the U.S. Environmental 

Protection Agency’s regulations at 40 CFR Part 191 and the Waste Isolation Pilot Plant Land Withdrawal Act, 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 2012). Table E-1 provides 

radionuclide action levels for waste characterization and reporting purposes. 
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specified to these ends is verified by DOE in a comprehensive performance assessment program, 

as noted previously. 

5.3.2 The Vessel Waste Packages

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 vessel waste packages to the Nevada National Security Site waste 

acceptance criteria and determined that it will meet these criteria. For example, the melter feed 

hold tank contained an estimated 16.8 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 (CHBWV 2011a).

5.3.3 The Vessels Meet 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 2012). 

As described in Section 2, DOE has packaged each vessel in a custom-built, steel-shielded IP-2

container and the internal void spaces have been filled with low-density cellular concrete. The lowdensity cellular concrete is compatible with the waste acceptance criteria and will help reduce the 

possibility of disposal cell subsidence in the area of the buried waste containers by eliminating the 

container void spaces. 

As noted previously, a waste profile package for disposal of the vessel waste packages (CHBWV 

2011a) was submitted to the Nevada National Security Site and approved conditional upon the 

determination that the vessels are not HLW and can be managed as LLW (DOE 2011). Table 5-1

compares the radionuclide concentrations in the vessels with the radionuclide action levels for 

waste characterization and reporting provided in the Nevada National Security Site Waste 

Acceptance Criteria (DOE 2012). Radionuclides with concentrations exceeding one percent of the 

action level are highlighted in the table.56

 

56 As noted previously, the waste acceptance criteria document (DOE 2012) require 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 subject vessels were rigorously characterized by the WVDP.
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Table 5-1. Vessel Radionuclide Concentrations in Bq/m3(1)

Nuclide AL(2) CFMT(3) MFHT(3) Nuclide AL(2) CFMT(3) MFHT(3)

H-3 6.2E+11 NA NA U-232 4.3E+10 NR(4) NR(4)

C-14 5.4E+15 NA NR(4) U-233 8.2E+10 NR(4) NR(4)

K-40 9.4E+10 NA NR(4) U-234 1.3E+10 NR(4) NR(4)

Mn-54 NA NA NR(4) U-235 1.1E+10 NA NR(4)

Fe-55 NA NA NA U-236 2.8E+11 NA NR(4)

Ni-59 1.7E+14 NA NA U-238 3.5E+11 NR(4) NR(4)

Co-60 1.6E+12 NR(4) NR(4) Np-237 3.4E+10 NR(4) NR(4)

Ni-63 3.2E+14 NA NR(4) Pu-238 1.8E+12 6.32E+06 1.60E+07

Sr-90 4.3E+11 1.36E-09 9.32E+09 Pu-239 5.1E+11 1.70E+06 3.98E+06

Zr-95 NA NA NR(4) Pu-240 5.2E+11 1.30E+06 3.04E+06

Tc-99 3.2E+09 NR(4) NR(4) Pu-241 5.8E+12 NR(4) NR(4)

I-129 3.4E+09 NA NA Pu-242 3.7E+11 NA NA

Cs-137 2.5E+11 1.16E+11 1.69E+11 Am-241 1.7E+11 3.02E+07 7.56E+07

Pm-147 NA NA NA Am-243 5.8E+10 3.10E+05 6.86E+05

Eu-154 1.7E+12 NR(4) NR(4) Cm-242 NA NR(4) NR(4)

Th-228 4.3E+13 NA NR(4) Cm-243 8.3E+11 1.89E+05 4.96E+05

Th-229 2.8E+10 NA NA Cm-244 3.4E+12 NR(4) NR(4)

Th-230 6.0E+07 NA NR(4) Cm-245 4.6E+10 NA NA

Th-232 8.1E+09 NR(4) NR(4) Cm-246 9.2E+10 NA NA

LEGEND: AL = action level, CFMT = concentrator feed makeup tank, MFHT = melter feed hold tank, NA = not available, NR 

= not required to be reported because the radionuclide is <1% of the action level or <1% of the total activity 

concentration. 

NOTES: (1) To convert Bq/m3to pCi/L multiply by 0.027.

(2) From the Table E-1 of the Waste Acceptance Criteria (DOE 2012). 

(3) From the vessel waste profile technical basis document (CHBWV 2011a). These values are based on the waste 

volume; values based on the waste package volume are lower.

(4) The waste profile package (CHBWV 2011a) includes high and low activity range values for these radionuclides, 

which are not reportable. 

Table 5-1 shows that the estimated concentrations of all radionuclides fall below the action 

levels. 

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 2012) contains additional requirements related to waste disposal, 

including total waste volume limitations and total activity limitations for certain radionuclides. Table 
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5-2 shows representative requirements compared to the related parameters for the vessel waste 

package.

TABLE 5-2. Key WCS Federal Facility Waste Disposal Facility License Requirements 

Requirement (Section)(1) License 

Limit(1) CFMT MFHT CFMT % of Limit

MFHT % 

of Limit

Total waste volume, ft3(§7.B) 8,100,000 2,885 (2) 2,396(2) 0.036 0.030

Total activity, curies (§7.B) 5,500,000 96.5(2)103(2)0.002 0.002

Total C-14, curies (§5.D) 180 NA 0.0004

(2)

 NA 0.0002

Total Tc-99, curies (§5.D) 35 0.0042(2)0.0008(2)0.012 0.0023

Total I-129, curies (§5.D) 0.15 NA NA NA NA

LEGEND: NA = not available 

NOTES: (1) From the WCS license (TCEQ 2012) 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 characterization report (WMG 2011).

Table 5-2 shows that the volume of each vessel 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 packages are small fractions of the WCS limits. 

If DOE were to elect to dispose of the subject vessels at the WCS Federal Facility Waste 

Disposal Facility, DOE would confirm that the waste packages meet 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 Form57.

 

57 Because the WCS facility is licensed to accept Class C LLW, DOE would expect that the vessel waste packages 

would be approved for disposal. However, because this waste stream 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 vessel waste packages.
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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 concentrator feed 

makeup tank and the melter feed hold tank waste packages are 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 concentrator feed 

makeup tank and the melter feed hold tank waste packages are 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 concentrator feed makeup tank and the melter feed hold tank waste 

packages are in a solid physical form.

 The radioactivity in the concentrator feed makeup tank and the melter feed 

hold tank waste packages do not exceed Class C concentration limits.

 The concentrator feed makeup tank and the melter feed hold tank waste 

packages will be managed and disposed of at an offsite low-level radioactive 

waste disposal facility in accordance with applicable requirements for lowlevel 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 concentrator feed makeup tank and the melter feed hold tank

have been packaged in custom-built, steel-shielded IP-2 shipping containers and voids in the 

vessels and the space between the vessels and the inside of the containers have been filled with 

low-density cellular concrete. Hence, the vessel waste packages are in a solid physical form. (The 

vessels themselves are already in a solid physical form, as noted previously.)

Because the subject vessels contain 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 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 subject vessels before packaging was as described in 

Section 2.5.3. Table 6-1 shows the results of the waste classification calculations, which show that 

the vessel waste packages do not exceed Class C limits.

Table 6-1. Vessel Waste Classification Results With Respect to Class C limits(1)

Waste Package

Percent of Class C Limit

Table 1 Table 2

Concentrator Feed Makeup Tank 4.515 1.960

Melter Feed Hold Tank 6.190 1.631

NOTE: (1) From WMG 2011, as calculated by the RADMAN code.

The calculations were performed using the weight and size of the vessels themselves; neither 

the grout nor the shipping container was considered for conservatism (WMG 2011) even though 

the mass of the grout – which was necessary for stabilization purposes and to encapsulate surface 
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contamination – could have been considered in accordance with applicable concentration averaging 

guidance (NRC 1995). The uncertainties associated with the data used to estimate the residual 

radioactivity associated with the two vessels and to calculate the sums of fractions were relatively 

small, as discussed in Section 2, so it is clear that the vessel waste packages do not exceed Class C 

concentration limits. 

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 

vessel waste packages 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 (DOE 2012). This Waste Profile (CHBWV 2011a) has been approved by the 

Nevada National Security Site (DOE 2011) if a final decision is made to send the vessel waste 

packages to that facility for disposal.

As noted previously, DOE may elect to send the vessels to the commercially-operated WCS 

Federal Facility Waste Disposal Facility in Texas. As demonstrated earlier in this draft evaluation, 

vessel waste packages are in a solid physical form (with voids in the vessel and the space between 

the vessel 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.58

In addition, the State of Texas Class C concentration limits mirror the concentration limits in 10 CFR 

61.55; consequently, disposal of the vessel waste packages in the WCS Federal Facility 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 vessel waste packages at the WCS Federal

Facility Waste Disposal Facility would meet the third criterion of DOE Manual 435.1-1, Section 

II.B.2(a). 

 

58 Texas Administrative Code, Title 30, Part 1, §336.362, Appendix E.
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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 it 

available for public review and comment, including comment by the States of Nevada and Texas where 

the subject vessels might be disposed of as LLW. 

DOE will consider NRC comments as well as comments from the public, including the States of 

Nevada and Texas, before finalizing the evaluation and before making any final determination as to 

whether the concentrator feed makeup tank and the melter feed hold tank meet the criteria in DOE 

Manual 435.1-1 for waste incidental to reprocessing, and thus are not HLW and are to be managed and 

disposed of as LLW 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 concentrator feed makeup tank and the melter feed hold tank are 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. 

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 Part 61, Subpart C, Licensing Requirements for Land Disposal of Radioactive Waste, Performance 

Objectives. 

10 CFR 61.55, Waste Classification.

10 CFR Part 830, Subpart A, Quality Assurance Requirements.

10 CFR Part 835, Occupational Radiation Protection.

40 CFR Part 61, National Emission Standards for Hazardous Air Pollutants.

40 CFR Part 191, Environmental Radiation Protection Standards for Management and Disposal of Spent 

Nuclear Fuel, High-Level and Transuranic Radioactive Wastes.

47 FR 40705, Long-Term Management of Liquid High-Level Radioactive Wastes Stored at the Western 

New York Nuclear Service Center, West Valley; Record of Decision. U.S. Department of Energy, 

Washington, D.C., September 15, 1982.

65 FR 10061, 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 Test Site. U.S. Department of Energy, West Valley, New York, February 25, 2000.

70 FR 35073, 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 20582, Record of Decision: Final Environmental Impact Statement for Decommissioning and/or 

Long-Term 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

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DOE Order 435.1, Radioactive Waste Management, Change 1. U.S. Department of Energy, Washington, 

D.C., August 28, 2001.

DOE Order 458.1, Radiation Protection of the Public and the Environment, Change 2. U.S. Department of 

Energy, Washington, D.C., Change 2, June 6, 2011.

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 Manual 435.1-1, Radioactive Waste Management Manual, Change 1. U.S. Department of Energy, 

Washington, D.C., June 19, 2001.
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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. 

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?tac_view=4&ti=30&pt=1&ch=336)

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documentation). CH2M Hill - Babcock & Wilcox West Valley, LLC, West Valley, New York, August

2011. 

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Texas, September 3, 2009.

WCS 2011, Updated Performance Assessment for the Low-Level Waste Facility, Radioactive Material 

License No. R04100, CN600616890/RN101702439. Waste Control Specialists LLC, Rosebud, 

Texas, October 17, 2011. 

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, West Valley CFMT Complete Package Assembly, Drawing 4005-DW-002, Revision 0. WMG, 

Inc., Peekskill, New York, June 18, 2004. 

WMG 2004c, West Valley CMFHT Complete Package Assembly, Drawing 4005-DW-003, Revision 0. WMG, 

Inc., Peekskill, New York, June 18, 2004. 
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WMG 2004d, Certificate of Conformance, WNSCO Purchase Order 19-104320-C-LH, West Valley 

Demonstration Project CFMT IP2 Package, Serial Number.:WVDP-TC-471. WMG, Inc., Peekskill, 

New York, August 30, 2004.

WMG 2004e, Certificate of Conformance, WNSCO Purchase Order 19-104320-C-LH, West Valley 

Demonstration Project MFHT IP2 Package, Serial Number.:WVDP-TC-472. WMG, Inc., Peekskill, 

New York, August 30, 2004.

WMG 2011, West Valley CFMT and MFHT Characterization, Report 4005-CA-041, Revision 3. WMG, Inc., 

Peekskill, New York, August 18, 2011. 

WVES 2010, WVDP Radiological Controls Manual, WVDP-010, Revision 30. Lachapelle, E.B., West Valley 

Nuclear Services Company, West Valley, New York and URS Corporation, Buffalo, New York, April 

27, 2010.

WVES 2011a, Waste Incidental to Reprocessing Determination, WV-929, Revision 9. West Valley 

Environmental Services LLC, West Valley, New York, May 2011.

WVES 2011b, Radiation and Contamination Survey Report for Melter, CFMT, and MFHT, report number 

162342. West Valley Environmental Services LLC, West Valley, New York, June 16, 2011.

WVNSCO 2001, Waste Incidental to Reprocessing (WIR) Evaluation for Vitrification Facility Expended 

Materials, Revision 1. West Valley Nuclear Services Company, West Valley, New York, October 

25, 2001.

WVNSCO 2002a, HLW Processing Systems Flushing Operations Run Plan, WVNS-IRP-005. Revision 4. 

West Valley Nuclear Services Company, West Valley, New York, March 29, 2002. 

WVNSCO 2002b, CFMT Water Flush, Work Order VFS-63499, completed documentation. West Valley 

Nuclear Services Company, West Valley, New York, May 13, 2002.

WVNSCO 2002c, CFMT Water Flush Number 2, Work Order VFS-65670, completed documentation. West 

Valley Nuclear Services Company, West Valley, New York, June 27, 2002.

WVNSCO 2002d, MFHT Water Flush, Work Order VFS-65708, completed documentation. West Valley 

Nuclear Services Company, West Valley, New York, July 26, 2002.

WVNSCO 2002e, Letter from J. Paul of WVNSCO to A. Williams of DOE-West Valley on completion of 

Fiscal Year 2002 contract milestone on Melter Feed Hold Tank Flushing, WV 48, WD:2002:0400.

West Valley Nuclear Services Company, West Valley, New York, July 30, 2002. 

WVNSCO 2002f, Report on Deployment of Miscellaneous Tanks and Piping Cleaning Equipment and 

Methodology, Tank Focus Area Milestone B.1.1 for Technical Task Plan (TTP) Oh0-0-WT-22. West 

Valley Nuclear Services Company, West Valley, New York, August 30, 2002.

WVNSCO 2002g, MFHT Heel Removal, Work Order VFS-71556, completed documentation. West Valley 

Nuclear Services Company, West Valley, New York, September 10, 2002.

WVNSCO 2002h, Concentrator Feed Makeup Tank Level Data. West Valley Nuclear Services Company, 

West Valley, New York, 2002.

WVNSCO 2002i, Batch 76 Summary – Handwritten Notes, from WVDP Vitrification Feed Data Package, 

Batch 76, Part B, page 8. West Valley Nuclear Services Company, West Valley, New York, 2002.

WVNSCO 2002j, CMFT Sample Analytical Data from Vitrification Records, 1999-2002. West Valley Nuclear 

Services Company, West Valley, New York, 2002.
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WVNSCO 2003, Addition of Caustic to the CFMT, Work Order VFS-98458 WIP, completed documentation. 

West Valley Nuclear Services Company, West Valley, New York, December 18, 2003.

WVNSCO 2004a, Analytical and Process Chemistry report of analysis for evacuated canister powder-dust, 

Vitrification Analytical Tracking System sample ID 04-0073. West Valley Nuclear Services 

Company, West Valley, New York, January 23, 2004.

WVNSCO 2004b, Analytical and Process Chemistry report of analysis for evacuated canister powder-dust, 

Vitrification Analytical Tracking System sample ID 04-0074. West Valley Nuclear Services 

Company, West Valley, New York, January 23, 2004.

WVNSCO 2004c, Characterization Rad Survey of Melter, CFMT, MFHT, HEME, and Various Equipment, 

Radiation and Contamination Survey Report 123427. West Valley Nuclear Services Company, 

West Valley, New York, February 5, 2004. 

WVNSCO 2004d, Preparation of Vitrification Vessels for Removal, Work Order VFS-102005, completed 

documentation. West Valley Nuclear Services Company, West Valley, New York, December 27, 

2004. 
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APPENDIX A

Comparability of DOE, NRC and Texas Requirements for LLW Disposal

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).

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 performance 

objectives for the waste disposal facility. 

 The Department of Energy performance objectives are described in DOE Manual 

435.1-1, Radioactive Waste Management Manual.

 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.

 The Department of Energy, the Nuclear Regulatory Commission, and the State of 

Texas all have provisions for imposing additional requirements for low-level 

waste disposal and the State of Texas has imposed additional requirements for 

the WCS low-level waste disposal facility.
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2.0 Applicable Performance Objectives

DOE Manual 435.1-1, Radioactive Waste Management Manual, 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. 

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:
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“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:

“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 implicit 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/m2/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 
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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.”59 

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 Part 20, Standards for Protection Against Radiation. 

Because NRC has not revised 10 CFR 61.41 to reflect the more current methodology in 10 CFR Part

20, DOE’s requirements and those in 10 CFR Part 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 Part 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.”

 

59 WCS also uses a performance criterion for radon gas flux emanating from the disposal facility cover 

of 20 pCi/m2/s based on a provision of 40 CFR Part 192 of 20 pCi/m2/s, although 40 CFR Part 192

does not apply to LLW disposal facilities. This is not a State of Texas requirement, but it is the same 

as DOE’s criterion in DOE Manual 435.1-1, Section IV.P(1) except for DOE’s separate limit at the 

facility boundary. 
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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. However, Texas has imposed on WCS a limit of 25 mrem per year (0.25 mSv) for 

inadvertent intruders (WCS 2011). 

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 

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. 

The State of Texas regulations mirror the NRC regulations. However, as noted above, Texas 

has imposed an additional requirement for a lower limit of 25 mrem per year dose limit for 

inadvertent intruders. Therefore the State of Texas requirement is more limiting than the DOE and 

NRC requirements.

60 The comparability of DOE, NRC, and State of Texas provisions for imposing 

additional requirements is discussed in Section 8 below. 

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 [now DOE Order 458.1].”

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 

 

60 Note that Paragraph 4.b(4) of DOE Order 435.1, Radioactive Waste Management, requires DOE to comply with 

applicable Federal, State, and local laws and regulations. Therefore, if the vessel waste packages were to be 

transported to the WCS LLW facility for disposal, the facility would have to meet the 25 mrem per year dose limit for 

inadvertent intruders and the waste packages would have to meet any associated waste acceptance criteria. Because 

the activity density (curies per cubic foot) in the vessels waste packages is much less than the average activity 

density used in the updated WCS performance assessment as discussed in Section 5.2.3 of the body of this draft 

evaluation, it is likely that disposal of the vessel waste packages would have a negligible impact, if any, on 

compliance with the 25 mrem per year intruder dose limit. 
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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. DOE requirements for occupational radiological protection are addressed in 10 CFR Part

835, and similar requirements for radiological protection of the public and the environment are 

addressed in DOE Order 458.1. The NRC 10 CFR 61.43 requirement references 10 CFR Part 20, 

Standards for Protection Against Radiation, which contains similar radiological protection standards 

for workers and the public. 

Appendix B 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 B. 

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:

“Disposal Facility Closure Plans (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 [now DOE Order 458.1].”

“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 [now DOE Order 458.1].”

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.”
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7.3 State of Texas

The State of Texas regulations (Rule §336.727) mirror as the NRC regulations in 10 CFR 

61.43.

7.4 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 Comparability Regarding Provisions for Imposing Additional Requirements

8.1 DOE

Section 4.d of DOE Order 435.1, Radioactive Waste Management, states that:

“DOE, within its authority, may impose such requirements, in addition to those established in 

this Order, as it deems appropriate and necessary to protect the public, workers, and the 

environment, or to minimize threats to property.”

8.2 NRC

NRC provisions for imposing additional requirements on the license for a LLW disposal facility 

are contained in 10 CFR 61.24(h), which states:

“(h) The Commission may incorporate in any license at the time of issuance, or thereafter, by 

appropriate rule, regulation or order, additional requirements and conditions with respect to the 

licensee's receipt, possession, and disposal of source, special nuclear or byproduct material as 

it deems appropriate or necessary in order to:

(1) Promote the common defense and security;

(2) Protect health or to minimize danger to life or property;

(3) Require reports and the keeping of records, and to provide for inspections of activities 

under the license that may be necessary or appropriate to effectuate the purposes of 

the Act and regulations thereunder.”

8.3 State of Texas

The Texas provisions for imposing additional requirements on the license for a low-level waste 

disposal facility are contained in Rule §336.716(g), which states:

“(g) The commission may incorporate in any license at the time of issuance, or thereafter, by 

appropriate rule or order, additional requirements and conditions with respect to the licensee's 

receipt, possession, and disposal of waste as it deems appropriate or necessary in order to: 

 (1) protect the health and safety of the public and the environment; and 
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 (2) require reports and recordkeeping and to provide for inspections of activities under 

the license that may be necessary or appropriate to effectuate the purposes of the 

TRCA [Texas Radiation Control Act] and rules thereunder.”

8.4 Discussion

The DOE requirement is broader in scope than the NRC and State of Texas requirements 

because the DOE requirement applies to all aspects of radioactive waste management while the 

NRC and State of Texas requirements apply to licenses for LLW disposal facilities. Otherwise, the 

requirements are comparable. 

9.0 References

Code of Federal Regulations

10 CFR Part 20, Standards for Protection Against Radiation.

10 CFR Part 61, Subpart C, Licensing Requirements for Land Disposal of Radioactive Waste, Performance Objectives. 

10 CFR Part 835, Occupational Radiation Protection.

40 CFR Part 192, Health and Environmental Standards for Uranium and Thorium Mill Tailings.

DOE Orders, Policies, and Manuals

DOE Order 435.1, Radioactive Waste Management

DOE Order 458.1, Radiation Protection of the Public and the Environment, Change 2. U.S. 

Department of Energy, Washington, D.C., June 6, 2011.

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 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?tac_view=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, 5003, February 1, 2002. 
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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.

WCS 2011, Updated Performance Assessment for the Low-Level Waste Facility, Radioactive 

Material License No. R04100, CN600616890/RN101702439. Waste Control Specialists LLC, 

Rosebud, Texas, October 17, 2011. 

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 B

Comparability of DOE, NRC, and Texas 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 concentrator feed makeup tank and melter feed hold tank waste packages for 

shipment at the WVDP and handling of those waste packages when they are received at either the 

Nevada National Security Site or the WCS LLW disposal facility in Texas for disposal, assuming that 

the waste packages will be sent to one of those facilities. 

Section 5.2.4 of the body of this evaluation briefly addressed protection of individuals during 

these operations at the WVDP, the Nevada National Security Site, and the WCS LLW disposal 

facility. Appendix A also addressed this matter. 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 

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 Part 835,

Occupational Radiation Protection, and in DOE Orders. 

 The Nuclear Regulatory Commission radiation dose standards appear in 10 CFR 

Part 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 – mirror the Nuclear Regulatory Commission 

dose standards. 
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governed by §61.41 of this part. Every reasonable effort shall be made to maintain 

radiation exposures as low as is reasonably achievable.”

This requirement references 10 CFR Part 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 Part 835, Occupational Radiation 

Protection, and those for radiological protection of the public and the environment are provided in 

DOE Order 458.1, 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 Part 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).61 These NRC dose limits correspond to the DOE dose limits in 10 CFR Part 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 

subject vessels 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 Dose Standard Comparison

Table B-1 provides a crosswalk of the NRC, DOE, and State of Texas dose standards. This 

table shows that the dose standards applicable to DOE for individual workers and members of 

the public are comparable to those of NRC and Texas. The following information is provided to 

help explain information included in the table. 

2.1 Air Emissions Limit for Individual Member of the Public 

The DOE is subject to and complies with the U.S. Environmental Protection Agency requirement in 

40 CFR 61.92.

62 As can be seen in the table, the annual limit of 10 mrem (0.1 mSv) is the same for 

each agency.

 

61 The “standards for radiation protection” in 10 CFR Part 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 Part 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 subject vessels. 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.

62 40 CFR 61.92 provides as follows: “Emissions of radionuclides 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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Table B-1: Dose Standard Crosswalk(1)

Topic DOE NRC Texas

Annual air emission limit for 

individual member of the 

public

10 mrem (0.1 Sv)

40 CFR 61.42 

10 mrem (0.1 Sv)

10 CFR 20.1101(d)

10 mrem (0.1 Sv)

§336.304

Annual total effective dose 

equivalent for adult workers 

5 rem (0.05 Sv)

10 CFR 835.202(a)(1) 

5 rem (0.05 Sv)

10 CFR 20.1201(a)

5 rem (0.05 Sv)

§336.305

Any individual organ or tissue 

annual dose limit for adult 

workers

50 rem (0.5 Sv)

10 CFR 835.202(a)(2)

50 rem (0.5 Sv)

10 CFR 20.1201(a)

50 rem (0.5 Sv)

§336.305

Annual dose limit to the lens of 

the eye for adult workers

15 rem (0.15 Sv)

10 CFR 835.202(a)(3)

15 rem (0.15 Sv)

§20.1201(a)

15 rem (0.15 Sv)

§336.305

Annual dose limit to the skin of 

the whole body and to the skin 

of the extremities for adult 

workers

50 rem (0.5 Sv)

10 CFR 835.202(a)(4)

50 rem (0.5 Sv)

10 CFR 20.1201(a)

50 rem (0.5 Sv)

§336.305

Limit on soluble uranium 

intake

2.4 mg/week 

29 CFR 1910, Subpart Z

10 mg/week 10 CFR 

20.1201(e)

10 mg/week

§336.305

Dose equivalent to embryo/ 

fetus

0.5 rem (5 mSv)

10 CFR 835.206(a)

0.5 rem (5 mSv)

10 CFR 20.1208(a)

0.5 rem (5 mSv)

§336.312

Dose limit for individual 

members of the public (total 

annual dose)

100 mrem (1 mSv)

DOE Order 458.1

100 mrem (1 mSv)

10 CFR 20.1301(a)

100 mrem (1 mSv)

§336.313

Dose limit for individual 

members of the public (dose 

rates in unrestricted areas)

0.05 mrem/hr

(0.0005 mSv)

10 CFR 835.602

2 mrem/hr 

(0.02 mSv)

10 CFR 20.1301(a)

2 mrem/hr 

(0.02 mSv)

§336.313

Dose limits for members of the 

public with access to 

controlled areas

0.1 rem (0.001 Sv)

10 CFR 835.208

0.1 rem (0.001 Sv)

10 CFR 20.1301(b)

0.1 rem (0.001 Sv)

§336.313

As low as reasonably 

achievable requirements

10 CFR 835.2, 

10 CFR 835.101 

10 CFR 20.1003 §336.2

NOTES: (1) Requirements from DOE’s 10 CFR Part 835, NRC’s 10 CFR Part 20, and State of Texas Administrative Code, Title 30, 

Part 1, Chapter 336, Subchapter D, along with the cited DOE Order. Dose limit differences appear in boldface.

2.2 Total Effective Dose Equivalent Limit for Adult Workers

As can be seen in the table, the annual limit of 5 rem (0.05 Sv) is the same for each agency. 

2.3 Any Individual Organ or Tissue Dose Limit for Adult Workers

As can be seen in the table, the annual limit of 50 rem (0.5 Sv) is the same for each agency. 

2.4 Annual Dose Limit to the Lens of the Eye for Adult Workers

As can be seen in the table, the annual limit of 15 rem (0.15 mSv) is the same for each agency. 
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2.5 Annual Dose Limit to the Skin of the Whole Body and to the Skin of the

Extremities for Adult Workers 

As can be seen in the table, the annual limit of 50 rem (0.5 Sv) is the same for each agency. 

2.6 Limit on Soluble Uranium Intake 

DOE uses the Occupational Safety and Health Administration permissible exposure limit for 

soluble uranium at 29 CFR 1910, Subpart Z (0.05 mg/m3). This limit equates to a soluble uranium 

intake of 2.4 mg/week as shown in the table, which is lower than the limit specified by the other 

two agencies. 

2.7 Dose Equivalent to an Embryo/Fetus 

As can be seen in the table, the dose limit of 0.5 rem (5 mSv) is the same for each agency. 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 dosimetry63

is provided and used to track exposure carefully.

2.8 Dose Limits for Individual Members of the Public (Total Annual Dose)

As can be seen in the table, the annual limit of 100 mrem (1 mSv) is the same for each agency. 

2.9 Dose Limits for Individual Members of the Public (Dose Rate in Unrestricted 

Areas)

DOE’s regulation in 10 CFR 835.602 establishes the expectation that the total effective dose 

equivalent to individuals who enter controlled areas, without entering radiological areas or 

radioactive material control 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

(0.05 mSv 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.

2.10 Dose Limits for Individual Members of the Public With Access to Controlled Areas64

As can be seen in the table, the annual limit of 0.1 rem (0.001 Sv) is the same for each agency. 

DOE requires training for individual members of the public before entry into controlled areas. In 

addition, to ensure no member of the public exceeds 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

2.11 As Low As Reasonably Achievable Requirements

As can be seen in the table, each agency has ALARA requirements.

 

63 The term dosimetry or 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.

64 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.
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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.” The DOE regulation at 10 CFR 835.101 requires a 

documented radiation protection program approved by DOE, which shall include formal plans 

and measures for applying the ALARA process to occupational exposure. 

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.”

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 with 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.”

Conclusion

The DOE, NRC, and State of Texas definitions of ALARA are comparable.

3.0 References

Code of Federal Regulations 

10 CFR Part 20, Standards for Protection Against Radiation.

10 CFR Part 835, Occupational Radiation Protection.

29 CFR 1910, Subpart Z, Occupational Safety and Health Standards, Toxic and Hazardous 

Substances.

40 CFR Part 61, National Emission Standards for Hazardous Air Pollutants.

DOE Orders and Policies

DOE Order 458.1, Radiation Protection of the Public and the Environment, Change 2. U.S. 

Department of Energy, Washington, D.C., June 6, 2011.

State Regulations

Texas Administrative Code, Title 30, Part 1, Chapter 336, Radiation Substance Rules.
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APPENDIX C

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 concentrator feed makeup tank and melter feed hold tank. 

Key Points

 Section 3116 of the Ronald W. Reagan National Defense Authorization Act for 

Fiscal Year 2005 does not apply to the two vessels. 

 However, disposal of the vessel waste packages 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 evaluation demonstrate that the concentrator feed makeup tank and 

melter feed hold tank waste packages 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.65

 

65 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. 
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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 subject vessels,

66 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 vessel waste packages 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 Subject Vessels Require 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 Chapter II.B.(2)(a) that the waste “will be managed as low-level 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

 

66 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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objectives, 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.” 67

That is not the case here. As demonstrated in Section 4 of this draft evaluation, key radionuclides 

have been removed from the subject vessels to the maximum extent technically and economically 

practical. Moreover, the vessel waste packages will be in a solid physical form and will not exceed the 

concentration limits for Class C LLW in 10 CFR 61.55, as described in Section 6. As explained in Section 5, 

management and disposal of the subject vessels 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 vessel waste packages 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 subject 

vessels do 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 vessels 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)1, 

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.”68

Section 4, Table 4-3, of this draft evaluation identifies key radionuclides for the subject vessels. 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 vessels have been 

removed to the maximum extent technically and economically practical, thus satisfying the DOE criterion 

and evincing consistency with the second criterion of 3116(a).

 

67 Basis for Section 3116 Determination for the Idaho Nuclear Technology and Engineering Center Tank Farm Facility (DOE 

2006).

68 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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5.0 Consideration of Radionuclide Concentration Limits and Waste Disposal Performance 

Objectives

The third criterion in section 3116(a)(3) concerns whether the waste meets 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.69 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 vessel waste packages do 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 vessels have been packaged in shielded shipping containers

which have been filled with low-density cellular concrete, and thus are 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 packages will meet safety requirements comparable to NRC performance objectives in 10 CFR 

Part 61, Subpart C. Given these considerations, management and disposal of the two vessels as planned 

meets the above-referenced 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, 

before finalizing the evaluation and before making any final determination as to whether the concentrator 

feed makeup tank and melter feed hold tank are or are not HLW. 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 Classification.

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.

 

69 Although not germane here, section 3116(a)(3) also provides that the waste be disposed of “pursuant to a Stateapproved closure plan or State issued permit” 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?tac_view=4&ti=30&pt=1&ch=336)

Other References

NRC 2005, In the Matter of Louisiana Energy Services, L.P. (National Enrichment Services), 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.